WorldWideScience

Sample records for reaktoru typu vver

  1. Safety of VVER-440 reactors

    CERN Document Server

    Slugen, Vladimir

    2011-01-01

    Safety of VVER-440 Reactors endeavours to promote an increase in the safety of VVER-440 nuclear reactors via the improvement of fission products limitation systems and the implementation of special non-destructive spectroscopic methods for materials testing. All theoretical and experimental studies performed the by author over the last 25 years have been undertaken with the aim of improving VVER-440 defence in depth, which is one of the most important principle for ensuring safety in nuclear power plants. Safety of VVER-440 Reactors is focused on the barrier system through which the safety pri

  2. VVER-1000 backfitting programs

    International Nuclear Information System (INIS)

    Zabka, H.; Milhem, J.L.

    1998-01-01

    Russia, Ukraine, and Bulgaria have nineteen nuclear generating units of the VVER-1000/V-320 (1000 MWe PWR) type in operation. Most of these plants were built in the eighties. Their design is based on Soviet standards of the seventies. In the early eighties and, in particular, after the Chernobyl accident, new safety principles and supplementary specific standards were introduced. However, they were taken into account only to a limited extent in the design and construction of the VVER-1000/V-320 plants. A number of nuclear power plants, whose construction was stopped after the political changes in the countries of the former USSR, now are to be completed with the financial assistance of the Commission of the European Union and other Western organizations, respectively. This Western support is dependent on the condition that these plants attain a level of engineered safeguards comparable to that of PWR plants currently in operation in Western Europe. (orig.) [de

  3. Overview of VVER water chemistry

    International Nuclear Information System (INIS)

    Ganesh, S.; Selvaraj, S.; Balasubramanian, M.R.; Selvavinayagam, P.; Sundar, R.S.

    2007-01-01

    Kudankulam Nuclear Power project is having twin units of 1000MWe of VVER type. This paper highlights the different analytical techniques that are followed to maintain the system chemistry within the technical specifications. This paper also briefs the different chemicals that are added to the systems and how they are monitored. Basic differences with respect to chemistry between a PHWR and VVER are also highlighted in this paper. (author)

  4. Advanced designs of VVER reactor plant

    International Nuclear Information System (INIS)

    Mokhov, V.A.

    2010-01-01

    The history of VVER reactors, current challenges and approaches to the challenges are highlighted. The VVER-1200 reactor of 3+ generation for AES-2006 units are under construction at the Leningrad 2 nuclear power plant (LNPP-2). The main parameters are listed and details are presented of the vessel, steam generator, and improved fuel. The issue of the NPP safety is discussed. Additional topics include the MIR-1200 reactor unit, VVER-600, and VVER-SCP (Generation 4). (P.A.)

  5. Fuel designs for VVER reactors

    International Nuclear Information System (INIS)

    Simonov, K.V.; Carbon, P.; Silberstein, A.

    1995-01-01

    That progresses in efficiency and safety through progresses in technology and better prediction with fully benchmarked upgraded computer codes is a common goal for on the one hand the original designer of the VVER reactors and their respective fuels and on the other hand for EVF a western company resulting from a combined force with highly diversified and complementary talents in reactor and fuel design and manufacturing. It can be expected that this new challenge and dialogue between the two Russian and European industrial ventures will be mutually beneficial and yield innovative and high quality products and as a consequence strong return will be produced for the best interest of utilities operating VVER reactors. (orig./HP)

  6. SCORPIO - VVER core surveillance system

    International Nuclear Information System (INIS)

    Hornaes, A.; Bodal, T.; Sunde, S.

    1998-01-01

    The Institutt 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 VVERs. The main differences between VVERs and typical western PWRs with respect to core surveillance requirements are outlined. The development of a VVER 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.(author)

  7. The FARC fuel archive of VVER

    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 VVER reactors is described. The objective of the archive is accumulation of fuel data, data storage and obtaining the fuel using characteristics. The working version of fuel archive on 01.07.98 is realised, in which the data tables for fuel assemblies for 169 VVER-440 cycles and 35 VVER-1000 cycles are stored. There are two different versions of fuel archive - for VVER-440 (FARC) and for VVER-1000 (FARC1000). A structure of some tables and the texts of programs for them differ. The algorithms and codes for checking integrity, reasonableness and reliability of fuel archive data are developed. (author)

  8. VVER fuel cycle development at Slovakia

    International Nuclear Information System (INIS)

    Darilek, P.; Chrapiak, V.; Majerik, J.

    1995-01-01

    Four VVER-440 units are now under exploitation at Bohunice-site in Slovakia. Fuel cycle development of Unit No.3 and No.4 (type 213) is discussed and compared with equilibrium cycles in this paper. (author)

  9. APROS multifunctional simulator applications for VVER-440

    International Nuclear Information System (INIS)

    Porkholm, K.; Kantee, H.; Tiihonen, O.

    2000-01-01

    Fortum Engineering Ltd and the Technical Research Centre of Finland have developed APROS simulation software since 1986. APROS is a multifunctional simulator, which is used for process and automation design, safety analysis and training simulator applications. APROS has unique features and models developed especially for VVER-440 reactors. At first the paper gives a short overview of APROS multifunctional simulator. The rest of the paper deals with different kind of applications of APROS in VVER-440 reactors' improvement and operation development. (author)

  10. Vliv typu vazby historického zdiva na jeho teplotní namáhání

    Czech Academy of Sciences Publication Activity Database

    Beran, Pavel

    2013-01-01

    Roč. 22, č. 10 (2013), s. 248-253 ISSN 1805-2576 R&D Projects: GA ČR(CZ) GBP105/12/G059 Institutional support: RVO:68378297 Keywords : thermal expansion * masonry * thermal stress * compatibility Subject RIV: JN - Civil Engineering http://www.stavebniobzor.cz/clanek/vliv-typu-vazby-historickeho-zdiva-na-jeho-teplotni-namahani/

  11. Nuclear fuel for VVER reactors. Actual state and trends

    International Nuclear Information System (INIS)

    Molchanov, V.

    2011-01-01

    The main tasks concerning development of FA design, development and modernization of structural materials, improvement of technology of structural materials manufacturing and FA fabrication and development of methods and codes are discussed in this paper. The main features and expected benefit of implementation of second generation and third generation fuel assembly for VVER-440 Nuclear Fuel are given. A brief review of VVER-440 and VVER-1000 Nuclear Fuel development before 1997 since 2010 is shown. A summary of VVER-440 and VVER-1000 Nuclear Fuel Today, including details about TVSA-PLUS, TVSA-ALFA, TVSA-12 and NPP-2006 Phase 2 tasks (2010-2012) is presented. In conclusion, as a result of large scope of R and D performed by leading enterprises of nuclear industry modern nuclear fuel for VVER reactors is developed, implemented and successfully operated. Fuel performance (burnup, lifetime, fuel cycles, operating reliability, etc.) meets the level of world's producers of nuclear fuel for commercial reactors

  12. CORONA project -contribution to VVER nuclear education and training

    International Nuclear Information System (INIS)

    Ilieva, M.; Miteva, R.; Takov, T.

    2016-01-01

    CORONA Project is established to stimulate the transnational mobility and lifelong learning amongst VVER end users. The project aims to provide a special purpose structure for training of specialists and to maintain the nuclear expertise by gathering the existing and generating new knowledge in the VVER area. CORONA Project consists of two parts: CORONA I (2011-2014) ''Establishment of a regional center of competence for VVER technology and Nuclear Applications'', co-financed by the Framework Program 7 of the European Union (EU) and CORONA II (2015-2018) ''Enhancement of training capabilities in VVER technology through establishment of VVER training academy'', co-financed by HORIZON 2020, EURATOM 2014-2015. The selected form of the CORONA Academy, together with the online availability of the training opportunities will allow trainees from different locations to access the needed knowledge on demand. The project will target also new-comers in VVER community like Vietnam, Turkey, Belarus, etc. (authors)

  13. Response of Soviet VVER-440 accident localization systems to overpressurization

    International Nuclear Information System (INIS)

    Kulak, R.F.; Fiala, C.; Sienicki, J.J.

    1989-01-01

    The Soviet designed VVER-440 model V230 and VVER-440 model V213 reactors do not use full containments to mitigate the effects of accidents. Instead, these VVER-440 units employ a sealed set of interconnected compartments, collectively called the accident localization system (ALS), to reduce the release of radionuclides to the atmosphere during accidents. Descriptions of the VVER accident localization structures may be found in the report DOE NE-0084. The objective of this paper is to evaluate the structural integrity of the VVER-440 ALS at the Soviet design pressure, and to determine their response to pressure loadings beyond the design value. Complex, three-dimensional, nonlinear, finite element models were developed to represent the major structural components of the localization systems of the VVER-440 models V230 and V213. The interior boundary of the localization system was incrementally pressurized in the calculations until the prediction of gross failure. 6 refs., 9 figs

  14. FP 6 EU - COVERS. Coordination action - VVER safety research

    International Nuclear Information System (INIS)

    Vasa, I.

    2008-01-01

    In this work research program of the European Union FP 6 - COVERS coordinated by the NRI Rez is presented. COVERS is designed to improve professional and communication environment in the specific area covering all aspects of safe and reliable operation of nuclear power plants with VVER-440 and VVER-1000 reactors. Project Consortium is composed of 26 research and development, engineering and technical support organisations of European VVER-operating and other EU and non-EU countries.

  15. Upper plenum break LOCA investigation in the ISB-VVER and PSB-VVER facilities

    Energy Technology Data Exchange (ETDEWEB)

    Blinkov, V. N.; Melikhov, O. I.; Lipatov, I. A.; Nikonov, S. M.; Dremin, G. I.; Galchanskaya, S. A.; Gashenko, M. P.; Rovnov, A. A.; Kapustin, A. V.; Elkin, I. V. [EREC, Moscow (Russian Federation)

    2003-07-01

    The capability to define the actual NPP transient/accident scenario depends to a great extent on facilities' scaling and reliability of the system thermalhydraulic codes which, in turn, are assessed against the experimental data taken in the same facilities. At the present time, it is received fact that the rigorous modeling of the cumulative set of all thermalhydraulic processes in the plant primary and secondary sides during accident is unfeasible. Therefore, the extrapolation of the facilities loops behavior to the actual systems constitutes a fundamental problem in this area. In the paper, some aspects for the problem have been discussed in the course of comparative analysis of the data derived from the 11 % upper plenum break LOCA tests performed in the PSB-VVER and ISB-VVER integral test facilities under the close scenarios. Both facilities, PSB-VVER and ISB-VVER, are modeled the same VVER-1000 reactor in different scales. The thermalhydraulic behavior of the primary systems in both facilities has been discussed in the paper, and shown to be similar. Also, the attention has been focused upon the discrepancies in the significant variables trends. The discrepancies are shown to be caused by influence of peculiarities of the facilities hardware and due to the scale factor. The scaling study is an important aspect of the thermalhydraulic codes verification procedure. Being qualified against the experimentally simulated accident sequence in two test facilities of different scales, the thermalhydraulic codes will be capable of evaluation of the prototype behavior to greater accuracy.

  16. Zaostrzenie przewlekłej choroby nerek w przebiegu pierwotnej hiperoksalurii typu 1 - prezentacja przypadku

    Directory of Open Access Journals (Sweden)

    Violetta Bochniewska

    2010-06-01

    Full Text Available Przewlekła choroba nerek (PChN zaliczana jest do schorzeń cywilizacyjnych XXI wieku. Jej częstość w ciągu ostatnich lat systematycznie wzrasta. Według danych epidemiologicznych na PChN choruje około 600 min ludzi na świecie i ponad 4 min osób w Polsce. Najczęstsze przyczyny przewlekłej choroby nerek to: cukrzyca, nadciśnienie tętnicze, przewlekłe kłębkowe zapalenie nerek przewlekłe odmiedniczkowe zapalenie nerek, kamica dróg moczowych, zwyrodnienie torbielowate nerek nowotwory układu moczowego, szpiczak mnogi, skrobiawica wtórna. Do rzadkich przyczyn PChN możemy zaliczyć choroby uwarunkowane genetycznie, wśród nich pierwotną hiperoksalurię typu 1. W pracy autorzy przedstawiają historię rozpoznania pierwotnej hiperoksalurii typu 1 u 4-letniej dziewczynki oraz wpływ tej choroby na rozwój przewlekłej choroby nerek. Podczas leczenia ostrego odmiedniczkowego zapalenia nerek u pacjentki rozpoznano niewydolność nerek, nefrokalcynozę oraz kamicę układu moczowego. Wdrożono leczenie urologiczne, które powtarzano przez 3 miesiące (2 x operacje, 2 x ESWL. Następnie wykonano diagnostykę metaboliczną kamicy i nefrokalcynozy, która ujawniła znacznie zwiększone wydalanie kwasu szczawiowego. Z tego powodu wysunięto podejrzenie pierwotnej hiperoksalurii, które następnie potwierdzono badaniem genetycznym (typowe mutacje genu AGXT. Wdrożenie leczenia przyczynowego pirydoksyną pozwoliło na zmniejszenie wydalania kwasu szczawiowego w moczu i poprawę funkcji nerek. Wnioski: Pierwotna hiperoksaluiia typu 1 może prowadzić do niewydolności nerek w okresie wczesnego dzieciństwa. Wczesna diagnostyka i rozpoznanie tej choroby pozwalają na wdrożenie odpowiedniego leczenia, co zwiększa szansę na opóźnienie jej postępu.

  17. VVER fuel. Results of post irradiation examination

    International Nuclear Information System (INIS)

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

    2005-01-01

    The present paper presents the main results of post-irradiation examination of more than 40 different fuel assemblies (FA) operated in the cores of VVER-1000 and VVER-440-type power reactors in a wide range of fuel burnup. The condition of fuel assembly components from the viewpoint of deformation, corrosion resistance and mechanical properties is described here. A serviceability of the FA design as a whole and interaction between individual FA components under vibration condition and mechanical load received primary emphasis. The reasons of FA damage fuel element failure in a wide range of fuel burnup are also analyzed. A possibility and ways of fuel burnup increase have been proved experimentally for the case of high-level serviceability maintenance of fuel elements to provide for advanced fuel cycles. (author)

  18. Core designs of modern VVER projects

    International Nuclear Information System (INIS)

    Vasilchenko, I.; Kushmanov, S.; Vjalitsyn, V.; Vasilchenko, R.

    2015-01-01

    The presented operational experience of TVS - 2M (pilot-commercial operation started in 2006 at Balakovo NPP -1) enables to use it as reference for new projects because of similarity in designs and operational conditions. In the paper main parameters of fuel cycles, stability to impact of damaging factors, pilot operation of MG, new alloys, ADF and NTMC, upgrade of FA - 2M for the further power uprating, profiling of Gd-fuel rods for 18-month Fuel Cycle (FC) and perfection of absorber element design are the discussed issues. At the end author concluded that: 1) Core designs of new projects AES-2006 and VVER-TOI are based on extensive successful operational experience of the close prototype of TVS - 2M. 2) All improvements both of technical and economic parameters of fuel are subjected to representative examination by pilot operation at the power units with VVER-1000 being close prototypes of new designs

  19. VVER-1000 RPV Head Examination Control System

    International Nuclear Information System (INIS)

    Erak, Z.; Gortan, K.

    2006-01-01

    This article presents the electronic system used for automated NDT examination of VVER-1000 Reactor Pressure Vessel Head (RPVH). The control system drives the inspection tool with end-effectors to needed position. When the final position is reached, the eddy current and ultra sound acquisition system performs the data acquisition. The system is composed of 3 layers. The first layer is the hardware layer consisting of motors driving the tool and end-effectors along with sensors needed to obtain the positioning data. The second layer is the MAC-8 control system performing basic monitoring and control routines as an interconnection between first and third layer. The third layer is the control software, running on PC, which is used as a human-machine-interface. Presentation contains details of examination techniques with focus on eddy current examination as well as details on manipulator and end effectors developed by Inetec for VVER-1000 RPVH examination.(author)

  20. VVER-1000: considering its strengths and weaknesses

    International Nuclear Information System (INIS)

    Laaksonen, J.

    1994-01-01

    The safety of currently operating VVER-1000 reactors is examined. The factors considered are deviations in operation, inherent safety, safety system design, protection against internal and external hazards, equipment quality, the approach to plant operations and the safety culture. On the basis of this evaluation it is concluded that the overall safety of a VVER-1000 cannot be at the level of a modern Western PWR though there is no sound basis to make a quantitative comparison. Many of the concerns raised are being adequately addressed in the Czech Temelin which is currently under construction and in new designs which are still at the drawing board stage. Extensive back fitting programmes are planned or underway in operating plants. The creation of independent responsible operating organizations, powerful regulation and an improved economic situation are advanced as necessary criteria for real improvements in safety. (UK)

  1. Vver-1000 Mox core computational benchmark

    International Nuclear Information System (INIS)

    2006-01-01

    The NEA Nuclear Science Committee has established an Expert Group that deals with the status and trends of reactor physics, fuel performance and fuel cycle issues related to disposing of weapons-grade plutonium in mixed-oxide fuel. The objectives of the group are to provide NEA member countries with up-to-date information on, and to develop consensus regarding, core and fuel cycle issues associated with burning weapons-grade plutonium in thermal water reactors (PWR, BWR, VVER-1000, CANDU) and fast reactors (BN-600). These issues concern core physics, fuel performance and reliability, and the capability and flexibility of thermal water reactors and fast reactors to dispose of weapons-grade plutonium in standard fuel cycles. The activities of the NEA Expert Group on Reactor-based Plutonium Disposition are carried out in close co-operation (jointly, in most cases) with the NEA Working Party on Scientific Issues in Reactor Systems (WPRS). A prominent part of these activities include benchmark studies. At the time of preparation of this report, the following benchmarks were completed or in progress: VENUS-2 MOX Core Benchmarks: carried out jointly with the WPRS (formerly the WPPR) (completed); VVER-1000 LEU and MOX Benchmark (completed); KRITZ-2 Benchmarks: carried out jointly with the WPRS (formerly the WPPR) (completed); Hollow and Solid MOX Fuel Behaviour Benchmark (completed); PRIMO MOX Fuel Performance Benchmark (ongoing); VENUS-2 MOX-fuelled Reactor Dosimetry Calculation (ongoing); VVER-1000 In-core Self-powered Neutron Detector Calculational Benchmark (started); MOX Fuel Rod Behaviour in Fast Power Pulse Conditions (started); Benchmark on the VENUS Plutonium Recycling Experiments Configuration 7 (started). This report describes the detailed results of the benchmark investigating the physics of a whole VVER-1000 reactor core using two-thirds low-enriched uranium (LEU) and one-third MOX fuel. It contributes to the computer code certification process and to the

  2. Modernizing the VVER-440/230

    International Nuclear Information System (INIS)

    Mink, F.J.

    1991-01-01

    The modernization of the VVER-440/230s is not fundamentally different from backfit projects on older pressurized water reactors which Westinghouse has completed elsewhere. However, carrying out such programmes only makes sense if the plants are expected to continue operation for their projected life or beyond. This clearly requires some licensing and political stability; both are essential if investors in the upgrading project are to be found. (author)

  3. In core-fuel management in approach to equilibrium of WWER-440 reactor; Prelazni rezim iskoriscenja goriva u nuklearnom reaktoru tipa VVER-440

    Energy Technology Data Exchange (ETDEWEB)

    Marinkovic, N [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

    1978-07-01

    For the need of in core fuel management and prediction of fuel cycle costs as well as operating of a nuclear power plant behaviour of main physical parameters and refueling scheme during approach to equilibrium operation are indispensable. An estimation of a refueling scheme during forst six years of exploitation for a commercially proven PWR reactor of WWER-440 type is shown in this paper. (author)

  4. Enhancement of Training Capabilities in VVER Technology Through Establishment of VVER Training Academy

    International Nuclear Information System (INIS)

    Ilieva, M.; Miteva, R.

    2015-01-01

    Education and training (E&T) have always been key factor to the sustainability of the nuclear industry. With regard to E&T it is still the challenge to raise the interest of qualified young people of studies and professions related to nuclear technologies. CORONA Project is established to provide a special purpose structure for training and for gathering the existing and generating new knowledge in the VVER area as well as to contribute to transnational mobility and lifelong learning amongst VVER operating countries. CORONA Project consists of two parts: CORONA I (2011–2014) “Establishment of a regional centre of competence for VVER technology and Nuclear Applications”, co-financed by the EC Framework Programme 7 and CORONA II “Enhancement of training capabilities in VVER technology through establishment of VVER training academy”, co-financed by the EURATOM 2014-2015 Working programme of HORIZON 2020. The project is focused on development of training schemes for VVER nuclear professionals, subcontractors, students and for non-nuclear specialists working in support of nuclear applications as civil engineers, physical protection employees, government employees, secondary school teachers, journalists. Safety culture and soft skills training are incorporated as an integral part of all training schemes because they require continuous consideration. It is vital for the acceptance of nuclear energy by the public and for the safe performance of the nuclear installations. CORONA II project is to proceed with the development of state-of-the-art virtual training centre — CORONA Academy. This objective will be realised through networking between universities, research organizations, regulatory bodies, industry and any other organizations involved in the application of nuclear science, ionising radiation and nuclear safety. It will bring together the most experienced trainers and will allow trainees from different locations to access the needed knowledge on demand

  5. Feasibility of VVER-440 type SFAT

    International Nuclear Information System (INIS)

    Kaartinen, J.; Tarvainen, M.

    1995-05-01

    Spent fuel attribute tester, SFAT, has been constructed and tested for gross defect verification of VVER-440 type spent fuel assemblies. Based on earlier optimisation studies, the VVER-440 SFAT is kept hanging from the mast of the fuel handling machine moved by the operator. The device tested includes a standard 2' x 2' NaI(T1) detector connected to a commercial MCA. The results achieved with normal VVER-440 spent fuel assemblies at the Loviisa npp in Finland in November 1994 show that the method is feasible. The design of the so-called fuel follower assemblies, however, prevents SFAT verification, at least with moderate measurement times. Verification of the presence of the assemblies based on the detection of the fission product 137 Cs (662 keV) is possible even in 10-30 seconds. Measurement times of the order of 1-2 minutes make it possible to draw also semi-quantitative conclusions of the burnup and cooling time of the operator declared data (consistency check). (orig.) (7 refs., 11 figs., 3 tabs.)

  6. Water chemistry experiences with VVERs at Kudankulam

    International Nuclear Information System (INIS)

    Rout, D.; Upadhyaya, T.C.; Ravindranath; Selvinayagam, P.; Sundar, R.S.

    2015-01-01

    Kudankulam Nuclear Power Project - 1 and 2 (Kudankulam NPP - 1 and 2) are pressurised water cooled VVERs of 1000 MWe each. Kudankulam NPP Unit - 1 is presently on its first cycle of operation and Kudankulam NPP Unit - 2 is on the advanced stage of commissioning with the successful completion of hot run related Functional tests. Water Chemistry aspects during various phases of commissioning of Kudankulam NPP Unit - 1 such as Hot Run, Boric acid flushing, initial fuel Loading (IFL), First approach to Criticality (FAC) are discussed. The main objectives of the use of controlled primary water chemistry programme during the hot functional tests are reviewed. The importance of the relevant water chemistry parameters were ensured to have the quality of the passive layer formed on the primary coolant system surfaces. The operational experiences during the 1 st cycle of operation of primary water chemistry, radioactivity transport and build-up are presented. The operational experience of some VVER units in the field of the primary water chemistry, radioactivity transport and build-up are presented as a comparison to VVER at Kudankulam NPP. The effects of the initial passivated layer formed on metal surfaces during hot run, activated corrosion products levels in the primary coolant under controlled water chemistry regime and the contamination/radiation situation are discussed. This report also includes the water chemistry related issues of secondary water systems. (author)

  7. Virtuální prohlídka staveniště termojaderného reaktoru aneb Tour de ITER

    Czech Academy of Sciences Publication Activity Database

    Řípa, Milan

    Únor (2017) ISSN 2464-7888 Institutional support: RVO:61389021 Keywords : fusion * ITER * tokamak * virtual * 2D * 3D * dron Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) http://www.3pol.cz/cz/rubriky/jaderna-fyzika-a-energetika/1977-virtualni-prohlidka-staveniste-termojaderneho-reaktoru-aneb-tour-de-iter

  8. Development of TVSA VVER-1000 fuel

    International Nuclear Information System (INIS)

    Samoilov, O.; Kaydalov, V.; Romanov, A.; Falkov, A.; Morozkin, O.; Sholin, E.

    2013-01-01

    The TVSA fuel assemblies with a rigid angle-piece skeleton operate at 21 VVER-1000 units of Kalinin NPP, and Ukrainian, and Czech and Bulgarian NPPs. The total of more than 6,000 TVSA fuel assemblies have been fabricated. High lifetime performance has been achieved, namely, the maximum FA burnup is 65 MW∙day/kgU; maximum fuel rod burnup is 72 MW∙day/kgU; the lifetime is 50,000 EFPH. The TVSA fuel assembly is being improved to enhance its technical and economic performance and competitiveness of the Russian fuel for the VVER-1000 reactor: 1) Reliability and safety are being enhanced; repairability is being ensured. 2) High burnup levels in fuel are being ensured. 3) The uranium content in FAs is being increased. 4) The operational life is being extended. 5) Thermal-technical characteristics of FAs are being improved. The basic TVSA fuel assembly design evolved into the TVSA-PLUS with the fuel column elongated by 150 mm. The TVSA-PLUS fuel assembly has been in operation since 2010 at Kalinin NPP power units; an eighteen-month cycle is implemented at the uprated power of 104%. The TVSA-12PLUS fuel assembly has been developed with an elongated fuel column, optimized spacer grid positions (the spacer grid pitch is 340 mm) and with ensuring higher rigidity for the skeleton. It is provided for that fuel rods with the elevated uranium content and mixing intensifier grids will be used. The TVSA-T is developed for VVER-1000 reactor cores at the Temelin NPP. The TVSA-T is characterized by a load-carrying skeleton formed with angle-pieces and combined spacer grids that incorporate mixer grids. The TVSA-T design won the international tender to supply fuel to the Temelin NPP in the Czech Republic, and currently Temelin NPP Unit 1 and 2 are operating with the cores fully loaded with TVSA-Ts

  9. Hydrogen mixing analyses for a VVER containment.

    Energy Technology Data Exchange (ETDEWEB)

    Sienicki, J.J.; Kostka, P.; Techy, Z.

    2002-02-25

    Hydrogen combustion may represent a threat to containment integrity in a VVER-440/213 plant owing to the combination of high pressure and high temperature. A study has been carried out using the GASFLOW 2.1 three-dimensional CFD code to evaluate the hydrogen distribution in the containment during a beyond design basis accident. The VVER-440/213 containment input model consists of two 3D blocks connected via one-dimensional (1D) ducts. One 3D block contains the reactor building and the accident localization tower with the suppression pools. Another 3D block models the air traps. 1D ducts represent the check valves connecting the accident localization tower with the air traps. The VVER pressure suppression system, called ''bubbler condenser,'' was modeled as a distributed heat sink with water thermodynamic properties. This model accounts for the energy balance. However, it is not currently possible to model dynamic phenomena associated with the water pools (e.g., vent clearing, level change). The GASFLOW 2.1 calculation gave detailed results for the spatial distribution of thermal-hydraulic parameters and gas concentrations. The range and trend of the parameters are reasonable and valuable. There are particularly interesting circulation patterns around the steam generators, in the bubbler tower and other primary system compartments. In case of the bubbler tower, concentration and temperature contour plots show an inhomogeneous distribution along the height and width, changing during the accident. Hydrogen concentrations also vary within primary system compartments displaying lower as well as higher (up to 13-20% and higher) values in some nodes. Prediction of such concentration distributions was not previously possible with lumped parameter codes. GASFLOW 2.1 calculations were compared with CONTAIN 1.2 (lumped parameter code) results. Apart from the qualitatively similar trends, there are, for the time being, quantitative differences between the

  10. VVER-specific features regarding core degradation - Status Report

    International Nuclear Information System (INIS)

    Hozer, Z.; Trambauer, K.; Duspiva, J.

    1999-01-01

    The objective of this report is to compare VVER reactors to pressurised water reactors (PWRs) of Western design from the point of view of core degradation phenomena using the terminology which was applied to the systematisation of severe accident phenomena in earlier CSNI reports. In the following the acronym 'PWR' is used for a PWR of Western design. The basic design features are described and the most important parameters are summarised in order to identify the differences between the two reactor types. In some specific cases the comparison shows more similarities with boiling water reactors (BWRs) than with PWRs. The known VVER experimental support is also summarised. RBMKs are not included in this report, as this reactor type is not operated in OECD countries, furthermore its design is completely different from those of VVERs and PWRs. The scope of this report is limited to in-vessel severe fuel damage phenomena. Neither thermal hydraulic processes involving no core degradation, nor containment phenomena, are discussed in detail. The VVER (water-cooled water-moderated power reactor) is a pressurised light water reactor of Soviet design. It operates on the same principles as a Western PWR reactor and uses similar technological systems. The primary coolant is pressurised water, which heats up in the reactor core and steam is produced on the secondary side of steam generators. The comparison of basic geometrical and technological parameters pointed out some differences between a PWR and a VVER, but it should be noted that differences exist even between two Western PWRs of different design. The VVER reactors are special types of PWRs, the most important design features of which are the horizontal steam generators and the hexagonal core structure. Similarity between PWR and VVER reactors was found in the comparison of dominant accidents sequences leading to core melt. The accident progression sequence consists of the same steps for VVERs and PWRs. The larger water

  11. Containment leak-tightness enhancement at VVER 440 NPPs

    International Nuclear Information System (INIS)

    Prandorfy, M.

    2001-01-01

    The hermetic compartments of VVER 440 NPPs fulfil the function of the containment used at NPPs all over the word. The purpose of the containment is to protect the NPP personal against radioactive impact as well as to prevent radioactive leakage to the environment during a lost of coolant accident. Leak-tightness enhancement in NPPs with VVER 440/213 and VVER 440/230 reactors is an important safety issue. New procedures, measures and methods were adopted at NPPs in Mochovce, J. Bohunice, Dukovany and Paks for leak identification and sealing works performed by VUEZ Levice. (authors)

  12. PSB-VVER experimental and analytical investigation of station blackout accident in VVER-1000

    Energy Technology Data Exchange (ETDEWEB)

    Lipatov, I.A.; Kapustin, A.V.; Nikonov, S.M.; Rovnov, A.A.; Basov, A.V. [Electrogorsk Research and Engineering Centre (EREC), Moscow Region (Russian Federation); Elkin, I.V. [NSI RRC, Kurchatov Institute, Moscow (Russian Federation)

    2007-07-01

    In November 2003, an experiment simulating station blackout accident was carried out in the PSB-VVER integral test facility at the Electrogorsk Research and Engineering Centre (Russia). The purpose of the experiment was to provide missing data for code validation as well as to investigate the VVER thermohydraulics in the blackout conditions. The experiment covers a wide range of phenomena relating not only to transients but also to small break loss-of-coolant accidents. The data gained in the test has been used to assess the RELAP5/MOD3.3 code. In this paper, a special attention has been paid to the code assessment regarding the mixture level and entrainment in steam generator secondary side. The analysis of the recorded transient has shown that the calculation of the heat transfer on the secondary side of steam generators is very sensitive to the steam generator nodalization. (authors)

  13. Innovative instrumentation for VVERs based in non-invasive techniques

    International Nuclear Information System (INIS)

    Jeanneau, H.; Favennec, J.M.; Tournu, E.; Germain, J.L.

    2000-01-01

    Nuclear power plants such as VVERs can greatly benefit from innovative instrumentation to improve plant safety and efficiency. In recent years innovative instrumentation has been developed for PWRs with the aim of providing additional measurements of physical parameters on the primary and secondary circuits: the addition of new instrumentation is made possible by using non-invasive techniques such as ultrasonics and radiation detection. These innovations can be adapted for upgrading VVERs presently in operation and also in future VVERs. The following innovative instrumentation for the control, monitoring or testing at VVERs is described: 1. instrumentation for more accurate primary side direct measurements (for a better monitoring of the primary circuit); 2. instrumentation to monitor radioactivity leaks (for a safer plant); 3. instrumentation-related systems to improve the plant efficiency (for a cheaper kWh)

  14. Estimation of material degradation of VVER-1000 baffle

    Science.gov (United States)

    Harutyunyan, Davit; Koš'ál, Michal; Vandlík, Stanislav; Hojná, Anna; Schulc, Martin; Flibor, Stanislav

    2017-09-01

    The planned lifetime of the first commercial VVER-1000 units were designed for 30 to 35 years. Most of the early VVER plants are now reaching and/or passing the 35-year mark. Service life extension for another 10 to 30 years is now under investigation. Life extension requires the evaluation of pressure vessel internals degradation under long-term irradiation. One of the possible limiting factors for the service life of VVERs is a void swelling of the Russian type titanium stabilized stainless 08Ch18N10T steel used to construct the baffle surrounding the core. This article aims to show first steps towards deeper analysis of the baffle degradation process and to demonstrate the possibilities of precise calculation and measurements on the VVER-1000 mock-up in LR-0 reactor.

  15. Validation of Advanced Computer Codes for VVER Technology: LB-LOCA Transient in PSB-VVER Facility

    Directory of Open Access Journals (Sweden)

    A. Del Nevo

    2012-01-01

    Full Text Available The OECD/NEA PSB-VVER project provided unique and useful experimental data for code validation from PSB-VVER test facility. This facility represents the scaled-down layout of the Russian-designed pressurized water reactor, namely, VVER-1000. Five experiments were executed, dealing with loss of coolant scenarios (small, intermediate, and large break loss of coolant accidents, a primary-to-secondary leak, and a parametric study (natural circulation test aimed at characterizing the VVER system at reduced mass inventory conditions. The comparative analysis, presented in the paper, regards the large break loss of coolant accident experiment. Four participants from three different institutions were involved in the benchmark and applied their own models and set up for four different thermal-hydraulic system codes. The benchmark demonstrated the performances of such codes in predicting phenomena relevant for safety on the basis of fixed criteria.

  16. Fuel for new Russian reactor VVER-1200

    Energy Technology Data Exchange (ETDEWEB)

    Vasilchenko, Ivan Nikitovich [GRPress, 21, Ordzhonikidze Street, 142103 Podolsk, Moscow region (Russian Federation)

    2009-06-15

    A great program is accepted in Russia on increasing the nuclear power capacities. The basis of the program is commissioning of VVER-1200 Units of AES-2006 design. This is largely an evolutionary project of VVER-1000 reactor plant. It is referred also to reactor core. The plant electric power is increased due to increase in the reactor thermal power and forcing the main parameters and the efficiency increase. With this, reactor pressure increases from 15,7 to 16,2 MPa. The reactor inlet temperature increases from 290 deg. C to 298 deg. C, and outlet temperature from 319 deg. C to 329 deg. C. In a set of the design for four Units (2 Units at Novovoronezh NPP and 2 Units at Leningrad NPP) two base fuel cycles are developed: 5 year and 3 year. To provide such fuel cycles the fuel loading is increased by 8 tons, as compared to VVER-1000 base design, due to fuel column increase by 200 mm and change of fuel pellet sizes. In the mentioned fuel cycles the average burnup in the unloaded batch will be {approx}57 MW.day/kg U and 52 MW.day/kg U (maximum burnup over FAs is 64,5 MW.day/kg U and 60,3 MW.day/kg U), respectively. Specific consumption of natural uranium will be reduced by 5% as compared to that reached at VVER-1000 reactor. In spite of increase in Unit power the limiting permissible fuel rod linear heat rate is decreased from 448 W/cm to 420 W/cm. Refueling pattern is used with small neutron escape. The safety criteria are used that were established for VVER-1000, except for those that did not comply with EUR. For instance, the number of leaky fuel rods under accident is limited. The more stringent requirements are stated on efficiency margin of CPS rods for reactor shutdown that is ensured by the increased number of CPS rods. The well-proved design of fuel assembly TVS-2 and its close modification TVS-2M, operated at Balakovo NPP and Rostov NPP, is laid down in the basis of the core design. The load-carrying component of this structure is a rigid skeleton formed by

  17. VVER-440 and VVER-1000 reactor dosimetry benchmark - BUGLE-96 versus ALPAN VII.0

    International Nuclear Information System (INIS)

    Duo, J. I.

    2011-01-01

    Document available in abstract form only, full text of document follows: Analytical results of the vodo-vodyanoi energetichesky reactor-(VVER-) 440 and VVER-1000 reactor dosimetry benchmarks developed from engineering mockups at the Nuclear Research Inst. Rez LR-0 reactor are discussed. These benchmarks provide accurate determination of radiation field parameters in the vicinity and over the thickness of the reactor pressure vessel. Measurements are compared to calculated results with two sets of tools: TORT discrete ordinates code and BUGLE-96 cross-section library versus the newly Westinghouse-developed RAPTOR-M3G and ALPAN VII.0. The parallel code RAPTOR-M3G enables detailed neutron distributions in energy and space in reduced computational time. ALPAN VII.0 cross-section library is based on ENDF/B-VII.0 and is designed for reactor dosimetry applications. It uses a unique broad group structure to enhance resolution in thermal-neutron-energy range compared to other analogous libraries. The comparison of fast neutron (E > 0.5 MeV) results shows good agreement (within 10%) between BUGLE-96 and ALPAN VII.O libraries. Furthermore, the results compare well with analogous results of participants of the REDOS program (2005). Finally, the analytical results for fast neutrons agree within 15% with the measurements, for most locations in all three mockups. In general, however, the analytical results underestimate the attenuation through the reactor pressure vessel thickness compared to the measurements. (authors)

  18. Research on loading pattern optimization for VVER reactor

    International Nuclear Information System (INIS)

    Tran Viet Phu; Nguyen Thi Mai Huong; Nguyen Huu Tiep; Ta Duy Long; Tran Vinh Thanh; Tran Hoai Nam

    2017-01-01

    A study on fuel loading pattern optimization of a VVER reactor was performed. In this study, a core physics simulator was developed based on a multi-group diffusion theory for the use in the problem of fuel loading optimization of VVER reactors. The core simulator could handle the triangular meshes of the core and the computational speed is fast. Verification of the core simulator was confirmed against a benchmark problem of a VVER-1000 reactor. Several optimization methods such as DS, SA, TS and a combination of them were investigated and implemented in coupling with the core simulator. Calculations was performed for optimizing the fuel loading pattern of the core using these methods based on a benchmark core model in comparison with the reference core. Comparison among these methods have shown that a combination of SA+TS is the most effective for the problem of fuel loading pattern optimization. Advanced methods are being researched continuously. (author)

  19. Corrosion products behaviour under VVER primary coolant conditions

    International Nuclear Information System (INIS)

    Grygar, T.; Zmitko, M.

    2002-01-01

    The aim of this work was to collect data on thermodynamic stability of Cr, Fe, and Ni oxides, mechanisms of hydrothermal corrosion of stainless steels and to compare the real observation with the theory. We found that the electrochemical potential and pH in PWR and VVER are close to the thermodynamic boundary between two fields of stable spinel type oxides. The ways of degradation of the passivating layers due to changes in water chemistry were considered and PWR and VVER systems were found to be potentially endangered by reductive attack. In certain VVER systems the characteristics of the passivating layer on steels and also concentration of soluble corrosion products seem to be in contradiction with the theoretical expectations. (author)

  20. VVER 1000-NPP Temelin safety upgrading

    International Nuclear Information System (INIS)

    Fleischhans, J.; Ubra, O.

    1995-01-01

    A modernisation program upgrading Temelin plant to meet internationally adopted standard has been implemented during plant design and construction phases. The initial Czech-Russian design (primary system was of Russian design, secondary system was of Czech design) has been extensively modified and adapted to present western safety criteria and operational requirements. The goals are to achieve a high level of safety, reliability, availability and load-following ability. The load-following ability and response to grid frequency changes are very important for the Czech Republic, since the nuclear capacity represents a high proportion of the overall electrical system there. On the basis of IAEA OSART missions and Halliburton NUS audit results and in compliance with recommendations of The State Office for Nuclear Safety, Czech Power Company and Czech scientists and researchers a modernisation program project for Temelin has been carried out. It includes three main groups of VVER1000 MW unit innovations: - Modernization and upgrading of the safety and control systems. - Fuel replacement and modification of the reactor core. - Innovation of some components of the primary and secondary systems. The tenders for instrumentation and control system, nuclear fuel, diagnostic system and radiation monitoring system were issued to the world-well known suppliers. The US company Westinghouse Electric >Corporation (WEC) was selected to submit contract for the delivery of instrumentation and control system primary side diagnostic system and for the delivery of nuclear fuel. The contract was signed in 1993

  1. VIPRE modeling of VVER-1000 reactor core for DNB analyses

    Energy Technology Data Exchange (ETDEWEB)

    Sung, Y.; Nguyen, Q. [Westinghouse Electric Corporation, Pittsburgh, PA (United States); Cizek, J. [Nuclear Research Institute, Prague, (Czech Republic)

    1995-09-01

    Based on the one-pass modeling approach, the hot channels and the VVER-1000 reactor core can be modeled in 30 channels for DNB analyses using the VIPRE-01/MOD02 (VIPRE) code (VIPRE is owned by Electric Power Research Institute, Palo Alto, California). The VIPRE one-pass model does not compromise any accuracy in the hot channel local fluid conditions. Extensive qualifications include sensitivity studies of radial noding and crossflow parameters and comparisons with the results from THINC and CALOPEA subchannel codes. The qualifications confirm that the VIPRE code with the Westinghouse modeling method provides good computational performance and accuracy for VVER-1000 DNB analyses.

  2. Interactive nuclear plant analyzer for VVER-440 reactor

    International Nuclear Information System (INIS)

    Shier, W.; Horak, W.; Kennett, R.

    1992-05-01

    This document discusses an interactive nuclear plant analyzer (NPA) which has been developed for a VVER-440, Model 213 reactor for use in the training of plant personnel, the development and verification of plant operating procedures, and in the analysis of various anticipated operational occurrences and accident scenarios. This NPA is operational on an IBM RISC-6000 workstation and utilizes the RELAP5/MOD2 computer code for the calculation of the VVER-440 reactor response to the interactive commands initiated by the NPA operator

  3. VVANTAGE 6 - an advanced fuel assembly design for VVER reactors

    International Nuclear Information System (INIS)

    Doshi, P.K.; DeMario, E.E.; Knott, R.P.

    1993-01-01

    Over the last 25 years, Westinghouse fuel assemblies for pressurized water reactors (PWR's) have undergone significant changes to the current VANTAGE 5. VANTAGE 5 PWR fuel includes features such as removable top nozzles, debris filter bottom nozzles, low-pressure-drop zircaloy grids, zircaloy intermediate flow mixing grids, optimized fuel rods, in-fuel burnable absorbers, and increased burnup capability to region average values of 48000 MWD/MTU. These features have now been adopted to the VVER reactors. Westinghouse has completed conceptual designs for an advanced fuel assembly and other core components for VVER-1000 reactors known as VANTAGE 6. This report describes the VVANTAGE 6 fuel assembly design

  4. Detailed modeling of KALININ-3 NPP VVER-1000 reactor pressure vessel by the coupled system code ATHLET/BIPR-VVER

    International Nuclear Information System (INIS)

    Nikonov, S.P.; Velkov, K.; Pautz, A.

    2011-01-01

    The paper gives an overview of the recent developments of a new reactor pressure vessel (RPV) model of VVER-1000 for the coupled system code ATHLET/BIPR-VVER. Based on the previous experience a methodology is worked out for modeling the RPV in a pseudo-3D way with the help of a multiple parallel thermal-hydraulic channel scheme that follows the hexagonal fuel assembly structure from the bottom to the top of the reactor. The results of the first application of the new modeling are discussed on the base of the OECD/NEA coupled code benchmark for Kalinin-3 NPP transient. Coolant mass flow distributions in reactor volume of VVER 1000 reactor are presented and discussed. It is shown that along the core height a mass flow re-distribution of the coolant takes place starting approximately at an axial layer located 1 meter below the core outlet. (author)

  5. Experience on KKNPP VVER 1000 MWe water chemistry

    International Nuclear Information System (INIS)

    Ganesh, S.; Selvaraj, S.; Balasubramanian, M.R.; Selvavinayagam, P.; Pillai, Suresh Kumar

    2015-01-01

    Kudankulam Nuclear Power Project consists of pressurized water reactor (VVER) 2 x 1000 MWe constructed in collaboration with Russian Federation at Kudankulam in Tirunelveli District, Tamilnadu. Unit - 1 attained criticality on July 13 th 2013 and the unit was synchronized to grid on 22 nd October 2013. This paper highlights experience gained on water chemistry regime for primary and secondary circuit. (author)

  6. ASTEC applications to VVER-440/V213 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Matejovic, Peter, E-mail: ivstt@nextra.sk; Barnak, Miroslav; Bachraty, Milan; Vranka, Lubomir

    2014-06-01

    Since the beginning of ASTEC development by IRSN and GRS the code was widely applied to VVER reactors. In this paper, at first specific features of VVER-440/V213 reactor design that are important from the modelling point of view are briefly described. Then the validation of ASTEC code with focus on its applicability to VVER reactors is briefly summarised and the results obtained with the ASTEC V2.0-rev1 version for the ISP-33 PACTEL natural circulation experiment are presented. In the next section the application of ASTEC V2.0-rev1 code in upgrade of VVER-440/V213 NPPs to cope with consequences of severe accidents is described. This upgrade includes adoption of in-vessel retention via external reactor vessel cooling and installation of large capacity passive autocatalytic recombiners. Results of analysis with focus on corium localisation and stabilisation inside reactor vessel, hydrogen control in confinement and prevention of long-term confinement pressurisation are presented.

  7. Investigations of SPND noise signals in VVER-440 reactors

    International Nuclear Information System (INIS)

    Kiss, S.; Lipcsei, S.; Hazi, G.

    2001-01-01

    This paper describes and characterises SPND noise measurements of an operating VVER-440 nuclear reactor. Characteristics of the signal can be radically influenced by the geometrical properties of the detector and the cable and by the measuring arrangement. Structure of phase spectra showing propagating perturbations measured on uncompensated SPN detectors is studied through models.(author)

  8. Primary water chemistry for NPP with VVER-TOI

    International Nuclear Information System (INIS)

    Susakin, S.N.; Brykov, S.I.; Zadonsky, N.V.; Bystrova, O.S.

    2012-09-01

    Nowadays within the framework of development of the nuclear power industry in Russia the VVER-TOI reactor is under designing (Standard optimized design). The given design provides for improvement of operation safety level, of technical-economic, operational and load-follow characteristics, and for the raise of competitive capacity of reactor plant and NPP as a whole. In VVER-TOI reactor plant design the primary water chemistry has been improved considering operation experience of VVER reactor plants and a possibility of RP operation under load-follow modes from the viewpoint of meeting the following requirements: - suppression of generation of oxidizing radiolytic products under power operation; - assurance of corrosion resistance of structural materials of equipment and pipelines throughout the NPP design service life; - minimization of deposits on surfaces of the reactor core fuel rods and on heat exchange surface of steam generators; - minimization of accumulation of activated corrosion products; - minimization of the amount of radioactive processing waste. In meeting these requirements an important role is devoted to suppression of generation of oxidizing radiolytic products owing to accumulation of hydrogen in the primary coolant. At NPP with VVER-1000 reactor the ammonia-potassium water chemistry is used wherein the hydrogen accumulation is provided at the expense of ammonia proportioning. Usage of ammonia leads to generation of additional amount of radioactive processing waste and to increased irregularity of maintaining the water chemistry under the daily load-follow modes. In VVER TOI design the primary water chemistry is improved by replacing the proportioning of ammonia with the proportioning of gaseous hydrogen. Different process schemes were considered that provide for a possibility of hydrogen accumulation and maintaining owing to direct proportioning of gaseous hydrogen. The obtained results showed that transition to the potassium water chemistry

  9. Mezivýměník tepla primárního okruhu reaktoru ESFR

    OpenAIRE

    Švihel, Miroslav

    2012-01-01

    Práce je zaměřena především na návrh mezivýměníku tepla primárního okruhu reaktoru ESFR. Výměník je počítán tepelně, hydraulicky i pevnostně a závěrem je zpracována část výkresové dokumentace. Jsou navrhnuty základní rozměry trubkového svazku a nádoby výměníku. Je zde obsažen i přehled koncepcí výměníků a doposud použitých typů mezivýměníku na jaderných elektrárnách s rychlými reaktory. Uvedeny jsou zde i základní parametry projektu ESFR a zhodnocena bezpečnost a provozní spolehlivost výměník...

  10. Severe accident management development program for VVER-1000 and VVER-440/213 based on the westinghouse owners group approach

    International Nuclear Information System (INIS)

    Felix, E.; Dessars, N.

    2003-01-01

    The development of the Westinghouse Owners Group Severe Accident Management Guidelines (WOG SAMG) between 1991 and 1994 was initiated in response to the U.S. Nuclear Regulatory Commission (NRC) requirement for addressing the regulatory severe accident concerns. Hence, the WOG SAMG is designed to interface with other existing procedures at the plant and is used in accident sequences that have progressed to the point where these other procedures are not applicable any longer, i.e. following core damage. The primary purpose of the WOG SAMG is to reach a controlled stable state, which can be declared when fission product releases are controlled, challenges to the confinement fission product boundary have been mitigated, and adequate heat removal is provided to the core and the containment. Although the WOG SAMG is a generic severe accident management guidance developed for use by the entirety of the operating Westinghouse PWR plants, provisions have been made in their development to address specific features of individual plants such as confinement type and the feasibility of reactor cavity flooding. Similarly, the generic SAMG does not address unique plant features and equipment, but rather allows for consideration of plant specific features and strategies. This adaptable approach has led to several SAMG development programs for VVER-1000 and VVER-440 type of power plants, under Westinghouse' s lead. The first of these programs carried out to completion was for Temelin NPP - VVER-1000 - in the first quarter of 2003. Other ongoing programs aim at providing a similar work for VVER-440 design, namely Dukovany, Mochovce and Bohunice NPPs. The challenge of adapting the existing generic WOG material to plants other than PWRs mainly arises for VVER-440 because of important differences in confinement design, making it more vulnerable to ex-vessel phenomena such as hydrogen burn. Also, for both eastern designs, cavity flooding strategy requires special consideration and

  11. Aplikace pro penetrační testování webových zranitelností typu Data Validation flaws

    OpenAIRE

    Němec, Václav

    2011-01-01

    Tato bakalářské práce se zabývá detekcí webových zranitelností typu Data validation flaws. V práci jsou popsány běžné útoky, obrana před těmito útoky a postupy při automatické detekci. Hlavním cílem je návrh a implementace nástroje pro automatickou detekci zranitelností typu Data validation flaws, jeho otestování a srovnání výsledků s podobnými nástroji jako například Paros Proxy nebo Burp Suite. This bachelor's thesis deals with detection of web vulnerabilities such as data validation fla...

  12. Vliv léčby kyselinou acetylosalycilovou na makrovaskulární reaktivitu u diabetiků 2.typu

    Czech Academy of Sciences Publication Activity Database

    Prázdný, M.; Hrach, Karel; Kasalová, Z.; Škrha, J.; Zvárová, Jana

    2005-01-01

    Roč. 8, 1 Suppl. (2005), s. 41 ISSN 1211-9326. [Diabetologické dny. 21.04.2005-23.04.2005, Luhačovice] R&D Projects: GA MŠk LN00B107 Institutional research plan: CEZ:AV0Z10300504 Keywords : diabetes mellitus 2. typu * mikrovaskulární reaktivita Subject RIV: BB - Applied Statistics, Operational Research

  13. An experimental investigation of 1% SBLOCA on PSB-VVER test facility

    Energy Technology Data Exchange (ETDEWEB)

    Lipatov, I.A.; Dremin, G.I.; Galtchanskaia, S.A.; Gorbunov, Yu.S. [Electrogorsk Research and Engineering Center, EREC, Electrogorsk (Russian Federation); Elkin, I.V. [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation)

    2001-07-01

    The paper presents the results of the three tests carried out in the PSB-VVER large-scale integral test facility. The PSB-VVER test facility is a four loop, full pressure scaled down model bearing structural similarities to the primary system of the NRP with VVER-1000 Russian design reactor. Volume-power scale is 1/300 while elevation scale is 1/1. (orig.)

  14. SCORPIO-VVER core monitoring and surveillance system with advanced capabilities

    International Nuclear Information System (INIS)

    Molnar, J.; Vocka, R.

    2010-01-01

    In this work authors present 12 years of operation experience of core monitoring and surveillance system with advanced capabilities on nuclear power plants on 6 unit of VVER-440 type of reactors at two different NPPs. The original version of the SCORPIO (Surveillance of reactor CORe by PIcture On-line display) system was developed for the western type of PWR reactors. The first version of the SCORPIO-VVER Core Monitoring System for Dukovany NPP (VVER-440 type of reactor, Czech Republic) was developed in 1998. For SCORPIO-VVER implementation at Bohunice NPP in Slovakia (2001) the system was enhanced with startup module KRITEX.

  15. Present status and recent improvements of water chemistry at Russian VVER plants

    International Nuclear Information System (INIS)

    Mamet, V.; Yurmanov, V.

    2001-01-01

    Water chemistry is an important contributor to reliable plant operation, safety barrier integrity, plant component lifetime, radiation safety, environmental impact. Primary and secondary water chemistry guidelines of Russian VVER plants have been modified to meet the new safety standards. At present 14 VVER units of different generation are in operation at 5 Russian NPPs. There are eight 4-loop pressurised water reactors VVER-1000 (1000 MWe) and six 6-loop pressurised water reactors VVER-440 (440 MWe). Generally, water chemistry at East European VVER plants (about 40 VVER-440 and VVER-1000 units in Ukraine, Bulgaria, Slovakia, Czech Republic, Hungary, Finland and Armenia) is similar to water chemistry at Russian VVER plants. Due to similar design and structural materials some water chemistry improvements were introduced at East European plants after they has been successfully implemented at Russian plants and vice versa. Some water chemistry improvements will be implemented at modern VVER plants under construction in Ukraine, Slovakia, Czech Republic, Iran, China, India. (R.P.)

  16. Dosimetry of VVER-1000 Reactor Pressure Vessel and Surveillance Specimens as a part of PLiM at Ukrainian NPPs

    International Nuclear Information System (INIS)

    Bukanov, V.N.; Diemokhin, V.L.; Grytsenko, O.V.; Ilkovych, V.V.; Pugach, A.M.; Pugach, S.M.; Vasylieva, O.G.; Vyshnevskyi, I.M.; Kasatkin, O.G.

    2012-01-01

    A regular surveillance program for VVER-1000 and its shortages are described. The Methodology for determination of neutron flux functionals on surveillance specimens of VVER-1000 pressure vessel is presented. The radiation exposure monitoring system for VVER-1000 pressure vessel is described. The main principles of an additional surveillance program for VVER-1000 are presented. The Dosimetry Experiment, which is already carrying out at Unit 3 of Rivne NPP, is described. (author)

  17. Seismic verification methods for structures and equipment of VVER-type and RBMK-type NPPs (summary of experiences)

    International Nuclear Information System (INIS)

    Masopust, R.

    2003-01-01

    The main verification methods for structures and equipment of already existing VVER-type and RBMK-type NPPs are briefly described. The following aspects are discussed: fundamental seismic safety assessment principles for VVER/RBMK-type NPPs (seismic safety assessment procedure, typical work plan for seismic safety assessment of existing NPPs, SMA (HCLPF) calculations, modified GIP (GIP-VVER) procedure, similarity of VVER/RBMK equipment to that included in the SQUG databases and seismic interactions

  18. Improving nuclear safety of VVER-440 units

    International Nuclear Information System (INIS)

    Nochev, T.; Sabinov, S.

    2001-01-01

    In this paper authors deals with improvement of nuclear safety of WWER-440 units in Kozloduy NPP. Main directions for improving nuclear safety of WWER-440 units were: - to expand number of the design accident; - to increase reliability of equipment important for the safety; - to decrease the probability of initiating events; - improvements the integrity of the primary circuit (application LBB concept, qualification of the pressure safety valves to avoid pressurized thermal shock); - improvement of the fire protection; - improvement of the operation including upgrading and improvement of operational documents, implementation of new system for training the operators and etc.; - reassessment of the seismic response of the plant. Main actions were made at NPP Kozloduy to increase nuclear safety of VVER-440 units. 1. Modernization of Emergency High Pressure Safety Injection System. The modernization includes dividing of independent channels with reservation of active elements. Pumps were exchanged with more effective and reliable ones. HPSIS was increased reliability in general through decrease number of active elements and exchanged with passive. 2. For the purpose of avoiding fast cooling at the primary circuit and obtaining thermal shock of reactor vessel, Main Safety Insulation Valves are installed at NPP Kozloduy. 3. Modernization of Emergency power supplies AC. Oil breakers VMP-10 are exchanged with gas FS-4. 4. Generator breakers are installed to decrease probability of loss power supply and blackout. They provide reliable power supply to the system important for the safety in case of failure on generator. 5. I and C system has been qualified and optimized. 6. Reassessments of Limiting Conditions of Operation and new scram signals have been introduced. 7. An operators-oriented Informational System has been developed. It includes ensuring and updating of equipment data, new informational support of operator and etc. 8. A new auxiliary independent system for

  19. Innovated feed water distributing system of VVER steam generators

    International Nuclear Information System (INIS)

    Matal, O.; Sousek, P.; Simo, T.; Lehota, M.; Lipka, J.; Slugen, V.

    2000-01-01

    Defects in feed water distributing system due to corrosion-erosion effects have been observed at many VVER 440 steam generators (SG). Therefore analysis of defects origin and consequently design development and testing of a new feed water distributing system were performed. System tests in-situ supported by calculations and comparison of measured and calculated data were focused on demonstration of long term reliable operation, definition of water flow and water chemical characteristics at the SG secondary side and their measurements and study of dynamic characteristics needed for the innovated feed water distributing system seismic features approval. The innovated feed water distributing system was installed in the SGs of two VVER units already. (author)

  20. Operational benchmark for VVER-1000, unit 6, Kozloduy NPP

    International Nuclear Information System (INIS)

    Apostolov, T.; Petrov, B.

    1999-01-01

    Benchmark calculations have been carried out using the 3D nodal code TRAPEZ. Global neutron-physics characteristics of the VVER-1000 core, Kozloduy NPP Unit 6, have been determined taking into account the real loading patterns and operational history of the first three cycles. The code TRLOAD has been used to perform the fuel reloading between any two cycles. The reactor and components descriptions as well as material compositions are given. The results presented include the critical boric acid concentration, the radial power distribution, the axial power distribution for the maximum overload assembly, and the burnup distribution at three different moments during each cycle. Calculated values have been compared with measured data. It is shown that the results obtained by the TRAPEZ code are in good agreement with the experimental data. The information presented could serve as a test case for validation of code packages designed for analyzing the steady-state operation of VVERs. (author)

  1. Interactive nuclear plant analyzer for the VVER-440 reactor

    International Nuclear Information System (INIS)

    Shier, W.; Kennett, R.

    1993-01-01

    An interactive nuclear plant analyzer (NPA) has been developed for a VVER-440 model 213 reactor for use in the training of plant personnel, the development and verification of plant operating procedures, and in the analysis of various anticipated operational occurrences and accident scenarios. This NPA is operational on an IBM RISC-6000 workstation and utilizes the RELAP5/MOD2 computer code for the calculation of the VVER-440 reactor response to the interactive commands initiated by the NPA operator. Results of the interactive calculation can be through the user-defined, digital display of various plant parameters and through color changes that reflect changes in primary system fluid temperatures, fuel and clad temperatures, and the temperatures of other metal structures. In addition, changes in the status of various components and system can be initiated and/or displayed both numerically and graphically on the mask

  2. Special features of embrittlement of welded joints in shells of VVER-type reactors

    International Nuclear Information System (INIS)

    Kasatkin, O.G.

    1999-01-01

    At present, the atomic power engineering of Russia and Ukraine is based on water-water energy reactors of the VVER-440 and VVER-1000 type, with the electric power of 440 and 1000 MW, respectively. The majority of the VVER-440 reactors are installed in Russia, and VVER-1000 reactors operate in Ukraine. The reactors' shell (RS) is produced from cylindrical shells and a dished end welded together by circular joints under a flux. The RS of the VVER-440 reactor is produced from 15Kh3MFA steel, and the VVER-1000 reactors are produced from 15Kh2NMFA steel. The shell of the VVER-1000 reactor has an internal austenite coating. The condition of the RS metal is determined mainly by the critical brittleness temperature T b at which the impact toughness of specimens with a sharp notch reaches 60 J/cm 2 . The energy reactors, working in western countries, are characterised by a service life of 40 years and discussion is being carried out to extend this lifetime to 60 years. The design service life of the domestic reactors varies from 30 (RS VVER-440) to 40 (RS VVER-1000) years. According to investigations, the service life of the shells of these reactors is restricted by the properties of welded joints which are characterised by higher susceptibility to embrittlement than that of the parent metal, especially due to a higher content in the weld of phosphorus (RS VVER-440) or nickel (RS VVER-1000). Therefore, some experts believe that the actual service life of the RS is shorter than the design life. The accurate evaluation of the service life of welded joints in the RS is very important for the safety of service and also in the economic aspects, because the unjustified decrease of the permissible service life and premature shutdown of units of the nuclear power station result in huge losses

  3. Evaluation of reliability of EC inspection of VVER SG tubes

    International Nuclear Information System (INIS)

    Stanic, D.

    2001-01-01

    Evaluation of eddy current data collected during inspection of VVER steam generators is very complex task because of numerous parameters which have affect on eddy current signals. That was the reason that recently ago INETEC has started related scientific project in order to evaluate the reliability of eddy current (EC) inspection of VVER steam generator (SG) tubing. In the scope of project the following objectives will be investigated: 1. Determination of POD (Probability of detection) of various types degradation cracks, where their basic parameters are variables (basic parameters are depth, length, width, orientation, number) on three different sets of tubes (clean ideal tubes, tubes with pilgering, tubes electroplated with copper) 2. Sizing quality (accuracy, repeatability) (same data sets as defined in 1.) 3. Effect of fill factor on POD and sizing quality. 4. Effect of tube bends on POD and sizing quality. 5. Effect of other tube geometry variations on POD and sizing quality (tube ovality, transition zone region, expanded (rolled) part of tube, dents, dings). Investigation will start with bobbin probe technique which is the most used technique for general purpose VVER tube examination. Since INETEC is the only world company which successfully developed and applied rotating probe technique for VVER SG tubes, scope of the project will be extended on rotating probe technique utilizing 'pancake' and 'point' coil. Method reliability will be investigated first on the huge set of EDM notches representing various defect morphologies and simulating different factors, and the second part will be investigated on sets of degradation defects obtained by artificial corrosion. In the scope of the project the measures for enhancing the method reliability have to be determined. This considers the proper definition of parameters of examination system, as well as establishment of the suitable analysis procedures. This article presents the temporary results of the first part of

  4. Application of a Russian nuclear reactor simulator VVER-1000

    International Nuclear Information System (INIS)

    Lopez-Peniche S, A.; Salazar S, E.

    2012-10-01

    The objective of the present work is to give to know the most important characteristics in the Russian nuclear reactor of pressurized light water VVER-1000, doing emphasis in the differences that has with the western equivalent the reactor PWR in the design and the safety systems. Therefore, a description of the computerized simulation of the reactor VVER-1000 developed by the company Eniko TSO that the International Atomic of Energy Agency distributes to the states members with academic purposes will take place. The simulator includes mathematical models that represent to the essential systems in the real nuclear power plant, for what is possible to reproduce common faults and transitory characteristic of the nuclear industry with a behavior sufficiently attached to the reality. In this work is analyzed the response of the system before a turbine shot. After the accident in the nuclear power plant of Three Mile Island (US) they have been carried out improvements in the design of the reactor PWR and their safety systems. To know the reach and the limitations of the program, the events that gave place to this accident will be reproduced in the simulator VVER-1000. With base to the results of the simulation we will conclude that so reliable is the response of the safety system of this reactor. (Author)

  5. Stresses in transition region of VVER-1000 reactor vessels

    Energy Technology Data Exchange (ETDEWEB)

    Namgung, I. [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of); Nguye, T.L. [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of); National Research Inst. of Mechanical Engineering, Hanoi City, Vietnam (China)

    2014-07-01

    Most of the western PWR reactor's bottom head is hemi-spherical shape, however for Russian designed VVER family of reactor it is ellipsoidal shape. The transition region from shell side to ellipsoidal head and transition top flange to cylindrical shell develop higher stress concentration than western PWR reactor vessel. This region can be modeled as conical shell with varying thickness. The theoretical derivation of stress in the thick-walled conical cylinder with varying thickness was developed and shown in detail. The results is applied to VVER-1000 reactor vessel of which shell to bottom ellipsoidal shell and shell to upper flange were investigated for stress field. The theoretical calculations were also compared with FEM solutions. An axisymmetric 3D model of VVER-1000 reactor vessel (without closure head) FEM model was created and internal hydrostatic pressure boundary condition was applied. The stress results from FEM and theoretical calculation were compared, and the discrepancies and accuracies of the theoretical results were discussed. (author)

  6. Stresses in transition region of VVER-1000 reactor vessels

    International Nuclear Information System (INIS)

    Namgung, I.; Nguye, T.L.

    2014-01-01

    Most of the western PWR reactor's bottom head is hemi-spherical shape, however for Russian designed VVER family of reactor it is ellipsoidal shape. The transition region from shell side to ellipsoidal head and transition top flange to cylindrical shell develop higher stress concentration than western PWR reactor vessel. This region can be modeled as conical shell with varying thickness. The theoretical derivation of stress in the thick-walled conical cylinder with varying thickness was developed and shown in detail. The results is applied to VVER-1000 reactor vessel of which shell to bottom ellipsoidal shell and shell to upper flange were investigated for stress field. The theoretical calculations were also compared with FEM solutions. An axisymmetric 3D model of VVER-1000 reactor vessel (without closure head) FEM model was created and internal hydrostatic pressure boundary condition was applied. The stress results from FEM and theoretical calculation were compared, and the discrepancies and accuracies of the theoretical results were discussed. (author)

  7. Greifswald and Rheinsberg: East European VVERs with a new mission

    International Nuclear Information System (INIS)

    Sterner, H.; Leushacke, D.; Rittscher, D.

    1995-01-01

    Plans for the decommissioning of the VVER reactors at Greifswald and Rheinsberg in the former German Democratic Republic are described. The decision to decommission the eight VVER-440s (two of which are still under construction) at Greifswald and the Russian prototype PWR, VVER-2, at Rheinsberg, was taken because of a lack of public acceptance and financial considerations. Three main phases are scheduled for the project. The first, the post-operation phase, has already commenced at both sites. It involves: the operation of all systems needed safely to store fuel elements, to remove them and to condition operational waste; dismantling of mainly inactive, systems not needed for fuel handling; and system decontamination. The second phase comprises the dismantling of contaminated systems, remote dismantling and conditioning of the dismantled material. Finally, in the site restoration phase, following the dismantling of remaining systems and building decontamination and demolition, the site will be adapted for other uses. Three projects for new site use at Greifswald being pursued. As a first step, an international Technology Centre is to be created to collect and evaluate information on decommissioning experience. The centre will require international financial support and co-ordination. New facilities are already being constructed to deal with decontamination of dismantled materials and the interim storage of radioactive waste and reactor fuel. (UK)

  8. Improving the VVER-440 fuel design and technology

    International Nuclear Information System (INIS)

    Aksenov, P.; Bondar, Y.; Kolosovsky, Y.; Kochergin, V.; Luzan, Y.; Malakhov, A.; Krapivtsev, V.; Bauman, N.; Shumeev, A.; Filippov, V.

    2009-01-01

    Operational performance of VVER-440 fuel has long been demonstrating good reliability of the fuel. However, assembly failures occur, and fuel suppliers should always take measures to maintain its reliability. For several years, OAO MSZ has been fabricating working assemblies with detachable shrouds and removable fuel rods. The next step is the supply of demountable assemblies to allow inspection or repair of fuel rods after removal of the shroud. With the help of corresponding program the Russian organizations have carried out research and development work to advance and study operational features of demountable VVER-440 CFAs. The main engineering solutions are consistent with the working assemblies. The pilot demountable CFAs are running in the Kola-4 core. The obtained results can be used when deciding on the demountable CFAs delivery issues. The experiment-calculated research results of coolant mixing in the present design VVER-440 have been analysed. It has been found out that coolant mixing in the WA head is incomplete and that is why leading to conservatism when determining the reactor operational limits. The proposed WA head design includes an upgraded bumper grid with additional planes intensifying coolant mixing in the head. The bumper grid drawing and a pilot model is available. The thermohydraulics and rigidity features of the proposed design have been studied by experiment-calculated methods

  9. Contributions of Modranska potrubni a.s. to the safety improvement of piping systems and valves of NPS type VVER 440 and VVER 1000

    International Nuclear Information System (INIS)

    Slach, J.

    2004-01-01

    The following activities are described: (i) Installation of pipe whip restraints on piping for high pressure and temperature steam and feed piping; (ii) Installation of air receivers for quick-acting valves with air actuator on VVER 440 units at the Jaslovske Bohunice V2 NPP; (iii) Replacement of the technical water distribution system material in the reactor hall of the Temelin VVER 1000 units; Installation of measuring nozzles on main steam piping DN 600 at the Temelin VVER 1000 units. (P.A.)

  10. Developing the European Center of Competence on VVER-Type Nuclear Power Reactors

    Science.gov (United States)

    Geraskin, Nikolay; Pironkov, Lyubomir; Kulikov, Evgeny; Glebov, Vasily

    2017-01-01

    This paper presents the results of the European educational projects CORONA and CORONA-II which are dedicated to preserving and further developing nuclear knowledge and competencies in the area of VVER-type nuclear power reactors technologies (Water-Water Energetic Reactor, WWER or VVER). The development of the European Center of Competence for…

  11. Diversification of the VVER fuel market in Eastern Europe and Ukraine

    Energy Technology Data Exchange (ETDEWEB)

    Kirst, Michael [Westinghouse EMEA, Brussels (Belgium); Benjaminsson, Ulf; Oenneby, Carina [Westinghouse Electric Sweden AB, Vaesteraes (Sweden)

    2015-03-15

    There are a total of 33 VVER active reactors in the EU and Ukraine, accounting for the largest percentage of the total electricity supply in the countries operating these. The responsible governments and utilities operating these units want too see an increased diversification of the nuclear fuel supply. Westinghouse is the only nuclear fuel producer outside Russia, which has taken the major steps to develop, qualify and manufacture VVER fuel designs - both for VVER-440 and VVER-1000 reactors. The company has delivered reloads of VVER-440 fuel to Loviisa 2 in Finland, VVER-1000 fuel for both the initial core and follow-on regions to Temelin 1-2 in the Czech Republic and more recently reloads of VVER-1000 fuel to South Ukraine 2-3. Technical challenges in form of mechanical interference with the resident fuel have been encountered in Ukraine, but innovative solutions have been developed and successfully implemented and today Ukraine has, for the first time in its history, a viable VVER-1000 fuel design alternative, representing a tremendous lever in energy security for the country.

  12. Validation of containment thermal hydraulic computer codes for VVER reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jiri Macek; Lubomir Denk [Nuclear Research Institute Rez plc Thermal-Hydraulic Analyses Department CZ 250 68 Husinec-Rez (Czech Republic)

    2005-07-01

    Full text of publication follows: The Czech Republic operates 4 VVER-440 units, two VVER-1000 units are being finalized (one of them is undergoing commissioning). Thermal-hydraulics Department of the Nuclear Research Institute Rez performs accident analyses for these plants using a number of computer codes. To model the primary and secondary circuits behaviour the system codes ATHLET, CATHARE, RELAP, TRAC are applied. Containment and pressure-suppression system are modelled with COCOSYS and MELCOR codes, the reactor power calculations (point and space-neutron kinetics) are made with DYN3D, NESTLE and CDF codes (FLUENT, TRIO) are used for some specific problems.An integral part of the current Czech project 'New Energy Sources' is selection of a new nuclear source. Within this and the preceding projects financed by the Czech Ministry of Industry and Trade and the EU PHARE, the Department carries and has carried out the systematic validation of thermal-hydraulic and reactor physics computer codes applying data obtained on several experimental facilities as well as the real operational data. One of the important components of the VVER 440/213 NPP is its containment with pressure suppression system (bubble condenser). For safety analyses of this system, computer codes of the type MELCOR and COCOSYS are used in the Czech Republic. These codes were developed for containments of classic PWRs or BWRs. In order to apply these codes for VVER 440 systems, their validation on experimental facilities must be performed.The paper provides concise information on these activities of the NRI and its Thermal-hydraulics Department. The containment system of the VVER 440/213, its functions and approaches to solution of its safety is described with definition of acceptance criteria. A detailed example of the containment code validation on EREC Test facility (LOCA and MSLB) and the consequent utilisation of the results for a real NPP purposes is included. An approach to

  13. Application of a Russian nuclear reactor simulator VVER-1000; Aplicacion de un simulador de reactor nuclear ruso VVER-1000

    Energy Technology Data Exchange (ETDEWEB)

    Lopez-Peniche S, A. [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04360 Mexico D. F. (Mexico); Salazar S, E., E-mail: alpsordo@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, 62250 Jiutepec, Morelos (Mexico)

    2012-10-15

    The objective of the present work is to give to know the most important characteristics in the Russian nuclear reactor of pressurized light water VVER-1000, doing emphasis in the differences that has with the western equivalent the reactor PWR in the design and the safety systems. Therefore, a description of the computerized simulation of the reactor VVER-1000 developed by the company Eniko TSO that the International Atomic of Energy Agency distributes to the states members with academic purposes will take place. The simulator includes mathematical models that represent to the essential systems in the real nuclear power plant, for what is possible to reproduce common faults and transitory characteristic of the nuclear industry with a behavior sufficiently attached to the reality. In this work is analyzed the response of the system before a turbine shot. After the accident in the nuclear power plant of Three Mile Island (US) they have been carried out improvements in the design of the reactor PWR and their safety systems. To know the reach and the limitations of the program, the events that gave place to this accident will be reproduced in the simulator VVER-1000. With base to the results of the simulation we will conclude that so reliable is the response of the safety system of this reactor. (Author)

  14. Primary LOCA in VVER-1000 by pressurizer PORV failure

    International Nuclear Information System (INIS)

    Sabotinov, L.; Lutsanych, S.; Kadenko, I.

    2013-01-01

    The paper presents the calculations and analysis of the design basis accident of a standard VVER-1000/V320 reactor with inadvertent opening and stuck in open position of the pressurizer pilot operated relief valve (PORV). The objective is the independent assessment of this accident applying the French best estimate thermal-hydraulic computer code CATHARE 2 and verification to meet the safety criteria for such kind of the accident. The 'Inadvertent opening and stuck in open position of PORV' is a design basis accident classified as Medium Break Loss of Coolant Accident (MB LOCA) with the equivalent diameter of the break D- 68 mm. This accident is particularly interesting to be calculated and analyzed, because it took place at operating NPP with VVER-1000 reactors (Rovno NPP) in 2009. The calculations have been carried out with conservative conditions as usual for DBA analysis. The NPP model corresponds to a real VVER-1000/V320 configuration and comprises all safety systems, adopted for one of the latest CATHARE 2 versions. The results of CATHARE 2 calculations are compared with available results of RELAP5 calculations. There is similarity of the thermal-hydraulic parameters behavior, but also some differences can be observed basically due to the break flow prediction, which causes differences in primary pressure evaluation. Both calculations show that there is no boiling crisis in the reactor core and reliable cooldown is achieved. The calculations performed with CATHARE2 code demonstrate the ability of the code to predict reasonably the break flow, pressures, temperatures etc. for considered LOCA scenario and to be applied for safety studies

  15. New code for VVER-440 loading pattern design

    International Nuclear Information System (INIS)

    Bajgl, J.; Lehmann, M.

    1999-01-01

    This paper describes the main attributes of a new computer program OPTIMAL used for loading pattern design in Dukovany NPP (4 reactors VVER-440). We have been developed this program in Nuclear Research Institute Rez since 1994 on the base of special contract between Dukovany NPP and Nuclear Research Institute Rez. General information about the optimisation methodology is given in the first part. The organisation of the optimisation process is described in part 2. Construction of the optimisation functional is shown in part 3. Procedures used during one-cycle optimisation are described in part 4. (Authors)

  16. Radionuclide inventory calculation in VVER and BWR reactor

    International Nuclear Information System (INIS)

    Bouhaddane, A.; Farkas, F.; Slugen, V.; Ackermann, L.; Schienbein, M.

    2014-01-01

    The paper shows different aspects in the radionuclide inventory determination. Precise determination of the neutron flux distribution, presented for a BRW reactor, is vital for the activation calculations. The precision can be improved utilizing variance reduction methods as importance treatment, weight windows etc. Direct calculation of the radionuclide inventory via Monte Carlo code is presented for a VVER reactor. Burn-up option utilized in this calculation appears to be proper for reactor internal components. However, it will not be probably effective outside the reactor core. Further calculations in this area are required to support the forth-set findings. (authors)

  17. Assessment of the recovery annealing efficiency for VVER-1000 materials' structure reset and lifetime extension

    International Nuclear Information System (INIS)

    Gurovich, B.; Kuleshova, E.; Prikhodko, K.; Fedotova, S.

    2011-01-01

    The results of the VVER-1000 reactor pressure vessels welds studies based on the surveillance specimens sets have revealed a high embrittlement rate of steel with high nickel content compared with predicted embrittlement determined from the Russian Guide. For these critical vessels further safe operation (even during design service life) is not allowed without additional measures (recovery annealing of the VVER-1000 welds as earlier for VVER- 440). The reason is that the rate of high nickel VVER-1000 welds embrittlement is significantly higher than that is for base metal. In order to solve a problem of VVER-1000 lifetime extension recovery annealing validation and accelerated reirradiation of specimens for prolonged operation period estimation after annealing were necessary. In this work comparison of electron-microscopy fine structure studies and fractographic studies of Charpy specimens fracture surface of the VVER-1000 high nickel welds in different states were carried out. It allows estimation of the recovery annealing effect on steels structure and its behavior at further operation. It is shown that both secondary and primary irradiation causes alike radiation-induced fine structure changes: dislocation loops and nano-size precipitates. Recovery annealing leads to full dislocation loops dissolution and significant nano-size precipitates solution but not to the initial values. The rate of radiation defects and radiation-induced precipitates accumulation at reirradiation weld after recovery annealing is lower than at primary irradiation and determine the lower secondary embrittlement rate of VVER-1000 weld. (authors)

  18. VVER-1000 small-medium break LOCAs predictions by ASTEC

    International Nuclear Information System (INIS)

    Georgieva, J.; Stefanova, A.; Atanasova, B.; Groudev, P.; Tusheva, P.; Mladenov, I.; Dimov, D.; Passalacqua, R.

    2005-01-01

    This paper deals with an assessment of ASTEC1.1v0 code in the simulation of small and medium break LOCAs (ranging from 30mm up to 70mm equivalent diameter). The reference power plant for this analysis is a VVER-1000/V320 (e.g. Units 5 and 6 at Kozloduy NPP). A preliminary comparison with MELCOR and RELAP-SCDAP severe accident codes will be discussed. This investigation has been performed in the framework of the SARNET project (under the Euratom 6 th framework program) by the FoBAUs group (Forum of Bulgarian ASTEC users). The FoBAUs group aims at the validation of the ASTEC code in the field of severe accidents. Future activities will target the ASTEC capability (as a PSA-level 2 tool) to simulate a large range of reactor accident scenarios with intervention of safety systems (either passive systems or operated by operators). The final target is to assess Severe Accident Management (SAM) procedures for VVER-1000 reactors. The ASTEC1.1v0 code version here used is the one released in June 2004 by the French IRSN (Institut de Radioprotection et de Surete Nucleaire) and the German GRS (Gesellschaft ReactorSicherheit mbH). (author)

  19. Microstructure and embrittlement of VVER 440 reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Hennion, A.

    1999-03-01

    27 VVER 440 pressurised water reactors operate in former Soviet Union and in Eastern Europe. The pressure vessel, is made of Cr-Mo-V steel. It contains a circumferential arc weld in front of the nuclear core. This weld undergoes a high neutron flux and contains large amounts of copper and phosphorus, elements well known for their embrittlement potency under irradiation. The embrittlement kinetic of the steel is accelerated, reducing the lifetime of the reactor. In order to get informations on the microstructure and mechanical properties of these steels, base metals, HAZ, and weld metals have been characterized. The high amount of phosphorus in weld metals promotes the reverse temper embrittlement that occurs during post-weld heat treatment. The radiation damage structure has been identified by small angle neutron scattering, atomic probe, and transmission electron microscopy. Nanometer-sized clusters of solute atoms, rich in copper with almost the same characteristics as in western pressure vessels steels, and an evolution of the size distribution of vanadium carbides, which are present on dislocation structure, are observed. These defects disappear during post-irradiation tempering. As in western steels, the embrittlement is due to both hardening and reduction of interphase cohesion. The radiation damage specificity of VVER steels arises from their high amount of phosphorus and from their significant density of fine vanadium carbides. (author)

  20. Primary water chemistry of VVERs-operating experience

    International Nuclear Information System (INIS)

    Kysela, Jan; Zmitko, Milan; Petrecky, Igor

    1998-01-01

    VVER units are operated in mixed boron-potassium-ammonia water chemistry. Several modifications of the water chemistry, differing in boron-potassium co-ordination and in the way how hydrogen concentration is produced and maintain in the coolant, is used. From the operational experience point of view VVER units do not show any significant problems connected with the primary coolant chemistry. The latest results indicate that dose rate levels are slowly returning to the former ones. An improvement of the radiation situation observed last two years is supported by the surface activity measurements. However, the final conclusion on the radiation situation can be made only after evaluation of the several following cycles. Further investigation is also needed to clarify a possible effect of modified water chemistry and shut-down chemistry on radioactivity build-up and dose rate level at Dukovany units. Structure materials composition has a significant effect on radiation situation in the units. It concerns mainly of cobalt content in SG material. There is no clear evidence of possible effect of the SG shut-down regimes on the radiation situation in the units even if the dose rate and surface activity data show wide spread for the individual reactor loops. (S.Y.)

  1. VVER Reactor Safety in Eastern Europe and Former Soviet Union

    Science.gov (United States)

    Papadopoulou, Demetra

    2012-02-01

    VVER Soviet-designed reactors that operate in Eastern Europe and former Soviet republics have heightened international concern for years due to major safety deficiencies. The governments of countries with VVER reactors have invested millions of dollars toward improving the safety of their nuclear power plants. Most of these reactors will continue to operate for the foreseeable future since they provide urgently-needed electrical power. Given this situation, this paper assesses the radiological consequences of a major nuclear accident in Eastern Europe. The paper also chronicles the efforts launched by the international nuclear community to improve the safety of the reactors and notes the progress made so far through extensive collaborative efforts in Armenia, Bulgaria, the Czech Republic, Hungary, Kazakhstan, Lithuania, Russia, Slovakia, and Ukraine to reduce the risks of nuclear accidents. Western scientific and technical staff collaborated with these countries to improve the safety of their reactor operations by strengthening the ability of the regulator to perform its oversight function, installing safety equipment and technologies, investing time in safety training, and working diligently to establish an enduring safety culture. Still, continued safety improvement efforts are necessary to ensure safe operating practices and achieve timely phase-out of older plants.

  2. Tightening unit EZ 250 for VVER 1000 type reactor pressure vessel head flange joints

    International Nuclear Information System (INIS)

    Ruchar, Miloslav; Nadenik, Tomas; Kroj, Ludek

    2010-01-01

    The programme of flange joints tightening by seals made of expanded graphite for VVER 440 and VVER 1000 reactor head flange joints is highlighted, and tightening units of row EZ 650 and EZ 650 TK and KNI for VVER 440 reactor head flange joints and EZ 250 tightening unit for VVER 1000 reactor head flange joints are described in detail. The main advantages of electronically controlled tightening units include: Precise and uniform compression of the gasket during the tightening procedure; Automated solution to the graphite relaxing problem during tightening; Possibility of diagnosis of the thread status of the joints being tightened; Alleviation of operator's tough work; Shorter time of tensioning associated with a lower collective doses; Quick preparation of tightening procedure report from the data measured; Calibration renewal is possible in advance at time of unit storage without the need to place it on a real joint. (P.A.)

  3. ANDREA 2.2 and 2.3. Advances in modelling of VVER cores

    Energy Technology Data Exchange (ETDEWEB)

    Havluj, Frantisek; Hejzlar, Jonatan; Vocka, Radim; Vysoudil, Jiri [UJV Rez, Husinec-Rez (Czech Republic)

    2017-09-15

    In 2016 a new version of code ANDREA for core design and reload safety analysis of VVER reactors has been released. The new code version includes several major improvements. The first of them is a seamless incorporation of short time kinetics calculations (without temperature feedback) into the code. This new feature accompanied by the possibility of excore detector signal predictions enables precise interpretation of dynamic measurements of control assembly weight during the reactor startup. Second important enhancement resides in new flexible format of cross section libraries and in new fuel temperature model based on results of TRANSURANUS fuel performance code. The new code version has been thoroughly tested and validated for both VVER440 and VVER-1000 reactors. Furthermore for the new version 2.3 which is to be released shortly we have implemented the possibility of fluent control assemblies' motion and of non-equidistant axial nodalization schemes in VVER-440 calculations.

  4. Natural circulation in a VVER reactor geometry: Experiments with the PACTEL facility and Cathare simulations

    Energy Technology Data Exchange (ETDEWEB)

    Raussi, P.; Kainulainen, S. [Lappeenranta Univ. of Technology, Lappeenranta (Finland); Kouhia, J. [VTT Energy, Lappeenranta (Finland)

    1995-09-01

    There are some 40 reactors based on the VVER design in use. Database available for computer code assessment for VVER reactors is rather limited. Experiments were conducted to study natural circulation behaviour in the PACTEL facility, a medium-scale integral test loop patterned after VVER pressurized water reactors. Flow behaviour over a range of coolant inventories was studied with a small-break experiment. In the small-break experiments, flow stagnation and system repressurization were observed when the water level in the upper plenum fell below the entrances to the hot legs. The cause was attributed to the hot leg loop seals, which are a unique feature of the VVER geometry. At low primary inventories, core cooling was achieved through the boiler-condenser mode. The experiment was simulated using French thermalhydraulic system code CATHARE.

  5. Assessment of Automated Data Analysis Application on VVER Steam Generator Tubing

    International Nuclear Information System (INIS)

    Picek, E.; Barilar, D.

    2006-01-01

    INETEC - Institute for Nuclear Technology has developed software package named EddyOne having an option of automated analysis of bobbin coil eddy current data. During its development and site use some features were noticed preventing the wide use automatic analysis on VVER SG data. This article discuss these specific problems as well evaluates possible solutions. With regards to current state of automated analysis technology an overview of advantaged and disadvantages of automated analysis on VVER SG is summarized as well.(author)

  6. Sensitivity study applied to the CB4 VVER-440 benchmark on burnup credit

    International Nuclear Information System (INIS)

    Markova, Ludmila

    2003-01-01

    A brief overview of four completed portions (CB1, CB2, CB3, CB3+, CB4) of the international VVER-440 benchmark focused on burnup credit and a sensitivity study as one of the final views of the benchmark results are presented in the paper. Finally, the influence of real and conservative VVER-440 fuel assembly models taken for the isotopics calculation by SCALE sas2 on the system k eff is shown in the paper. (author)

  7. Typical design/qualification acceptance criteria for newly installed pipelines and equipment components of VVER-type NPPs

    International Nuclear Information System (INIS)

    Masopust, R.

    2003-01-01

    This paper describes in general the typical design/qualification acceptance criteria and seismic acceptance criteria in particular that are applicable for important to safety newly installed pipelines and equipment components of VVER-type already existing NPPs, specifically during the design verification phase of this newly installed equipment. These criteria are currently used for VVER 440-213 and VVER 1000 NPPs in Czech Republic and in Slovakia. The similar criteria are also used in Hungary. (author)

  8. Metformina – mechanizmy działania i zastosowanie w terapii cukrzycy typu 2[i][/i

    Directory of Open Access Journals (Sweden)

    Marzena Grzybowska

    2011-01-01

    Full Text Available Metformina jest obecnie najczęściej zalecanym lekiem w terapii cukrzycy typu 2. Mimo iż ta pochodna biguanidu jest stosowana od ponad 50 lat, mechanizm jej działania nie został dokładnie poznany. W pracy przedstawiono najnowsze doniesienia o mechanizmach antyhiperglikemicznego działania metforminy. Obejmują one: zmniejszenie wchłaniania glukozy w jelicie cienkim, zwiększony transport glukozy do komórek, obniżenie osoczowego stężenia wolnych kwasów tłuszczowych oraz hamowanie glukoneogenezy. Szczególną rolę w tych procesach odgrywa aktywacja kinazy białkowej aktywowanej przez AMP. Najnowsze odkrycia umożliwiły poznanie mechanizmów działania przeciwmiażdżycowego, hipotensyjnego i przeciwnowotworowego metforminy oraz jej wpływu na czynność śródbłonka naczyń. Plejotropowe działanie metforminy obejmuje wpływ na profil lipidowy osocza, zmniejszenie stresu oksydacyjnego, a także zwiększenie aktywności fibrynolitycznej osocza. Mimo że metformina nie jest metabolizowana, najnowsze badania wykazały, że jest aktywnie transportowana do hepatocytów, a także do komórek nabłonka kanalików nerkowych, odpowiednio przez OCT1 (organic cation transporter 1, kodowany przez gen SLC22A1 oraz OCT2 (kodowany przez [i]SLC22A2[/i]. Z kolei transporter MATE1 (multidrug and toxin extrusion 1 protein, kodowany przez gen [i]SLC47A1[/i] umożliwia wydzielanie metforminy z tych komórek do żółci lub moczu. Polimorfizm genów transporterów metforminy może się przyczynić do istotnych różnic w reakcji na lek.

  9. Minimum throttling feedwater control in VVER-1000 and PWR NPPs

    International Nuclear Information System (INIS)

    Symkin, B.E.; Thaulez, F.

    2004-01-01

    This paper presents an approach for the design and implementation of advanced digital control systems that use a minimum-throttling algorithm for the feedwater control. The minimum-throttling algorithm for the feedwater control, i.e. for the control of steam generators level and of the feedwater pumps speed, is applicable for NPPs with variable speed feedwater pumps. It operates in such a way that the feedwater control valve in the most loaded loop is wide open, steam generator level in this loop being controlled by the feedwater pumps speed, while the feedwater control valves in the other loops are slightly throttling under the action of their control system, to accommodate the slight loop imbalances. This has the advantage of minimizing the valve pressure losses hence minimizing the feedwater pumps power consumption and increasing the net MWe. The benefit has been evaluated for specific plants as being roughly 0.7 and 2.4 MW. The minimum throttling mode has the further advantages of lowering the actuator efforts with potential positive impact in actuator life and of minimizing the feedwater pipelines vibrations. The minimum throttling mode of operation has been developed by the Ukrainian company LvivORGRES. It has been applied with great deal of success on several VVER-1000 NPPs, six units of Zaporizhzha in Ukraine plus, with participation of Westinghouse, Kozloduy 5 and 6 in Bulgaria and South Ukraine 1 to 3 in Ukraine. The concept operates with both ON-OFF valves and true control valves. A study, jointly conducted by Westinghouse and LvivORGRES, is ongoing to demonstrate the applicability of the concept to PWRs having variable speed feedwater pumps and having, or installing, digital feedwater control, standalone or as part of a global digital control system. The implementation of the algorithm at VVER-1000 plants provided both safety improvement and direct commercial benefits. The minimum-throttling algorithm will similarly increase the performance of PWRs. The

  10. In vessel retention for VVER 1000 - Experimental work

    International Nuclear Information System (INIS)

    Batek, D.

    2015-01-01

    After Fukushima accident, the nuclear community realized that it is necessary to have strategy and solution for severe accident management. In Vessel Retention (IVR) of corium is an important strategy to mitigate the consequences of a severe accident. In this poster the author reviews the present status of experimental works made by UJV (Czech Republic) from 2012 until now, on the IVR strategy specifically applied for the VVER 1000 unit. The BESTH 1 experiment was prepared to test the behavior of the RPV (Reactor Pressure Vessel) surface under 2 configurations: clean and corroded. BESTH 2 experiment is a modification of BESTH 1 experiment in order to get greater thermal fluxes. The BESTH 3 facility is a large scale experiment that is under extensive design (2016-2017) whose main objective will be to investigate the results of vast analytical works made by experts with specialization of severe accident phenomenology

  11. Accident loads for a VVER-440/213 containment

    Energy Technology Data Exchange (ETDEWEB)

    Techy, Z. [Institute for Electric Power Research (VEIKI), Budapest (Hungary); Lajtha, G. [Institute for Electric Power Research (VEIKI), Budapest (Hungary); Taubner, R. [Institute for Electric Power Research (VEIKI), Budapest (Hungary)

    1995-08-01

    Specific features of the VVER-440/213 containment are the subdivided rectangular building and the localization system including the bubbler trays and air traps. Accident loads are calculated for a large break loss of coolant accident (LBLOCA). The maximum pressure and temperature loads are calculated with different codes during the blowdown phase of the LBLOCA. The uncertainty margins of the maximum pressure are given in this case. Sensitivity studies are performed for different leakage rates and hydraulic data of the containment. The effects of the active and passive spray systems on the depressurization are presented in this paper. The maximum pressure loads are also examined in case of degraded conditions of the localization system. (orig.).

  12. DRAGON analysis of MOX fueled VVER cell benchmarks

    International Nuclear Information System (INIS)

    Marleau, G.; Foissac, F.

    2002-01-01

    The computational unit-cell benchmarks problems for LEU and MOX fueled VVER-1000 ('water-water energetic reactor') have been analyzed using the code DRAGON with ENDF/B-V and ENDF/B-VI based WIMS-AECL cross section libraries. The results obtained were compared with those generated using the SAS2H module of the SCALE-4.3 computational code system and with the code HELIOS. Good agreements between DRAGON and HELIOS were obtained when the ENDF/B-VI based library was considered while the ENDF/B-V DRAGON results were generally closer to those obtained using SAS2H. This study was useful for the verification of the DRAGON code and confirms that HELIOS and DRAGON have a similar behavior when compatible cross sections library are used. (author)

  13. Accident management strategies for VVER-1000 reactors. Part 1: text

    International Nuclear Information System (INIS)

    Sdouz, G.; Sonneck, G.; Pachole, M.

    1994-10-01

    This report describes the effect of different accident management strategies on the onset, development and end of the core-concrete-interaction as well as on the containment integrity for a TMLB'-type sequence in a Pressurized Water Reactor of the type VVER- 1000. Using the computer code MARCH3 the following strategies were investigated: (1) One or more Spray and LP ECC Systems available with and without coolers after 10 hours (2) Inclusion of the reactor pressure vessel testing facility room to the cavity (3) Containment venting (4) External water supply and (5) Different electric power restoration times. The results show that some of these accident management measures will maintain the containment integrity and reduce the source term drastically, others will reduce the source term rate. For some measures final conclusions can only be given after complete source term calculations have been performed. (authors)

  14. Method of the characteristics for calculation of VVER without homogenization

    Energy Technology Data Exchange (ETDEWEB)

    Suslov, I.R.; Komlev, O.G.; Novikova, N.N.; Zemskov, E.A.; Tormyshev, I.V.; Melnikov, K.G.; Sidorov, E.B. [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

    2005-07-01

    The first stage of the development of characteristics code MCCG3D for calculation of the VVER-type reactor without homogenization is presented. The parallel version of the code for MPI was developed and tested on cluster PC with LINUX-OS. Further development of the MCCG3D code for design-level calculations with full-scale space-distributed feedbacks is discussed. For validation of the MCCG3D code we use the critical assembly VENUS-2. The geometrical models with and without homogenization have been used. With both models the MCCG3D results agree well with the experimental power distribution and with results generated by the other codes, but model without homogenization provides better results. The perturbation theory for MCCG3D code is developed and implemented in the module KEFSFGG. The calculations with KEFSFGG are in good agreement with direct calculations. (authors)

  15. Fuel Cycle of VVER-1000: technical and economic aspects

    International Nuclear Information System (INIS)

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

    2009-01-01

    The paper contains estimations of dependences of technical and economic characteristics of VVER-1000 fuel cycle on number of charged FAs and their enrichment. In the study following restrictions were used: minimum quantity of loaded fresh FAs is equal 36 FAs, a maximum one - 78 (79) FAs and fuel enrichment is limited by value 4,95 %. The following technical and economic characteristics are discussed: cycle length, average burnup of spent fuel, specific consumption of natural uranium, specific quantity of separative work, annual production of thermal energy, fuel component of electrical energy cost, electricity generation cost. Results of estimations are presented as dependences of researched characteristics on cycle length, quantity of loaded FAs and their enrichments. The presented information allows to show tendencies and ranges of technical and economic characteristics at change of fuel cycle parameters. This information can be useful for definition of the fuel cycle parameters which satisfy the requirements of power system and exploiting organizations. (authors)

  16. Profilaktyka cukrzycy typu 2 w grupie osób starszych = Type 2 diabetes prevention among elderly people

    Directory of Open Access Journals (Sweden)

    Anna Mirczak

    2016-12-01

    Cukrzyca typu 2 jest poważną chorobą przewlekłą występującą powszechnie w grupie osób starszych. Późna diagnoza oraz niewłaściwe leczenie prowadzi do poważnych  powikłań takich jak choroby układu krążenia, neuropatia, zespół stopy cukrzycowej i amputacji, niewydolność nerek oraz ślepota, których leczenie jest bardzo kosztowne. Cukrzyca typu 2 jest uwarunkowana czynnikami genetycznymi oraz środowiskowymi (niewłaściwy styl życia. Coraz powszechniejsze wstępowanie cukrzycy w społeczeństwie wynika jednak głównie z nadwagi, otyłości brzusznej oraz niskiej aktywności fizycznej. Badania pokazują, że poprzez modyfikacje stylu życia (dieta, aktywność fizyczna można opóźniać i zapobiegać występowaniu cukrzycy w grypach ryzyka. W tym procesie niezwykle ważne są trzypoziomowe działania profilaktyczne. Profilaktyka pierwszej fazy oznacza zapobieganie cukrzycy poprzez zmianę stylu życia u osób zdrowych, profilaktyka drugiej fazy to zapobieganie występowaniu powikłaniom cukrzycy u osób z grupy ryzyka, natomiast profilaktyka trzeciorzędowa skierowana jest do osób chorych i polega na wczesnym wykrywaniu i odpowiednim leczeniu powikłań cukrzycy. Jednakże, pomimo iż cukrzyca jest chorobą często występującą wśród seniorów, wciąż brakuje badań klinicznych dotyczących specyfiki leczenia w tej grupie wiekowe, jak również doniesień na temat działań prewencyjnych. Celem pracy było przedstawienie zasad profilaktyki cukrzycy w grupie osób starszych na podstawie analizy dostępnej literatury. Dotychczasowe badania wskazują, że skuteczną metodą profilaktyki cukrzycy wśród seniorów jest optymalizacja poziomu glukozy we krwi poprzez zdrową dietę, aktywność fizyczną, kontrolę wartości trójglicerydów i ciśnienia tętniczego krwi, zmniejszenie masy ciała i unikanie używek.   Abstract Type 2 diabetes is one of the most common chronic diseases among the elderly and is also a very serious health

  17. Radioactive release from VVER-1000 reactors after a terror attack

    International Nuclear Information System (INIS)

    Sdouz, G.

    2005-01-01

    Full text: One of the terror scenarios for nuclear power plants is a severe damage of the reactor containment caused by a plane crash or a missile. Due to the loss of electric power the cooling of the core is not maintained leading to a core melt accident. Normally in the course of severe accidents an intact containment has the ability to retain a large part of the radioactive inventory. The goal of this work is the investigation of the behavior of the radioactive release from a VVER-1000-type reactor during a severe accident with a large containment leak from the beginning of the accident. The results are compared with the release in a severe accident via a very small leakage due to the untightness of the containment. This work supplements a series of studies investigating the behavior of a VVER-1000-type reactor during severe accidents under different accident management strategies. The focus in this study is on the 'station blackout'-sequence (or TMLB' in the WASH-1400 nomenclature). The calculations were performed using the Source Term Code Package (STCP), hydrogen explosions are not considered. Up to the melt-through of the cavity bottom the thermal-hydraulics phenomena are almost identical to the TMLB'-case with an intact containment from the beginning. The phenomena occur slightly delayed due to the large containment leak. When the core-concrete-interaction begins the resulting gases leave the containment through the large leak and do not cause a pressure increase. The containment pressure remains at ambient pressure. Due to the different behavior and to the different release times of the nuclides the deviations to the scenario with an intact containment show a great variety. From this comparison it can be shown that the intact containment retains the nuclides up to a factor of 6000. (author)

  18. NNSA / IAEA VVER reactor safety workshops. May 2002 - April 2003. Executive summary

    International Nuclear Information System (INIS)

    Evans, M.; Petri, M. C.

    2003-01-01

    Over the past year, the U.S. National Nuclear Security Administration (NNSA) has sponsored four workshops to compare the probabilistic risk assessments (PRAs) of Soviet-designed VVER power plants. The ''International Workshop on Safety of First-Generation VVER-440 Nuclear Power Plants'' was held on May 20-25, 2002, in Piestany, Slovakia. A short follow-on workshop was held in Bratislava, Slovakia, on November 5-6, 2002, to complete the work begun in May. Piestany was the location also for the ''International Workshop on Safety of Second-Generation VVER-440 Nuclear Power Plants'' (September 9-14, 2002) and the ''International Workshop on Safety of VVER-1000 Nuclear Power Plants'' (April 7-12, 2003). The four workshops were held in cooperation with the International Atomic Energy Agency (IAEA), the Nuclear Regulatory Authority of Slovakia (UJD), the Center for Nuclear Safety in Central and Eastern Europe (CENS), and Argonne National Laboratory (ANL). The objectives of the workshops were to identify the impact of the improvements on the core damage frequency; the contribution to the PRA results of different assumptions about events that can occur at the plants; and to understand, identify, and prioritize potential improvements in hardware and plant operation of VVER nuclear power plants. These objectives were achieved based on insights gained from recent PRAs completed by the plants and their technical support organizations. Nine first-generation VVER-440 plants (nominally of the VVER-440/230 design) are currently operating in Armenia, Bulgaria, Russia, and Slovakia. Sixteen VVER-440/213 plants are currently operating in the Czech Republic, Hungary, Russia, Slovakia, and Ukraine. Twenty-three VVER-1000 plants are currently operating in Bulgaria, the Czech Republic, Russia, and Ukraine. Eleven addition plants are in the advanced stages of construction in various parts of the world. The workshops reviewed the current configuration and safety status of each plant

  19. PSB-VVER counterpart experiment simulating a small cold leg break LOCA

    International Nuclear Information System (INIS)

    Blinkov, V.N.; Melikhov, O.I.; Kapustin, A.V.; Lipatov, I.A.; Dremin, G.I.; Nikonov, S.M.; Rovnov, A.A.; Elkin, I.V.; Pylev, S.S.

    2005-01-01

    Full text of publication follows: An experiment simulating a small break LOCA has been performed in PSB-VVER facility, under PSB-VVER OECD Project. The test is intended to be a counterpart one to an experiment performed in the LOBI integral test facility. The objectives of the PSB-VVER small cold leg break test are: to study VVER-1000 thermal hydraulic response following a small break in the cold leg, to provide data for code assessment regarding phenomena indicated in the VVER-1000 code validation matrix and to study the scaling effect. The scenario for the PSB-VVER experiment has been designed taking the LOBI BL-34 test as reference. The ratio of primary system volumes (without volume of the pressurizer and the surge line) has been chosen to scale the reference experiment conditions and to generate the conditions of PSB-VVER cold leg break experiment. The resulting conditions are compared with the LOBI cold leg break test conditions by means of different counterpart test criteria. Comparing the two experiments in terms of the criteria shows that basic requirements to the counterpart test are fulfilled. A pretest analysis with RELAP5/MOD3.2 code has shown that the PSBVVER small break experiment is expected to show the same relevant phenomena and main events as the LOBI BL-34 test. The predicted PSB-VVER primary pressure is very close to that measured in the LOBI facility. The measured pressure in the PSB-VVER primary system has turned out to be very close to that registered in LOBI BL-34 test. This verifies the approach used for developing the conditions of the PSB-VVER counterpart test. The experiment results and the RELAP5/MOD3.2 pretest calculation are in good agreement. A posttest calculation of the experiment with RELAP5/MOD3.2 code has been performed in order to assess the codes capability to simulate the phenomena relevant to the test. The code has shown a reasonable prediction of the phenomena measured in the experiment. (authors)

  20. PSB-VVER counterpart experiment simulating a small cold leg break LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Blinkov, V.N.; Melikhov, O.I.; Kapustin, A.V.; Lipatov, I.A.; Dremin, G.I.; Nikonov, S.M.; Rovnov, A.A. [Elektrogorsk Research and Engineering Center, EREC, Bezymiannaja Street, 6, Elektrogorsk, Moscow Region, 142530 (Russian Federation); Elkin, I.V.; Pylev, S.S. [NSI RRC ' Kurchatov Institute' , Kurchatov Sq., 1, Moscow, 123182 (Russian Federation)

    2005-07-01

    Full text of publication follows: An experiment simulating a small break LOCA has been performed in PSB-VVER facility, under PSB-VVER OECD Project. The test is intended to be a counterpart one to an experiment performed in the LOBI integral test facility. The objectives of the PSB-VVER small cold leg break test are: to study VVER-1000 thermal hydraulic response following a small break in the cold leg, to provide data for code assessment regarding phenomena indicated in the VVER-1000 code validation matrix and to study the scaling effect. The scenario for the PSB-VVER experiment has been designed taking the LOBI BL-34 test as reference. The ratio of primary system volumes (without volume of the pressurizer and the surge line) has been chosen to scale the reference experiment conditions and to generate the conditions of PSB-VVER cold leg break experiment. The resulting conditions are compared with the LOBI cold leg break test conditions by means of different counterpart test criteria. Comparing the two experiments in terms of the criteria shows that basic requirements to the counterpart test are fulfilled. A pretest analysis with RELAP5/MOD3.2 code has shown that the PSBVVER small break experiment is expected to show the same relevant phenomena and main events as the LOBI BL-34 test. The predicted PSB-VVER primary pressure is very close to that measured in the LOBI facility. The measured pressure in the PSB-VVER primary system has turned out to be very close to that registered in LOBI BL-34 test. This verifies the approach used for developing the conditions of the PSB-VVER counterpart test. The experiment results and the RELAP5/MOD3.2 pretest calculation are in good agreement. A posttest calculation of the experiment with RELAP5/MOD3.2 code has been performed in order to assess the codes capability to simulate the phenomena relevant to the test. The code has shown a reasonable prediction of the phenomena measured in the experiment. (authors)

  1. Aktywność fizyczna osób z przewlekłym wirusowym zapaleniem wątroby typu C

    Directory of Open Access Journals (Sweden)

    Marta Estera Kowalska

    2016-02-01

    Independent Epidemiology Unit, Medical University of Lublin   Słowa kluczowe: wirusowe zapalenie wątroby typu C, aktywność fizyczna, edukacja zdrowotna. Key words: hepatitis C, physical activity, health education.     Streszczenie Wstęp Wirusowe zapalenie wątroby typu C stanowi jeden z głównych problemów zdrowia publicznego ze względu na długi okres bezobjawowy zakażenia wirusem HCV (hepatitis C virus oraz brak skutecznej szczepionki. Cel pracy Celem pracy była ocena poziomu aktywności fizycznej u osób z przewlekłym wirusowym zapaleniem wątroby typu C. Materiał i metody Materiałem wykorzystanym do badań były odpowiedzi na pytania zamknięte zawarte w autorskim kwestionariuszu ankiety. W badaniu wzięło udział 200 osób z potwierdzonym zakażeniem HCV. Wyniki Większość badanych podejmuje aktywność ruchową kilka razy w roku lub też sezonowo. Stwierdzono również, iż w badanej grupie im starszy wiek tym ankietowani rzadziej deklarują podejmowanie form aktywności fizycznej. Ponadto wykazano, iż aktywność fizyczna zmienia się wraz z poziomem wykształcenia. Im wyższy poziom wykształcenia, tym badani częściej podejmują aktywność fizyczną. Wnioski W grupie badanych pacjentów z wirusowym zapaleniem wątroby typu C występuje niska częstotliwość podejmowania aktywności fizycznej. Tylko młodzi ludzie oraz posiadający wyższe wykształcenie deklarują aktywność fizyczną prowadzoną kilka razy w roku.     Abstract Introduction and aim: Hepatitis C is one of the major public health problems due to the long period of asymptomatic infection with HCV and the lack of an effective vaccine. The aim of this study was to evaluate the level of physical activity in patients with chronic hepatitis C. Materials and methods: Material used for the study were responses to closed questions contained in the author's survey interview. The study involved 200 people with confirmed HCV infection. Results: Most of the respondents take

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

    International Nuclear Information System (INIS)

    Mikolas, P.

    2009-01-01

    Joint AER Working Group A on 'Improvement, extension and validation of parameterized few-group libraries for VVER-440 and VVER-1000' and AER Working group B on 'Core design' eighteenth meeting was hosted by Skoda JS a.s. in Plzen (Czech Republic) during the period of 4 to 6 May 2009. There were present altogether 16 participants from 6 member organizations and 13 presentations were read. Objectives of the meeting of WG A are: Issues connected with spectral calculations and few-groups libraries preparation, their accuracy and validation. Presentations were devoted to some aspects of few group libraries preparations and to the benchmark dealing with VVER-440 follower modeling in calculations. Gy. Hegyi gave some new information about NURESIM-NURISP EU project (ZR-6), R. Zajac spoke about the development of data libraries for codes BIPR-7 and PERMAK, P. Darilek compared FA's with Gd during burning process and Yu. Bilodid described further development of plutonium-based burnup history modeling in DYN3D burnup calculations. G. Hordosy presented results of control rod follower induced local power peaking computational benchmark and J. Svarny described Monte Carlo VVER-440 control rod follower benchmark computations. Future activities are also shortly described in the end of the paper. (author)

  3. Qualification of UT methods and systems used for in-service inspections of VVER 440 vessels

    International Nuclear Information System (INIS)

    Skala, Z.; Vit, J.

    2003-01-01

    SKODA JS has been performing automated in-service inspections VVER reactor pressure vessels for more than twenty years. All of these inspections were performed by ultrasonic pulse echo method, combined from 1996 with eddy current testing. The Time of Flight Diffraction Method (TOFD) is one of modern methods of ultrasonic testing. The accuracy of sizing the through wall extent of a flaw by TOFD is much better than the accuracy achievable by the pulse echo method. A series of laboratory tests were performed by SKODA JS and confirmed the suitability of TOFD method for VVER reactor parts testing. The Czech Atomic law demands the qualification of systems and methods used for the in-service inspections of nuclear reactors. The qualification is done in accordance with ENIQ methodology and consists of preparation of the Technical Justification and practical tests made under the surveillance of Qualification Body. SKODA JS intends to qualify systems and methods used for the automated ultrasonic testing of VVER 440 and VVER 1000 reactor components from the inner as well as from the outer surface. The accuracy of the flaw through wall extent sizing by TOFD was confirmed by the qualification of methods and systems used for the testing of VVER 440 vessel circumferential weld and so the TOFD method shall be used routinely by SKODA JS for the inspection of vessel circumferential welds root area and for sizing of flaws exceeding the acceptance level. (author)

  4. Control of selected VVER components life time SKODA JS a.s. experience

    International Nuclear Information System (INIS)

    Zdebor, J.; Pribulla, E.

    2005-01-01

    Experience from the operation of nuclear power plants with type VVER reactors has shown that the life time management of a number of nuclear reactor components is technically as well as with respect to safety substantiated even at the time which exceeds their originally designed life time. To accept such solution it was necessary to develop and implement a number of programs based on which it was possible to evaluate the actual condition of monitored equipment. It is a condition allowing to adopt solution concerning the possibility of their further operation. Experience from the manufacture of 21 sets of reactor equipment for VVER 440 and 3 sets VVER 1000 has been gathered in SKODA JS a.s. which have been completed by experience from service activities performed at operated nuclear power plants. Analyses of manufacturing data and in-service inspection results completed by a lot of laboratory tests have become the basis for the development of selected VVER components life time management programs. The paper focuses on the life time management of those VVER components which SKODA JS a.s. has most experience with. (authors)

  5. SCORPIO-VVER core monitoring and surveillance system with advanced capabilities

    International Nuclear Information System (INIS)

    Molnar, J.; Vocka, R.

    2009-01-01

    The SCORPIO-VVER system includes following features: 1) Validation of plant measurements and identification of sensor failures. 2) Optimum combination of measurements and calculations to obtain precise values of important parameters. 3) On-line 3D power distribution calculation with pin power reconstruction. 4) Limit checking and thermal margin calculation allowing for surveillance of VVER core limits such as DNBR, Sub-cooling margin, FdH and FQ peeking factors. 5) Integrated modules for monitoring fuel performance and coolant activity for identification of fuel failures. 6) Predictive capabilities and strategy planning, offering the possibility to check the consequences of operational manoeuvres in advance, prediction of critical parameters, etc. 7) Convenient monitoring of approach to criticality during reactor start-up. 8) Automated transition between cycles (fuel reload). The SCORPIO-VVER core monitoring system with its flexible and modular framework successfully responses to the plant operating needs and advances in nuclear fuel cycle strategies and fuel design. Modular framework allows for easy modifications of the system and implementation of new methods in physical modules. Even if the system is installed only on VVER-440 reactors, it could be adapted for VVER-1000 needs

  6. Verification of the Korsar code on results of experiments executed on the PSB-VVER facility

    International Nuclear Information System (INIS)

    Roginskaya, V.L.; Pylev, S.S.; Elkin, I.V.

    2005-01-01

    Full text of publication follows: Paper represents some results of computational research executed within the framework of verification of the KORSAR thermal hydraulic code. This code was designed in the NITI by A.P. Aleksandrov (Russia). The general purpose of the work was development of a nodding scheme of the PSB-VVER integral facility, scheme testing and computational modelling of the experiment 'The PSB-VVER Natural Circulation Test With Stepwise Reduction of the Primary Inventory'. The NC test has been performed within the framework of the OECD PSB-VVER Project (task no. 3). This Project is focused upon the provision of experimental data for codes assessment with regard to VVER analysis. Paper presents a nodding scheme of the PSB-VVER facility and results of pre- and post-test calculations of the specified experiment, obtained with the KORSAR code. The experiment data and the KORSAR pre-test calculation results are in good agreement. A post-test calculation of the experiment with KORSAR code has been performed in order to assess the code capability to simulate the phenomena relevant to the test. The code showed a reasonable prediction of the phenomena measured in the experiment. (authors)

  7. A 25% double-ended LOCA in the PSB-VVER facility

    International Nuclear Information System (INIS)

    Blinkov, V.N.; Lipatov, I.A.; Dremin, G.I.

    2003-01-01

    The paper presents the results of the experimental investigation in the PSB-VVER facility and post-test analysis with RELAP5/MOD3.3 of the thermal-hydraulic response of the PSB-VVER to a 25% double-ended Hot Leg Break (HLB). The test scenario included loss of off-site power concurrently with the scram-signal and the safety system operation as described in the reference VVER-1000 operational manual in the case of this type of accident assuming one diesel-generator failure. The key transient parameter trends as well as sequence of events and phenomena are given in the paper. RELAP5/MOD3.3 a post-test analysis has been performed using the experimental data gained as a base. The reasonable qualitative agreement between the key calculated and measured variables has been shown. The quantitative code accuracy evaluation has shown that the total average amplitude of the main parameters' deviations AA tot tot < 0.28 that corresponds to satisfactory quality of the VVER-1000 hot leg guillotine break LOCA modeling in the PSB-VVER. (author)

  8. Verification of the Korsar code on results of experiments executed on the PSB-VVER facility

    Energy Technology Data Exchange (ETDEWEB)

    Roginskaya, V.L.; Pylev, S.S.; Elkin, I.V. [NSI RRC ' Kurchatov Institute' , Kurchatov Sq., 1, Moscow, 123182 (Russian Federation)

    2005-07-01

    Full text of publication follows: Paper represents some results of computational research executed within the framework of verification of the KORSAR thermal hydraulic code. This code was designed in the NITI by A.P. Aleksandrov (Russia). The general purpose of the work was development of a nodding scheme of the PSB-VVER integral facility, scheme testing and computational modelling of the experiment 'The PSB-VVER Natural Circulation Test With Stepwise Reduction of the Primary Inventory'. The NC test has been performed within the framework of the OECD PSB-VVER Project (task no. 3). This Project is focused upon the provision of experimental data for codes assessment with regard to VVER analysis. Paper presents a nodding scheme of the PSB-VVER facility and results of pre- and post-test calculations of the specified experiment, obtained with the KORSAR code. The experiment data and the KORSAR pre-test calculation results are in good agreement. A post-test calculation of the experiment with KORSAR code has been performed in order to assess the code capability to simulate the phenomena relevant to the test. The code showed a reasonable prediction of the phenomena measured in the experiment. (authors)

  9. Investigation of station blackout scenario in VVER440/v230 with RELAP5 computer code

    Energy Technology Data Exchange (ETDEWEB)

    Gencheva, Rositsa Veselinova, E-mail: roseh@mail.bg; Stefanova, Antoaneta Emilova, E-mail: antoanet@inrne.bas.bg; Groudev, Pavlin Petkov, E-mail: pavlinpg@inrne.bas.bg

    2015-12-15

    Highlights: • We have modeled SBO in VVER440. • RELAP5/MOD3 computer code has been used. • Base case calculation has been done. • Fail case calculation has been done. • Operator and alternative operator actions have been investigated. - Abstract: During the development of symptom-based emergency operating procedures (SB-EOPs) for VVER440/v230 units at Kozloduy Nuclear Power Plant (NPP) a number of analyses have been performed using the RELAP5/MOD3 (Carlson et al., 1990). Some of them investigate the response of VVER440/v230 during the station blackout (SBO). The main purpose of the analyses presented in this paper is to identify the behavior of important VVER440 parameters in case of total station blackout. The RELAP5/MOD3 has been used to simulate the SBO in VVER440 NPP model (Fletcher and Schultz, 1995). This model was developed at the Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences (INRNE-BAS), Sofia, for analyses of operational occurrences, abnormal events and design based scenarios. The model provides a significant analytical capability for specialists working in the field of NPP safety.

  10. Issues and remedies for secondary system of PWR/VVER

    International Nuclear Information System (INIS)

    Nordmann, Francis; Odar, Suat; Rochester, Dewey

    2012-09-01

    Secondary side degradation of steam generators (SG) and Flow Accelerated Corrosion (FAC) in the secondary system have been for a long time important issues in PWR and VVER types of Nuclear Power Plants. With the evolution of the design, the most important issues are progressively moving from secondary side corrosion of Alloy 600 SG tubing, which is being replaced, to a larger variety of risks associated with potential inadequate chemistries. As far as FAC of carbon steel is concerned, the evolution of treatment selection for minimizing corrosion products transport toward the SG, as well as progressive replacement of components in the feedwater train, decreases the risk of dramatic failures which have occurred in the past. After having briefly explained the reason for the past problems encountered in the secondary system of PWR and VVER, this paper evaluates the risk associated with various impurities or contaminants that may be present in the secondary system and how to mitigate them in the most appropriate, efficient, economical and environmental friendly way. The covered species are sodium, calcium, magnesium, chloride, sulfate and sulfur compounds, fluorides, organic compounds, silica, oxygen, lead, ion exchange resins. This paper also proposes the best remedies for mitigating the new issues that may be encountered in operating plants or units under construction. These are mainly: - Selecting a steam water treatment able to minimize the quantity of corrosion products transported toward the SG; - Mitigating the risk of Flow Induced Vibration by a proper control of deposits in sensitive areas; - Minimizing the risk of concentration of impurities in local areas where they may induce corrosion; - Avoiding the presence of abnormal quantities of some species in SG, such as the detrimental presence of lead and ion exchange resin debris or the controversial presence of organic compounds; - Optimizing costs of maintenance activities (SG mechanical, chemical cleaning

  11. Three-dimensional reactor dynamics code for VVER type nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kyrki-Rajamaeki, R. [VTT Energy, Espoo (Finland)

    1995-10-01

    A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.).

  12. Three-dimensional reactor dynamics code for VVER type nuclear reactors

    International Nuclear Information System (INIS)

    Kyrki-Rajamaeki, R.

    1995-10-01

    A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.)

  13. Numerical investigation of the coolant mixing during fast deboration transients for VVER-440 type reactors

    International Nuclear Information System (INIS)

    Hoehne, T.; Rhode, U.

    2000-01-01

    The VVER-440 (440 MW) V-230 was considered for analyzing the flow field and mixing processes. The V-230 has no elliptical sieve plate in the lower plenum. Previously, the 3D flow distribution in the downcomer and the lower plenum of the VVER-440 reactor have been calculated by means of the CFD code CFX-4 for operational conditions. The CFX-calculations were compared with the experimental data and the analytical mixing model. In this paper, CFD calculations for the start-up of the first main coolant pump in a VVER-440 type reactor are reported about. This scenario is important in case that there is a plug of lower borated water in one of the primary coolant loops. (orig.)

  14. Test facility of the VVER-440 condensation-type pressure suppression system

    International Nuclear Information System (INIS)

    Wolff, H.; Arndt, S.

    2004-01-01

    Since the early nineties, GRS has supported regulatory authorities in Central and Eastern Europe in performing safety assessments of nuclear power plants. Especially studies of the condensation-type pressure suppression system of VVER-440/V-213-type plants have been important in this respect. Major steps in demonstrating complete functioning of the condensation-type pressure suppression system under accident conditions by experiments run in the Russian large scale test facility, BC V-213, have been completed in the past two years within the framework of various international experimental programs. The test results were used to validate specifically for power plants with VVER-400/V-213 reactors the COCOSYS GRS computer code, which is used in the safety assessments. The results of recalculations of the C02 EREC test, which simulates a break of a main steam pipe, demonstrate the present state of validation of COCOSYS for VVER condensation-type pressure suppression systems. (orig.) [de

  15. Component vibration of VVER-reactors - diagnostics and modelling

    International Nuclear Information System (INIS)

    Altstadt, E.; Scheffler, M.; Weiss, F.-P.

    1995-01-01

    Flow induced vibrations of reactor pressure vessel (RPV) internals (control element and core barrel motions) at VVER-440 reactors have led to the development of dedicated methods for on-line monitoring. These methods need a certain developed stage of the faults to be detected. To achieve a real sensitive early detection of mechanical faults of RPV internals, a theoretical vibration model was developed based on finite elements. The model comprises the whole primary circuit including the steam generators (SG). By means of that model all eigenfrequencies up to 30 Hz and the corresponding mode shapes were calculated for the normal vibration behaviour. Moreover the shift of eigenfrequencies and of amplitudes due to the degradation or to the failure of internal clamping and spring elements could be investigated, showing that a recognition of such degradations even inside the RPV is possible by pure excore vibration measurements. A true diagnostic, that is the identification of the failed component, might become possible because different faults influence different and well separated eigenfrequencies. (author)

  16. Investigation on accident management measures for VVER-1000 reactors

    International Nuclear Information System (INIS)

    Tusheva, P.; Schaefer, F.; Rohde, U.; Reinke, N.

    2009-01-01

    A consequence of a total loss of AC power supply (station blackout) leading to unavailability of major active safety systems which could not perform their safety functions is that the safety criteria ensuring a secure operation of the nuclear power plant would be violated and a consequent core heat-up with possible core degradation would occur. Currently, a study which examines the thermal-hydraulic behaviour of the plant during the early phase of the scenario is being performed. This paper focuses on the possibilities for delay or mitigation of the accident sequence to progress into a severe one by applying Accident Management Measures (AMM). The strategy 'Primary circuit depressurization' as a basic strategy, which is realized in the management of severe accidents is being investigated. By reducing the load over the vessel under severe accident conditions, prerequisites for maintaining the integrity of the primary circuit are being created. The time-margins for operators' intervention as key issues are being also assessed. The task is accomplished by applying the GRS thermal-hydraulic system code ATHLET. In addition, a comparative analysis of the accident progression for a station blackout event for both a reference German PWR and a reference VVER-1000, taking into account the plant specifics, is being performed. (authors)

  17. Issues and remedies for secondary system of PWR/VVER

    International Nuclear Information System (INIS)

    Nordmann, F.; Odar, S.; Rochester, D.

    2012-01-01

    Secondary side degradation of steam generators (SG) tubing with Alloy 600 MA and flow accelerated corrosion (FAC) of carbon steel have been for a long time important issues for the secondary system of PWR and VVER. With the beneficial evolution of the design (for instance the replacement of Alloy 600 SG tubing), the most important issues are progressively moving to a larger variety of risks associated to potential inadequate chemistries. The best remedies for mitigating the new concerns are: -) selecting a steam water treatment able to minimize the quantity of corrosion products transported to the steam generator, -) mitigating the risk of flow induced vibration by a proper control of deposits in sensitive areas, -) minimizing the risk of concentration of impurities in local areas where they may induce corrosion. The paper also explains: -) the benefit of eliminating or by pass of condensate polishers, -) the absence of need for expensive lead investigation, if no specific pollution occurred, -) the absence of need for very low oxygen in the condensate water, and -) the necessary and optimum number of on-line monitors

  18. NRC Support for the Kalinin (VVER) probabilistic risk assessment

    International Nuclear Information System (INIS)

    Bley, D.; Diamond, D.J.; Chu, T.L.; Azarm, A.; Pratt, W.T.; Johnson, D.; Szukiewicz, A.; Drouin, M.; El-Bassioni, A.; Su, T.M.

    1998-01-01

    The US Nuclear Regulatory Commission (NRC) and the Federal Nuclear and Radiation Safety Authority of the Russian Federation have been working together since 1994 to carry out a probabilistic risk assessment (PRA) of a VVER-1000 in the Russian Federation. This was a recognition by both parties that this technology has had a profound effect on the discipline of nuclear reactor safety in the West and that the technology should be transferred to others so that it can be applied to Soviet-designed plants. The NRC provided funds from the Agency for International Development and technical support primarily through Brookhaven National Laboratory and its subcontractors. The latter support was carried out through workshops, by documenting the methodology to be used in a set of guides, and through periodic review of the technical activity. The result of this effort to date includes a set of procedure guides, a draft final report on the Level 1 PRA for internal events (excluding internal fires and floods), and progress reports on the fire, flood, and seismic analysis. It is the authors belief that the type of assistance provided by the NRC has been instrumental in assuring a quality product and transferring important technology for use by regulators and operators of Soviet-designed reactors. After a thorough review, the report will be finalized, lessons learned will be applied in the regulatory and operational regimes in the Russian Federation, and consideration will be given to supporting a containment analysis in order to complete a simplified Level 2 PRA

  19. Determination of mixing factors for VVER-440 fuel assembly head

    Energy Technology Data Exchange (ETDEWEB)

    Tóth, S., E-mail: toth@reak.bme.hu [Institute of Nuclear Techniques, Budapest University of Technology and Economics, Műegyetem rkp. 9, H-1111 Budapest (Hungary); Aszódi, A. [Institute of Nuclear Techniques, Budapest University of Technology and Economics, Műegyetem rkp. 9, H-1111 Budapest (Hungary)

    2013-11-15

    CFD models have been developed for the heads of the old, the present and the new type VVER-440 fuel assemblies using the experience of a former validation process. With these models temperature distributions are investigated in the heads of some typical assemblies and the in-core thermocouple signals are calculated. The analyses show that the coolant mixing is intensive but not-perfect in the assembly heads. The difference between the thermocouple signal and the cross-sectional average temperature at the measurement level depends on the assembly type. Using the results of these CFD calculations the weight factors of the rod bundle regions for the in-core thermocouple have been determined. With these factors the thermocouple signals are estimated and the results are statistically tested using the registered data of the Hungarian nuclear power plant. This test shows that the deviations between the measured and the calculated temperatures can be significantly decreased and consequently monitoring uncertainties can be reduced with using the weight factors.

  20. CASTOR - Advanced System for VVER Steam Generator Inspection

    International Nuclear Information System (INIS)

    Mateljak, Petar

    2014-01-01

    From the safety point of view, steam generator is a very important component of a nuclear power plant. Only a thin tube wall prevents leakage of radioactive material from the primary side into the environment. Therefore, it is very important to perform inspections in order to detect pipe damage and apply appropriate corrective actions during outage. Application of the nondestructive examination (NDE) technique, that can locate degradation and measure its size and orientation, is an integral part of nuclear power plant maintenance. The steam generator inspection system is consisted of remotely controlled manipulator, testing instrument and software for data acquisition and analysis. Recently, the inspection systems have evolved to a much higher level of automation, efficiency and reliability resulting in a lower cost and shorter outage time. Electronic components have become smaller and deal with more complex algorithms. These systems are very fast, precise, reliable and easy to handle. The whole inspection, from the planning, examination, data analysis and final report, is now a highly automated process, which makes inspection much easier and more reliable. This paper presents the new generation of INETEC's VVER steam generator inspection system as ultimate solution for steam generator inspection and repair. (author)

  1. The prediction problems of VVER fuel element cladding failure theory

    International Nuclear Information System (INIS)

    Pelykh, S.N.; Maksimov, M.V.; Ryabchikov, S.D.

    2016-01-01

    Highlights: • Fuel cladding failure forecasting is based on the fuel load history and the damage distribution. • The limit damage parameter is exceeded, though limit stresses are not reached. • The damage parameter plays a significant role in predicting the cladding failure. • The proposed failure probability criterion can be used to control the cladding tightness. - Abstract: A method for forecasting of VVER fuel element (FE) cladding failure due to accumulation of deformation damage parameter, taking into account the fuel assembly (FA) loading history and the damage parameter distribution among FEs included in the FA, has been developed. Using the concept of conservative FE groups, it is shown that the safety limit for damage parameter is exceeded for some FA rearrangement, though the limits for circumferential and equivalent stresses are not reached. This new result contradicts the wide-spread idea that the damage parameter value plays a minor role when estimating the limiting state of cladding. The necessary condition of rearrangement algorithm admissibility and the criterion for minimization of the probability of cladding failure due to damage parameter accumulation have been derived, for using in automated systems controlling the cladding tightness.

  2. Neutronic study of nanofluids application to VVER-1000

    Energy Technology Data Exchange (ETDEWEB)

    Hadad, K., E-mail: hadad@email.arizona.ed [School of Engineering, Shiraz University, Shiraz 7134554115 (Iran, Islamic Republic of); Aerospace and Mechanical Engineering, University of Arizona, Tucson, AZ 85721 (United States); Hajizadeh, A.; Jafarpour, K. [School of Engineering, Shiraz University, Shiraz 7134554115 (Iran, Islamic Republic of); Ganapol, B.D. [Aerospace and Mechanical Engineering, University of Arizona, Tucson, AZ 85721 (United States)

    2010-11-15

    The change in neutronic parameters of the VVER-1000 nuclear reactor core attributable to the use of nanoparticle/water (nanofluid) as coolant is presented in this paper. Optimization of type and volume fraction of nanoparticles in water that affect the safety enhancement of core primary parameters is intended in this study. Reactivity change, radial and axial local peaking factors (LPPF), and the consequence of nanoparticle deposition on fuel clad are investigated. We considered five nanoparticles which have been studied extensively for their heat transfer properties including Alumina, Aluminum, Copper oxide, Copper and Zirconia. The results of our study show that at low concentration (0.001 volume fraction) Alumina is optimum nanoparticle for normal operation. The maximum radial and axial LPPF were found to be invariant to the type of nanofluid at low volume fractions. With an increase in nanoparticle deposition thickness on fuel clad, a flux and K{sub eff} depression occurs and Al{sub 2}O{sub 3} has the lowest rate of drop off.

  3. Structural mechanisms of the flux effect for VVER-1000 reactor pressure vessel materials

    International Nuclear Information System (INIS)

    Gurovich, B.; Kuleshova, E.; Fedotova, S.; Maltsev, D.; Zabusov, O.; Frolov, A.; Erak, D.; Zhurko, D.

    2015-01-01

    To justify the lifetime extension of VVER-1000 reactor pressure vessels (RPV) up to 60 years and more it is necessary to expand the existing surveillance samples database to beyond design fluence by means of accelerated irradiation in a research reactor. Herewith since the changes in mechanical properties of materials under irradiation are due to occurring structural changes, correct analysis of the data obtained at accelerated irradiation of VVER-1000 RPV materials requires a clear understanding of the structural mechanisms that are responsible for the flux effect in VVER-1000 RPV steels. Two mechanisms are responsible for radiation embrittlement of VVER-1000 RPV steels: the hardening one (radiation hardening due to formation of radiation-induced Ni-based precipitates and radiation defects) and non-hardening one (due to formation of impurities segregations at grain boundaries - reversible temper brittleness). In this context for an adequate interpretation of the mechanical tests results when justifying the lifetime extension of existing units a complex of comparative structural studies (TEM, SEM and AES) of VVER-1000 RPV materials irradiated in different conditions (in research reactor IR-8 and within surveillance samples) was performed. It is shown that the flux effect is observed for materials with high nickel content (weld metals with Ni content > 1.35%) and it is mostly due to the contribution of non-hardening mechanism of radiation embrittlement (the difference in the accumulation kinetics of grain boundary phosphorus segregation) and somewhat contribution of the hardening mechanism (the difference in density of radiation-induced precipitates). Therefore when analyzing the results obtained from the accelerated irradiation of VVER-1000 WM the correction for the flux effect should be made. (authors)

  4. Results of operation of VVER-1000 FAs manufactured at PJSC NCCP

    International Nuclear Information System (INIS)

    Davidov, D.; Brovkin, O.; Bezborodov, Y.

    2015-01-01

    Fuel Assemblies manufactured at PJSC NCCP are in operation at 27 VVER-1000 power units at 11 NPPs in Russia, Ukraine, Bulgaria, China, Iran and India. Basic results of operation of PJSC NCCP VVER-1000 FAs during 2007-2014 are presented. The operation results confirm the design characteristics of fuel, i.e.: average fuel burnup up to 55 MW*day/kgU in FAs; safe and reliable FA operation, with low leaking rate (in the order of 10-6). The achieved operation characteristics of TVSA and TVS-2M Fuel Assemblies prove the quality, reliability and competitiveness of FAs manufactured at PJSC NCCP

  5. Simulation of VVER MCCI reactor test case with ASTEC V2/MEDICIS computer code

    International Nuclear Information System (INIS)

    Stefanova, A.; Grudev, P.; Gencheva, R.

    2011-01-01

    This paper presents an application of the ASTEC v2, module MEDICIS for simulation of VVER Molten core concrete interaction test (MCCI) case without water injection. The main purpose of performed calculation is verification and improvement of module MEDICIS/ASTECv2 for better simulation of core concrete interaction processes. The VVER-1000 reference nuclear power plant was chosen as SARNET2 benchmark MCCI test-case. The initial conditions for MCCI test are taken after SBO scenario calculated with ASTEC version 1.3R2 by INRNE. (authors)

  6. Implementation of the SCDAP/RELAP5 Mod. 3.3 and MAAP/VVER codes

    International Nuclear Information System (INIS)

    Duspiva, J.; Vokac, P.; Dienstbier, J.

    2001-05-01

    The SR5 code was installed on a Hewlett/Packard workstation, and test problems, supplied with the software, were solved. Finally, the tool for graphical processing of the calculation results was prepared and tested. The MAAP/VVER code was installed on a HP J210 workstation and, in particular, on PC. The code was tested on two problems, supplied with the software. The transformation of the output from MAAP/VVER to the graphical format was carried out by using the support tools obtained as well as by using tools that have been in use at the Institute for other codes to analyze severe accidents. (P.A.)

  7. Conservative ground of qualification BRU-A VVER-1000 in modes of instability of diphasic environment

    International Nuclear Information System (INIS)

    Skalozubov, V.I.; Klyuchnikov, A.A.; Haj Farajallah Dabbach

    2010-01-01

    The article first presents grounds and conditions of origin of hydraulic shocks in the VVER system of safety relief valves, caused interchannel heat hydrodynamic instability of biphasic medium. It is supposed conservatively that origin of hydraulic shocks caused instability of biphasic stream determines the unavailability to close of safety relief valves. It is established that the modes of hydraulic shocks in safety relief valves of VVER 1000 (B-320) at the fully opened valves are not typical for the conditions of accidents with intercontour leakages.

  8. Modernization of Cross Section Library for VVER-1000 Type Reactors Internals and Pressure Vessel Dosimetry

    Directory of Open Access Journals (Sweden)

    Voloschenko Andrey

    2016-01-01

    Full Text Available The broad-group library BGL1000_B7 for neutron and gamma transport calculations in VVER-1000 internals, RPV and shielding was carried out on a base of fine-group library v7-200n47g from SCALE-6 system. The comparison of the library BGL1000_B7 with the library v7-200n47g and the library BGL1000 (the latter is using for VVER-1000 calculations is demonstrated on several calculation and experimental tests.

  9. Analysis of VVER-1000 large and small break LOCA experiments with RELAP5/MOD3.2

    International Nuclear Information System (INIS)

    Rychkov, M.; Chikkanagoudar, U.; Sehgal, B.R.

    2004-01-01

    A RELAP5 model for the analysis of the PSB-VVER test facility was developed by EREC in Russia. The PSB-VVER is a large-scale integral test facility to model the VVER-1000 type NPP. The volume and power scale in this test facility is 1:300 and the elevation scale is 1:1, which corresponds to the elevation mark of the reactor prototype. At the Division of Nuclear Safety, we have modified the PSB-VVER facility's RELAP5 model in order to analyze two of the transient tests performed on the PSB-VVER facility, which serve as the validation matrix described by NEA/CSNI. The objective of the work conducted was to validate the results obtained from RELAP5's calculation with the supplied experimental data from the PSB-VVER test facility. Two accident scenarios have been calculated and analyzed. After being verified against the '11% UP LOCA' test data, the RELAP5/MOD3.2 model was used for a so-called 'blind' transient calculation of the test '2*25% HL LOCA' and the results obtained were compared with the experimental data provided after the calculation. From the results of the qualitative and quantitative comparison of the 2 test calculations and the experimental data, we can state that the RELAP5/MOD3.2 code gives a satisfactory modeling of the PSB-VVER facility' thermal hydraulic phenomena

  10. Neutron magnetic scattering experiments performed by VIKRA group at the RA reactor in Vinca; Eksperimenti grupe VIKRA na reaktoru RA u Vinci sa rasejanjem neutrona na magnetnim materijalima

    Energy Technology Data Exchange (ETDEWEB)

    Kranicki, S; Rzany, H; Wanic, A [Institute of Nuclear Physics, Krakov (Poland); Szytula, A [Jagiellonian University, Krakov (Poland); Todorovic, J; Dimitrijevic, Z; Radenkovic, S [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1978-05-15

    A short communication about investigations of magnetic materials by means of the neutron magnetic scattering performed by Polish-Yugoslav group VIKRA (Vinca-Krakov) at the RA reactor in Vinca in the period 1962-1977 will be presented. In the field of neutron inelastic scattering information about magnetic interactions (exchange integrals) and their temperature behaviour in such substances as Fe{sub 7}S{sub 8}, {alpha}-Fe{sub 2}O{sub 3}, CR{sub 2}O{sub 3}, {alpha}-Co, Cu{sub 2}MnAl and Mn{sub 1.88}Cr{sub 0.12}Sb was obtained by measuring magnon dispersion relations. In the field of neutron elastic scattering valuable results concerning crystallographic and magnetic structures of oxyhydroxides, several Heusler alloys and intermetallic compounds of 3d metals (for instance of NiMnGe, CoMnSi) were obtained. A short description of the KSN spectrometer used in all experiments will also be given. (author) Dato je kratko saopstenje o istrazivanjima magnetnih materijala magnetnim neutronskim rasejanjem izvrsenim od strane poljsko-jugoslovenske grupe VIKRA (Vinca-Krakov) na reaktoru RA u Vinci u periodu od 1962-1977. U oblasti neelasticnog rasejanja neutrona, obavestenja o magnetnim uzajamnim dejstvima (integralima izmene) i njihovim temperaturskim ponasanjima su dobijeni merenjem disperzione relacije magnona u takvim materijalima kao sto su Fe{sub 7}S{sub 8}, {alpha}-Fe{sub 2}O{sub 3}, CR{sub 2}O{sub 3}, {alpha}-Co, Cu{sub 2}MnAl and Mn{sub 1.88}Cr{sub 0.12}Sb. U oblasti neutronskog elasticnog rasejanja dobijeni su korisni rezultati o kristalografskim i magnetnim strukturama oksihidroksida gvozdja, niza Heuslerovih legura i intermetalnih jedinjenja 3d metala (napr. NiMnGe, Co,MnSi). Dat je kratak opis KSN spektrometra upotrebljenog u svim ovim eksperimentima. (author)

  11. Modelling and analysis of severe accidents for VVER-1000 reactors

    International Nuclear Information System (INIS)

    Tusheva, Polina

    2012-01-01

    Accident conditions involving significant core degradation are termed severe accidents /IAEA: NS-G-2.15/. Despite the low probability of occurrence of such events, the investigation of severe accident scenarios is an important part of the nuclear safety research. Considering a hypothetical core melt down scenario in a VVER-1000 light water reactor, the early in-vessel phase focusing on the thermal-hydraulic phenomena, and the late in-vessel phase focusing on the melt relocation into the reactor pressure vessel (RPV) lower head, are investigated. The objective of this work is the assessment of severe accident management procedures for VVER-1000 reactors, i.e. the estimation of the maximum period of time available for taking appropriate measures and particular decisions by the plant personnel. During high pressure severe accident sequences it is of prime importance to depressurize the primary circuit in order to allow for effective injection from the emergency core cooling systems and to avoid reactor pressure vessel failure at high pressure that could cause direct containment heating and subsequent challenge to the containment structure. Therefore different accident management measures were investigated for the in-vessel phase of a hypothetical station blackout accident using the severe accident code ASTEC, the mechanistic code ATHLET and the multi-purpose code system ANSYS. The analyses performed on the PHEBUS ISP-46 experiment, as well as simulations of small break loss of coolant accident and station blackout scenarios were used to contribute to the validation and improvement of the integral severe accident code ASTEC. Investigations on the applicability and the effectiveness of accident management procedures in the preventive domain, as well as detailed analyses on the thermal-hydraulic phenomena during the early in-vessel phase of a station blackout accident have been performed with the mechanistic code ATHLET. The results of the simulations show, that the

  12. Simulation of mixing effects in a VVER-1000 reactor

    International Nuclear Information System (INIS)

    Ulrich Bieder; Gauthier Fauchet; Sylvie Betin; Nikola Kolev; Dimitar Popov

    2005-01-01

    Full text of publication follows: The work presented has been performed in the framework of the OECD/NEA thermalhydraulic benchmark V1000CT-2. This benchmark is related to fluid mixing in the reactor vessel during a MSLB accident scenario in a VVER-1000 reactor. The purpose of the first exercise is to test the capability of CFD codes to represent the coolant mixing in the reactor vessel, in particular in the downcomer and the lower plenum. Coolant mixing in a VVER-1000 V320 reactor was investigated in plant experiments during the commissioning of Kozloduy Unit 5 and 6. Starting from nearly symmetric states, asymmetric loop operation in different combinations was caused by disturbing the steam flow from one or more steam generators. Non-uniform and asymmetric loop flow mixing in the reactor vessel has been observed in the event of asymmetric loop operation. For certain flow patterns there is a shift (swirl) of the main loop flows with respect to the cold leg axes. This azimuthal shift as well as mixing coefficients from cold legs to the fuel assembly inlets have been measured. The presented reference problem is a pure TH problem with given boundary conditions and power distributions. During a stabilization phase, the thermal power of the reactor was 281 MW i.e. 9.36% of the nominal power according to primary balance. Then, a transient was initiated by closing the steam isolation valve of the steam generator one (SG-1) and isolating SG-1 from feed water. The coolant temperature in the cold and hot legs of Loop no 1 rose by 13-13.5 C. After about 20 minutes a stabilized state was reached which is considered as 'final state'. This final state has been analysed with the Trio-U code. Trio-U is a CFD code developed by the CEA Grenoble, aimed to supply an efficient computational tool to simulate transient thermalhydraulic mono-phase turbulent flows encountered in nuclear systems as well as in industrial processes. For the presented study, a LES approach was used. Therefore

  13. Simulation of a nuclear accident by an academic simulator of a VVER-1000 reactor; Simulacion de un accidente nuclear, mediante un simulador academico de un reactor VVER-1000

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez G, L. [UNAM, Facultad de Ingenieria, Ciudad Universitaria, 04510 Mexico D. F. (Mexico); Salazar S, E., E-mail: laurahg42@gmail.com [UNAM, Facultad de Ingenieria, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, 62250 Jiutepec, Morelos (Mexico)

    2014-10-15

    This work is planned to simulate a scenario in which the same conditions that caused the accident at the Fukushima Daichi nuclear power plant are present, using a simulator of a nuclear power plant with VVER-1000 reactor, a different type of technology to the NPP where the accident occurred, which used BWR reactors. The software where it will take place the simulation was created and distributed by the IAEA for academic purposes, which contains the essential systems that characterize this type of NPP. The simulator has tools for the analysis of the characteristic phenomena of a VVER-1000 reactor in the different systems together and planned training tasks. This makes possible to identify the function of each component and how connects to other systems, thus facilitating the visualization of possible failures and the consequences that they have on the general behavior of the reactor. To program the conditions in the simulator, is necessary to know and synthesize a series of events occurred in Fukushima in 2011 and the realized maneuvers to reduce the effects of the system failures. Being different technologies interpretation of the changes that would suffer the VVER systems in the scenario in question will be developed. The Fukushima accident was characterized by the power loss of regular supply and emergency of the cooling systems which resulted in an increase in reactor temperature and subsequent fusion of their nuclei. Is interesting to reproduce this type of failure in a VVER, and extrapolate the lack of power supply in the systems that comprise, as well as pumping systems for cooling, has a pressure regulating system which involves more variables in the balance of the system. (Author)

  14. Training operators of VVER-1000 units in Eastern Europe

    International Nuclear Information System (INIS)

    Normand, X.; Nabet, E.; Hauesberger, P.

    1996-01-01

    The VVER 1000 is the most recent nuclear reactor designed in the former Soviet Union. Its design and operation principles are close to Western four-loop reactors in the 1000- to 1500-MW class; therefore, the Western simulation technology is usually directly applicable to the simulation of these units. Moreover, the current number of state-of-the-art training simulators in operation is very limited. A total of 19 units are in operation, while only 2 modern simulators are available (full-scope type) in Balakovo and Zaporozhe. Access to these simulators is practically limited to the respective plants' trainees, which means that the other units have to be satisfied with hands-on training. Facing this situation and taking into account the predicted lifetime of these plants (15 to 25 yr to go, maybe more), a lot of effort has been made in recent years to provide the plants with modern simulators. The major hurdles to this action were obviously financial and technical (availability of codes, computers, software tools). Today, one full-scope project (Kalinin) is almost complete, and three have been announced (Novovoronezh, Khmelnitsky, Kozloduy). Full-scope simulators are clearly the ultimate target of a concerned power plants. However, all users do realize the advantages of the complementary approach with compact simulators: 1. They can be available quickly for starting the training process. 2. They cover a training field that is not (or partly) addressed by full-scope simulators, i.e., the demonstration of physical phenomena in normal and accidental situations

  15. Automatic loading pattern optimization tool for Loviisa VVER-440 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kuopanportti, Jaakko [Fortum Power and Heat, Fortum (Finland). Nuclear Competence Center

    2013-09-15

    An automatic loading pattern optimization tool called ALPOT has been developed for Loviisa VVER-440 reactors. The ALPOT code utilizes combination of three different optimization methods. The first method is the imitation of the equilibrium pattern that is the optimized pattern in case the cycle length and the operation conditions are constant and the same shuffling pattern is repeated from cycle to cycle. In practice, the algorithm imitates assemblies' operation year distribution of the equilibrium pattern stochastically. The function of the imitation algorithm is to provide initial patterns quickly for the next optimization phase, which is performed either with the stochastic guided binary search algorithm or the deterministic burnup kernel method depending on the choice of the user. The former is a modified version of the standard binary search. The standard version goes through all possible swaps of the assemblies and chooses the best swap at each iteration round. The guided version chooses one assembly, tries to swap it with every other possible assembly and performs the best swap at each iteration round. The search is guided so that the algorithm chooses the assemblies at or near the most restrictive fuel assembly first. The kernel method creates burnup kernel functions to estimate burnup variations that are required to achieve desired changes in the power distribution of the reactor. The idea of the kernel method is first determine the optimal burnup distribution that minimizes the maximum relative assembly power using the created kernel functions and a common solver routine. Then, the burnups of the available fuel assemblies are matched with the obtained burnup distribution. (orig.)

  16. Automatic loading pattern optimization tool for Loviisa VVER-440 reactors

    International Nuclear Information System (INIS)

    Kuopanportti, Jaakko

    2013-01-01

    An automatic loading pattern optimization tool called ALPOT has been developed for Loviisa VVER-440 reactors. The ALPOT code utilizes combination of three different optimization methods. The first method is the imitation of the equilibrium pattern that is the optimized pattern in case the cycle length and the operation conditions are constant and the same shuffling pattern is repeated from cycle to cycle. In practice, the algorithm imitates assemblies' operation year distribution of the equilibrium pattern stochastically. The function of the imitation algorithm is to provide initial patterns quickly for the next optimization phase, which is performed either with the stochastic guided binary search algorithm or the deterministic burnup kernel method depending on the choice of the user. The former is a modified version of the standard binary search. The standard version goes through all possible swaps of the assemblies and chooses the best swap at each iteration round. The guided version chooses one assembly, tries to swap it with every other possible assembly and performs the best swap at each iteration round. The search is guided so that the algorithm chooses the assemblies at or near the most restrictive fuel assembly first. The kernel method creates burnup kernel functions to estimate burnup variations that are required to achieve desired changes in the power distribution of the reactor. The idea of the kernel method is first determine the optimal burnup distribution that minimizes the maximum relative assembly power using the created kernel functions and a common solver routine. Then, the burnups of the available fuel assemblies are matched with the obtained burnup distribution. (orig.)

  17. MOCA, Criticality of VVER Reactor Hexagonal Fuel Assemblies

    International Nuclear Information System (INIS)

    KYNCL, Jan

    1994-01-01

    1 - Description of program or function: Criticality problem in neutron transport for hexagonal fuel assembly in VVER nuclear reactor. The assembly is assumed to be either arranged in an infinite hexagonal array or placed in vacuum. The problem is solved in three- dimensional geometry, using standard energy group formalism and assuming that effective scattering cross sections are presented as Legendre polynomial expansions. The code evaluates ten different physical quantities, e.g. multiplication factor, neutron flux per energy group and spatial zone, integrated over angle and power in any zone of the assembly. 2 - Method of solution: Monte Carlo method of successive generations is applied. Computation proceeds according to an analog random process. The code is organized into three blocks: In the first block, the input data are converted to quantities for use in the Monte Carlo calculation. An initial neutron distribution is calculated, which corresponds to a fission spectrum uniform in spatial and angular variables. The main calculations are carried out in the second block (subroutine PROC2). This block is subdivided into geometrical and physical parts. Neutron tracks in individual zones and groups as well as probabilities for the formation of secondary neutrons are calculated. In the third block (subroutine PROC3), the results are evaluated statistically. Effective multiplication coefficients, the neutron flux per group and zone, and respective errors are computed. These quantities serve as a basis for the evaluation of other quantities. The results are either printed or stored for future evaluations. 3 - Restrictions on the complexity of the problem: In the PC version of the program, the maximum number of neutrons is 1000, the maximum number of energy groups is 4, and the maximum number of material compositions is 15. Angular expansion of scattering cross sections is allowed up to P10. These restrictions can easily be removed by increasing input parameters and

  18. Insights from Severe Accident Analyses for Verification of VVER SAMG

    Energy Technology Data Exchange (ETDEWEB)

    Gaikwad, A. J.; Rao, R. S.; Gupta, A.; Obaidurrahaman, K., E-mail: avinashg@aerb.gov.in [Nuclear Safety Analysis Division, Atomic Energy Regulatory Board, Mumbai (India)

    2014-10-15

    The severe accident analyses of simultaneous rupture of all four steam lines (case-a), simultaneous occurrence of LOCA with SBO (case-b) and Station blackout (case-c) were performed with the computer code ASTEC V2r2 for a typical VVER-1000. The results obtained will be used for verification of sever accident provisions and Severe Accident Management Guidelines (SAMG). Auxiliary feed water and emergency core cooling systems are modelled as boundary conditions. The ICARE module is used to simulate the reactor core, which is divided into five radial regions by grouping similarly powered fuel assemblies together. Initially, CESAR module computes thermal hydraulics in primary and secondary circuits. As soon as core uncovery begins, the ICARE module is actuated based on certain parameters, and after this, ICARE module computes the thermal hydraulics in the core, bypass, downcomer and the lower plenum. CESAR handles the remaining components in the primary and secondary loops. CPA module is used to simulate the containment and to predict the thermal-hydraulic and hydrogen behaviour in the containment. The accident sequences were selected in such a way that they cover low/high pressure and slow/fast core damage progression events. Events simulated included slow progression events with high pressure and fast accident progression with low primary pressure. Analysis was also carried out for the case of SBO with the opening of the PORVs when core exit temperature exceeds certain value as part of SAMG. Time step sensitivity study was carried out for LOCA with SBO. In general the trends and magnitude of the parameters are as expected. The key results of the above analyses are presented in this paper. (author)

  19. Safety design bases collation for Czech VVER NPPs

    International Nuclear Information System (INIS)

    Kadecka, P.; Krhounek, V.; Samohyl, P.; Zdarek, J.

    2004-01-01

    Goals of Safety Design Bases (SDB) Collation for Czech VVER NPPs are following: (i) Collation of SDB up to the component level. (ii) Preparation of necessary supporting information. (iii) The use of the effective knowledge management system DART (product WEE) for work organization and data storage. (iv) Use of the computer network between cooperating organizations with access to DART database. (v) Storage of DB information in format convenient for configuration, change and plant life management. It is concluded that SDB Collation for Czech NPPs continue very well. NPPs Management gives projects high priority and necessary amount of founds covering also active participation of NPPs staff. Current results of the projects are following: (i) DART (Knowledge management system) and other tools work on computer network link together all project participants. All information are stored to regularly backuped databases. (ii) Overall processes for SDB collation were designed and tested. Detail methodologies and working procedures for SDB collation of mechanical, civil, electro and I and C SSC were prepared. (iii) Detail workflow for all steps of the process was designed to organize different working groups of authors (write information) and reviewers (can annotate only or annotate and approve information). (iv) Pilot HP ECCS DBDs were prepared for both NPPs, it means whole methodology and procedures were tested. (v) Activity transfer trees, barrier integrity trees and fault trees for mechanical and civil SSC are finished. (vi) Fault trees for Dukovany NPP electro and I and C SSC were finished, for Temelin are under preparation. (vii) Fault gates description and functional requirements for Temelin mechanical and civil SSC are finished, for Dukovany NPP are under preparation. (P.A.)

  20. Radiolysis of the VVER-1000 reactor coolant: An experimental study and mathematical modeling

    International Nuclear Information System (INIS)

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

    1995-01-01

    Variations in the composition of the coolant for the primary circuit of a VVER-1000 reactor of the Kalinin nuclear power plant upon transition from power-level operation to shutdown was studied experimentally. The data obtained were used for verification of the MORAVA-H2 program developed earlier for simulation of the coolant state in pressurized-water power reactors

  1. Hextran-Smabre calculation of the VVER-1000 coolant transient benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Elina Syrjaelahti; Anitta Haemaelaeinen [VTT Processes, P.O.Box 1604, FIN-02044 VTT (Finland)

    2005-07-01

    Full text of publication follows: The VVER-1000 Coolant Transient benchmark is intended for validation of couplings of the thermal hydraulic codes and three dimensional neutron kinetic core models. It concerns a switching on a main coolant pump when the other three main coolant pumps are in operation. Problem is based on experiment performed in Kozloduy NPP in Bulgaria. In addition to the real plant transient, two extreme scenarios concerning control rod ejection after switching on a main coolant pump were calculated. In VTT the three-dimensional advanced nodal code HEXTRAN is used for the core kinetics and dynamics, and thermohydraulic system code SMABRE as a thermal hydraulic model for the primary and secondary loop. Parallelly coupled HEXTRAN-SMABRE code has been in production use since early 90's, and it has been extensively used for analysis of VVER NPPs. The SMABRE input model is based on the standard VVER-1000 input used in VTT. Last plant specific modifications to the input model have been made in EU projects. The whole core calculation is performed in the core with HEXTRAN. Also the core model is based on earlier VVER-1000 models. Nuclear data for the calculation was specified in the benchmark. The paper outlines the input models used for both codes. Calculated results are introduced both for the coupled core system with inlet and outlet boundary conditions and for the whole plant model. Sensitivity studies have been performed for selected parameters. (authors)

  2. Hextran-Smabre calculation of the VVER-1000 coolant transient benchmark

    International Nuclear Information System (INIS)

    Elina Syrjaelahti; Anitta Haemaelaeinen

    2005-01-01

    Full text of publication follows: The VVER-1000 Coolant Transient benchmark is intended for validation of couplings of the thermal hydraulic codes and three dimensional neutron kinetic core models. It concerns a switching on a main coolant pump when the other three main coolant pumps are in operation. Problem is based on experiment performed in Kozloduy NPP in Bulgaria. In addition to the real plant transient, two extreme scenarios concerning control rod ejection after switching on a main coolant pump were calculated. In VTT the three-dimensional advanced nodal code HEXTRAN is used for the core kinetics and dynamics, and thermohydraulic system code SMABRE as a thermal hydraulic model for the primary and secondary loop. Parallelly coupled HEXTRAN-SMABRE code has been in production use since early 90's, and it has been extensively used for analysis of VVER NPPs. The SMABRE input model is based on the standard VVER-1000 input used in VTT. Last plant specific modifications to the input model have been made in EU projects. The whole core calculation is performed in the core with HEXTRAN. Also the core model is based on earlier VVER-1000 models. Nuclear data for the calculation was specified in the benchmark. The paper outlines the input models used for both codes. Calculated results are introduced both for the coupled core system with inlet and outlet boundary conditions and for the whole plant model. Sensitivity studies have been performed for selected parameters. (authors)

  3. Impact of burnable absorber Gd on nuclide composition for VVER-440 fuel (Gd-2)

    International Nuclear Information System (INIS)

    Zajac, R.; Chrapciak, V.

    2010-01-01

    The latest version of Russian fuel VVER-440 includes burnable absorber in 6 pins. In this article is impact of burnable absorber on nuclide composition and criticality analyzed. In part 1 was analyzed whole burnup interval 0-50 MWd/kgU. In present part 2 are detailed analysis only for first cycle (burnup 0-10 MWd/kgU). (Authors)

  4. Water chemistry regimes for VVER-440 units: water chemistry influence on fuel cladding behaviour

    International Nuclear Information System (INIS)

    Zmitko, M.

    1999-01-01

    In this lecture next problems of water chemistry influence on fuel cladding behaviour for VVER-440 units are presented: primary coolant technologies; water chemistry specification and control; fuel integrity considerations; zirconium alloys cladding corrosion (corrosion versus burn-up; water chemistry effect; crud deposition; hydrogen absorption; axial offset anomaly); alternatives for the primary coolant regimes

  5. VVER operational safety improvements: lessons learnt from European co-operation and future research needs

    International Nuclear Information System (INIS)

    Pazdera, F.; Vasa, I.; Zd'arek, J.

    2003-01-01

    The paper summarises involvement of Nuclear Research Institute Rez (NRI) in the areas which are directly related to Reactor Operational Safety and Plant Life Management, it also gives an idea how results of the research projects can be used to enhance safety of VVER reactors. These issues are for many years subject of a wide international co-operation effort, covered by such programmes as PHARE, OECD/NEA TACIS, 5th Framework Programme. Nuclear Research Institute participated in the majority of these programmes and projects, which allowed us to evaluate benefits (especially for VVER reactors) of the projects already finalised or running, as well as to formulate so-called 'future research needs', which possibly may be pursued within 6th Framework Programme. The paper highlights the main features of some projects our Institute was and is involved in, emphasising the most important results, expectations and future needs. It also very briefly, deals with some general and particular lessons learnt within these projects and their application to VVER reactors, especially as to their safety improvement. The paper also mentions VVER-focused projects and activities, co-ordinated by the OECD, which should enable to extend multilateral contacts already existing between organisations of the EU countries to include organisations from Russia, USA, Japan and possibly some other countries

  6. Engineering Margin Factors Used in the Design of the VVER Fuel Cycles

    Science.gov (United States)

    Lizorkin, M. P.; Shishkov, L. K.

    2017-12-01

    The article describes methods for determination of the engineering margin factors currently used to estimate the uncertainties of the VVER reactor design parameters calculated via the KASKAD software package developed at the National Research Center Kurchatov Institute. These margin factors ensure the meeting of the operating (design) limits and a number of other restrictions under normal operating conditions.

  7. Inspection qualification as a tool to risk based ET ISI of VVER type SG tubes

    International Nuclear Information System (INIS)

    Horacek, L.

    2002-01-01

    A Pilot study on Eddy current inspection qualification of VVER 440 steam generator tubes, discussed in this paper, followed the ENIQ methodology principles and covered briefly the assumed scope of ET qualification, relevant elaborated qualification documents, known ISI limitations and a review of input information on component and defects determined for Eddy current inspection qualification of VVER 440 steam generator tubes. The information includes the fabrication of the test blocks with SG tube segments provided by intended defect simulations of realistic SCC type and basic data on the realistic SCC type defects manufacturing technology. Lessons learned from the development of manufacturing technology of SSC type of defects, regional blind tests, elaboration of the preliminary technical justification for Eddy current automated inspections, potential optimisation of inspection procedures, laboratory and practical open trials are summarised in the paper. The results of the Pilot study also especially in relation to POD curve being determined seem to be useful for practical operational ISI programme and Risk informed ISI decisions and the establishment of plugging criteria of VVER 440 and VVER 1000 type steam generator tubes. (orig.)

  8. Investigation of primary-to-secondary leakage accident on the PSB-VVER integral test facility

    International Nuclear Information System (INIS)

    Lipatov, I.A.; Dremin, G.I.; Galtchanskaya, S.A.; Chmal, I.I.; Moloshnikov, A.S.; Gorbunov, Y.S.; Antonova, A.I.; Elkin, I.V.

    2001-01-01

    The full text follows. The paper presents the main results from the test on primary-to-secondary leakage of 100 mm in equivalent diameter. The test was performed on the PSB-VVER integral test facility. PSB-VVER is a 4-loops scaled down model of primary system of NPP with VVER-1000 Russian type reactor. Volume - power scale is about 1/300 while elevation scale is 1/1. All components of the primary system of the reference NPP are modeled on PSB-VVER. Both passive (accumulators) and active (high and low pressure) ECCSs, pressurizer spray and relief circuits, feed water system and atmospheric dumping system (ADS) as well as the primary circuit gas remove emergency system are also simulated. The primary-to-secondary leakage was simulated using an external break line which connects the upper part of the hot header to SG water volume. The break line included a break nozzle (a cylindrical channel d = 5.8 mm, l/d = 10 with sharp inlet edge), quick-acting valve and two-phase mass flow rate measurement system. In addition loss of off-site power at the moment when a scram-signal is generated was assumed in the experiment. Thus the accident is to be considered as a beyond-design-basic one. The loss of off-site power results in the following: -main circulation pump shutdown; -pressurizer heaters switching off; -HPIS water cooling flow rate and number of points of water injection are reduced The study focuses on the adequacy of the associated accident management (AM) procedure developed by EDO ''GIDROPRESS'' as a General Designer of VVER-type reactors. The AM-procedure was adopted to the PSB-VVER test facility conditions using CATHARE (France) and DINAMIKA (Russia) codes analysis. The AM-procedure in PSB-VVER is as follows: after about 30 min of the onset of the accident, when the accident type and the localization of the SG affected become evident for the operator, he closes all the main steam isolation valves, inhibits the ADS actuation in the affected SG and begins to remove

  9. Investigation of primary-to-secondary leakage accident on the PSB-VVER integral test facility

    Energy Technology Data Exchange (ETDEWEB)

    Lipatov, I.A.; Dremin, G.I.; Galtchanskaya, S.A.; Chmal, I.I.; Moloshnikov, A.S.; Gorbunov, Y.S.; Antonova, A.I. [Electrogorsk Research and Engineering Center, EREC, Moscow (Russian Federation); Elkin, I.V. [RRC ' ' Kurchatov Institute, Moscow (Russian Federation)

    2001-07-01

    The full text follows. The paper presents the main results from the test on primary-to-secondary leakage of 100 mm in equivalent diameter. The test was performed on the PSB-VVER integral test facility. PSB-VVER is a 4-loops scaled down model of primary system of NPP with VVER-1000 Russian type reactor. Volume - power scale is about 1/300 while elevation scale is 1/1. All components of the primary system of the reference NPP are modeled on PSB-VVER. Both passive (accumulators) and active (high and low pressure) ECCSs, pressurizer spray and relief circuits, feed water system and atmospheric dumping system (ADS) as well as the primary circuit gas remove emergency system are also simulated. The primary-to-secondary leakage was simulated using an external break line which connects the upper part of the hot header to SG water volume. The break line included a break nozzle (a cylindrical channel d = 5.8 mm, l/d = 10 with sharp inlet edge), quick-acting valve and two-phase mass flow rate measurement system. In addition loss of off-site power at the moment when a scram-signal is generated was assumed in the experiment. Thus the accident is to be considered as a beyond-design-basic one. The loss of off-site power results in the following: -main circulation pump shutdown; -pressurizer heaters switching off; -HPIS water cooling flow rate and number of points of water injection are reduced The study focuses on the adequacy of the associated accident management (AM) procedure developed by EDO ''GIDROPRESS'' as a General Designer of VVER-type reactors. The AM-procedure was adopted to the PSB-VVER test facility conditions using CATHARE (France) and DINAMIKA (Russia) codes analysis. The AM-procedure in PSB-VVER is as follows: after about 30 min of the onset of the accident, when the accident type and the localization of the SG affected become evident for the operator, he closes all the main steam isolation valves, inhibits the ADS actuation in the affected SG

  10. Safety enhancement concept for NPP of new generation with VVER reactors

    International Nuclear Information System (INIS)

    Bezlepkin, V.; Kukhtevich, I.; Semashko, S.; Svetlov, S.; Solodovnikov, A.

    2004-01-01

    With the present day conditions, in order to successfully promote new NPP designs in the electric power markets, it is necessary to ensure enhanced technical/economic performances provided that international safety requirements are properly adhered to. When compared with high-powered nuclear power plants, NPP VVER-640 design (medium powered) possesses a number of advantages for the regions with undeveloped energy systems. Reduced specific energy intensity of the core adopted in this type of reactor allows to ensure the emergency cooldown of the reactor plant by passive means and to minimize the 'human factor' risk and external effects and provide sound substantiations as to how to retain corium inside RPV in case of severe accidents. At the same time, high-powered NPPs seem to be promising for regions with developed energy systems. Among such designs, NPP VVER-1000 and VVER-1500 designs are the most desirable. Configuration of new generation NPP with VVER-1500 is to be selected based on the gained experience in designing NPPs of previous generations considering the latest safety requirements and situation in the domestic and global energy markets for the time being and in the short run. Recent IAEA publications and latest EUR requirements insist that the following key safety indices should be established for new NPP designs: - aggregated frequency of core melting is 10 -6 (1/year); - frequency of maximum accident release is 10 -7 (1/year). To meet the aforementioned criteria, it is necessary to implement some safety assurance principles recommended by IAEA (in-depth defence, single failure, redundancy, diversity, etc.), application of deterministic and probabilistic methods for selection of safety assurance activities and means and use of reasonable combination of active and passive systems. Application of VVER-640 concept to high-powered NPPs seems to be a formidable task due to a number of reasons, namely, it is quite difficult to carry out cooldown process

  11. PSB-VVER simulation of Kozloduy NPP 'loss of feed water transient'

    Energy Technology Data Exchange (ETDEWEB)

    Groudev, P.P. [Institute for Nuclear Research and Nuclear Energy, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)]. E-mail: pavlinpg@inrne.bas.bg; Stefanova, A.E. [Institute for Nuclear Research and Nuclear Energy, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)]. E-mail: antoanet@inrne.bas.bg; Gencheva, R.V. [Institute for Nuclear Research and Nuclear Energy, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)]. E-mail: roseh@inrne.bas.bg; Pavlova, M.P. [Institute for Nuclear Research and Nuclear Energy, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)]. E-mail: pavlova@inrne.bas.bg

    2005-04-01

    This paper provides a comparison between the PSB test facility experimental results obtained during the simulation of loss of feed water transient (LOFW) and the calculation results received by INRNE computer model of the same test facility. Integral thermal-hydraulic PSB-VVER test facility located at Electrogorsk Research and Engineering Center on NPPs Safety (EREC) was put in operation in 1998. The structure of the test facility allows experimental studies under steady state, transient and accident conditions. RELAP5/MOD3.2 computer code has been used to simulate the loss of feed water transient in a PSB-VVER model. This model was developed at the Institute for Nuclear Research and Nuclear Energy for simulation of loss of feed water transient. The objective of the experiment 'loss of feed water', which has been performed at PSB-VVER test facility is simulation of Kozloduy NPP LOFW transient. One of the main requirements to the experiment scenario has been to reproduce all main events and phenomena that occurred in Kozloduy NPP during the LOFW transient. Analyzing the PSB-VVER test with a RELAP5/MOD3.2 computer code as a standard problem allows investigating the phenomena included in the VVER code validation matrix as 'integral system effects' and 'natural circulation'. For assessment of the RELAP5 capability to predict the 'Integral system effect' phenomenon the following RELAP5 quantities are compared with external trends: the primary pressure and the hot and cold leg temperatures. In order to assess the RELAP5 capability to predict the 'Natural circulation' phenomenon the hot and cold leg temperatures behavior have been investigated. This report was possible through the participation of leading specialists from Kozloduy NPP and with the support of Argonne National Laboratory (ANL), under the International Nuclear Safety Program (INSP) of the United States Department of Energy.

  12. Results of 15 years experiments in the PMK-2 integral-type facility for VVERs

    Energy Technology Data Exchange (ETDEWEB)

    Szabados, L.; Ezsoel, G.; Perneczky, L. [KFKI Atomic Energy Research Institute, Budapest (Hungary)

    2001-07-01

    Due to the specific features of the VVER-440/213-type reactors the transient behaviour of such a reactor system is different from the usual PWR system behaviour. To provide an experimental database for the transient behaviour of VVER systems the PMK integral-type facility, the scaled down model of the Paks NPP was designed and constructed in the early 1980's. Since the start-up of the facility 48 experiments have been performed. It was confirmed through the experiments that the facility is a suitable tool for the computer code validation experiments and to the identification of basic thermal-hydraulic phenomena occurring during plant accidents. High international interest was shown by the four Standard Problem Exercises of the IAEA and by the projects financed by the EU-PHARE. A wide range of small- and medium-size LOCA sequences have been studied to know the performance and effectiveness of ECC systems and to evaluate the thermal-hydraulic safety of the core. Extensive studies have been performed to investigate the one- and two-phase natural circulation, the effect of disturbances coming from the secondary circuit and to validate the effectiveness of accident management measures like bleed and feed. The VVER-specific case, the opening of the SG collector cover was also extensively investigated. Examples given in the report show a few results of experiments and the results of calculation analyses performed for validation purposes of codes like RELAP5, ATHLET and CATHARE. There are some other white spots in Cross Reference Matrices for VVER reactors and, therefore, further experiments are planned to perform tests primarily in further support of accident management measures at low power states of plants to facilitate the improved safety management of VVER-440-type reactors. (authors)

  13. Results of 15 years experiments in the PMK-2 integral-type facility for VVERs

    International Nuclear Information System (INIS)

    Szabados, L.; Ezsoel, G.; Perneczky, L.

    2001-01-01

    Due to the specific features of the VVER-440/213-type reactors the transient behaviour of such a reactor system is different from the usual PWR system behaviour. To provide an experimental database for the transient behaviour of VVER systems the PMK integral-type facility, the scaled down model of the Paks NPP was designed and constructed in the early 1980's. Since the start-up of the facility 48 experiments have been performed. It was confirmed through the experiments that the facility is a suitable tool for the computer code validation experiments and to the identification of basic thermal-hydraulic phenomena occurring during plant accidents. High international interest was shown by the four Standard Problem Exercises of the IAEA and by the projects financed by the EU-PHARE. A wide range of small- and medium-size LOCA sequences have been studied to know the performance and effectiveness of ECC systems and to evaluate the thermal-hydraulic safety of the core. Extensive studies have been performed to investigate the one- and two-phase natural circulation, the effect of disturbances coming from the secondary circuit and to validate the effectiveness of accident management measures like bleed and feed. The VVER-specific case, the opening of the SG collector cover was also extensively investigated. Examples given in the report show a few results of experiments and the results of calculation analyses performed for validation purposes of codes like RELAP5, ATHLET and CATHARE. There are some other white spots in Cross Reference Matrices for VVER reactors and, therefore, further experiments are planned to perform tests primarily in further support of accident management measures at low power states of plants to facilitate the improved safety management of VVER-440-type reactors. (authors)

  14. The corrosion and corrosion mechanical properties evaluation for the LBB concept in VVERs

    Energy Technology Data Exchange (ETDEWEB)

    Ruscak, M.; Chvatal, P.; Karnik, D.

    1997-04-01

    One of the conditions required for Leak Before Break application is the verification that the influence of corrosion environment on the material of the component can be neglected. Both the general corrosion and/or the initiation and, growth of corrosion-mechanical cracks must not cause the degradation. The primary piping in the VVER nuclear power plant is made from austenitic steels (VVER 440) and low alloy steels protected with the austenitic cladding (VVER 1000). Inspection of the base metal and heterogeneous weldments from the VVER 440 showed that the crack growth rates are below 10 m/s if a low oxygen level is kept in the primary environment. No intergranular cracking was observed in low and high oxygen water after any type of testing, with constant or periodic loading. In the framework of the LBB assessment of the VVER 1000, the corrosion and corrosion mechanical properties were also evaluated. The corrosion and corrosion mechanical testing was oriented predominantly to three types of tests: stress corrosion cracking tests corrosion fatigue tests evaluation of the resistance against corrosion damage. In this paper, the methods used for these tests are described and the materials are compared from the point of view of response on static and periodic mechanical stress on the low alloyed steel 10GN2WA and weld metal exposed in the primary circuit environment. The slow strain rate tests and static loading of both C-rings and CT specimens were performed in order to assess the stress corrosion cracking characteristics. Cyclic loading of CT specimens was done to evaluate the kinetics of the crack growth under periodical loading. Results are shown to illustrate the approaches used. The data obtained were evaluated also from the point of view of comparison of the influence of different structure on the stress corrosion cracking appearance. The results obtained for the base metal and weld metal of the piping are presented here.

  15. Comparison of problems and experience of core operation with distorted fuel element assemblies in VVER-1000 and PWR reactors

    International Nuclear Information System (INIS)

    Afanas'ev, A.

    1999-01-01

    The main reactors leading to distortion of fuel element assemblies during reactor operation were studied. A series of actions which compensate this effect was proposed. Criteria of operation limitation in VVER-1000 and PWR reactors are described

  16. Burnup Estimation of Rhodium Self-Powered Neutron Detector Emitter in VVER Reactor Core Using Monte Carlo Simulations

    OpenAIRE

    Khrutchinsky, А. А.; Kuten, S. A.; Babichev, L. F.

    2011-01-01

    Estimation of burn-up in a rhodium-103 emitter of self-powered neutron detector in VVER-1000 reactor core has been performed using Monte Carlo simulations within approximation of a constant neutron flux.

  17. Application of the Defense-in-Depth Concept in the Projects of New-Generation NPPs Equipped with VVER Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shvyryaev, Yu. V.; Morozov, V. B.; Kuchumov, A.Yu., E-mail: morozov@aep.ru [JSC Atomenergoproekt, Moscow (Russian Federation)

    2014-10-15

    The projects of new-generation NPPs equipped with VVER reactors are developed as projects the safety level of which is superior to that of NPPs that are currently in operation. The main design solutions adopted for implementing the defence-in-depth (DiD) concept in the projects of new-generation NPPs equipped with VVER reactors are briefly characterized in the paper. (author)

  18. Enhancing VVER annular proliferation resistance fuel with minor actinides

    International Nuclear Information System (INIS)

    Chang, G. S.

    2007-01-01

    to the reactivity control of the systems into which they are incorporated. In the study, a typical pressurized water reactor (PWR) VVER-1000 annular fuel unit lattice cell model with UO 2 fuel pins will be used to investigate the effectiveness of minor actinide reduction approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance in the intermediate term goal for future nuclear energy systems

  19. Evaluation of carbon-14 life cycle in reactors VVER-1000

    International Nuclear Information System (INIS)

    Lysakova, Katerina; Neumann, Jan; Vonkova, Katerina

    2012-09-01

    This work is aimed at the evaluation of carbon-14 life cycle in light water reactors VVER-1000. Carbon-14 is generated as a side product in different systems of nuclear reactors and has been an issue not only in radioactive waste management but mainly in release into the environment in the form of gaseous effluents. The principal sources of this radionuclide are in primary cooling water and fuel. Considerable amount of C-14 is generated by neutron reactions with oxygen 17 O and nitrogen 14 N present in water coolant and fuel. The reaction likelihood and consequently volume of generated radioisotope depends on several factors, especially on the effective cross-section, concentrations of parent elements and conditions of power plant operating strategies. Due to its long half-life and high capability of integration into the environment and thus into the living species, it is very important to monitor the movement of carbon-14 in all systems of nuclear power plant and to manage its release out of NPP. The dominant forms of radioactive carbon-14 are the hydrocarbons owing to the combinations with hydrogen used for absorption of radiolytic oxygen. These organic compounds, such as formaldehyde, methyl alcohol, ethyl alcohol and formic acid can be mostly retained on ion exchange resins used in the system for purifying primary cooling water. The gaseous carbon compounds (CH 4 and CO 2 ) are released into the atmosphere via the ventilation systems of NPP. Based on the information and data obtained from different sources, it has been designed a balance model of possible carbon-14 pathways throughout the whole NPP. This model includes also mass balance model equations for each important node in system and available sampling points which will be the background for further calculations. This document is specifically not to intended to describe the best monitoring program attributes or technologies but rather to provide evaluation of obtained data and find the optimal way to

  20. Formation of radiation induced precipitates in VVER RPV materials

    International Nuclear Information System (INIS)

    Platonov, P.A.; Chernobaeva, A.A.

    2016-01-01

    This paper presents an analysis of experimental results received in course of research of copper-enriched precipitates (Cu-precipitates) and nickel-manganese-silicon clusters (Ni-Mn-Si clusters), which are formed in steels of VVER-type reactor pressure vessels (RPVs) under neutron irradiation. Based on this analysis, a hypothetical model is suggested for cluster formation in course of evolution of a cascade region. The model presumes cluster formation in two stages. At the first stage, in course of cascade region crystallization, a stable cluster is formed in the center of the cascade region, which consists of vacancies and Cu atoms following the mechanism of the inverse Kirkendall effect. At the second stage, diffusion of Ni, Mn and P atoms with a flow of vacancies from the matrix takes place to form a cluster. The size of a cluster is limited by a balance of vacancies' flows entering and leaving the cluster. The paper also considers a possibility of stabilization of atomic-vacancy cluster due to uneven distribution of Ni, Mn and P atoms, which explains dependence of cluster density on the content of these elements. Kinetics of cluster formation and evolution presumed by suggested model is analyzed. It is demonstrated that a fall in cluster density and an increase in their size under high irradiation doses may be caused by a decrease of matrix supersaturation with vacancies resulting from high density of dislocation loops. - Highlights: • The analysis of the mechanism of formation of radiation-induced clusters in RPV steels has been done. • Radiation-induced clusters are formed after the mechanism based on the inverse Kirkendall effect in two stages. • At post-dynamic stage a flow of vacancies moving to the center of the cascade entrains Cu atoms contained and forms a stable atom-vacancies cluster. • At the 2nd stage Cu, Ni, Mn, Si atoms forming complexes with vacancies diffuse into a cluster driving out Fe and Cr atoms from the cluster. • The cluster

  1. Investigation of accident management procedures related to loss of feedwater and station blackout in PSB-VVER integral test facility

    Energy Technology Data Exchange (ETDEWEB)

    Bucalossi, A. [EC JRC, (JRC F.5) PO Box 2, 1755 ZG Petten (Netherlands); Del Nevo, A., E-mail: alessandro.delnevo@enea.it [ENEA, C.R. Brasimone, 40032 Camugnano (Italy); Moretti, F.; D' Auria, F. [GRNSPG, Universita di Pisa, via Diotisalvi 2, 56100 Pisa (Italy); Elkin, I.V.; Melikhov, O.I. [Electrogorsk Research and Engineering Centre, Electrogorsk, Moscow Region (Russian Federation)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer Four integral test facility experiments related to VVER-1000 reactor. Black-Right-Pointing-Pointer TH response of the VVER-1000 primary system following total loss of feedwater and station blackout scenarios. Black-Right-Pointing-Pointer Accident management procedures in case of total loss of feedwater and station blackout. Black-Right-Pointing-Pointer Experimental data represent an improvement of existing database for TH code validation. - Abstract: VVER 1000 reactors have some unique and specific features (e.g. large primary and secondary side fluid inventory, horizontal steam generators, core design) that require dedicated experimental and analytical analyses in order to assess the performance of safety systems and the effectiveness of possible accident management strategies. The European Commission funded project 'TACIS 2.03/97', Part A, provided valuable experimental data from the large-scale (1:300) PSB-VVER test facility, investigating accident management procedures in VVER-1000 reactor. A test matrix was developed at University of Pisa (responsible of the project) with the objective of obtaining the experimental data not covered by the OECD VVER validation matrix and with main focus on accident management procedures. Scenarios related to total loss of feed water and station blackout are investigated by means of four experiments accounting for different countermeasures, based on secondary cooling strategies and primary feed and bleed procedures. The transients are analyzed thoroughly focusing on the identification of phenomena that will challenge the code models during the simulations.

  2. Poddawanie się badaniom profilaktycznym przez osoby z przewlekłym wirusowym zapaleniem wątroby typu C = Submission to preventive examinations by people with chronic hepatitis C

    Directory of Open Access Journals (Sweden)

    Marta Estera Kowalska

    2016-07-01

    2.      SKN przy Samodzielnej Pracowni Epidemiologii Uniwersytetu Medycznego w Lublinie     Streszczenie Wstęp Wirusowe zapalenie wątroby typu C stanowi jeden z głównych problemów zdrowia publicznego ze względu na długi okres bezobjawowy zakażenia wirusem HCV. Cel pracy Celem pracy była analiza zachowania zdrowotnego jakim jest poddawanie się badaniom profilaktycznym przed i po wykryciu choroby osób z wirusowym zapaleniem wątroby typu C. Materiał i metody Materiałem wykorzystanym do badań były odpowiedzi na pytania zamknięte zawarte w autorskim kwestionariuszu ankiety. W badaniu wzięło udział 200 osób z potwierdzonym zakażeniem HCV. Wyniki Przed wykryciem choroby badana grupa deklarowała, że poddawała się badaniom profilaktycznym sporadycznie, jedynie gdy zaistniały niepokojące objawy.   Słowa kluczowe: wirusowe zapalenie wątroby typu C, profilaktyka.       Summary Introduction Hepatitis C is a major public health problems due to the long latency period of infection with HCV (hepatitis C virus. Aim The aim of the study was to analyze the behavior of health which is to submission to preventive tests before and after the detection of the disease people with hepatitis C. Material and methods Material used for the study were answers to closed questions included in the original questionnaire survey. The study involved 200 people with chronic hepatitis C. Results Before detection of of the disease studied group declared that they give up preventive testing occasionally, only when any worrying symptoms. Keywords: hepatitis C, prevention.

  3. Status and prospects of the core surveillance system SCORPIO-VVER in Czech Republic and Slovakia

    International Nuclear Information System (INIS)

    Molnar, J.; Vocka, R.

    2011-01-01

    The SCORPIO-VVER core monitoring system has proved since the first installation at Dukovany NPP in 1999 to be a valuable tool for the reactor operators and reactor physicists. It is now installed on four units of Dukovany NPP (Czech Republic) and two units of Bohunice NPP (Slovak Republic) replacing the original Russian VK3 system. By both Czech and Slovak nuclear regulatory bodies it was licensed as a Technical Specification Surveillance tool. Since it's first installation, the development of SCORPIO-VVER system continues along with the changes in WWER reactors operation. The system is being adapted according the utility needs and several notable improvements in physical modules of the system were introduced. The latest most significant changes were done in connection with implementation of a new digital I and C system, loading of the optimized gadolinium bearing Gd2 fuel assemblies, improvements in the area of core design (neutron physics, core thermal hydraulics and fuel thermal mechanics), adaptation of the system to up-rated unit conditions, in design and methodology of the limit and technical specifications checking (implementation of the on-line shutdown margin calculation to the system) and improvements in the predictive part of the system (Strategy Generator). After the currently finished upgrades (Upgrade 2 at EBO Slovakia in 2009, Upgrade 5 at EDU Czech Republic in 2010) the SCORPIO-VVER is still in focus of Central European nuclear power plants with the road map of modification and implementation up to 2015. In April 2011 the Upgrade 3 at EBO Slovakia has been contracted to support the changed reactor technical specification and for support of the new type of fuel planned to load in 2012. During the summer of 2011 the discussions started with EDU NPP in Czech Republic regarding to the future development of the SCORPIO-VVER system up to 2015. Parallel with the support of current installations at NPPs the project of new installations is ongoing. During

  4. Przygotowanie pacjentów z cukrzycą typu 2 do profilaktyki zespołu stopy cukrzycowej = Preparing patients with type 2 diabetes to prophylaxis of diabetic foot

    OpenAIRE

    Kropornicka, Beata; Baczewska, Bożena; Dadej, Aleksandra; Krzyżanowska, Ewa; Łuczyk, Robert; Olszak, Cecylia; Nowicka, Elżbieta; Lorencowicz, Regina; Muraczyńska, Bożena

    2016-01-01

    Kropornicka Beata, Baczewska Bożena, Dadej Aleksandra, Krzyżanowska Ewa, Łuczyk Robert, Olszak Cecylia, Nowicka Elżbieta, Lorencowicz Regina, Muraczyńska Bożena. Przygotowanie pacjentów z cukrzycą typu 2 do profilaktyki zespołu stopy cukrzycowej = Preparing patients with type 2 diabetes to prophylaxis of diabetic foot. Journal of Education, Health and Sport. 2016;6(11):311-324. eISSN 2391-8306. DOI http://dx.doi.org/10.5281/zenodo.167141 http://ojs.ukw.edu.pl/index.php/johs/article/view/3...

  5. OECD/DOE/CEA VVER-1000 coolant transient (V1000CT) benchmark for assessing coupled neutronics/thermal-hydraulics system codes for VVER-1000 RIA analysis

    International Nuclear Information System (INIS)

    Ivanov, B.; Ivanov, K.; Aniel, S.; Royer, E.; Kolev, N.; Groudev, P.

    2004-01-01

    The present paper describes the two phases of the OECD/DOE/CEA VVER-1000 coolant transient benchmark labeled as V1000CT. This benchmark is based on a data from the Bulgarian Kozloduy NPP Unit 6. The first phase of the benchmark was designed for the purpose of assessing neutron kinetics and thermal-hydraulic modeling for a VVER-1000 reactor, and specifically for their use in analyzing reactivity transients in a VVER-1000 reactor. Most of the results of Phase 1 will be compared against experimental data and the rest of the results will be used for code-to-code comparison. The second phase of the benchmark is planned for evaluation and improvement of the mixing computational models. Code-to-code and code-to-data comparisons will be done based on data of a mixing experiment conducted at Kozloduy-6. Main steam line break will be also analyzed in the second phase of the V1000CT benchmark. The results from it will be used for code-to-code comparison. The benchmark team has been involved in analyzing different aspects and performing sensitivity studies of the different benchmark exercises. The paper presents a comparison of selected results, obtained with two different system thermal-hydraulics codes, with the plant data for the Exercise 1 of Phase 1 of the benchmark as well as some results for Exercises 2 and 3. Overall, this benchmark has been well accepted internationally, with many organizations representing 11 countries participating in the first phase of the benchmark. (authors)

  6. Task 08/41, Low temperature loop at the RA reactor, Review I of materials planned for irradiation within the five-year plan of nuclear power development in Federal Peoples Republic of Yugoslavia; Zadatak 08/41, Niskotemperaturna petlja u reaktoru 'A', Pregled I materijala predvidjenih za ozracivanje petogodisnjim planom razvoja nuklearne energije FNRJ

    Energy Technology Data Exchange (ETDEWEB)

    Milenkovic, S; Milasin, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    For the purpose of designing the low-temperature loop at the RA reactor, a list of materials that would be irradiated in the reactor was made. This review includes materials which should be irradiated in for isotope production and materials to be irradiated for research and testing purposes. The neutron flux needed for materials testing should be much higher than presently available neutron flux of 10{sup 12} n/cm{sup 2} sec. [Serbo-Croat] Za potrebe projektovanja niskotemperaturne petlje na reaktoru RA nacinjen je pregled materijala za ozracivanje u reaktoru. Pregled sadrzi materijale za ozracivanje u cilju proizvodnje izotopa i materijale za specijalna ozracivanja u cilju ispitivanja materijala. Za ispitivanje materijala potreban bi bio fluks neutron koji je znatno veci od raspolozivog fluksa od 10{sup 12} n/cm{sup 2} sec.

  7. Devices for irradiation of materials in the fast neutron flux at the RA heavy water reactor in Vinca; Uredjaji za ozracivanje materijala u fluksu brzih neutrona na teskovodnom reaktoru RA u Vinci

    Energy Technology Data Exchange (ETDEWEB)

    Pavicevic, M [The Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1965-07-01

    Full text: This paper covers the concept, design description and problems of constructing special experimental devices at the RA reactor in Vinca. These devices were used for irradiation of construction materials (graphite, Mg and Al oxides, steel and zircaloy) in the integral fast neutron flux of 2 x 10{sup 20} n/cm{sup 2}. Temperature of samples cooled by heavy water circulating through the hollow uranium fuel elements was measured by thermocouples. Construction of the device, preparation of samples, related flux measurements as well as irradiation were done in cooperation with the nuclear center in Saclay, France. Brief review of the experiences gained in using these experimental spaces at the RA reactor as well as problems of transporting the irradiated capsules from Vinca to France is included. [Serbo-Croat] Pun tekst: Rad obuhvata koncepciju, tehnicki opis i probleme realizacije specijalnih eksperimentalnih uredjaja u reaktoru RA u Vinci, u kojima je tokom 1962. godine izvrseno, na temperaturama ispod 100 deg C, ozracivanje konstrukcionih materijala (grafita, Mg i Al-oksida, celika i cirkaloja) u integralnom fluksu brzih neutrona vrednosti 2 x 10{sup 20} n/cm{sup 2}. Temperatura uzoraka hladjenih teskom vodom, koja protice kroz suplje uranske elemente, merena je termoparovima. Realizacija uredjaja, priprema uzoraka, odgovarajuca merenja fluksa kao i sama ozracivanja vrsena su u saradnji sa Nuklearnim centrom u Saclay-u, Francuska. Dat je kraci pregled iskustava iz eksploatacije ovih novih eksperimentalnih prostora u reaktoru RA, kao i neki problemi vezani za transport aktivnih kapsula sa uzorcima od Vince do Saclay-a (author)

  8. Assessment of computer codes for VVER-440/213-type nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Szabados, L.; Ezsol, Gy.; Perneczky [Atomic Energy Research Institute, Budapest (Hungary)

    1995-09-01

    Nuclear power plant of VVER-440/213 designed by the former USSR have a number of special features. As a consequence of these features the transient behaviour of such a reactor system should be different from the PWR system behaviour. To study the transient behaviour of the Hungarian Paks Nuclear Power Plant of VVER-440/213-type both analytical and experimental activities have been performed. The experimental basis of the research in the PMK-2 integral-type test facility , which is a scaled down model of the plant. Experiments performed on this facility have been used to assess thermal-hydraulic system codes. Four tests were selected for {open_quotes}Standard Problem Exercises{close_quotes} of the International Atomic Energy Agency. Results of the 4th Exercise, of high international interest, are presented in the paper, focusing on the essential findings of the assessment of computer codes.

  9. Inspection qualification programme for VVER reactors and review of round robin test results

    International Nuclear Information System (INIS)

    Horacek, L.; Zdarek, J.

    1998-01-01

    Experience obtained, especially from in-service inspections of VVER 440-type reactor pressure vessels and from the Czech round test trials with international participation of ultrasonic teams, has highlighted the need for an in-service inspection qualification programme in the Czech Republic focused on NDT procedures, equipment and personnel. Recently, several national and international regional projects included in the PHARE programme (projects 4.1.2/93 and 1.02/94), briefly described, have been initiated. These projects are to cover step by step the programme of the in-service inspection qualification in view of technical justification as well as of practical assessment-performance demonstration-for all the main VVER-type primary circuit components. (orig.)

  10. Analysing The Thermalhydraulic Parameters Of VVER-1000 Reactor For The Accident Of Steam Generator Tube Rupture

    International Nuclear Information System (INIS)

    Luu Nam Hai; Truong Cong Thang

    2011-01-01

    To ensure the safety operation of nuclear power plant (NPP), a lot of postulated accident scenarios were considered and analysed. This research chose and analysed the accident of steam generator tube rupture (SGTR) under the actual plant conditions by using the simulation program PCTRAN. The SGTR accident is happen when the NPP is under operation with the steady state condition (power of 3000 MWth, primary pressure of 157 bar and secondary pressure of 63 bar). The accident is initiated by creating a break with equivalent diameter of 100 mm in the area of lower row heat exchanging tubes. The result of analysis is compared with the calculation of the Shariz University, Iran using the thermal hydraulics code RELAP5/mod3.2 and the report in the PSAR data of VVER-1000. This comparison shows that it is possible for using PCTRAN to analyse accidents of VVER-1000 reactor. (author)

  11. Core designs for new VVER reactors and operational experience of immediate prototypes

    International Nuclear Information System (INIS)

    Vasilchenko, I.; Mokhov, V.; Ryzhov, S.

    2011-01-01

    The paper covers the recent improvements analyzed in order to implement the enhanced core performances. AES-2006 reactor core design is considered from the point of view of its application and improvement in the planned VVER-TOI project and of the possibilities of using the basic engineering solutions for the cores with spectral control. The discussion of several types of mixing grids considered in the paper involves a preliminary assessment of their efficiency and the information on their introduction into pilot operation at the VVER-1000 Units. Special attention is given to the results of the operation of immediate prototypes (TVS-2 and TVS-2M) that corroborate the reliability of the design both with regard for the core geometrical stability and fuel cladding tightness

  12. Fusion of eastern and western technology in VVER 1000 NPP upgrade

    International Nuclear Information System (INIS)

    Ubra, O.; Fleischhans, J.; Kveton, M.

    1997-01-01

    An extensive modernization program upgrading two units of VVER 1000 type of the Czech nuclear power plant (NPP) Temelin to meet the latest international standards is presented. The program is based primarily on combination of eastern and western technology and it has been implemented during plant construction. The NPP Temelin was originally designed according to the standards of the former Soviet Union. After a series of reviews in the 1990s, a decision was made by the Temelin management of upgrade the design of the plant, including the supply of fuel and instrumentation and control system by a western company. The adoption of western technology and practices has helped to solve a large number of IAEA safety issues related to design and operation of VVER 1000 NPP. Details on the current Temelin design and other related safety matters are presented

  13. Control of radiation fields at the RA reactor (Radiological data relevant for the period when irregularities on the fuel elements were observed), Part 2, Annex 1; Kontrola radijacionih polja na reaktoru RA (Neki podaci relevantni za ocenu radijacione situacije na reaktoru za period u kome su otkrivene neregularnosti na reaktorskom gorivu), Prilog 1, Deo 2

    Energy Technology Data Exchange (ETDEWEB)

    Ninkovic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1980-10-15

    period in question to prove the above conclusion. [Serbo-Croat] U sklopu programa mera zastite od zracenja u radnoj sredini na reaktoru RA vrsi se kontrola i merenje osnovnih karakteristika radijacionih polja: nivoi gama zracenja na karaktoristicnim tackama reaktorskih sistema, nivoi gamma zracenja i neutrona u pojedinim radnim i tehnoloskim prostorijama, koncentracija radioaktivnih aerosola i sadrzaj radioaktivnih gasova u vazduhu radnih i tehnoloskih prostorija, nivoi kontaminacije radnih povrsina, alata, opreme i odece zaposlenih, kontrola individualnog izlaganja zracenju, evidencija i kontrola unutrasnjeg i spoljasnjeg manipulisanja kontaminiranim i aktiviranim predmetima, aktiviranim materijalima, radioaktivnim izvorima i radioaktivnim otpadnim materijalom. Svi podaci o merenjima karakteristika radijacionih polja registruju se u dnevnicima koje vodi dezurno osoblje i u periodicnim izvestajima. Uvidom u pomenute dnevnike i izvestaje, nisu uocena nikakva znacajna odstupanja u nivoima zracenja i merenim karakteristikama radijacionih polja za period februar-mart u odnosu na vrednosti iz prethodnih godina. U trenutku kada su otkrivene neregularnosti na gorivnim elementima iz poslednje sarze goriva u reaktoru, krajem marta ove godine, postavilo se pitanje da li doslo do isticanja fisionih produkata iz nekog elementa. Zahvaljujuci iskustvu stecenom prilikom prvog i do sada jedinog prskanja kosuljice gorivnog elementa, pre oko devet godina, moglo se sa visokim stepenom sigurnosti tvrditi da do toga nije doslo. Da je doslo do isticanja fisionih produkata znacajno bi se povecali nivoi zracenja na karakteristicnim tackama reaktorskih sistema, i u radnoj sredini povecao bi se i relativni sadrzaj fisionih produkata u rashadnom mediju, moguce bi bilo registrovati sveze fisione produkte kao sto je {sup 131} I, u vecim kolicinama. Medjutim, nista od toga nije registrovano. Moze se zakljuciti da nijedan gorini elemenat nije procurio, i da se sve standardne operacije manipulisanja

  14. Temperature and boron dependencies of buckling and radial reflector saving for VVER lattices

    International Nuclear Information System (INIS)

    Alvarez, C.

    1990-01-01

    The temperature and boron dependencies of buckling and radial reflectors savings are analyzed in this paper on the basis of the results from the calculations ZR-6M critical assembly. These dependencies are related to the physical behavior of temperature and boron reactivity coefficients for the cores of VVER-type critical facilities. As a byproduct, the parameter was also investigated and its dependence on water density was determined

  15. Influence of geometrical parameters of the VVER-1000 reactor construction elements to internals irradiation conditions

    Directory of Open Access Journals (Sweden)

    О. M. Pugach

    2015-07-01

    Full Text Available Investigations to determine the influences of geometrical parameters of the calculational VVER-1000 reactor model to the results of internal irradiation condition determination are carried out. It is shown that the values of appropriate sensitivity matrix elements are not dependent on a height coordinate for any core level, but there is their azimuthal dependence. Maximum possible relative biases of neutron fluence due to inexact knowledge of internal geometrical parameters are obtained for the baffle and the barrel.

  16. Experimental study of hydrodynamically induced vibrational processes in VVER-440 fuel assemblies

    International Nuclear Information System (INIS)

    Solonin, V.I.; Perevezentsev, V.V.; Rekshnya, N.F.; Krapivtsev, V.G.

    2000-01-01

    Investigations are described of hydrodynamically induced vibrations in a single fuel assembly of a VVER-440 reactor, performed on a full-scale model installed in a closed loop filled with distilled water; the model fuel elements contained simulators of fuel pellets. Data on hydrodynamic loads were obtained by measuring pressure oscillations along the height of the fuel assembly case. Results of the measurements are presented in graphs and are discussed in some detail. (A.K.)

  17. Influence of Bypass on Thermo-Hydraulics of VVER 440 Fuel Assembly

    Directory of Open Access Journals (Sweden)

    Jakubec Jakub

    2017-04-01

    Full Text Available The paper deals with CFD modelling and simulation of coolant flow within the nuclear reactor VVER 440 fuel assembly. The influence of coolant flow in bypass on the temperature distribution at the outlet of the fuel assembly and pressure drop was investigated. Only steady-state analyses were performed. Boundary conditions are based on operating conditions. ANSYS CFX is chosen as the main CFD software tool, where all analyses are performed.

  18. The RELAP5-Based NPA of the VVER Type Paks NPP

    International Nuclear Information System (INIS)

    Guba, A.; Toth, I.; Mandy, C.; Stubbe, E.

    1999-01-01

    NPA is a data driven interactive graphical tool for visualisation of different plant conditions. Data generated by the analysis code RELAP5/MOD3.2 are processed and displayed on a computer monitor. The NPA model of Paks NPP Unit 3 was developed with the aim to demonstrate the phenomena occurring in different transient/accident scenarios. This VVER-specific NPA development is a result of a cooperation between BELGATOM and KFKI-AEKI. (author)

  19. High energy pipe line break postulations and their mitigation - examples for VVER nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Zdarek, J.; Pecinka, L.; Kadecka, P.; Dotrel, J. [Nuclear Res. Inst., Rez (Czech Republic)

    1998-11-01

    The concept and the proposals for the protection and reinforcement of equipment against the effects of postulated rupture of the high-energy piping, in VVER Plant, are presented. The most recent version of the US NRC Guidelines has been used. The development of the legislation, the basic approach and selection of criteria for the assessment of the rupture of high energy piping, provide the basis for the application of the separation concept in the overall safety philosophy. (orig.)

  20. Model of nuclear reactor type VVER-1000/V-320 built by computer code ATHLET-CD

    International Nuclear Information System (INIS)

    Georgiev, Yoto; Filipov, Kalin; Velev, Vladimir

    2014-01-01

    A model of nuclear reactor type VVER-1000 V-320 developed for computer code ATHLET-CD2.1A is presented. Validation of the has been made, in the analysis of the station blackout scenario with LOCA on fourth cold leg is shown. As the calculation has been completed, the results are checked through comparison with the results from the computer codes ATHLET-2.1A, ASTEC-2.1 and RELAP5mod3.2

  1. Steam oxidation of Zr 1% Nb clads of VVER fuels in high temperature

    International Nuclear Information System (INIS)

    Solyanyj, V.I.; Bibilashvili, Yu.K.; Dranenko, V.V.; Levin, A.Ya.; Izrajlevskij, L.B.; Morozov, A.M.

    1984-01-01

    In a wide range of accident conditions processes of clad corrosion effected by steam are rather intensive and in many respects influence the safety of NPP and the after-accident dismantling of a reactor core. This paper discusses the results of comprehensive studies into corrosion behaviour of Zr 1%Nb clads of VVER-type fuels at high temperatures. These studies are a continuation of previous work and the base for the design modelling of corrosion processes

  2. Floor response spectra for seismic qualification of Kozloduy VVER 440-230 NPP

    Energy Technology Data Exchange (ETDEWEB)

    Kostov, M.K. [Bulgarian Academy of Sciences, Sofia (BG). Central Lab. for Seismic Mechanics and Earthquake Engineering; Ma, D.C. [Argonne National Lab., IL (United States); Prato, C.A. [Univ. of Cordoba (AR); Stevenson, J.D. [Stevenson and Associates, Cleveland, OH (US)

    1993-08-01

    In this paper the floor response spectra generation methodology for Kozloduy NPP, Unit 1-2 of VVER 440-230 is presented. The 2D coupled soil-structure interaction models are used combined with a simplified correction of the final results for accounting of torsional effects. Both time history and direct approach for in-structure spectra generation are used and discussion of results is made.

  3. Floor response spectra for seismic qualification of Kozloduy VVER 440-230 NPP

    International Nuclear Information System (INIS)

    Kostov, M.K.; Prato, C.A.; Stevenson, J.D.

    1993-01-01

    In this paper the floor response spectra generation methodology for Kozloduy NPP, Unit 1-2 of VVER 440-230 is presented. The 2D coupled soil-structure interaction models are used combined with a simplified correction of the final results for accounting of torsional effects. Both time history and direct approach for in-structure spectra generation are used and discussion of results is made

  4. Study of the flux effect nature for VVER-1000 RPV welds with high nickel content

    Energy Technology Data Exchange (ETDEWEB)

    Kuleshova, E.A., E-mail: evgenia-orm@yandex.ru [National Research Center “Kurchatov Institute”, Kurchatov Sq.1, 123182, Moscow (Russian Federation); National Research Nuclear University, “MEPhI” (Moscow Engineering Physics Institute), Kashirskoe Highway 31, 115409, Moscow (Russian Federation); Gurovich, B.A.; Lavrukhina, Z.V.; Maltsev, D.A.; Fedotova, S.V.; Frolov, A.S.; Zhuchkov, G.M. [National Research Center “Kurchatov Institute”, Kurchatov Sq.1, 123182, Moscow (Russian Federation)

    2017-01-15

    This work extends the research of the basic regularities of segregation processes in the grain boundaries (GB) of VVER-1000 reactor pressure vessel (RPV) steels. The paper considers the influence of irradiation with different fast neutron fluxes on the structure, yield strength and ductile-to-brittle transition temperature (T{sub K}) changes as well as on changes of the share of brittle intergranular fracture and development of segregation processes in the VVER-1000 RPV weld metal (WM). The obtained experimental results allow to separate the contribution of the hardening and non-hardening mechanisms to mechanical properties degradation of material irradiated at the operating temperature. It is shown that the difference in T{sub K} shift in WM irradiated to the same fluence with different fast neutron fluxes is mainly due to the difference in the GB accumulation kinetics of impurities and only to a small extent due to the material hardening. Phosphorus bulk diffusion coefficients were evaluated for the temperature exposure, accelerated irradiation and irradiation within surveillance specimens (SS) using a kinetic model of phosphorus GB accumulation in low-alloyed low-carbon steels under the influence of operational factors. The correlation between the GB segregation level of phosphorus and nickel, and the T{sub K} shift - in WM SS was obtained experimentally and indicates the non-hardening mechanism contribution to the total radiation embrittlement of VVER-1000 RPV steels throughout its extended lifetime. - Highlights: • Structural elements in high Ni welds are studied at different irradiation fluxes. • AES study demonstrated different P GB kinetics at different irradiation fluxes. • Hardening and non-hardening mechanism contributions to the flux effect are assessed. • Correlation between ΔT{sub K} and P and Ni GB content is shown for VVER-1000 RPV welds.

  5. 3D analysis of the reactivity insertion accident in VVER-1000

    Energy Technology Data Exchange (ETDEWEB)

    Abdullayev, A. M.; Zhukov, A. I.; Slyeptsov, S. M. [NSC Kharkov Inst. for Physics and Technology, 1, Akademicheskaya Str., Kharkov 61108 (Ukraine)

    2012-07-01

    Fuel parameters such as peak enthalpy and temperature during rod ejection accident are calculated. The calculations are performed by 3D neutron kinetics code NESTLE and 3D thermal-hydraulic code VIPRE-W. Both hot zero power and hot full power cases were studied for an equilibrium cycle with Westinghouse hex fuel in VVER-1000. It is shown that the use of 3D methodology can significantly increase safety margins for current criteria and met future criteria. (authors)

  6. High energy pipe line break postulations and their mitigation - examples for VVER nuclear power plants

    International Nuclear Information System (INIS)

    Zdarek, J.; Pecinka, L.; Kadecka, P.; Dotrel, J.

    1998-01-01

    The concept and the proposals for the protection and reinforcement of equipment against the effects of postulated rupture of the high-energy piping, in VVER Plant, are presented. The most recent version of the US NRC Guidelines has been used. The development of the legislation, the basic approach and selection of criteria for the assessment of the rupture of high energy piping, provide the basis for the application of the separation concept in the overall safety philosophy. (orig.)

  7. Sequence of decommissioning of the main equipment in a central type VVER 440 V-230

    International Nuclear Information System (INIS)

    Andres, E.; Garcia Ruiz, R.

    2014-01-01

    IBERDROLA Ingenieria y Construccion S.A.U., leader of consortium with Empresarios Agrupados and INDRA, has developed the Basic Engineering for the decommissioning of contaminated systems and building of a VVER 440 V-230 Nuclear Power Plant, establishing the sequence and methodology for the main equipment fragmentation. For that, it has been designed dry and wet cutting zones to be set up in the area where steam generators, main cooling pumps and pressurizer are located; these components will be dismantled previously. (Author)

  8. Regulatory use the classification security systems of I and C in VVER type reactors

    International Nuclear Information System (INIS)

    Ilizastegui Perez, F.

    1998-01-01

    Presently work the author proposes a classification to the system I and C to the VVER 440 type reactor in categories the regulatory control with a view to establishing the degree to the attention that the regulator should pay to these systems, leaving the importance that have the same ones for the security the installation, during the execution the works that are carried out with this equipment in the stages construction, setting in service and exploitation

  9. Temperature and boron dependencies of buckling and radial reflector savings for VVER lattices

    International Nuclear Information System (INIS)

    Alvarez, C.

    1990-01-01

    The temperature and boron dependencies of buckling and radial reflector savings are analyzed in this paper on the basis of the results from the calculations for the ZR-6M critical assembly. These dependencies are related to he physical behaviour of temperature and boron reactivity coefficients for the cores of VVER-type critical facilities. As a byproduct, the dp/dBg 2 parameter was also investigated and its dependence on water density was determined

  10. Simulation of a nuclear accident by an academic simulator of a VVER-1000 reactor

    International Nuclear Information System (INIS)

    Hernandez G, L.; Salazar S, E.

    2014-10-01

    This work is planned to simulate a scenario in which the same conditions that caused the accident at the Fukushima Daichi nuclear power plant are present, using a simulator of a nuclear power plant with VVER-1000 reactor, a different type of technology to the NPP where the accident occurred, which used BWR reactors. The software where it will take place the simulation was created and distributed by the IAEA for academic purposes, which contains the essential systems that characterize this type of NPP. The simulator has tools for the analysis of the characteristic phenomena of a VVER-1000 reactor in the different systems together and planned training tasks. This makes possible to identify the function of each component and how connects to other systems, thus facilitating the visualization of possible failures and the consequences that they have on the general behavior of the reactor. To program the conditions in the simulator, is necessary to know and synthesize a series of events occurred in Fukushima in 2011 and the realized maneuvers to reduce the effects of the system failures. Being different technologies interpretation of the changes that would suffer the VVER systems in the scenario in question will be developed. The Fukushima accident was characterized by the power loss of regular supply and emergency of the cooling systems which resulted in an increase in reactor temperature and subsequent fusion of their nuclei. Is interesting to reproduce this type of failure in a VVER, and extrapolate the lack of power supply in the systems that comprise, as well as pumping systems for cooling, has a pressure regulating system which involves more variables in the balance of the system. (Author)

  11. Proposal of a benchmark for core burnup calculations for a VVER-1000 reactor core

    International Nuclear Information System (INIS)

    Loetsch, T.; Khalimonchuk, V.; Kuchin, A.

    2009-01-01

    In the framework of a project supported by the German BMU the code DYN3D should be further validated and verified. During the work a lack of a benchmark on core burnup calculations for VVER-1000 reactors was noticed. Such a benchmark is useful for validating and verifying the whole package of codes and data libraries for reactor physics calculations including fuel assembly modelling, fuel assembly data preparation, few group data parametrisation and reactor core modelling. The benchmark proposed specifies the core loading patterns of burnup cycles for a VVER-1000 reactor core as well as a set of operational data such as load follow, boron concentration in the coolant, cycle length, measured reactivity coefficients and power density distributions. The reactor core characteristics chosen for comparison and the first results obtained during the work with the reactor physics code DYN3D are presented. This work presents the continuation of efforts of the projects mentioned to estimate the accuracy of calculated characteristics of VVER-1000 reactor cores. In addition, the codes used for reactor physics calculations of safety related reactor core characteristics should be validated and verified for the cases in which they are to be used. This is significant for safety related evaluations and assessments carried out in the framework of licensing and supervision procedures in the field of reactor physics. (authors)

  12. Thermal aging effects of VVER-1000 weld metal under operation temperature

    International Nuclear Information System (INIS)

    Chernobaeva, A.A.; Kuleshova, E.A.; Gurovich, B.A.; Erak, D.Y.; Zabusov, O.O.; Maltsev, D.A.; Zhurko, D.A.; Papina, V.B.; Skundin, M.A.

    2015-01-01

    The VVER-1000 thermal aging surveillance specimen sets are located in the reactor pressure vessel (RPV) under real operation conditions. Thermal aging surveillance specimens data are the most reliable source of the information about changing of VVER-1000 RPV materials properties because of long-term (hundred thousand hours) exposure at operation temperature. A revision of database of VVER-1000 weld metal thermal aging surveillance specimens has been done. The reassessment of transition temperature (T t ) for all tested groups of specimens has been performed. The duration of thermal exposure and phosphorus contents have been defined more precisely. The analysis of thermal aging effects has been done. The yield strength data, study of carbides evolution show absence of hardening effects due to thermal aging under 310-320 C degrees. Measurements of phosphorus content in grain boundaries segregation in different states have been performed. The correlation between intergranular fracture mode in Charpy specimens and transition temperature shift under thermal aging at temperature 310-320 C degrees has been revealed. All these data allow developing the model of thermal aging. (authors)

  13. CORONA ACADEMY, Opportunities for Enhancement of Training Capabilities in VVER Technology

    International Nuclear Information System (INIS)

    Ilieva, M.; Dieguez Porras, P.; Klepakova, A.

    2016-01-01

    Full text: The general objective of the project CORONA II is to enhance the safety of nuclear installations through further improvement of the training capabilities for providing the necessary personnel competencies in VVER area. More specific objective of the project is to continue the development of a state-of-the-art regional training network for VVER competence called CORONA Academia. The project aims at continuation of the European cooperation and support in this area for preservation and further development of expertise in the nuclear field by improvement of higher education and training. The consortium is focusing its effort on using the most advanced ways of providing training to the trainees, saving cost and time–distance learning and e-learning approaches which will be tested in CORONA II Project. The knowledge management portal will integrate the information on VVER web into a single communication system and develop and implement a semantic web structure to achieve mutual recognition of authentication information with other databases. That will enable the partners to share the materials available in each specific training center. (author

  14. Isothermal and thermal–mechanical fatigue of VVER-440 reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Fekete, Balazs, E-mail: fekete.mm.bme@gmail.com [College of Dunaujvaros, Tancsics 1A, Dunaujvaros H-2400 (Hungary); Department of Applied Mechanics, Budapest University of Technology and Economics, Muegyetem 5, Budapest H-1111 (Hungary); Trampus, Peter [College of Dunaujvaros, Tancsics 1A, Dunaujvaros H-2400 (Hungary)

    2015-09-15

    Highlights: • We aimed to determine the thermomechanical behaviour of VVER reactor steels. • Material tests were developed and performed on GLEEBLE 3800 physical simulator. • Coffin–Manson curves and parameters were derived. • High accuracy of the strain energy based evaluation was found. • The observed dislocation evolution correlates with the mechanical behaviour. - Abstract: The fatigue life of the structural materials 15Ch2MFA (CrMoV-alloyed ferritic steel) and 08Ch18N10T (CrNi-alloyed austenitic steel) of VVER-440 reactor pressure vessel under completely reserved total strain controlled low cycle fatigue tests were investigated. An advanced test facility was developed for GLEEBLE-3800 physical simulator which was able to perform thermomechanical fatigue experiments under in-service conditions of VVER nuclear reactors. The low cycle fatigue results were evaluated with the plastic strain based Coffin–Manson law, and plastic strain energy based model as well. It was shown that both methods are able to predict the fatigue life of reactor pressure vessel steels accurately. Interrupted fatigue tests were also carried out to investigate the kinetic of the fatigue evolution of the materials. On these samples microstructural evaluation by TEM was performed. The investigated low cycle fatigue behavior can provide reference for remaining life assessment and lifetime extension analysis.

  15. The estimation of the control rods absorber burn-up during the VVER-1000 operation

    Energy Technology Data Exchange (ETDEWEB)

    Bolshagin, Sergey N.; Gorodkov, Sergey S.; Sukhino-Khomenko, Evgeniya A. [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation)

    2013-09-15

    The isotopic composition of the control rods absorber changes under the neutron flux influence, so the control rods efficiency can decrease. In the VVER-1000 control rods boron carbide and dysprosium titanate are used as absorbing materials. In boric part the efficiency decreases due to the {sup 10}B isotope burn-up. Dysprosium isotopes turn into other absorbing isotopes, so the absorbing properties of dysprosium part decrease to a lesser degree. Also the control rod's shells may be deformed as a consequence of boron carbide radiation swelling. This fact should be considered in substantiation of control rods durability. For the estimation of the control rods absorber burn-up two models are developed: VVER-1000 3-D fuel assembly with control rods partially immersed (imitation of the control rods operation in the working group) and VVER-1000 3-D fuel assembly with control rods, located at the upper limit switch (imitation of the control rods operation in groups of the emergency shutdown system). (orig.)

  16. Multiple recycle of REMIX fuel at VVER-1000 operation in closed fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Alekseev, P. N.; Bobrov, E. A., E-mail: evgeniybobrov89@rambler.ru; Chibinyaev, A. V.; Teplov, P. S.; Dudnikov, A. A. [National Research Center Kurchatov Institute (Russian Federation)

    2015-12-15

    The basic features of loading the VVER-1000 core with a new variant of REMIX fuel (REgenerated MIXture of U–Pu oxides) are considered during its multiple recycle in a closed nuclear fuel cycle. The fuel composition is produced on the basis of the uranium–plutonium regenerate extracted at processing the spent nuclear fuel (SNF) from a VVER-1000, depleted uranium, and the fissionable material: {sup 235}U as a part of highly enriched uranium (HEU) from warheads superfluous for defense purposes or {sup 233}U accumulated in thorium blankets of fusion (electronuclear) neutron sources or fast reactors. Production of such a fuel assumes no use of natural uranium in addition. When converting a part of the VVER-1000 reactors to the closed fuel cycle based on the REMIX technology, the consumption of natural uranium decreases considerably, and there is no substantial degradation of the isotopic composition of plutonium or change in the reactor-safety characteristics at the passage from recycle to recycle.

  17. VVER operational experience - effect of preconditioning and primary water chemistry on radioactivity build-up

    International Nuclear Information System (INIS)

    Zmitko, M.; Kysela, J.; Dudjakova, K.; Martykan, M.; Janesik, J.; Hanus, V.; Marcinsky, P.

    2004-01-01

    The primary coolant technology approaches currently used in VVER units are reviewed and compared with those used in PWR units. Standard and modified water chemistries differing in boron-potassium control are discussed. Preparation of the VVER Primary Water Chemistry Guidelines in the Czech Republic is noted. Operational experience of some VVER units, operated in the Czech Republic and Slovakia, in the field of the primary water chemistry, and radioactivity transport and build-up are presented. In Mochovce and Temelin units, a surface preconditioning (passivation) procedure has been applied during hot functional tests. The main principles of the controlled primary water chemistry applied during the hot functional tests are reviewed and importance of the water chemistry, technological and other relevant parameters is stressed regarding to the quality of the passive layer formed on the primary system surfaces. The first operational experience obtained in the course of beginning of these units operation is presented mainly with respect to the corrosion products coolant and surface activities. Effect of the initial passivation performed during hot functional tests and the primary water chemistry on corrosion products radioactivity level and radiation situation is discussed. (author)

  18. Transient thermal-hydraulic/neutronic analysis in a VVER-1000 reactor core

    International Nuclear Information System (INIS)

    Seyed khalil Mousavian; Mohammad Mohsen Ertejaei; Majid Shahabfar

    2005-01-01

    Full text of publication follows: Nowadays, coupled thermal-hydraulic and three-dimensional neutronic codes in order to consider different feedback effects is state of the art subject in nuclear engineering researches. In this study, RELAP5/COBRA and WIMS/CITATION codes are implemented to investigate the VVER-1000 reactor core parameters during Large Break Loss of Coolant Accident (LB-LOCA). In a LB-LOCA, the primary side pressure, coolant density and fuel temperature strongly decrease but the cladding temperature experiences a strong peak. For this purpose, the RELAP5 Best Estimate (BE) system code is used to simulate the LB-LOCA analysis in VVER-1000 nuclear thermal-hydraulic loops. Also, the modified COBRA-IIIc software as a sub-channel analysis code is applied for modeling of VVER-1000 reactor core. Moreover, WIMS and CITATION as a cross section and 3-D neutron flux codes are coupled with thermal-hydraulic codes with the aim of consider the spatial effects through the reactor core. For this reason, suitable software is developed to link and speed up the coupled thermalhydraulic and three-dimensional neutronic calculations. This software utilizes of external coupling concept in order to integrate thermal-hydraulic and neutronic calculations. (authors)

  19. Microstructure and embrittlement of VVER 440 reactor pressure vessel steels; Microstructure et fragilisation des aciers de cuve des reacteurs nucleaires VVER 440

    Energy Technology Data Exchange (ETDEWEB)

    Hennion, A

    1999-03-15

    27 VVER 440 pressurised water reactors operate in former Soviet Union and in Eastern Europe. The pressure vessel, is made of Cr-Mo-V steel. It contains a circumferential arc weld in front of the nuclear core. This weld undergoes a high neutron flux and contains large amounts of copper and phosphorus, elements well known for their embrittlement potency under irradiation. The embrittlement kinetic of the steel is accelerated, reducing the lifetime of the reactor. In order to get informations on the microstructure and mechanical properties of these steels, base metals, HAZ, and weld metals have been characterized. The high amount of phosphorus in weld metals promotes the reverse temper embrittlement that occurs during post-weld heat treatment. The radiation damage structure has been identified by small angle neutron scattering, atomic probe, and transmission electron microscopy. Nanometer-sized clusters of solute atoms, rich in copper with almost the same characteristics as in western pressure vessels steels, and an evolution of the size distribution of vanadium carbides, which are present on dislocation structure, are observed. These defects disappear during post-irradiation tempering. As in western steels, the embrittlement is due to both hardening and reduction of interphase cohesion. The radiation damage specificity of VVER steels arises from their high amount of phosphorus and from their significant density of fine vanadium carbides. (author)

  20. Shaping of the axial power density distribution in the core to minimize the vapor volume fraction at the outlet of the VVER-1200 fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Savander, V. I.; Shumskiy, B. E., E-mail: borisshumskij@yandex.ru [National Research Nuclear University MEPhI (Russian Federation); Pinegin, A. A. [National Research Center Kurchatov Institute (Russian Federation)

    2016-12-15

    The possibility of decreasing the vapor fraction at the VVER-1200 fuel assembly outlet by shaping the axial power density field is considered. The power density field was shaped by axial redistribution of the concentration of the burnable gadolinium poison in the Gd-containing fuel rods. The mathematical modeling of the VVER-1200 core was performed using the NOSTRA computer code.

  1. Validation of coupled neutronic / thermal-hydraulic codes for VVER reactors. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Mittag, S.; Grundmann, U.; Kliem, S.; Kozmenkov, Y.; Rindelhardt, U.; Rohde, U.; Weiss, F.-P.; Langenbuch, S.; Krzykacz-Hausmann, B.; Schmidt, K.-D.; Vanttola, T.; Haemaelaeinen, A.; Kaloinen, E.; Kereszturi, A.; Hegyi, G.; Panka, I.; Hadek, J.; Strmensky, C.; Darilek, P.; Petkov, P.; Stefanova, S.; Kuchin, A.; Khalimonchuk, V.; Hlbocky, P.; Sico, D.; Danilin, S.; Ionov, V.; Nikonov, S.; Powney, D.

    2004-08-01

    In recent years, the simulation methods for the safety analysis of nuclear power plants have been continuously improved to perform realistic calculations. Therefore in VALCO work package 2 (WP 2), the usual application of coupled neutron-kinetic / thermal-hydraulic codes to VVER has been supplemented by systematic uncertainty and sensitivity analyses. A comprehensive uncertainty analysis has been carried out. The GRS uncertainty and sensitivity method based on the statistical code package SUSA was applied to the two transients studied earlier in SRR-1/95: A load drop of one turbo-generator in Loviisa-1 (VVER-440), and a switch-off of one feed water pump in Balakovo-4 (VVER-1000). The main steps of these analyses and the results obtained by applying different coupled code systems (SMABRE - HEXTRAN, ATHLET - DYN3D, ATHLET - KIKO3D, ATHLET - BIPR-8) are described in this report. The application of this method is only based on variations of input parameter values. No internal code adjustments are needed. An essential result of the analysis using the GRS SUSA methodology is the identification of the input parameters, such as the secondary-circuit pressure, the control-assembly position (as a function of time), and the control-assembly efficiency, that most sensitively affect safety-relevant output parameters, like reactor power, coolant heat-up, and primary pressure. Uncertainty bands for these output parameters have been derived. The variation of potentially uncertain input parameter values as a consequence of uncertain knowledge can activate system actions causing quite different transient evolutions. This gives indications about possible plant conditions that might be reached from the initiating event assuming only small disturbances. In this way, the uncertainty and sensitivity analysis reveals the spectrum of possible transient evolutions. Deviations of SRR-1/95 coupled code calculations from measurements also led to the objective to separate neutron kinetics from

  2. Experimental studies of resistance fretting-wear of fuel rods for VVER-1000 and TVS-KVADRAT fuel assemblies

    International Nuclear Information System (INIS)

    Makarov, V.; Afanasiev, A.; Egorov, Yu.; Matvienko, I.

    2015-01-01

    The paper covers the results of the studies performed to justify the wear resistance of fuel rods in contact with the spacer grids of TVS VVER-1000 fuel assembly and TVS-KVADRAT square fuel assembly of Russian design for PWR-900 reactor. The presented results of three testing stages comprise: Testing of mockup fuel rods of VVER TVS fuel assembly for fretting wear under the conditions of the water chemistry of VVER reactor; Testing models of different design embodiments of the fuel rods for VVER TVS fuel assembly for fretting wear in still cold water; Testing mockup fuel rods of TVS-KVADRAT square fuel assembly for PWR reactor for frettingwear under the conditions of PWR water chemistry. The effect of structural and operational factors was determined (amplitudes, fuel rod vibration frequencies, values of cladding-to-spacer grid cell gap for the depth of fuel rod cladding wear etc.), an assessment was made of the threshold values of fuel rod vibration parameters, which, if not exceeded, provide the absence of the fuel rod cladding fretting wear in the fuel rod-to spacer grid contact area. Key words: fretting wear, fuel rod, spacer grid, VVER, PWR (author)

  3. Validation cases of CATHARE 2 for VVER-1000 main steam line break analysis

    International Nuclear Information System (INIS)

    Kolev, Nikolay P.; Petrov, Nikolay; Donov, Jordan; Sabotinov, Luben; Nikonov, Sergey

    2008-01-01

    Recent coupled code benchmarks identified coolant mixing in the reactor vessel as an unresolved issue in the analysis of complex plan transients with reactivity insertion. Thus, Phase 2 of the OECD VVER-1000 Coolant Transient Benchmark (V1000CT-2) was defined. The benchmark includes calculation of vessel mixing tests and main steam line break (MSLB) analysis. The reference plant is Kozloduy-6 in Bulgaria. The general objective is the assessment of system codes for VVER safety analysis and specifically for their use in the analysis of reactivity transients. A specific objective is the testing of different scale mixing models (mixing matrix, multi-1D, coarse-3D and CFD), and analysis of MSLB transients with improved vessel thermal hydraulic models. The benchmark is sponsored by CEA-France and OECD and is jointly prepared by CEA and INRNE, in collaboration with the Kozloduy NPP, IRSN and PSU. This paper summarizes CATHARE2 code assessment calculations using multi-1D vessel thermal hydraulics with cross flow. Test cases are the OECD V1000CT-1 pump start-up benchmark and the V1000CT-2 benchmarks. Emphasis is put on vessel mixing aspects. Separate effects in the lower plenum as well as component and integral system tests are considered. The comparison shows that a six-sector vessel mixing model informed by plant data or validated CFD calculations in the initial state was able to correctly reproduce the channel average temperatures at the core inlet as well as the vessel outlet temperatures. Testing at system level including code-to-experiment and CATHARE-ATHLET comparison shows that the considered CATHARE VVER-1000 system model is capable of MSLB simulation. (author)

  4. Development of a VVER-1000 core loading pattern optimization program based on perturbation theory

    International Nuclear Information System (INIS)

    Hosseini, Mohammad; Vosoughi, Naser

    2012-01-01

    Highlights: ► We use perturbation theory to find an optimum fuel loading pattern in a VVER-1000. ► We provide a software for in-core fuel management optimization. ► We consider two objectives for our method (perturbation theory). ► We show that perturbation theory method is very fast and accurate for optimization. - Abstract: In-core nuclear fuel management is one of the most important concerns in the design of nuclear reactors. Two main goals in core fuel loading pattern design optimization are maximizing the core effective multiplication factor in order to extract the maximum energy, and keeping the local power peaking factor lower than a predetermined value to maintain the fuel integrity. Because of the numerous possible patterns of fuel assemblies in the reactor core, finding the best configuration is so important and challenging. Different techniques for optimization of fuel loading pattern in the reactor core have been introduced by now. In this study, a software is programmed in C language to find an order of the fuel loading pattern of a VVER-1000 reactor core using the perturbation theory. Our optimization method is based on minimizing the radial power peaking factor. The optimization process launches by considering an initial loading pattern and the specifications of the fuel assemblies which are given as the input of the software. The results on a typical VVER-1000 reactor reveal that the method could reach to a pattern with an allowed radial power peaking factor and increases the cycle length 1.1 days, as well.

  5. Extension and Verification of the Cross-Section Library for the VVER-1000 Surveillance Specimen Region

    International Nuclear Information System (INIS)

    Kirilova, D.; Belousov, S.; Ilieva, K.

    2011-01-01

    The objective of this work is a generation of new version of the BGL multigroup cross-section to extend the region of its applicability. The existing library version is problem oriented for VVER-1000 type of reactors and was generated by collapsing of the VITAMIN-B6 problem independent cross-section fine-group library applying the VVER-1000 reactor middle plane spectrum in cylindrical geometry. The new version BGLex additionally contains cross-sections averaged on the corresponding spectra of the surveillance specimen's (SS) region for VVER-1000 type of reactors. Comparative analysis of the neutron spectra for different one-dimensional geometry models that could be applied for the cross-section collapsing using the software package SCALE, showed a high sensitivity of the results to the geometry model. That is why a neutron importance assessment was done for the SS region using the adjoint solution calculated by the two-dimensional code DORT and problem-independent library VITAMIN-B6. The one-dimensional geometry model applied to the cross-section collapsing were determined by the material limits above the reactor core in axial direction z as for every material a homogenization in radial direction was done. The material homogenization in radial direction was done by material weighing taking into account the adjoint solution as well as the neutron source. The one-dimensional geometry model comprising the homogenized weighed materials was applied for the cross-section generation of the fine-group library VITAMIN-B6 to the broad-group structure of BGL library. The new version BGLex was extended with cross-sections for the SS region. Verification and validation of the new version BGLex is forthcoming. It includes comparison between the calculated results with the new version BGLex and the libraries BGL and VITAMIN-B6 and comparison with experimental results. (author)

  6. Extension and Verification of the Cross-Section Library for the VVER- 1000 Surveillance Specimen Region

    International Nuclear Information System (INIS)

    Kirilova, D.; Belousov, S.; Ilieva, K.

    2011-01-01

    The objective of this work is a generation of new version of the BGL multigroup cross-section to extend the region of its applicability. The existing library version is problem oriented for VVER-1000 type of reactors and was generated by collapsing of the VITAMIN-B6 problem independent cross-section fine-group library applying the VVER-1000 reactor middle plane spectrum in cylindrical geometry. The new version BGLex additionally contains cross-sections averaged on the corresponding spectra of the surveillance specimen's (SS) region for VVER-1000 type of reactors. Comparative analysis of the neutron spectra for different one-dimensional geometry models that could be applied for the cross-section collapsing using the software package SCALE, showed a high sensitivity of the results to the geometry model. That is why a neutron importance assessment was done for the SS region using the adjoint solution calculated by the two-dimensional code DORT and problem-independent library VITAMIN-B6. The one-dimensional geometry model applied to the cross-section collapsing were determined by the material limits above the reactor core in axial direction z as for every material a homogenization in radial direction was done. The material homogenization in radial direction was done by material weighing taking into account the adjoint solution as well as the neutron source. The one-dimensional geometry model comprising the homogenized weighed materials was applied for the cross-section generation of the fine-group library VITAMIN-B6 to the broad-group structure of BGL library. The new version BGLex was extended with cross-sections for the SS region. Verification and validation of the new version BGLex is forthcoming. It includes comparison between the calculated results with the new version BGLex and the libraries BGL and VITAMIN-B6 and comparison with experimental results. (author)

  7. CATHARE-2 prediction of large primary to secondary leakage (PRISE) at PSB-VVER experimental facility

    Energy Technology Data Exchange (ETDEWEB)

    Sabotinov, L.; Chevrier, P. [Institut de Radioprotection et de Surete Nucleaire, 92 - Fontenay aux Roses (France)

    2007-07-01

    The large primary to secondary leakage (PRISE) is a specific loss-of-coolant accident in VVER reactors, related to the break of the steam generator collector cover, leading to loss of primary mass inventory and possible direct radioactive release to atmosphere. The best estimate thermal-hydraulic computer code CATHARE-2 Version 2.5-1 was used for post-test analysis of a PRISE experiment, conducted at the large scale test facility PSB-VVER in Russia. The PSB rig is 1:300 scaled model of VVER-1000 NPP. The accident is calculated with a 1.4% break size, which corresponds to 100 mm leak from primary to secondary side in the real NPP. A computer model has been developed for CATHARE-2 V2.5-1, taking into account all important components of the PSB facility: reactor model (lower plenum, core, bypass, upper plenum, downcomer), 4 separate loops, pressurizer, horizontal multi-tube steam generators, break section. The secondary side is presented by recirculation model. A large number of sensitivity calculations has been performed regarding break modeling, reactor pressure vessel modeling, counter current flow modeling, hydraulic losses, heat losses, steam generator level regulation. Comparison between calculated and experimental results shows good prediction of the basic thermal-hydraulic phenomena and parameters such as primary and secondary pressures, temperatures, loop flows, etc. Some discrepancies were observed in the calculations of primary mass inventory and loop seal clearance. Nevertheless the final core heat up, which is one of the most important safety criteria, was correctly predicted. (authors)

  8. The strength of the reactor cavity of VVER-1000 NPP against steam explosion

    International Nuclear Information System (INIS)

    Varpasuo, P.

    1995-01-01

    The reactor cavity of VVER-1000 NPP is a thick-walled, cylindrical reinforced concrete structure. In case of molten core-water reaction during the severe accident the load carrying capacity of the cavity structure is of interest against the short impulse type loading caused by the steam explosion phenomenon. The assumed size of the impulse was 20 kPa-s and the duration was 10 ms. The static analysis of the structure used the ABAQUS/STANDARD and ANSYS codes. The material properties in both runs were specified to be elasto-plastic, and the cracking of concrete was taken into account. (author). 2 refs., 5 figs

  9. Application of the thermal-hydraulic codes in VVER-440 steam generators modelling

    Energy Technology Data Exchange (ETDEWEB)

    Matejovic, P.; Vranca, L.; Vaclav, E. [Nuclear Power Plant Research Inst. VUJE (Slovakia)

    1995-12-31

    Performances with the CATHARE2 V1.3U and RELAP5/MOD3.0 application to the VVER-440 SG modelling during normal conditions and during transient with secondary water lowering are described. Similar recirculation model was chosen for both codes. In the CATHARE calculation, no special measures were taken with the aim to optimize artificially flow rate distribution coefficients for the junction between SG riser and steam dome. Contrary to RELAP code, the CATHARE code is able to predict reasonable the secondary swell level in nominal conditions. Both codes are able to model properly natural phase separation on the SG water level. 6 refs.

  10. Design and implementation of the control system for nuclear plant VVER-1000. Instrumentation (program technical complexes)

    International Nuclear Information System (INIS)

    Siora, A.; Tokarev, V.; Bakhmach, E.

    2004-01-01

    Program-technical complexes (PTC) are designed as control and protection systems in water-moderated atomic reactors, including emergency and preventive systems, automatic control, unloading, reactor capacity limitation and accelerated preventive protection systems. Utilization of programmable logic integrated circuits from world leading manufacturers makes the complexes simple in structure, compact, with low energy demands and mutually independent for key and supporting functions The results of PTC assessment and implementation in Ukraine are outlined. Opportunities for a future development of RADIJ company in the area of control and protection systems for VVER reactors are also discussed

  11. Proposal of criteria for evaluation of engineering safety factors of VVER 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 (K eng ). The AER countries use different approaches to K eng 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 VVER 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)

  12. Application of the thermal-hydraulic codes in VVER-440 steam generators modelling

    Energy Technology Data Exchange (ETDEWEB)

    Matejovic, P; Vranca, L; Vaclav, E [Nuclear Power Plant Research Inst. VUJE (Slovakia)

    1996-12-31

    Performances with the CATHARE2 V1.3U and RELAP5/MOD3.0 application to the VVER-440 SG modelling during normal conditions and during transient with secondary water lowering are described. Similar recirculation model was chosen for both codes. In the CATHARE calculation, no special measures were taken with the aim to optimize artificially flow rate distribution coefficients for the junction between SG riser and steam dome. Contrary to RELAP code, the CATHARE code is able to predict reasonable the secondary swell level in nominal conditions. Both codes are able to model properly natural phase separation on the SG water level. 6 refs.

  13. Main results of post-irradiation examinations of new-generation fuel assemblies VVER-1000

    International Nuclear Information System (INIS)

    Zvir, E.; Markov, D.; Polenok, V.; Zhitelev, V.; Kobylyansky, G.

    2009-01-01

    To increase the competitiveness of Russian nuclear fuel at the foreign market and to improve its technical and economic performance in order to provide a necessary level of safety, it is necessary to solve certain important tasks: Increase of fuel burn-up; Extension of operational lifetime of fuel assemblies and operational reliability of nuclear fuel; Introduction of cost-beneficial and flexible fuel cycles. Alternative fuel assemblies TVSA VVER-1000 and TVS-2 are used as a basis to optimize the nuclear fuel and develop advanced fuel cycles for nuclear power plants with VVER-1000 reactor types. Four fuel assemblies TVSA operated during 1 and up to 6 reactor cycles, reference fuel assembly TVS-2 operated during three reactor cycles and achieved an average fuel burnup of 48MW·day/kgU as well as failed fuel assembly TVS-2 operated during one cycle were examined at RIAR in recent years. The main objectives of these examinations were to obtain experimental data in support of operational integrity of products or to find out reasons of their failure. The performed post-irradiation examinations confirmed the operational integrity of alternative fuel assemblies TVSA including their geometrical stability up to the average fuel burnup of 55 MW·day/kgU over the fuel assembly (FA) (up to the maximal fuel burnup of ∼73 MW·day/kgU in fuel rods) and of TVS-2 up to the average fuel burnup of 48 MW·day/kgU over the fuel assembly. The changes introduced in the design of VVER-1000 fuel assembly during the development of alternative fuel assembly TVSA and TVS-2 did not make any negative effect on fuel rods. It was proved that causes of fuel rod failure were not related to design features of fuel assemblies. The design features and operating conditions of fuel assemblies under examinations are briefly described. Post-irradiation examinations proved the geometrical stability of fuel assemblies TVSA and TVS-2 under operation up to the fuel burnup of ∼50 MW day/kgU, as for the

  14. System analysis of nuclear safety of VVER reactor with MOX fuel

    Energy Technology Data Exchange (ETDEWEB)

    Klimov, A.D.; Zharkov, V.P.; Suslov, I.R. [Russia, Moscow Malaya Krasnoselskaya St. (Russian Federation); Boyarinov, V.F.; Kevrolev, V.V.; Tchibinyaev, A.V.; Tsibulskiy, V.F. [RRC KI, Russia, Moscow (Russian Federation); Kochurov, B.P. [ITEP, Russia, Moscow (Russian Federation); Giovanni, B. [NFPSC, FRAMATOME (France)

    2005-07-01

    The report presents a short summary of the results achieved in the ISTC (International Science and Technology Center) project 'System analysis of nuclear safety of VVER reactor with MOX fuel' (April 2005). The studies within the project are of a systematic character and include the solutions of 15 tasks. The report gives an overview of the major blocks of these tasks: neutron transport equation solution; calculations of isotopic vectors, analysis of the impact of uncertainties on predicted reactor functionals. The calculation methods, the verification results and the corresponding codes are briefly described. (authors)

  15. Hydrodynamics around a spacer of a VVER-440 fuel rod bundle

    International Nuclear Information System (INIS)

    Mayer, G.; Hazi, G.; Kavran, P.

    2004-01-01

    Recently, an intensive research has been started in our institute, focusing on the hydrodynamics of fuel rod bundles. Numerical computations have been planed to be carried out in a three level bottom-up hierarchy, using direct numerical simulation, large eddy simulation and Reynolds averaged Navier-Stokes approach. Here, we give a description of the numerical method applied for direct numerical and large eddy simulation. We present some preliminary results obtained by the simulation of the flow around a spacer of a VVER-440 fuel rod bundle. (author)

  16. Effect of uncompensated SPN detector cables on neutron noise signals measured in VVER-440 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kiss, S. E-mail: kisss@sunserv.kfki.hu; Lipcsei, S. E-mail: lipcsei@sunserv.kfki.hu; Hazi, G. E-mail: gah@sunserv.kfki.hu

    2003-03-01

    The Self Powered Neutron Detector (SPND) noise measurements of an operating VVER-440 nuclear reactor are described and characterised. Signal characteristics may be radically influenced by the geometrical properties of the detector and the cable, and by the measuring arrangement. Simulator is used as a means of studying the structure of those phase spectra that show propagating perturbations measured on uncompensated SPN detectors. The paper presents measurements with detectors of very different sizes (i.e. 20 cm length SPNDs and the 200 cm length compensation cables), where the ratios of the global and local component differ significantly for the different detector sizes. This phenomenon is used up for signal compensation.

  17. Calculation with MCNP of capture photon flux in VVER-1000 experimental reactor.

    Science.gov (United States)

    Töre, Candan; Ortego, Pedro

    2005-01-01

    The aim of this study is to obtain by Monte Carlo method the high energy photon flux due to neutron capture in the internals and vessel layers of the experimental reactor LR-0 located in REZ, Czech Republic, and loaded with VVER-1000 fuel. The calclated neutron, photon and photon to neutron flux ratio are compared with experimental measurements performed with a multi-parameter stilbene detector. The results show clear underestimation of photon flux in downcomer and some overestimation at vessel surface and 1/4 thickness but a good fitting for deeper points in vessel.

  18. Standard and hydrazine water chemistry in primary circuit of VVER 440 units

    International Nuclear Information System (INIS)

    Burclova, J.

    1992-01-01

    Standard ammonia-potassium-boron water chemistry of 8 units with VVER 440 in CSFR is discussed as well as the corrosion product activity in the coolant during steady state and shut-down period and surface activity, dose rate build-up and occupational radiation exposure. Available data on hydrazine application (USSR, Hungary) indicate the possibility of the radiation field decreasing. Nevertheless the detailed analysis of 55 cycles of operation under standard water chemistry in Czechoslovakia allows to expect the comparable results for both water chemistries. (author)

  19. Ukrainian Nuclear Society International Conference 'Modernization of the NPP with VVER reactor' (abstracts)

    International Nuclear Information System (INIS)

    Bar'yakhtar, V.G.

    1999-01-01

    Abstracts of the papers presented at International conference of the Ukrainian Nuclear Society 'Modernization of the NPP with VVER reactor'. The following problems are considered: improving the NPP's safety and reliability; reactor modernization, the lifetime prolongation; increasing of the reactor operating characteristics; methods of capacity factor increasing: refueling control, maintenance control; technical and economical aspects of NPP modernization; modernization of the automated control system of the fuel process at the NPP's; technical features and methods for the continued radiation and technology control at the NPP's; training, increasing the staff qualification and NPP modernization

  20. Validation matrix for the assessment of thermal-hydraulic codes for VVER LOCA and transients. A report by the OECD support group on the VVER thermal-hydraulic code validation matrix

    International Nuclear Information System (INIS)

    2001-06-01

    This report deals with an internationally agreed experimental test facility matrix for the validation of best estimate thermal-hydraulic computer codes applied for the analysis of VVER reactor primary systems in accident and transient conditions. Firstly, the main physical phenomena that occur during the considered accidents are identified, test types are specified, and test facilities that supplement the CSNI CCVMs and are suitable for reproducing these aspects are selected. Secondly, a list of selected experiments carried out in these facilities has been set down. The criteria to achieve the objectives are outlined. The construction of VVER Thermal-Hydraulic Code Validation Matrix follows the logic of the CSNI Code Validation Matrices (CCVM). Similar to the CCVM it is an attempt to collect together in a systematic way the best sets of available test data for VVER specific code validation, assessment and improvement, including quantitative assessment of uncertainties in the modelling of phenomena by the codes. In addition to this objective, it is an attempt to record information which has been generated in countries operating VVER reactors over the last 20 years so that it is more accessible to present and future workers in that field than would otherwise be the case. (authors)

  1. Department of Energy's team's analyses of Soviet designed VVERs (water-cooled water-moderated atomic energy reactors)

    Energy Technology Data Exchange (ETDEWEB)

    1989-09-01

    This document contains apprendices A through P of this report. Topics discussed are: a cronyms and technical terms, accident analyses reactivity control; Soviet safety regulations; radionuclide inventory; decay heat; operations and maintenance; steam supply system; concrete and concrete structures; seismicity; site information; neutronic parameters; loss of electric power; diesel generator reliability; Soviet codes and standards; and comparisons of PWR and VVER features. (FI)

  2. A Roadmap and Discussion of Issues for Physics Analyses Required to Support Plutonium Disposition in VVER-1000 Reactors

    International Nuclear Information System (INIS)

    Primm, R.T.; Drischler, J.D.; Pavlovichev, A.M.; Styrine, Y.A.

    2000-01-01

    The purpose of this report is to document the physics analyses that must be performed to successfully disposition weapons-usable plutonium in VVER-1000 reactors in the Russian Federation. The report is a document to support programmatic and financial planning. It does not include documentation of the technical procedures by which physics analyses are performed, nor are the results of any analyses included

  3. State of the VVER-1000 spent U-Gd fuel rods based on the results of post-irradiation examinations

    International Nuclear Information System (INIS)

    Shevlyakov, G.; Zvir, E.; Strozhuk, A.; Polenok, V.; Sidorenko, O.; Volkova, I.; Nikitin, O.

    2015-01-01

    The present paper is devoted to post-irradiation examinations (PIE) of U-Gd fuel rods with different geometry of the fuel pellets irradiated as part of the VVER-1000 fuel assembly. As evidenced by their PIE data, they did not exhaust their service life based on the main parameters (geometrical dimensions, corrosion state, and release of fission product gases). (author)

  4. Shut-down margin study for the next generation VVER-1000 reactor including 13 x 13 hexagonal annular assemblies

    International Nuclear Information System (INIS)

    Faghihi, Farshad; Mirvakili, S. Mohammad

    2011-01-01

    Highlights: → Shut-Down Margin (SDM) for the next generation annular fuel core of typical VVER-1000, 13 x 13 assemblies are calculated. → The MCNP-5 code is run for many cases with different core burn up at various core temperatures. → There is a substantial drop in SDM in the case of annular fuel for the same power level. → SDM for our proposed VVER-1000 annular pins is calculated for specific average fuel burn up values at the BOC, MOC, and EOC. - Abstract: Shut-Down Margin (SDM) for the next generation annular fuel core of typical VVER-1000, 13 x 13 assemblies are calculated as the main aim of the present research. We have applied the MCNP-5 code for many cases with different values of core burn up at various core temperatures, and therefore their corresponding coolant densities and boric acid concentrations. There is a substantial drop in SDM in the case of annular fuel for the same power level. Specifically, SDM for our proposed VVER-1000 annular pins is calculated when the average fuel burn up values at the BOC, MOC, and EOC are 0.531, 11.5, and 43 MW-days/kg-U, respectively.

  5. Analysis of the VVER-440 reactor steam generator secondary side with the RELAP5/MOD3 code

    International Nuclear Information System (INIS)

    Tuunanen, J.

    1993-01-01

    Nuclear Engineering Laboratory of the Technical Research Centre of Finland has widely used RELAP5/MOD2 and -MOD3 codes to simulate horizontal steam generators. Several models have been developed and successfully used in the VVER-safety analysis. Nevertheless, the models developed have included only rather few nodes in the steam generator secondary side. The secondary side has normally been divided into about 10 to 15 nodes. Since the secondary side at the steam generators of VVER-440 type reactors consists of a rather large water pool, these models were only roughly capable to predict secondary side flows. The paper describes an attempt to use RELAP5/MOD3 code to predict secondary side flows in a steam generator of a VVER-440 reactor. A 2D/3D model has been developed using RELAP5/MOD3 codes cross-flow junctions. The model includes 90 volumes on the steam generator secondary side. The model has been used to calculate steady state flow conditions in the secondary side of a VVER-440 reactor steam generator. (orig.) (1 ref., 9 figs., 2 tabs.)

  6. OECD/DOE/CEA VVER-1000 Coolant Transient Benchmark. Summary Record of the First Workshop (V1000-CT1)

    International Nuclear Information System (INIS)

    2003-01-01

    The first workshop for the VVER-1000 Coolant Transient Benchmark TT Benchmark was hosted by the Commissariat a l'Energie Atomique, Centre d'Etudes de Saclay, France. The V1000CT benchmark defines standard problems for validation of coupled three-dimensional (3-D) neutron-kinetics/system thermal-hydraulics codes for application to Soviet-designed VVER-1000 reactors using actual plant data without any scaling. The overall objective is to access computer codes used in the safety analysis of VVER power plants, specifically for their use in reactivity transient simulations in a VVER-1000. The V1000CT benchmark consists of two phases: V1000CT-1 - simulation of the switching on of one main coolant pump (MCP) while the other three MCP are in operation, and V1000CT- 2 - calculation of coolant mixing tests and Main Steam Line Break (MSLB) scenario. Further background information on this benchmark can be found at the OECD/NEA benchmark web site . The purpose of the first workshop was to review the benchmark activities after the Starter Meeting held last year in Dresden, Germany: to discuss the participants' feedback and modifications introduced in the Benchmark Specifications on Phase 1; to present and to discuss modelling issues and preliminary results from the three exercises of Phase 1; to discuss the modelling issues of Exercise 1 of Phase 2; and to define work plan and schedule in order to complete the two phases

  7. Reflooding and boil-off experiments in a VVER-440 like rod bundle and analyses with the CATHARE code

    International Nuclear Information System (INIS)

    Korteniemi, V.; Haapalehto, T.; Puustinen, M.

    1995-01-01

    Several experiments were performed with the VEERA facility to simulate reflooding and boil-off phenomena in a VVER-440 like rod bundle. The objective of these experiments was to get experience of a full-scale bundle behavior and to create a database for verification of VVER type core models used with modern thermal-hydraulic codes. The VEERA facility used in the experiments is a scaled-down model of the Russian VVER-440 type pressurized water reactors used in Loviisa, Finland. The test section of the facility consists of one full-scale copy of a VVER-440 reactor rod bundle with 126 full-length electrically heated rod simulators. Bottom and top-down reflooding, different modes of emergency core cooling (ECC) injection and the effect of heating power on the heat-up of the rods was studied. In this paper the results of calculations simulating two reflood and one boil-off experiment with the French CATHARE2 thermal-hydraulic code are also presented. Especially the performance of the recently implemented top-down reflood model of the code was studied

  8. Results of VVER fuel rods tests in the MIR.M1 reactor under power cycling conditions

    International Nuclear Information System (INIS)

    Burukin, A.; Izhutov, A.; Ovchinnikov, V.; Kalygin, V.; Markov, D.; Pimenov, Y.; Novikov, V.; Medvedev, A.; Nesterov, B.

    2011-01-01

    The paper presents the main results of the 50 ... 60 MWd/kgU burnup VVER fuel rods tests performed in the MIR.M1 reactor loop facilities under power cycling. The non-destructive PIE results are presented as well. A series of experiments was performed, including overall measurement of fuel rod parameters test, in one of which 300 cycles were done. Irradiation under power cycling conditions and PIE of high-burnup VVER fuel rods showed the following: 1) all fuel rods claddings preserved their integrity under irradiation at linear heat rate (LHR) higher than the NPP operating one; 2) experimental data were obtained on the axial and radial cladding strain and fission gas release (FGR) from 50 ... 60 MWd/kgU burnup VVER-440 and VVER-1000 fuel rods as well as on the kinetics of the change in these parameters and fuel temperature under the power cycling; 3) non-destructive PIE results are in a satisfactory correlation with the data obtained by means of in-pile measurement gages during irradiation. (authors)

  9. Multi codes and multi-scale analysis for void fraction prediction in hot channel for VVER-1000/V392

    International Nuclear Information System (INIS)

    Hoang Minh Giang; Hoang Tan Hung; Nguyen Huu Tiep

    2015-01-01

    Recently, an approach of multi codes and multi-scale analysis is widely applied to study core thermal hydraulic behavior such as void fraction prediction. Better results are achieved by using multi codes or coupling codes such as PARCS and RELAP5. The advantage of multi-scale analysis is zooming of the interested part in the simulated domain for detail investigation. Therefore, in this study, the multi codes between MCNP5, RELAP5, CTF and also the multi-scale analysis based RELAP5 and CTF are applied to investigate void fraction in hot channel of VVER-1000/V392 reactor. Since VVER-1000/V392 reactor is a typical advanced reactor that can be considered as the base to develop later VVER-1200 reactor, then understanding core behavior in transient conditions is necessary in order to investigate VVER technology. It is shown that the item of near wall boiling, Γ w in RELAP5 proposed by Lahey mechanistic method may not give enough accuracy of void fraction prediction as smaller scale code as CTF. (author)

  10. A Roadmap and Discussion of Issues for Physics Analyses Required to Support Plutonium Disposition in VVER-1000 Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Primm, R.T.; Drischler, J.D.; Pavlovichev, A.M. Styrine, Y.A.

    2000-06-01

    The purpose of this report is to document the physics analyses that must be performed to successfully disposition weapons-usable plutonium in VVER-1000 reactors in the Russian Federation. The report is a document to support programmatic and financial planning. It does not include documentation of the technical procedures by which physics analyses are performed, nor are the results of any analyses included.

  11. Reflooding and boil-off experiments in a VVER-440 like rod bundle and analyses with the CATHARE code

    Energy Technology Data Exchange (ETDEWEB)

    Korteniemi, V.; Haapalehto, T. [Lappeenranta Univ. of Technology (Finland); Puustinen, M. [VTT Energy, Lappeenranta (Finland)

    1995-09-01

    Several experiments were performed with the VEERA facility to simulate reflooding and boil-off phenomena in a VVER-440 like rod bundle. The objective of these experiments was to get experience of a full-scale bundle behavior and to create a database for verification of VVER type core models used with modern thermal-hydraulic codes. The VEERA facility used in the experiments is a scaled-down model of the Russian VVER-440 type pressurized water reactors used in Loviisa, Finland. The test section of the facility consists of one full-scale copy of a VVER-440 reactor rod bundle with 126 full-length electrically heated rod simulators. Bottom and top-down reflooding, different modes of emergency core cooling (ECC) injection and the effect of heating power on the heat-up of the rods was studied. In this paper the results of calculations simulating two reflood and one boil-off experiment with the French CATHARE2 thermal-hydraulic code are also presented. Especially the performance of the recently implemented top-down reflood model of the code was studied.

  12. Experimental Testing of Innovative Cold-Formed "GEB" Section / Badania Eksperymentalne Innowacyjnego Kształtownika Giętego Na Zimno Typu "Geb"

    Science.gov (United States)

    Łukowicz, Agnieszka; Urbańska-Galewska, Elżbieta; Gordziej-Zagórowska, Małgorzata

    2015-03-01

    One of the major advantages of light gauge steel structures made of cold-formed steel sections is their low weight so the production of typical single-storey steel structures of this kind of profiles is still rising. The well known profiles, e.o. Z-sections, C-sections and the so called hat-sections studied and described in the literature, are used mainly as purlins or truss components. A new profile GEB was patented for the use for primary load-bearing member in fabricated steel frames. According to the code [1] every novel cross section should be tested to assign the deformation shape and bearing capacity. The paper deals with the numerical and experimental research of bearing capacity of cold formed GEB profiles. The deformation shape and limit load was obtained from bending tests. The GEB cross section bearing capacity was also determined according to codes [1, 2]. Jedną z najważniejszych zalet lekkich konstrukcji metalowych, wytwarzanych z kształtowników giętych na zimno, jest ich mała masa, dlatego też, producenci coraz częściej wykorzystują możliwości profili giętych do wytwarzania typowych konstrukcji halowych w budownictwie systemowym. Proces gięcia na zimno, pozwala na formowanie różnego rodzaju przekrojów poprzecznych, które mogą być wykorzystywane jako elementy konstrukcji. Typowe kształty elementów. tzn. Z, C oraz tzw. przekroje kapeluszowe, które zostały przebadane i opisane w literaturze, wykorzystuje się głównie jako płatwie lub części składowe wiązarów kratowych. Nowo opatentowany przekrój typu GEB ma być wykorzystany jako element nośny konstrukcji ramowych. W związku z tym innowacyjny kształt oraz parametry geometryczne przekroju takiego kształtownika, związane z możliwością jego wyprodukowania oraz z warunkami nośności, stateczności oraz sztywności, muszą być optymalne. Według normy PN-EN 1993-1-3, każdy nowo uformowany przekrój powinien być przebadany pod kątem nośności elementu i formy

  13. Ocena wpływu relaksacji poizometrycznej na napięcia mięśniowe u chorych z zawrotami głowy typu szyjnego

    Directory of Open Access Journals (Sweden)

    Hanna Zielińska-Bliźniewska

    2012-11-01

    Full Text Available Wprowadzenie: Celem pracy była ocena wpływu relaksacji poizometrycznej na napięcia mięśniowe u chorych z zawrotami głowy typu szyjnego. Materiał i metody: Badania przeprowadzono na grupie losowo wybranych 100 chorych, w tym 65 kobiet w wieku 20–76 lat i 35 mężczyzn w wieku 20–73 lat, leczonych w Klinice Otolaryngologii i Onkologii Laryngologicznej Uniwersyteckiego Szpitala Klinicznego im. WAM w Łodzi. Pacjentów podzielono na 2 grupy: I – badaną, liczącą 50 chorych z zawrotami głowy pochodzenia szyjnego, II – porównawczą, składającą się z 50 zdrowych osób, bez zawrotów głowy. U wszystkich chorych przeprowadzono szczegółowy wywiad, badanie przedmiotowe otolaryngologiczne, otoneurologiczne, fizykalne oraz rutynowe badania laboratoryjne. Każdy chory był konsultowany neurologicznie, okulistycznie i internistycznie oraz miał wykonywane USG naczyń doczaszkowych, tomografię komputerową odcinka szyjnego kręgosłupa i głowy. U wszystkich pacjentów zastosowano indywidualnie dobrany cykl ćwiczeń, uwzględniający dotychczasowy przebieg choroby oraz ewentualne przeciwwskazania, obejmujący relaksację poizometryczną mięśni okołokręgosłupowych w odcinku szyjnym przez okres 2 miesięcy. Obiektywna ocena skuteczności zastosowanej terapii odbywa- ła się (przed rozpoczęciem terapii oraz po 2 tygodniach, po miesiącu i po 2 miesiącach za pomocą liniowych pomiarów czynnego zakresu ruchomości szyjnego odcinka kręgosłupa oraz siły mięśniowej według testu Lovetta w skali punktowej i oceny zawrotów głowy według kryteriów Silvoniemiego. Wyniki: Na podstawie przeprowadzonych badań stwierdzono, że pod wpływem kompleksowych ćwiczeń w grupie badanej nastąpiła znaczna poprawa ruchomości odcinka szyjnego kręgosłupa oraz siły mięśniowej. Wnioski: Zarówno w badaniach obiektywnych (pomiar ruchomości czynnej szyjnego odcinka kręgosłupa oraz siły mięśniowej, jak i subiektywnych (ocena wg

  14. Analysis of transients for NPP with VVER-440 using the code SiTAP

    International Nuclear Information System (INIS)

    Kalinenko, V.

    1994-06-01

    The report contains analysis of transients ''Loop connection'' and ''Steam generator tube rupture'' for nuclear power plants (NPP) with VVER-440. To obtain more detailed information about NPP's dynamic characteristics, various variants of initial and boundary conditions are considerd. Calculation of these transients was performed using the SiTAP code developed at the Nuclear Safety Institute of the Russian Research Centre ''Kurchatov Institute''. SiTAP code is a multifunctional computer tool for fast analysis of transient and accidental processes of VVER type reactors for engineers working in the field of NPP dynamics. SiTAP can be used form comparative analysis of several variants of accident scenarios to find out the conditions leading to most serious consequences from a safety point of view. In such cases, additional analyses using best-estimate codes should be carried out. The results of SiTAP for a faulty loop connection leading to a boron dilution accident are intended to be used as boundary conditions for a more detailed anlaysis with the aid of the three-dimensional reactor core model DYN3D, developed in the Research Centre Rossendorf for the simulation of reactivity initiated accidents. (orig.)

  15. CFD Analysis of a Slug Mixing Experiment Conducted on a VVER-1000 Model

    Directory of Open Access Journals (Sweden)

    F. Moretti

    2009-01-01

    Full Text Available A commercial CFD code was applied, for validation purposes, to the simulation of a slug mixing experiment carried out at OKB “Gidropress” scaled facility in the framework of EC TACIS project R2.02/02: “Development of safety analysis capabilities for VVER-1000 transients involving spatial variations of coolant properties (temperature or boron concentration at core inlet.” Such experimental model reproduces a VVER-1000 nuclear reactor and is aimed at investigating the in-vessel mixing phenomena. The addressed experiment involves the start-up of one of the four reactor coolant pumps (the other three remaining idle, and the presence of a tracer slug on the starting loop, which is thus transported to the reactor pressure vessel where it mixes with the clear water. Such conditions may occur in a boron dilution scenario, hence the relevance of the addressed phenomena for nuclear reactor safety. Both a pretest and a posttest CFD simulations of the mentioned experiment were performed, which differ in the definition of the boundary conditions (based either on nominal quantities or on measured quantities, resp.. The numerical results are qualitatively and quantitatively analyzed and compared against the measured data in terms of space and time tracer distribution at the core inlet. The improvement of the results due to the optimization of the boundary conditions is evidenced, and a quantification of the simulation accuracy is proposed.

  16. The in-core fuel management code system for VVER reactors

    International Nuclear Information System (INIS)

    Cada, R.; Krysl, V.; Mikolas, P.; Sustek, J.; Svarny, J.

    2004-01-01

    The structure and methodology of a fuel management system for NPP VVER 1000 (NPP Temelin) and VVER 440 (NPP Dukovany) is described. It is under development in SKODA JS a.s. and is followed by practical applications. The general objectives of the system are maximization of end of cycle reactivity, the minimization of fresh fuel inventory for the minimization of fed enrichment and minimization of burnable poisons (BPs) inventory. They are also safety related constraints in witch minimization of power peaking plays a dominant role. General structure of the system consists in preparation of input data for macrocode calculation, algorithms (codes) for optimization of fuel loading, calculation of fuel enrichment and BPs assignment. At present core loading can be calculated (optimized) by Tabu search algorithm (code ATHENA), genetic algorithm (code Gen1) and hybrid algorithm - simplex procedure with application of Tabu search algorithm on binary shuffling (code OPAL B ). Enrichment search is realized by the application of simplex algorithm (OPAL B code) and BPs assignment by module BPASS and simplex algorithm in OPAL B code. Calculations of the real core loadings are presented and a comparison of different optimization methods is provided. (author)

  17. Isothermal and thermal-mechanical fatigue of VVER-440 reactor pressure vessel steels

    Science.gov (United States)

    Fekete, Balazs; Trampus, Peter

    2015-09-01

    The fatigue life of the structural materials 15Ch2MFA (CrMoV-alloyed ferritic steel) and 08Ch18N10T (CrNi-alloyed austenitic steel) of VVER-440 reactor pressure vessel under completely reserved total strain controlled low cycle fatigue tests were investigated. An advanced test facility was developed for GLEEBLE-3800 physical simulator which was able to perform thermomechanical fatigue experiments under in-service conditions of VVER nuclear reactors. The low cycle fatigue results were evaluated with the plastic strain based Coffin-Manson law, and plastic strain energy based model as well. It was shown that both methods are able to predict the fatigue life of reactor pressure vessel steels accurately. Interrupted fatigue tests were also carried out to investigate the kinetic of the fatigue evolution of the materials. On these samples microstructural evaluation by TEM was performed. The investigated low cycle fatigue behavior can provide reference for remaining life assessment and lifetime extension analysis.

  18. Thermal-hydraulic studies on the safety of VVER-440 type nuclear power plants

    International Nuclear Information System (INIS)

    Tuunanen, J.

    1994-01-01

    The thesis includes several thermal-hydraulic analyses related to the Loviisa VVER-440 nuclear power plant units. The work consists of experimental studies, analysis of the experiments, analysis of some plant transients and development of a calculational model for calculation of boric concentrations in the reactor. In the first part of thesis, in the case of simulation of boric acid solution behaviour during long-term cooling period of LOCAs, experiments were performed in scaled-down test facilities. The experimental data together with the results of RELAP5/MOD3 simulations were used to develop a model for calculations of boric acid concentrations in the reactor during LOCAs. In the second part, in the case of simulation of horizontal generators, experiments were performed with PACTEL integral test loop to simulate loss of feedwater transients. The PACTEL experiments as well as earlier REWETT-III natural circulation tests, were analyzed with RELAP5/MOD3 Version 5m5 code. The third part of the work consists of simulations of Loviisa VVER reactor pump trip transients with RELAP5/MOD1-Eur, RELAP5/MOD3 and CATHARE codes. (56 refs., 9 figs.)

  19. Study of Fast Transient Pressure Drop in VVER-1000 Nuclear Reactor Using Acoustic Phenomenon

    Directory of Open Access Journals (Sweden)

    Soroush Heidari Sangestani

    2018-01-01

    Full Text Available This article aims to simulate the sudden and fast pressure drop of VVER-1000 reactor core coolant, regarding acoustic phenomenon. It is used to acquire a more accurate method in order to simulate the various accidents of reactor core. Neutronic equations should be solved concurrently by means of DRAGON 4 and DONJON 4 coupling codes. The results of the developed package are compared with WIMS/CITATION and final safety analysis report of Bushehr VVER-1000 reactor (FSAR. Afterwards, time dependent thermal-hydraulic equations are answered by employing Single Heated Channel by Sectionalized Compressible Fluid method. Then, the obtained results were validated by the same transient simulation in a pressurized water reactor core. Then, thermal-hydraulic and neutronic modules are coupled concurrently by use of producing group constants regarding the thermal feedback effect. Results were compared to the mentioned transient simulation in RELAP5 computer code, which show that mass flux drop is sensed at the end of channel in several milliseconds which causes heat flux drop too. The thermal feedback resulted in production of some perturbations in the changes of these parameters. The achieved results for this very fast pressure drop represent accurate calculations of thermoneutronic parameters fast changes.

  20. Optimizing a gap conductance model applicable to VVER-1000 thermal–hydraulic model

    International Nuclear Information System (INIS)

    Rahgoshay, M.; Hashemi-Tilehnoee, M.

    2012-01-01

    Highlights: ► Two known conductance models for application in VVER-1000 thermal–hydraulic code are examined. ► An optimized gap conductance model is developed which can predict the gap conductance in good agreement with FSAR data. ► The licensed thermal–hydraulic code is coupled with the gap conductance model predictor externally. -- Abstract: The modeling of gap conductance for application in VVER-1000 thermal–hydraulic codes is addressed. Two known models, namely CALZA-BINI and RELAP5 gap conductance models, are examined. By externally linking of gap conductance models and COBRA-EN thermal hydraulic code, the acceptable range of each model is specified. The result of each gap conductance model versus linear heat rate has been compared with FSAR data. A linear heat rate of about 9 kW/m is the boundary for optimization process. Since each gap conductance model has its advantages and limitation, the optimized gap conductance model can predict the gap conductance better than each of the two other models individually.

  1. Experimental study of asymmetric boron dilution at VVER-1000 of Kudankulam NPP and its simulation

    Energy Technology Data Exchange (ETDEWEB)

    Tsyganov, Sergey V.; Kotsarev, Alexander V.; Baykov, Alexander V. [National Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

    2017-09-15

    The Kudankulam NPP units contain additional and unique for VVER Quick Boron Injection System (QBIS) for beyond-design-basis accident management without scram. During the physical start-up stage at hot zero power of both Kudankulam units, special tests were performed to assess the efficiency of the system. In the course of test three out of four QBIS tanks had been promptly opened and it led to the asymmetrical injection of boric acid into the core. The scenario of the tests may address to the inhomogeneous boron dilution process that is now an essential part of safety analysis of pressurised water reactors. The simulation of the process, including ex-core ion chambers readings, has been accomplished using ATHLET/BIPR-VVER code. Behaviour of some reactor parameters in the course of the test and some results of the simulation are discussing in the paper. Authors believe the process of the asymmetrical injection of boric acid may be useful for verification and validation of coupled neutronic and thermo-hydraulic codes widely used for safety analysis, including analysis of boron dilution accident.

  2. Verification of the enrichment of fresh VVER-440 fuel assemblies at NPP Paks

    Energy Technology Data Exchange (ETDEWEB)

    Almasia, I.; Hlavathya, Z.; Nguyena, C. T. [Institute of Isotopes, Hungarian Academy of Sciences, Budapest, (Hungary); others, and

    2012-06-15

    A Non Destructive Analysis (NDA) method was developed for the verification of {sup 235}U enrichment of both homogeneous and profiled VVER-440 reactor fresh fuel assemblies by means of gamma spectrometry. A total of ca. 30 assemblies were tested, five of which were homogeneous, with {sup 235}U enrichment in the range 1,6% to 3,6%, while the others were profiled with pins of 3,3% to 4,4% enrichment. Two types of gamma detectors were used for the test measurements: 2 coaxial HPGe detectors and a miniature CdZnTe (CZT) detector fitting into the central tube of the assemblies. It was therefore possible to obtain information from both the inside and the outside of the assemblies. It was shown that it is possible to distinguish between different types of assemblies within a reasonable measurement time (about 1000 sec). For the HPGe measurements the assemblies had to be lifted out from their storage rack, while for the CZT detector measurements the assemblies could be left at their storage position, as it was shown that the neighbouring assemblies do not affect measurement inside the assemblies' central tube. The measured values were compared to Monte Carlo simulations carried out using the MCNP code, and a recommendation for the optimal approach to verify the {sup 235}U enrichment of fresh VVER-440 reactor fuel assemblies is suggested.

  3. Physical startup tests for VVER-1200 of Novovoronezh NPP. Advanced technique and some results

    Energy Technology Data Exchange (ETDEWEB)

    Afanasiev, Dmitry A.; Kraynov, Yury A.; Pinegin, Anatoly A.; Tsyganov, Sergey V. [National Research Centre, Moscow (Russian Federation). Kurchatov Inst.

    2017-09-15

    The intention of the startup physics tests was to confirm design characteristics of the core loading and their compliance with safety analysis preconditions. The program of startup tests for the leading unit is usually composed in such a way that is is possible to study as much neutron-physical characteristics as possible in the safest condition of zero power. State-of-the-art safety analysis is including computer codes that use three dimensional neutron kinetics and thermohydraulics models. For the substantiation of such models, for its validation and verification there is a need in reactor experiments that implementing spatially distributed transients. We based on such statements when composing hot zero power physical startup program for the new VVER-1200 unit of Novovoronezh NPP. Several tests unconventional for VVER were developed for that program. It includes measuring the worth for each of control rod groups and measuring of single rod worth from the inserted groups - test that models rod ejection event in some sense.

  4. Neutron dosimetry in EDF experimental surveillance programme for VVER-440 nuclear power plants

    International Nuclear Information System (INIS)

    Brumovsky, M.; Erben, O.; Novosad, P.; Zerola, L.; Hogel, J.; Trollat, C.

    2001-01-01

    Fourteen chains containing experimental surveillance material specimens of the VVER 440/213 nuclear power reactor pressure vessels were irradiated in the surveillance channels of the Nuclear Power Plant Dukovany in the Czech Republic. The irradiation periods were one, two or three cycles. The chains contained different number and types of containers, the omitted ones were replaced by chain elements. All of the containers were instrumented with wire neutron fluence detectors, some of the containers in the chain had spectrometric sets of neutron fluence monitors. For the absolute fluence values evaluation it was taken into account time history of the reactor power and local changes of the neutron flux along the reactor core height, correction factors due to the orientation of monitors with respect to the reactor core centre. Unfolding programs SAND-II or BASA-CF were used. The relative axial fluence distribution was obtained from the O-wire measurements. Neutron fluence values above 0.5 MeV energy and above 1.0 MeV energy in the container axis on the axial positions of the sample centres and fluence values in the geometric centre of the samples was calculated making use the exponential attenuation model of the incident neutron beam. Received fast neutron fluence values can be used as reference values to all VVER-440 type 213 nuclear power plant reactors. (author)

  5. Numerical investigation of passive heat removal system via steam generator in VVER 1200

    International Nuclear Information System (INIS)

    Dinh Anh Tuan; Duong Thanh Tung; Tran Chi Thanh; Nguyen Van Thai

    2015-01-01

    Passive heat removal system (PHRS) via Steam Generator is an important part in VVER design. In case of Design Basic Accidents such as blackout, failure of feed water supply to steam generator or coolant leakage with failure of emergency core cooling at high pressure. PHRS is designed to remove the residual heat from reactor core through steam generator to heat exchanger which is placed outside reactor vessel. In order to evaluate the passive system, a numerical investigation using a CFD code is performed. However, PHRS has complex geometry for using CFD simulation. Thus, RELAP5 is applied to provide the wall heat flux of tube in the heat exchanger tank. The natural convection in the heat exchanger tank is investigated in this report. Numerical results show temperature and velocity distribution in the heat exchanger tank are calculated with different wall heat flux corresponding to various transient conditions. The calculated results contribute to the capacity analysis of passive heat removal system and giving valuable information for safe operation of VVER 1200. (author)

  6. A successful approach for the implementation of symptom-based emergency operating procedures for VVER reactors

    International Nuclear Information System (INIS)

    Lhoest, V.; Prior, R.; Pascal, G.

    2000-01-01

    The paper provides an overview of the organization, the progress and the results of the various Emergence Operating Procedure (EOP) development programs for VVER type reactors conducted by Westinghouse so far. The detailed working process is presented through the solutions to some major plant issues. The EOPs have been developed for the Temelin, Dukovany, Bohunice, Mochovce and Paks VVER nuclear power plants. The procedures are developed in working teams of experts from the utility and Westinghouse. The completion of the programs constitute an indication of the overall success of this approach. This is further reinforced by the general acceptance of the new procedures by the plant personnel, together with the good results obtained so far from procedure testing. This is also confirmed by a new PSA-level 1 analysis for Dukovany plant, which shows a significant improvement in the overall plant safety. This means a 20% reduction in the Core Damage Frequency due to the introduction of the new EOPs. The fact that some modifications have been implemented to the plants to solve design weaknesses identified in the course of this programs also constitute a positive result

  7. Perspectives for practical application of the combined fuel kernels in VVER-type reactors

    International Nuclear Information System (INIS)

    Baranov, V.; Ternovykh, M.; Tikhomirov, G.; Khlunov, A.; Tenishev, A.; Kurina, I.

    2011-01-01

    The paper considers the main physical processes that take place in fuel kernels under real operation conditions of VVER-type reactors. Main attention is given to the effects induced by combinations of layers with different physical properties inside of fuel kernels on these physical processes. Basic neutron-physical characteristics were calculated for some combined fuel kernels in fuel rods of VVER-type reactors. There are many goals in development of the combined fuel kernels, and these goals define selecting the combinations and compositions of radial layers inside of the kernels. For example, the slower formation of the rim-layer on outer surface of the kernels made of enriched uranium dioxide can be achieved by introduction of inner layer made of natural or depleted uranium dioxide. Other potential goals (lower temperature in the kernel center, better conditions for burn-up of neutron poisons, better retention of toxic materials) could be reached by other combinations of fuel compositions in central and peripheral zones of the fuel kernels. Also, the paper presents the results obtained in experimental manufacturing of the combined fuel pellets. (authors)

  8. Development and qualification of a thermal-hydraulic nodalization for modeling station blackout accident in PSB-VVER test facility

    Energy Technology Data Exchange (ETDEWEB)

    Saghafi, Mahdi [Department of Energy Engineering, Sharif University of Technology, Azadi Avenue, Tehran (Iran, Islamic Republic of); Ghofrani, Mohammad Bagher, E-mail: ghofrani@sharif.edu [Department of Energy Engineering, Sharif University of Technology, Azadi Avenue, Tehran (Iran, Islamic Republic of); D’Auria, Francesco [San Piero a Grado Nuclear Research Group (GRNSPG), University of Pisa, Via Livornese 1291, San Piero a Grado, Pisa (Italy)

    2016-07-15

    Highlights: • A thermal-hydraulic nodalization for PSB-VVER test facility has been developed. • Station blackout accident is modeled with the developed nodalization in MELCOR code. • The developed nodalization is qualified at both steady state and transient levels. • MELCOR predictions are qualitatively and quantitatively in acceptable range. • Fast Fourier Transform Base Method is used to quantify accuracy of code predictions. - Abstract: This paper deals with the development of a qualified thermal-hydraulic nodalization for modeling Station Black-Out (SBO) accident in PSB-VVER Integral Test Facility (ITF). This study has been performed in the framework of a research project, aiming to develop an appropriate accident management support tool for Bushehr nuclear power plant. In this regard, a nodalization has been developed for thermal-hydraulic modeling of the PSB-VVER ITF by MELCOR integrated code. The nodalization is qualitatively and quantitatively qualified at both steady-state and transient levels. The accuracy of the MELCOR predictions is quantified in the transient level using the Fast Fourier Transform Base Method (FFTBM). FFTBM provides an integral representation for quantification of the code accuracy in the frequency domain. It was observed that MELCOR predictions are qualitatively and quantitatively in the acceptable range. In addition, the influence of different nodalizations on MELCOR predictions was evaluated and quantified using FFTBM by developing 8 sensitivity cases with different numbers of control volumes and heat structures in the core region and steam generator U-tubes. The most appropriate case, which provided results with minimum deviations from the experimental data, was then considered as the qualified nodalization for analysis of SBO accident in the PSB-VVER ITF. This qualified nodalization can be used for modeling of VVER-1000 nuclear power plants when performing SBO accident analysis by MELCOR code.

  9. Structural strength during severe reactor accidents of the VVER- 91 nuclear power plant

    International Nuclear Information System (INIS)

    Varpasuo, P.

    1999-12-01

    The report summarises the studies carried out in Fortum Engineering (formerly IVO Power Engineering) between the years of 1992 and 1997 concerning ultimate strength of structures designed to mitigate and contain the consequences of various core melt accident scenarios. The report begins with the description of containment loading situations arising from core melt accidents. These situations are divided to fast and slow loads. Fast loads include ex-vessel steam explosions, steam spikes, hydrogen burns, direct containment heating and missiles. Slow loads are connected with pressure rise inside the containment in case when the containment heat removal system is not functioning. First part of report describes the analyses of reactor cavity based on axi-symmetric load assumptions. These studies are performed with various models like one degree of freedom idealisation, axi-symmetric modelling of geometry and full three-dimensional modelling of geometry. Second part of report describes the analyses of cavity based on non-axi-symmetric load assumptions. Here full 3D- geometry model is used combined with various physical models for the behaviour of reinforced concrete. Third part of report gives short account of the analysis of containment ultimate pressure capacity. The containment model in this case includes pre-stressing tendons and mild steel reinforcing bars. The load is assumed to axi-symmetric internal static pressure. The capacity of the reactor cavity against the ex-vessel steam explosion scenarios for VVER-91 plant concept is established for both axi-symmetric and non-axi-symmetric load models using ANACAP structural analysis code. The validation of the cavity response to ex-vessel steam explosion load using different commercially available codes gave mixed results for both axisymmetric and non-axi-symmetric load presentations.The ultimate static overpressure capacity of the VVER-91 reactor cavity structure was established to be of the order of 10 MPa. This result

  10. Investigation of reliability of EC method for inspection of VVER steam generator tubes

    International Nuclear Information System (INIS)

    Corak, Z.

    2004-01-01

    Complete and accurate non-destructive examinations (NDE) data provides the basis for performing mitigating actions and corrective repairs. It is important that detection and characterization of flaws are done properly at an early stage. EPRI Document PWR Steam Generator Examination Guidelines recommends an approach that is intended to provide the following: Ensure accurate assessment of steam generator tube integrity; Extend the reliable, cost effective, operating life of the steam generators, and Maximize the availability of the unit. Steam Generator Eddy Current Data Analysis Performance Demonstration represents the culmination of the intense two-year industry effort in the development of a performance demonstration program for eddy current testing (ECT) of steam generator tubing. It is referred to as the Industry Database (IDB) and provides a capability for individual organizations to implement SG ECT performance demonstration programs in accordance with the requirements specified in Appendices G and H of the ISI Guidelines. The Appendix G of EPRI Document PWR Steam Generator Examination Guidelines specifies personnel training and qualification requirements for NDE personnel who analyze NDE data for PWR steam generator tubing. Its purpose is to insure a continuing uniform knowledge base and skill level for data analysis. The European methodology document is intended to provide a general framework for development of qualifications for the inspection of specific components to ensure they are developed in a consistent way throughout Europe while still allowing qualification to be tailored in detail to meet different nation requirements. In the European methodology document one will not find a detailed description of how the inspection of a specific component should be qualified. A recommended practice is a document produced by ENIQ to support the production of detailed qualification procedures by individual countries. VVER SG tubes are inspected by EC method but a

  11. Axial stability of VVER-1000 reactor with control with minimum standard deviation

    International Nuclear Information System (INIS)

    Afanas'ev, A.M.; Torlin, B.Z.

    1980-01-01

    Results are given of investigations on the stability of a reactor which has, in addition to an automatic controller, a height distribution regulator (HDR) based on an auxiliary control rod (CR) or a special shortened absorption rod (SAR). The HDR was controlled by using either a special ionization chamber (IC), generating an imbalance signal which sets the CR in motion, or two ionization chambers whose difference signal causes a displacement of the SAR. Since data from numerous pickups can be used to control the height field of the VVER-1000, it is of interest to analyze how this would affect the stability of the reactor. The analysis was carried out with the improved IRINA programs. 11 refs

  12. Uranium oxide fuel cycle analysis in VVER-1000 with VISTA simulation code

    Science.gov (United States)

    Mirekhtiary, Seyedeh Fatemeh; Abbasi, Akbar

    2018-02-01

    The VVER-1000 Nuclear power plant generates about 20-25 tons of spent fuel per year. In this research, the fuel transmutation of Uranium Oxide (UOX) fuel was calculated by using of nuclear fuel cycle simulation system (VISTA) code. In this simulation, we evaluated the back end components fuel cycle. The back end component calculations are Spent Fuel (SF), Actinide Inventory (AI) and Fission Product (FP) radioisotopes. The SF, AI and FP values were obtained 23.792178 ton/y, 22.811139 ton/y, 0.981039 ton/y, respectively. The obtained value of spent fuel, major actinide, and minor actinide and fission products were 23.8 ton/year, 22.795 ton/year, 0.024 ton/year and 0.981 ton/year, respectively.

  13. Basket criticality design of a dual purpose cask for VVER 1000 spent fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Rezaeian, Mahdi [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of); Kamali, Jamshid [Atomic Energy Organization of Iran (AEOI), Tehran (Iran, Islamic Republic of)

    2016-12-15

    Dual purpose cask technology is one of the most prominent options for interim storage of spent fuels following their removal from reactors. Criticality safety of the spent fuel assemblies are ensured by design of the basket within these casks. In this study, a set of criticality design calculations of a dual purpose cask for 12 VVER 1000 spent fuel assemblies of Bushehr nuclear power plant were carried out. The basket material of borated stainless steel with 0.5 to 2.5 wt% of boron and Boral (Al-B{sub 4}C) with 1.5 to 40 wt% of boron carbide, were investigated and the minimum required receptacle pitch of the basket was determined. Using the calculated receptacle pitch of the basket, the minimum required diameter of the cavity could be established.

  14. AER Working Group D on VVER safety analysis - report of the 2009 meeting

    International Nuclear Information System (INIS)

    Kliem, S.

    2009-01-01

    The AER Working Group D on VVER reactor safety analysis held its 18-th meeting in Rez, Czech Republic, during the period 18-19 May, 2009. The meeting was hosted by the Nuclear Research Institute Rez. Altogether 17 participants attended the meeting of the working group D, 16 from AER member organizations and 1 guest from a non-member organization. The co-ordinator of the working group, 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: 1) Code validation and benchmarking; 2) Safety analysis and code developments; 3) Reactor pressure vessel thermal hydraulics; 4) Future activities including discussion on the participation in the OECD/NEA Benchmark for the Kalinin-3 NPP

  15. Uncertainty and sensitivity analysis applied to coupled code calculations for a VVER plant transient

    International Nuclear Information System (INIS)

    Langenbuch, S.; Krzykacz-Hausmann, B.; Schmidt, K. D.

    2004-01-01

    The development of coupled codes, combining thermal-hydraulic system codes and 3D neutron kinetics, is an important step to perform best-estimate plant transient calculations. It is generally agreed that the application of best-estimate methods should be supplemented by an uncertainty and sensitivity analysis to quantify the uncertainty of the results. The paper presents results from the application of the GRS uncertainty and sensitivity method for a VVER-440 plant transient, which was already studied earlier for the validation of coupled codes. For this application, the main steps of the uncertainty method are described. Typical results of the method applied to the analysis of the plant transient by several working groups using different coupled codes are presented and discussed The results demonstrate the capability of an uncertainty and sensitivity analysis. (authors)

  16. Development of some operations in technological flowsheet for spent VVER fuel reprocessing at a pilot plant

    International Nuclear Information System (INIS)

    Lazarev, L.N.; Galkin, B.Ya; Lyubtsev, R.I.; Romanovskii, V.N.; Velikhov, E.P.

    1981-01-01

    The fuel reprocessing pilot plants for high active materials would permit the study and development or particular processing steps and flowsheet variations; in some cases, these experimental installations realize on a small scale practically all technological chains of large reprocessing plants. Such a fuel reprocessing pilot plant with capacity of 3 kg U/d has been built at V. G. Khlopin Radium Institute. The pilot plant is installed in the hot cell of radiochemical compartment, and is composed of the equipments for fuel element cutting and dissolving, the preparation of feed solution (clarification, correction), extraction reprocessing and the production of uranium, plutonium and neptunium concentrates, the complex processing of liquid and solid wastes and a special unit for gas purification and analysis. In the last few years, a series of experiments have been carried out on the reprocessing of spent VVER fuel. (J.P.N.)

  17. VVER-440 training simulators upgrades - Experience of CORYS T.E.S.S

    International Nuclear Information System (INIS)

    Bartak, J.; Fallon, B.

    2006-01-01

    The paper presents recent projects of upgrading screen operated simulators of VVER-440 nuclear power plants to full scale replica simulators, implemented by CORYS TESS. Control room replica full scope simulators were built for the Bohunice NPP in Slovakia and the Novovoronezh NPP in Russia. The scope of simulation was extended to reflect the current status of the units, which have undergone significant modernization programs over the last few years. The paper describes the software and hardware adaptations and evolutions of the existing simulators, the implementation in the simulator of modern supervision systems as well as of systems and equipment designed in the seventies and still used on the reference units. The training benefits of parallel use of control room replica and screen-operated simulators in the training process are discussed. (author)

  18. Technology of repair of selected equipment in the power plant type VVER 440

    Energy Technology Data Exchange (ETDEWEB)

    Barborka, J.; Magula, V. [Welding Research Inst. (WRI), Bratislava (Slovakia)

    1998-11-01

    This article is divided in two parts: The first part is studying the effect of individual parameters by the usual and pulsed welding of 15CH2MFA steel. It can be concluded that by use of mechanized or automatic TIG process in PC position with addition of a cold wire with high nickel content the desired quality of repair welded joints of a pressure vessel of VVER 440 reactor can be achieved. Based on the results of the second laboratory study of the renovation technology applied for the rotary surfaces of pressure-tight cover and spindle of the main closing armature type DN 500 it can be concluded, that the developed technology for surfacing the sealing surfaces by TIG process with addition of a high-nickel cold wire the functional capability of the mentioned parts can be fully restored.

  19. Design study on steam generator integration into the VVER reactor pressure vessel

    International Nuclear Information System (INIS)

    Hort, J.; Matal, O.

    2004-01-01

    The primary circuit of VVER (PWR) units is arranged into loops where the heat generated by the reactor is removed by means of main circulating pumps, loop pipelines and steam generators, all located outside the reactor pressure vessel. If the primary circuit and reactor core were integrated into one pressure vessel, as proposed, e.g., within the IRIS project (WEC), a LOCA situation would be limited by the reactor pressure vessel integrity only. The aim of this design study regarding the integration of the steam generator into the reactor pressure vessel was to identify the feasibility limits and some issues. Fuel elements and the reactor pressure vessel as used in the Temelin NPP were considered for the analysis. From among the variants analyzed, the variant with steam generators located above the core and vertically oriented circulating pumps at the RPV lower bottom seems to be very promising for future applications

  20. Technology of repair of selected equipment in the power plant type VVER 440

    International Nuclear Information System (INIS)

    Barborka, J.; Magula, V.

    1998-01-01

    This article is divided in two parts: The first part is studying the effect of individual parameters by the usual and pulsed welding of 15CH2MFA steel. It can be concluded that by use of mechanized or automatic TIG process in PC position with addition of a cold wire with high nickel content the desired quality of repair welded joints of a pressure vessel of VVER 440 reactor can be achieved. Based on the results of the second laboratory study of the renovation technology applied for the rotary surfaces of pressure-tight cover and spindle of the main closing armature type DN 500 it can be concluded, that the developed technology for surfacing the sealing surfaces by TIG process with addition of a high-nickel cold wire the functional capability of the mentioned parts can be fully restored

  1. Inter-assembly gap deviations in VVER-1000: Accounting for effects on engineering margin factors

    Energy Technology Data Exchange (ETDEWEB)

    Shishkov, Lev; Gorodkov, Sergey; Mikailov, Eldar; Sukhino-Khomenko, Evgenia [Nuclear Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

    2015-09-15

    Jacketless fuel assemblies change their form in the course of operation. Often they bow lengthwise. Primarily, these fuel assembly (FA) bows threaten to reduce the control rods' fall rate, but at the same time they change (e.g. increase) the amount of moderator in inter-assembly gaps, thus producing additional power surges. Gap sizes vary randomly and their impact is accounted for with the help of engineering margin factors. For VVER-1000, this account of engineering margin factors increases the fuel component of electricity generation cost by 3 - 5 %, and a half of this increase is due to inter- assembly gap variations. This paper discusses the technique used to account for the impact produced by these gaps on fuel rod power; gives numerical values of sensitivity factors for power variations vs. gap sizes depending on the computational model assumed; and discusses the interference of gap effects and the account of power and coolant temperature feedbacks.

  2. KUPOL-M code for simulation of the VVER's accident localization system under LOCA conditions

    International Nuclear Information System (INIS)

    Efanov, A.D.; Lukyanov, A.A.; Shangin, N.N.; Zajtsev, A.A.; Solov'ev, S.L.

    2004-01-01

    Computer code KUPOL-M is developed for analysis of thermodynamic parameters of medium within full pressure containment for NPPs with VVER under LOCA conditions. The analysis takes into account the effects of non-stationary heat-mass transfer of gas-drop mixture in the containment compartments with natural convection, volume and surface steam condensation in the presence of noncondensables, heat-mass exchange of the compartment atmosphere with water in the sumps. The operation of the main safety systems like a spray system, hydrogen catalytic recombiners, emergency core cooling pumps, valves and a fan system is simulated in KUPOL-M code. The main results of the code verification including the ones of the participation in ISP-47 International Standard Problem on containment thermal-hydraulics are presented. (author)

  3. Economical Feedback of Increasing Fuel Enrichment on Electricity Cost for VVER-1000

    Directory of Open Access Journals (Sweden)

    Mohammed Saad Dwiddar

    2015-08-01

    Full Text Available A methodology of evaluating the economics of the front-end nuclear fuel cycle with a price change sensitivity analysis for a VVER-1000 reactor core as a case study is presented. The effect of increasing the fuel enrichment and its corresponding reactor cycle length on the energy cost is investigated. The enrichment component was found to represent the highly expenses dynamic component affecting the economics of the front-end fuel cycle. Nevertheless, the increase of the fuel enrichment will increase the reactor cycle length, which will have a positive feedback on the electricity generation cost (cent/KWh. A long reactor operation time with a cheaper energy cost set the nuclear energy as a competitive alternative when compared with other energy sources.

  4. Increase in VVER type reactor critical heat fluxes due to placing the mixing grids

    International Nuclear Information System (INIS)

    Bezrukov, Y.; Lisenkov, E.; Vasilchenko, I.

    2011-01-01

    The report deals with the results of studies of critical heat fluxes (CHF) on the models of VVER type reactor fuel assembly models equipped with the 'Vihr' intensifiers-grids. The models are the seven-rod bundles with the uniform and non-uniform axial power that correspond to two periods of FA operation i.e. beginning of cycle and end of cycle. The experiments performed showed that the mixing grids of this type are capable of increasing the FA burnout power. The power ascension rate depends on both coolant pressure and steam quality value in the CHF point. Placing the mixing grids in the bundle upper spans results in shifting the point of DNB occurrence downward along the FA height. The experimental data obtained will be used to develop the correlations for determining the CHF in the FA equipped with the mixing grids. (authors)

  5. Application of the SCANAIR code for VVER RIA conditions - Boron dilution accident

    International Nuclear Information System (INIS)

    Arffman, A.; Cazalis, B.

    2010-01-01

    This paper consists of two parts. In part A, RIA pulse tests conducted at the Russian BIGR reactor are being analysed at IRSN with SCANAIR V6 fuel performance code as a part of the code validation for VVER fuel. Recently a new version of the SCANAIR code was made available to VTT Technical Research Centre of Finland, and part B of the paper covers the introduction of the code version at VTT by a calculation of a hypothetical boron dilution accident in a VVER-440 power reactor. Concerning part A, it appears that the SCANAIR V6 version, including a BIGR/NSRR heat transfer model, validated by Japanese NSRR experiments, and a Norton viscoplastic clad mechanical behaviour, is able to simulate the rod thermal behaviour in BIGR tests. Concerning the clad mechanics, it has been seen that a pellet swelling model is able to simulate the average rod deformation. Nonetheless, the current clad creep model associated with the free volume equilibrium assumption is not suited to predict the maximum clad deformation and the possible post DNB rod failure because they do not simulate local balloons. Furthermore, it has been shown that the clad deformation is strongly dependent on transient gas transfer. Concerning part B, a boron dilution accident previously calculated with SCANAIR V2 was recalculated with SCANAIR V6. A limited amount of result parameters were compared with the results of VTT's neutronics code TRAB. Divergence problems encountered previously when reaching the DNB limit were not present anymore. Fuel and cladding temperatures produced by SCANAIR were in good agreement with those calculated with TRAB

  6. Plasma, a plant safety monitoring and assessment system for VVER-440 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hornaes, A.; Hulsund, J. E. [Institutt for energiteknikk (IFE), OECD Halden Reactor Project, Halden (Norway); Lipcsei, S.; Major, Cs.; Racz, A.; Vegh, J. [KFKI, Atomic Energy Research Institute, Budapest (Hungary); Eiler, J. [Paks, Nuclear Power Plant Ltd, Paks (Hungary)

    1999-05-15

    The objective with the Plant Safety Monitoring and Assessment System (PLASMA) is to develop an operator support system to support the execution of new symptom-based Emergency Operating Procedures for application in VVER reactors, with the Paks NPP in Hungary as the target plant. Many of the VVER reactors are rewriting their EOPs to comply more with Western standards of symptom-based EOPs. In this connection it is desirable to improve the data validation, information integration and presentation for operators when executing the EOPs. The entry-point to a symptom-oriented procedure is defined by the occurrence of a well-defined reactor operation status, with all its symptoms. However, the application of the EOF benefits from an operator support system, which performs plant status and symptom identification reliably and accurately. The development of the PLASMA system is a joint venture between Institutt for energiteknikk (IFE) and KFKI with the NPP Paks as the target plant. The project has been initiated and partly funded by the Science and Technology Agency (STA), Japan through the OECD NEA assistance program. In Hungary, considerable effort has concentrated on the safety reassessment of the Paks NPP and new EOPs are being written, but no comprehensive Operator Support System (OSS) for plant safety assessment is installed. Some safety parameter display functions are incorporated into diverse operator support systems, but an online 'plant safety monitoring and assessment system' is still missing. The present project comprises designing, constructing, testing and installing such an OSS, which to a great extent could support plant operators in their safety assessment work (author) (ml)

  7. Assessment of In-vessel corium retention for VVER-440/V213

    International Nuclear Information System (INIS)

    Matejovic, P.; Barnak, M.; Bachraty, M.; Berky, R.

    2011-01-01

    In-vessel corium retention (IVR) via external reactor vessel cooling (ERVC) has been recognised as a feasible and promising severe accident management strategy for VVER-440/V213 reactors. In general, the avoiding of boiling crisis on outer (cooled) RPV (reactor pressure vessel) surface is sufficient condition for preserving the RPV integrity. The crucial point of the proposed IVR concept for VVER-440/V213 is the narrow gap between elliptical lower head and thermal and biological shield. In the cold conditions the width of this gap is only about 2 cm and would be even lower in hot IVR conditions, when the reactor wall is subjected to large thermal gradients due to temperature difference between the hot inner surface (loaded by corium) and cold outer surface (which is cooled by water in flooded cavity). Sufficient gap should remain free for coolant flow for the success of the proposed IVR concept. Thus, realistic estimation of thermal load and corresponding deformations of reactor wall and their impact on gap width are of primarily importance. Two different approaches were used for the estimation of the thermal load: a conservative approach and a transient approach, both were computed with the ASTEC code. The structural analysis of RPV subjected to IVR load was performed using the finite element method (FEM) code ANSYS release 10.0. From the results obtained it follows, that even when the RPV is subjected to limiting loading conditions during severe accident, there should be sufficient gap width (∼ 1 cm) between RPV wall and thermal/biological shield for the coolant flow in natural circulation regime alongside the outer surface of the RPV wall

  8. Some benchmark calculations for VVER-1000 assemblies by WIMS-7B code

    International Nuclear Information System (INIS)

    Sultanov, N.V.

    2001-01-01

    Our aim in this report is to compare of calculation results, obtained with the use of different libraries, which are in the variant of the WIMS7B code. We had the three libraries: the 1986 library is based on the UKNDL files, the two 1996 libraries are based on the JEF-2.2 files, the one having the 69 group approximation, the other having the 172 group approximation. We wanted also to have some acquaintance with the new option of WIMS-7B - CACTUS. The variant of WIMS-7B was placed at our disposal by the code authors for a temporal use for 9 months. It was natural to make at comparisons with analogous values of TVS-M, MCU, Apollo-2, Casmo-4, Conkemo, MCNP, HELIOS codes, where the other different libraries were used. In accordance with our aims the calculations of unprofiled and profiled assemblies of the VVER-1000 reactor have been carried out by the option CACTUS. This option provides calculations by the characteristics method. The calculation results have been compared with the K ∞ values obtained by other codes in work. The conclusion from this analysis is such: the methodical parts of errors of these codes have nearly the same values. Spacing for K eff values can be explained of the library microsections differences mainly. Nevertheless, the more detailed analysis of the results obtained is required. In conclusion the calculation of a depletion of VVER-1000 cell has been carried out. The comparison of the dependency of the multiply factor from the depletion obtained by WIMS-7B with different libraries and by the TVS-M, MCU, HELIOS and WIMS-ABBN codes in work has been performed. (orig.)

  9. Development of a new chemical technology for cleanup of VVER steam generators

    International Nuclear Information System (INIS)

    Smykov, V.B.; Yermolaev, N.P.; Ivanov, V.N.

    2002-01-01

    As shows the maintenance experience of SG's, the long-time maintenance them without chemical cleanup on secondary-side results in accumulation of considerable amounts of depositions of oxides of iron with a high content of copper on outside of tubes. The deposit accumulation creates conditions for concentrating of salts which promote corrosion and, then, the loosing of inter-contour tightness. Therefore the experts do not have any doubts in necessity of chemical cleanups and the chemical cleanups were carried out at some NPP's with VVER during last years. However it is possible to say, that these cleanups were carried out not by the best mode - the same main reagents had been used in order to dissolve the copper and iron oxides. For example, all cleanups at Balakovo NPP in 1996-1997 years had the common deficiency - even during 5. final stage of process the copper prolongs to be washed. By our opinion, the reasons of it are the poor scientific and technical justification of this process. Therefore at various NPP's with VVER cleanups realize by various techniques. The process of chemical cleanup, close to offered in the present work, was repeated many times utilized at BN-600 Belojarsk NPP and at BN-350 Shevtchenko NPP. The purposes of the present work are: 1. Research the behaviours of physicochemical processes during dissolution of components of depositions and their mixtures with use of the various formulas; 2. Analysis of the carried out chemical cleanups of PGV-1000M at an example of Balakovo NPP; 3. Development of a new process of SG's cleanup on the base of experimental researches and analysis; 4. Check of this process on the samples of full-scale depositions from SG Balakovo NPP. (authors)

  10. SCORPIO-VVER core monitoring and surveillance system with advanced capabilities

    International Nuclear Information System (INIS)

    Molnar, Jozef; Vocka, Radim

    2010-01-01

    The SCORPIO (SCORPIO-VVER) core monitoring system, its basic features and history of implementation at Czech NPPs are described. The most important improvements in the area of neutron physics, core thermal analysis and operation support are as follows: Moving to the 42 axial nodes across the whole system (2004); Implementation of new cross section library to support mixed reactor core with differences in axial geometry of used fuel types and enhancement of Core Simulator boundary conditions model, to properly address the 'wild' geometry in axial direction; Adjusting the thermohydraulic and neutron-physical models regarding to the Gd2 fuel needs; Support up to 5 types of FAs and 2 types of SPND (Posit, IST); Extension of form functions for pin-wise reconstruction to improve pin-power prediction in control rod coupler region; System adaptation to the new upgraded digital I and C unit system; Integration of the SCORPIO-VVER system and its workstation into the plant redundant in-core system; Implementation of new On-Line form function generation to module RECON; New design of the Strategy Generator with advanced predictions; Adaptation of the system to support the new up-rated reactor thermal power; Adding new online SDM calculation function into to system; Implementation of the new 3D power reconstruction with SPND interpretation; Extending the limit checking to the 'full core' checking. The control of margins to the technical specification: Extended to full core - all FA is controlled individually in core; The limits are definable up to 59 FA (1/6 symmetry); 4 limited parameters are controlled - Kr, qlin, Tout-fa, dTfa; 2 additional parameters are monitored - dTsat, DNBR; New MMI are developed to present the limited and controlled parameters in core. Upgrade 3 is planned for the Slovak Bohunice NPP in 2011-2012. (P.A.)

  11. Delayed Neutron Fraction (beta-effective) Calculation for VVER 440 Reactor

    International Nuclear Information System (INIS)

    Hascik, J.; Michalek, S.; Farkas, G.; Slugen, V.

    2008-01-01

    Effective delayed neutron fraction (β eff ) is the main parameter in reactor dynamics. In the paper, its possible determination methods are summarized and a β eff calculation for a VVER 440 power reactor as well as for training reactor VR1 using stochastic transport Monte Carlo method based code MCNP5 is made. The uncertainties in determination of basic delayed neutron parameters lead to the unwished conservatism in the reactor control system design and operation. Therefore, the exact determination of the β eff value is the main requirement in the field of reactor dynamics. The interest in the delayed neutron data accuracy improvement started to increase at the end of 80-ties and the beginning of 90-ties, after discrepancies among the results of experiments and measurements what do you mean differences between different calculation approaches and experimental results. In consequence of difficulties in β eff experimental measurement, this value in exact state is determined by calculations. Subsequently, its reliability depends on the calculation method and the delayed neutron data used. An accurate estimate of β eff is essential for converting reactivity, as measured in dollars, to an absolute reactivity and/or to an absolute k eff . In the past, k eff has been traditionally calculated by taking the ratio of the adjoint- and spectrum-weighted delayed neutron production rate to the adjoint- and spectrum-weighted total neutron production rate. An alternative method has also been used in which β eff is calculated from simple k-eigenvalue solutions. The summary of the possible β eff determination methods can be found in this work and also a calculation of β eff first for the training reactor VR1 in one operation state and then for VVER 440 power reactor in two different operation states are made using the prompt method, by MCNP5 code.(author)

  12. Review of operational requirements with respect to PCMI in a VVER and the corresponding developments in the trans uranus code

    International Nuclear Information System (INIS)

    Van Uffelen, P.; Lassman, K.; Schubert, A.; Van der Laar, J.; Gyori, C.; Elenkov, D.; Hatala, B.

    2005-01-01

    Since the mid-90's, a version of the TRANSURANUS code has been under development for the analysis of the fuel rod performance in Russian-type VVER reactors. This required, among other things, the implementation of specific thermal and mechanical properties for Nb-containing cladding. The first part of the paper summarises the present status of the models for normal operating conditions. Further refinements will include the correlation between the effective creep strain rate and the effective stress. In the second part of the paper we consider accident conditions for which new correlations have been developed, including plastic deformation, high-temperature oxidation and burst of the cladding. These conditions have been implemented in TRANSURANUS and verified by means of burst tests for as-received, oxidised and irradiated cladding specimens. Finally, an outlook of the planned activities for code development and validation, including experiments regarding PCMI-related safety criteria for VVER reactors, is presented. (author)

  13. Evaluation of the applicability of cladding deformation model in RELAP5/MOD3.2 code for VVER-1000 fuel

    International Nuclear Information System (INIS)

    Vorob'ev, Yu.; Zhabin, O.

    2015-01-01

    Applicability of cladding deformation model in RELAP5/MOD3.2 code is analyzed for VVER-1000 fuel cladding from Zr+1%Nb alloy. Experimental data and calculation model of fuel assembly channel of the core are used for this purpose. The model applicability is tested for the cladding temperature range from 600 to 1200 deg C and pressure range from 1 to 12 MPa. Evaluation results demonstrate limited applicability of built-in RELAP5/MOD3.2 cladding deformation model to the estimation of Zr+1%Nb cladding rupture conditions. The limitations found shall be considered in application of RELAP5/MOD3.2 cladding deformation model in the design-basis accident analysis of VVER reactors

  14. Calculations of 3D full-scale VVER fuel assembly and core models using MCU and BIPR-7A codes

    Energy Technology Data Exchange (ETDEWEB)

    Aleshin, Sergey S.; Bikeev, Artem S.; Bolshagin, Sergey N.; Kalugin, Mikhail A.; Kosourov, Evgeniy K.; Pavlovichev, Aleksandr M.; Pryanichnikov, Aleksandr V.; Sukhino-Khomenko, Evgenia A.; Shcherenko, Anna I.; Shcherenko, Anastasia I.; Shkarovskiy, Denis A. [Nuclear Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

    2015-09-15

    Two types of calculations were made to compare BIPR-7A and MCU results for 3D full-scale models. First EPS (emergency protection system) efficiency and in-core power distributions were analyzed for an equilibrium fuel load of VVER-1000 assuming its operation within an 18-month cycle. Computations were performed without feedbacks and with fuel burnup distributed over the core. After 3D infinite lattices of full-scale VVER-1000 fuel assemblies (A's) with uranium fuel 4.4% enrichment and uranium-erbium fuel 4.4% enrichment and Er{sub 2}O{sub 3} 1 % wt were considered. Computations were performed with feedbacks and fuel burnup at the constant power level. For different time moments effective multiplication factor and power distribution were obtained. EPS efficiency and reactivity effects at chosen time moments were analyzed.

  15. Kinetics Parameters of VVER-1000 Core with 3 MOX Lead Test Assemblies To Be Used for Accident Analysis Codes

    International Nuclear Information System (INIS)

    Pavlovitchev, A.M.

    2000-01-01

    The present work is a part of Joint U.S./Russian Project with Weapons-Grade Plutonium Disposition in VVER Reactor and presents the neutronics calculations of kinetics parameters of VVER-1000 core with 3 introduced MOX LTAs. MOX LTA design has been studied in [1] for two options of MOX LTA: 100% plutonium and of ''island'' type. As a result, zoning i.e. fissile plutonium enrichments in different plutonium zones, has been defined. VVER-1000 core with 3 introduced MOX LTAs of chosen design has been calculated in [2]. In present work, the neutronics data for transient analysis codes (RELAP [3]) has been obtained using the codes chain of RRC ''Kurchatov Institute'' [5] that is to be used for exploitation neutronics calculations of VVER. Nowadays the 3D assembly-by-assembly code BIPR-7A and 2D pin-by-pin code PERMAK-A, both with the neutronics constants prepared by the cell code TVS-M, are the base elements of this chain. It should be reminded that in [6] TVS-M was used only for the constants calculations of MOX FAs. In current calculations the code TVS-M has been used both for UOX and MOX fuel constants. Besides, the volume of presented information has been increased and additional explications have been included. The results for the reference uranium core [4] are presented in Chapter 2. The results for the core with 3 MOX LTAs are presented in Chapter 3. The conservatism that is connected with neutronics parameters and that must be taken into account during transient analysis calculations, is discussed in Chapter 4. The conservative parameters values are considered to be used in 1-point core kinetics models of accident analysis codes

  16. Development of a cross-section methodology and a real-time core model for VVER-1000 simulator application

    Energy Technology Data Exchange (ETDEWEB)

    Georgieva, Emiliya Lyudmilova

    2016-06-06

    The novel academic contributions are summarized as follows. A) A cross-section modelling methodology and a cycle-specific cross-section update procedure are developed to meet fidelity requirements applicable to a cycle-specific reactor core simulation, as well as particular customer needs and practices supporting VVER-1000 operation and safety. B) A real-time version of the Nodal Expansion Method code is developed and implemented into Kozloduy 6 full-scope replica control room simulator.

  17. Analytical validation of operator actions in case of primary to secondary leakage for VVER-1000/V320

    Energy Technology Data Exchange (ETDEWEB)

    Andreeva, M., E-mail: m_andreeva@inrne.bas.bg; Groudev, P., E-mail: pavlinpg@inrne.bas.bg; Pavlova, M., E-mail: pavlova@inrne.bas.bg

    2015-12-15

    Highlights: • We validate operator actions in case of primary to secondary leakage. • We perform four scenarios related to SGTR accident for VVER-1000/V320. • The reference power plant for the analyses is Unit 6 at Kozloduy NPP. • The RELAP5/MOD 3.2 computer code is used in performing the analyses. • The analyses confirm the effectiveness of operator actions during PRISE. - Abstract: This paper presents the results of analytical validation of operator actions in case of “Steam Generator Tube Rupture” (SGTR) for VVER-1000/V320 units at Kozloduy Nuclear Power Plant (KNPP), done during the development of Symptom Based Emergency Operating Procedures (SB EOPs) for this plant. The purpose of the analyses is to demonstrate the ability to terminate primary to secondary leakage and to indicate an effective strategy for preventing secondary leakage to the environment and in this way to prevent radiological release to the environment. Following depressurization and cooldown of reactor coolant system (RCS) with isolation of the affected steam generator (SG), in these analyses are validated options for post-SGTR cooldown by: • back up filling the ruptured SG; • using letdown system in the affected SG and • by opening Fast Acting Isolation Valve (FAIV) and using Steam Dump Facility to the Condenser (BRU-K). The results of the thermal-hydraulic analyses have been used to assist KNPP specialists in analytical validation of EOPs. The RELAP5/MOD3.2 computer code has been used for the analyses in a VVER-1000 Nuclear Power Plant (NPP) model. A model of VVER-1000 based on Unit 6 of Kozloduy NPP has been developed for the thermal-hydraulics code RELAP5/MOD3.2 at the Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Sciences (INRNE-BAS). This paper is possible through the participation of leading specialists from KNPP.

  18. The virtual digital nuclear power plant: A modern tool for supporting the lifecycle of VVER-based nuclear power units

    Science.gov (United States)

    Arkadov, G. V.; Zhukavin, A. P.; Kroshilin, A. E.; Parshikov, I. A.; Solov'ev, S. L.; Shishov, A. V.

    2014-10-01

    The article describes the "Virtual Digital VVER-Based Nuclear Power Plant" computerized system comprising a totality of verified initial data (sets of input data for a model intended for describing the behavior of nuclear power plant (NPP) systems in design and emergency modes of their operation) and a unified system of new-generation computation codes intended for carrying out coordinated computation of the variety of physical processes in the reactor core and NPP equipment. Experiments with the demonstration version of the "Virtual Digital VVER-Based NPP" computerized system has shown that it is in principle possible to set up a unified system of computation codes in a common software environment for carrying out interconnected calculations of various physical phenomena at NPPs constructed according to the standard AES-2006 project. With the full-scale version of the "Virtual Digital VVER-Based NPP" computerized system put in operation, the concerned engineering, design, construction, and operating organizations will have access to all necessary information relating to the NPP power unit project throughout its entire lifecycle. The domestically developed commercial-grade software product set to operate as an independently operating application to the project will bring about additional competitive advantages in the modern market of nuclear power technologies.

  19. Phase and structural transformations in VVER-440 RPV base metal after long-term operation and recovery annealing

    Science.gov (United States)

    Kuleshova, E. A.; Gurovich, B. A.; Maltsev, D. A.; Frolov, A. S.; Bukina, Z. V.; Fedotova, S. V.; Saltykov, M. A.; Krikun, E. V.; Erak, D. Yu; Zhurko, D. A.; Safonov, D. V.; Zhuchkov, G. M.

    2018-04-01

    This study was carried out to evaluate the possibility of 1st generation VVER-440 reactors lifetime extension by recovery re-annealing with the respect to base metal (BM). Comprehensive studies of the structure and properties of BM templates (samples cut from the inner surface of the shells in beltline region) of operating VVER-440 reactor (after primary standard recovery annealing 475 °C/150 h and subsequent long-term re-irradiation within reactor pressure vessel (RPV)) were conducted. These templates were also subjected to laboratory re-annealing 475 °C/150 h. TEM, SEM and APT studies of BM after laboratory re-annealing revealed significant recovery of radiation-induced hardening elements (Cu-rich precipitates and dislocation loops). Simultaneously a process of strong phosphorus accumulation at grain boundaries occurs since annealing temperature corresponds to the maximum reversible temper brittleness development. The latter is not observed for VVER-440 weld metal (WM). Comparative assessment of the properties return level for the beltline BM templates after recovery re-annealing 475 °C/150 h showed that it does not reach the one typical for beltline WM after the same annealing.

  20. The status of the Bubbler Condenser Containment System for the Reactors of the VVER-440/213 Type

    International Nuclear Information System (INIS)

    Karwat, H.; Rosinger, H.E.

    1998-01-01

    VVER-440/213 Pressurized Water Reactors have a pressure-suppression containment structure called a 'Bubbler Condenser' tower which can reduce the design pressure of the entire containment following a design basis accident (DBA), such as a loss-of-coolant accident (LOCA). The bubbler condenser pressure suppression system provides reduction of the LOCA containment pressure by the condensation of released steam in a water pool. World-wide there are 14 nuclear power plants of the VVER-440/213 type in Eastern Europe and Russia. One of the safety concerns for the VVER-440/213 reactors relates to the ability of the bubbler condenser containment system to function satisfactorily and to maintain its integrity following certain postulated accidents and thus limit the release of radioactive material to the environment. The complicated geometry of the bubbler condenser unit, and the dependence on several moving devices and interlocks are the main doubts expressed by different specialists with regard to the design. General description of the bubbler condenser containment system, the physical processes, concerns and design assessment of the bubbler condenser containment system, presentation of the OECD's Unified Bubbler Condenser Research Project (UBCRP) and the European Commission PHARE/TACIS project. Recent utility investigations are also discussed

  1. The Procedure for Determination of Special Margin Factors to Account for a Bow of the VVER-1000 Fuel Assemblies

    International Nuclear Information System (INIS)

    Tsyganov, Sergey V.; Marin, Stanislav V.; Shishkov, Lev K.

    2008-01-01

    Starting from 1980's, the problem of bow of the VVER-1000 reactor FAs 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 FAs that eliminated the problems of control rods. However, bow of the VVER-1000 reactor FAs 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 FAs of state-of-the-art designs. This technique is employed in the VVER-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)

  2. The Procedure for Determination of Special Margin Factors to Account for a Bow of the VVER-1000 Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Tsyganov, Sergey V.; Marin, Stanislav V.; Shishkov, Lev K. [Russian Research Center ' Kurchatov Institute' , 1., Kurchatov sq., 123182 Moscow (Russian Federation)

    2008-07-01

    Starting from 1980's, the problem of bow of the VVER-1000 reactor FAs 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 FAs that eliminated the problems of control rods. However, bow of the VVER-1000 reactor FAs 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 FAs of state-of-the-art designs. This technique is employed in the VVER-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)

  3. The choice of the fuel assembly for VVER-1000 in a closed fuel cycle based on REMIX-technology

    International Nuclear Information System (INIS)

    Bobrov, E.; Alekseev, P.; Chibinyaev, A.; Teplov, P.; Dudnikov, A.

    2016-01-01

    REMIX (Regenerated Mixture) fuel is produced directly from a non-separated mix of recycled uranium and plutonium from reprocessed used fuel and the fabrication technology of such fuel is called REMIX-technology. This paper shows basic features of different fuel assembly (FA) application for VVER-1000 in a closed fuel cycle based on REMIX-technology. This investigation shows how the change in the water-fuel ratio in the VVER FA affects the fuel characteristics produced by REMIX technology during multiple recycling. It is shown that for for the traditional REMIX-fuel it does not make sense to change anything in the design of VVER FA, because there are no advantages in the fuel feed consumption. The natural uranium economy by the fifth cycle reached about 29%. In the case of the REMIX fuel based on uranium-plutonium from SNF MOX fuel, it would be appropriate to use fuel assemblies with a water-fuel ratio of 1.5

  4. Experimental Investigation of Operation of VVER Steam Generator in Condensation Mode in the Event of the Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

    Morozov, Andrey [Institute for Physics and Power Engineering by A.I. Leypunsky, 1 Bondarenko sq. Obninsk, 249033 (Russian Federation)

    2008-07-01

    For new Russian nuclear power plants with VVER-1200 reactor in the event of a beyond design basis accident, provision is made for the use of passive safety systems for necessary core cooling. These safety systems include the passive heat removal system (PHRS). In the case of leakage in the primary circuit this system assures the transition of steam generators (SG) to operation in the mode of condensation of the primary circuit steam. As a result, the condensate from SG arrives at the core providing its additional cooling. To investigate the condensation mode of VVER SG operation, a large scale HA2M-SG test facility was constructed. The rig incorporates: buffer tank, SG model with scale is 1:46, PHRS heat exchanger. Experiments at the test facility have been performed to investigate condensation mode of operation of SG model at the pressure 0.4 MPa, correspond to VVER reactor pressure at the last stage of the beyond design basis accident. The report presents the test procedure and the basic obtained test results. (authors)

  5. PMK-2, the First Integral Thermal-Hydraulics Tests for the Safety Evaluation of VVER-440/213 Nuclear Power Plants

    Directory of Open Access Journals (Sweden)

    Gy. Ézsöl

    2012-01-01

    Full Text Available The PMK-2 facility is a full-pressure thermal-hydraulic model of the primary and partly the secondary circuit of the VVER-type units of Paks NPP. The facility was the first integral-type facility for VVERs. The PMK-2 was followed later by the PACTEL (for VVER-440, the ISB, and PSB for VVER-1000. Since the startup of the facility in 1985, 55 experiments have been performed primarily in international frameworks with the participation of experts from 29 European and overseas countries forming a scientific school to better understand VVER system behaviour and reach a high level of modelling of accident sequences. The ATHLET, CATHARE, and RELAP5 codes have been validated including both qualitative and quantitative assessments. The former was almost exclusively applied to the early phase of validation by integral experiments, while the quantitative assessments have been performed by the Fast Fourier Transform Based Method. Paper gives comprehensive information on the design features of PMK-2 facility with a special respect to the representativeness of phenomena, the experiments performed, and the results of the validation of ATHLET, CATHARE, and RELAP5 codes. Safety significance of the PMK-2 projects is also discussed.

  6. PMK-2, the First Integral Thermal-Hydraulics Tests for the Safety Evaluation of VVER-440/213 Nuclear Power Plants

    International Nuclear Information System (INIS)

    Ezsol, G.; Perneczky, L.; Szabados, L.; Toth, I.

    2012-01-01

    The PMK-2 facility is a full-pressure thermal-hydraulic model of the primary and partly the secondary circuit of the VVER-type units of Paks NPP. The facility was the first integral-type facility for VVERs. The PMK-2 was followed later by the PACTEL (for VVER-440), the ISB, and PSB for VVER-1000. Since the startup of the facility in 1985, 55 experiments have been performed primarily in international frameworks with the participation of experts from 29 European and overseas countries forming a scientific school to better understand VVER system behaviour and reach a high level of modelling of accident sequences. The ATHLET, CATHARE, and RELAP5 codes have been validated including both qualitative and quantitative assessments. The former was almost exclusively applied to the early phase of validation by integral experiments, while the quantitative assessments have been performed by the Fast Fourier Transform Based Method. Paper gives comprehensive information on the design features of PMK-2 facility with a special respect to the representativeness of phenomena, the experiments performed, and the results of the validation of ATHLET, CATHARE, and RELAP5 codes. Safety significance of the PMK-2 projects is also discussed

  7. Heuristic rules embedded genetic algorithm to solve VVER loading pattern optimization problem

    International Nuclear Information System (INIS)

    Fatih, Alim; Kostandi, Ivanov

    2006-01-01

    Full text: Loading Pattern (LP) optimization is one of the most important aspects of the operation of nuclear reactors. A genetic algorithm (GA) code GARCO (Genetic Algorithm Reactor Optimization Code) has been developed with embedded heuristic techniques to perform optimization calculations for in-core fuel management tasks. GARCO is a practical tool that includes a unique methodology applicable for all types of Pressurized Water Reactor (PWR) cores having different geometries with an unlimited number of FA types in the inventory. GARCO was developed by modifying the classical representation of the genotype. Both the genotype representation and the basic algorithm have been modified to incorporate the in-core fuel management heuristics rules so as to obtain the best results in a shorter time. GARCO has three modes. Mode 1 optimizes the locations of the fuel assemblies (FAs) in the nuclear reactor core, Mode 2 optimizes the placement of the burnable poisons (BPs) in a selected LP, and Mode 3 optimizes simultaneously both the LP and the BP placement in the core. This study describes the basic algorithm for Mode 1. The GARCO code is applied to the VVER-1000 reactor hexagonal geometry core in this study. The M oby-Dick i s used as reactor physics code to deplete FAs in the core. It was developed to analyze the VVER reactors by SKODA Inc. To use these rules for creating the initial population with GA operators, the worth definition application is developed. Each FA has a worth value for each location. This worth is between 0 and 1. If worth of any FA for a location is larger than 0.5, this FA in this location is a good choice. When creating the initial population of LPs, a subroutine provides a percent of individuals, which have genes with higher than the 0.5 worth. The percentage of the population to be created without using worth definition is defined in the GARCO input. And also age concept has been developed to accelerate the GA calculation process in reaching the

  8. Study on safety analysis of VVER-1200/V491 in scenario of Loss of Coolant Accidents along with partly failure of ECCS using RELAP5 code

    International Nuclear Information System (INIS)

    Hoang Minh Giang; Ha Thi Anh Dao; Hoang Tan Hung; Bui Thi Hoa; Nguyen Thi Tu Oanh; Dinh Anh Tuan; Pham Tuan Nam

    2017-01-01

    The advanced VVER-1200/V491 reactor designed with passive safety systems to deal with design extension conditions is primarily selected as priority candidate for Ninh Thuan 1 nuclear power plant project. So that, in order to enhance competence of nuclear safety and toward participation on review Safety Analysis Report (SAR) of Ninh Thuan nuclear Power project the study on safety analysis of VVER-1200/V491 in scenario of Loss of Coolant Accidents along with partly failure of ECCS is implemented. As requirement of the study, the input deck file of VVER-1200/V491 for RELAP5 and analysis report for some special case of LOCAs along with partly failure of ECCS are issued. (author)

  9. Generation of multigroup cross-sections from micro-group ones in code system SUHAM-U used for VVER-1000 reactor core calculations with MOX loading

    Energy Technology Data Exchange (ETDEWEB)

    Boyarinov, V.F.; Davidenko, V.D.; Polismakov, A.A.; Tsybulsky, V.F. [RRC Kurchatov Institute, Moscow (Russian Federation)

    2005-07-01

    At the present time, the new code system SUHAM-U for calculation of the neutron-physical processes in nuclear reactor core with triangular and square lattices based both on the modern micro-group (about 7000 groups) cross-sections library of code system UNK and on solving the multigroup (up to 89 groups) neutron transport equation by Surface Harmonics Method is elaborated. In this paper the procedure for generation of multigroup cross-sections from micro-group ones for calculation of VVER-1000 reactor core with MOX loading is described. The validation has consisted in computing VVER-1000 fuel assemblies with uranium and MOX fuel and has shown enough high accuracy under corresponding selection of the number and boundaries of the energy groups. This work has been fulfilled in the frame of ISTC project 'System Analyses of Nuclear Safety for VVER Reactors with MOX Fuels'.

  10. Fast Neutron Transport in the Biological Shielding Model and Other Regions of the VVER-1000 Mock-Up on the LR-0 Research Reactor

    Directory of Open Access Journals (Sweden)

    Košťál Michal

    2016-01-01

    Full Text Available A set of benchmark experiments was carried out in the full scale VVER-1000 mock-up on the reactor LR-0 in order to validate neutron transport calculation methodologies and to perform the optimization of the shape and locations of neutron flux operation monitors channels inside the shielding of the new VVER-1000 type reactors. Compared with previous experiments on the VVER-1000 mock-up on the reactor LR-0, the fast neutron spectra were measured in the extended neutron energy interval (0.1–10 MeV and new calculations were carried out with the MCNPX code using various nuclear data libraries (ENDF/B VII.0, JEFF 3.1, JENDL 3.3, JENDL 4, ROSFOND 2009, and CENDL 3.1. Measurements and calculations were carried out at different points in the mock-up. The calculation and experimental data are compared.

  11. Qualification of coupled 3D neutron kinetic/thermal hydraulic code systems by the calculation of a VVER-440 benchmark. Re-connection of an isolated loop

    Energy Technology Data Exchange (ETDEWEB)

    Kotsarev, Alexander; Lizorkin, Mikhail [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation); Bencik, Marek; Hadek, Jan [UJV Rez, a.s., Rez (Czech Republic); Kozmenkov, Yaroslav; Kliem, Soeren [Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany)

    2016-09-15

    The 7th AER dynamic benchmark is a continuation of the efforts to validate the codes systematically for the estimation of the transient behavior of VVER type nuclear power plants. The main part of the benchmark is the simulation of the re-connection of an isolated circulation loop with low temperature in a VVER-440 plant. This benchmark was calculated by the National Research Centre ''Kurchatov Institute'' (with the code ATHLET/BIPR-VVER), UJV Rez (with the code RELAP5-3D {sup copyright}) and HZDR (with the code DYN3D/ATHLET). The paper gives an overview of the behavior of the main thermal hydraulic and neutron kinetic parameters in the provided solutions.

  12. The design of PSB-VVER experiments carried-out inside the TACIS contract N. 30303

    International Nuclear Information System (INIS)

    Del Nevo, A.; D'Auria, F.; Mazzini, M.; Bykov, M.; Elkin, I.V.; Suslov, A.

    2007-01-01

    Integral Test Facility (ITF) experimental programs are relevant for validating the Best Estimate (BE) Thermal Hydraulic codes (TH) used for transient and accident analyses, design of Accident Management (AM) procedures, licensing of Nuclear Power Plants (NPP), etc. The validation process is based on well designed experiments. It consists in the comparison of the measured and calculated parameters and the determination whether a computer code has an adequate capability in predicting the major phenomena expected to occur for transient and/or accidents. University of Pisa (UNIPI) was responsible of the numerical design of the 12 experiments executed in PSB-VVER facility, operated at Electrogorsk Research and Engineering Center (EREC), in the framework of the TACIS Contract 3.03.03 Part A. This paper describes the methodology adopted at UNIPI, starting form the scenarios foreseen in the final Test Matrix (TM) until the execution of the experiments. This process considers three key topics: a) the scaling issue and the simulation, with unavoidable distortions, of the expected performance of the reference NPP, b) the code assessment process involving the identification of phenomena challenging the code models, c) the features of the concerned ITF (scaling limitations, control logics, data acquisition system, instrumentation, etc.). An overview of all the activities performed in this respect is provided focusing the discussion on the relevance of the heat losses. This issue is particularly relevant for addressing the scaling approach related to the power and volume of the facility. (author)

  13. PCA-based ANN approach to leak classification in the main pipes of VVER-1000

    International Nuclear Information System (INIS)

    Hadad, Kamal; Jabbari, Masoud; Tabadar, Z.; Hashemi-Tilehnoee, Mehdi

    2012-01-01

    This paper presents a neural network based fault diagnosing approach which allows dynamic crack and leaks fault identification. The method utilizes the Principal Component Analysis (PCA) technique to reduce the problem dimension. Such a dimension reduction approach leads to faster diagnosing and allows a better graphic presentation of the results. To show the effectiveness of the proposed approach, two methodologies are used to train the neural network (NN). At first, a training matrix composed of 14 variables is used to train a Multilayer Perceptron neural network (MLP) with Resilient Backpropagation (RBP) algorithm. Employing the proposed method, a more accurate and simpler network is designed where the input size is reduced from 14 to 6 variables for training the NN. In short, the application of PCA highly reduces the network topology and allows employing more efficient training algorithms. The accuracy, generalization ability, and reliability of the designed networks are verified using 10 simulated events data from a VVER-1000 simulation using DINAMIKA-97 code. Noise is added to the data to evaluate the robustness of the method and the method again shows to be effective and powerful. (orig.)

  14. Completion of the VVER 440/213 NPP Mochovce incorporation enhanced safety features

    International Nuclear Information System (INIS)

    Charbonneau, S.; Eckert, G.

    1996-01-01

    The cooperation between the western countries and the countries of ex-eastern block in the field of nuclear safety is recent and still limited. The main reasons for this situation are limited or non existent capabilities of these countries for financing as well as non acceptable legal conditions concerning the third party nuclear liability in this part of Europe. Nevertheless, Framatome and Siemens associated in the consortium named EUCOM, have signed in April 1996 the contract of about 100 million US dollars with Slovak electricity company (SLOVENSKE ELEKTRARNE-SE) for upgrading the Units 1 and 2 of Mochovce Nuclear Power Plant according to the western safety standards. This is the first important project involving west-european companies in the modernisation of Russian type of pressurized water reactor (VVER 440/213). The consortium will cooperate with other partners involved in the project: Slovak, Czech and Russian. The financing of the project will be provided mainly form Slovak and Czech sources. The safety upgrading will be financed through French and German buyer credits. French company Electricite de France (EDF) will be the consultant for SE. The safety upgrading measures have been elaborated taking into account the recommendation of Vienna International Atomic Energy Agency (IAEA) and the evaluation of the safety realised by RISKAUDIT, the common organization of German and French safety authorities (GSR and IPSN). Hence all guaranties have been taken to fulfil the western safety criteria for Nuclear Power Plant Mochovce. (author)

  15. Fire protection upgrading of four Russian 440/230 VVER units

    International Nuclear Information System (INIS)

    Corsini, G.; Yelfimov, S.

    1995-01-01

    The main goal of TACIS 3.6 a project funded by the Commission of the European Communities (CEC), was the front-end engineering for upgrading the Fire Protection System (FPS) of the safety-related equipment of Novovoronezh, Units 3 and 4, and Kola, Units 1 and 2, VVER 440/230 nuclear power plants. As a first step, all the safety-related equipment had to be identified, evaluation criteria had to be established and the existing FPS reviewed against the criteria. In the second step, the selection of the upgrading measures, depending on feasibility and cost estimate, has been accomplished, room by room. The third step, carried out on schedule and completed end July 95, has been essentially the preparation of the Technical Specifications for procurement of the needed equipment including remaining detail engineering. The Russian sub-contactor Atom Energo Project (AEP), who have been the designers of these older NPP s, have done the work with the Italian Ansaldo as the consultants of their Russian colleagues. Practical aspects of the engineering work are discussed and examples of improvements selected for retrofitting described. (author)

  16. Design issues concerning Iran's Bushehr nuclear power plant VVER-1000 conversion

    International Nuclear Information System (INIS)

    Carson, C.F.

    1996-01-01

    On January 8, 1995, the Atomic Energy Organization of Iran (AEOI) signed a contract for $800 million with the Russian Federation Ministry for Atomic Energy (Minatom) to complete Bushehr nuclear power plant (BNPP) unit 1. The agreement called for a Russian VVER-1000/320 pressurized water reactor (PWR) to be successfully installed into the existing German-built BNPP facilities in 5 yr. System design differences, bomb damage, and environmental exposure are key issues with which Minatom must contend in order to fulfill the contract. The AEOI under the Shah of Iran envisioned Bushehr as the first of many nuclear power plants, with Iran achieving 24 GW(electric) by 1993 and 34 GW(electric) by 2000. Kraftwerk Union AG (KWU) began construction of the two-unit plant near the Persian Gulf town of Halileh in 1975. Unit 1 was ∼80% complete and unit 2 was ∼50% complete when construction was interrupted by the 1979 Iranian Islamic revolution. Despite repeated AEOI attempts to lure KWU and other companies back to Iran to complete the plant, Western concerns about nuclear proliferation in Iran and repeated bombings of the plant during the 1980-1988 Iran-Iraq war dissuaded Germany from resuming construction

  17. Possibility of implementation of 6-year fuel cycle at NPP with VVER-440 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Heraltova, L., E-mail: lenka.heraltova@fjfi.cvut.cz [UJV Rez a.s., Hlavni 130, 250 68 Husinec-Rez (Czech Republic); Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, Brehova 7, 115 19 Praha 1 (Czech Republic)

    2015-12-15

    Highlights: • Possibility of extension of fuel cycle. • Increase of enrichment above 5% {sup 235}U. • Core properties calculated by diffusion code ANDREA. • Back end fuel cycle characteristic. - Abstract: This paper discusses possibility of an extension of a fuel cycle at a VVER-440 reactor for up to 6 years. The prolongation of a fuel cycle was realized by optimization of a fuel design and increasing of a fuel enrichment. The modified design of the fuel assembly covers change of pellet geometry, decreasing of parasitic absorption in construction materials, improved moderation of fuel pins and also increase of enrichment. Fuel assemblies with enrichment up to 7% {sup 235}U are considered for prolonged fuel batches. Three different batch lengths were considered for evaluation of core properties – 12, 18 and 24 months, and two types of burnable absorbers were included – Gd{sub 2}O{sub 3} and Er{sub 2}O{sub 3}. Comparison of proposed fuel assemblies was realized by length of a batch, average burnup, maximal power of fuel assembly or fuel pin, control fuel assembly worth, reactivity coefficients, and effective delayed neutrons fraction. Comparison of characteristics of a burned fuel discharged from a reactor core is discussed in the last part of the paper.

  18. Burnup influence on the VVER-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 the 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 VVER-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 1/4 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)

  19. The role of PRA in the safety assessment of VVER Nuclear Power Plants in Ukraine

    International Nuclear Information System (INIS)

    Kot, C.

    1999-01-01

    Ukraine operates thirteen (13) Soviet-designed pressurized water reactors, VVERS. All Ukrainian plants are currently operating with annually renewable permits until they update their safety analysis reports (SARs), in accordance with new SAR content requirements issued in September 1995, by the Nuclear Regulatory Authority and the Government Nuclear Power Coordinating Committee of Ukraine. The requirements are in three major areas: design basis accident (DBA) analysis, probabilistic risk assessment (PRA), and beyond design-basis accident (BDBA) analysis. The last two requirements, on PRA and BDBA, are new, and the DBA requirements are an expanded version of the older SAR requirements. The US Department of Energy (USDOE), as part of its Soviet-Designed Reactor Safety activities, is providing assistance and technology transfer to Ukraine to support their nuclear power plants (NPPs) in developing a Western-type technical basis for the new SARs. USDOE sponsored In-Depth Safety Assessments (ISAs) are in progress at three pilot nuclear reactor units in Ukraine, South Ukraine Unit 1, Zaporizhzhya Unit 5, and Rivne Unit 1, and a follow-on study has been initiated at Khmenytskyy Unit 1. The ISA projects encompass most areas of plant safety evaluation, but the initial emphasis is on performing a detailed, plant-specific Level 1 Internal Events PRA. This allows the early definition of the plant risk profile, the identification of risk significant accident sequences and plant vulnerabilities and provides guidance for the remainder of the safety assessments

  20. Severe damage analysis of VVER 1000 following large break LOCA using Astec code

    International Nuclear Information System (INIS)

    Chatterjee, B.; Mukhopadhyay, D.; Lele, H.G.; Ghosh, A.K.; Kushwaha, H.S.

    2007-01-01

    Severe accident analysis of a reactor is an important aspect in the evaluation of source term. This in turn helps in emergency planning. An analysis has been carried out for VVER-1000 (V320) reactor following Large Break LOCA (loss of coolant accident) along with Station Blackout (SBO). Computer code ASTEC (jointly developed by IRSN, France, and GRS, Germany) is used for analyzing the transient. This integral code has been designed to be used as reference code for PSA2 studies. Severe accident analysis is carried out for an accident initiated by Large break LOCA along with SBO. Two cases have been analysed with the version ASTEC V1.2-rev1. In the first case hydro-accumulators are considered not available while the second case has been analysed with hydro accumulators. In this paper, ASTEC predictions have been studied for the in-vessel phase of the accident till vessel failure. The vessel failure was observed at 6979 s when accumulators were assumed not available. The vessel failure was quite delayed (19294 s) with operating accumulators. The hydrogen production was found to be very large (22% of total Zr inventory) in the case with accumulators compared to the case without accumulators (1.5% of total Zr inventory)

  1. In situ electrochemical impedance spectroscopy of Zr-1%Nb under VVER primary circuit conditions

    International Nuclear Information System (INIS)

    Nagy, Gabor; Kerner, Zsolt; Pajkossy, Tamas

    2002-01-01

    Oxide layers were grown on tubular samples of Zr-1%Nb under conditions simulating those in VVER-type pressurised water reactors, viz. in near-neutral borate solutions in an autoclave at 290 deg. C. These samples were investigated using electrochemical impedance spectroscopy which was found to be suitable to follow in situ the corrosion process. A -CPE ox parallel R ox - element was used to characterise the oxide layer on Zr-1%Nb. Both the CPE ox coefficient, σ ox , and the parallel resistance, R ox , were found to be thickness dependent. The layer thickness, however, can only be calculated after a calibration procedure. The temperature dependence of the CPE ox element was also found to be anomalous while the temperature dependence of R ox indicates that the oxide layer has semiconductor properties. The relaxation time - defined as (R ox σ ox ) 1/α - was found to be quasi-independent of oxidation time and temperature; thus it is characteristic to the oxide layer on Zr-1%Nb

  2. Economical aspects of multiple plutonium and uranium recycling in VVER reactors

    Energy Technology Data Exchange (ETDEWEB)

    Alekseev, P.N.; Bobrov, E.A.; Dudnikov, A.A.; Teplov, P.S. [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation)

    2016-09-15

    The basic strategy of Russian Nuclear Energy development is the formation of the closed fuel cycle based on fast breeder and thermal reactors, as well as the solution of problems of spent nuclear fuel accumulation and availability of resources. Three options of multiple Pu and U recycling in VVER reactors are considered in this work. Comparison of MOX and REMIX fuel recycling approaches for the closed fuel cycle involving thermal reactors is presented. REMIX fuel is supposed to be fabricated from non-separated mixture of uranium and plutonium obtained in spent fuel reprocessing with further makeup by enriched U. These options make it possible to recycle several times the total amount of Pu and U obtained from spent fuel. The main difference is the full or partial fuel loading of the core by assemblies with recycled Pu. The third option presents the concept of heterogeneous arrangement of fuel pins made of enriched uranium and MOX in one fuel assembly. It should be noted that fabrication of all fuel assemblies with Pu requires the use of expensive manufacturing technology. These three options of core loading can be balanced with respect to maximum Pu and U involvement in the fuel cycle. Various physical and economical aspects of Pu and U multiple recycling for selected options are considered in this work.

  3. Feasibility and usefulness of reconstructing obsolete power blocks of VVER-440 reactors

    International Nuclear Information System (INIS)

    Kirichenko, A.M.; Krushenik, S.D.; Sigal, M.V.; Kustov, V.P.

    1993-01-01

    At the present time, in Russia and in the East European countries there are atomic power stations with first-generation VVER-440 reactors built according to specification which no longer satisfy the more rigorous modern safety standards. Among these power stations are, in particular, the Novovoronezh and the Armenian Atomic Power Station and two blocks of the Kola Atomic Power Station. The search for technical solutions for modernizing these power blocks is complicated because two conditions which are hard to reconcile must be fulfilled: an acceptable safety level must be obtained and the rebuilding must be economically justifiable (particularly since the time of operation of a power block until its standard service life is over is short). Research work undertaken in the All-Union Scientific Research Institute of Atomic Power Stations has shown that one way of overcoming these difficulties may involve changing the operating conditions of the reactor assembly to a less demanding mode of operation. This solution implies an economically justified minimum of structural improvements, provides the required safety level, and prolongs the service life of the power block. The reduction of the thermal power, and consequently, the necessary transfer of a power block to another option

  4. Posttest examination of the VVER-1000 fuel rod bundle CORA-W2

    International Nuclear Information System (INIS)

    Sepold, L.

    1995-06-01

    The bundle meltdown experiment CORA-W2, representing the behavior of a Russian type VVER-1000 fuel element, with one B 4 C/stainless steel absorber rod was selected by the OECD/CSNI as International Standard Problem (ISP-36). The experimental results of CORA-W2 serve as data base for comparison with analytical predictions of the high-temperature material behavior by various code systems. The first part of the experimental results is described in KfK 5363 (1994), the second part is documented in this report which contains the destructive post-test examination results. The metallographical and analytical (SEM/EDX) post-test examinations were performed in Germany and Russia and are summarized in five individual contributions. The upper half of the bundle is completely oxidized, the lower half has kept the fuel rods relatively intact. The post-test examination results show the strong impact of the B 4 C absorber rod and the stainless steel grid spacers on the ''low-temperature'' bundle damage initiation and progression. The B 4 C absorber rod completely disappeared in the upper half of the bundle. The multicomponent melts relocated and formed coolant channel blockages on solidification with a maximum extent of about 30% in the lower part of the bundle. At temperatures above the melting point of the ZrNb1 cladding extensive fuel dissolution occurred. (orig.) [de

  5. A new optimization method based on cellular automata for VVER-1000 nuclear reactor loading pattern

    International Nuclear Information System (INIS)

    Fadaei, Amir Hosein; Setayeshi, Saeed

    2009-01-01

    This paper presents a new and innovative optimization technique, which uses cellular automata for solving multi-objective optimization problems. Due to its ability in simulating the local information while taking neighboring effects into account, the cellular automata technique is a powerful tool for optimization. The fuel-loading pattern in nuclear reactor cores is a major optimization problem. Due to the immensity of the search space in fuel management optimization problems, finding the optimum solution requires a huge amount of calculations in the classical method. The cellular automata models, based on local information, can reduce the computations significantly. In this study, reducing the power peaking factor, while increasing the initial excess reactivity inside the reactor core of VVER-1000, which are two apparently contradictory objectives, are considered as the objective functions. The result is an optimum configuration, which is in agreement with the pattern proposed by the designer. In order to gain confidence in the reliability of this method, the aforementioned problem was also solved using neural network and simulated annealing, and the results and procedures were compared.

  6. Living PSA program for VVER 440/213 in the Czech Republic

    International Nuclear Information System (INIS)

    Husak, S.; Patrik, M.

    2000-01-01

    The paper presents an overview of a Living PSA concept in the Czech Republic for the VVER 440/213 NPP Dukovany unit. The first step of PSA program was a Level 1 basic study for Unit No. 1 which was completed in 1995. The main objective of the study was to determine the risk level of full power operation and its contributors as well as to reveal the weak points of the plant. Living PSA program for a Level 1 study has been afterwards established as a framework for all activities related to risk assessment and risk based decision-making support in NPP Dukovany. The basic parts of the project are: a management of PSA models and studies to implement design and procedures, modifications or new data inputs from data collection; continuous improvement based of new analyses, experiments or more detailed models; an extensions of the scope (external events, all plant operating modes, other sources of radioactive releases). The Living PSA program in NPP Dukovany provides basis for three kinds of PSA activities: risk assessment applications, risk monitoring and risk assessment of operational. (author)

  7. Ambition to reach zero level failure in VVER 1000 with russian fuel

    International Nuclear Information System (INIS)

    Mečíř, V.

    2015-01-01

    The purpose of “The Zero Failure Level Project” is to bring to real operation of VVER 1000 units the dream of all utilities such as problem free and cost effective operation. This essentially turns into requirement on failure free fuel operation. At the same time the general requirements such as safety, cost effectiveness, operational flexibility, fuel cycle and fuel flexibility need to be satisfied. Several specific tasks were performed and many of them are still in process. Specific failure tree was developed in a format, which allows step by step failure tree improvement. Fuel types and its modifications, taking into account manufacturing conditions, were specified. In parallel with fuel types classification, real operational conditions were evaluated based on approximately 280 parameters by fuel assembly design features, operational procedures and practices and about 250 reactor unit parameters. As a result of this stage, groups of units with similar fuel operational conditions should be revealed and experience sharing database created. It is also recognized a need for consistent methods of operational data and data from pool side fuel assembly inspection. In the area of Foreign Material Exclusion activities closer cooperation between utility and supplier should be established including foreign material classification and improvement in root cause investigation

  8. Aspects of using a best-estimate approach for VVER safety analysis in reactivity initiated accidents

    Energy Technology Data Exchange (ETDEWEB)

    Ovdiienko, Iurii; Bilodid, Yevgen; Ieremenko, Maksym [State Scientific and Technical Centre on Nuclear and Radiation, Safety (SSTC N and RS), Kyiv (Ukraine); Loetsch, Thomas [TUEV SUED Industrie Service GmbH, Energie und Systeme, Muenchen (Germany)

    2016-09-15

    At present time, Ukraine faces the problem of small margins of acceptance criteria in connection with the implementation of a conservative approach for safety evaluations. The problem is particularly topical conducting feasibility analysis of power up-rating for Ukrainian nuclear power plants. Such situation requires the implementation of a best-estimate approach on the basis of an uncertainty analysis. For some kind of accidents, such as loss-of-coolant accident (LOCA), the best estimate approach is, more or less, developed and established. However, for reactivity initiated accident (RIA) analysis an application of best estimate method could be problematical. A regulatory document in Ukraine defines a nomenclature of neutronics calculations and so called ''generic safety parameters'' which should be used as boundary conditions for all VVER-1000 (V-320) reactors in RIA analysis. In this paper the ideas of uncertainty evaluations of generic safety parameters in RIA analysis in connection with the use of the 3D neutron kinetic code DYN3D and the GRS SUSA approach are presented.

  9. Determination of power density in VVER-1000 Mock-Up in LR-0 reactor

    Directory of Open Access Journals (Sweden)

    Košál Michal

    2017-01-01

    Full Text Available The pin power density is an important quantity which has to be monitored during the reactor operation, for two main reasons. Firstly, it is part of the limits and conditions of safe operation and, secondly, it is source term in neutron transport calculations used for the adequate assessing of the state of core structures and pressure vessel material. It is often calculated using deterministic codes which may have problems with an adequate definition of boundary conditions in subcritical regions. This may lead to overestimation of real situation, and therefore the validation of the utility codes contributes not only to better fuel utilization, but also to more precise description of radiation situation in structural components of core. Current paper presents methods developed at LR-0 reactor, as well as selected results for pin power density measurement in peripheral regions of VVER-1000 mock-up. The presented data show that the results of a utility diffusion code at core boundary overestimate the measurement. This situation, however satisfactory safe, may lead to unduly conservative approach in the determination of radiation damage of core structures.

  10. Coupled neutronic core and subchannel analysis of nanofluids in VVER-1000 type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zarifi, Ehsan; Sepanloo, Kamran [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Reactor and Nuclear Safety School; Jahanfarnia, Golamreza [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering, Science and Research Branch

    2017-05-15

    This study is aimed to perform the coupled thermal-hydraulic/neutronic analysis of nanofluids as the coolant in the hot fuel assembly of VVER-1000 reactor core. Water-based nanofluid containing various volume fractions of Al{sub 2}O{sub 3} nanoparticle is analyzed. WIMS and CITATION codes are used for neutronic simulation of the reactor core, calculating neutron flux and thermal power distribution. In the thermal-hydraulic modeling, the porous media approach is used to analyze the thermal behavior of the reactor core and the subchannel analysis is used to calculate the hottest fuel assembly thermal-hydraulic parameters. The derived conservation equations for coolant and conduction heat transfer equation for fuel and clad are discretized by Finite volume method and solved numerically using visual FORTRAN program. Finally the analysis results for nanofluids and pure water are compared together. The achieved results show that at low concentration (0.1 percent volume fraction) alumina is the optimum nanoparticles for normal reactor operation.

  11. Neutronics feasibility of using Gd2O3 particles in VVER-1000 fuel assembly

    International Nuclear Information System (INIS)

    Hoang Van Khanh; Hoang Thanh Phi Hung; Tran Hoai Nam

    2016-01-01

    Neutronics feasibility of using Gd 2 O 3 particles for controlling excess reactivity of VVER-1000 fuel assembly has been investigated. The motivation is that the use of Gd 2 O 3 particles would increase the thermal conductivity of the UO 2 +Gd 2 O 3 fuel pellet which is one of the desirable characteristics for designing future high burnup fuel. The calculation results show that the Gd 2 O 3 particles with the diameter of 60 µm could control the reactivity similarly to that of homogeneous mixture with the same amount of Gd 2 O 3 . The power densities at the fuel pin with Gd 2 O 3 particles increase by about 10-11%, leading to the decrease of the power peak and a slightly flatter power distribution. The power peak appears at the periphery pins at the beginning of burnup process which is decreased by 0.9 % when using Gd 2 O 3 particles. Further work and improvement are being planned to optimize the high power peaking at the beginning of burnup. (author)

  12. The most extensive reconstruction of nuclear power plant with VVER 440/V230 reactor

    International Nuclear Information System (INIS)

    Ferenc, M.

    2000-01-01

    The nuclear power plant V-1 Bohunice consists of two VVER-440 units with V-230 reactors. Unit 1 was commissioned in 1978 and Unit 2 in 1980. Large experience and knowledge from the operation of previous units with V-230 reactors were incorporated into the V-1 design, which resulted in a higher level of safety and operational reliability of these units. The Siemens company which won an international bidding process developed these basic goals for the Gradual Upgrading into the so called Basic Engineering (BE). For the implementation of the Gradual Upgrading in line with the BE, Rekon consortium was established consisting of Siemens and VUJE. The implementation of the Gradual Upgrading is scheduled for the time period of 1996 - 2000. Siemens was responsible for the upgrading strategy - based on the approved results of the basic engineering phase and the PSAR, the engineering and realization of all I and C improvements, and also for the seismic upgrade. VUJE's responsibility covered the detailed engineering and implementation of mechanical, electrical and civil part of upgrading measures as well as overall organisation and evaluation of verification tests. The consortium awarded contracts for final planning and design, installation services and commissioning to other Slovakian subcontractors in order to ensure the largest possible local content. The gradual reconstruction of the V-1 Bohunice with V230 reactors represents a comprehensive reconstruction of safety-related systems and equipment. Following its completion, the units will be operated with a safety level accepted internationally. (author)

  13. In-core nuclear fuel management optimization of VVER1000 using perturbation theory

    International Nuclear Information System (INIS)

    Hosseini, Mohammad; Vosoughi, Naser

    2011-01-01

    In-core nuclear fuel management is one of the most important concerns in the design of nuclear reactors. The two main goals in core fuel loading pattern design optimization are maximizing the core effective multiplication factor in order to extract the maximum energy, and keeping the local power peaking factor lower than a predetermined value to maintain fuel integrity. Because of the numerous possible patterns of the fuel assemblies in the reactor core, finding the best configuration is so important and complex. Different methods for optimization of fuel loading pattern in the core have been introduced so far. In this study, a software is programmed in C ⧣ language to find an order of the fuel loading pattern of the VVER-1000 reactor core using the perturbation theory. Our optimization method is based on minimizing the radial power peaking factor. The optimization process lunches by considering the initial loading pattern and the specifications of the fuel assemblies which are given as the input of the software. It shall be noticed that the designed algorithm is performed by just shuffling the fuel assemblies. The obtained results by employing the mentioned method on a typical reactor reveal that this method has a high precision in achieving a pattern with an allowable radial power peaking factor. (author)

  14. Evaluating the consequences of loss of flow accident for a typical VVER-1000 nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mirvakili, S.M.; Safaei, S. [Shiraz Univ., Shiraz (Iran, Islamic Republic of). Dept. of Nuclear Engineering, School of Mechanical Engineering; Faghihi, F. [Shiraz Univ., Shiraz (Iran, Islamic Republic of). Safety Research Center

    2010-07-01

    The loss of coolant flow in a nuclear reactor can result from a mechanical or electrical failure of the coolant pump. If the reactor is not tripped promptly, the immediate effect is a rapid increase in coolant temperature, decrease in minimum departure from nucleate boiling ratio (DNBR) and fuel damage. This study evaluated the shaft seizure of a reactor coolant pump in a VVER-1000 nuclear reactor. The locked rotor results in rapid reduction of flow through the affected reactor coolant loop and in turn leads to an increase in the primary coolant temperature and pressure. The analysis was conducted with regard for superimposing loss of power to the power plant at the initial accident moment. The required transient functions of flow, pressure and power were obtained using system transient calculations applied in COBRA-EN computer code in order to calculate the overall core thermal-hydraulic parameters such as temperature, critical heat flux and DNBR. The study showed that the critical period for the locked rotor accident is the first few seconds during which the maximum values of pressure and temperature are reached. 10 refs., 1 tab., 3 figs.

  15. Structural Integrity Assessment of VVER-1000 RPV under Accidental Cool down Transients

    International Nuclear Information System (INIS)

    Shrivastav, V.; Sen, R.N.; Yadav, R.S.

    2012-01-01

    Corrosion, Fatigue and Irradiation embrittlement are the major degradation mechanisms responsible for ageing of RPV (and its internals) of a Pressurized Water Reactor. While corrosion and fatigue can generate cracks, irradiation damage can lead to brittle fracture initiating from these cracks. Ageing in nuclear power plants needs to be managed so as to ensure that design functions remain available throughout the life of the plant. From safety perspective, this implies that ageing degradation of systems, structures and components important to safety remain within acceptable limits. Reactor Pressure Vessel has been identified as the highest priority key component in plant life management for Pressurized Water Reactors. Therefore special attention is required to ensure its structural integrity during its lifetime. In this paper, structural integrity assessment for typical VVER-1000 RPV is carried out under severe accidental cool down transients using the Finite Element Method. Three different accidental scenarios are postulated and safety of the vessel is conservatively assessed under these transients using the Linear Elastic Fracture Mechanics approach. Transient thermo mechanical stress analysis of the core belt region of the RPV is carried out in presence of postulated cracks and stress intensity factors are calculated and compared with the material fracture toughness to assess the structural integrity of the vessel. The paper also include some parametric analyses to justify the methodology. (author)

  16. Pressure loadings of Soviet-designed VVER [Water-Cooled, Water-Moderated Energy Reactor] reactor release mitigation structures from large-break LOCAs

    International Nuclear Information System (INIS)

    Sienicki, J.J.; Horak, W.C.

    1989-01-01

    Analyses have been carried out of the pressurization of the accident release mitigation structures of Soviet-designed VVER (Water-Cooled, Water-Moderated Energy Reactor) pressurized water reactors following large-break loss-of-coolant accidents. Specific VVER systems for which calculations were performed are the VVER-440 model V230, VVER-440 model V213, and VVER-1000 model V320. Descriptions of the designs of these and other VVER models are contained in the report DOE/NE-0084. The principal objective of the current analyses is to calculate the time dependent pressure loadings inside the accident localization or containment structures immediately following the double-ended guillotine rupture of a primary coolant pipe. In addition, the pressures are compared with the results of calculations of the response of the structures to overpressure. Primary coolant system thermal hydraulic conditions and the fluid conditions at the break location were calculated with the RETRAN-02 Mod2 computer code (Agee, 1984). Pressures and temperatures inside the building accident release mitigation structures were obtained from the PACER (Pressurization Accompanying Coolant Escape from Ruptures) multicompartment containment analysis code developed at Argonne National Laboratory. The analyses were carried out using best estimate models and conditions rather than conservative, bounding-type assumptions. In particular, condensation upon structure and equipment was calculated using correlations based upon analyses of the HDR, Marviken, and Battelle Frankfurt containment loading experiments. The intercompartment flow rates incorporate an effective discharge coefficient and liquid droplet carryover fraction given by expressions of Schwan determined from analyses of the Battelle Frankfurt and Marviken tests. 5 refs., 4 figs

  17. Developing a computational tool for predicting physical parameters of a typical VVER-1000 core based on artificial neural network

    International Nuclear Information System (INIS)

    Mirvakili, S.M.; Faghihi, F.; Khalafi, H.

    2012-01-01

    Highlights: ► Thermal–hydraulics parameters of a VVER-1000 core based on neural network (ANN), are carried out. ► Required data for ANN training are found based on modified COBRA-EN code and then linked each other using MATLAB software. ► Based on ANN method, average and maximum temperature of fuel and clad as well as MDNBR of each FA are predicted. -- Abstract: The main goal of the present article is to design a computational tool to predict physical parameters of the VVER-1000 nuclear reactor core based on artificial neural network (ANN), taking into account a detailed physical model of the fuel rods and coolant channels in a fuel assembly. Predictions of thermal characteristics of fuel, clad and coolant are performed using cascade feed forward ANN based on linear fission power distribution and power peaking factors of FAs and hot channels factors (which are found based on our previous neutronic calculations). A software package has been developed to prepare the required data for ANN training which applies a modified COBRA-EN code for sub-channel analysis and links the codes using the MATLAB software. Based on the current estimation system, five main core TH parameters are predicted, which include the average and maximum temperatures of fuel and clad as well as the minimum departure from nucleate boiling ratio (MDNBR) for each FA. To get the best conditions for the considered ANNs training, a comprehensive sensitivity study has been performed to examine the effects of variation of hidden neurons, hidden layers, transfer functions, and the learning algorithms on the training and simulation results. Performance evaluation results show that the developed ANN can be trained to estimate the core TH parameters of a typical VVER-1000 reactor quickly without loss of accuracy.

  18. Duckweed does not improve the efficiency of municipal wastewater treatment in lemna system plants / Rzęsa nie poprawia efektywności oczyszczania ścieków komunalnych w oczyszczalniach typu Lemna System

    Directory of Open Access Journals (Sweden)

    Ozimek Teresa

    2015-09-01

    Full Text Available Badania prowadzono w trzech czyszczalniach hydrofi towych z powierzchniowym przepływem ścieków typu Lemna System, zbudowanych według projektu Lemna Corporation. Składają się one ze stawu napowietrzanego i stawu rzęsowego wyposażonego w system barier mających na celu równomierne rozmieszczenie roślin. Stawy poprzedzone są osadnikiem wstępnym. Według projektantów odpowiednia praca tych oczyszczalni związana jest z występowaniem rzęsy na powierzchni stawu rzęsowego. Badano efektywność oczyszczania ścieków komunalnych w trzech oczyszczalniach Lemna System usytuowanych w centralnej Polsce ze szczególnym uwzględnieniem roli stawu rzęsowego i samej rzęsy. Oczyszczalnie różniły się między sobą występowaniem rzęsy. W dwóch z nich (w Rakowie i Bąkowcu w trakcie całego okresu badań rzęsa nie występowała, natomiast w trzeciej (w Falęcinie Starym w okresie wegetacyjnym pokrywała do 90%.powierzchni stawu. Efektywność oczyszczanie ścieków nie różniła się w oczyszczalni z rzęsą i bez rzęsy. Rzęsa nie wpływała istotnie na efektywność oczyszczania ścieków. We wszystkich oczyszczalniach główną rolę w redukcji zanieczyszczeń odgrywał staw napowietrzany

  19. OECD/DOE/CEA VVER-1000 Coolant Transient Benchmark. Summary Record of the Third Workshop (V1000-CT3)

    International Nuclear Information System (INIS)

    2005-01-01

    The overall objective of the VVER-1000 coolant transient (V1000CT) benchmark is to assess computer codes used in the safety analysis of VVER power plants, specifically for their use in analysis of reactivity transients in a VVER-1000. The V1000CT benchmark consists of two phases: V1000CT-1 is a simulation of the switching on of one main coolant pump (MCP) when the other three MCPs are in operation, and V1000CT-2 concerns calculation of coolant mixing tests and main steam line break (MSLB) scenarios. Each of the two phases contains three exercises. The reference problem chosen for simulation in Phase 1 is a MCP switching on when the other three main coolant pumps are in operation in a VVER-1000. This event is characterized by rapid increase in the flow through the core resulting in a coolant temperature decrease, which is spatially dependent. This leads to insertion of spatially distributed positive reactivity due to the modelled feedback mechanisms and non-symmetric power distribution. Simulation of the transient requires evaluation of core response from a multi-dimensional perspective (coupled three-dimensional neutronics/core thermal-hydraulics) supplemented by a one-dimensional simulation of the remainder of the reactor coolant system. Three exercises are defined in the framework of Phase 1: a) Exercise 1 - Point kinetics plant simulation; b) Exercise 2 - Coupled 3-D neutronics/core thermal-hydraulics response evaluation; c) Exercise 3 - Best-estimate coupled 3-D core/plant system transient modelling. In addition to the measured (experiment) scenario, extreme calculation scenarios were defined in the frame of Exercise 3 for better testing 3-D neutronics/thermal-hydraulics techniques. The proposals concerned: rod ejection simulations with scram set points at two different power levels. The technical topics presented at this workshop were: Review of the benchmark activities after the 2. Workshop; - Discussion of participant's feedback and introduced modifications

  20. Experimental support of the bleed and feed accident management measures for VVER-440/213 type reactors

    International Nuclear Information System (INIS)

    Szabados, L.

    2002-01-01

    In the original design of the VVER-440/213 type nuclear power plants event oriented emergency operating procedures (EOP) were implemented. In the last years, however, new symptom based procedures of Westinghouse-type have been developed and partly implemented for the plants in Central Europe including the Paks Nuclear Power Plant. Paper gives a short review of the experiments performed in the PMK-2 facility to study the effectiveness of the bleed and feed strategies and to get experimental data bases for the validation of thermohydraulic system codes like RELAP5, ATHLET and CATHARE.(author)

  1. CFX-10 and RELAP5-3D simulations of coolant mixing phenomena in RPV of VVER-1000 reactors

    International Nuclear Information System (INIS)

    Terzuoli, F.; Moretti, F.; Melideo, D.; D'Auria, F.; Shkarupa, O.

    2006-01-01

    The present paper deals with numerical analyses of coolant mixing in the reactor pressure vessel of a VVER-1000 reactor, performed with the ANSYS CFX-10 CFD code and with the RELAP5-3D system code. In particular, the attention focused on the 'swirl' effect that has been observed to take place in the downcomer of such kind of reactor, with the aim of assessing the capability of the codes to predict that effect, and to understand the reasons for its occurrence. The results have been compared against experimental data from V1000CT-2 Benchmark. (author)

  2. Calculation of the source term for a S1B-sequence at a VVER-1000 type reactor. Part 1

    International Nuclear Information System (INIS)

    Sdouz, G.

    1990-10-01

    The behaviour of the source term in a VVER-1000 type reactor is calculated using the 'Source Term Code Package' (STCP). The input data are based on the russian plant Zaporozhye-5. The selected accident sequence is a small break LOCA in the hot leg followed by loss offsite and onsite electric power (S 1 B-sequence). According to the course of the calculation the results are presented and analyzed for each program. Except for the noble gases all release fractions are lower than 10 -4 . 18 refs., 10 tabs. (Author)

  3. Three-dimensional analysis of the coolant flow characteristics in the fuel assemblies of VVER-1000 reactors

    International Nuclear Information System (INIS)

    Dinh Van Thin; Tran Thi Nhan

    2015-01-01

    Computational Fluid Dynamics (CFD) is a widely used method around the world for complex flow and heat industrial problems. In this paper, the coolant flow parameters were investigated in subchannels of VVER-1000 reactor’s fuel assemblies by ANSYS V14.5 programme. The different mesh solutions and turbulence models were carried out to deal with the water flow problems such as velocity distribution, streamline, temperature and pressure change as well as the hydraulic resistances of the spacer grids. The obtained results are good agreement with the measured values and the published reports from other authors. (author)

  4. The influence of changes in the VVER-1000 fuel assembly shape during operation on the power density distribution

    Energy Technology Data Exchange (ETDEWEB)

    Shishkov, L. K., E-mail: Shishkov-LK@nrcki.ru; Gorodkov, S. S.; Mikailov, E. F.; Sukhino-Homenko, E. A.; Sumarokova, A. S., E-mail: Sumarokova-AS@nrcki.ru [National Research Center Kurchatov Institute (Russian Federation)

    2016-12-15

    A new approach to calculation of the coefficients of sensitivity of the fuel pin power to deviations in gap sizes between fuel assemblies of the VVER-1000 reactor during its operation is proposed. It is shown that the calculations by the MCU code should be performed for a full-size model of the core to take the interference of the gap influence into account. In order to reduce the conservatism of calculations, the coolant density and coolant temperature feedbacks should be taken into account, as well as the fuel burnup.

  5. Possible emission of radioactive fission products during off-design accidents at a nuclear power plant with VVER-1000 reactor

    International Nuclear Information System (INIS)

    Dubkov, A.P.; Kozlov, V.F.; Luzanova, L.M.

    1995-01-01

    It is well known that eight nuclear power plants with VVER-1000 reactors have been constructed in Russia, Ukraine, and in the Republic of Belarus and they have been operating successfully without any serious accidents since 1980. These facilities have been analyzed for various accident scenarios, and measures have been incorporated which will prevent core damage during these possible events. However, an off-design accident can occur, and in such a case, the radiological consequences would exceed the worst design accidents. This paper reviews a number of potential off-design accidents in order to develop an accident plan to mitigate the consequences of such an accident

  6. Experience in modernization of safety I and C in VVER 440 nuclear power plants Bohunice V1 and Paks

    International Nuclear Information System (INIS)

    Martin, M.

    2000-01-01

    For nuclear power plants which have been in operation for more than 15 years, backfitting or even complete replacement of the instrumentation and control (I and C) equipment becomes an increasingly attractive option, motivated not only by the objective to reduce the cost of I and C system maintenance and repair but also to prolong or at least to safeguard the plant life-time: optimized spare-part management through use of standard equipment; reduction of number and variety of different items of equipment by implementing functions stepwise in application software; adding new functionality in the application software which was not possible in the old technology due to lack of space; safeguarding of long-term After-Sales-Service. Some years ago Bohunice V1 NPP, Slovak Republic and Paks NPP, Hungary intended to replace parts of their Safety I and C, mainly the Reactor Trip System, the Reactor Limitation System and the Neutron Flux Excore Instrumentation and Monitoring Systems. After a Basic Engineering Phase in Bohunice V1 and a Feasibility Study in Paks both plants decided to use the Siemens' Digital Safety I and C System TELEPERM XS to modernize their plants. Both Bohunice, Unit 2 and Paks, Unit 1 have been back on line for over six months with the new Digital Safety I and C. At the present time Bohunice, Unit 1 and within the next few months Paks, Unit 2 will be replaced. Trouble-free start-ups and no major problems under operation in the first two plants were based on: thorough understanding of the VVER 440 technology; comprehensive planning together with the plant operators and authorities; the possibility to adapt TELEPERM XS to every plant type; the execution of extensive pre-operational tests. Regarding these modernization measures Siemens, as well as the above Operators, have gained considerable experience in the field of I and C Modernization in VVER 440 NPPs. Important aspects of this experience are: how to transfer the VVER technology to TELEPERM XS; how to

  7. 3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors

    Energy Technology Data Exchange (ETDEWEB)

    Langenbuch, S.; Velkov, K. [GRS, Garching (Germany); Lizorkin, M. [Kurchatov-Institute, Moscow (Russian Federation)] [and others

    1997-07-01

    This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed.

  8. Thermal-hydraulic modeling of nanofluids as the coolant in VVER-1000 reactor core by the porous media approach

    International Nuclear Information System (INIS)

    Jahanfarnia, G.; Zarifi, E.; Veysi, F.

    2013-01-01

    The aim of this study was to perform a thermal-hydraulic analysis of nanofluids as coolant in the Bushehr VVER-1000 reactor core using the porous media approach. Water-based nanofluids containing various volume fractions of Al 2 O 3 and TiO 2 nanoparticles were analyzed. The conservation equations were discretized by the finite volume method and solved by numerical methods. To validate the approaches applied in this study, the results of the model and COBRA-EN code were compared for pure water. The achieved results show that the temperature of the coolant increases with the concentration of the nanoparticles. (authors)

  9. CFD analysis of flow distribution of reactor core and temperature rise of coolant in fuel assembly for VVER reactor

    International Nuclear Information System (INIS)

    Du Daiquan; Zeng Xiaokang; Xiong Wanyu; Yang Xiaoqiang

    2015-01-01

    Flow field of VVER-1000 reactor core was investigated by using computational fluid dynamics code CFX, and the temperature rise of coolant in hot assembly was calculated. The results show that the maximum value of flow distribution factor is 1.12 and the minimum value is 0.92. The average value of flow distribution factor in hot assembly is 0.97. The temperature rise in hot assembly is higher than current warning limit value ΔT t under the deviated operation condition. The results can provide reference for setting ΔT t during the operation of nuclear power plant. (authors)

  10. Insights from the U.S. department of Energy plant safety evaluation program of VVER and RBMK reactors

    International Nuclear Information System (INIS)

    Petri, M.C.; Binder, J.L.; Pasedag, W.F.

    2001-01-01

    Throughout the years 1990 the U.S. Department of Energy has worked build capability in countries of the former Soviet Union to assess the safety of their VVER and RBMK reactors. Through this Plant Safety Evaluation Program, deterministic and probabilistic analyses have been used to provide a documented plant risk profile to support safe plant operation and to set priorities for safety upgrades. Work has been sponsored at thirteen nuclear power plant sites in eight countries. The Plant Safety Evaluation Program has resulted in immediate and long-term safety benefits for the Soviet-designed nuclear plants. (author)

  11. Several perspectives on water-chemical cycles for nuclear power stations equipped with type VVER and RBMK reactors

    International Nuclear Information System (INIS)

    Mamet, A.P.; Mamet, V.A.; Pashevich, V.I.; Nazarenko, P.N.

    1982-01-01

    Water-chemical cycles for loops I and II of VVER reactors are discussed. These cycles are mixed ammonia-sodium with a variable concentration of boric acid and ammonia hydrazine with a pH factor of 9.1 +/- 0.1. New water-chemical cycles are considered for use in both existing and new nuclear power plants. Application of these new water-chemical cycles showed produce a significant improvement in operating conditions of nuclear power plants. Upon accumulation of sufficient operating experience with these cycles, it should be possible to raise the issue of revising applicable standard documentation

  12. The design of PSB-VVER experiments relevant to accident management

    International Nuclear Information System (INIS)

    Del Nevo, Alessandro; D'auria, Francesco; Mazzini, Marino; Bykov, Michael; Elkin, Ilya V.; Suslov, Alexander

    2008-01-01

    Experimental programs carried-out in integral test facilities are relevant for validating the best estimate thermal-hydraulic codes, which are used for accident analyses, design of accident management procedures, licensing of nuclear power plants, etc. The validation process, in fact, is based on well designed experiments. It consists in the comparison of the measured and calculated parameters and the determination whether a computer code has an adequate capability in predicting the major phenomena expected to occur in the course of transient and/or accidents. University of Pisa was responsible of the numerical design of the 12 experiments executed in PSB-VVER facility, operated at Electrogorsk Research and Engineering Center (Russia), in the framework of the TACIS 2.03/97 Contract 3.03.03 Part A, EC financed. The paper describes the methodology adopted at University of Pisa, starting form the scenarios foreseen in the final test matrix until the execution of the experiments. This process considers three key topics: a) the scaling issue and the simulation, with unavoidable distortions, of the expected performance of the reference nuclear power plants; b) the code assessment process involving the identification of phenomena challenging the code models; c) the features of the concerned integral test facility (scaling limitations, control logics, data acquisition system, instrumentation, etc.). The activities performed in this respect are discussed, and emphasis is also given to the relevance of the thermal losses to the environment. This issue affects particularly the small scaled facilities and has relevance on the scaling approach related to the power and volume of the facility. (author)

  13. Safety regulations regarding to accident monitoring and accident sampling at Russian NPPs with VVER type reactors

    International Nuclear Information System (INIS)

    Sharafutdinov, Rachet; Lankin, Michail; Kharitonova, Nataliya

    2014-01-01

    The paper describes a tendency by development of regulatory document requirements related to accident monitoring and accident sampling at Russia's NPPs. Lessons learned from the Fukushima Daiichi accident pointed at the importance and necessary to carry out an additional safety check at Russia's nuclear power plants in the preparedness for management of severe accidents at NPPs. Planned measures for improvement of severe accidents management include development and implementation of the accident instrumentation systems, providing, monitoring, management and storage of information in a severe accident conditions. The draft of Safety Guidelines <VVER reactors' prepared by Scientific and Engineering Centre for Nuclear and Radiation Safety (SEC NRS) established the main criteria for accident monitoring instrumentation that can monitor relevant plant parameters in the reactor and inside containment during and after a severe accident in nuclear power plants. Development of these safety guidelines is in line with the recommendations of IAEA Action Plan on Nuclear Safety in response to the Fukushima Daiichi event and recommendations of the IAEA Nuclear Energy series Report <

  14. Results of international standard problem No. 36 severe fuel damage experiment of a VVER fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Firnhaber, M. [Gesellschaft fuer Anlagen-und Reaktorsicherheit, Koeln (Germany); Yegorova, L. [Nuclear Safety Institute of Russian Research Center, Moscow (Russian Federation); Brockmeier, U. [Ruhr-Univ. of Bochum (Germany)] [and others

    1995-09-01

    International Standard Problems (ISP) organized by the OECD are defined as comparative exercises in which predictions with different computer codes for a given physical problem are compared with each other and with a carefully controlled experimental study. The main goal of ISP is to increase confidence in the validity and accuracy of analytical tools used in assessing the safety of nuclear installations. In addition, it enables the code user to gain experience and to improve his competence. This paper presents the results and assessment of ISP No. 36, which deals with the early core degradation phase during an unmitigated severe LWR accident in a Russian type VVER. Representatives of 17 organizations participated in the ISP using the codes ATHLET-CD, ICARE2, KESS-III, MELCOR, SCDAP/RELAP5 and RAPTA. Some participants performed several calculations with different codes. As experimental basis the severe fuel damage experiment CORA-W2 was selected. The main phenomena investigated are thermal behavior of fuel rods, onset of temperature escalation, material behavior and hydrogen generation. In general, the calculations give the right tendency of the experimental results for the thermal behavior, the hydrogen generation and, partly, for the material behavior. However, some calculations deviate in important quantities - e.g. some material behavior data - showing remarkable discrepancies between each other and from the experiments. The temperature history of the bundle up to the beginning of significant oxidation was calculated quite well. Deviations seem to be related to the overall heat balance. Since the material behavior of the bundle is to a great extent influenced by the cladding failure criteria a more realistic cladding failure model should be developed at least for the detailed, mechanistic codes. Regarding the material behavior and flow blockage some models for the material interaction as well as for relocation and refreezing requires further improvement.

  15. Characteristics and properties of cladding tubes for VVER-1000 higher Uranium content fuel rods

    International Nuclear Information System (INIS)

    Peregud, M.; Markelov, A.; Novikov, V.; Gusev, A.; Konkov, V.; Pimenov, Y.; Agapitov, V.; Shtutsa, M.

    2009-01-01

    To improve the fuel cycle economics and to further increase the VVER fuel usability the work programme is under way to design novel improved fuel, fuel rods and fuel assemblies. Longer FA operation time that is needed to increase the fuel burnup and the related design developments of novel fuel assemblies resulted not only in changing types and sizes of Zirconium items and fuel assembly components but also altered the requirements placed on their technical characteristics. To use fuel rods having a larger charge of fuel, to improve their behaviour in LOCA, to reduce fuel rod damage ability during assembling the work was carried out to perfect the characteristics of both the cladding (reduced wall thickness and more rigid tolerances for geometry) and its material. To meet the more rigid requirements for the geometry dimensions of cladding tubes an improved process flow sheet has been designed and employed for their fabrication and also the finishing treatment of tube surfaces has been improved. The higher and stable properties of the cladding materials were managed through using the special purity in terms of Hafnium Zirconium (not higher than 100 ppm Hf) as a base of the E110 alloy and maintaining within the valid specifications for the alloy the optimized contents of Oxygen and Iron at the levels of (600 - 990) ppm and (250 - 700) ppm, respectively. The work was under way in 2004 - 2008 years; during this period the technology and materials science solutions were mastered that were phased-in introduced into the production of the cladding tubes for the fuels loaded into the of the Kalinin NPP Unit 1

  16. CFD evaluation of hydrogen risk mitigation measures in a VVER-440/213 containment

    Energy Technology Data Exchange (ETDEWEB)

    Heitsch, Matthias, E-mail: Matthias.Heitsch@ec.europa.e [Institute for Energy, Joint Research Centre, PO Box 2, 1755 ZG Petten (Netherlands); Huhtanen, Risto [VTT Technical Research Centre of Finland, PO Box 1000, FI-02044 VTT (Finland); Techy, Zsolt [VEIKI Institute for Electric Power Research Co., PO Box 80, H-1251 Budapest (Hungary); Fry, Chris [Serco, Winfrith Technology Centre, Dorchester, Dorset DT2 8DH (United Kingdom); Kostka, Pal [VEIKI Institute for Electric Power Research Co., PO Box 80, H-1251 Budapest (Hungary); Niemi, Jarto [VTT Technical Research Centre of Finland, PO Box 1000, FI-02044 VTT (Finland); Schramm, Berthold [Gesellschaft fuer Anlagen- und Reaktorsicherheit, GRS mbH, Schwertnergasse 1, 50667 Koeln (Germany)

    2010-02-15

    In the PHARE project 'Hydrogen Management for the VVER440/213' (HU2002/000-632-04-01), CFD (Computational Fluid Dynamics) calculations using GASFLOW, FLUENT and CFX were performed for the Paks NPP (Nuclear Power Plant), modelling a defined severe accident scenario which involves the release of hydrogen. The purpose of this work is to demonstrate that CFD codes can be used to model gas movement inside a containment during a severe accident. With growing experience in performing such analyses, the results encourage the use of CFD in assessing the risk of losing containment integrity as a result of hydrogen deflagrations. As an effective mitigation measure in such a situation, the implementation of catalytic recombiners is planned in the Paks NPP. In order to support these plans both unmitigated and recombiner-mitigated simulations were performed. These are described and selected results are compared. The codes CFX and FLUENT needed refinement to their models of wall and bulk steam condensation in order to be able to fully simulate the severe accident under consideration. Several CFD codes were used in parallel to model the same accident scenario in order to reduce uncertainties in the results. Previously it was considered impractical to use CFD codes to simulate a full containment subject to a severe accident extending over many hours. This was because of the expected prohibitive computing times and missing physical capabilities of the codes. This work demonstrates that, because of developments in the capabilities of CFD codes and improvements in computer power, these calculations have now become feasible.

  17. IVO participation in IAEA benchmark for VVER-type nuclear power plants seismic analysis and testing

    International Nuclear Information System (INIS)

    Varpasuo, P.

    1997-12-01

    This study is a part of the IAEA coordinated research program 'Benchmark study for the Seismic Analysis and Testing of VVER Type NPPs'. The study reports the numerical simulation of the blast test for Paks and Kozloduy nuclear power plants beginning from the recorded free-field response and computing the structural response at various points inside the reactor building. The full-scale blast tests of the Paks and Kozloduy NPPs took place in December 1994 and in July 1996. During the tests the plants operated normally. The instrumentation for the tests consisted of 52 recording channels with 200 Hz sampling rate. Detonating 100 kg charges in 50-meter deep boreholes at 2.5-km distance from the plant carried out the blast tests. The 3D structural models for both reactor buildings were analyzed in the frequency domain. The number of modes extracted in both cases was about 500 and the cut-off frequency was 25 Hz. In the response history run the responses of the selected points were evaluated. The input values for response history run were the three components of the excitation, which were transformed from time domain to the frequency domain with the aid of Fourier transform. The analysis was carried out in frequency domain and responses were transferred back to time domain with inverse Fourier transform. The Paks and Kozloduy blast tests produced a wealth of information on the behavior of the nuclear power plant structures excited by blast type loads containing also the low frequency wave train if albeit with small energy content. The comparison of measured and calculated results gave information about the suitability of the selected analysis approach for the investigated blast type loading

  18. Analyses of SBO sequence of VVER1000 reactor using TRACE and MELCOR codes

    International Nuclear Information System (INIS)

    Mazzini, Guido; Kyncl, Milos; Miglierini, Bruno; Kopecek, Vit

    2015-01-01

    In response to the Fukushima accident, the European Commission ordered to perform stress tests to all European Nuclear Power Plants (NPPs). Due to shortage of time a number of conclusions in national stress tests reports were based on engineering judgment only. In the Czech Republic, as a follow up, a consortium of Research Organizations and Universities has decided to simulate selected stress tests scenarios, in particular station Black-Out (SBO) and Loss of Ultimate Sink (LoUS), with the aim to verify conclusions made in the national stress report and to analyse time response of respective source term releases. These activities are carried out in the frame of the project 'Prevention, preparedness and mitigation of consequences of Severe Accident (SA) at Czech NPPs in relation to lessons learned from stress tests after Fukushima' financed by the Ministry of Interior. The Research Centre Rez has been working on the preparation of a MELCOR model for VVER1000 NPP starting with a plant systems nodalization. The basic idea of this paper is to benchmark the MELCOR model with the validated TRACE model, first comparing the steady state and continuing in a long term SBO plus another event until the beginning of the severe accident. The presented work focuses mainly on the preliminary comparison of the thermo-hydraulics of the two models created in MELCOR and TRACE codes. After that, preliminary general results of the SA progression showing the hydrogen production and the relocation phenomena will be shortly discussed. This scenario is considered closed after some seconds to the break of the lower head. (author)

  19. Importance of ECP in the prediction of radiation fields in PWR and VVER primary circuits

    International Nuclear Information System (INIS)

    Urquidi-Macdonald, M.; Jacesko, S.L.; Macdonald, Digby D.; Salter-Williams, M.

    2002-01-01

    A model has been developed for predicting mass and activity transport in the primary coolant circuits of PWRs and VVERs with the objective of demonstrating and quantifying the importance of the electrochemical corrosion potential (ECP) in determining the impact of both processes on reactor operation. The model initially employs a radiolysis/mixed potential code to calculate the ECP at four locations (core, hot leg, steam generator, cold leg) and the ECP is then used to estimate the local magnetite solubility. The solubility is then averaged around the loop to yield the ''background'' solubility. Comparison of the background solubility with the local solubility determines whether precipitation or dissolution will occur at any given point in the circuit under any given set of conditions. It is further assumed that the concentration of 59 Co in the coolant is given by the isotopic fraction of this species compared with iron averaged over all materials and weighted by the respective wetted areas. Activation of 59 Co to 60 Co is assumed to occur in the coolant phase by fast, epithermal, and thermal neutron capture. The calculated activity is then used to train an artificial neural network (ANN) to establish relationships between activity at any given location and the operating properties of the reactor, including coolant pH, ECP, temperature, power level, etc. The model predicts that during shut down, magnetite (and hence 59 Co) migrates to the core, where it is irradiated and activated, particularly during subsequent start-up. During start-up, the magnetite (and hence 60 Co) migrates from the core to out-of-core surfaces where it establishes the radiation fields. (authors)

  20. Application of an optimized AM procedure following a SBO in a VVER1000

    International Nuclear Information System (INIS)

    Cherubini, Marco; D'Auria, Francesco; Petrangeli, Gianni; Muellner, Nikolaus

    2006-01-01

    The University of Pisa was involved in investigations of an Accident Management procedure based on passive feed water injection. Some experiments were performed to validate this possibility (e.g. in LOBI and Bethsy facilities) and fully analyzed by thermal hydraulic system codes. Recent activities in which the University of Pisa is engaged (also as leader) are focused on VVER-1000 safety analyses. The idea is now to use the acquired knowledge to explore if a procedure based on passive feed water injection is applicable and can provide any benefits to the Russian design pressurized plant. The postulated accident is a station blackout, in such a way only passive systems are available. The proposed AM is based on secondary and primary side depressurisation in sequence. The secondary side depressurisation performed by the BRU-A valves has the scope to feed passively the SGs with the water left in the feed water lines and in the deaerators. The primary side depressurisation, via the PORV, is foreseen to keep the plant at the lowest pressure (to reduce the energy of the system) and to maximize the 'grace time' of the plant. Three cases are here considered: no operator action, application of the optimized AM sequence, application of the AM procedure at the last time when it is effective. The intention of this paper is to show that in case of an unlikely event such a SBO the implementation of a strategy based on systems not designed for specific safety application can have a large impact on the 'grace time' of the plant. (author)

  1. The Design of PSB-VVER Experiments Relevant to Accident Management

    Science.gov (United States)

    Nevo, Alessandro Del; D'Auria, Francesco; Mazzini, Marino; Bykov, Michael; Elkin, Ilya V.; Suslov, Alexander

    Experimental programs carried-out in integral test facilities are relevant for validating the best estimate thermal-hydraulic codes(1), which are used for accident analyses, design of accident management procedures, licensing of nuclear power plants, etc. The validation process, in fact, is based on well designed experiments. It consists in the comparison of the measured and calculated parameters and the determination whether a computer code has an adequate capability in predicting the major phenomena expected to occur in the course of transient and/or accidents. University of Pisa was responsible of the numerical design of the 12 experiments executed in PSB-VVER facility (2), operated at Electrogorsk Research and Engineering Center (Russia), in the framework of the TACIS 2.03/97 Contract 3.03.03 Part A, EC financed (3). The paper describes the methodology adopted at University of Pisa, starting form the scenarios foreseen in the final test matrix until the execution of the experiments. This process considers three key topics: a) the scaling issue and the simulation, with unavoidable distortions, of the expected performance of the reference nuclear power plants; b) the code assessment process involving the identification of phenomena challenging the code models; c) the features of the concerned integral test facility (scaling limitations, control logics, data acquisition system, instrumentation, etc.). The activities performed in this respect are discussed, and emphasis is also given to the relevance of the thermal losses to the environment. This issue affects particularly the small scaled facilities and has relevance on the scaling approach related to the power and volume of the facility.

  2. New treatment centers for radioactive waste from Russian designed VVER-reactors

    International Nuclear Information System (INIS)

    Chrubasik, A.

    1997-01-01

    The nuclear power plants using Russian designed VVER-type reactors, were engineered and designed without any wastes treatment facilities. The liquid and solid waste were collected in storage tanks and shelters. After many years of operation, the storage capabilities are exhausted. The treatment of the stored and still generated waste represents a problem of reactor safety and requires a short term solution. NUKEM has been commissioned to design and construct several new treatment centers to remove and process the stored waste. This paper describes the process and lessons learned on the development of this system. The new radioactive waste treatment center (RWTC) includes comprehensive systems to treat both liquid and solid wastes. The process includes: 1) treatment of evaporator concentrates, 2) treatment of ion exchange resins, 3) treatment of solid burnable waste, 4) treatment of liquid burnable waste, 5) treatment of solid decontaminable waste, 6) treatment of solid compactible waste. To treat these waste streams, various separate systems and facilities are needed. Six major facilities are constructed including: 1. A sorting facility with systems for waste segregation. 2. A high-force compactor facility for volume reduction of non-burnable waste. 3. An incinerator facility for destruction of: 1) solid burnable waste, 2) liquid burnable waste, 3) low level radioactive ion exchange resins. 4. A facility for melting of incineration residue. 5. A cementation facility for stabilization of: 1) medium level radioactive ion exchange resins, 2) solid non compactible waste, 3) compacted solid waste. 6. Separation of radionuclides from evaporator concentrates. This presentation will address the facilities, systems, and lessons learned in the development of the new treatment centers. (author)

  3. Influence of taking into account in-pressurizer convective heat- and mass transfer influence effects at the transients in VVER with code RELAP 5/MOD 3.2

    International Nuclear Information System (INIS)

    Konovalyuk, L.N.; Shevelev, D.V.; Kravchenko, V.G.

    2003-01-01

    PRZ model is proposed which allows taking into account in pressurizer convective heat- and mass transfer influence effects at the transients in VVER (PWR) Type Reactors case when calculations performed with using 1D thermohydraulic codes. The theoretical backgrounds are given to define the transients with the convective coolant instability in PRZ. The instability threshold is given for real PRZ geometry

  4. Construction of the Plant RT-2 as a way for solving the problem of VVER-1000 spent fuel management in Russia

    International Nuclear Information System (INIS)

    Lazarev, L.N.; Lyubtsev, R.I.; Egorov, N.N.; Lebedev, V.A.; Revenko, Y.A.; Fedosov, Y.G.; Dubrovskii, V.M.

    1993-01-01

    Nuclear power in the Russian Federation in the future will be based on the VVER-1000 and it's modifications. To manage the spent fuels from this plant, the Plant RT-2 was designed to process the spent fuel. Plant construction was started in 1984 and stopped in 1989 due to economic difficulties. The necessity of the continuation of the plant is discussed

  5. United States Department of Energy's reactor core protection evaluation methodology for fires at RBMK and VVER nuclear power plants. Revision 1

    International Nuclear Information System (INIS)

    1997-06-01

    This document provides operators of Soviet-designed RBMK (graphite moderated light water boiling water reactor) and VVER (pressurized light water reactor) nuclear power plants with a systematic Methodology to qualitatively evaluate plant response to fires and to identify remedies to protect the reactor core from fire-initiated damage

  6. Comparison of radioactive doses after the last protection layer insight the reactor structure for Russian VVER-1000 and German PWR-1300 reactors

    International Nuclear Information System (INIS)

    Rahimi, A.; Mansourshaiflu, N.; Alizadeh, M. R.

    2004-01-01

    In pressurized reactors (VVER and PWR), various protections layers are used for reducing the output core doses. At any protection layer, some amount of neutron and gamma doses is reduced. In this project the axial flux of neutron and gamma beams have been evaluated at various protection layers in the operation state the German PWR-1300 and Russian VVER-1000 reactors by the MCNP computer code. For the purpose of effective use of the MCNP code and assuring its correct performance about of fluxed beams common and series of scientific answers and bench marks should be considered and the results obtained by the MCNP code, be compared with this answers. Then by using appropriate method, for reducing the flux variants of neutron and gamma beams at various protection layers of German PWR-1300 and Russian VVER-1000 reactors of the operation state of both reactors have been accelerated. In this projects, bench marks are computations and numbers existing in PSAR's present at Bushehr nuclear power plant. At the end, by using the results obtained and the standard doses, the time which a person can have work activity at the reactor wall (after the last protection layer), was compared for the operation status of the German PWR-1300 and Russian VVER-1000 reactors

  7. Accounting for the inertia of the thermocouples' measurements by modelling of a NPP Kalinin-3 transient with the coupled system code ATHLET-BIPR-VVER

    International Nuclear Information System (INIS)

    Nikonov, S.; Velkov, K.

    2008-01-01

    The ATHLET-BIPR-VVER coupled system code is applied for performing of safety analysis for different WWER reactors. During the last years its validation matrix is continuously being enlarged. The measurements performed during the commissioning phase of NPP Kalinin Unit 3 for the transient 'Switching-off of one Main Circulation Pump at nominal power' are very well documented and have a variety of recorded integral and local thermo-hydraulic and neutron-physic parameters including the measurements' errors. This data is being used for further validation of the coupled code system ATHLET-BIPR-VVER. In the paper are discussed the problems and our solutions by the correct interpretation of the measured thermocouples' records at NPP Kalinin-3 and the comparison with the predicted results by the coupled thermal-hydraulic/neutron-kinetic code ATHLET-BIPR-VVER. Of primary importance by such comparisons is the correct accounting of the fluid mixing process that take place in the surrounding of the measuring sensors and also the consideration of the time delay (inertia term) of the measuring devices. On the bases of previous experience and many simulations of the defined transient a method is discussed and proposed to consider correctly the inertia term of the thermocouples' measurements. The new modelling is implemented in the coupled system code ATHLET-BIPR-VVER for further validation. (Author)

  8. Development of fuel cycles with new fuel with 8.9 mm external diameter for VVER-440. Preliminary assessment of operating efficiency

    Energy Technology Data Exchange (ETDEWEB)

    Gagarinskiy, Alexey [National Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

    2017-09-15

    Since the introduction of VVERs-440, their fuel assemblies are subject to ongoing improvements. Until now, the basic structural parameters of fuel, such as rod diameter of 9.1 mm, have never changed. This paper focuses on computational estimates of basic neutronic parameters of the fuel cycle that involves assemblies consisting of fuel rods with diameter reduced to 8.9 mm.

  9. The calculational VVER burnup Credit Benchmark No.3 results with the ENDF/B-VI rev.5 (1999)

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez Gual, Maritza [Centro de Tecnologia Nuclear, La Habana (Cuba). E-mail: mrgual@ctn.isctn.edu.cu

    2000-07-01

    The purpose of this papers to present the results of CB3 phase of the VVER calculational benchmark with the recent evaluated nuclear data library ENDF/B-VI Rev.5 (1999). This results are compared with the obtained from the other participants in the calculations (Czech Republic, Finland, Hungary, Slovaquia, Spain and the United Kingdom). The phase (CB3) of the VVER calculation benchmark is similar to the Phase II-A of the OECD/NEA/INSC BUC Working Group benchmark for PWR. The cases without burnup profile (BP) were performed with the WIMS/D-4 code. The rest of the cases have been carried with DOTIII discrete ordinates code. The neutron library used was the ENDF/B-VI rev. 5 (1999). The WIMS/D-4 (69 groups) is used to collapse cross sections from the ENDF/B-VI Rev. 5 (1999) to 36 groups working library for 2-D calculations. This work also comprises the results of CB1 (obtained with ENDF/B-VI rev. 5 (1999), too) and CB3 for cases with Burnup of 30 MWd/TU and cooling time of 1 and 5 years and for case with Burnup of 40 MWd/TU and cooling time of 1 year. (author)

  10. Post-Test Analysis of 11% Break at PSB-VVER Experimental Facility using Cathare 2 Code

    Science.gov (United States)

    Sabotinov, Luben; Chevrier, Patrick

    The best estimate French thermal-hydraulic computer code CATHARE 2 Version 2.5_1 was used for post-test analysis of the experiment “11% upper plenum break”, conducted at the large-scale test facility PSB-VVER in Russia. The PSB rig is 1:300 scaled model of VVER-1000 NPP. A computer model has been developed for CATHARE 2 V2.5_1, taking into account all important components of the PSB facility: reactor model (lower plenum, core, bypass, upper plenum, downcomer), 4 separated loops, pressurizer, horizontal multitube steam generators, break section. The secondary side is represented by recirculation model. A large number of sensitivity calculations has been performed regarding break modeling, reactor pressure vessel modeling, counter current flow modeling, hydraulic losses, heat losses. The comparison between calculated and experimental results shows good prediction of the basic thermal-hydraulic phenomena and parameters such as pressures, temperatures, void fractions, loop seal clearance, etc. The experimental and calculation results are very sensitive regarding the fuel cladding temperature, which show a periodical nature. With the applied CATHARE 1D modeling, the global thermal-hydraulic parameters and the core heat up have been reasonably predicted.

  11. Safety related investigations of the VVER-1000 reactor type by the coupled code system TRACE/PARCS

    International Nuclear Information System (INIS)

    Jaeger, Wadim; Lischke, Wolfgang; Sanchez Espinoza, Victor Hugo

    2007-01-01

    This study was performed at the Institute of Reactor Safety at the Research Center Karlsruhe. It is embedded in the ongoing investigations of the international code application and maintenance program (CAMP) for qualification and validation of system codes like TRACE [1] and PARCS [2]. The predestinated reactor type for the validation of these two codes was the Russian designed VVER-1000 because the OECD/NEA VVER-1000 Coolant Transient Benchmark Phase 2 [3] includes detailed information of the Bulgarian nuclear power plant (NPP) Kozloduy unit 6. The posttest-investigations of a coolant mixing experiment have shown that the predicted parameters (coolant temperature, pressure drop, etc.) are in good agreement to the measured data. The coolant mixing pattern especially in the downcomer has been also reproduced quiet well by TRACE. The coupled code system TRACE/PARCS which was applied on a postulated main steam line break (MSLB) provides good results compared to reference values and the ones of other participants of the benchmark. It can be pointed out that the developed three-dimensional nodalisation of the reactor pressure vessel (RPV) is appropriate for the description of transients where the thermal-hydraulics and the neutronics are strongly linked. (author)

  12. A study of different cases of VVER reactor core flooding in a large break loss of coolant accident

    International Nuclear Information System (INIS)

    Bezrukov, Y.A.; Schekoldin, V.I.; Zaitsev, S.I.; Churkin, A.N.; Lisenkov, E.A.

    2016-01-01

    The paper covers a brief review of reflooding studies performed in different countries and the relevant tests performed in OKB GIDROPRESS (Russia) are discussed in more detail. The OKB GIDROPRESS test facility simulates the primary circuit of the VVER-440 reactor, with a full-scale fuel assembly (FA) mockup as the core simulator. The VVER core reflooding was studied in a FA mockup containing 126 fuel rod simulators with axial power peaking. The experiments were performed for two types of flooding. The first type is top flooding of the empty (steamed) FA mockup. The second type is bottom flooding of the FA mockup with level of boiling water. The test parameters are as follows: the range of the supplied power to the bundle is from 40 to 320 kW, the cooling water flow rate is from 0.04 to 1.1 kg/s, the maximum temperature of the fuel rod simulator is 800 C. degrees and the linear heat flux is from 0.1 to 1.0 kW/m. The test results were used for computer code validation, especially for the TRAP package code. The experiments show that as the transverse dimension of the FA mockup increases, the flow choking of the water supplied from the top by the steam flow significantly decreases

  13. Reloading pattern optimization of VVER-1000 reactors in transient cycles using genetic algorithm

    International Nuclear Information System (INIS)

    Rahmani, Yashar

    2017-01-01

    Highlights: • The genetic algorithm (GA) and the innovative weighting factors method were used. • The coupling of WIMSD5-B and CITATION-LDI2 neutronic codes with the thermohydraulic WERL code was employed. • Optimization of reloading patterns was carried out in two states. • First an arrangement with satisfactory excess reactivity and the flattest power distribution was searched. • Second, it is tried to obtain an arrangement with satisfactory safety threshold and the maximum K_e_f_f. - Abstract: The present paper proposes application of the genetic algorithm (GA) and the innovative weighting factor method to optimize the reloading pattern of Bushehr VVER-1000 reactor in the second cycle. To estimate the composition of fuel assemblies remaining from the first cycle and precisely calculate the objective parameters of each reloading pattern in the second cycle, coupling of WIMSD5-B and CITATION-LDI2 codes in the neutronic section and the WERL code in the thermo-hydraulic section was employed. Optimization of the reloading patterns was carried out in two states. To meet the mentioned objective, with application of the weighting factor method in the first state, the type and quantity of the loadable fresh assemblies were determined to enable the reactor core to maintain the core criticality over the entire cycle length. Afterwards, the genetic algorithm was used to optimize the reloading pattern of the reactor to obtain an arrangement with flat radial power distribution. In the second state, the optimization algorithm was free to select the type and number of fresh fuel assemblies to be able to search for an arrangement with the maximum effective multiplication factor and the safe power peaking factor. In addition, in order to ensure the safety and desirability of the proposed patterns in both states, a time-dependent examination of the thermo-neutronic behavior of the reactor core was carried out during the second cycle. With consideration of the new

  14. Radioactive sludge and wastewater analysis and treatment in the Hungarian VVER-440/213-type NPP

    International Nuclear Information System (INIS)

    Patzay, G.; Weiser, L.; Feil, F.; Schunk, J.; Patek, G.; Pinter, T.

    2010-01-01

    It is well known that in the Hungarian VVER-type nuclear power plant Paks the radioactive waste waters are collected in common tanks. These water streams contain radioactive isotopes in ultra-low concentration and inactive compounds as major components (borate 1.7 g/dm 3 , sodium-nitrate 0.4 g/dm 3 , sodium-hydroxide 0.16 g/dm 3 , and oxalate 0.25 g/dm 3 ). These low salinity solutions were evaporated by adding sodium-hydroxide, until 400 g/dm 3 salt content is reached. There is about 6000 m 3 concentrated evaporator bottom residues in the tanks of the reactor. There are some tanks at the power plant containing sludge type radioactive waste containing more or less liquid phase too. The general physical and chemical characteristics (density, pH, total solid, dissolved solid etc.) and chemical and radiochemical composition are important information for volume reduction and solidification treatment of these wastes. We have investigated and constructed a complex analysis system for the radioactive sludge and supernatant analysis, including the physical, as well as the chemical and radiochemical analysis methods. Using well known analysis techniques as ion chromatography, ICP-MS, AAS, gamma-and alpha-spectrometry and chemical alkaline fusion digestion and acidic dissolution methods we could analyze the main inorganic, organic and radioactive components of the sludges and supernatants. Determination of the mass and charge balance for the sludge samples were more difficult then for the supernatant samples. Not only are there assumptions required about the chemical form and the oxidation state of the species present in the sludge, but many of the compounds in the sludge are mixed oxides which are not directly measured. Also, the sludge is actually a slurry with a high water content. The interstitial liquid is in close contact with the sludge, and there are many ionic solubility equilibriums. The anion data for the sludge samples are based on the water soluble anions that

  15. Adoption of in-vessel retention concept for VVER-440/V213 reactors in Central European Countries

    Energy Technology Data Exchange (ETDEWEB)

    Matejovic, Peter, E-mail: peter.matejovic@ivstt.sk [Inzinierska Vypoctova Spolocnost (IVS), Jana Holleho 5, 91701 Trnava (Slovakia); Barnak, Miroslav; Bachraty, Milan; Vranka, Lubomir [Inzinierska Vypoctova Spolocnost (IVS), Jana Holleho 5, 91701 Trnava (Slovakia); Berky, Robert [Integrita a Bezpecnost Ocelovych Konstrukcii, Rybnicna 40, 831 07 Bratislava (Slovakia)

    2017-04-01

    Highlights: • Design of in-vessel retention concept for VVER-440/V213 reactors. • Thermal loads acting on the inner reactor surface. • Structural response of reactor pressure vessel. • External reactor vessel cooling. - Abstract: An in-vessel retention (IVR) concept was proposed for standard VVER-440/V213 reactors equipped with confinement made of reinforced concrete and bubbler condenser pressure suppression system. This IVR concept is based on simple modifications of existing plant technology and thus it was attractive for plant operators in Central European Countries. Contrary to the solution that was adopted before at Loviisa NPP in Finland (two units of VVER-440/V213 reactor with steel confinement equipped with ice condenser), the coolant access to the reactor pressure vessel from flooded cavity is enabled via closable hole installed in the centre of thermal shield of the reactor lower head instead of lowering this massive structure in the case of severe accident. As a consequence, the crucial point of this IVR concept is narrow gap between torispherical lower head and thermal and biological shield. Here the highest thermal flux is expected in the case of severe accident. Thus, realistic estimation of thermal load and corresponding deformations of reactor wall and their impact on gap width for coolant flow are of primarily importance. In this contribution the attention is paid especially to the analytical support with emphasis to the following points: 1) {sup ∗}Estimation of thermal loads acting on the inner reactor surface; 2) {sup ∗}Estimation of structural response of reactor pressure vessel (RPV) with emphasis on the deformation of outer reactor surface and its impact on the annular gap between RPV wall and thermal/biological shield; 3) {sup ∗}Analysis of external reactor vessel cooling. For this purpose the ASTEC code was used for performing analysis of core degradation scenarios, the ANSYS code for structural analysis of reactor vessel

  16. Mechanisms of radiation embrittlement of VVER-1000 RPV steel at irradiation temperatures of (50–400)°C

    Energy Technology Data Exchange (ETDEWEB)

    Kuleshova, E.A., E-mail: evgenia-orm@yandex.ru [National Research Center “Kurchatov Institute”, Kurchatov Sq. 1, Moscow 123182 (Russian Federation); National Research Nuclear University “MEPhI” (Moscow Engineering Physics Institute), Kashirskoe Highway 31, Moscow 115409 (Russian Federation); Gurovich, B.A.; Bukina, Z.V.; Frolov, A.S.; Maltsev, D.A.; Krikun, E.V.; Zhurko, D.A.; Zhuchkov, G.M. [National Research Center “Kurchatov Institute”, Kurchatov Sq. 1, Moscow 123182 (Russian Federation)

    2017-07-15

    This work summarizes and analyzes our recent research results on the effect of irradiation temperature within the range of (50–400)°C on microstructure and properties of 15Kh2NMFAA class 1 steel (VVER-1000 reactor pressure vessel (RPV) base metal). The paper considers the influence of accelerated irradiation with different temperature up to different fluences on the carbide and irradiation-induced phases, radiation defects, yield strength changes and critical brittleness temperature shift (ΔT{sub K}) as well as on changes of the fraction of brittle intergranular fracture and segregation processes in the steel. Low temperature irradiation resulted solely in formation of radiation defects – dislocation loops of high number density, the latter increased with increase in irradiation temperature while their size decreased. In this regard high embrittlement rate observed at low temperature irradiation is only due to the hardening mechanism of radiation embrittlement. Accelerated irradiation at VVER-1000 RPV operating temperature (∼300 °C) caused formation of radiation-induced precipitates and dislocation loops, as well as some increase in phosphorus grain boundary segregation. The observed ΔT{sub K} shift being within the regulatory curve for VVER-1000 RPV base metal is due to both hardening and non-hardening mechanisms of radiation embrittlement. Irradiation at elevated temperature caused more intense phosphorus grain boundary segregation, but no formation of radiation-induced precipitates or dislocation loops in contrast to irradiation at 300 °C. Carbide transformations observed only after irradiation at 400 °C caused increase in yield strength and, along with a contribution of the non-hardening mechanism, resulted in the lowest ΔT{sub K} shift in the studied range of irradiation temperature and fluence. - Highlights: •Structural elements in RPV steel are studied at different irradiation temperatures. •Highest number density dislocation loops are

  17. Analysis of the VVER Standard Problem INSC-PSBV1 '11% Coolant Leak from Upper Plenum' with RELAP5/MOD3.2

    International Nuclear Information System (INIS)

    Melikhov, O.; Melikhov, V.; Parfenov, Yu.; Gavritenkova, O.; Lipatov, I.; Elkin, I.; Bayless, P.

    2004-01-01

    Analyses of a loss-of-coolant experiment carried out at the PSB-VVER test facility with the RELAP5/MOD3.2 code have been performed independently by analysts at the Electrogorsk Research and Engineering Center (EREC) and the Idaho National Engineering and Environmental Laboratory (INEEL). The PSB-VVER facility is a full-height scale model of a VVER 1000 reactor that is approximately 1/300 scale in volume and power. VVER Standard Problem INSC-PSBV1 represents an 11% leak from the upper plenum of the PSB-VVER facility, simulating the rupture of one of the accumulator injection lines. The safety-significant thermalhydraulic phenomena occurring in VVER type reactors addressed by this experiment were identified in the test validation matrix. Most of the phenomena of the validation matrix were reasonably simulated by RELAP5/MOD3.2 in both calculations. The major differences between the test and the calculations were the timing of the core heatup, and the thermal response to the accumulator injection cycles in both calculations. The INEEL calculation had a more extensive axial heatup, with most of the core experiencing small heat-ups. The accumulator injection was more effective in quenching the core in the test than in the INEEL calculation. This difference is attributed to the liquid distribution in the core, rather than to the heat transfer models in the code. The code calculation had a more uniform axial distribution of the liquid in the core, and the accumulator injection did not have much impact on the core liquid inventory. In the EREC calculation, only one heatup of the cladding temperature was observed for upper and middle section of the fuel rods before the final heatup. The small heat-ups were not reproduced in EREC calculation. The difference could be attributed to differences in liquid distribution, namely the core region in the EREC calculation contains more liquid over most of the transient than in the experiment. The distribution of liquid in the core in

  18. Determination of the Effectiveness of Control Rods in the VVER Reactor Fuel Assemblies; Determination de l'Efficacite des Barres de Reglage dans les Ensembles Combustibles du reacteur VVER; Opredelenie ehffektivnosti reguliruyushchikh sterzhnej v sborkakh reaktora VVEHR; Determinacion de la Eficacia de las Barras de Control en los Conjuntos de Elementos Combustibles del Reactor VVER

    Energy Technology Data Exchange (ETDEWEB)

    Semenov, V. N.; Lunin, G. L.; Komissarov, L. V.; Kamyshan, A. N.; Halizev, V. I.; Andrianov, G. Ja.; Voznesenskij, V. A.; Kuz' micheva, V. A.; Lebedev, V. I. [Ordena Lenina Institut Atomnoj Energii Im. I.V. Kurchatova, Moskva, SSSR (Russian Federation)

    1964-06-15

    The paper describes experiments done in homogeneous mock-ups of the fuel assemblies from the VVER Reactor (at one level of enrichment) to determine the effectiveness of absorbing systems comprising shim fuel assemblies or water cavities and of absorbing rods clad in jackets made of differing materials. The paper also gives data on some experiments that have been done in mock-ups of assemblies with differing levels of enrichment. These experiments make it possible to verify the methods used in calculation and to evaluate the prospects of using them for heterogeneous reactors. (author) [French] Le memoire decrit les experiences qui ont ete faites pour determiner l 'efficacite des absorbants contenus dans les barres de compensation, l'effet cavitaire et l 'efficacite des absorbants gaines de materiaux divers, au moyen d'assemblages homogenes de cartouches de combustible du reacteur VVER (reacteur de puissance ralenti et refroidi a l 'eau ayant le meme taux d'enrichissement. On y trouve en outre des donnees sur certaines experiences executees a l 'aide d'assemblages de cartouches de combustible taux d'enrichissement differents. Ces travaux permettent de verifier la methode de calcul et d'evaluer ses possibilites d'application aux reacteurs non homogenes. (author) [Spanish] Se describen en la memoria experimentos para determinar la eficacia de los materiales absorbentes contenidos en las barras de compensacion, el efecto de cavitacion y la eficacia de los materiales absorbentes revestidos de diversos materiales, realizados con ayuda de los conjuntos homogeneos de elementos combustibles del reactor VVER (reactor de potencia moderado y refrigerado por agua) con un solo grado de enriquecimiento. Ademas, se exponen datos sobre los experimentos efectuados con ayuda de conjuntos de grados de enriquecimientos; variados. Tales experimentos permiten verificar el metodo de calculo teorico, utilizad o y evaluar la posibilidad de aplicarlo a los reactores no homogeneos. (author

  19. A phenomenological method of mechanical properties definition of reactor pressure vessels (RPV) steels VVER according to the ball indentation diagram

    International Nuclear Information System (INIS)

    Bakirov, M. B.; Potapov, V.V.; Massoud, J.P.

    2002-01-01

    This work presents specimen-free methods of a standard uniaxial tension diagram construction and RPV (reactor pressure vessel) steels VVER strength properties definition out of a continuous ball indentation diagram. A similarity phenomenon of uniaxial tension strain curves at a hardening area and an area of a ball indentation constitutes the ground of the methods. The methods are developed on the basis of the uniform graphic representation of elasto-plastic strain processes by indentation and tension and with the reception of the unified yield curve at a hardening area. The calculation results on the phenomenological method conducted for a wide range of RPV steels conditions of nuclear reactors have shown a good precision as far as strain curves construction by the uniaxial tension out of the elasto-plastic indentation diagram is concerned. (authors)

  20. The effect of boron dilution transient on the VVER-1000 reactor core using MCNP and COBRA-EN codes

    Energy Technology Data Exchange (ETDEWEB)

    Jafari, Naser; Talebi, Saeed [Amirkabir Univ. of Technology, Tehran Polytechnic (Iran, Islamic Republic of). Dept. of Energy Engineering and Physics

    2017-07-15

    In this paper, the effect of boron dilution transient, as a consequence of the malfunction of the boron control system, was investigated in a VVER-1000 reactor, and then an appropriate setpoint was determined for the actuation of the emergency protection system to the reactor shutdown. In order to simulate the boron dilution, first, the whole reactor core was simulated by MCNPX code to compute the radial and axial power distribution. Then, the COBRA-EN code was employed using calculated power distribution for analyzing the thermal-hydraulic of hot fuel assembly and for extracting the safety parameters. For the safe operation of the reactor, certain parameters must be in defined specified ranges. Comparison between our results and FSARs data shows that the present modeling provides a good prediction of boron dilution transient with the maximum relative difference about 4%.

  1. 'AER working group D on VVER safety analysis' minutes of the meeting in Rossendorf, Germany, 10-12 May 1999

    International Nuclear Information System (INIS)

    Siltanen, P.

    1999-01-01

    AER Working Group D on VVER reactor safety analysis held its eighth meeting in Research Centre Rossendorf during the period 10-12 May 1999. There were altogether 14 participants from eight member organizations. In addition to the general information exchange on recent activities, the topics of the meeting included: Final conclusions and lessons from the solution of the AER Dynamic Benchmark Problem No. 5 on a main steam header break accident; Validation of dynamics codes and the definition of a 6th dynamic benchmark; Safety analyses and reactivity issues: approaches and results; Advances in development of models and codes for reactor dynamics applications; Documentation of dynamic benchmark problems and solutions for the AER Benchmark Book; Future activities. (author)

  2. Optimization of fuel core loading pattern design in a VVER nuclear power reactors using Particle Swarm Optimization (PSO)

    International Nuclear Information System (INIS)

    Babazadeh, Davood; Boroushaki, Mehrdad; Lucas, Caro

    2009-01-01

    The two main goals in core fuel loading pattern design optimization are maximizing the core effective multiplication factor (K eff ) in order to extract the maximum energy, and keeping the local power peaking factor (P q ) lower than a predetermined value to maintain fuel integrity. In this research, a new strategy based on Particle Swarm Optimization (PSO) algorithm has been developed to optimize the fuel core loading pattern in a typical VVER. The PSO algorithm presents a simple social model by inspiration from bird collective behavior in finding food. A modified version of PSO algorithm for discrete variables has been developed and implemented successfully for the multi-objective optimization of fuel loading pattern design with constraints of keeping P q lower than a predetermined value and maximizing K eff . This strategy has been accomplished using WIMSD and CITATION calculation codes. Simulation results show that this algorithm can help in the acquisition of a new pattern without contravention of the constraints.

  3. Improvement of MSLB transient analysis for VVER by the coupled code system KIKO3D/ATHLET

    International Nuclear Information System (INIS)

    Hegyi, Gy.; Kereszturi, A.; Trosztel, I.

    2001-01-01

    An overview is given on the investigations of the Main Steam Line Break transient in a VVER- 440 NPP by using the KIKO3D/ATHLET 1.2.A coupled code system. Special attention was paid for the influence of modeling the outcore detector signals and the malfunctioning of the emergency control system (scram with stuck rod). The conservatism of the calculations was assured even in the case of application of the 3D best estimate KIKO3D code. The consequence of MSLB accident is investigated at the end of cycle (EOC), at full power (FP) and shut down initial conditions. Even if very strong conservative assumptions were applied, dangerous hot spots were not found in the supposed scenarios.(author)

  4. Results of the QUENCH-12 experiment on reflood of a VVER-type bundle

    International Nuclear Information System (INIS)

    Stuckert, J.; Grosse, M.; Heck, M.; Schanz, G.; Sepold, L.; Stegmaier, U.; Steinbrueck, M.

    2008-09-01

    The QUENCH experiments are to investigate the hydrogen source term resulting from the water injection into an uncovered core of a Light-Water Reactor. The QUENCH test bundle with a total length of approximately 2.5 m usually consists of 21 fuel rod simulators of Western PWR (Pressurized Water Reactor) geometry. The QUENCH-12 test bundle, however, which was set up to investigate the effects of VVER materials and bundle geometry (hexagonal lattice) on core reflood consisted of 31 fuel rod simulators. 18 rods of which were electrically heated using tungsten heaters in the rod center. All claddings, corner rods and grid spacers were made of Zr1%Nb (E110) and the shroud of Zr2.5%Nb (E125). For comparison, the QUENCH-06 test (ISP-45) with Western PWR geometry (square lattice) was chosen as reference. QUENCH-12 conducted at the Forschungszentrum Karlsruhe (FZK, Karlsruhe Research Center) on 27 September, 2006 in the frame of the EC-supported ISTC program 1648.2 was proposed by FZK together with RIAR Dimitrovgrad and IBRAE Moscow (Russia), and supported by pretest calculations performed by PSI (Switzerland) and the Kurchatov Institute Moscow (Russia) together with IRSN Cadarache (France). It had been preceded by a low-temperature (maximum 1073 K) pretest on 25 August, 2006 to characterize the bundle thermal hydraulic performance and to provide data to assess the code models used for pretest calculational support. After a stabilization period at 873 K pre-oxidation took place at ∝1470 K for ∝3400 s to achieve a maximum oxide thickness of about 200 μm. A transient phase followed with a temperature rise to ∝2050 K. Then quenching of the bundle by a water flow of 48 g/s was initiated cooling the bundle to ambient temperature in ∝5 min. Following reflood initiation, a moderate temperature excursion of ∝50 K was observed, over a longer period than in QUENCH-06. The temperatures at elevations between 850 mm and 1050 mm exceeded the melting temperature of β-Zr, i

  5. Results of the QUENCH-12 experiment on reflood of a VVER-type bundle

    Energy Technology Data Exchange (ETDEWEB)

    Stuckert, J.; Grosse, M.; Heck, M.; Schanz, G.; Sepold, L.; Stegmaier, U.; Steinbrueck, M. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Materialforschung, Programm Nukleare Sicherheitsforschung; Goryachev, A.; Ivanova, I. [RIAR (FSUE SSC-RIAR) Dimitrovgrad (Russian Federation)

    2008-09-15

    The QUENCH experiments are to investigate the hydrogen source term resulting from the water injection into an uncovered core of a Light-Water Reactor. The QUENCH test bundle with a total length of approximately 2.5 m usually consists of 21 fuel rod simulators of Western PWR (Pressurized Water Reactor) geometry. The QUENCH-12 test bundle, however, which was set up to investigate the effects of VVER materials and bundle geometry (hexagonal lattice) on core reflood consisted of 31 fuel rod simulators. 18 rods of which were electrically heated using tungsten heaters in the rod center. All claddings, corner rods and grid spacers were made of Zr1%Nb (E110) and the shroud of Zr2.5%Nb (E125). For comparison, the QUENCH-06 test (ISP-45) with Western PWR geometry (square lattice) was chosen as reference. QUENCH-12 conducted at the Forschungszentrum Karlsruhe (FZK, Karlsruhe Research Center) on 27 September, 2006 in the frame of the EC-supported ISTC program 1648.2 was proposed by FZK together with RIAR Dimitrovgrad and IBRAE Moscow (Russia), and supported by pretest calculations performed by PSI (Switzerland) and the Kurchatov Institute Moscow (Russia) together with IRSN Cadarache (France). It had been preceded by a low-temperature (maximum 1073 K) pretest on 25 August, 2006 to characterize the bundle thermal hydraulic performance and to provide data to assess the code models used for pretest calculational support. After a stabilization period at 873 K pre-oxidation took place at {proportional_to}1470 K for {proportional_to}3400 s to achieve a maximum oxide thickness of about 200 {mu}m. A transient phase followed with a temperature rise to {proportional_to}2050 K. Then quenching of the bundle by a water flow of 48 g/s was initiated cooling the bundle to ambient temperature in {proportional_to}5 min. Following reflood initiation, a moderate temperature excursion of {proportional_to}50 K was observed, over a longer period than in QUENCH-06. The temperatures at elevations

  6. On detection of the possible use of VVERs for unreported production of plutonium. Final report for the period July 1988 - December 1989

    International Nuclear Information System (INIS)

    Simov, R.; Nelov, N.; Stoyanova, I.; Kovachev, N.; Yonchev, P.

    1989-01-01

    The study includes an analysis of the feasibility of unreported production of plutonium-239 in VVER-440 reactors. It is shown that for VVER-440 reactors 36 natural uranium oxide fuel assemblies in the peripheral region of the core need to be loaded to produce 8 kg of extra plutonium in one cycle. Substituting the peripheral fuel assemblies with natural uranium oxide fuel assemblies, the changes in the power peaking are negligible and do not affect reactor safety. Unreported production outside the core is not practical due to physical and mechanical constraints, low flux level, etc. The feasibility of unreported removal of irradiated material in spent fuel cask has been also assessed. After about a month cooling time, still within the refueling period, the irradiated natural uranium fuel assemblies could be removed off-site without significant health hazard to the workers. To improve the effectiveness of the safeguards objectives, additional inspection activities are suggested. 10 figs

  7. Study, analysis, assess and compare the nuclear engineering systems of nuclear power plant with different reactor types VVER-1000, namely AES-91, AES-92 and AES-2006

    International Nuclear Information System (INIS)

    Le Van Hong; Tran Chi Thanh; Hoang Minh Giang; Le Dai Dien; Nguyen Nhi Dien; Nguyen Minh Tuan

    2015-01-01

    On November 25, 2009, in Hanoi, the National Assembly had been approved the resolution about policy for investment of nuclear power project in Ninh Thuan province which include two sites, each site has two units with power around 1000 MWe. For the nuclear power project at Ninh Thuan 1, Vietnam Government signed the Joint-Governmental Agreement with Russian Government for building the nuclear power plant with reactor type VVER. At present time, the Russian Consultant proposed four reactor technologies can be used for Ninh Thuan 1 project, namely: AES-91, AES-92, AES-2006/V491 and AES-2006/V392M. This report presents the main reactor engineering systems of nuclear power plants with VVER-1000/1200. The results from analysis, comparison and assessment between the designs of AES-91, AES-92 and AES-2006 are also presented. The obtained results show that the type AES-2006 is appropriate selection for Vietnam. (author)

  8. Issues in the use of Weapons-Grade MOX Fuel in VVER-1000 Nuclear Reactors: Comparison of UO2 and MOX Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, J.J.

    2005-05-27

    The purpose of this report is to quantify the differences between mixed oxide (MOX) and low-enriched uranium (LEU) fuels and to assess in reasonable detail the potential impacts of MOX fuel use in VVER-1000 nuclear power plants in Russia. This report is a generic tool to assist in the identification of plant modifications that may be required to accommodate receiving, storing, handling, irradiating, and disposing of MOX fuel in VVER-1000 reactors. The report is based on information from work performed by Russian and U.S. institutions. The report quantifies each issue, and the differences between LEU and MOX fuels are described as accurately as possible, given the current sources of data.

  9. Top-Level Software for VVER-1000 In-core Monitoring System under Implementation of Expanded Nuclear Fuel Diversification Program in Ukraine

    International Nuclear Information System (INIS)

    Khalimonchuk, V.A.

    2015-01-01

    The paper considers the possibility and expediency of developing mathematical software for VVER-1000 ICMS in Ukraine. This mathematical software is among the most important conditions for implementation of the expanded nuclear fuel diversification program. The top-level software is to be developed based on SSTC own studies in the development of codes for power distribution recovery, which were successfully used previously for RBMK-1000 safety analysis

  10. Development of a qualified nodalization for small-break LOCA transient analysis in PSB-VVER integral test facility by RELAP5 system code

    Energy Technology Data Exchange (ETDEWEB)

    Shahedi, S. [Department of Energy Engineering, Sharif University of Technology, Azadi Street, Tehran (Iran, Islamic Republic of); Jafari, J., E-mail: jalil_jafari@yahoo.co [Reactors and Accelerators R and D School, Nuclear Science and Technology Research Institute, North Kargar Street, Tehran (Iran, Islamic Republic of); Boroushaki, M. [Department of Energy Engineering, Sharif University of Technology, Azadi Street, Tehran (Iran, Islamic Republic of); D' Auria, F. [DIMNP, University of Pisa, Via Diotisalvi 2, 56126 Pisa (Italy)

    2010-10-15

    This paper deals with development and qualification of a nodalization for modeling of the PSB-VVER integral test facility (ITF) by RELAP5/MOD3.2 code and prediction of its primary and secondary systems behaviors at steady state and transient conditions. The PSB-VVER is a full-height, 1/300 volume and power scale representation of a VVER-1000 NPP. A RELAP5 nodalization has been developed for PSB-VVER modeling and a nodalization qualification process has been applied for the developed nodalization at steady state and transient levels and a qualified nodalization has been proposed for modeling of the PSB ITF. The 11% small-break loss-of-coolant-accident (SBLOCA), i.e. rupture of one of the hydroaccumulators (HA) injection lines in the upper plenum (UP) region of reactor pressure vessel (RPV) below the hot legs (HL), inlets has been considered for nodalization qualification process. The influence of the different steam generator (SG) nodalizations on the RELAP5 results and on the nodalization qualification process has been examined. The 'steady state' qualification level includes checking the correctness of the initial and boundary conditions and geometrical fidelity. In the 'transient' qualification level, the time dependent results of the code calculation are compared with the experimental time trends from both the qualitative and quantitative point of view. For quantitative assessment of the results, a Fast Fourier Transform Based Method (FFTBM) has been used. The FFTBM was used to establish a range in which the steam generators nodalizations can vary.

  11. Conducting water chemistry of the secondary coolant circuit of VVER-based nuclear power plant units constructed without using copper containing alloys

    Science.gov (United States)

    Tyapkov, V. F.

    2014-07-01

    The secondary coolant circuit water chemistry with metering amines began to be put in use in Russia in 2005, and all nuclear power plant units equipped with VVER-1000 reactors have been shifted to operate with this water chemistry for the past seven years. Owing to the use of water chemistry with metering amines, the amount of products from corrosion of structural materials entering into the volume of steam generators has been reduced, and the flow-accelerated corrosion rate of pipelines and equipment has been slowed down. The article presents data on conducting water chemistry in nuclear power plant units with VVER-1000 reactors for the secondary coolant system equipment made without using copper-containing alloys. Statistical data are presented on conducting ammonia-morpholine and ammonia-ethanolamine water chemistries in new-generation operating power units with VVER-1000 reactors with an increased level of pH. The values of cooling water leaks in turbine condensers the tube system of which is made of stainless steel or titanium alloy are given.

  12. Study of long-term loss of all AC power supply sources for VVER-1000/V320 in connection with application of new engineering safety features for SAMG

    International Nuclear Information System (INIS)

    Borisov, Evgeni; Grigorov, Dobrin; Mancheva, Kaliopa

    2013-01-01

    Highlights: • In this study we presented analysis for a new SAMG approach. • The approach is applicable for all PWR reactors from 2nd generation. • We investigated two scenarios with total black out. • The RELAP/MOD 3.2 computer code is used in performing the analyses. - Abstract: This paper presents the results of analysis for application of a new Severe Accident Management Guideline (SAMG) approach which is specifically applied for VVER-1000/B320 reactor installations. In general, this innovative approach is fully applicable for all the pressurized water reactors from second generation. The purposes of the analysis for the new SAMG approach application are as follows: • To represent suggestions for new engineering safety features application for SAMG strategies. • To assess the applicability of the new engineering safety features and means for SAMG strategies in case of loss of all off-site power supply sources for VVER-1000/B320 reactor installations. • To represent important operator actions and to analyse the effectiveness of these actions for accidents management in compliance with the new approach. • The RELAP5/MOD3.3 computer code has been used in performing the analyses in a VVER-1000 Nuclear Power Plant (NPP) model. The input data deck for the analysis is optimized, verified and validated

  13. TAREG 2.01/00 Project, ''Validation of neutron embrittlement for VVER 1000 and 440/213 RPVs, with emphasis on integrity assessment''

    International Nuclear Information System (INIS)

    Ahlstrand, R.; Margolin, B.; Kostylev, V.; Yurchenko, E.; Akbashev, I.; Piminov, V.; Nikolaev, Y.; Koshkin, V.; Kharshenko, V.; Chyrko, L.; Bukhanov, V.; Comsa, O.

    2012-01-01

    The irradiation embrittlement and integrity of the VVER reactors has been an important issue in many EC supported TACIS and PHARE projects since 1990. In the EC annual program 2000 two TACIS projects (TAREG 2.01/00 and 2.01/03) were approved on the issue in order to improve the neutron irradiation embrittlement databases, elaborate new trend curves for the embrittlement and to assess the integrity of the RPVs (Reactor Pressure Vessel) by analysing PTS transients (Pressurized Thermal Shock) for some selected Russian and Ukrainian VVER 1000 and 440/213 NPPs. In this paper the TAREG 2.01/00 project is briefly described with some details from the twin project 2.01/03, which served as a materials testing project, providing inputs for the 1st project. As a result of the project new trend curves for neutron irradiation embrittlement were elaborated, based on upgraded and more reliable surveillance results databases. The PTS study shows that the integrity of the selected VVER RPVs can be ensured to the end of RPV design life. (author)

  14. OECD/DOE/CEA VVER-1000 Coolant Transient Benchmark. Summary Record of the Fourth Workshop (V100-CT4)

    International Nuclear Information System (INIS)

    2006-01-01

    The overall objective of the VVER-1000 coolant transient (V1000CT) benchmark is to assess computer codes used in the safety analysis of VVER power plants, specifically for their use in analysis of reactivity transients in a VVER-1000. The V1000CT benchmark consists of two phases: V1000CT-1 is a simulation of the switching on of one main coolant pump (MCP) when the other three MCPs are in operation, and V1000CT-2 concerns calculation of coolant mixing tests and main steam line break (MSLB) scenarios. Each of the two phases contains three exercises. The reference problem chosen for simulation in Phase 1 is a MCP switching on when the other three main coolant pumps are in operation in a VVER-1000. This event is characterized by rapid increase in the flow through the core resulting in a coolant temperature decrease, which is spatially dependent. This leads to insertion of spatially distributed positive reactivity due to the modelled feedback mechanisms and non-symmetric power distribution. Simulation of the transient requires evaluation of core response from a multi-dimensional perspective (coupled three-dimensional neutronics/core thermal-hydraulics) supplemented by a one-dimensional simulation of the remainder of the reactor coolant system. Three exercises are defined in the framework of Phase 1: a) Exercise 1 - Point kinetics plant simulation; b) Exercise 2 - Coupled 3-D neutronics/core thermal-hydraulics response evaluation; c) Exercise 3 - Best-estimate coupled 3-D core/plant system transient modelling. In addition to the measured (experiment) scenario, extreme calculation scenarios were defined in the frame of Exercise 3 for better testing 3-D neutronics/thermal-hydraulics techniques. The proposals concerned: rod ejection simulations with scram set points at two different power levels. Since the previous coupled code benchmarks indicated that further development of the mixing computation models in the integrated codes is necessary, a coolant mixing experiment and

  15. Measures for ensuring hydrogen fire and explosion safety for VVER-440/230

    International Nuclear Information System (INIS)

    Bezlepkin, V.; Semashko, S.; Svetlov, S.; Sidorov, V.; Ivkov, I.; Krylov, Yu.; Kukhtevich, V.

    2004-01-01

    This paper deals with the findings of calculation analysis as regards the release of mass, energy and hydrogen during beyond-design-basis accident (BDBA) at Kola NPP equipped with VVER-440 reactor (B-230 design) and in respect of distribution of hydrogen throughout NPP tight compartments. The analysis figures out the number and locations of passive catalytic hydrogen recombiners and of the sensors of the hydrogen concentration monitoring system. In order to prove the hydrogen safety of the design, it has been necessary to review accidents accompanied by maximum emissions (both peak and integral ones) of hydrogen into the tight area. During design-basis accident (DBA), no steam/zirconium reactions occur in the reactor core. Out of BDBA, the severe accidents with damage to the core accompanied oxidative reactions between zirconium and steel with emission of hydrogen are regarded as the most dangerous ones. Assessment of additional hydrogen sources shows that the contribution of such sources to the total amount of hydrogen that may emit during a severe accident is insignificant. Calculations have been made for the following scenarios of severe accidents, which seem to be the most important in terms of hydrogen safety analysis: - 20 mm leak from the primary circuit in combination with a failure of the emergency makeup system; - 500 mm PCP rupture in the vicinity of reactor inlet branch with bi-lateral leakage of coolant. Releases of mass and energy during the aforesaid scenarios, changes of medium parameters within the tight compartments and analysis of possible fire conditions have been analyzed by means of Russian computer codes RATEG/SVECHA/HEFEST, KUPOL-M and LIMITS. The said analysis shows that the large break accident (500 mm), i.e. PCP rupture in the vicinity of the reactor branch with bi-lateral leakage of coolant is of the keen interest in terms of hydrogen safety. This accident typifies powerful short-term release of hydrogen at a significantly lesser

  16. Time versus frequency domain calculation of the main building complex of the VVER 440/213 NPP PAKS

    International Nuclear Information System (INIS)

    Katona, T.; Ratkai, S.; Halbritter, A.; Krutzik, N.J.; Schuetz, W.

    1995-01-01

    Various dynamic analyses were conducted for the main building complex of the VVER 440/213 PAKS in order to determine the dynamic response and assess the aseismic capacity of this nuclear power plant. Different types of mathematical models for idealizing the soil and the building structures were used. The main goal of the study presented here was to demonstrate the effects of different procedures for consideration of soil-structure interaction on the dynamic response of the structures mentioned above. The analyses were based on appropriate mathematical models of the coupled vibration structures (reactor building, turbine hall, intermediate building structures) and the layered soil. On the basis of this study, it can be concluded that substructure models using frequency-independent impedances and cut-off of modal damping usually provide conservative results. Complex models which allow the soil-soil and the structure or by frequency-dependent impedances) provide more accurate results. The latter approach results in more efficient designs which are not only safe but also economical. (author). 7 refs., 15 figs

  17. Calculation of spatial weighting functions for ex-core detectors of VVER-440 reactors by Monte Carlo method

    International Nuclear Information System (INIS)

    Berki, T.

    2003-01-01

    The signal of ex-core detectors depends not only on the total power of a reactor but also on the power distribution. The spatial weighting function establishes correspondence between the power distribution and the detector signal. The weighting function is independent of the power distribution. The weighting function is used for detector-response analyses, for example in the case of rod-drop experiments. (1) The paper describes the calculation and analysis of the weighting function of a VVER-440. The three-dimensional Monte Carlo code MCNP is used for the evaluation. Results from forward and adjoint calculations are compared. The effect of the change in the concentration of boric acid is also investigated. The evaluation of the spatial weighting function is a fixed-source neutron transport problem, which can be solved much faster by adjoint calculation, however forward calculations provide more detailed results. It is showed that the effect of boric acid upon the weighting function is negligible. (author)

  18. Recent results of three-dimensional CFD simulations of coolant mixing in VVER-440/213 reactor pressure vessel

    International Nuclear Information System (INIS)

    Kiss, B.; Boros, I.; Aszodi, A.

    2008-01-01

    The Budapest University of Technology and Economics, Institute of Nuclear Techniques has been working since 2001 on the three-dimensional CFD model of the reactor pressure vessel of the VVER-440 type reactor. During this time period - due to the development of the available computational capacity - a very complex and detailed model of the RPV has been developed. The aim of the construction of the new model is to describe further internal structures of the RPV (e.g. correct modeling of brake tubes, or internals in the upper mixing chamber) and to perform an extensive sensitivity analysis on the different modeling and calculation parameters (e.g. porous region models vs. detailed modeling, or n different turbulence models). The new model can be applied for steady state calculation during normal operational condition and for different transient analyses as well. One interesting application is the participation in a planned benchmark exercise on the start-up of the sixth main coolant pump, which is aimed to compare the capabilities of mixing models of one-dimensional system codes with the results of CFD simulation. (authors)

  19. Almost twenty years' search of transuranium isotopes in effluents discharged to air from nuclear power plants with VVER reactors.

    Science.gov (United States)

    Hölgye, Z; Filgas, R

    2006-04-01

    Airborne effluents of 5 stacks (stacks 1-5) of three nuclear power plants, with 9 pressurized water reactors VVER of 4,520 MWe total power, were searched for transuranium isotopes in different time periods. The search started in 1985. The subject of this work is a presentation of discharge data for the period of 1998-2003 and a final evaluation. It was found that 238Pu, 239,240Pu, 241Am, 242Cm, and 244Cm can be present in airborne effluents. Transuranium isotope contents in most of the quarterly effluent samples from stacks 2, 4 and 5 were not measurable. Transuranium isotopes were present in the effluents from stack l during all 9 years of the study and from stack 3 since the 3rd quarter of 1996 as a result of a defect in the fuel cladding. A relatively high increase of transuranium isotopes in effluents from stack 3 occurred in the 3rd quarter of 1999, and a smaller increase occurred in the 3rd quarter of 2003. In each instance 242Cm prevailed in the transuranium isotope mixtures. 238Pu/239,240Pu, 241Am/239,240Pu, 242Cm/239,240Pu, and 244Cm/239,240Pu ratios in fuel for different burn-up were calculated, and comparison of these ratios in fuel and effluents was performed.

  20. Comparison of hydrogen generation for TVSM and TVSA fuel assemblies for water water energy reactor (VVER)-1000

    International Nuclear Information System (INIS)

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

    2009-01-01

    This paper presents the results received during investigation of hydrogen generation for both types fuel assemblies-the old modernistic type of fuel assemblies (TVSM) and recently installed new one alternative type of fuel assemblies (TVSA) in case of severe accident. There are some differences between both types FAs. They have different geometry as well as different burnable poisons. To investigate behavior of new fuel assemblies during the severe conditions it have been performed comparison of fuel behavior of old type TVSM fuel assembly to new one TVSA. To perform this investigation it has been used MELCOR 'input model' for Kozloduy Nuclear Power Plant (KNPP) VVER 1000. The model was developed by Institute for Nuclear Research and Nuclear Energy-Bulgarian Academy of Sciences (INRNE-BAS) for investigation of severe accident scenarios and Probabilistic Safety Analyses (PSA) level 2. The model provides a significant analytical capability for the Bulgarian technical specialists, working in the field of the NPP safety, for analysis of core and containment damaged states and the estimation of radionuclides release outside fuel cladding. It was accepted criteria for vessel integrity about hydrogen concentration to be 8%. This criterion was based on the decision of RSK (Germany commission for reactor safety). Generally based on the received results it was made conclusion that using both types of fuel assemblies it was not disturbance safety conditions of NPP

  1. Progress in the U.S. department of energy sponsored in-depth safety assessments of VVER and RBMK reactors

    International Nuclear Information System (INIS)

    Binder, J.L.; Petri, M.C.; Pasedag, W.F.

    2001-01-01

    Since the disastrous accident at Chernobyl Nuclear Power Plant Unit 4 in 1986, there has been international recognition of the safety concerns posed by the operation of 67 Soviet-designed commercial nuclear reactors. These reactors are operated in eight countries from the former Soviet Union and its former satellite states in Central and Eastern Europe. The majority of these plants are in the Russian Federation (30 units) and Ukraine (14 units). New plants are in various stages of construction. U.S. support to improve the safety of Soviet-designed reactors over the past decade has been intended to enhance operational safety, provide for risk-reduction measures, and enhance regulatory capability. The U.S. approach to improving the safety of Soviet-designed reactors has matured into a large multi-year program known as the Soviet-Designed Reactor Safety Program that is managed by the U.S. Department of Energy (US DOE). The mission of the program is to implement a self-sustaining nuclear safety improvement program that would lead to internationally accepted safety practices at the plants. Those practices would create a safety culture that would be reflected in the operation, regulation, and professional attitudes of the designers, operators, and regulators of the nuclear facilities. A key component of this larger program has been the Plant Safety Evaluation Program, which supports in-depth safety assessments of VVER and RBMK plants. (author)

  2. Accident analysis for new reactor concepts and VVER type reactor design with advanced fuel. STC with Russia. Final report

    International Nuclear Information System (INIS)

    Grundmann, U.; Kliem, S.; Mittag, S.; Rohde, U.; Seidel, A.

    2000-10-01

    In the frame of a project on scientific-technical cooperation funded by BMBF/BMWi, the 3D reactor dynamics code DYN3D developed at Forschungszentrum Rossendorf (FZR), has been transferred to the Institute of Physics and Power Engineering (IPPE) Obninsk in Russia and integrated into the software package of IPPE. DYN3D has been coupled to a thermohydraulic system code used in IPPE making available 3D neutron kinetics within this software package. A new macroscopic cross section library has been created using a modified version of the WIMS/D4 code. This library includes data for modernized fuel design containing burnable absorbers in different concentrations, which is tested in VVER-1000 type reactors. The cross section library has been connected to DYN3D. Calculations were performed to check the library in comparison with other data libraries and codes. The code DYN3D and the coupled 3D neutron kinetics/thermal hydraulics code system were used to perform analyses of Anticipated Transients Without Scram (ATWS) for the reactor design ABV-67, an integral reactor concept with small power developed under participation of IPPE. The fluid dynamics code DINCOR developed at IPPE was transferred to FZR. It was used in validation calculations on test problems for the short-term core melt behaviour (CORVIS experiments). (orig.) [de

  3. Implementation of New Reactivity Measurement System and New Reactor Noise Analysis Equipment in a VVER-440 Nuclear Power Plant

    Science.gov (United States)

    Vegh, János; Kiss, Sándor; Lipcsei, Sándor; Horvath, Csaba; Pos, István; Kiss, Gábor

    2010-10-01

    The paper deals with two recently developed, high-precision nuclear measurement systems installed at the VVER-440 units of the Hungarian Paks NPP. Both developments were motivated by the reactor power increase to 108%, and by the planned plant service time extension. The first part describes the RMR start-up reactivity measurement system with advanced services. High-precision picoampere meters were installed at each reactor unit and measured ionization chamber current signals are handled by a portable computer providing data acquisition and online reactivity calculation service. Detailed offline evaluation and analysis of reactor start-up measurements can be performed on the portable unit, too. The second part of the paper describes a new reactor noise diagnostics system using state-of-the-art data acquisition hardware and signal processing methods. Details of the new reactor noise measurement evaluation software are also outlined. Noise diagnostics at Paks NPP is a standard tool for core anomaly detection and for long-term noise trend monitoring. Regular application of these systems is illustrated by real plant data, e.g., results of standard reactivity measurements during a reactor startup session are given. Noise applications are also illustrated by real plant measurements; results of core anomaly detection are presented.

  4. VVER-1000 main steam line break analysis using the coupled code system DYN3D/ATHLET

    International Nuclear Information System (INIS)

    Kliem, Soeren; Hoehne, Thomas; Rohde, Ulrich; Weiss, Frank-Peter; Kozmenkov, Yaroslav

    2008-01-01

    Calculations using the coupled code system DYN3D/ATHLET were performed in the frame of the OECD/NEA MSLB benchmark for a VVER-1000 reactor. The coolant mixing inside the reactor pressure vessel was treated using a validated empirical mixing model implemented into the DYN3D/ATHLET code. Using very conservative boundary conditions (reduced scram worth, two stuck rods, running MCP throughout the whole transient) a return-to-power was predicted. For the assessment of the empirical mixing model a time dependent calculation using the computational fluid dynamics code CFX-10 was performed. For that analysis, a detailed model of the reactor pressure vessel consisting of the inlets nozzles, downcomer, lower plenum and a part of the core and having 4.67 million unstructured tetra cell elements was used. For the considered case with running main coolant pumps, this calculation shows a sector formation at the core inlet with a certain amount of mixing at the edges of the sector. A core calculation using these CFX results as boundary conditions predicted also a return-to-power with a maximum value being about 200 MW lower than in the coupled code calculation. This variation calculation confirms the applicability of the empirical mixing model. The comparison shows also, that in this way results with a reasonable degree of conservatism can be obtained. (authors)

  5. Evaluation of an experiment modelling heat transfer from the melt pool for use in VVER 440/213 reactors

    International Nuclear Information System (INIS)

    Skop, J.

    2003-12-01

    The strategy of confining core melt within the reactor vessel is among promising strategies to mitigate severe accidents of VVER 440/213 reactors. This strategy consists in residual heat removal from the melt by external vessel cooling from the outside, using water from the flooded reactor downcomer. This approach can only be successful if the critical heat flux on the external vessel surface is not exceeded. This can be assessed based on the parameters of heat transfer from the core melt pool in the conditions of natural circulation within the pool. Those parameters are the subject of the report. A basic description of the terms and physical basis of the strategy of confining core melt inside the vessel is given in Chapter 2, which also briefly explains similarity theory, based on which the results obtained on experimental facilities, using simulation materials, can be related to the actual situation inside a real reactor. Chapter 3 presents an overview of experimental work addressing the characteristics of heat transfer from the core melt pool in natural circulation conditions and a description of the experimental facilities. An overview of the results emerging from the experiments and their evaluation with respect to their applicability to reactors in Czech nuclear power plants are given in Chapter 4

  6. Operation of Nuclear Fuel Based on Reprocessed Uranium for VVER-type Reactors in Competitive Nuclear Fuel Cycles

    Energy Technology Data Exchange (ETDEWEB)

    Troyanov, V.; Molchanov, V.; Tuzov, A. [TVEL Corporation, 49 Kashirskoe shosse, Moscow 115409 (Russian Federation); Semchenkov, Yu.; Lizorkin, M. [RRC ' Kurchatov Institute' (Russian Federation); Vasilchenko, I.; Lushin, V. [OKB ' Gidropress' (Russian Federation)

    2009-06-15

    Current nuclear fuel cycle of Russian nuclear power involves reprocessed low-enriched uranium in nuclear fuel production for some NPP units with VVER-type LWR. This paper discusses design and performance characteristics of commercial nuclear fuel based on natural and reprocessed uranium. It presents the review of results of commercial operation of nuclear fuel based on reprocessed uranium on Russian NPPs-unit No.2 of Kola NPP and unit No.2 of Kalinin NPP. The results of calculation and experimental validation of safe fuel operation including necessary isotope composition conformed to regulation requirements and results of pilot fuel operation are also considered. Meeting the customer requirements the possibility of high burn-up achieving was demonstrated. In addition the paper compares the characteristics of nuclear fuel cycles with maximum length based on reprocessed and natural uranium considering relevant 5% enrichment limitation and necessity of {sup 236}U compensation. The expedience of uranium-235 enrichment increasing over 5% is discussed with the aim to implement longer fuel cycles. (authors)

  7. Analysis of core damage frequency: Nuclear power plant Dukovany, VVER/440 V-213 Unit 1, internal events. Volume 1: Main report

    International Nuclear Information System (INIS)

    Pugila, W.J.

    1994-01-01

    This report presents the final results from the Level 1 probabilistic safety assessment (PSA) for the Dukovany VVER/440 V-213 nuclear power plant, Unit 1. Section 1.1 describes the objectives of this study. Section 1.2 discusses the approach that was used for completing the Dukovany PSA. Section 1.3 summarizes the results of the PSA. Section 1.4 provides a comparison of the results of the Dukovany PSA with the results of other PSAs for different types of reactors worldwide. Section 1.5 summarizes the conclusions of the Dukovany PSA

  8. Using of the Serpent code based on the Monte-Carlo method for calculation of the VVER-1000 fuel assembly characteristics

    Directory of Open Access Journals (Sweden)

    V. V. Galchenko

    2016-12-01

    Full Text Available The description of calculation scheme of fuel assembly for preparation of few-group characteristics is considered with help of Serpent code. This code uses the Monte-Carlo method and energy continuous microscopic data libraries. Serpent code is devoted for calculation of fuel assembly characteristics, burnup calculations and preparation of few-group homogenized macroscopic cross-sections. The results of verification simulations in comparison with other codes (WIMS, HELIOS, NESSEL etc., which are used for neutron-physical analysis of VVER type fuel, are presented.

  9. Specification of a VVER-1000 SFAT device prototype. Interim report on Task FIN A 1073 of the Finnish Support Programme to IAEA Safeguards

    International Nuclear Information System (INIS)

    Nikkinen, M.; Tiitta, A.; Iievlev, S.; Dvoeglazov, M.; Lopatin, S.

    1999-01-01

    The project to specify the optimal design of the Spent Fuel Attribute Tester (SFAT) for Ukrainian VVER-1000 facilities was run under Finnish Support Programme for IAEA Safeguards under the task FIN A1073. This document illustrates the optimum design and takes into account the special conditions at the Ukrainian facilities. The requirement presented here takes into account the needs of the user (IAEA), nuclear safety authority (NRA) and facilities. This document contains the views of these parties. According to this document, the work to design the optimal SFAT device can be started. This document contains also consideration for the operational procedures, maintenance and safety. (orig.)

  10. Analysis of the VVER-1000 coolant transient benchmark phase 1 with the code system RELAP5/PARCS

    International Nuclear Information System (INIS)

    Victor Hugo Sanchez Espinoza

    2005-01-01

    Full text of publication follows: As part of the reactor dynamics activities of FZK/IRS, the qualification of best-estimate coupled code systems for reactor safety evaluations is a key step toward improving their prediction capability and acceptability. The VVER-1000 Coolant Transient Benchmark Phase 1 represents an excellent opportunity to validate the simulation capability of the coupled code system RELAP5/PACRS regarding both the thermal hydraulic plant response (RELAP5) using measured data obtained during commissioning tests at the Kozloduy nuclear power plant unit 6 and the neutron kinetics models of PARCS for hexagonal geometries. The Phase 1 is devoted to the analysis of the switching on of one main coolant pump while the other three pumps are in operation. It includes the following exercises: (a) investigation of the integral plant response using a best-estimate thermal hydraulic system code with a point kinetics model (b) analysis of the core response for given initial and transient thermal hydraulic boundary conditions using a coupled code system with 3D-neutron kinetics model and (c) investigation of the integral plant response using a best-estimate coupled code system with 3D-neutron kinetics. Already before the test, complex flow conditions exist within the RPV e.g. coolant mixing in the upper plenum caused by the reverse flow through the loop-3 with the stopped pump. The test is initiated by switching on the main coolant pump of loop-3 that leads to a reversal of the flow through the respective piping. After about 13 s the mass flow rate through this loop reaches values comparable with the one of the other loops. During this time period, the increased primary coolant flow causes a reduction of the core averaged coolant temperature and thus an increase of the core power. Later on, the power stabilizes at a level higher than the initial power. In this analysis, special attention is paid on the prediction of the spatial asymmetrical core cooling during

  11. Investigations of the VVER-1000 coolant transient benchmark phase 1 with the coupled code system RELAP5/PARCS

    International Nuclear Information System (INIS)

    Sanchez-Espinoza, Victor Hugo

    2008-07-01

    As part of the reactor dynamics activities of FZK/IRS, the qualification of best-estimate coupled code systems for reactor safety evaluations is a key step toward improving their prediction capability and acceptability. The VVER-1000 Coolant Transient Benchmark Phase 1 represents an excellent opportunity to validate the simulation capability of the coupled code system RELAP5/PACRS regarding both the thermal hydraulic plant response (RELAP5) using measured data obtained during commissioning tests at the Kozloduy nuclear power plant unit 6 and the neutron kinetics models of PARCS for hexagonal geometries. The Phase 1 is devoted to the analysis of the switching on of one main coolant pump while the other three pumps are in operation. It includes the following exercises: (a) investigation of the integral plant response using a best-estimate thermal hydraulic system code with a point kinetics model (b) analysis of the core response for given initial and transient thermal hydraulic boundary conditions using a coupled code system with 3D-neutron kinetics model and (c) investigation of the integral plant response using a best-estimate coupled code system with 3D-neutron kinetics. Already before the test, complex flow conditions exist within the RPV e.g. coolant mixing in the upper plenum caused by the reverse flow through the loop-3 with the stopped pump. The test is initiated by switching on the main coolant pump of loop-3 that leads to a reversal of the flow through the respective piping. After about 13 s the mass flow rate through this loop reaches values comparable with the one of the other loops. During this time period, the increased primary coolant flow causes a reduction of the core averaged coolant temperature and thus an increase of the core power. Later on, the power stabilizes at a level higher than the initial power. In this analysis, special attention is paid on the prediction of the spatial asymmetrical core cooling during the test and its effects on the

  12. Investigations of the VVER-1000 coolant transient benchmark phase 1 with the coupled code system RELAP5/PARCS

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez-Espinoza, Victor Hugo

    2008-07-15

    As part of the reactor dynamics activities of FZK/IRS, the qualification of best-estimate coupled code systems for reactor safety evaluations is a key step toward improving their prediction capability and acceptability. The VVER-1000 Coolant Transient Benchmark Phase 1 represents an excellent opportunity to validate the simulation capability of the coupled code system RELAP5/PACRS regarding both the thermal hydraulic plant response (RELAP5) using measured data obtained during commissioning tests at the Kozloduy nuclear power plant unit 6 and the neutron kinetics models of PARCS for hexagonal geometries. The Phase 1 is devoted to the analysis of the switching on of one main coolant pump while the other three pumps are in operation. It includes the following exercises: (a) investigation of the integral plant response using a best-estimate thermal hydraulic system code with a point kinetics model (b) analysis of the core response for given initial and transient thermal hydraulic boundary conditions using a coupled code system with 3D-neutron kinetics model and (c) investigation of the integral plant response using a best-estimate coupled code system with 3D-neutron kinetics. Already before the test, complex flow conditions exist within the RPV e.g. coolant mixing in the upper plenum caused by the reverse flow through the loop-3 with the stopped pump. The test is initiated by switching on the main coolant pump of loop-3 that leads to a reversal of the flow through the respective piping. After about 13 s the mass flow rate through this loop reaches values comparable with the one of the other loops. During this time period, the increased primary coolant flow causes a reduction of the core averaged coolant temperature and thus an increase of the core power. Later on, the power stabilizes at a level higher than the initial power. In this analysis, special attention is paid on the prediction of the spatial asymmetrical core cooling during the test and its effects on the

  13. Measurements at the RA Reactor related to the VISA-2 project - Part 1, Start-up of the RA reactor and measurement of new RA reactor core parameters; Fizicka merenja na reaktoru RA u vezi projekta VISA-2 - I deo, Pustanje u rad reaktora RA i merenje fizickih parametara novog jezgra reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Markovic, H [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1962-07-15

    The objective of the measurements was determining the neutron flux in the RA reactor core. Since the number of fuel channels is increased from 56 to 68 within the VISA-2 project, it was necessary to attain criticality of the RA reactor and measure the neutron flux properties. The 'program of RA reactor start-up' has been prepared separately and it is included in this report. Measurements were divided in two phases. First phase was measuring of the neutron flux after the criticality was achieved but at zero power. During phase two measurements were repeated at several power levels, at equilibrium xenon poisoning. This report includes experimental data of flux distributions and absolute values of the thermal and fast neutron flux in the RA reactor experimental channels and values of cadmium ratio for determining the neutron epithermal flux. Data related to calibration of regulatory rods for cold un poisoned core are included. [Serbo-Croat] Svrha merenja je odredjivanje neutronskog fluksa u reaktoru RA. S obzirom na uvecani broj tehnoloskih kanala of 56 na 68 u vezi projekta VISA-2, bilo je potrebno ponovo dovesti reaktora RA do kriticnosti i izvrsiti merenja karakteristika fluksa neutrona. Posebno je pripremljen 'program pustanja u pogon reaktora RA', koji je sadrzan u ovom dokumentu. Program merenja bio je podeljen na dve faze. Prva faza je merenje fluksa pre podizanju reaktora na nominalnu snagu. Slicna merenja vrsena su i na vecim snagama u drugoj fazi, pod uslovima ravnoteznog zatrovanja reaktora ksenonom, jer se tada pokazuju izvesne promene u odgovarajucim karakteristikama fluksa neutrona. Ovaj izvestaj sadrzi merene vrednosti raspodele fluksa i apsolutne vrednosti termalnih i brzih neutrona kao i kadmijumskih odnosa koji su korisceni za odredjivanje fluksa epitermalnih neutrona. Opisana je kalibracija regulacionih sipki za hladan nezatrovan reaktor.

  14. Design and neutronic investigation of the Nano fluids application to VVER-1000 nuclear reactor with dual cooled annular fuel

    International Nuclear Information System (INIS)

    Ansarifar, G.R.; Ebrahimian, M.

    2016-01-01

    Highlights: • The change in neutronic parameters to the use of nanofluid as coolant is presented. • Nanoparticle deposition on fuel clad is investigated. • Radial and axial local power peaking factors are presented. • ZrO 2 and Al 2 O 3 have the lowest rate of K eff drop off. - Abstract: Nowadays, many efforts have been made to improve the efficiency of nuclear power plants. One of which is use of the dual cooled annular fuel which is an internally and externally cooled annular fuel with many advantages in heat transfer characteristics. Another is the use of nanoparticle/water (nanofluid) as coolant. In this paper, by combining these two methods, the change in neutronic parameters of the VVER-1000 nuclear reactor core with dual cooled annular fuel attributable to the use of nanoparticle/water (nanofluid) as coolant is presented. Optimization of type and volume fraction of nanoparticles in water that affect the safety enhancement of core primary parameters is intended in this study. Reactivity change, radial and axial local power peaking factors (LPPF), and the consequence of nanoparticle deposition on fuel clad are investigated. As a result of changing the effective multiplication factor and PPF calculations for six types of nanoparticles which have been studied extensively for their heat transfer properties including Alumina, Aluminum, Copper oxide, Copper, Titania, and Zirconia with different volume fractions, it can be concluded that at low concentration (0.03 volume fraction), Zirconia and Alumina are the optimum nanoparticles for normal operation. The maximum radial and axial PPF are found to be invariant to the type of nanofluid at low volume fractions. With an increase in nanoparticle deposition thickness on the outer and inner clad, a flux and K eff depression occurred and ZrO 2 and Al 2 O 3 have the lowest rate of drop off.

  15. The review of the reactor physics experiments carried out on the LR-0 research reactor NRI Rez plc for reactors of the VVER type

    International Nuclear Information System (INIS)

    Hudec, Frantisek; Jansky, Bohumil; Juricek, Vlastimil; Mikus, Jan; Novak, Evzen; Osmera, Bohumil; Posta, Severin; Rypar, Vojtech; Svadlenkova, Marie

    2010-01-01

    LR-0 is an experimental zero power reactor mainly used for the determination of the neutron-physical characteristics of WWER and PWR type reactor lattices and shielding with UO2 or MOX fuel. Its major assets include capability to design and operate multizone cores, i.e. substituted cores, with an inner inserted part in hexagonal or square geometry (driven by LR-0 standard assemblies); Standard and special supporting plates for mock-up experiments; special supporting plates, which enables the triangular symmetrical assembly arrangement with an arbitrary pitch; Modeling neutron field parameters of power reactors; Wide range benchmarking possibilities, with high reproducibility of the benchmark design parameters; Wide range of measurement techniques including equipment and experienced personal; Flexible rearrangements of the core. The main experiments included: Pin wise flux distribution measurements; VVER-440 and VVER-1000 mock-ups; compact spent fuel storage; space kinetics experiment; core parameters experimental determination; experiment with new design fuel assembly; WWER-440 control assembly influence; and burnable absorber study. International research projects are also described. (P.A.)

  16. Application of Integral Ex-Core and Differential In-Core Neutron Measurements for Adjustment of Fuel Burn-Up Distributions in VVER-1000

    Science.gov (United States)

    Borodkin, Pavel G.; Borodkin, Gennady I.; Khrennikov, Nikolay N.

    2010-10-01

    The paper deals with calculational and semi-analytical evaluations of VVER-1000 reactor core neutron source distributions and their influence on measurements and calculations of the integral through-vessel neutron leakage. Time-integrated neutron source distributions used for DORT calculations were prepared by two different approaches based on a) calculated fuel burn-up (standard routine procedure) and b) in-core measurements by means of SPD & TC (new approach). Taking into account that fuel burn-up distributions in operating VVER may be evaluated now by analytical methods (calculations) only it is needed to develop new approaches for testing and correction of calculational evaluations. Results presented in this paper allow to consider a reverse task of alternative estimation of fuel burn-up distributions. The approach proposed is based on adjustment (fitting) of time-integrated neutron source distributions, and hence fuel burn-up patterns in some part of reactor core, on the base of ex-core neutron leakage measurement, neutron-physical calculation and in-core SPD & TC measurement data.

  17. RELAP5/MOD3.2 investigation of reactor vessel YR line capabilities for primary side depressurization during the TLFW in VVER1000/V320

    International Nuclear Information System (INIS)

    Gencheva, Rositsa V.; Stefanova, Antoaneta E.; Groudev, Pavlin P.

    2005-01-01

    During the development of Symptom Based Emergency Operating Procedures (SB-EOPs) for VVER-1000/V320 units at Kozloduy Nuclear Power Plant (NPP), a number of analyses have been performed using the RELAP5/MOD3.2 computer code. One of them is 'Investigation of reactor vessel YR line capabilities for primary side depressurization during the Total Loss of Feed Water (TLFW)'. The main purpose of these calculations is to evaluate the capabilities of YR line located at the top of the reactor vessel for primary side depressurization to the set point of High Pressure Injection System (HPIS) actuation and the abilities for successful core cooling after Feed and Bleed procedure initiation. For the purpose of this, operator action with 'Reactor vessel off-gas valve - 0.032 m' opening has been investigated. RELAP5/MOD3.2 computer code has been used to simulate the TLFW transient in VVER-1000 NPP model. This model was developed at Institute for Nuclear Research and Nuclear Energy - Bulgarian Academy of Sciences (INRNE-BAS), Sofia, for analyses of operational occurrences, abnormal events, and design basis scenarios. The model provides a significant analytical capability for the specialists working in the field of NPP safety

  18. The Plinius/Colima CA-U3 test on fission-product aerosol release over a VVER-type corium pool

    International Nuclear Information System (INIS)

    Journeau, Ch.; Piluso, P.; Correggio, P.; Godin-Jacqmin, L.

    2007-01-01

    In a hypothetical case of severe accident in a PWR type VVER-440, a complex corium pool could be formed and fission products could be released. In order to study aerosols release in terms of mechanisms, kinetics, nature or quantity, and to better precise the source term of VVER-440, a series of experiments have been performed in the Colima facility and the test Colima CA-U3 has been successfully performed thanks to technological modifications to melt a prototypical corium at 2760 C degrees. Specific instrumentation has allowed us to follow the evolution of the corium melt and the release, transport and deposition of the fission products. The main conclusions are: -) there is a large release of Cr, Te, Sr, Pr and Rh (>95%w), -) there is a significant release of Fe (50%w), -) there is a small release of Ba, Ce, La, Nb, Nd and Y (<90%w), -) there is a very small release of U in proportion (<5%w) but it is one of the major released species in mass, and -) there is no release of Zr. The Colima experimental results are consistent with previous experiments on irradiated fuels except for Ba, Fe and U releases. (A.C.)

  19. Evaluation of VVER-1200/V-491 reactor pressure vessel integrity during large break LOCA along with SBO using MELCOR 1.8.6

    International Nuclear Information System (INIS)

    Bui Thi Hoa; Tran Chi Thanh

    2015-01-01

    After Fukushima accident and stress test recommended by IAEA for existing reactors, higher safety requirements are enforced upon nuclear power plants during design extension and severe accident conditions. Based on those arguments, Vietnam Government requests a lot of effective safety solutions, in designs proposed for the nuclear power plants in Ninh Thuan province of Vietnam, which can prevent the accident progression toward severe accidents and mitigate severe accident consequences. One of safety requirements is related to delay time of core melt during design extension condition. Especially, if the worst case of accidents occurs, the reactor vessel integrity must be maintained at least 24 hours from the beginning of the accident. With the aim at investigation of Reactor Pressure Vessel (RPV) integrity, in this study, MELCOR 1.8.6 code is used to evaluate the integrity of RPV lower head for VVER-1200/V-491 reactor during a Large Break Loss of Coolant Accident (LBLOCA) in combination with Station Blackout (SBO) event. The study figures out several parameters related to melt down progress such as: rupture position and rupture timing, the amount of hydrogen generated. Availability of the second stage hydro-accumulators (HA2) in the VVER-1200/V-491 is assumed as an additional improvement to delay the timing of core melt as well as to maintain the vessel integrity for long-term. (author)

  20. The experimental definition of the acoustic standing wave series shapes, formed in the coolant of the primary circuit of VVER-440 type reactor

    International Nuclear Information System (INIS)

    Bulavin, V.V.; Pavelko, V.I.

    1995-01-01

    On the basis of pressure fluctuation measurements in some primary circuit loops at 2 nd Unit of Kola NPP with VVER-440 type reactors, the shapes of acoustic standing waves (ASW) were determined at frequencies corresponding to four minimal oscillation eigenfrequencies in the primary circuit coolant. On identification of the ASW modes and properties, experimental results based on six circulating loops in symmetric arrangement allowed determination of the three-dimensional space structure of the wave nodes and antinodes inside and outside of the reactor vessel (RV). As part of this analysis, the geometric features of the primary circuit that caused the formation of these standing waves were identified. Differences in each ASW shape were shown to cause different individual effects on the neutron field in the reactor core and on fuel assembly vibration. This has been partially confirmed by ex-core neutron ionization chamber noise analysis. One type of ASW, possessing an antinode inside the RV, can be used for measurement of the pressure coefficient of reactivity. However, this must be done with care to avoid the potential for incorrect results in some cases. The results presented in this paper can be readily extended to other VVER type reactors with both odd and even number of loops. (author)

  1. Three-dimensional neutron kinetics-thermal-hydraulics VVER 1000 main steam line break analysis by RELAP5-3D code

    International Nuclear Information System (INIS)

    Frisani, A.; Parisi, C.; D'Auria, F.

    2007-01-01

    After the development and the assessment of Three-Dimensional (3D) Neutron Kinetics (NK) - 1D Thermal-Hydraulics (TH) coupled codes analyses methods, deterministic nuclear safety technology is nowadays producing noticeable efforts for the validation of 3D NK - 3D TH coupled codes analyses methods too. Thus, the purpose of this work was to address the capability of the RELAP5-3D 3D NK-3D TH code to reproduce VVER 1000 Nuclear Power Plant (NPP) core dynamic in simulating the mixing effects that could happen in the vessel downcomer and lower plenum during some scenarios. The work was developed in three steps. The first step dealt with the 3D TH modeling of the Kozloduy-6 VVER 1000 reactor pressure vessel. Then this model was validated following a Steam Generator Isolation transient. The second step has been the development of a 3D NK nodalization for the reactor core region. Then the 3D NK model was directly coupled with the previously developed 3D TH model. The third step was the calculation of a Main Steam Line Break (MSLB) transient. The 3D NK global nuclear parameters were then compared with the 0-D results showing a good agreement; nevertheless only the 3D NK- 3D TH model allowed the calculation of each single assembly power trend for this strong NK-TH asymmetric transient. (author)

  2. Investigation of circulating temperature fluctuations of the primary coolant in order to develop an enhanced MTC estimator for VVER-440 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kiss, Sandor; Lipcsei, Sandor [Hungarian Academy of Sciences, Budapest (Hungary). Centre for Energy Research - MTA

    2017-09-15

    Our aim was to develop a method based on noise diagnostics for the estimation of the moderator temperature coefficient of reactivity (MTC) for the Paks VVER-440 units in normal operation. The method requires determining core average neutron flux and temperature fluctuations. The circulation period of the primary coolant, transfer properties of the steam generators, as well as the source and the propagation of the temperature perturbations and the proportions of the perturbation components were investigated in order to estimate the feedback caused by the circulation of the primary coolant. Finally, after developing the new MTC estimator, determining its frequency range and optimal parameters, trends were produced based on an overall evaluation of measurements made with standard instrumentation during a one-year-long period at Paks NPP.

  3. Main results on pilot operation during 5 years of the 3rd generation fuel in VVER-440 reactors of Kola NPP

    International Nuclear Information System (INIS)

    Saprykin, V.; Sumarokov, M.; Gagarinskiy, A.; Sumarokova, A.; Adeev, V.

    2015-01-01

    In the report the results of comparison of main neutron-physical data of exploitation of nuclear fuel are presented for the average enrichment (on U - 235) of 4.87 for the 2nd and 3rd (12 piece) generations with the results of calculations by the complex of the programs KASKAD for 5 fuel loadings of Kola NPP Unit 4 with the reactor VVER- 440. The basic feature of fuel of the 3rd generation as compared with the 2nd is a presence of ribs of inflexibility at corners instead of cover of the fuel assembly and also the increased amount of uranium. The arrangement of fuel rods with different enrichment in fuel assemblies of the 2nd and 3rd generations is chosen identical for the convenient comparison of neutronic and thermohydraulic characteristics of the fuel of different generations. The fuel of 3rd generation was situated in the core symmetrically to the fuel of 2nd one

  4. AER Working Group D on VVER safety analysis minutes of the meeting in Rez, Czech Republic 18-20 May 1998

    International Nuclear Information System (INIS)

    Siltanen, P.

    1998-01-01

    AER Working Group D on VVER reactor safety analysis held its seventh meeting in Hotel Vltava in Rez near Prague during the period 18-20 May 1998. There were altogether 11 participants from 8 member organisations. The coordinator for the working group, Mr. P. Siltanen (IVO) served as chairman. In addition to the general information exchange on recent activities, the topics of the meeting included: First review of solutions to the 3-dimensional AER Dynamic Benchmark Problem No. 5 on a steam line break accident. This benchmark involves a break of the main steam header. Safety analysis of reactivity events. Recent code development work and fuel behaviour. Coolant mixing calculations and experiments related to diluted slugs. A list of participants and a list of handouts distributed at the meeting are attached to the minutes. (author)

  5. Effect of a dual-purpose cask payload increment of spent fuel assemblies from VVER 1000 Bushehr Nuclear Power Plant on basket criticality

    International Nuclear Information System (INIS)

    Rezaeian, M.; Kamali, J.

    2017-01-01

    Dual-purpose casks can be utilized for dry interim storage and transportation of the highly radioactive spent fuel assemblies (SFAs) of Bushehr Nuclear Power Plant (NPP). Criticality safety analysis was carried out using the MCNP code for the cask containing 12, 18, or 19 SFAs. The basket materials of borated stainless steel and Boral (Al-B_4C) were investigated, and the minimum required receptacle pitch of the basket was determined. - Highlights: • Criticality safety analysis for a dual purpose cask was carried out. • The basket material of borated stainless steel and Boral were investigated. • Minimum receptacle pitch was determined for 12, 18, or 19 VVER 1000 spent fuel assemblies.

  6. Problems and prospects connected with development of high-temperature filtration technology at nuclear power plants equipped with VVER-1000 reactors

    Science.gov (United States)

    Shchelik, S. V.; Pavlov, A. S.

    2013-07-01

    Results of work on restoring the service properties of filtering material used in the high-temperature reactor coolant purification system of a VVER-1000 reactor are presented. A quantitative assessment is given to the effect from subjecting a high-temperature sorbent to backwashing operations carried out with the use of regular capacities available in the design process circuit in the first years of operation of Unit 3 at the Kalinin nuclear power plant. Approaches to optimizing this process are suggested. A conceptual idea about comprehensively solving the problem of achieving more efficient and safe operation of the high-temperature active water treatment system (AWT-1) on a nuclear power industry-wide scale is outlined.

  7. Simulation of long-term cooling in the VVER-640 power plant after a large break LOKA on the PACTEL facility

    International Nuclear Information System (INIS)

    Banati, J.

    2000-01-01

    The present report gives a short introduction to the safety features of the new Russian VVER-640 reactor design. In order to analyze the complex thermal hydraulic phenomena during long-term cooling after a large-break LOCA, experiments will be carried out in the PACTEL facility. For preparation, pre-test calculations were performed using the RELAPS/MOD3.2 computer code. The main part of the report discusses the results obtained by the program. The structure and options used in the input deck, as well as the efforts of code application to the simulation of proposed experiments are reviewed. A short sensitivity study is provided on the calculated results. Finally, conclusions are drawn for the code capabilities to represent the expectable trends in the upcoming tests. (orig.)

  8. Występowanie cech kofeinizmu u młodzieży i młodych dorosłych deklarujących częste spożycie bezalkoholowych napojów typu cola zawierających kofeinę

    Directory of Open Access Journals (Sweden)

    Sandra Gustek

    2011-12-01

    Full Text Available Wprowadzenie: W ciągu ostatnich lat obserwuje się znaczny wzrost spożycia napojów zawierających kofeinę, szczególnie wśród dzieci i młodzieży. Tendencja ta jest niepokojąca, a liczne badania wskazują na negatywny wpływ nadmiernej podaży kofeiny na różne aspekty funkcjonowania człowieka. W celu głębszej analizy tego zjawiska przeprowadzono badanie własne i dokonano próby oceny problemu uzależnienia od kofeiny wśród młodzieży i młodych dorosłych ze szczególnym uwzględnieniem cech kofeinizmu u osób deklarujących czę- ste spożycie napojów typu cola. Materiał i metodyka: Na podstawie przeglądu literatury dotyczącej kofeinizmu skonstruowano autorski kwestionariusz ankiety. Określono następujące kryteria włączenia do badania: 1 deklaracja o spożywaniu napojów typu coca-cola, 2 wiek 13-30 lat. W badaniu metodą sondażu diagnostycznego wzięło udział 118 osób. Dodatkowo kontrolowano także zmienne demograficzne, tj. wiek, płeć oraz wzrost i masę ciała. Wyniki: Najczęściej deklarowane przez respondentów potencjalne objawy kofeinizmu to: bóle głowy, drażliwość i wybuchowość, problemy ze snem, przyspieszona praca serca i palpitacje, przy czym symptomy te uzależnione były od wieku. Wnioski: Nadmierne spożycie produktów zawierających kofeinę może być ważnym czynnikiem środowiskowym w kształtowaniu i utrzymywaniu się chorób przewlekłych oraz uzależnień, a także wiązać się ze złymi zachowaniami żywieniowymi. Dlatego tak istotne są profilaktyka i zwiększanie świadomości dzieci, młodzieży oraz ich rodziców na temat szkodliwości nadmiernej podaży produktów zawierających kofeinę, a także dalsze badania nad wpływem kofeiny na funkcjonowanie psychofizyczne tej grupy wiekowej.

  9. The Plinius/Colima CA-U3 test on fission-product aerosol release over a VVER-type corium pool; L'essai Plinius/Colima CA-U3 sur le relachement des aerosols de produits de fission au-dessus d'un bain de corium de type VVER

    Energy Technology Data Exchange (ETDEWEB)

    Journeau, Ch.; Piluso, P.; Correggio, P.; Godin-Jacqmin, L

    2007-07-01

    In a hypothetical case of severe accident in a PWR type VVER-440, a complex corium pool could be formed and fission products could be released. In order to study aerosols release in terms of mechanisms, kinetics, nature or quantity, and to better precise the source term of VVER-440, a series of experiments have been performed in the Colima facility and the test Colima CA-U3 has been successfully performed thanks to technological modifications to melt a prototypical corium at 2760 C degrees. Specific instrumentation has allowed us to follow the evolution of the corium melt and the release, transport and deposition of the fission products. The main conclusions are: -) there is a large release of Cr, Te, Sr, Pr and Rh (>95%w), -) there is a significant release of Fe (50%w), -) there is a small release of Ba, Ce, La, Nb, Nd and Y (<90%w), -) there is a very small release of U in proportion (<5%w) but it is one of the major released species in mass, and -) there is no release of Zr. The Colima experimental results are consistent with previous experiments on irradiated fuels except for Ba, Fe and U releases. (A.C.)

  10. Próba ukierunkowania rehabilitacji u pacjentów ze stwardnieniem rozsianym w zależności od typu klinicznego choroby = The attempt conditions for the rehabilitation of patients with multiple sclerosis, depending on the type of clinical disease

    Directory of Open Access Journals (Sweden)

    Wojciech Garczyński

    2016-07-01

    rednim czasie trwania choroby z podziałem na formy kliniczne (odpowiednio p<0,001 i p<0,01. Na podstawie testu Kruskala - Wallisa stwierdzono istotne statystycznie różnice w poziomie niewydolności ruchowej - EDSS (p<0,001, zmęczenia – MFIS (p<0,05, równowagi – EQUI SCALE (p<0,001 i niesprawności – MDST (p<0,01 w typach klinicznych stwardnienia rozsianego. WNIOSKI. 1. Poziom zmęczenia, zakres zaburzeń równowagi, poziom niewydolności ruchowej oraz niesprawności zależą od typu klinicznego SM. 2. Poziom objawów lękowych i depresyjnych nie wykazuje zróżnicowania pod względem typu klinicznego SM. 3. Pomimo występowania pewnego zróżnicowania poziomu wybranych funkcji fizycznych i społecznych zależnie od typu SM, trudnym wydaje się ukierunkowanie rehabilitacji w oparciu o te założenia.       INTRODUCTION. Multiple sclerosis (eng. Multiple sclerosis, MS is an inflammatory-demyelinating disease of the central nervous system, wherein the multifocal changes occur. Etiopathogenesis is related to genetics and immune. Lesions are characterized by rapid progression and are mainly found in the white matter, brainstem, cerebellum, spinal cord and optic nerve. In remission, remyelination occurs, following which the inflammation subsides. Multiple sclerosis can run with or without relapses. Clinical form of primary progressive (PPMS, multiple sclerosis occurring without relapses, other clinical forms: relapsing - remitting (RRMS, secondary progressive (SPMS and progressive - relapsing (PRMS are characterized by the appearance of relapses, a form of benign (BNMS may be present without relapses. OBJECTIVES. The main objective of this study was to examine correlations between the diversity of physical and social functions in patients with multiple sclerosis and the type and form of the disease. Possible to find such relations, in consequence,  would enable an early orientation of rehabilitation process for needs of patients with a particular type and form of the

  11. Experimental investigation of am measures and effect of hydro-accumulator initial pressure for VVER-440 plants

    International Nuclear Information System (INIS)

    Ivan Toth; Gyorgy Ezsol; Attila Guba; Laszlo Perneczky

    2005-01-01

    Full text of publication follows: A series of experiments were performed at the PMK-2 test facility within the IMPAM-VVER project of the EU 5. Framework Programme. The PMK-2 integral-type facility is a scaled down model of the Paks NPP with a volume and power scaling of 1:2070. Transients can be started from nominal operating conditions. The ratio of elevations is 1:1 except for the lower plenum and pressurizer. The six loops of the plant are modelled by a single active loop. The main objective of the project was to address different problems encountered during the development of EOPs for the Paks NPP in Hungary. Two of the six PMK tests addressed the investigation of starting criteria for primary and secondary bleed during a small break LOCA without HPIS: - a 'base case', with bleed actions following the plant procedures; - a run with secondary and primary bleed started as early as possible. Further two tests investigated the effect of nominal and reduced initial hydro-accumulator pressures on the process, the main question being, whether the starting pressure of the LPIS can be reached without significant overheating of the fuel. These latter were run from lowered initial system pressure in order to be compared to similar tests performed in the project at the PACTEL facility. The two first tests confirmed tendencies shown by earlier plant calculations that neither the secondary nor the primary bleed is effective enough to reduce the pressure, even if their earliest possible actuation is envisaged. As a consequence, low pressure injection could not be started in time to avoid severe fuel rod heat-up and the core power had to be cut in both tests. Comparing the results of tests 3 and 4 the beneficial effect of lowered HA pressure could be analysed. Although heater rod temperatures started to rise also in this test after hydro-accumulators were empty, the secondary and primary bleed actions resulted in the primary pressure dropping to 0.7 MPa and LPIS injection

  12. Qualification of the nuclear reactor core model DYN3D coupled to the thermohydraulic system code ATHLET, applied as an advanced tool for accident analysis of VVER-type reactors. Final report

    International Nuclear Information System (INIS)

    Grundmann, U.; Kliem, S.; Krepper, E.; Mittag, S; Rohde, U.; Schaefer, F.; Seidel, A.

    1998-03-01

    The nuclear reactor core model DYN3D with 3D neutron kinetics has been coupled to the thermohydraulic system code ATHLET. In the report, activities on qualification of the coupled code complex ATHLET-DYN3D as a validated tool for the accident analysis of russian VVER type reactors are described. That includes: - Contributions to the validation of the single codes ATHLET and DYN3D by the analysis of experiments on natural circulation behaviour in thermohydraulic test facilities and solution of benchmark tasks on reactivity initiated transients, - the acquisition and evaluation of measurement data on transients in nuclear power plants, the validation of ATHLET-DYN3D by calculating an accident with delayed scram and a pump trip in VVER plants, - the complementary improvement of the code DYN3D by extension of the neutron physical data base, implementation of an improved coolant mixing model, consideration of decay heat release and xenon transients, - the analysis of steam leak scenarios for VVER-440 type reactors with failure of different safety systems, investigation of different model options. The analyses showed, that with realistic coolant mixing modelling in the downcomer and the lower plenum, recriticality of the scramed reactor due to overcooling can be reached. The application of the code complex ATHLET-DYN3D in Czech Republic, Bulgaria and the Ukraine has been started. Future work comprises the verification of ATHLET-DYN3D with a DYN3D version for the square fuel element geometry of western PWR. (orig.) [de

  13. VVER-1000 dominance ratio

    International Nuclear Information System (INIS)

    Gorodkov, S.

    2009-01-01

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

  14. A study of the effects of changing burn-up and gap gaseous compound on the gap convection coefficient (in a hot fuel pin) in VVER-1000 reactor

    International Nuclear Information System (INIS)

    Rahgoshay, M.; Rahmani, Y.

    2007-01-01

    In this article we worked on the result and process of calculation of the gap heat transfer coefficient for a hot fuel pin in accordance with burn-up changes in the VVER-1000 reactor at the Bushehr nuclear power plant (Iran). With regard to the fact that in calculating the fuel gap heat transfer coefficient, various parameters are effective and the need for designing a model is being felt, therefore, in this article we used Ross and Stoute gap model to study impacts of different effective parameters such as thermal expansion and gaseous fission products on the h gap change rate. Over time and with changes in fuel burn-up some gaseous fission products such as xenon, argon and krypton gases are released to the gas mixture in the gap, which originally contained helium. In this study, the composition of gaseous elements in the gap volume during different times of reactor operation was found using ORIGEN code. Considering that the thermal conduction of these gases is lower than that of helium, and by using the Ross and Stoute gap model, we find first that the changes in gaseous compounds in the gap reduce the values of gap thermal conductivity coefficient, but considering thermal expansion (due to burn-up alterations) of fuel and clad resulting in the reduction of gap thickness we find that the gap heat transfer coefficient will augment in a broad range of burn-up changes. These changes result in a higher rate of gap thickness reduction than the low rate of decrease of heat conduction coefficient of the gas in the gap during burn-up. Once these changes have been defined, we can proceed with the analysis of the results of calculations based on the Ross and Stoute model and compare the results obtained with the experimental results for a hot fuel pin as presented in the final safety analysis report of the VVER-1000 reactor at Bushehr. It is noteworthy that the results of accomplished calculations based on the Ross and Stoute model correspond well with the existing

  15. Analysis of noncondensable effect during small break transient in VVER-440 geometry with CATHARE V1.3L. Preliminary results

    International Nuclear Information System (INIS)

    Sarrette, C.

    1996-11-01

    The report presents a study of the transport and dissolution-release of non-condensable gas into the fluid of the primary loop for the VVER-440 geometry. The analysis has been done using a new model developed for the CATHARE thermal hydraulic code. Results are presented, obtained from calculations of small break loss-of-coolant (SBLOCA) accidents for the Loviisa nuclear power plant (NPP) geometry. The influence of nitrogen dissolved in the water of the accumulators of the emergency core coolant system (ECCS) on natural circulation is discussed. Possibilities of formation of nitrogen bubbles in the main vessels upper plenum, top of the downcomer, steam generators collectors, and upper structures of RCP's are investigated. First results show that there is potentiality for interruption, mainly due to the presence of nitrogen in the top of the downcomer and the upper parts of the RCP's. These preliminary results should be confirmed by carrying out calculations now prematurely stopped for numerical reasons. (8 refs.)

  16. Behavior of a VVER fuel element tested under severe accident conditions in the CORA facility. Test results of experiment CORA-W1

    International Nuclear Information System (INIS)

    Hagen, S.; Hofmann, P.; Noack, V.; Schanz, G.; Schumacher, G.; Sepold, L.

    1994-01-01

    Test bundle CORA-W1 was without absorber material. As in the earlier CORA tests the test bundles were subjected to temperature transients of a slow heatup rate in a steam environment. The transient phases of the test were initiated with a temperature ramp rate of 1 K/s. With these conditions a so-called small-break LOCA was simulated. The temperature escalation due to the exothermal zirconium/niobium-steam reaction started at about 1200 C, leading the bundle to a maximum temperature of approximately 1900 C. With the movement of the melt also heat is transported to the lower region. Below 300 mm elevation the test bundle remained intact due to the axial temeprature distribution. W2 ist characterized by a strong oxidation above 300 mm elevation. Besides the severe oxidation the test bundle resulted in considerable fuel dissolution by ZrNb1/UO 2 interaction in the upper part, complete spacer destruction at 600 mm due to chemical interactions between steel and the ZSrNb1 cladding. Despite some specific features the material behavior of the VVER-1000 bundle is comparable to that observed in the PWR and BWR test using fuel elements typical for Western countries. (orig./HP) [de

  17. Start-up of a cold loop in a VVER-440, the 7th AER benchmark calculation with HEXTRAN-SMABRE-PORFLO

    International Nuclear Information System (INIS)

    Hovi, Ville; Taivassalo, Veikko; Haemaelaeinen, Anitta; Raety, Hanna; Syrjaelahti, Elina

    2017-01-01

    The 7 th dynamic AER benchmark is the first in which three-dimensional thermal hydraulics codes are supposed to be applied. The aim is to get a more precise core inlet temperature profile than the sector temperatures available typically with system codes. The benchmark consists of a start-up of the sixth, isolated loop in a VVER-440 plant. The isolated loop initially contains cold water without boric acid and the start-up leads to a somewhat asymmetrical core power increase due to feedbacks in the core. In this study, the 7 th AER benchmark is calculated with the three-dimensional nodal reactor dynamics code HEXTRAN-SMABRE coupled with the porous computational fluid dynamics code PORFLO. These three codes are developed at VTT. A novel two-way coupled simulation of the 7 th AER benchmark was performed successfully demonstrating the feasibility and advantages of the new reactor analysis framework. The modelling issues for this benchmark are reported and some evaluation against the previously reported comparisons between the system codes is provided.

  18. Start-up of a cold loop in a VVER-440, the 7{sup th} AER benchmark calculation with HEXTRAN-SMABRE-PORFLO

    Energy Technology Data Exchange (ETDEWEB)

    Hovi, Ville; Taivassalo, Veikko; Haemaelaeinen, Anitta; Raety, Hanna; Syrjaelahti, Elina [VTT Technical Research Centre of Finland Ltd, VTT (Finland)

    2017-09-15

    The 7{sup th} dynamic AER benchmark is the first in which three-dimensional thermal hydraulics codes are supposed to be applied. The aim is to get a more precise core inlet temperature profile than the sector temperatures available typically with system codes. The benchmark consists of a start-up of the sixth, isolated loop in a VVER-440 plant. The isolated loop initially contains cold water without boric acid and the start-up leads to a somewhat asymmetrical core power increase due to feedbacks in the core. In this study, the 7{sup th} AER benchmark is calculated with the three-dimensional nodal reactor dynamics code HEXTRAN-SMABRE coupled with the porous computational fluid dynamics code PORFLO. These three codes are developed at VTT. A novel two-way coupled simulation of the 7{sup th} AER benchmark was performed successfully demonstrating the feasibility and advantages of the new reactor analysis framework. The modelling issues for this benchmark are reported and some evaluation against the previously reported comparisons between the system codes is provided.

  19. Comparison of ASTEC 1.3 and ASTEC 1.3 R2 calculations in case of SBO for VVER-1000 reactor

    International Nuclear Information System (INIS)

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

    2009-01-01

    The report presents the results from severe accident analyses performed with the both versions of ASTEC v1.3 and ASTEC v1.3R2 computer code for a VVER 1000 type of reactor. The purpose of this analysis is to assess the progress of ASTEC code modeling of main phenomena arising during hypothetical severe accidents. The final target of these analyses is to estimate the behaviour of the ASTEC code, its capability for simulation of severe accidents, including safety systems and Severe Accident Management (SAM) procedures. The analyses have been performed assuming a station blackout with simultaneous loss of HPIS, LPIS (ECCSs), EFWS and spray system due to failure of DGs. Hydro accumulators are not available. In the calculation it is assumed opening and stuck-open of PRZ relief valves. It has been organized the Fission Products path through the SEMPELL valve. It should be said that this investigation was limited to the 'in-vessel' phase of the sequence; therefore the effect of sprays on containment atmosphere has not been studied. (authors)

  20. Evaluation of containment peak pressure and structural response for a large-break loss-of-coolant accident in a VVER-440/213 NPP

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, B.W.; Sienicki, J.J.; Kulak, R.F.; Pfeiffer, P.A. [Argonne National Lab., IL (United States); Voeroess, L.; Techy, Z. [VEIKI Inst. for Electric Power Research, Budapest (Hungary); Katona, T. [Paks Nuclear Power Plant (Hungary)

    1998-07-01

    A collaborative effort between US and Hungarian specialists was undertaken to investigate the response of a VVER-440/213-type NPP to a maximum design-basis accident, defined as a guillotine rupture with double-ended flow from the largest pipe (500 mm) in the reactor coolant system. Analyses were performed to evaluate the magnitude of the peak containment pressure and temperature for this event; additional analyses were performed to evaluate the ultimate strength capability of the containment. Separate cases were evaluated assuming 100% effectiveness of the bubbler-condenser pressure suppression system as well as zero effectiveness. The pipe break energy release conditions were evaluated from three sources: (1) FSAR release rate based on Soviet safety calculations, (2) RETRAN-03 analysis and (3) ATHLET analysis. The findings indicated that for 100% bubbler-condenser effectiveness the peak containment pressures were less than the containment design pressure of 0.25 MPa. For the BDBA case of zero effectiveness of the bubbler-condenser system, the peak pressures were less than the calculated containment failure pressure of 0.40 MPa absolute.

  1. Application of the Fast Fourier Transform Based Method to assist in the qualification process for the PSB-VVER1000 RELAP5 nodalisation

    International Nuclear Information System (INIS)

    Muellner, N.; Seidelberger, E.; Del Nevo, A.; D'Auria, F.

    2005-01-01

    One dimensional Thermal-Hydraulic-System (TH-SYS) codes like RELAP5 provide a degree of freedom that is significantly greater than desired. An undisciplined code user with some experience usually can achieve any pre-set results by tuning the nodalization. To take some freedom away from the user and achieve code user independent results several strategies were adopted. The approach of the UNIPI is to develop a multi purpose nodalization which must pass a rigorous nodalization qualification process. A qualified nodalization is also the basis to apply the Uncertainty Methodology based on Accuracy Extrapolation (UMAE) or to develop the accuracy database and to apply the Code with capability of Internal Assessment of Uncertainty (CIAU). An important part of the nodalization qualification is to verify the results of the nodalization approach against experimental data. In this context the Fast Fourier Transform Based Method (FFTBM) provides an independent tool to assess the quantitative accuracy of the analysis. This paper will present a series of RELAP5 calculations, each assessed by the FFTBM, which analyze an experiment at the PSB-VVER1000 facility This experiment is a 0.7% Small Break (SB) Loss Of Coolant Accident (LOCA) in the Cold Leg (CL) near the Reactor Pressure Vessel (RPV). The FFTBM was used to establish a range in which parameters like power, break area or total heat losses can vary, while the nodalization is still qualified from a quantitative point of view. (author)

  2. VVER-1000 coolant transient benchmark. Phase 1 (V1000CT-1). Vol. 3: summary results of exercise 2 on coupled 3-D kinetics/core thermal-hydraulics

    International Nuclear Information System (INIS)

    2007-01-01

    In the field of coupled neutronics/thermal-hydraulics computation there is a need to enhance scientific knowledge in order to develop advanced modelling techniques for new nuclear technologies and concepts, as well as current applications. (authors) Recently developed best-estimate computer code systems for modelling 3-D coupled neutronics/thermal-hydraulics transients in nuclear cores and for the coupling of core phenomena and system dynamics need to be compared against each other and validated against results from experiments. International benchmark studies have been set up for this purpose. The present volume is a follow-up to the first two volumes. While the first described the specification of the benchmark, the second presented the results of the first exercise that identified the key parameters and important issues concerning the thermal-hydraulic system modelling of the simulated transient caused by the switching on of a main coolant pump when the other three were in operation. Volume 3 summarises the results for Exercise 2 of the benchmark that identifies the key parameters and important issues concerning the 3-D neutron kinetics modelling of the simulated transient. These studies are based on an experiment that was conducted by Bulgarian and Russian engineers during the plant-commissioning phase at the VVER-1000 Kozloduy Unit 6. The final volume will soon be published, completing Phase 1 of this study. (authors)

  3. Application of the coupled code system KIKO3D/ATHLET to the boron dilution transients in VVER-440 Type NPP

    International Nuclear Information System (INIS)

    Hegyi, G.; Kereszturi, A.; Trosztel, I.

    2003-01-01

    The transient caused by a perturbation of boron concentration and coolant temperature at the inlet of a Russian developed reactor (VVER-440) is reanalysed as a part of the modernisation (introduction of a new type profiled fuel) and power upgrading (up to 108 %) project. This task is one of the basis cases to be investigated in the safety analysis of the pressurized water reactor (PWR) and the criteria of Anticipated Operational Occurrences (AOO) have to be fulfilled for it. First detailed planning calculations were performed with the thermal hydraulic system code ATHLET and neutron physical code system KARATE-440 to find out the appropriate initial parameter set taking into account the active safety system of the NPP. Finally the most reactive case is analysed by the KIKO3D/ATHLET coupled system code. Whereas the investigation is done for safety analysis, conservative assumptions are imposed on reactivity characteristics. Moreover at the core inlet no-mixing is supposed from the unaffected loops. The presented calculations show, how the coupled code system with a detailed description of plant functions and core behaviour can help to understand better the local phenomena in the study of a potential risk of dilution accident, as it offers the possibility to evaluate the plant safety in a more realistic and versatile manner. (author)

  4. Comparisons with measured data of the simulated local core parameters by the coupled code ATHLET-BIPR-VVER applying a new enhanced model of the reactor pressure vessel

    International Nuclear Information System (INIS)

    Nikonov, S.; Pasichnyk, I.; Velkov, K.; Pautz, A.

    2011-01-01

    The paper describes the performed comparisons of measured and simulated local core data based on the OECD/NEA Benchmark on Kalinin-3 NPP: 'Switching off of one of the four operating main circulation pumps at nominal reactor power'. The local measurements of in core self-powered neutron detectors (SPND) in 64 fuel assemblies on 7 axial levels are used for the comparisons of the assemblies axial power distributions and the thermocouples readings at 93 fuel assembly heads are applied for the fuel assembly coolant temperature comparisons. The analyses are done on the base of benchmark transient calculations performed with the coupled system code ATHLET/BIPR-VVER. In order to describe more realistically the fluid mixing phenomena in a reactor pressure vessel a new enhanced nodalization scheme is being developed. It could take into account asymmetric flow behaviour in the reactor pressure vessel structures like downcomer, reactor core inlet and outlet, control rods' guided tubes, support grids etc. For this purpose details of the core geometry are modelled. About 58000 control volumes and junctions are applied. Cross connection are used to describe the interaction between the fluid objects. The performed comparisons are of great interest because they show some advantages by performing coupled code production pseudo-3D analysis of NPPs applying the parallel thermo-hydraulic channel methodology (or 1D thermo-hydraulic system code modeling). (Authors)

  5. Using a combination of weighting factor method and imperialist competitive algorithm to improve speed and enhance process of reloading pattern optimization of VVER-1000 reactors in transient cycles

    Energy Technology Data Exchange (ETDEWEB)

    Rahmani, Yashar, E-mail: yashar.rahmani@gmail.com [Department of Physics, Faculty of Engineering, Islamic Azad University, Sari Branch, Sari (Iran, Islamic Republic of); Shahvari, Yaser [Department of Computer Engineering, Payame Noor University (PNU), P.O. Box 19395-3697, Tehran (Iran, Islamic Republic of); Kia, Faezeh [Golestan Institute of Higher Education, Gorgan 49139-83635 (Iran, Islamic Republic of)

    2017-03-15

    Highlights: • This article was an attempt to optimize reloading pattern of Bushehr VVER-1000 reactor. • A combination of weighting factor method and the imperialist competitive algorithm was used. • The speed of optimization and desirability of the proposed pattern increased considerably. • To evaluate arrangements, a coupling of WIMSD5-B, CITATION-LDI2 and WERL codes was used. • Results reflected the considerable superiority of the proposed method over direct optimization. - Abstract: In this research, an innovative solution is described which can be used with a combination of the new imperialist competitive algorithm and the weighting factor method to improve speed and increase globality of search in reloading pattern optimization of VVER-1000 reactors in transient cycles and even obtain more desirable results than conventional direct method. In this regard, to reduce the scope of the assumed searchable arrangements, first using the weighting factor method and based on values of these coefficients in each of the 16 types of loadable fuel assemblies in the second cycle, the fuel assemblies were classified in more limited groups. In consequence, the types of fuel assemblies were reduced from 16 to 6 and consequently the number of possible arrangements was reduced considerably. Afterwards, in the first phase of optimization the imperialist competitive algorithm was used to propose an optimum reloading pattern with 6 groups. In the second phase, the algorithm was reused for finding desirable placement of the subset assemblies of each group in the optimum arrangement obtained from the previous phase, and thus the retransformation of the optimum arrangement takes place from the virtual 6-group mode to the real mode with 16 fuel types. In this research, the optimization process was conducted in two states. In the first state, it was tried to obtain an arrangement with the maximum effective multiplication factor and the smallest maximum power peaking factor. In

  6. Neutron converter at reactor RB; Konvertor neutrona na reaktoru RB

    Energy Technology Data Exchange (ETDEWEB)

    Strugar, P; Sotic, P; Ninkovic, M; Pesic, M [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

    1977-07-01

    A neutron converter at Reactor RB in the 'Boris Kidric' Institute of Nuclear Sciences - Vinca has been constructed. Preliminary measurements have been shown that the converted neutron spectrum is very similar to the fission neutron spectrum. For the same integral reactor power, the measured neutron radiation dose has been for about ten times larger with the neutron converter. The neutron converter offers wide possibilities, as in investigations in the reactor physics, where the fission neutron spectra have been required, as well as in the field of neutron dosimetry and biological irradiations (author)

  7. Parní generátor reaktoru ESFR

    OpenAIRE

    Bátěk, David

    2012-01-01

    Tato diplomová práce se zabývá parním generátorem pro reaktory ESFR (Evropský Rychlý Sodíkem Chlazený Reaktor), jenž je vyhříván tekutým sodíkem. V úvodních kapitolách jsou teoretické informace o parametrech ESFR a porovnání s výměníky tepla v jaderných elektrárnách pracujících na stejném principu (sodík jako chladivo). Následuje návrhová část, kam patří úvod do problematiky výpočtu, volba materiálu a koncepce výměníku a samotná výpočtová část, která zahrnuje tepelný, hydraulický a pevnostní ...

  8. The passive system for reflooding of the VVER reactor core from the second-stage hydro-accumulators: design and basic design solutions

    International Nuclear Information System (INIS)

    Alexandr D Efanov; Sergey G Kalyakin; Andrey V Morozov; Oleg V Remizov; Vladimir M Berkovich; Victor N Krushelnitskiy; Vladimir G Peresadko; Yuri G Dragunov; Alexey K Podshibyakin; Sergey I Zaitcev

    2005-01-01

    Full text of publication follows: The fundamental difference in the safety assurance of the operating NPPs and those under design implies that the safety in the existing NPPs is achieved by energy-dependent (active) systems and depends on the proficiency of attending personnel. To provide safety, the new NPP designs use the physical processes proceeding in the facility without power supply; and they are unaffected by human errors. As to the safety level, the design of the new generation nuclear power plant NPP-92 relates to the class of the improved NPPs; and it applies a principle of diversity in the structure of systems responsible for critical safety functions. In accordance with the above-mentioned safety concept, the design development required a complex of experimental investigations and numerical modeling to be conducted. Among the passive safety systems of the NPP with RP-392 is the system of the second stage hydro-accumulators (GE-2). The system of the second-stage hydro-accumulators consists of four groups of hydro-accumulating tanks with a total coolant volume of 960 m 3 . The system is intended for the core flooding with coolant during 24 hours. In each group of the hydro-accumulators, the graded coolant flowrate is provided, which depends on residual heat in the reactor. The special check valves are tuned to open at the pressure drop in the circuit below 1.5 MPa. The paper presents the thermalhydraulic substantiation of the serviceability of the second-stage hydro-accumulators system for passive heat removal from the VVER reactor core and the basic design solutions on the GE-2 system. (authors)

  9. VVER-1000 SFAT-specification of an industrial prototype. Interim report on Task FIN A 1073 of the Finnish Support Programme to IAEA Safeguards

    Energy Technology Data Exchange (ETDEWEB)

    Tiitta, A. [VTT Chemical Technology, Espoo (Finland); Dvoyeglazov, A.M.; Iievlev, S.M. [State Scientific and Technical Centre for Nuclear and Radiation Safety, Kiev (Ukraine); Tarvainen, M.; Nikkinen, M. [Radiation and Nuclear Safety Authority, Helsinki (Finland)

    2000-05-01

    The project to develop a Spent Fuel Attribute Tester (SFAT) for Ukrainian VVER-1000 facilities is going on under the Task FIN A 1073 of the Finnish Support Programme to the IAEA safeguards. In the SFAT method the verification is based on an unambiguous detection of gamma radiation of the fission products. This is implemented by detecting the radiation emitted by a fuel assembly with a mobile gamma-spectroscopic instrument consisting of a collimator arrangement and a detector unit. The fuel assemblies stored in a wet storage are not moved during the verification measurement. The principal target is the radiation characteristic to {sup 137}Cs. For short cooled assemblies also {sup 144}Pr can be used as the target fission product nuclide. The generic IAEA SFAT concept has been adapted to the special conditions at the Ukrainian facilities. The requirements of the End User (IAEA), the State Nuclear Safety Authority (NRA) and the facilities have been taken into account and included in the specifications. Since the issuance of the first interim report, additional measurements were conducted at the Zaporozhye NPP to ensure the feasibility of the suggested measurement geometry and to test whether the SFAT device could be operated using the refuelling machine. A clear answer to the optimal measurement geometry and the detector choice was also obtained during this first phase of the task. Basing on the measurement results and the operational experience, the technical specifications for an industrial SFAT prototype are formulated. The technical specifications presented in this report and in the previous report have been approved by the Ukrainian State Authority and one of the facility operators, the Zaporozhye NPP. A procedure has been started for getting the approval of the other Ukrainian operators. (orig.)

  10. Fluktuacja dynamiki i dymorfizm płciowy cech somatycznych, typu budowy i otłuszczenia ciała dzieci i młodzieży w wieku od 4 do 18 lat środowiska wiejskiego regionu warmińsko-mazurskiego

    Directory of Open Access Journals (Sweden)

    Mirosław Mrozkowiak

    2015-07-01

    Mirosław Mrozkowiak   Mirosław Mrozkowiak Bioergonsport, Nowa Biała. Polska e-mail: magmar54@interia.pl strona: http://wadypostawy.republika.pl   Słowa kluczowe: wysokość i masa ciała, wskaźnik BMI, IR.   Streszczenie Wstęp. Mimo sekularnego zwiększenia wysokości ciała we wszystkich grupach i płciach, dystanse międzyśrodowiskowe nie wykazywały tendencji malejących, jednak wszystkie populacje dążą do pewnej stabilizacji. Powszechnie stosowany wskaźnik wagowo-wzrostowym BMI umożliwił diagnostykę otyłości i nadwagi, a wskaźnik IR określenie typu budowy ciała. Celem badań jest określenie dynamiki zmian i dymorfizmu płciowego cech somatycznych, typu budowy i otłuszczenia ciała dzieci i młodzieży w wieku od 4 do 18 lat środowiska wiejskiego, regionu warmińsko-mazurskiego Polski. Metoda i materiał. Z badań uzyskano 8270 obserwacji w tym 3384 wśród dziewcząt i 3786 chłopców. Pomiarów dokonano na wadze lekarskiej z dokładnością do 0,5 cm i 100 g. Wyniki. Statystykę opisową wskaźnika BMI przedstawiono w tab. 2 oraz ryc. 1. Dynamikę zmian i istotność różnic między płciami w kategoriach wiekowych w tab. 3 i 4 oraz ryc. 2. Statystykę opisową wskaźnika IR przedstawiono w tab. 5 i ryc. 3. Dynamikę zmian i istotność różnic między płciami w kategoriach wiekowych w tab. 6 i 7 oraz ryc. 4. W tab. 8 przedstawiono przyjętą interpretację wskaźnika IR, w 9 wskaźnika BMI, w tab. 10, 11 i 12 oraz ryc. od 5 do 9 przedstawiony jest odsetek typów budowy i otłuszczenia występujący w badanej kohorcie. Wnioski. 1. Średnia wysokość i masa ciała chłopców w wieku od 4 do 18 r.ż. jest większa niż dziewcząt w tym samym wieku. 2. Wśród dziewcząt i chłopców dominuje smukły typ budowy ciała. Wśród chłopców w wieku 17 lat występuje tylko typ smukły, a w 18 r.ż. średni. 3. Odsetek z niedowagą utrzymuje się na wysokim poziomie wśród dziewcząt i chłopców do 8 r.ż., później sukcesywnie obni

  11. Modelling of thermal mechanical behaviour of high burn-Up VVER fuel at power transients with special emphasis on the impact of fission gas induced swelling of fuel pellets

    International Nuclear Information System (INIS)

    Novikov, V.; Medvedev, A.; Khvostov, G.; Bogatyr, S.; Kuzetsov, V.; Korystin, L.

    2005-01-01

    This paper is devoted to the modelling of unsteady state mechanical and thermo-physical behaviour of high burn-up VVER fuel at a power ramp. The contribution of the processes related to the kinetics of fission gas to the consequences of pellet-clad mechanical interaction is analysed by the example of integral VVER-440 rod 9 from the R7 experimental series, with a pellet burn-up in the active part at around 60 MWd/kgU. This fuel rod incurred ramp testing with a ramp value ΔW 1 ∼ 250 W/cm in the MIR research reactor. The experimentally revealed residual deformation of the clad by 30-40 microns in the 'hottest' portion of the rod, reaching a maximum linear power of up to 430 W/cm, is numerically justified on the basis of accounting for the unsteady state swelling and additional degradation of fuel thermal conductivity due to temperature-induced formation and development of gaseous porosity within the grains and on the grain boundaries. The good prediction capability of the START-3 code, coupled with the advanced model of fission gas related processes, with regard to the important mechanical (residual deformation of clad, pellet-clad gap size, central hole filling), thermal physical (fission gas release) and micro-structural (profiles of intra-granular concentration of the retained fission gas and fuel porosity across a pellet) consequences of the R7 test is shown. (authors)

  12. Validation of computer codes and modelling methods for giving proof of nuclear saefty of transport and storage of spent VVER-type nuclear fuels. Part 1. Purposes and goals of the project. Final report

    International Nuclear Information System (INIS)

    Buechse, H.; Langowski, A.; Lein, M.; Nagel, R.; Schmidt, H.; Stammel, M.

    1995-01-01

    The report gives the results of investigations on the validation of computer codes used to prove nuclear safety during transport and storage of spent VVER - fuel of NPP Greifswald and Rheinsberg. Characteristics of typical spent fuel (nuclide concentration, neutron source strength, gamma spectrum, decay heat) - calculated with several codes - and dose rates (e.g. in the surrounding of a loaded spent fuel cask) - based on the different source terms - are presented. Differences and their possible reasons are discussed. The results show that despite the differences in the source terms all relevant health physics requirements are met for all cases of source term. The validation of the criticality code OMEGA was established by calculation of appr. 200 critical experiments of LWR fuel, including VVER fuel rod arrangements. The mean error of the effective multiplication factor k eff is -0,01 compared to the experiment for this area of applicability. Thus, the OMEGA error of 2% assumed in earlier works has turned out to be sufficiently conservative. (orig.) [de

  13. Cesta k neo-modernímu typu smrti

    Directory of Open Access Journals (Sweden)

    Lenka Beranová

    2017-10-01

    Full Text Available This text aims to trace the thematisation of death and dying in Sociology, which in the 1990’ evolves into constituion of independent sociological sub-discipline: “sociology of dying and death”. The main intention then is to go beyond the commonly held belief that the death is still a taboo. The arcticle consists of six thematic parts: first part summarizes the occurence of problems of dying and death in sociology, reflecting recent developments not only in social sciences, but also in the society as such. Theretofore the death is interpreted by religion but has been medicalized so that nowadays medicine represents the determinative paradigm of death (second part. This development is in accordance with the most important framework in the intensely individualized society centered around the notion of “self”, closely connected with the concept of body and embodiment (third part. Fourth part summarizes the evolution of new sub-discip-line and the fifth part is concerned with the revival concept of death in the work of British sociologist Tony Walter. On the basis of the comparison of traditional, modern and neo-modern model of death we comment on social phenomena associated with other relevant conceptions. Sixth part is focused on the problems of “good death”.

  14. Application of Concentration-Number and Concentration-Volume Fractal Models to Recognize Mineralized Zones in North Anomaly Iron Ore Deposit, Central Iran / Zastosowanie Modeli Fraktalnych Typu K-L (Koncentracja-Liczba), Oraz K-O (Koncentracja Objętość) Do Rozpoznawania Stref Występowania Surowców Mineralnych W Regionie Złóż Rud Żelaza North Anomaly, W Środkowym Iranie

    Science.gov (United States)

    Afzal, Peyman; Ghasempour, Reza; Mokhtari, Ahmad Reza; Haroni, Hooshang Asadi

    2015-09-01

    Identification of various mineralized zones in an ore deposit is essential for mine planning and design. This study aims to distinguish the different mineralized zones and the wall rock in the Central block of North Anomaly iron ore deposit situated in Bafq (Central Iran) utilizing the concentration-number (C-N) and concentration-volume (C-V) fractal models. The C-N model indicates four mineralized zones described by Fe thresholds of 8%, 21%, and 50%, with zones 50% Fe representing wall rocks and highly mineralized zone, respectively. The C-V model reveals geochemical zones defined by Fe thresholds of 12%, 21%, 43% and 57%, with zones 50% zawartości żelaza, co odpowiada pasmu skalnemu oraz strefie o wysokim stopniu zawartości rudy. Model koncentracja-objętość wskazuje na istnienie stref geochemicznych określonych poprzez progowe wartości zawartości żelaza: 12%, 21%, 43% i 57 % oraz strefy <12%, co odpowiada ścianie skalnej. Obydwa modele stwierdzają obecność stref o wysokim stopniu zawartości surowca w środkowej i zachodniej części złoża. Wyniki walidacji modeli fraktalnych przy użyciu modeli geologicznych wskazują, ze model fraktalny koncentracja-liczba lepiej odwzorowuje obecność stref o wysokiej zawartości rud niż model fraktalny typu koncentracja-objętość.

  15. OECD/DOE/CEA VVER-1000 coolant transient (V1000CT) benchmark - a consistent approach for assessing coupled codes for RIA analysis

    International Nuclear Information System (INIS)

    Boyan D Ivanov; Kostadin N Ivanov; Eric Royer; Sylvie Aniel; Nikola Kolev; Pavlin Groudev

    2005-01-01

    Full text of publication follows: The Rod Ejection Accident (REA) and Main Steam Line Break (MSLB) are two of the most important Design Basis Accidents (DBA) for VVER-1000 exhibiting significant localized space-time effects. A consistent approach for assessing coupled three-dimensional (3-D) neutron kinetics/thermal hydraulics codes for these Reactivity Insertion Accidents (RIA) is to first validate the codes using the available plant test (measured) data and after that perform cross code comparative analysis for REA and MSLB scenarios. In the framework of joint effort between the Nuclear Energy Agency (NEA) of OECD, the United States Department of Energy (US DOE), and the Commissariat a l'Energie Atomique (CEA), France a coupled 3-D neutron kinetics/thermal hydraulics benchmark was defined. The benchmark is based on data from the Unit 6 of the Bulgarian Kozloduy Nuclear Power Plant (NPP). In performing this work the PSU, USA and CEA-Saclay, France have collaborated with Bulgarian organizations, in particular with the KNPP and the INRNE. The benchmark consists of two phases: Phase 1: Main Coolant Pump Switching On; Phase 2: Coolant Mixing Tests and MSLB. In addition to the measured (experiment) scenario, an extreme calculation scenario was defined for better testing 3-D neutronics/thermal-hydraulics techniques: rod ejection simulation with control rod being ejected in the core sector cooled by the switched on MCP. Since the previous coupled code benchmarks indicated that further development of the mixing computation models in the integrated codes is necessary, a coolant mixing experiment and MSLB transients are selected for simulation in Phase 2 of the benchmark. The MSLB event is characterized by a large asymmetric cooling of the core, stuck rods and a large primary coolant flow variation. Two scenarios are defined in Phase 2: the first scenario is taken from the current licensing practice and the second one is derived from the original one using aggravating

  16. Severe accident experiments on PLINIUS platform. Results of first experiments on COLIMA facility related to VVER-440. Presentation of planned VULCANO and KROTOS tests

    International Nuclear Information System (INIS)

    Piluso, P.; Boccaccio, E.; Bonnet, J.-M.; Journeau, C.; Fouquart, P.; Magallon, D.; Ivanov, I.; Mladenov, I.; Kalchev, S.; Grudev, P.; Alsmeyer, H.; Fluhrer, B.; Leskovar, M.

    2005-01-01

    In the hypothetical case of a nuclear reactor severe accident, the reactor core could melt and form a mixture of nuclear fuel (UO 2 + Fission Products), metallic or oxidized cladding + steel, called c orium , of highly refractory oxides (UO 2 , ZrO 2 ) and metallic or oxidized steel, that could eventually flow out of the vessel and mix with the substrate decomposition products (generally oxides such as SiO 2 , Al 2 O 3 , CaO, Fe 2 O 3 ). The French Atomic Energy Commission (CEA) has launched a R and D programme aimed at providing the tools for improving the mastering of severe accidents. It encompasses the development of models and codes, performance of experiments in simulant and prototypic materials and the analysis of international experiments. The experiments with prototypic corium (i.e. material containing depleted UO 2 ) are performed in the PLINIUS experimental platform at CEA Cadarache. It comprises the VULCANO facility for 50-100 kg tests (corium-material interactions, corium solidification etc.), the COLIMA facility for smaller scale (∼1 kg) experiments, the VITI facility for corium properties measurement and the KROTOS facility for corium-water interaction (a few kg). In the framework of the 5 th European Framework Programme, free trans-national access to these facilities has been offered to EU and Associated States researchers. For the first PLINIUS access, COLIMA experiments have been conducted with a Bulgarian Team (TU/SOFIA, BAS/INRNE and NPP/KOZLODUY). This series of tests was devoted to experimental studies on fission products release and corium behaviour in the late phase in a hypothetic case of severe accident in a PWR type VVER-440. The COLIMA experimental results are consistent with previous experiments on irradiated fuels (VERCORS, PHEBUS) with small differences for some fission products and show new results for the remaining corium. For the second visit, scientific users from FZK in Germany were selected to validate the COMET core

  17. RELAP5 / MOD3.2 analysis of INSC standard problem INSCSP - V4 : investigation of heat transfer for partly uncovered VVER-1000 core at the test facility KS (RRC K1)

    International Nuclear Information System (INIS)

    Tentner, A.; Ahrens, J. W.

    2002-01-01

    The RELAP5/MOD3.2 computer program has been used to analyze a series of tests investigating heat-transfer from a partly uncovered VVER-1000 core in the KS test facility at the Russian Research Center ''Kurchatov Institute'' (RRC-KI). The analysis documented represents VVER Standard Problem 4 in Joint Project 6, which is the investigation of Computer Code Validation for Transient Analysis of RBMK and VVER Reactors, between the United States and Russian International Nuclear Safety Centers. The experiment facility and data, RELAP5 nodalization, and results are shown for one of the six tests defined in Standard Problem 4. Only part of the data was analyzed due to our conclusion that the available experimental data is not sufficient to allow the modeling of the actual experiment sequence. The experiment initial conditions were reached through a series of transient processes, about which no quantitative information was available. This has required the modeling of an arbitrary computational transient, with the goal of reaching initial conditions similar to those observed during the experiment. The use of an arbitrary transient introduces many degrees of freedom in the analysis, i.e. initial computational values that influence the entire sequence of events, including the loop behavior during the experiment time window. Reasonable agreement between RELAP5 and the experiment data can be obtained by manipulating a number of initial computational values, including the liquid level in the fuel assembly model, the liquid level in the annular region, the quality of the saturated vapor in the voided loop regions, etc. Our study has focused on exploring the sensitivity of results to changes in these initial values which are not based on experimental information, but are selected with the goal of matching the experimentally observed behavior during the experiment time window. We have shown that changes in several initial arbitrary values can lead to similar changes in the

  18. CFD investigations of natural circulation between the RPV and the cooling pond of VVER-440 type reactors in incidental conditions during maintenance performed with the code CFX-4.3

    International Nuclear Information System (INIS)

    Legradi, G.; Aszodi, A.

    2002-01-01

    During the annual maintenance of the VVER-440 type reactors, the RPV, the cooling pond and the transfer pond form a connected flow domain. The reactor is cooled by the natural circulation, which develops in one or two main loops. The cooling pond has its own cooling loops. CFD calculations have been performed with the CFX-4.3 code to investigate whether it is possible to cool the reactor core in case the main loops are lost and other emergency systems are not available. The results point out that the cooling system of the cooling pond is not capable of cooling the reactor core with the present connection. Therefore, modifications of the cooling system are investigated which would make it suitable for removing the remanent heat from the core.(author)

  19. Calculation of the fuel composition and the thermo-neutronic parameters of the Bushehr’s VVER-1000 reactor during the initial startup and the first cycle using the WIMSD5-B, CITATION-LDI2 and WERL codes

    International Nuclear Information System (INIS)

    Rahmani, Yashar; Pazirandeh, Ali; Ghofrani, Mohammad B.; Sadighi, Mostafa

    2013-01-01

    Highlights: ► In this paper, the changes of the thermo-neutronic parameters of a VVER 1000 reactor were studied during the first cycle. ► The coupling of neutronic and thermo-hydraulic codes was utilized. ► A computational program (WERL code) was designed to calculate the temperature distribution of the reactor core. ► To estimate the concentration of the released gaseous fission products, the Weisman model was used. ► The results of this study enjoyed the desirable accuracy. - Abstract: In this paper, the concentrations of fission products and fuel isotopes as well as the changes of the thermo-neutronic parameters of the Bushehr’s VVER-1000 reactor were studied during the initial startup and the first cycle. In order to perform the time-dependent cell calculations and obtain the concentration of fuel elements, the WIMSD5-B code was used. Besides, by utilizing the CITATION-LDI2 code, the effective multiplication factor and the thermal power distribution of the reactor were calculated. A computer program (WERL code) was designed in order to perform accurate calculation of the temperature distribution of the reactor core. For this purpose, the Ross–Stoute, Weisman, and Lee–Kesler models were used for calculating of the gap conductance coefficient, fission gas release and gap pressure, respectively. The results demonstrated that in designing the startup process, in addition to the role considered for overcoming the power defects and in preparing the required conditions for performing the safety-assurance tests, the flattening of the reactor’s power must be taken into account. Comparison between the results of this modeling and the final safety analysis report of this reactor showed that the results presented in this paper are satisfactorily accurate

  20. Behaviour of a VVER-1000 fuel element with boron carbide/steel absorber tested under severe fuel damage conditions in the CORA facility (Results of experiment CORA-W2)

    International Nuclear Information System (INIS)

    Hagen, S.; Hofmann, P.; Noack, V.; Schanz, G.; Schumacher, G.; Sepold, L.

    1994-10-01

    The 'Severe Fuel Damage' (SFD) experiments of the Kernforschungszentrum Karlsruhe (KfK), Federal Republic of Germany, were carried out in the out-of-pile facility 'CORA' as part of the international Severe Fuel Damage (SFD) research. The experimental program was set up to provide information on the failure mechanisms of Light Water Reactor (LWR) fuel elements in a temperature range from 1200 C to 2000 C and in few cases up to 2400 C. Between 1987 and 1992 a total of 17 CORA experiments with two different bundle configurations, i.e. PWR (Pressurized Water Reactor) and BWR (Boiling Water Reactor) bundles were performed. These assemblies represented 'Western-type' fuel elements with the pertinent materials for fuel, cladding, grid spacer, and absorber rod. At the end of the experimental program two VVER-1000 specific tests were run in the CORA facility with identical objectives but with genuine VVER-type materials. The experiments, designated CORA-W1 and CORA-W2 were conducted on February 18, 1993 and April 21, 1993, respectively. Test bundle CORA-W1 was without absorber material whereas CORA-W2 contained one absorber rod (boron carbide/steel). As in the earlier CORA tests the test bundles were subjected to temperature transients of a slow heatup rate in a steam environment. The transient phases of the tests were initiated with a temperature ramp rate of 1 K/s. With these conditions a so-called small-break LOCA was simulated. The temperature escalation due to the exothermal zircon/niobium-steam reaction started at about 1200 C, leading the bundles to maximum temperatures of approximately 1900 C. The thermal response of bundle CORA-W2 is comparable to that of CORA-W1. In test CORA-W2, however, the temperature front moved faster from the top to the bottom compared to test CORA-W1 [de

  1. Techniques and results of examination of fission product release from VVER fuel rods with artificial defects and a burnup of ∼60 MWd/kgU at the MIR loop facility

    International Nuclear Information System (INIS)

    Burukin, A.; Goryachev, A.; Ilyenko, S.; Izhutov, A.; Konyashov, V.; Shishin, V.; Shulimov, V.; Luzanova, L.; Miglo, V.

    2009-01-01

    Complex of equipment and several techniques for examination of radioactive fission product release from defective fuel rods were developed, prepared and tested at the PV-1 loop facility of the MIR reactor. During the first test, which was conducted at the PV-1 loop facility and aimed at testing of developed equipment and techniques, measurement of radioactive fission product release from an experimental re-fabricated fuel rod with a burnup of ∼60 MWd/kgU and an artificial defect was performed under design-basis steady-state operating conditions of the VVER-1000 reactor. PIE of all main parameters of the experimental defective fuel rod did not reveal any state peculiarities which could be caused by the artificial defect, i.e. fuel and cladding characteristics in the defect area did not differ from the initial ones (before testing) as well as their characteristics in areas distant from the defect; they are typical for fuel rods with a similar irradiation history in the VVER NPP. The gap in the experimental fuel rod was bridged due to close contact between fuel and cladding at increased fuel burnup; it can appreciable reduce release of radioactive fission products into the PV-1 primary coolant. This suggestion and quantitative characteristics of effect of gap bridging in a high-burnup fuel rod on radioactive fission product release should be investigated during the next tests performed at the PV-1 loop facility. Values of radioactive fission product release measured during the first test at the PV-1 loop facility in the MIR reactor will be used for development of an empirical engineering model in order to take into account high burnup effects and their impact on fission product release from fuel and defective fuel rods

  2. Validation of computer codes and modelling methods for giving proof of nuclear safety of transport and storage of spent VVER-type nuclear fuels. Pt. 2. Criticality safety during transport and storage of spent VVER fuel elements. Final report; Einschaetzung von Rechenprogrammen und Methoden zum Nachweis der nuklearen Sicherheit bei Transport und Lagerung von WWER-Kernbrennstoffen. T. 2. Kritikalitaetssicherheit bei Transport und Lagerung von WWER-Brennelementen. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Buechse, H.; Langowski, A.; Lein, M.; Nagel, R.; Schmidt, H.; Stammel, M.

    1995-03-15

    The report gives the results of investigations on the validation of computer codes used to prove nuclear safety during transport and storage of spent VVER - fuel of NPP Greifswald and Rheinsberg. Characteristics of typical spent fuel (nuclide concentration, neutron source strength, gamma spectrum, decay heat) - calculated with several codes - and dose rates (e.g. in the surrounding of a loaded spent fuel cask) - based on the different source terms - are presented. Differences and their possible reasons are discussed. The results show that despite the differences in the source terms all relevant health physics requirements are met for all cases of source term. The validation of the criticality code OMEGA was established by calculation of appr. 200 critical experiments of LWR fuel, including VVER fuel rod arrangements. The mean error of the effective multiplication factor k{sub eff} is -0,01 compared to the experiment for this area of applicability. Thus, the OMEGA error of 2% assumed in earlier works has turned out to be sufficiently conservative. (orig.) [Deutsch] Der Bericht enthaelt die Ergebnisse von Untersuchungen zur Validierung von Rechenprogrammen, welche zum Nachweis der nukelaren Sicherheit bei Transport und Lagerung von WWER-Kernbrennstoff der KKW Greifswald und Rheinsberg eingesetzt wurden. Es werden eine Charakteristik des abgebrannten Brennstoffs (Nuklidkonzentrationen, Neutronenquellstaerke, Gammaspektrum, Nachzerfallsleistung) - berechnet mit verschiedenen Programmen - und Ortsdosisleistungen (z.B. in der Umgebung eines Transportbehaelters) - basierend auf den verschiedenen Quelltermen - angegeben. Differenzen und Ursachen werden diskutiert. Die Ergebnisse zeigen, dass trotz der Differenzen in den Quelltermen alle strahlenschutztechnisch relevanten Aussagen unbeeinflusst bleiben. Fuer die Einschaetzung des Gueltigkeitsbereiches des Monte-Carlo-Programms OMEGA wurden ca. 200 kritische Experimente mit LWR-Brennstoff unter besonderer Beruecksichtigung

  3. State of fuel rods spent in the VVER-1000 reactor up to a fuel burnup of 75 MW·Day/KgU

    International Nuclear Information System (INIS)

    Markov, D.; Zvir, E.; Polenok, V.; Zhitelev, V.; Strozhuk, A.; Volkova, I.

    2011-01-01

    The presented material contains the data on change in form, corrosion state and mechanical properties of fuel rod claddings, change in fuel structure and release of gaseous fission products (GFP) under the cladding. The results of PIEs of the VVER-1000 fuel rods with the high burnup of fuel (average value is 72.3 MW·day/kgU and maximum is 75 MW·day/kgU) carried out in JSC 'SSC RIAR' show that by the basic operational characteristics the lifetime of fuel rods with such burnup of fuel is not exhausted. The state of fuel rods is characterized by following key parameters. The fuel-to-cladding gap on the most part of the fuel meat is absent. With the burnup growth, diameter of the fuel rod increases due to fuel meat swelling. In so doing, the reverse strain achieves the values of 0.40-0.47 %. Ridges on the cladding are formed practically along the entire length of the fuel meat, average height of ridges makes up 25 μm, maximum - 40 μm. At burnups exceeding 55 MW·day/kgU, the rate of the fuel rod elongation is less than at low and average burnups. So if within a burnup range of 20-55 MW·day/kgU, the rate of the fuel rod elongation makes up about 0.330mm per 1 MW·day/kgU, at burnups exceeding 55 MW·day/kgU it is only 0.085mm per 1 MW·day/kgU. Corrosion state of the claddings of fuel rods with high burnup of fuel is satisfactory. The oxide film, as a rule, is uniform, dense, without cracks and exfoliation, its thickness on the external surface does not exceed 13 μm, while on the internal surface - 15 μm. Hydrogenation is insignificant, mass fraction of hydrogen does not exceed 0.01 %. Interaction of fuel rods with spacer grids does not result in significant fretting-corrosion. Based of the results of tests, short-term mechanical properties of the claddings of fuel rods with high burnup of fuel remain at high level. The state of fuel is characterized by absence of the fuel-to-cladding gap on the most part of the fuel meat, fuel is tightly fixed to the cladding

  4. SCORPIO - VVER core surveillance system

    International Nuclear Information System (INIS)

    Zalesky, K.; Svarny, J.; Novak, L.; Rosol, J.; Horanes, A.

    1997-01-01

    The Halden Project 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. So far the system has only been implemented on western PWRs but the basic concept is applicable to a wide range of reactor 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 was initiated in cooperation with the Nuclear Research Institute at Rez and industry partners in the Czech Republic. The first system will be installed at the Dukovany NPP. (author)

  5. Qualification of the core model DYN3D coupled with the code ATHLET as an advanced tool for the accident analysis of VVER type reactors. Pt. 2. Final report

    International Nuclear Information System (INIS)

    Grundmann, U.; Kliem, S.; Rohde, U.

    2002-10-01

    Benchmark calculations for the validation of the coupled neutron kinetics/thermohydraulic code complex DYN3D-ATHLET are described. Two benchmark problems concerning hypothetical accident scenarios with leaks in the steam system for a VVER-440 type reactor and the TMI-1 PWR have been solved. The first benchmark task has been defined by FZR in the frame of the international association 'Atomic Energy Research' (AER), the second exercise has been organized under the auspices of the OECD. While in the first benchmark the break of the main steam collector in the sub-critical hot zero power state of the reactor was considered, the break of one of the two main steam lines at full reactor power was assumed in the OECD benchmark. Therefore, in this exercise the mixing of the coolant from the intact and the defect loops had to be considered, while in the AER benchmark the steam collector break causes a homogeneous overcooling of the primary circuit. In the AER benchmark, each participant had to use its own macroscopic cross section libraries. In the OECD benchmark, the cross sections were given in the benchmark definition. The main task of both benchmark problems was to analyse the re-criticality of the scrammed reactor due to the overcooling. For both benchmark problems, a good agreement of the DYN3D-ATHLET solution with the results of other codes was achieved. Differences in the time of re-criticality and the height of the power peak between various solutions of the AER benchmark can be explained by the use of different cross section data. Significant differences in the thermohydraulic parameters (coolant temperature, pressure) occurred only at the late stage of the transient during the emergency injection of highly borated water. In the OECD benchmark, a broader scattering of the thermohydraulic results can be observed, while a good agreement between the various 3D reactor core calculations with given thermohydraulic boundary conditions was achieved. Reasons for the

  6. Impact of cross-section generation procedures on the simulation of the VVER 1000 pump startup experiment in the OECD/DOE/CEA V1000CT benchmark by coupled 3-D thermal hydraulics/ neutron kinetics models

    International Nuclear Information System (INIS)

    Boyan D Ivanov; Kostadin N Ivanov; Sylvie Aniel; Eric Royer

    2005-01-01

    Full text of publication follows: In the framework of joint effort between the Nuclear Energy Agency (NEA) of OECD, the United States Department of Energy (US DOE), and the Commissariat a l'Energie Atomique (CEA), France a coupled 3-D thermal hydraulics/neutron kinetics benchmark was defined. The overall objective OECD/NEA V1000CT benchmark is to assess computer codes used in analysis of VVER-1000 reactivity transients where mixing phenomena (mass flow and temperature) in the reactor pressure vessel are complex. Original data from the Kozloduy-6 Nuclear Power Plant are available for the validation of computer codes: one experiment of pump start-up (V1000CT-1) and one experiment of steam generator isolation (V1000CT-2). Additional scenarios are defined for code-to-code comparison. As a 3D core model is necessary for a best-estimate computation of all the scenarios of the V1000CT benchmark, all participants were asked to develop their own core coupled 3-D thermal hydraulics/ neutron kinetics models based on the data available in the benchmark specifications. The first code to code comparisons based on the V1000CT-1 Exercise 2 specifications exhibited unacceptable discrepancies between 2 sets of results, one of them being close to experimental results. The present paper focuses first on the analysis of the observed discrepancies. The VVER 1000 3-D thermal hydraulics/neutron kinetics models are based on thermal-hydraulic and neutronic data homogenized at the assembly scale. The neutronic data, provided as part of the benchmark specifications, consist thus in a set of parametrized 2 group cross sections libraries representing the different assemblies and the reflectors. The origin of the high observed discrepancies was found to lie in the use of these neutronic libraries. The concern was then to find a way to provide neutronic data, compatible with all the benchmark participants neutronic models, that enable also comparisons with experimental results. An analysis of the

  7. Stress corrosion cracking (Standard Astm G 30-90) in stainless steel 08X18H10T of swimming-pool that contain nuclear fuel in reactors V.V.E.R.-440

    International Nuclear Information System (INIS)

    Zamora R, L.; Herrera, V.

    1998-01-01

    The standard recommended practice for making and using 'U' bend stress corrosion test specimens; Designation G30-90 has been used as a laboratory tool to study the susceptibility of austenitic stainless steels and the other materials of test of intergranular stress corrosion cracking (IGSCC). The experiment has been development in a similar conditions of the chemical regime, the swimming-pool that containing nuclear fuel in borated water reactors VVER-440 in general this cladding by two films, one of carbon steel (04T26) and other with austenitic stainless steel 08X18HT (similar type 321) stabilized with titanium, the thickness of filler metals was to 4 to 8 mm. The specimens was prepare one plate with this characteristics, the welding was put in the part central with the following measurements of 160x15x5 mm. The specimens strips bent approximately 180 degrees around radius of curvature of R=14.5 mm and ε 1 = 17.2% and maintained in this plastically deformed condition during the test. And then preparing metallographically and exposure in environment of 12 and 40 gr./l of H 3 BO 3 70 Centigrade with or noting contaminants of NaCl. The results showed the initial cracks. (Author)

  8. Rolls-Royce successful modernization of safety-critical Instrumentation and Control (I and C) equipment at the Dukovany VVER 440/213 Nuclear Power Plant, based on SPINLINE 3 platform

    International Nuclear Information System (INIS)

    Rebreyend, P.; Burel, J.P.; Spoc, J.; Karasek, A.

    2010-01-01

    Rolls-Royce has provided on-time delivery of a substantial safety-critical I and C overhaul for four Nuclear reactors operated by Czech Republic utility, CEZ a.s. This nine-year project is considered to be one of the largest I and C modernization projects in the world. The Dukovany VVER 440 I and C modernization project and its key success factors are profiled in this paper. The project is in the final stages with the last unit to be completed in 2009. Beginning in September 2000, the project is in compliance with the initial schedule. Rolls-Royce has been designing and manufacturing I and C solutions dedicated to the implementation of safety and safety-related functions in nuclear power plants (NPPs) for more than 30 years. Though the early solutions were non-software-based, since 1984 software-based solutions for safety I and C functions have been deployed in operating NPPs across France and 15 other countries. The Rolls-Royce platform is suitable for implementation of safety I and C functions in new NPPs, as well as in the modernization of safety equipment in existing plants. CEZ a.s. is a major electricity supplier for the national grid. At Dukovany, CEZ a.s. operates four units of VVER-440/213-type reactors producing one quarter of CEZ a.s. electricity production. The first of these units was connected to the grid in 1985. Since the year 2000, the nine-year modernization program has been underway at Dukovany, at a cost of more than 200 million Euros. The equipment replacement was implemented during regular, planned outages of the original equipment and systems. After an international bidding phase, CEZ a.s. awarded a contract to Skoda JS for general engineering and project management. Individual subcontracts were then signed between Skoda JS and a consortium between Rolls-Royce and Areva for modernization of the safety systems, including the Reactor Protection System (RPS), the Reactor Control System (RCS), and the Post-Accident Monitoring System (PAMS). Two

  9. Feasibility study of chabazite absorber tube utilization in online absorption of released gaseous fission products and substitution of burnable absorber rods with chabazite absorber tubes in VVER-1000 reactor series

    International Nuclear Information System (INIS)

    Rahmani, Yashar

    2017-01-01

    Highlights: • Chabazite tubes are used for online removal of the released gaseous fission products. • The feasibility of using chabazite tubes instead of burnable absorber rods was studied. • A computational cycle was designed using the WIMSD5-B, CITATION-LDI2 and WERL codes. • In modeling fission gas release, the Weisman, Booth, Mason and T.S. models were used. • By this method, it is possible to increase cycle length and enhance heat transfer. - Abstract: As gaseous fission products, e.g. xenon and krypton have adverse effects such as reducing the rate of heat transfer in fuel rods and adding negative reactivity to the reactor core, the present manuscript was dedicated to development of a novel method for improving these defects. In the proposed method, chabazite absorber tubes were used for online removal of the released gaseous fission products from gaseous gap spaces. Moreover, in this research, feasibility of using chabazite absorber tubes instead of burnable absorber rods was examined. To perform the required modeling and calculations to successfully meet the mentioned objectives, a thermo-neutronic computational cycle was designed using the coupling of WIMSD5-B and CITATION-LDI2 codes in the neutronic section and the WERL code in the thermo-hydraulic calculations. In addition, in modeling the release process of gaseous fission products, the Weisman, Booth, Mason, and T.S. models were examined. It is worth mentioning that in this research, calculations and modeling procedures were based on the first cycle of Bushehr’s VVER-1000 reactor to study the feasibility of the proposed solution. The obtained results revealed that with application of the proposed method in this research, it is possible to increase cycle length, improve safety thresholds, and enhance heat transfer in the core of nuclear reactors.

  10. Microstructure alterations in the base material, heat affected zone and weld metal of a 440-VVER-reactor pressure vessel caused by high fluence irradiation during long term operation: material: 15 Ch2MFA {approx} 0, 15 C-2,5 Cr-0, 7Mo-0,3 V; Veraenderungen der Mikrostruktur in Grundwerkstoff, WEZ und Schweissgut eines 440-VVER-Reaktordruckbehaelters, verursacht durch Neutronenbestrahlung im langzeitigen Betrieb; Werkstoff: 15 Ch2MFA {approx} 0,15 C-2,5 Cr-0, 7Mo-0,3 V

    Energy Technology Data Exchange (ETDEWEB)

    Maussner, G; Scharf, L; Langer, R [Siemens AG Energieerzeugung KWU, Erlangen (Germany); Gurovich, B [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation)

    1998-11-01

    Within the scope of the Tacis `91/1.1 project of the European Community, ``Reactor Vessel Embrittlement``, specimens were taken from the heavily irradiated circumferential welds of a VVER pressure vessel. The cumulated fast neutron fluence in the specimens amounts to up to 6.5 x 10{sup 19} cm{sup -}2 (E > 0.5 MeV). For the multi-laboratory, coordinated study, the specimens were cutted for mechanical testing as well as analytical, microstructural and microanalytical examinations in the base metal, HAZ and weld metal with respect to the effects of reactor operatio and post-irradiation annealing as well as thermal treatment (475 C, 560 C). The analytical transmission electron microscopy (200 kV) revealed the alterations found in the mechanical properties to be due to the formation of black dots and irradiation-induced segregations and accumulations of copper and carbides. These effects, caused by operation, (neutron radiation, temperature), are much more significant in the HAZ than in the base metal. (orig./CB) [Deutsch] Im Rahmen des von der Europaeischen Union beauftragten Tacis `91/1.1 Programms `Reactor Vessel Embrittlement` wurden Bohrkerne aus dem hochbestrahlten Rundnahtbereich eines VVER-Reaktordruckbehaelters entnommen. Die kumulierte schnelle Neutronenfluenz in diesen Proben betraegt bis zu 6,5 x 10{sup 19} cm{sup -2} (E>0,5 MeV). In einer gemeinschaftlichen Untersuchung wurden mechanisch-technologische, chemische sowie mirkostrukturelle Untersuchungen an Grundwerkstoff-, WEZ- und Schweissgutproben im vergleichbaren Ausgangs-, bestrahlten und anschliessend waermebehandelten (475 C, 560 C) Werkstoffzustand durchgefuehrt. Die analytische Durchstrahlelektronenmikroskopie (200 kV) laesst als Ursache fuer die festgestellten Veraenderungen der mechanischen Eigenschaften die Bildung von Versetzungsringen (black dots) sowie von bestrahlungsinduzierten Ausscheidungen und Anreicherungen von Kupfer in den Karbiden erkennen. Diese Effekte, als Folge der betrieblichen

  11. Vliv hydrodynamiky na HDS testy ve zkrápěném reaktoru

    OpenAIRE

    Tukač, V.

    2012-01-01

    The study was fosussed on identification of flow parameters of gas and liquid phases during hydrodesulphurization in laboratory trickle bed reactor in order to minimize transfer phenomena effects on experimental measurement.

  12. Hydrogenace styrenu v pilotním zkrápěném reaktoru

    Czech Academy of Sciences Publication Activity Database

    Šimíčková, M.; Tukač, V.; Chyba, V.; Lederer, J.; Kolena, J.; Šimek, J.; Hanika, Jiří

    2005-01-01

    Roč. 1, - (2005), 117 /6P10/ ISSN 1336-7242. [Zjazd chemických spoločností /57./. 04.09.2005-08.09.2005, Tatranské Matliare] Institutional research plan: CEZ:AV0Z40720504 Keywords : styrene * hydrogenation * trickle bed reactor Subject RIV: CI - Industrial Chemistry, Chemical Engineering

  13. Measurements of neutron flux in the RA reactor; Merenje karakteristika neutronskog fluksa u reaktoru RA

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    This report includes the following separate parts: Thermal neutron flux in the experimental channels od RA reactor; Epithermal neutron flux in the experimental channels od RA reactor; Fast neutron flux in the experimental channels od RA reactor; Thermal neutron flux in the thermal column and biological experimental channel; Neutronic measurements in the RA reactor cell; Temperature reactivity coefficient of the RA reactor; design of the device for measuring the activity of wire. [Serbo-Croat] Ovaj izvestaj sadrzi sledece referate: Fluks termalnih neutrona u eksperimentalnim kanalima reaktora RA; Fluks epitermalnih neutrona u eksperimentalnim kanalima reaktora RA; Fluks brzih neutrona u eksperimentalnim kanalima reaktora RA; Fluks termalnih neurona u termalnoj koloni i bioloskom eksperimentalnom kanalu; Neutronska merenja u elementarnoj celiji reaktora RA; Temperaturni koeficijent reaktivnosti reaktora RA; Projekat uredjaja za merenje radioaktivnosti zice.

  14. Fire damp gas in a heavy water reactor; Praskavi gas u teskovodnom reaktoru

    Energy Technology Data Exchange (ETDEWEB)

    Nikolic, V D [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Yugoslavia)

    1963-07-01

    This document describes the process of fire damp gas creation in the reactor core and dependence of the gas percentage on the temperature, i.e. reactor power. It contains a detailed plan for measuring the the percent of fire damp gas at the RA reactor: before start-up, after longer shut-down periods, immediately after safety shutdown, periodically during operation campaign.

  15. Measurements of neutron flux in the RA reactor; Merenje karakteristika neutronskog fluksa u reaktoru RA

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    This report includes results of the following measurements performed at the RA reactor: thermal neutron flux in the experimental channels, epithermal and fast neutron flux, neutron flux in the biological shield, neutron flux distribution in the reactor cell.

  16. Instrumentation and control doctrine for VVER upgrading

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  17. Operating problems of the thermocouples in VVER

    International Nuclear Information System (INIS)

    Timonin, A.S.

    1997-01-01

    In WWER reactors, the coolant temperature at the outlet of the majority of fuel assemblies is measured with chromel-alumel cable thermocouples. The components of systematic errors in temperature measurements are discussed. Errors due to calibration drift can be avoided by periodical calibrations performed during the heating and hot test runs after reactor refueling. Errors due to radiation heating and response time can be estimated and thus eliminated. Errors due to flow stratification of the coolant can also be eliminated by an estimation of correction factors. The effects of the aging of the thermocouples are also discussed. The removal of thermocouples from their coverings for replacement presents some difficulties, which thus determine the service life of the thermocouples. (A.K.)

  18. VVER NPPs fuel handling machine control system

    International Nuclear Information System (INIS)

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

    2002-01-01

    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 Fuel Handling Machine (FHM) Control System units have been already supplied for Temelin NPP and others supply are in process for the Atommash company, which has in charge the supply of FHMs for NPPs located in Russia, Ukraine and China.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 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 and 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 of 1090 Work-stations (APMS - Advanced Plant Monitoring System, or Tenore NT) has been successfully used to interface the operator with the control of the fuel handling machine

  19. Examination of VVER-1000 Reactor Pressure Vessel

    International Nuclear Information System (INIS)

    Matokovic, A.; Picek, E.; Markulin, K.

    2008-01-01

    The increasing demand of a higher level of safety in the operation of the nuclear power plants requires the utilisation of more precise automated equipment to perform in-service inspections. That has been achieved by technological advances in computer technology, in robotics, in examination probe technology with the development of the advanced inspection technique and has also been due to the considerable and varied experience gained in the performance of such inspections. In-service inspection of reactor pressure vessel, especially Russian-designed WWER-1000 presents one of the most important and extensive examination of nuclear power plants primary circuit components. Such examination demand high standards of inspection technology, quality and continual innovation in the field of non-destructive testing advanced technology. A remote underwater contact ultrasonic technique is employed for the examination of the base metal of vessel and reactor welds, whence eddy current method is applied for clad surface examinations. Visual testing is used for examination of the vessel interior. The movement of inspection probes and data positioning are assured by using new reactor pressure vessel tool concept that is fully integrated with inspection systems. The successful performance of reactor pressure vessel is attributed thorough pre-outage planning, training and successful performance demonstration qualification of chosen non-destructive techniques on the specimens with artificial and/or real defects. Furthermore, use of advanced approach of inspection through implementation the state-of-the-art examination equipment significantly reduced the inspection time, radiation exposure to examination personnel, shortening nuclear power plant outage and cutting the total inspection costs. This paper presents advanced approach in the reactor pressure vessel in-service inspections and it is especially developed for WWER-1000 nuclear power plants.(author)

  20. Calculus of a reactor VVER-1000 benchmark

    International Nuclear Information System (INIS)

    Dourougie, C.

    1998-01-01

    In the framework of the FMDP (Fissile Materials Disposition Program between the US and Russian, a benchmark was tested. The pin cells contain low enriched uranium (LEU) and mixed oxide fuels (MOX). The calculations are done for a wide range of temperatures and solute boron concentrations, in accidental conditions. (A.L.B.)

  1. ABPM a hypertenze u dětí s diabetem mellitem typu 1

    OpenAIRE

    Šuláková, Terezie

    2010-01-01

    Objective: to evaluate the diagnostic efficacy of the office blood pressure (OBP) and ambulatory blood pressure monitoring (ABPM) in the assessment of hypertension (HTN) in children with diabetes mellitus type 1 (DM1). Methods: We analyzed OBP and ABPM measurements in 84 diabetic children (43 boys) obtained at a median age of 14.9 years and 6.3±3.5 years after diagnosis of DM1. OBP and ABPM results were converted into standard deviation scores (SDS). In addition, we analyzed blood pressure lo...

  2. Biologická charakteristika půdního typu černozem

    OpenAIRE

    Krajčíková, Martina

    2014-01-01

    In bachelor thesis we evaluated biological parameters of selected subtypes of Chernozems. Object of study were -- Haplic Chernozem (Unčovice), Haplic Chernozem (Praha-Ruzyně) and Calcaric Chernozem (Velešovice). We determined total organic carbon content, humic substances content, microbial biomass and soil basal respiration. Further we evaluated the influence of tillage system and compost application onto selected parameters. Obtained results showed that compost application effected soil bio...

  3. Riziková analýza vodovodu obecního typu

    OpenAIRE

    Potyšová, Petra

    2012-01-01

    Cílem diplomové práce bylo analyzovat vodovodní síť v obci Klobouky u Brna pomocí softwarové aplikace WaterRisk za použití jednoduché a komplexní metodiky. Na vodovodní síti bylo provedeno měření zákalu jako praktická podpora nežádoucích stavů. Rovněž byl proveden podrobný průzkum obce Klobouky u Brna. Přidanou hodnotou diplomové práce byla konzultace s provozním technikem ve společnosti Vodovody a kanalizace Hodonín, a.s. o zdroji vody a úpravně vody, ze které se voda přebírá pro obec Klobou...

  4. Osobowość pacjentek z cukrzycą typu 2

    OpenAIRE

    Kocemba, Iwona

    2003-01-01

    The aim of this study was to estimate the relationship between personality of 66 healthy women and 66 women with type 2 diabetes. The average age of both groups was 63. I use the 16th Personality Factors Test by Cattell. A lower level of intelligence characterizes the personality of a diabetic, they are more submissive, effusive and responsible. They are calmer and more practical in their behavior. Healthy women are more self-sufficient and unconventional.

  5. Welding of Prosthetic Alloys / Spawanie Stopów Protetycznych Typu Co-Cr

    Directory of Open Access Journals (Sweden)

    Wojciechowska M.

    2015-12-01

    Full Text Available This paper presents the techniques of joining metal denture elements, used in prosthetic dentistry: the traditional soldering technique with a gas burner and a new technique of welding with a laser beam; the aim of the study was to make a comparative assessment of the quality of the joints in view of the possibility of applying them in prosthetic structures. Fractographic examinations were conducted along with tensile strength and impact strength tests, and the quality of the joints was assessed compared to the solid metal. The experiments have shown that the metal elements used to make dentures, joined by the technique which employs a laser beam, have better strength properties than those achieved with a gas burner.

  6. Qualification of the nuclear reactor core model DYN3D coupled to the thermohydraulic system code ATHLET, applied as an advanced tool for accident analysis of VVER-type reactors. Final report; Qualifizierung des Kernmodells DYN3D im Komplex mit dem Stoerfallcode ATHLET als fortgeschrittenes Werkzeug fuer die Stoerfallanalyse von WWER-Reaktoren. T. 1. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Grundmann, U.; Kliem, S.; Krepper, E.; Mittag, S; Rohde, U.; Schaefer, F.; Seidel, A.

    1998-03-01

    The nuclear reactor core model DYN3D with 3D neutron kinetics has been coupled to the thermohydraulic system code ATHLET. In the report, activities on qualification of the coupled code complex ATHLET-DYN3D as a validated tool for the accident analysis of russian VVER type reactors are described. That includes: - Contributions to the validation of the single codes ATHLET and DYN3D by the analysis of experiments on natural circulation behaviour in thermohydraulic test facilities and solution of benchmark tasks on reactivity initiated transients, - the acquisition and evaluation of measurement data on transients in nuclear power plants, the validation of ATHLET-DYN3D by calculating an accident with delayed scram and a pump trip in VVER plants, - the complementary improvement of the code DYN3D by extension of the neutron physical data base, implementation of an improved coolant mixing model, consideration of decay heat release and xenon transients, - the analysis of steam leak scenarios for VVER-440 type reactors with failure of different safety systems, investigation of different model options. The analyses showed, that with realistic coolant mixing modelling in the downcomer and the lower plenum, recriticality of the scramed reactor due to overcooling can be reached. The application of the code complex ATHLET-DYN3D in Czech Republic, Bulgaria and the Ukraine has been started. Future work comprises the verification of ATHLET-DYN3D with a DYN3D version for the square fuel element geometry of western PWR. (orig.) [Deutsch] Das Reaktorkernmodell DYN3D mit 3D Neutronenkinetik wurde an den Thermohydraulik-Systemcode ATHLET angekoppelt. Im vorliegenden Bericht werden Arbeiten zur Qualifizierung des gekoppelten Codekomplexes zu einem validierten Hilfsmittel fuer Stoerfallablaufanalysen zu Reaktoren des russischen Typs WWER dargestellt. Diese umfassten im einzelnen: - Beitraege zur Validierung der Einzelcodes ATHLET und DYN3D anhand der Nachrechnung von Experimenten zum

  7. Susceptibility to stress corrosion in stainless steels type AISI 321 and 12X18H10T used in PWR type reactors (WWER); Susceptibilidad a la corrosion bajo esfuerzo de barras de acero inoxidable AISI 321 y 12X18H10T en ambientes utilizados en reactores VVER

    Energy Technology Data Exchange (ETDEWEB)

    Matadamas C, N

    1996-12-31

    Titanium stabilized stainless steels have been utilized in sovietic pressurized water reactors (VVER) for avoid the susceptibility to Intergranular Corrosion (IGC) present in other austenitic stainless steels. However the Intergranular Corrosion resistance of this kind of materials has been questioned because of Intergranular Stress Corrosion Cracking failures (IGSCC) have been reported. This paper study the electrochemical behavior of the AISI 321 stainless steel in a H{sub 3}BO{sub 3} Solution contaminated with chlorides and its susceptibility to Intergranular Corrosion.Electrochemical prediction diagrams of the stainless steels AISI 321 and 12X18H10T (sovietic) sensitized (600 Centigrade, 3 h.) were compared. Cylindrical and conical samples were used in Slow Strain Rate Tests (SSRT), to determine the susceptibility to Stress Corrosion Cracking (SCC) in AISI 321 and 12X18H10T stainless steels. The results obtained showed that the temperature of the solution is a very important factor to detect this susceptibility. Fractography studies on the fracture surfaces of the samples obtained in the SSRT at high temperature were realized. Corrosion velocities of both AISI 321 and 12X18H10T stainless steels were determined using conical samples in the CERT system at high temperature. E.D.A.X. analysis was employed in both AISI 321 and 12X18H10T stainless steels in order to explain the degree of sensitization. (Author).

  8. Calculus of a reactor VVER-1000 benchmark; Calcul d'un benchmark de reacteur VVER-1000

    Energy Technology Data Exchange (ETDEWEB)

    Dourougie, C

    1998-07-01

    In the framework of the FMDP (Fissile Materials Disposition Program between the US and Russian, a benchmark was tested. The pin cells contain low enriched uranium (LEU) and mixed oxide fuels (MOX). The calculations are done for a wide range of temperatures and solute boron concentrations, in accidental conditions. (A.L.B.)

  9. The results of postirradiation examinations of VVER-1000 and VVER-440 fuel rods

    Science.gov (United States)

    Dubrovin, K. P.; Ivanov, E. G.; Strijov, P. N.; Yakovlev, V. V.

    1991-02-01

    The paper presents the results of postirradiation examination of the fuel rods having different fuel-cladding gaps, pellet densities, pellet inner diameters and so on. The fuel rods were irradiated in the material science reactor (MR) of the Kurchatov Institute of Atomic Energy and at 4 unit of the Novo-Voronezh nuclear powerplant. Some data on fission gas release and rod geometry and compared with computer code predictions.

  10. Stochastic processes analysis in nuclear reactor using ARMA models; Analiza stohastichkih procesa u nuklearnom reaktoru korishcjenjem ARMA modela

    Energy Technology Data Exchange (ETDEWEB)

    Zavaljevski, N [Institute of Nuclear Sciences Boris Kidric VINCA, Belgrade (Yugoslavia)

    1990-07-01

    The analysis of ARMA model derived from general stochastic state equations of nuclear reactor is given. The dependence of ARMA model parameters on the main physical characteristics of RB nuclear reactor in Vinca is presented. Preliminary identification results are presented, observed discrepancies between theory and experiment are explained and the possibilities of identification improvement are anticipated. (author)

  11. Contribution to solving the problem of neutron thermalization in heterogeneous reactor; Prilog resavanju problema termalizacije neutron u heterognom reaktoru

    Energy Technology Data Exchange (ETDEWEB)

    Pop-Jordanov, J P [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1963-12-15

    A method for calculating of neutron termalization in heterogeneous rector core was developed. It is more precise than the diffusion method but more complcated. Concerning accuracy it is comparable to non-diffusion methods. Sonce the approach was analytical need for powerful computer is avoided and the description of physical phenomena is more transparent. Convergence is satsfactory. Constraints of the proposed method are: low neutron absorption in the moderator, negligible slowing down in the fuel, and big lattice pitch. The method is applicable for heavy water and graphite moderator systems. Based on the application of this method, procedures were developed for calculating thermal utilzation and neutron temperature. Since 1/v dependence of cross sections is not estimated this metof could be used for long-term reactivity changes.

  12. Coupled fast-thermal system at the RB, masters thesis; Spregnuti brzo-termicki sistem na reaktoru RB, magistarski rad

    Energy Technology Data Exchange (ETDEWEB)

    Pesic, M [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

    1984-05-15

    Coupled fast-thermal system at the RB reactor was formed owing to availability of highly enriched fuel. This paper deals with reactor parameters calculations and measurements of coupled core taking into account safety constraints. Validity of applied calculation methods was confirmed. The following parameters were analyzed: critical height of the core; reactivity dependent on heavy water level in the core; fast neutron spectrum in the fast region channel; spatial distribution of thermal. epithermal and fat neutrons in the fast region channel; reactivity of safety rods; neutron and gamma absorption doses in the center of the coupled core.

  13. Nuclear fuel burnup calculation in a Voronezh type reactor; Analiza izgaranja nuklearnog goriva u reaktoru tipa Voronjez

    Energy Technology Data Exchange (ETDEWEB)

    Matausek, M; Marinkovic, N; Kocic, A [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

    1977-07-01

    In order to summarize and present our abilities to perform a complex computation of the nuclear fuel burn-up, a systematic review of the available methods, algorithms and computer programmes is given in this paper. The computer programmes quoted have all been developed, modified and tested in our department, so that they can be successfully used in the analysis of nuclear power plants from both physics and economic points of view. For a commercially proven nuclear reactor - reactor of the Voronezh type - an illustrative computation of the fuel burn-up is performed. The typical results are presented and discussed. The conclusion concerns the completion of a modular scheme for the fuel burn-up calculation and the fuel cycle analysis (author)

  14. Safety shutdowns and failures of the RA reactor equipment; Sigurnosna zaustavljanja i kvarovi opreme na reaktoru RA

    Energy Technology Data Exchange (ETDEWEB)

    Mitrovic, S [Institut za nuklearne nauke ' Boris Kidric' , Vinca, Belgrade (Yugoslavia)

    1966-07-01

    This report is an attempt of statistical analysis of the failures occurred during RA reactor operation. A list of failures occurred on the RA equipment during 1965 is included. Failures were related to the following systems: dosimetry system (22%), safety and control system (7%), heavy water system (2%), technical water (4%), helium system (2%), measuring instruments (30%), transport, ventilation, power supply systems (32%). A review of safety shutdowns from 1962 to 1966 is included as well, as a comparison with three similar reactors. Although the number of events used for statistical analysis was not adequate, it has been concluded that RA reactor operation was stable and reliable.

  15. Program of critical experiment and measurements at the RA reactor; Program kriticnih eksperimenata i merenja na reaktoru RA

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1960-04-14

    Program described in this document describes in detail the following experiments: critical experiments with two reactor core lattices with 38 and 44 fuel channels, initial heavy water level being 1300 mm, criticality is achieved by adding heavy water; preliminary analysis of heavy water quality and verification of the fuel isotopic contents; experiment with the initial core which contains 56 fuel channels with maximum heavy water level according to the Russian proposal; measurement of neutron flux by Dy and In foils; measurement of reactivity excess dependent on the heavy water level and number of fuel rods; measurement of reactor period for determined reactivity change; measurement of moderator temperature coefficient; measurement of absolute flux. [Serbo-Croat] Program sadrzan u ovom dokumentu opisuje detaljno sledece eksperimente: kriticni eksperiment sa dve konfiguracije jezgra reaktora, sa 38 i 44 gorivna kanala, pocetni nivo teske vode je 1300 mm, kriticnost se dostize dodavanjem teske vode; prethodno izvrsenom analizom teske vode i proverom izotopskog sastava goriva; eksperiment sa pocetnom resetkom koja prema ruskom predlogu sadrzi 56 gorivnih kanala i maksimalnom visinom teske vode; merenje raspodele neutronskog fluksa folijama Dy i In; kalibracija regulacionih sipki; merenje viska reaktivnosti sa promenom visine nivoa teske vode i promenom broja sipki; merenje periode reaktora za odredjenu promenu reaktivnosti; merenje temperaturnog koeficijenta za vodu; merenje apsolutnog fluksa.

  16. New neutron and gamma dosimetry equipment at the RB reactor; Nova merna neutronska i gama dozimetrijska oprema na reaktoru RB

    Energy Technology Data Exchange (ETDEWEB)

    Pesic, M; Stefanovic, D; Jevremovic, M; Petronijevic, M; Vranic, S; Ilic, I [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1992-07-15

    In the frame of bilateral cooperation between Germany and Yugoslavia, complete control, safety and dosimetry equipment of the shut-down SNEAK reactor was donated to Vinca Institute and transported to be installed at the RB reactor. This report contains detailed description of instrumentation components including detectors, electronic components and electronic circuits. Experimental data which verified correct functioning of the installed devices are part of this document. The objective of the RB reactor staff is to achieve new safety and dosimetry system in order to improve the reliability and availability of the RB reactor for future experiments.

  17. Engineering documentation of the GM counter at the RB reactor; Tehnicka dokumentacija GM brojaca na reaktoru RB

    Energy Technology Data Exchange (ETDEWEB)

    Pesic, M; Petronijevic, M; Vranic, S; Jevremovic, M; Ilic, I [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1986-07-01

    A GM counting device was completed. It is meant for absolute and relative measurements of beta and gamma activities of the irradiated samples as well as for determining the radiation transmission properties of the materials. This report shows technical and operation characteristics of the device. Realizovan je GM brojacki uredjaj koji je namenjen za apsolutna i relativan merenja beta i gama akrivnosti ozracenih uzoraka kao i za odredjivanje transmisionih karakteristika materijala za ova zracenja. U radu su prokazane tehnicke i radne karakteristike uredjaja. (author)

  18. Model typu “czarna skrzynka“ s charakterystyka kwadraowa v analizie wlašciwošci pomp i dmuchaw

    Czech Academy of Sciences Publication Activity Database

    Tesař, Václav; Peszyński, K.

    2007-01-01

    Roč. 4, č. 65 (2007), s. 27-32 ISSN 1426-6644 R&D Projects: GA AV ČR IAA200760705 Institutional research plan: CEZ:AV0Z20760514 Keywords : pump * ventilator * characteristic Subject RIV: BK - Fluid Dynamics

  19. Analýza licencí typu open source a jejich použití v praxi

    OpenAIRE

    Neumann, Martin

    2010-01-01

    The aim of this bachelor thesis is to analyse the most commonly used open source/free software licenses. After brief introduction to the history and defining related terms, there is a classification of licenses by the method of distribution. An analysis of the most common licenses is brought in, followed by presentation of the main points of licensing conditions in the Czech language. The main purpose of this work is to create a transparent methodology for selecting the proper licenses accord...

  20. Laboratorní úloha Virtuální sítě typu OpenVPN

    OpenAIRE

    Kortus, Jiří

    2012-01-01

    Tato práce v úvodu popisuje principy virtuálních privátních sítí (VPN) a stručně zmiňuje hlavní protokoly a přístupy pro jejich realizaci. Následně se podrobněji zaměřuje na seznámení čtenáře s charakterem a vlastnostmi sítí OpenVPN. Uvádí jednotlivé rysy a možnosti, diskutuje jejich výhody a nevýhody. Dále se věnuje návrhu laboratorní práce zaměřené na použití OpenVPN. Diskutuje možnosti návrhu práce zejména z technického a didaktického hlediska. Na jejich základě představuje koncept laborat...

  1. Základní pojmy a principy konstrukce modelů typu věk-období-kohorta

    Czech Academy of Sciences Publication Activity Database

    Reissigová, Jindra; Rychtaříková, J.

    2015-01-01

    Roč. 57, č. 1 (2015), s. 21-39 ISSN 0011-8265 Grant - others:GA ČR(CZ) GAP404/12/0883 Institutional support: RVO:67985807 Keywords : Lexis diagram * age-period-cohort * identification problem * generalised linear model * prediction * male mortality * Czech Republic Subject RIV: BB - Applied Statistics, Operational Research

  2. SYSTEM WYŚWIETLANIA PARAMETRÓW LOTU DLA ŚMIGŁOWCA TYPU Mi-17 – WYNIKI BADAŃ

    Directory of Open Access Journals (Sweden)

    Bęczkowski Grzegorz

    2015-08-01

    Full Text Available Przedmiotem artykułu są badania w locie nowo opracowanego systemu wyświetlania parametrów lotu (SWPL. Omawiany system przeznaczony jest do zobrazowania na wyświetlaczu przeziernym umieszczonym przed okiem pilota pełnej informacji pilotażowo-nawigacyjnej, ostrzeganiu o sytuacji niebezpiecznej na pokładzie śmigłowca i sygnalizacji błędów pracy systemów pokładowych. System przystosowany jest do pracy w warunkach dziennych i nocnych, przy czym w warunkach nocnych współpracuje z goglami noktowizyjnymi stosowanymi w lotnictwie Sił Zbrojnych RP.

  3. Unit information system operational displays for VVER-1000 reactors

    International Nuclear Information System (INIS)

    Anikanov, S.S.; Carrera, J.P.; Gordon, P.

    1997-01-01

    The role of high level operational displays is explained as well as the principles of the design of such displays. The tasks of WWER operating personnel are described and the support provided by operational displays is highlighted. The architecture of the displays is also dealt with. (A.K.)

  4. Upgrading the safety of VVER-440/V-230

    International Nuclear Information System (INIS)

    Kelm, P.; Wenk, W.

    1995-01-01

    Besides measures seeking to restore the status as laid down in the project design, especially backfitting measures must be mentioned which serve to ensure component and pipe integrity. Ensuring component integrity is a problem not only of RPV embrittlement, but also of failure prevention. This aspect was not always taken into account properly. Further activities in the field of component integrity will focus on backing the brittle fracture evaluation of the RPV; qualifying the leak-before-breack criterion for the main pipes and in areas with screwed connections; qualifying the program of in-service inspections. Several operators are currently in the process of drafting backfitting programs. The upgrading measures envisaged must be checked as to their balanced nature. In certain plants, the integrity of the RPV coud turn out to be the weak spot in upgrading measures. As a consquence, concepts seeking to achieve upgrading for long periods of time may differ from one location to the next and even between units. Extensive modifications in systems engineering and building structures generally must be evaluated against the expected improvement in safety of the whole plant. (orig.) [de

  5. Testing of VVER reactor pressure vessels by TOFD method

    International Nuclear Information System (INIS)

    Skala, Z.; Vit, J.

    2002-01-01

    The Time of Flight Diffraction Method (TOFD) - one of the new testing methods capable to obtain the real dimensions of flaws - is presented in the paper.The laboratory experiments on samples with artificial flaws and samples with artificially prepared cracks confirmed the high accuracy of flaw through wall extent sizing by TOFD. This accuracy was confirmed by qualification of methods and systems used by Skoda JS for the in-service inspections of WWER 440 vessel circumferential weld. The qualification also confirmed the ability of TOFD to detect reliably flaws, which can are not reliably detected by standard pulse echo testing. Based on the result of experiments and qualification, the TOFD method shall be used routinely by Skoda JS for the inspection of vessel circumferential welds root area and for sizing of flaws exceeding the acceptance level

  6. Corrosion particles in the primary coolant of VVER-440 reactors

    International Nuclear Information System (INIS)

    Vajda, N.; Molnar, Z.; Macsik, Z.; Szeles, E.; Hargittai, P.; Csordas, A.; Pinter, T.; Pinter, T.

    2010-01-01

    Corrosion and activity build-up processes are of major concern in ageing and life-extension of nuclear power reactors. Researches to study the migration of radioactive corrosion particles have been initiated at Paks Nuclear Power Plant (NPP), Hungary in order to better understand the corrosion of the primary circuit surfaces, the transport and activation of the particles of corrosion origin and their deposition on in-core and out-of-core surfaces. Radioactive corrosion particles were collected from the primary coolant and the steam generator surfaces of the 4 reactor units and subjected to detailed microanalytical and radioanalytical investigations. Scanning electron microscopy and energy dispersive X-ray microanalysis (SEM-EDX) were used to study the morphology and the composition of the matrix elements in the particles and the deposited corrosion layers. Particles identified by SEM-EDX were re-located under optical microscope by means of a coordinate transformation algorithm and were separated with a micromanipulator for further studies. Activities of γ emitting radionuclides were determined by high resolution γ spectrometry, and those of β decaying isotopes were measured by liquid scintillation (LS) spectrometry after radiochemical processing. High sensitivity of the nuclear measuring techniques allowed us to determine the low activity concentrations of the long-lived radionuclides, i.e. 60 Co, 54 Mn, 63 Ni, 55 Fe in the individual particles. Finally, high resolution sector-field inductively coupled plasma mass spectrometry (SF-ICP-MS) was applied to determine the ultralow concentrations of Co, Fe, Ni in the same particles. Specific activities of 60 Co/Co, 54 Mn/Fe, 55 Fe/Fe and 63 Ni/Ni were derived from the measured activity and concentration data. Specific activities of the radioactive corrosion products reveal the history of activity buildup processes in the particle. Typically, Fe-Cr-Ni oxide particles formed as a result of corrosion of the steel surfaces are released into the coolant and activated during the residence time (τ) in the reactor core that varies in a wide range. Smallest τ values are obtained if the particle circulates without deposition in the core and high values reflect a long deposition time in the core, probably on the fuel cladding surfaces. Residence times (τ) were calculated from the measured specific activities under various migration-activation conditions. Conclusions about particle transport-activation processes in case of the 4 reactor units have been drawn. The residence time of the migrating corrosion particles is usually small, less than 50 days. Periods of stable operation, shutdown and startup were comparatively evaluated. The 4 reactor units of the plant which have different operation history (including steam generator decontamination in certain cases) were also compared with each other. Significant differences in the Fe/Cr/Ni ratio of the oxide layers of the steam generators were observed. Correlation between the behavior of the migrating particles and particles deposited on the steam generator surfaces were also studied. (author)

  7. Coolers of the emergency cooling system for VVER 1000 Temelin

    International Nuclear Information System (INIS)

    Prchal, J.

    1997-01-01

    The fabrication of 6 heat exchangers for the new nuclear power plant at Temelin was successfully finished by the Kralovopolska Company in August 1993. All parts of the emergency cooling exchangers that will get into contact with the water coolant will be fitted with two-phase steel. The exchanger design is based on the AO 257 572 heat exchanger with helical partitions. The advantage of this patented solution is a significant intensification of the heat exchange at a minimum loss in pressure, which makes it possible to use tubes of a smaller diameter. The simplicity of this solution leads to a reduction in the number and length of welded joints and thus increases the reliability. (M.D.)

  8. Nuclear fuel for VVER reactors. Actual state and trends

    International Nuclear Information System (INIS)

    Molchanov, V.

    2009-01-01

    The implementation of 2nd generation of FA with the following performance characteristics: 1) Average FP failure factor during operation is less than 1x10 -6 ; 2) Fuel burn-up: up to 60 GWxd/tHM; 3) 5-year fuel cycle; 4) Unit thermal power uprate up to 110% N nom ; 5) Load Follow Mode (97,5±2,5% and 100-75-100% N nom ) is presented. A new generation of FA (TVSA and TVS-2) is developed, implemented and successfully operated at Russian, Ukrainian and Bulgarian NPPs providing: Safe and reliable operation cycle of 6 years; Assembly burn-up - up to 60 GWxd/tHM; Load Follow Mode. For satisfaction of customer needs regarding nuclear fuel performance the following development of FA design is under way: Power uprate up to 104% N nom and Fuel cycles length up to 18 months

  9. Unique systems for inspection and repair of VVER steam generators

    International Nuclear Information System (INIS)

    Kovalyk, Adrian; Pilat, Peter; Jablonicky, Pavol

    2014-01-01

    During over 30 years, VUJE Trnava has developed a series of remote control manipulators for in-service inspection and maintenance of steam generator collectors. The manipulators have been widely used on WWER type reactors in Slovakia and abroad. Some of the manipulators for non-destructive testing are shown.

  10. VVER-440 loading patterns optimization using ATHENA code

    International Nuclear Information System (INIS)

    Katovsky, K.; Sustek, J.; Bajgl, J.; Cada, R.

    2009-01-01

    In this paper the Czech optimization state-of-the-art, new code system development goals and OPAL optimization system are briefly mentioned. The algorithms, maths, present status and future developments of the ATHENA code are described. A calculation exercise of the Dukovany NPP cycles, on increased power using ATHENA, starting with on-coming 24th cycle (303 FPD) continuing with 25th (322 FPD), and 26th (336 FPD); for all cycles K R ≤1.54 is presented

  11. Current problems of VVER-1000 reactor core operation in Ukraine

    International Nuclear Information System (INIS)

    Bykov, A.

    2000-01-01

    Planned control rod drop time registration was passed two times a year per reactor unit. In 1992-1993 some control rods at almost all WWER-1000 units exceed the prescribed 4 second time limit. More than 7000 individual control rod time tests were made from the main correction measures. The following conclusions have been made from the current statistics data: (a) The main role of the vibration factor is proven in the fuel assembly (FA) bowing process. The greatest drop times and maximum of bowing values are concentrated at the vibration zone (2-4 FA rows from the reactor partition). The first FA row seems to be stable due to the interaction with the reactor partition; (b) Bowing relaxation will proceed during several fuel cycles (estimated value is 4-6), and depends on previous FA use history. It seems to be proven that previously bowed FAs effect the new FA, so previously bowed FAs are straightened until the middle of the fuel cycle. At some reactor units small drop time reduction is observed up to half of the fuel cycle from the start time values; (c) Control rod drop medium time (t) has almost linear dependence on operation time (τ) (t=k x τ + b]. Estimated by the method of least squares, values of k and b differ from unit to unit and from cycle to cycle. Values of k and b are in following ranges: b=2.0 - 2.6 seconds, k=5-50x10 -4 seconds per effective operation day; (d) Control rod drop time distribution changes through operation time. The position of maximum starts to shift after 240 effective days, and the form of the distribution start to change at the same time. Before 240 effective days, the distribution essentially does not change. To guarantee that the control rod system reliability is now within prescribed limits, we should continue testing. Additional analysis is needed. Test frequency can be reduced to avoid additional unreasonable transients. (authors)

  12. Nuclear safety evaluation of the VVER 440, Type 213

    International Nuclear Information System (INIS)

    Urbancik, L.

    1997-01-01

    The supervisory activities of the State Office for Nuclear Safety at the Dukovany nuclear power plant are described. No event resulting in an inpermissible radioactivity leak into the environment occurred at the plant in 1996. From among the 76 failures and events having occurred, only 4 were classified as level 1 on the International Nuclear Event Scale. Changes in the technology of radioactive waste bituminization were proposed. The Interim Spent Fuel Storage Facility at the Dukovany site was in test operation in 1996. Selected physical parameters of this facility were monitored. Seven international transports of spent fuel were accomplished in 1996. The dose rates in the surroundings of the Dukovany plant are monitored constantly by a teledosimetric system operated by the nuclear power plant. Periodical sampling and radionuclide activity measurements in the environment are also performed. (M.D.)

  13. Four-quadrant characteristics of Psb-VVER pumps

    International Nuclear Information System (INIS)

    Lipatov, I.A.; Elkin, I.V.; Antonova, A.I.; Dremin, G.I.; Kapustin, A.V.; Nikonov, S.M.; Rovnov, A.A.; Gudkov, V.I.

    2005-01-01

    This paper represents description of determination of Tunis-1620 pump head and torque characteristics of the integral thermophysical test facility Psb-Ver, obtained for single-phase coolant. Test procedure and main results obtained are described in the paper. (author)

  14. Extended Station Blackout Analysis for VVER-1000 MWe Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gaikwad, A. J.; Rao, R. S.; Lakshmanan, S. P.; Gupta, A., E-mail: avinashg@aerb.gov.in [Atomic Energy Regulatory Board, Mumbai (India)

    2014-10-15

    Post Fukushima, the plant behaviour for an extended station black-out (ESBO) scenario with only passive system availability for about 7 days has become imperative. Thermal hydraulic analysis of ESBO with the availability of passive heat removal system (PHRS), passive first stage and second stage hydro accumulators were carried out to demonstrate the design capabilities. Two different cases having primary leak rates of 2.2 tons/hr and 6.6 tons/hr were analyzed to study sustenance of natural circulation. For the study, out of 4 PHRS trains, one PHRS train was assumed to be in failure mode. The objective here is to predict the core cooling capability for a period of 7 days under ESBO conditions with the available water inventories from first and second stage hydroaccumulators only. Over simplified energy balance studies cannot ascertain sustenance of natural circulation in the primary system, steam generators (SGs) and PHRS. The analysis was carried out by using system thermal hydraulic safety code RELAP5/SCDAP/MOD 3.4. It is inferred that the inventory in the first stage accumulators and second stage accumulators compensate the leak and decay heat is removed effectively with the help of passive heat removal systems. It is also observed that even after 7 days of ESBO a large inventory is still available in the second stage accumulators and the primary system remains subcooled. (author)

  15. VVER-440 fuel cycles possibilities using modified FA design

    International Nuclear Information System (INIS)

    Mikolas, P.; Svarny, J.; Razym, V.; Dostal, M.; Jenik, J.; Krupar, P.

    2009-01-01

    A nearly equilibrium five-year cycle has been achieved at Dukovany NPP over the last years. This means that working fuel assemblies (WFA) with an average enrichment of 4.25 w% (control assemblies (CA) with an average enrichment of 3.82 w%) are normally loaded and reloaded for five years. Operation at uprated thermal power (105% of the original one, increase from 1375 MW t to 1444 MW t ) is being prepared by use of WFA with an average enrichment of 4.38 w% (CA with an average enrichment of 4.25 w%). With the aim of fuel cycle economy improvement, the fuel residence time in the core has to be prolonged up to six years with one cycle duration time up to 18 months and preserving loadings with very low leakage. In order to achieve this goal, at least neutron-physical characteristics of FA must be improved and such changes should be evaluated from other viewpoints. Some particular changes have already been analyzed earlier. Designs of new fuel assemblies with higher (and in the central part of a FA the highest possible, i.e. 4.95 w%) enrichment with preserving low pin power non-uniformity are described in the presented paper. An FA with an average enrichment of 4.66 w% (lower than originally evaluated) containing six fuel pins with 3.35 w% Gd 2 O 3 content was selected in the end. Fuel pins have bigger pellet diameter, bigger pin pitch and thinner FA shroud. A newly designed FA was evaluated from the viewpoint of physics (pin power non-uniformity, criticality of fuel at transport and storage and determination of basic quantities for spent fuel storage purposes by ORIGEN code), thermo-hydraulics (comparison of subchannel output temperatures and the departure from nucleate boiling ratio - DNBR) and mechanical properties. The purpose of this study was to simulate an FA subject to the loads during its six- year lifetime whereas normal working conditions were taken into account. There are presented two models with different shroud thickness undergoing these analyses. Both models also undergo buckling shroud analyses. Next the maximum inner excessive pressure limits (as a consequence of accident conditions) were determined. Furthermore, the low shroud thickness limit for loads representing normal working conditions was assessed. The model, whose shroud thickness was reduced down to 1.0 mm, was subjected to a low cycle fatigue analysis. Possibilities of fuel cycles are evaluated on model loadings with the newly designed FA, where the base are loadings for 27 th - 34 th cycles of the third unit of Dukovany NPP for uprated power. These cycles were prolonged (from approx 330 FPD to 370 FPD) using FA with higher enrichment. Moreover, newly optimized loadings of a length of up to 500 FPD (18 months) were considered. The transient process started from the last of the set of loadings (27 th - 34 th ) for uprated power. Newly designed fuel assemblies were loaded regularly in 18-month cycles. Average enrichment of CA was 4.38 w%. Transient loadings are formed by cycles 35-37 and the equilibrium cycle is created by cycles 38 and 39. Each cycle was optimized individually and fuel assemblies intended for unloading were determined for each cycle separately. The 'equilibrium' cycle is realized by three consecutive loadings with 16 fresh WFA and 2 fresh CA. Basic characteristics of a reference cycle and an 18-month cycle were compared. Optimization was performed by the OPAL - B code on the basis of 3D n-ph calculations of the MOBY-DICK code with the target function F dh < 1.51. Consecutive thermo-hydraulic calculations were executed following the core neutron-physical analysis that had been carried out by the MOBY-DICK code in the 1/6 core symmetry. These thermo-hydraulic calculations were executed for loadings of both existing and newly designed fuel assemblies. Fast neutron fluences (calculated by TORT transport code based on neutron sources calculated by MOBY-DICK code) onto the reactor pressure vessel for proposed the 12-month and 18-month cycles were also calculated and compared. The analyses performed confirmed that fuel cycles using newly designed FA and fulfilling basic safety criteria can be designed together with fuel cycle economy improvement. (authors)

  16. Safety analyses of the electrical systems on VVER NPP

    International Nuclear Information System (INIS)

    Andel, J.

    2004-01-01

    Energoprojekt Praha has been the main entity responsible for the section on 'Electrical Systems' in the safety reports of the Temelin, Dukovany and Mochovce nuclear power plants. The section comprises 2 main chapters, viz. Offsite Power System (issues of electrical energy production in main generators and the link to the offsite transmission grid) and Onsite Power Systems (AC and DC auxiliary system, both normal and safety related). In the chapter on the off-site system, attention is paid to the analysis of transmission capacity of the 400 kV lines, analysis of transient stability, multiple fault analyses, and probabilistic analyses of the grid and NPP power system reliability. In the chapter on the on-site system, attention is paid to the power balances of the electrical sources and switchboards set for various operational and accident modes, checks of loading and function of service and backup sources, short circuit current calculations, analyses of electrical protections, and analyses of the function and sizing of emergency sources (DG sets and UPS systems). (P.A.)

  17. Component vibration of VVER-reactors - diagnostics and modelling

    International Nuclear Information System (INIS)

    Altstadt, E.; Scheffler, M.; Weiss, F.P.

    1994-01-01

    The model comprises the whole primary circuit, including steam generators, loops, coolant pumps, main isolating valves and certainly the reactor pressure vessel and its internals. It was developed using the finite-element-code ANSYS. The model has a modular structure, so that various operational and assembling states can easily be considered. (orig./DG)

  18. Implementation of safety parameter display system at VVER-440 NPPs

    International Nuclear Information System (INIS)

    Manninen, T.

    1997-01-01

    Furnishing WWER-440 nuclear power plant units with a safety parameter display system (SPDS) fulfilling the requirements of internationally recognized standards and guidelines has been ranked high on the lists of proposed safety improvement projects. Technically such an SPDS system can be implemented either as a separate stand-alone system or as a more or less closely integrated part of a process information system of the plant unit. In the paper examples of these approaches are presented. Functionally all these examples include the well proven SPDS concept developed by IVO Power Engineering Ltd, Finland. The functional design basis, the general requirements for the system platform, experience with implementation and expansion possibilities of the systems are discussed. (author)

  19. Containment leak-tightness enhancement at VVER 440 NPPs

    International Nuclear Information System (INIS)

    Prandorfy, M.

    2000-01-01

    The hermetic compartments of WWER 440 NPPs fulfil the function of the containment used at NPPs all over the world. The purpose of the containment is to protect the NPP personnel against radioactive impact as well as to prevent radioactive leakage to the. environ ent during a lost of coolant accident. Leak-tightness enhancement in NPPs with WWER 440/213 and WWER 440/230 reactors is an important safety issue. New procedures, measures and methods were adopted at NPPs in Mochovce, Jaslovske Bohunice, Dukovany and PAKS for leak identification and sealing works performed by VUEZ Levice. (authors)

  20. Feasibility study of applying reactor oscillator phase method at the RB reactor; Razmatranje mogucnosti primene fazne metode reaktorskog oscilatora na reaktoru RB

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, M; Kocic, A; Markovic, V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1965-11-15

    This paper decsribes the principles of amplitude and phase methods for applying reactor oscillator; experimental procedure and choice of optimum parameters for usractor oscillator at the RB reactor, dependent on the values of absorption properties of moderator and construction materials. Short description of the oscillator and the electronic equipment is included.