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Sample records for vver reactor pressure

  1. 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)

  2. 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)

  3. 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)

  4. 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)

  5. 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

  6. 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.

  7. 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.

  8. 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.

  9. 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.)

  10. 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

  11. 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)

  12. 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)

  13. 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

  14. 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.)

  15. 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)

  16. 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)

  17. 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.

  18. 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

  19. 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)

  20. Ensuring the nuclear safety of VVER-440 reactor pressure vessels in Skoda, Concern Enterprise, Plzen

    International Nuclear Information System (INIS)

    Hrbek, Z.

    1985-01-01

    Various types of routine inspections are described of reactor pressure vessels with the aim of identifying residual lifetime and overall safety. The inspection programme includes: choice of systems and instruments, type of tests, test frequency, safety criteria, measures to be taken in case of unsatisfactory results, documentation. The criteria are given for periodical inspections and requirements listed for instruments and equipment. The main three groups of tests are: visual inspection and dimension tests, surface inspection and volumetric inspection. Briefly described is some of the equipment used. (M.D.)

  1. 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)

  2. 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

  3. 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)

  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. 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)

  8. 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)

  9. 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

  10. 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.

  11. 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

  12. 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

  13. 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

  14. 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)

  15. 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)

  16. Design and Computational Fluid Dynamics Optimization of the Tube End Effector for Reactor Pressure Vessel Head Type VVER-1000

    International Nuclear Information System (INIS)

    Novosel, D.

    2006-01-01

    In this paper is presented development and optimization of the tube end effector design which should consist of 4 ultrasonic transducers, 4 Eddy Current's transducers and Radiation Proof Dot Camera. Basically, designing was conducted by main input requests, such as: inner diameter of a tested reactor pressure vessel head penetration tube, dimensions of a transducers and maximum allowable vertical movement of a manipulator connection rod in order to cover all inner tube surface. As is obvious, for ultrasonic testing should be provided the thin layer of liquid material (in our case water was chosen) which is necessary to make physical contact between transducer surface and investigated inner tube surface. By help of Computational Fluid Dynamics, determined were parameters of geometry, as the most important factor of transducer housing, hydraulically parameters for water supply and primary drain together implemented into this housing, movement of the end effectors (vertical and cylindrical) and finally, necessary equipment which has to provide all hydraulically and pneumatic requirements. As the cylindrical surface of the inner tube diameter was liquefied and contact between transducer housing and tested tube wasn't ideally covered, water leakage could occur in downstream direction. To reduce water leakage, which is highly contaminated, developed was second water drain by diffuser assembly which is driven by Venturi pipe, commercially called vacuum generator. Using the Computational Fluid Dynamic, obtained was optimized geometry of diffuser control volume with the highest efficiency, in other words, unobstructed fluid flux. Afterwards, the end effectors system was synchronized to the existing operable system for NDT methods all invented and designed by INETEC. (author)

  17. 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)

  18. 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

  19. 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

  20. 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.

  1. 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)

  2. 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.

  3. 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)

  4. 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)

  5. 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)

  6. 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

  7. 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

  8. 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)

  9. 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)

  10. 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

  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. 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…

  14. 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)

  15. 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.

  16. 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.)

  17. 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

  18. EPR (European Pressurized Reactor)

    International Nuclear Information System (INIS)

    2015-01-01

    This document presents the EPR (European Pressurized Reactor), a modernised version of PWRs which uses nuclear fission. It indicates to which category it belongs (third generation). It briefly describes its operation: recalls on nuclear fission, electricity production in a nuclear reactor. It presents and comments its characteristics: power, thermal efficiency, redundant systems for safety control, double protective enclosure, expected lifetime, use of MOX fuel, modular design. It discusses economic stakes (expected higher nuclear electricity competitiveness, but high construction costs), and safety challenges (design characteristics, critics by nuclear safety authorities about the safety data processing system). It presents the main involved actors (Areva, EDF) and competitors in the field of advanced reactors (Rosatom with its VVER 1200, General Electric with its ABWR and its ESBWR, Mitsubishi with its APWR, Westinghouse with its AP100) while outlining the importance of certifications and delays to obtain them. After having evoked key data on EPR fuel consumption, it indicates reactors under construction, evokes potential markets and perspectives

  19. 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

  20. 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

  1. Possible pressurized thermal shock events during large primary to secondary leakage. The Hungarian AGNES project and PRISE accident scenarios in VVER-440/V213 type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Perneczky, L. [KFKI Atomic Energy Research Inst., Budabest (Hungary)

    1997-12-31

    Nuclear power plants of WWER-440/213-type have several special features. Consequently, the transient behaviour of such a reactor system should be different from the behaviour of the PWRs of western design. The opening of the steam generator (SG) collector cover, as a specific primary to secondary circuit leakage (PRISE) occurring in WWER-type reactors happened first time in Rovno NPP Unit I on January 22, 1982. Similar accident was studied in the framework of IAEA project RER/9/004 in 1987-88 using the RELAP4/mod6 code. The Hungarian AGNES (Advanced General and New Evaluation of Safety) project was performed in the period 1991-94 with the aim to reassess the safety of the Paks NPP using state-of-the-art techniques. The project comprised three type of analyses for the primary to secondary circuit leakages: Design Basis Accident (DBA) analyses, Pressurized Thermal Shock (PTS) study and deterministic analyses for Probabilistic Safety Analysis (PSA). Major part of the thermohydraulic analyses has been performed by the RELAP5/mod2.5/V251 code version with two input models. 32 refs.

  2. Possible pressurized thermal shock events during large primary to secondary leakage. The Hungarian AGNES project and PRISE accident scenarios in VVER-440/V213 type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Perneczky, L [KFKI Atomic Energy Research Inst., Budabest (Hungary)

    1998-12-31

    Nuclear power plants of WWER-440/213-type have several special features. Consequently, the transient behaviour of such a reactor system should be different from the behaviour of the PWRs of western design. The opening of the steam generator (SG) collector cover, as a specific primary to secondary circuit leakage (PRISE) occurring in WWER-type reactors happened first time in Rovno NPP Unit I on January 22, 1982. Similar accident was studied in the framework of IAEA project RER/9/004 in 1987-88 using the RELAP4/mod6 code. The Hungarian AGNES (Advanced General and New Evaluation of Safety) project was performed in the period 1991-94 with the aim to reassess the safety of the Paks NPP using state-of-the-art techniques. The project comprised three type of analyses for the primary to secondary circuit leakages: Design Basis Accident (DBA) analyses, Pressurized Thermal Shock (PTS) study and deterministic analyses for Probabilistic Safety Analysis (PSA). Major part of the thermohydraulic analyses has been performed by the RELAP5/mod2.5/V251 code version with two input models. 32 refs.

  3. 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

  4. 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)

  5. 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.

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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.)

  11. 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.)

  12. 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)

  13. 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

  14. 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

  15. 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.)

  16. 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)

  17. 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

  18. 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)

  19. 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)

  20. 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.

  1. 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)

  2. 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.

  3. 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

  4. 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.

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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)

  11. 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)

  12. Probabilistic approach to the analysis of reactor pressure vessel integrity during a pressurized thermal shock

    International Nuclear Information System (INIS)

    Adamec, P.

    2000-12-01

    Following a general summary of the issue, an overview of international experience (USA; Belgium, France, Germany, Russia, Spain, Sweden, The Netherlands, and the UK; and probabilistic PTS assessment for the reactor pressure vessel at Loviisa-1, Finland) is presented, and the applicable computer codes (VISA-II, OCA-P, FAVOR, ZERBERUS) are highlighted and their applicability to VVER type reactor pressure vessels is outlined. (P.A.)

  13. 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

  14. 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

  15. 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.

  16. 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.

  17. 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)

  18. 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.)

  19. 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)

  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. 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

  2. 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.

  3. 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)

  4. 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

  5. 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

  6. 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)

  7. Research on the reactor physics and reactor safety of VVER reactors. AER Symposium 2016

    Energy Technology Data Exchange (ETDEWEB)

    Kliem, S.

    2017-09-15

    The selected paperscan be attributed to the following main subjects: Reactor start-up tests and use of corresponding data for code validation, code development and application, approaches for safety analyses, closure of nuclear fuel cycle, prospective reactor concepts.

  8. 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.)

  9. 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)

  10. 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

  11. 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)

  12. 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

  13. 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.

  14. 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)

  15. 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.

  16. 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)

  17. Pressure tube type reactors

    International Nuclear Information System (INIS)

    Komada, Masaoki.

    1981-01-01

    Purpose: To increase the safety of pressure tube type reactors by providing an additional ECCS system to an ordinary ECCS system and injecting heavy water in the reactor core tank into pressure tubes upon fractures of the tubes. Constitution: Upon fractures of pressure tubes, reduction of the pressure in the fractured tubes to the atmospheric pressure in confirmed and the electromagnetic valve is operated to completely isolate the pressure tubes from the fractured portion. Then, the heavy water in the reactor core tank flows into and spontaneously recycles through the pressure tubes to cool the fuels in the tube to prevent their meltdown. By additionally providing the separate ECCS system to the ordinary ECCS system, fuels can be cooled upon loss of coolant accidents to improve the safety of the reactors. (Moriyama, K.)

  18. State space modeling of reactor core in a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ashaari, A.; Ahmad, T.; M, Wan Munirah W. [Department of Mathematical Science, Faculty of Science, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Shamsuddin, Mustaffa [Institute of Ibnu Sina, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Abdullah, M. Adib [Swinburne University of Technology, Faculty of Engineering, Computing and Science, Jalan Simpang Tiga, 93350 Kuching, Sarawak (Malaysia)

    2014-07-10

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  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. 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.

  1. 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

  2. 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.

  3. 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

  4. Reactor pressure vessel support

    International Nuclear Information System (INIS)

    Butti, J.P.

    1977-01-01

    A link and pin support system provides the primary vertical and lateral support for a nuclear reactor pressure vessel without restricting thermally induced radial and vertical expansion and contraction. (Auth.)

  5. Pressurized water reactor systems

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1975-01-01

    Design and mode of operation of the main PWR components are described: reactor core, pressure vessel and internals, cooling systems with pumps and steam generators, ancillary systems, and waste processing. (TK) [de

  6. Reactor pressure vessel design

    International Nuclear Information System (INIS)

    Foehl, J.

    1998-01-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. In chapter 2, the general principles of reactor pressure vessel design are elaborated. Crack and fracture initiation and propagation are treated in some detail

  7. 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 <reactor coolant and containment air and fluid samples that support accurate analytical results for the parameters of interest are considered. The radiological and chemistry parameters to be monitored for primary coolant and sump and for containment air are specified. (author)

  8. Reactor pressure tank

    International Nuclear Information System (INIS)

    Dorner, H.; Scholz, M.; Jungmann, A.

    1975-01-01

    In a reactor pressure tank for a nuclear reactor, self-locking hooks engage a steel ring disposed over the removable cover of the steel vessel. The hooks exert force upon the cover to maintain the cover in a closed position during operation of the reactor pressure tank. The force upon the removal cover is partly the result of the increasing temperature and thermal expansion of the steel vessel during operation. The steel vessel is surrounded by a reinforced-concrete tank. (U.S.)

  9. Reactor pressure boundary materials

    International Nuclear Information System (INIS)

    Hong, Jun Hwa; Chi, S. H.; Lee, B. S.

    2002-04-01

    With a long-term operation of nuclear power plants, the component materials are degraded under severe reactor conditions such as neutron irradiation, high temperature, high pressure and corrosive environment. It is necessary to establish the reliable and practical technologies for improving and developing the component materials and for evaluating the mechanical properties. Especially, it is very important to investigate the technologies for reactor pressure boundary materials such as reactor vessel and pipings in accordance with their critical roles. Therefore, this study was focused on developing and advancing the microstructural/micro-mechanical evaluation technologies, and on evaluating the neutron irradiation characteristics and radiation effects analysis technology of the reactor pressure boundary materials, and also on establishing a basis of nuclear material property database

  10. 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.)

  11. Pressure tube reactor

    International Nuclear Information System (INIS)

    Seki, Osamu; Kumasaka, Katsuyuki.

    1988-01-01

    Purpose: To remove the heat of reactor core using a great amount of moderators at the periphery of the reactor core as coolants. Constitution: Heat of a reactor core is removed by disposing a spontaneous recycling cooling device for cooling moderators in a moderator tank, without using additional power driven equipments. That is, a spontaneous recycling cooling device for cooling the moderators in the moderator tank is disposed. Further, the gap between the inner wall of a pressure tube guide pipe disposed through the vertical direction of a moderator tank and the outer wall of a pressure tube inserted through the guide pipe is made smaller than the rupture distortion caused by the thermal expansion upon overheating of the pressure tube and greater than the minimum gap required for heat shiels between the pressure tube and the pressure tube guide pipe during usual operation. In this way, even if such an accident as can not using a coolant cooling device comprising power driven equipment should occur in the pressure tube type reactor, the rise in the temperature of the reactor core can be retarded to obtain a margin with time. (Kamimura, M.)

  12. 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.

  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. 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)

  15. 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'.

  16. 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.

  17. 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.

  18. Pressurized-water reactors

    International Nuclear Information System (INIS)

    Bush, S.H.

    1983-03-01

    An overview of the pressurized-water reactor (PWR) pressure boundary problems is presented. Specifically exempted will be discussions of problems with pumps, valves and steam generators on the basis that they will be covered in other papers. Pressure boundary reliability is examined in the context of real or perceived problems occurring over the past 5 to 6 years since the last IAEA Reliability Symposium. Issues explicitly covered will include the status of the pressurized thermal-shock problem, reliability of inservice inspections with emphasis on examination of the region immediately under the reactor pressure vessel (RPV) cladding, history of piping failures with emphasis on failure modes and mechanisms. Since nondestructive examination is the topic of one session, discussion will be limited to results rather than techniques

  19. 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

  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. Coolant mixing in pressurized water reactors. Proceedings

    International Nuclear Information System (INIS)

    Hoehne, T.; Grunwald, G.; Rohde, U.

    1998-10-01

    For the analysis of boron dilution transients and main steam like break scenarios the modelling of the coolant mixing inside the reactor vessel is important. The reactivity insertion due to overcooling or deboration depends strongly on the coolant temperature and boron concentration. The three-dimensional flow distribution in the downcomer and the lower plenum of PWR's was calculated with a computational fluid dynamics (CFD) code (CFX-4). Calculations were performed for the PWR's of SIEMENS KWU, Westinghouse and VVER-440 / V-230 type. The following important factors were identified: exact representation of the cold leg inlet region (bend radii etc.), extension of the downcomer below the inlet region at the PWR Konvoi, obstruction of the flow by the outlet nozzles penetrating the downcomer, etc. The k-ε turbulence model was used. Construction elements like perforated plates in the lower plenum have large influence on the velocity field. It is impossible to model all the orifices in the perforated plates. A porous region model was used to simulate perforated plates and the core. The porous medium is added with additional body forces to simulate the pressure drop through perforated plates in the VVER-440. For the PWR Konvoi the whole core was modelled with porous media parameters. The velocity fields of the PWR Konvoi calculated for the case of operation of all four main circulation pumps show a good agreement with experimental results. The CFD-calculation especially confirms the back flow areas below the inlet nozzles. The downcomer flow of the Russian VVER-440 has no recirculation areas under normal operation conditions. By CFD calculations for the downcomer and the lower plenum an analytical mixing model used in the reactor dynamic code DYN3D was verified. The measurements, the analytical model and the CFD-calculations provided very well agreeing results particularly for the inlet region. The difficulties of analytical solutions and the uncertainties of turbulence

  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. 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

  4. 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.

  5. 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)

  6. 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)

  7. 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

  8. 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.)

  9. Pressure tube reactor

    International Nuclear Information System (INIS)

    Susuki, Akira; Murata, Shigeto; Minato, Akihiko.

    1993-01-01

    In a pressure tube reactor, a reactor core is constituted by arranging more than two units of a minimum unit combination of a moderator sealing pipe containing a calandria tube having moderators there between and a calandria tube and moderators. The upper header and a lower header of the calandria tank containing moderators are communicated by way of the moderator sealing tube. Further, a gravitationally dropping mechanism is disposed for injecting neutron absorbing liquid to a calandria gas injection portion. A ratio between a moderator volume and a fuel volume is defined as a function of the inner diameter of the moderator sealing tube, the outer diameter of the calandria tube and the diameter of fuel pellets, and has no influence to intervals of a pressure tube lattice. The interval of the pressure tube lattice is enlarged without increasing the size of the pressure tube, to improve production efficiency of the reactor and set a coolant void coefficient more negative, thereby enabling to improve self controllability and safety. Further, the reactor scram can be conducted by injecting neutron absorbing liquid. (N.H.)

  10. Pressure tube reactor

    International Nuclear Information System (INIS)

    Matsumoto, Tomoyuki; Fujino, Michihira.

    1980-01-01

    Purpose: To equalize heavy water flow distribution by providing a nozzle for externally injecting heavy water from a vibration preventive plate to the upper portion to feed the heavy water in a pressure tube reactor and swallowing up heavy water in a calandria tank to supply the heavy water to the reactor core above the vibration preventive plate. Constitution: A moderator injection nozzle is mounted on the inner wall of a calandria tank. Heavy water is externally injected above the vibration preventive plate, and heavy water in the calandria tank is swallowed up to supply the heavy water to the core reactor above the vibration preventive plate. Therefore, the heavy water flow distribution can be equalized over the entire reactor core, and the distribution of neutron absorber dissolved in the heavy water is equalized. (Yoshihara, H.)

  11. 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)

  12. 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)

  13. 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

  14. 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)

  15. 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)

  16. Reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Van De Velde, J.; Fabry, A.; Van Walle, E.; Chaouuadi, R.

    1998-01-01

    Research and development activities related to reactor pressure vessel steels during 1997 are reported. The objectives of activities of the Belgian Nuclear Research Centre SCK/CEN in this domain are: (1) to develop enhanced surveillance concepts by applying micromechanics and fracture-toughness tests to small specimens, and by performing damage modelling and microstructure characterization; (2) to demonstrate a methodology on a broad database; (3) to achieve regulatory acceptance and industrial use

  17. Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Van de Velde, J.; Fabry, A.; Van Walle, E.; Chaoudi, R

    1998-07-01

    SCK-CEN's R and D programme on Reactor Pressure Vessel (RPV) Steels in performed in support of the RVP integrity assessment. Its main objectives are: (1) to develop enhanced surveillance concepts by applying micromechanics and fracture-toughness tests to small specimens, and by performing damage modelling and microstructure characterization; (2) to demonstrate the applied methodology on a broad database; (3) to achieve regulatory acceptance and industrial use. Progress and achievements in 1999 are reported.

  18. The pressurized water reactor

    International Nuclear Information System (INIS)

    Gallagher, J.L.

    1987-01-01

    Pressurized water reactor technology has reached a maturity that has engendered a new surge of innovation, which in turn, has led to significant advances in the technology. These advances, characterized by bold thinking but conservative execution, are resulting in nuclear plant designs which offer significant performance and safety improvements. This paper describes the innovations which are being designed into mainstream PWR technology as well as the desings which are resulting from such innovations. (author)

  19. 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.

  20. 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.

  1. 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

  2. Pressure tube reactors

    International Nuclear Information System (INIS)

    Natori, Hisahide.

    1981-01-01

    Purpose: To improve the electrical power generation efficiency in a pressure tube reactor in which coolants and moderators are separated by feedwater heating with heat generated in heavy water and by decreasing the amount of steams to be extracted from the turbine. Constitution: A heat exchanger and a heavy water cooler are additionally provided to a conventional pressure tube reactor. The heat exchanger is disposed at the pre-stage of a low pressure feedwater heater series. High temperature heavy water heated in the core is passed through the primary side of the exchanger, while feedwater is passed through the secondary side. The cooler is disposed on the downstream of the heat exchanger in the flowing direction of the heavy water, in which heavy water from the heat exchanger is passed through the primary side and the auxiliary equipment cooling water is sent to the secondary side thereof. Accordingly, since extraction of heating steams is no more necessary, the steam can be used for the rotation of the turbine, and the electrical power generation efficiency can be improved. (Seki, T.)

  3. 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)

  4. 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

  5. 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)

  6. Reactor pressure vessel embrittlement

    International Nuclear Information System (INIS)

    1992-07-01

    Within the framework of the IAEA extrabudgetary programme on the Safety of WWER-440/230 NPPs, a list of safety issues requiring broad studies of generic interest have been agreed upon by an Advisory Group who met in Vienna in September 1990. The list was later revised in the light of the programme findings. The information on the status of the issues, and on the amount of work already completed and under way in the various countries, needs to be compiled. Moreover, an evaluation of what further work is required to resolve each one of the issues is also necessary. In view of this, the IAEA has started the preparation of a series of status reports on the various issues. This report on the generic safety issue ''Reactor Pressure Vessel Embrittlement'' presents a comprehensive survey of technical information available in the field and identifies those aspects which require further investigation. 39 refs, 21 figs, 4 tabs

  7. Pressurized water reactor with reactor pressure vessel

    International Nuclear Information System (INIS)

    Werres, L.

    1985-01-01

    The pressure vessel has a cylindrical jacket with a domed floor. A guide is arranged on the domed floor to even out the flow in the core. It consists of a cylindrical jacket, whose lower end has slots and fins. These fins are welded to the domed floor. (orig./PW)

  8. Pressurized water reactor with reactor pressure vessel

    International Nuclear Information System (INIS)

    Werres, L.

    1980-01-01

    The pressure vessel has a cylindrical jacket with a domed floor. A guide is arranged on the domed floor to even out the flow in the core. It consists of a cylindrical jacket, whose lower end has slots and fins. These fins are welded to the domed floor. (DG) [de

  9. 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)

  10. 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)

  11. 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)

  12. 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

  13. 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

  14. Fundamentals of pressurized water reactors

    International Nuclear Information System (INIS)

    Murray, L.

    1982-01-01

    In many countries, the pressurized water reactor (PWR) is the most widely used, even though it requires enrichment of the uranium to about 3% in U-235 and the moderator-coolant must be maintained at a high pressure, about 2200 pounds per square inch. Our objective in this series of seven lectures is to describe the design and operating characteristics of the PWR system, discuss the reactor physics methods used to evaluate performance, examine the way fuel is consumed and produced, study the instrumentation system, review the physics measurements made during initial startup of the reactor, and outline the administrative aspects of starting up a reactor and operating it safely and effectively

  15. 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.

  16. Reactor vessel pressure transient protection for pressurized water reactors

    International Nuclear Information System (INIS)

    Zech, G.

    1978-09-01

    During the past few years the NRC has been studying the issue of protection of the reactor pressure vessels at Pressurized Water Reactors (PWRs) from transients when the vessels are at a relatively low temperature. This effort was prompted by concerns related to the safety margins available to vessel damage as a result of such events. Nuclear Reactor Regulation Category A Technical Activity No. A-26 was established to set forth the NRC plan for resolution of the generic aspects of this safety issue. The purpose of the report is to document the completion of this generic technical activity

  17. Pressurized water reactor flow arrangement

    International Nuclear Information System (INIS)

    Gibbons, J.F.; Knapp, R.W.

    1980-01-01

    A flow path is provided for cooling the control rods of a pressurized water reactor. According to this scheme, a small amount of cooling water enters the control rod guide tubes from the top and passes downwards through the tubes before rejoining the main coolant flow and passing through the reactor core. (LL)

  18. 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)

  19. 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)

  20. Effects of irradiation at lower temperature on the microstructure of Cr-Mo-V-alloyed reactor pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Grosse, M; Boehmert, J; Gilles, R [Hahn-Meitner-Institut Berlin GmbH (Germany)

    1998-10-01

    The microstructural damage process due to neutron irradiation [1] proceeds in two stages: - formation of displacement cascades - evolution of the microstructure by defect reactions. Continuing our systematic investigation about the microstructural changes of Russian reactor pressure vessel steel due to neutron irradiation the microstructure of two laboratory heats of the VVER 440-type reactor pressure vessel steel after irradiation at 60 C was studied by small angle neutron scattering (SANS). 60 C-irradiation differently changes the irradiation-induced microstructure in comparison with irradiation at reactor operation temperature and can, thus, provide new insights into the mechanisms of the irradiation damage. (orig.)

  1. 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)

  2. 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

  3. Reactor pressure vessel status report

    International Nuclear Information System (INIS)

    Strosnider, J.; Wichman, K.; Elliot, B.

    1994-12-01

    This report gives a brief description of the reactor pressure vessel (RPV), followed by a discussion of the radiation embrittlement of RPV beltline materials and the two indicators for measuring embrittlement, the end-of-license (EOL) reference temperature and the EOL upper-shelf energy. It also summarizes the GL 92-01 effort and presents, for all 37 boiling water reactor plants and 74 pressurized water reactor plants in the United States, the current status of compliance with regulatory requirements related to ensuring RPV integrity. The staff has evaluated the material data needed to predict neutron embrittlement of the reactor vessel beltline materials. These data will be stored in a computer database entitled the reactor vessel integrity database (RVID). This database will be updated annually to reflect the changes made by the licensees in future submittals and will be used by the NRC staff to assess the issues related to vessel structural integrity

  4. 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)

  5. Reactor Structural Materials: Reactor Pressure Vessel Steels

    International Nuclear Information System (INIS)

    Chaouadi, R.

    2000-01-01

    The objectives of SCK-CEN's R and D programme on Rector Pressure Vessel (RPV) Steels are:(1) to complete the fracture toughness data bank of various reactor pressure vessel steels by using precracked Charpy specimens that were tested statically as well as dynamically; (2) to implement the enhanced surveillance approach in a user-friendly software; (3) to improve the existing reconstitution technology by reducing the input energy (short cycle welding) and modifying the stud geometry. Progress and achievements in 1999 are reported

  6. Borate compound content reduction in liquid radioactive waste from nuclear power plants with VVER reactor

    International Nuclear Information System (INIS)

    Szalo, A.; Zatkulak, M.

    2000-01-01

    This paper describes the current status of liquid waste (evaporator concentrates) inventory at V-1 and V-2 NPPs in Jaslovske Bohunice and the intention to separate boron from them with respect to waste minimisation and improvement of physical and chemical properties for further waste treatment and conditioning. Preliminary results of laboratory experiments concerned to borate crystallisation after pH adjustment with nitric or formic acid performed in the 1998 are given. At the present time laboratory experiments continuing - next acids, coagulation with carbon oxide, electrolytic process, ion exchange resin, study of decontamination factors, immobilization of boric acid, decrease radioactivity, purification of boron-contained compounds. Slovenske Elektrarne have accumulated 7,000 m 3 of evaporator concentrates containing 100-180 g/l borate. In order to make more storage space available, it is proposed to remove some of the borate in the liquor by precipitation as sodium tetraborate and immobilise in either cement of bitumen. The supernate can be further volume reduced by evaporation and returned to the tanks. Slovenske Elektrarne are currently evaluating acid addition to the pH 12-13 concentrate to reduce the borate solubility. However, this adds to the salt burden of the waste through this chemical addition -thus creating future increases in conditioning and disposal costs. Boric acid is used in pressurized water as a soluble neutron poison to control reactivity and also to assure a safety margin in the spent fuel pool and during refuelling operations. Boric acid is also present in the water reserved for injection into the reactor in the event of postulated accidents. (author)

  7. 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

  8. 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)

  9. 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.

  10. 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)

  11. Power reactor pressure vessel benchmarks

    International Nuclear Information System (INIS)

    Rahn, F.J.

    1978-01-01

    A review is given of the current status of experimental and calculational benchmarks for use in understanding the radiation embrittlement effects in the pressure vessels of operating light water power reactors. The requirements of such benchmarks for application to pressure vessel dosimetry are stated. Recent developments in active and passive neutron detectors sensitive in the ranges of importance to embrittlement studies are summarized and recommendations for improvements in the benchmark are made. (author)

  12. Experience with safety assessment of digital upgrading of IandC in VVER type reactors

    International Nuclear Information System (INIS)

    Wach, D.; Mulka, B.; Schnuerer, G.

    1997-01-01

    The digital upgrading of IandC systems important to safety in WWER type reactors requires a broad expertise in various knowledge fields. The approach of the Institute for safety Technology to the qualification and categorization of safety-critical software systems is highlighted. The role of the Institute in the qualification of the Teleperm XS and the type testing of its components is described. The aspects of the safety assessment of digital IandC systems in WWER type reactors is discussed in some detail. (A.K.)

  13. Pressure tube reactor

    International Nuclear Information System (INIS)

    Kanazawa, Nobuhiro; Kaneto, Kunikazu.

    1979-01-01

    Purpose: To attain uniform fluid poison distribution in a calandria tank by downwardly projecting, at an equal distance to the reactor core, a spacer wall from the periphery of an anti-vibration plate in the vicinity of a heavy water flow passage in the periphery of the anti-vibration plate, thereby decrease the amount of heavy water flowing into the heavy water flow passage. Constitution: A projecting wall concentrical with a calandria tank is suspended vertically from the boundary side at the peripheral portion of an anti-vibration plate to a water heavy flow passage in the periphery of the anti-vibration plate. The projecting wall has such a vertical length as about equal to the width of the heavy water flow passage, prevents heavy water flowing through apertures of a control rod guide tube from entering into the heavy water passage and increases the ratio of heavy water that flows through the heavy water flow passage in the anti-vibration plate. Consequently, if the liquid poison density in heavy water is varied, the ununiform poison density in the calandria tank can be prevented. (Seki, T.)

  14. Pressure tube type research reactor

    International Nuclear Information System (INIS)

    Ueda, Hiroshi.

    1975-01-01

    Object: To permit safe and reliable replacement of primary pipes by providing a reactor container so as to surround a pressure pipe, with upper portions of the two separably coupled together, and coupling the pressure pipe and primary piping by joint coupling above and below the reactor container, with the lower coupling joint surrounded by drain receptacle. Structure: At the time of replacement of a pressure pipe, a partition valve is opened to exhaust primary cooling water within pressure pipe and upper and lower portions of the primary piping and replace the decelerator within the reactor container with water of the same quality as that of pool water within an upper shield pool. Thereafter, the entire space above the drain receptacle is filled with pool water by closing a partition valve and opening a water supply valve. Then, upper portion seal cover, pool bottom lid, upper joint and upper portion primary piping are removed, then bolts and nuts are loosened, and the pressure pipe is taken out together with the shield block. (Kamimura, M.)

  15. Pressurized water reactor inspection procedures

    International Nuclear Information System (INIS)

    Heinrich, D.; Mueller, G.; Otte, H.J.; Roth, W.

    1998-01-01

    Inspections of the reactor pressure vessels of pressurized water reactors (PWR) so far used to be carried out with different central mast manipulators. For technical reasons, parallel inspections of two manipulators alongside work on the refueling cavity, so as to reduce the time spent on the critical path in a revision outage, are not possible. Efforts made to minimize the inspection time required with one manipulator have been successful, but their effects are limited. Major reductions in inspection time can be achieved only if inspections are run with two manipulators in parallel. The decentralized manipulator built by GEC Alsthom Energie and so far emmployed in boiling water reactors in the USA, Spain, Switzerland and Japan allows two systems to be used in parallel, thus reducing the time required for standard inspection of a pressure vessel from some six days to three days. These savings of approximately three days are made possible without any compromises in terms of positioning by rail-bound systems. During inspection, the reactor refueling cavity is available for other revision work without any restrictions. The manipulator can be used equally well for inspecting standard PWR, PWR with a thermal shield, for inspecting the land between in-core instrumentation nozzles, BWR with and without jet pumps (complementary inspection), and for inspecting core support shrouds. (orig.) [de

  16. 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

  17. The European pressurized water reactor

    International Nuclear Information System (INIS)

    Leny, J.C.

    1993-01-01

    The present state of development of the European Pressurized Water Reactor (EPR) is outlined. During the so-called harmonization phase, the French and German utilities drew up their common requirements and evaluated the reactor concept developed until then with respect to these requirements. A main result of the harmonization phase was the issue, in September 1993, of the 'EPR Conceptual Safety Feature Review File' to be jointly assessed by the safety authorities in France and Germany. The safety objectives to be met by the EPR are specified in the second part of the paper, and some details of the primary and secondary side safety systems are given. (orig.) [de

  18. Pressurized water reactor with a reactor pressure vessel

    International Nuclear Information System (INIS)

    Werres, L.

    1979-01-01

    The core barrel is suspended from a flange by means of a grid. The coolant enters the barrel from below through the grid. In order to get a uniform flow over the reactor core there is provided for a guiding device below the grid. It consists of a cylindrical shell with borings uniformly distributed around the shell as well as fins on the inner surface of the shell and slots at the bottom facing the pressure vessel. (GL) [de

  19. Pressure vessel for nuclear reactors

    International Nuclear Information System (INIS)

    1975-01-01

    The invention applies to a pressure vessel for nuclear reactors whose shell, made of cast metal segments, has a steel liner. This liner must be constructed to withstand all operational stresses and to be easily repairable. The invention solves this problem by installing the liner at a certain distance from the inner wall of the pressure vessel shell and by filling this clearance with supporting concrete. Both the concrete and the steel liner must have a lower prestress than the pressure vessel shell. In order to avoid damage to the liner when prestressing the pressure vessel shell, special connecting elements are provided which consist of welded-on fastening elements projecting into recesses in the cast metal segments of the pressure vessel. Their design is described in detail. (TK) [de

  20. 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)

  1. 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.

  2. 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.

  3. Analytical and experimental justification of safe operation of fuel loads of VVER reactors at Rovno NPP

    International Nuclear Information System (INIS)

    Andrianov, A.; Zagrebelny, L.

    2011-01-01

    The main task during the nuclear fuel operation us ensuring of the larger fuel burnup and as a result - reduction of the fuel constituent in the electricity production cost. The neutron-physic calculations are performed using the qualified codes BIPR-7A and PERMAK developed by Kurchatov Institute. The limits for calculated parameters are set by the Ukrainian regulations. Calculation results are documented in reports subject for independent expert review requested by the regulatory authority. In this report the following item have presented: 1) metrological check and calibration of measuring channels; 2) fuel cycles at Rivne NPP; 3) determination of experimental values of thermal-physic and neutron-physic parameters; 4) ICIS equipment check, execution of the program confirming correct connection of the temperature and neutron flux monitoring sensors; 5) monitoring of the core TPh and NPh parameters in all operating modes; 6) monitoring of the fuel condition in the core and 7) FE leak tightness monitoring at the operating reactor

  4. 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.

  5. 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

  6. INETEC new system for inspection of PWR reactor pressure vessel head

    International Nuclear Information System (INIS)

    Nadinic, B.; Postruzin, Z.

    2004-01-01

    INETEC Institute for Nuclear Technology developed new equipment for inspection of PWR and VVER reactor pressure vessel head. The new advances in inspection technology are presented in this article, as the following: New advance manipulator for inspection of RPVH with high speed of inspection possibilities and total automated work; New sophisticated software for manipulator driving which includes 3D virtual presentation of manipulator movement and collision detection possibilities; New multi axis controller MAC-8; New end effector system for inspection of penetration tube and G weld; New eddy current and ultrasonic probes for inspection of G weld and penetration tube; New Eddy One Raster scan software for analysis of eddy current data with mant advanced features which allows easy and quick data analysis. Also the results of laboratory testing and laboratory qualification are presented on reactor pressure vessel head mock, as well as obtained speed of inspection and quality of collected data.(author)

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

    International Nuclear Information System (INIS)

    Martin, M.

    1997-01-01

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

  8. 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

  9. Non-Proliferative, Thorium-Based, Core and Fuel Cycle for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Todosow, M.; Raitses, G.; Galperin, A.

    2009-01-01

    Two of the major barriers to the expansion of worldwide adoption of nuclear power are related to proliferation potential of the nuclear fuel cycle and issues associated with the final disposal of spent fuel. The Radkowsky Thorium Fuel (RTF) concept proposed by Professor A. Radkowsky offers a partial solution to these problems. The main idea of the concept is the utilization of the seed-blanket unit (SBU) fuel assembly geometry which is a direct replacement for a 'conventional' assembly in either a Russian pressurized water reactor (VVER-1000) or a Western pressurized water reactor (PWR). The seed-blanket fuel assembly consists of a fissile (U) zone, known as seed, and a fertile (Th) zone known as blanket. The separation of fissile and fertile allows separate fuel management schemes for the thorium part of the fuel (a subcritical 'blanket') and the 'driving' part of the core (a supercritical 'seed'). The design objective for the blanket is an efficient generation and in-situ fissioning of the U233 isotope, while the design objective for the seed is to supply neutrons to the blanket in a most economic way, i.e. with minimal investment of natural uranium. The introduction of thorium as a fertile component in the nuclear fuel cycle significantly reduces the quantity of plutonium production and modifies its isotopic composition, reducing the overall proliferation potential of the fuel cycle. Thorium based spent fuel also contains fewer higher actinides, hence reducing the long-term radioactivity of the spent fuel. The analyses show that the RTF core can satisfy the requirements of fuel cycle length, and the safety margins of conventional pressurized water reactors. The coefficients of reactivity are comparable to currently operating VVER's/PWR's. The major feature of the RTF cycle is related to the total amount of spent fuel discharged for each cycle from the reactor core. The fuel management scheme adopted for RTF core designs allows a significant decrease in the

  10. 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

  11. Some aspects of reactor pressure vessel integrity

    International Nuclear Information System (INIS)

    Korosec, D.; Vojvodic, G.J.

    1996-01-01

    Reactor pressure vessel of the pressurized water reactor nuclear power plant is the subject of extreme interest due to the fact that presents the pressure boundary of the reactor coolant system, which is under extreme thermal, mechanical and irradiation effects. Reactor pressure vessel by itself prevents the release of fission products to the environment. Design, construction and in-service inspection of such component is governed by strict ASME rules and other forms of administrative control. The reactor pressure vessel in nuclear power plant Kriko is designed and constructed in accordance with related ASME rules. The in-service inspection program includes all requests presented in ASME Code section XI. In the present article all major requests for the periodic inspections of reactor pressure vessel and fracture mechanics analysis are discussed. Detailed and strict fulfillment of all prescribed provisions guarantee the appropriate level of nuclear safety. (author)

  12. 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)

  13. 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

  14. 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.)

  15. 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.

  16. Ultrasound periodic inspections of reactor pressure vessels

    International Nuclear Information System (INIS)

    Haniger, L.

    1980-01-01

    Two versions are described of ultrasonic equipment for periodic inspections of reactor pressure vessels. One uses the principle of exchangeable programmators with solid-state logic while the other uses programmable logic with semiconductor memories. The equipment is to be used for inspections of welded joints on the upper part of the V-1 reactor pressure vessel. (L.O.)

  17. Influence of gamma irradiation on the deterioration of reactor pressure vessel materials and on reactor dosimetry measurements. Final report

    International Nuclear Information System (INIS)

    Boehmer, B.; Konheiser, J.; Kumpf, H.; Noack, K.; Vladimirov, P.

    2002-10-01

    Radiation embrittlement of pressure vessel steel in mixed neutron-gamma fields is mostly determined by neutrons, but in some cases also by gamma-radiation. Depending on the reactor type, gamma radiation can influence evaluations of lead factors of surveillance specimens, effect the interpretation of results of irradiation experiments and finally, it can result in changed pressure vessel lifetime evaluations. The project aimed at the evaluation of the importance of gamma radiation for RPV steel damage for several types of light-water reactors. Absolute neutron and gamma fluence rate spectra had been calculated for the Russian PWR types VVER-440 and two core loading variants of VVER-1000, for a German 1300 MW PWR and a German 900 MW BWR. Based on the calculated spectra several flux integrals and radiation damage parameters were derived for the region of the azimuthal flux maxima in the mid-planes for different radial positions between core and biological shield, especially, at the inner and outer surfaces of the PV walls, at the (1/4)-PV-thickness and at the surveillance positions. Fissionable materials are often used as activation detectors for neutron fluence measurements. To get the real value the analysis demands to take into account the gamma induced fissions. Therefore, the part of these fissions in the total number of fissions was estimated for the detector reactions 237 Np(n,f) and 238 U(n,f) in the calculated neutron/gamma fields. It has been found that considerable corrections of the neutron fluence measurements can be necessary, especially in case of 238 U(n,f). Most of the calculations were performed using a three-dimensional synthesis of 2D/1D-flux distributions obtained by the S N -code DORT with the BUGLE-96T group cross-section library. (orig.) [de

  18. Computerized reactor pressure vessel materials information system

    International Nuclear Information System (INIS)

    Strosnider, J.; Monserrate, C.; Kenworthy, L.D.; Tether, C.D.

    1980-10-01

    A computerized information system for storage and retrieval of reactor pressure vessel materials data was established, as part of Task Action Plan A-11, Reactor Vessel Materials Toughness. Data stored in the system are necessary for evaluating the resistance of reactor pressure vessels to flaw-induced fracture. This report includes (1) a description of the information system; (2) guidance on accessing the system; and (3) a user's manual for the system

  19. 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)

  20. 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.

  1. Reactor water spontaneous circulation structure in reactor pressure vessel

    International Nuclear Information System (INIS)

    Takahashi, Kazumi

    1998-01-01

    The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

  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. Reactor pressure vessel thermal annealing

    International Nuclear Information System (INIS)

    Lee, A.D.

    1997-01-01

    The steel plates and/or forgings and welds in the beltline region of a reactor pressure vessel (RPV) are subject to embrittlement from neutron irradiation. This embrittlement causes the fracture toughness of the beltline materials to be less than the fracture toughness of the unirradiated material. Material properties of RPVs that have been irradiated and embrittled are recoverable through thermal annealing of the vessel. The amount of recovery primarily depends on the level of the irradiation embrittlement, the chemical composition of the steel, and the annealing temperature and time. Since annealing is an option for extending the service lives of RPVs or establishing less restrictive pressure-temperature (P-T) limits; the industry, the Department of Energy (DOE) and the Nuclear Regulatory Commission (NRC) have assisted in efforts to determine the viability of thermal annealing for embrittlement recovery. General guidance for in-service annealing is provided in American Society for Testing and Materials (ASTM) Standard E 509-86. In addition, the American Society of Mechanical Engineers (ASME) Code Case N-557 addresses annealing conditions (temperature and duration), temperature monitoring, evaluation of loadings, and non-destructive examination techniques. The NRC thermal annealing rule (10 CFR 50.66) was approved by the Commission and published in the Federal Register on December 19, 1995. The Regulatory Guide on thermal annealing (RG 1.162) was processed in parallel with the rule package and was published on February 15, 1996. RG 1.162 contains a listing of issues that need to be addressed for thermal annealing of an RPV. The RG also provides alternatives for predicting re-embrittlement trends after the thermal anneal has been completed. This paper gives an overview of methodology and recent technical references that are associated with thermal annealing. Results from the DOE annealing prototype demonstration project, as well as NRC activities related to the

  4. Pressure vessel for nuclear reactors

    International Nuclear Information System (INIS)

    Schulten, R.; Kugeler, K.; Kugeler, M.; Petersen, K.; Decken, C.B. von der.

    1983-01-01

    This construction of a container, which is pressure-relieved by axial-central tensioning cables or reinforcing cables distributed over the circumference, makes a reduction of the wall thickness for the floor and roof, which was previously 2.5 metres by about 40% possible, and thus reduce manufacturing and cost problems. This is achieved by an appreciable increase of the prestressing exerted by the tensioning cables as this is taken up, not by the elasticity of the roof and floor, but instead by an intermediate part of pressure-resisting material. Such a container consists of a vertical cylindrical jacket of, for example, 20 metres diameter and 18 metres height, of a roof and floor of, for example, 1.50 metres thickness each and the intermediate part, which keeps the spacing of floor and roof as a central piece. This intermediate part which is taken through seals through the container can be imagined as a double tube of outside tube diameter of, for example, 4 metres and inside tube diameter of 2 metres with both tubes having thick walls. 4 tensioning cables displaced vertically by 900 run in the cylindrical annulus between the outer and inner tubes which are brought to the required pretension, e.g. 80,000 tonnes by nuts situated on the outside. The inner tube projects through the floor and roof. Its openings act as manholes and for the introduction of pipelines. These can, for example, carry a cooling medium for a reactor core via further ducts into the inside of the container. Container wall, floor and roof and the intermediate part in the form of a double tube are made up of cast steel segments or sectors in several layers. (RW)

  5. 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

  6. 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

  7. Reactor pressure vessel. Status report

    International Nuclear Information System (INIS)

    Elliot, B.J.; Hackett, E.M.; Lee, A.D.

    1996-10-01

    This report describes the issues raised as a result of the staffs review of Generic Letter (GL) 92-01, Revision 1, responses and plant-specific reactor pressure vessel (RPV) assessments and the actions taken or work in progress to address these issues. In addition, the report describes actions taken by the staff and the nuclear industry to develop a thermal annealing process for use at U.S. commercial nuclear power plants. This process is intended to be used as a means of mitigating the effects of neutron radiation on the fracture toughness of RPV materials. The Nuclear Regulatory Commission (NRC) issued GL 92-01, Revision 1, Supplement 1, to obtain information needed to assess compliance with regulatory requirements and licensee commitments regarding RPV integrity. GL 92-01, Revision 1, Supplement 1, was issued as a result of generic issues that were raised in the NRC staff's reviews of licensee responses to GL 92-01, Revision 1, and plant-specific RPV evaluations. In particular, an integrated review of all data submitted in response to GL 92-01, Revision 1, indicated that licensees may not have considered all relevant data in their RPV assessments. This report is representative of submittals to and evaluations by the staff as of September 30, 1996. An update of this report will be issued at a later date

  8. 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

  9. Functional systems of a pressurized water reactor

    International Nuclear Information System (INIS)

    Heinzel, V.

    1982-01-01

    The main topics, discussed in the present paper, are: - Principle design of the reactor coolant system - reactor pressure vessel with internals - containment design - residual heat removal and emergency cooling systems - nuclear component cooling systems - emergency feed water systems - plant electric power supply system. (orig./RW)

  10. Results of reactor pressure vessels ISI

    International Nuclear Information System (INIS)

    Cepcek, S.

    1994-01-01

    To find out the possible influence of the annealing process to reactor pressure vessel integrity, a large in-service inspection programme has been implemented as an associated activity to reactor pressure vessel annealing. In this paper the approach to the RPV in-service inspection is shown. Also, the main results and conclusions following in-service inspection are presented. (author). 3 refs, 1 fig

  11. 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)

  12. 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)

  13. Physics of pressurized water reactors

    International Nuclear Information System (INIS)

    Gruen, A.

    1980-01-01

    The objective of this lecture is to demonstrate typical problems and solutions encountered in the design and operation of PWR power plants. The examples selected for illustration refer to PWR's of KWU design and to results of KWU design methods. In order to understand the physics of a power reactor it is necessary to have some knowledge of the structure and design of the power plant system of which the reactor is a part. It is therefore assumed that the reader is familiar with the design of the more important components and systems of a PWR, such as fuel assemblies, control assemblies, core lay-out, reactor coolant system, instrumentation. (author)

  14. Supercritical-pressure light water cooled reactors

    CERN Document Server

    Oka, Yoshiaki

    2014-01-01

    This book focuses on the latest reactor concepts, single pass core and experimental findings in thermal hydraulics, materials, corrosion, and water chemistry. It highlights research on supercritical-pressure light water cooled reactors (SCWRs), one of the Generation IV reactors that are studied around the world. This book includes cladding material development and experimental findings on heat transfer, corrosion and water chemistry. The work presented here will help readers to understand the fundamental elements of reactor design and analysis methods, thermal hydraulics, materials and water

  15. 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

  16. Irradiation, Annealing, and Reirradiation Effects on American and Russian Reactor Pressure Vessel Steels

    International Nuclear Information System (INIS)

    Chernobaeva, A.A.; Korolev, Y.N.; Nanstad, R.K.; Nikolaev, Y.A.; Sokolov, M.A.

    1998-01-01

    One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPVs) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. Even though a postirradiation anneal may be deemed successful, a critical aspect of continued RPV operation is the rate of embrittlement upon reirradiation. There are insufficient data available to allow for verification of available models of reirradiation embrittlement or for the development of a reliable predictive methodology. This is especially true in the case of fracture toughness data. Under the U.S.-Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS), Working Group 3 on Radiation Embrittlement, Structural Integrity, and Life Extension of Reactor Vessels and Supports agreed to conduct a comparative study of annealing and reirradiation effects on RPV steels. The Working Group agreed that each side would irradiate, anneal, reirradiate (if feasible ), and test two materials of the other. Charpy V-notch (CVN) and tensile specimens were included. Oak Ridge National Laboratory (ORNL) conducted such a program (irradiation and annealing, including static fracture toughness) with two weld metals representative of VVER-440 and VVER-1000 RPVs, while the Russian Research Center-Kurchatov Institute (RRC-KI) conducted a program (irradiation, annealing, reirradiation, and reannealing) with Heavy-Section Steel Technology (HSST) Program Plate 02 and Heavy-Section Steel Irradiation (HSSI) Program Weld 73W. The results for each material from each laboratory are compared with those from the other laboratory. The ORNL experiments with the VVER welds included irradiation to about 1 x 10 19 n/cm 2 (>1 MeV), while the RRC-KI experiments with the U.S. materials included irradiations from about 2 to 18 x 10 19 n/cm 2 (>l MeV). In both cases, irradiations were conducted at ∼290 C and annealing treatments were conducted at ∼454 C. The ORNL and RRC

  17. Pressurized water reactors: the EPR project

    International Nuclear Information System (INIS)

    Py, J.P.; Yvon, M.

    2007-01-01

    EPR (originally 'European pressurized water reactor', and now 'evolutionary power reactor') is a model of reactor initially jointly developed by French and German engineers which fulfills the particular safety specifications of both countries but also the European utility requirements jointly elaborated by the main European power companies under the initiative of Electricite de France (EdF). Today, two EPR-based reactors are under development: one is under construction in Finland and the other, Flamanville 3 (France), received its creation permit decree on April 10, 2007. This article presents, first, the main objectives of the EPR, and then, describes the Flamanville 3 reactor: reactor type and general conditions, core and conditions of operation, primary and secondary circuits with their components, main auxiliary and recovery systems, man-machine interface and instrumentation and control system, confinement and serious accidents, arrangement of buildings. (J.S.)

  18. Method of repairing pressure tube type reactors

    International Nuclear Information System (INIS)

    Asada, Takashi.

    1983-01-01

    Purpose: To enable to re-start the reactor operation in a short time, upon occurrence of failures in a pressure tube, as well as directly examine the cause for the failures in the pressure tube. Method: The pressure tube reactor main body comprises a calandria tank of a briquette form, pressure tubes, fuel assemblies and an iron-water shielding body. If failure is resulted to a pressure tube, the reactor operation is at first shutdown and nuclear fuel assemblies are extracted to withdraw from the pressure tube. Then, to an inlet pipe way and an outlet pipeway connected to the failed pressure tube, are attached plugs by means of welding or the like at the appropriate position where the radiation exposure dose is lower and the repairing work can be performed with ease. The pressure tube is disconnected to withdraw from the inlet pipeway and the outlet pipeway and, instead, radiation shielding plug tube is inserted and shield cooling device is actuated if required, wherein the reactor is actuated to re-start the operation. (Yoshino, Y.)

  19. Pressurized water reactor fuel rod design methodology

    International Nuclear Information System (INIS)

    Silva, A.T.; Esteves, A.M.

    1988-08-01

    The fuel performance program FRAPCON-1 and the structural finite element program SAP-IV are applied in a pressurized water reactor fuel rod design methodology. The applied calculation procedure allows to dimension the fuel rod components and characterize its internal pressure. (author) [pt

  20. How to replace a reactor pressure vessel

    International Nuclear Information System (INIS)

    Huber, R.

    1996-01-01

    A potential life extending procedure for a nuclear reactor after, say, 40 years of service life, might in some circumstances be the replacement of the reactor pressure vessel. Neutron induced degradation of the vessel might make replacement by one of a different material composition desirable, for example. Although the replacement of heavy components, such as steam generators, has been possible for many years, the pressure vessel presents a much more demanding task if only because it is highly irradiated. Some preliminary feasibility studies by Siemens are reported for the two removal strategies that might be considered. These are removal of the entire pressure vessel in one piece and dismantling it into sections. (UK)

  1. Reactor Safety Commission Code of Practice for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    1990-01-01

    The Reactor Safety Commission of the Federal German Republic has summarized in the form of Official Guidelines the safety requirements which, in the Commission's view, have to be met in the design, construction and operation of a nuclear power station equipped with a pressurized water reactor. The Third Edition of the RSK Guidelines for pressurized water reactors dated 14.10.81. is a revised and expanded version of the Second Edition dated 24.1.79. The Reactor Safety Commission will with effect from October 1981 use these Guidelines in consultations on the siting of and safety concept for the installation approval of future pressurized water reactors and will assess these nuclear power stations during their erection in the light of these Guidelines. They have not however been immediately conceived for the adaptation of existing nuclear power stations, whether under construction or in operation. The scope of application of these Guidelines to such nuclear power stations will have to be examined for each individual case. The main aim of the Guidelines is to simplify the consultation process within the reactor Safety Commission and to provide early advice on the safety requirements considered necessary by the Commission. (author)

  2. Pressurized water reactor simulator. Workshop material

    International Nuclear Information System (INIS)

    2003-01-01

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21 'WWER-1000 Reactor Simulator' (2002). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23 'Boiling Water Reactor Simulator' (2003). This report consists of course material for workshops using a pressurized water reactor simulator

  3. 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

  4. 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.

  5. Towards EPR (European pressurized reactor)

    International Nuclear Information System (INIS)

    Anon.

    2003-01-01

    According to the French industry minister, it is nonsense continuing delaying the construction of an EPR prototype because France needs it in order to renew timely its park of nuclear reactors. The renewing is expected to begin in 2020 and will be assured with third generation reactors like EPR. A quick launching of the EPR prototype is necessary to have it being in service by 2012, the feedback operating experience that will be accumulated over the 8 years that will follow will be necessary to optimize the industrial version and to have it ready by 2020. The EPR reactor has indisputable assets: modern, safer, more competitive and it will produce less wastes than present nuclear reactors. The construction cost of an EPR prototype is estimated to 3 milliard Euros and the nuclear industry operators propose to finance it completely. The EPR prototype does not jeopardize the ambitious French program about renewable energy sources, France is committed to produce 21% of its electricity from renewable energies by 2010 and 10 milliard Euros will be invested over this period on wind energy. Nuclear energy and alternative energies must be considered as 2 aspects of a diversified energy policy. (A.C.)

  6. 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)

  7. Pressure releasing device for reactor container

    International Nuclear Information System (INIS)

    Takeda, Mika.

    1994-01-01

    In the present invention, dose rate to public caused by radioactive rare gases can be decreased. That is, a reactor container contains a reactor pressure vessel incorporating a reactor core. There are disposed a pressure releasing system for releasing the pressure in the reactor pressure vessel to the outside, and a burning device for burning gases released from the pressure releasing system. An exhaustion pipe is disposed to the pressure releasing system. A burning device is disposed to the exhaustion pipe. It is effective to dispose a ventilation port at a portion of the exhaustion pipe upstream of the burning device. In addition, the burning device may preferably be disposed in a multi-stage in the axial direction of the exhaustion pipe. With such procedures, hydrogen in gases discharged along with the release of the pressure in the container is burned. Buoyancy is caused to the exhaustion gases by heat energy upon burning. Since the exhaustion gases can reach a higher level by the buoyancy, the dose rate due to the rare gases can be reduced. (I.S.)

  8. N Reactor pressure tube 1350 postirradiation examination

    International Nuclear Information System (INIS)

    Cook, D.J.

    1977-01-01

    The N Reactor pressure tubes were fabricated from Zircaloy-2 primarily due to the excellent corrosion resistance, low neutron absorption, and high strength properties of this alloy. Irradiation damage mechanisms increase the strength and decrease the ductility of the Zircaloy-2. Irradiation data available at the time the tubes were installed indicated that fast neutron irradiation damage mechanisms would not decrease the ductility to unacceptable levels over the estimated plant life of 25 to 30 years. However, because the tubes are a primary coolant system component and only limited data are available on irradiation effects at high fluences, a Postirradiation Examination (PIE) program was developed to assure that service factors do not compromise pressure tube integrity essential to reactor safety. The PIE program requires that a pressure tube be periodically removed from the reactor for destructive testing. The N Reactor Technical Specifications specify that the frequency of pressure tube removal and examination be based upon the previous PIE test results. Four pressure tubes were examined before tube 1350, and the test results were summarized in individual reports. PIE results on tube 1350 were summarized along with the test results on the previous four tubes in a previous report. The purpose of this report is to present in detail the results on PIE of pressure tube 1350, and, in particular, document the technique by which the fracture toughness of the pressure tube was determined

  9. 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)

  10. N Reactor pressure tube 2566 postirradiation examination

    International Nuclear Information System (INIS)

    Scott, K.V.

    1978-01-01

    Pressure tube 2566 was removed from N Reactor in July, 1977 to initiate the postirradiation examination program required by the Technical Specifications. Destructive examination of the pressure tube, after a maximum accumulated fluence of 4.6 x 10 21 n/cm 2 (E > 1 MeV), was conducted at the Hanford Engineering Development Laboratory to determine the effects of reactor service on the mechanical properties and hydrogen absorption and corrosion characteristics of the pressure tube. Tube 2566 is the sixth tube removed for destructive examination since the initial reactor startup. Evaluation of test results reveal that no significant detrimental changes have occurred in the parameters studied, since the last tube was removed in 1974

  11. Proof, interpretation and evaluation of radiation-induced microstructural changes in WWER reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Boehmert, J.; Gokhman, A.; Grosse, M.; Ulbricht, A.

    2003-06-01

    Neutron embrittlement is a special issue for the VVER-type reactors. One of the fundamentals for a reliable assessment of the current material state is knowledge of the causes and mechanisms of neutron embrittlement. The aim of the project is to understand and to quantify the microstructural appearances due to neutron radiation in VVER-type reactor pressure vessel steels. The material base is a broad variation of irradiation probes taken from the irradiation programme Rheinsberg, surveillance programmes of Russian, Ukrainian or Hungarian NPPs or irradiation experiments with mockup-alloys. The microstructure was investigated by different methods. The small angle neutron scattering (SANS) proved to be the most suitable method. A procedure was developed to determine mean diameter, size distribution and volume fraction of irradiation-induced microstructure from SANS experiments in a reliable and comparable manner. With this method microstructural parameters were systematically determined and the main factors of influence were identified. Apart from the neutron fluence the volume fraction of radiation defects mainly changes with the copper or nickel content whereas phosphorus is hardly relevant. Annealing remedies the radiation-induced microstructural appearances. The ratio between nuclear and magnetic neutron scattering provides information on the type of radiation defects. This leads to the conclusion that the material composition changes the radiation defects. The change occurs gradually rather than abruptly. The radiation defects detected by SANS correlate with the radiation hardening and embrittlement. Generally, the results suggest a bimodal mechanism due to radiation-enhanced and radiation-induced defect evolution. A kinetic model on base of the rate theory approach was established. (orig.)

  12. Thermohydraulic analysis of pressurized water reactors

    International Nuclear Information System (INIS)

    Veloso, M.A.

    1980-01-01

    The computer program PANTERA is applied in the thermo-hydraulic analysis of Pressurized Water Reactor Cores (PWR). It is a version of COBRA-IIIC in which a new thermal conduction model for fuel rods was introduced. The results calculated by this program are compared with experimental data obtained from bundles of fuel rods, simulating reactor conditions. The validity of the new thermal model is checked too. The PANTERA code, through a simplified procedure of calculation, is used in the thermo-hydraulic analysis of Indian Point, Unit 2, reactor core, in stationary conditions. The results are discussed and compared with design data. (Autor) [pt

  13. 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

  14. Pressure test method for reactor pressure vessel in construction field

    International Nuclear Information System (INIS)

    Takeda, Masakado; Ushiroda, Koichi; Miyahara, Ryohei; Takano, Hiroshi; Matsuura, Tadashi; Sato, Keiya.

    1998-01-01

    Plant constitutional parts as targets of both of a primary pressure test and a secondary pressure test are disposed in communication with a reactor pressure vessel, and a pressure of the primary pressure test is applied to the targets of both tests, so that the primary pressure test and the second pressure test are conducted together. Since the number of pressure tests can be reduced to promote construction, and the number of workers can also be reduced. A pressure exceeding the maximum pressure upon use is applied to the pressure vessel after disposing the incore structures, to continuously conduct the primary pressure test and the secondary pressure test joined together and an incore flowing test while closing the upper lid of the pressure vessel as it is in the construction field. The number of opening/closing of the upper lid upon conducting every test can be reduced, and since the pressure resistance test is conducted after arranging circumference conditions for the incore flowing test, the tests can be conducted collectively also in view of time. (N.H.)

  15. Integrity of Magnox reactor steel pressure vessels

    International Nuclear Information System (INIS)

    Flewitt, P.E.J.; Williams, G.H.; Wright, M.B.

    1992-01-01

    The background to the safety assessment of the steel reactor pressure vessels for Magnox power stations is reviewed. The evolved philosophy adopted for the 1991 safety cases prepared for the continued operation of four Magnox power stations operated by Nuclear Electric plc is described, together with different aspects of the multi-legged integrity argument. The main revisions to the materials mechanical property data are addressed together with the assessment methodology adopted and their implications for the overall integrity argument formulated for the continued safe operation of these reactor pressure vessels. (author)

  16. Pressure tube type research reactor

    International Nuclear Information System (INIS)

    Ueda, Hiroshi.

    1976-01-01

    Object: To prevent excessive heat generation due to radiation of a pressure tube vessel. Structure: A pressure tube encasing therein a core comprises a dual construction comprising inner and outer tubes coaxially disposed. High speed cooling water is passed through the inner tube for cooling. In addition, in the outer periphery of said outer tube there is provided a forced cooling tube disposed coaxially thereto, into which cooling fluid, for example, such as moderator or reflector is forcibly passed. This forced cooling tube has its outer periphery surrounded by the vessel into which moderator or reflector is fed. By the provision of the dual construction of the pressure tube and the forced cooling tube, the vessel may be prevented from heat generation. (Ikeda, J.)

  17. Design and realization experience of Advanced Control Rod Group and Individual Control System (GIC) for VVER-1000 reactors

    International Nuclear Information System (INIS)

    Cerny, V.; Novy, L.; Janour, J.; Ris, M.; Zidek, P.

    1997-01-01

    During the reactor refueling outage of unit 1 of the South Ukrainian nuclear power plant in mid-1996, full replacement of the reactor's group and individual control (GIC) system was performed. The main functions of the GIC system are briefly characterized. The structure of the advanced GIC system is described and shown by means of a diagram. The criteria used in deciding on the upgrading strategy are discussed in some detail. The implementation of the replacement is also dealt with, as is the testing and commissioning of the system. (A.K.)

  18. Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part I: BWR/NWC conditions

    International Nuclear Information System (INIS)

    Ritter, S.; Seifert, H.P.; Devrient, B.; Roth, A.; Ehrnsten, U.; Ernestova, M.; Zamboch, M.; Foehl, J.; Weissenberg, T.; Gomez-Briceno, D.; Lapena, J.

    2004-01-01

    One of the ageing phenomena of pressure boundary components of light water reactors (LWR) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It was focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of EAC crack growth behaviour/mechanism of LAS in high-temperature water under steady-state power operation (constant load) and transient operating conditions (e.g., start-up/shut-down, transients in water chemistry and load). Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurised water reactor (VVER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (VVER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarises the most important crack growth results obtained under simulated BWR/NWC conditions. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

  19. Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part I: BWR/NWC conditions

    Energy Technology Data Exchange (ETDEWEB)

    Ritter, S.; Seifert, H.P. [Paul Scherrer Institute, PSI, Villigen (Switzerland); Devrient, B.; Roth, A. [Framatome ANP GmbH, Erlangen (Germany); Ehrnsten, U. [VTT Industrial Systems, Espoo (Finland); Ernestova, M.; Zamboch, M. [Nuclear Research Institute, NRI, Rez (Czech Republic); Foehl, J.; Weissenberg, T. [Staatliche Materialpruefungsanstalt, MPA, Stuttgart (Germany); Gomez-Briceno, D.; Lapena, J. [Centro de Investigaciones Energeticas Medioambientales y Tecnologicas, CIEMAT, Madrid (Spain)

    2004-07-01

    One of the ageing phenomena of pressure boundary components of light water reactors (LWR) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It was focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of EAC crack growth behaviour/mechanism of LAS in high-temperature water under steady-state power operation (constant load) and transient operating conditions (e.g., start-up/shut-down, transients in water chemistry and load). Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurised water reactor (VVER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (VVER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarises the most important crack growth results obtained under simulated BWR/NWC conditions. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

  20. 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.

  1. Pressure suppression facility for reactor container

    International Nuclear Information System (INIS)

    Fujii, Tadashi; Fukui, Toru; Kataoka, Yoshiyuki; Tominaga, Kenji.

    1993-01-01

    In a nuclear reactor comprising heat transfer surfaces from a pressure suppression pool at the inside to the outer circumferential pool at the outside, a means for supplying water from a water supply source at the outside of the container to the pools is disposed. Then, a heat transfer means is disposed between the pressure suppression chamber and the water cooling pool. The water supply means comprises a pressurization means for applying pressure to water of the water supply source and a water supply channel. Water is supplied into the pressure suppression pool and the outer circumferential pool to elevate the water level and extend the region of heat contact with the water cooling heat transfer means. In addition, since dynamic pressure is applied to the feedwater, for example, by pressurizing the water surface of the water supply source, water can be supplied without using dynamic equipments such as pumps. Then, since water-cooling heat transfer surface can be extended after occurrence of accident, enlargement of a reactor container and worsening of earthquake proofness can be avoided as much as possible, to improve function for suppressing the pressure in the container. Further, since water-cooling heat transfer region can be extended, the arrangement of the water source and the place to which water is supplied is made optional without considering the relative height therebetween, to improve earthquake proofness. (N.H.)

  2. Nuclear reactor of pressurized liquid coolant type

    International Nuclear Information System (INIS)

    Costes, D.

    1976-01-01

    The reactor comprises a vertical concrete pressure vessel, a bell-housing having an open lower end and disposed coaxially with the interior of the pressure vessel so as to delimit therewith a space filled with gas under pressure for the thermal insulation of the internal vessel wall, a pressurizing device for putting the coolant under pressure within the bell-housing and comprising a volume of control gas in contact with a large free surface of coolant in order that an appreciable variation in volume of liquid displaced within the coolant circuit inside the bell-housing should correspond to a small variation in pressure of the control gas. 9 claims, 3 drawing figures

  3. Construction management of Indian pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Bohra, S.A.; Sharma, P.D.

    2006-01-01

    Pandit Jawaharlal Nehru and Dr. Homi J. Bhabha, the visionary architects of Science and Technology of modern India foresaw the imperative need to establish a firm base for indigenous research and development in the field of nuclear electricity generation. The initial phase has primarily focused on the technology development in a systematic and structured manner, which has resulted in establishment of strong engineering, manufacturing and construction base. The nuclear power program started with the setting up of two units of boiling light water type reactors in 1969 for speedy establishment of nuclear technology, safety culture, and development of operation and maintenance manpower. The main aim at that stage was to demonstrate (to ourselves, and indeed to the rest of the world) that India, inspite of being a developing country, with limited industrial infrastructure and low capacity power grids, could successfully assimilate the high technology involved in the safe and economical operation of nuclear power reactors. The selection of a BWR was in contrast to the pressurized heavy water reactors (PHWR), which was identified as the flagship for the first stage of India's nuclear power program. The long-term program in three stages utilizes large reserves of thorium in the monazite sands of Kerala beaches in the third stage with first stage comprising of series of PHWR type plants with a base of 10,000 MW. India has at present 14 reactors in operation 12 of these being of PHWR type. The performance of operating units of 2720 MW has improved significantly with an overall capacity factor of about 90% in recent times. The construction work on eight reactor units with installed capacity of 3960 MW (two PHWRs of 540 MW each, four PHWRs of 220 MW each and two VVERs of 1000 MW each) is proceeding on a rapid pace with project schedules of less than 5 years from first pour of concrete. This is being achieved through advanced construction technology and management. Present

  4. Pressure Tube and Pressure Vessel Reactors; certain comparisons

    Energy Technology Data Exchange (ETDEWEB)

    Margen, P H; Ahlstroem, P E; Pershagen, B

    1961-04-15

    In a comparison between pressure tube and pressure vessel type reactors for pressurized D{sub 2}O coolant and natural uranium, one can say that reactors of these two types having the same net electrical output, overall thermal efficiency, reflected core volume and fuel lattice have roughly the same capital cost. In these circumstances, the fuel burn-up obtainable has a significant influence on the relative economics. Comparisons of burn-up values made on this basis are presented in this report and the influence on the results of certain design assumptions are discussed. One of the comparisons included is based on the dimensions and ratings proposed for CANDU. Moderator temperature coefficients are compared and differences in kinetic behaviour which generally result in different design philosophies for the two types are mentioned, A comparison of different methods of obtaining flux flattening is presented. The influence of slight enrichment and other coolants, (boiling D{sub 2}O and gases) on the comparison between pressure tube and pressure vessel designs is discussed and illustrated with comparative designs for 400 MW electrical output. This paper was presented at the EAES Enlarged Symposium on Heterogeneous Heavy Water Power Reactors, Mallorca, October 10 - 14, 1960.

  5. Pressure Tube and Pressure Vessel Reactors; certain comparisons

    International Nuclear Information System (INIS)

    Margen, P.H.; Ahlstroem, P.E.; Pershagen, B.

    1961-04-01

    In a comparison between pressure tube and pressure vessel type reactors for pressurized D 2 O coolant and natural uranium, one can say that reactors of these two types having the same net electrical output, overall thermal efficiency, reflected core volume and fuel lattice have roughly the same capital cost. In these circumstances, the fuel burn-up obtainable has a significant influence on the relative economics. Comparisons of burn-up values made on this basis are presented in this report and the influence on the results of certain design assumptions are discussed. One of the comparisons included is based on the dimensions and ratings proposed for CANDU. Moderator temperature coefficients are compared and differences in kinetic behaviour which generally result in different design philosophies for the two types are mentioned, A comparison of different methods of obtaining flux flattening is presented. The influence of slight enrichment and other coolants, (boiling D 2 O and gases) on the comparison between pressure tube and pressure vessel designs is discussed and illustrated with comparative designs for 400 MW electrical output. This paper was presented at the EAES Enlarged Symposium on Heterogeneous Heavy Water Power Reactors, Mallorca, October 10 - 14, 1960

  6. Calculation of neutron fluence in the region of the pressure vessel for the history of different reactors by using the Monte-Carlo-method

    International Nuclear Information System (INIS)

    Barz, H.U.; Bertram, W.

    1992-01-01

    Embrittlement of pressure vessel material caused by neutron irradiation is a very important problem for VVER-440 reactors. For the estimation of the fracture risk highly reliable neutron fluence values are necessary. For this reason a special theoretical determination of space dependent neutron fluences has been performed mainly on the basis of Monte-Carlo calculations. The described method allows the accurate calculation of neutron fluences near the pressure vessel in the height of the core region for all reactor histories and loading cycles in an efficient manner. To illustrate the accuracy of the suggested method a comparison with experimental results was done. The calculated neutron fluence values can be used for planning the loading schemes of each reactor according to the safety requirements against brittle fracture. (orig.)

  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. 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.)

  9. Leak detector for reactor pressure vessel

    International Nuclear Information System (INIS)

    Morimoto, Mikio.

    1991-01-01

    A branched pipe is disposed to a leak off pipeline led from a flange surface which connects the main body and the upper lid of a reactor pressure vessel. An exhaust pump is disposed to the branched pipe and a moisture gage is disposed on the side of the exhaustion and a dry air supplier is connected to the branched pipe. Upon conducting a pressure-proof leak test for the reactor pressure vessel, the exhaust pump is operated and an electromagnet valve disposed at the upstream of the dry air supplier is opened and closed repeatedly. The humidity of air sucked by the exhaust pump is detected by the moisture gage. If leaks should be caused in the joining surface of the flange, leaked water is diffused as steams. Accordingly, occurrence of leak can be detected instantly based on the comparison with the moisture level of the dry air as a standard. In this way, a leak test can be conducted reliably in a short period of time with no change of for the reactor pressure container itself. (I.N.)

  10. Dismantling id the reactor pressure vessel insulation and dissecting of the MZFR reactor pressure vessel

    International Nuclear Information System (INIS)

    Loeb, Andreas; Stanke, Dieter; Thoma, Markus; Eisenmann, Beata; Prechtl, Erwin; Dehnke, Burckhard

    2008-01-01

    The MZFR reactor was decommissioned in 1984. The authors describe the dismantling of the reactor pressure vessel insulation that consists of asbestos containing mineral fiber wool. The appropriate remote handling and cutting tools had to be adapted with respect to the restrained space in the containment. The dismantling of the reactor pressure vessel has been completed, the dissected parts have been packaged into 200 containers for the final repository Konrad. During the total project time no reportable events and no damage to persons occurred.

  11. EPR (European Pressurized water Reactor) The advanced nuclear reactor

    International Nuclear Information System (INIS)

    2004-01-01

    Nuclear energy, which provides a steady supply of electricity at low cost, has its rightful place in the energy mix of the 21. century, which puts the emphasis on sustainable development. The EPR is the only 3. generation reactor under construction today. It is an evolutionary reactor that represents a new generation of pressurized water reactors with no break in the technology used for the most recent models. The EPR was developed by Framatome and Siemens, whose nuclear activities were combined in January 2001 to form Framatome ANP, a subsidiary of AREVA and Siemens. EDF and the major German electricity companies played an active part in the project. The safety authorities of the two countries joined forces to bring their respective safety standards into line and draw up joint design rules for the new reactor. The project had three objectives: meet the requirements of European utilities, comply with the safety standards laid down by the French safety authority for future pressurized water reactors, in concert with its German counterpart, and make nuclear energy even more competitive than energy generated using fossil fuels. The EPR can guarantee a safe, inexpensive electricity supply, without adding to the greenhouse effect. It meets the requirements of the safety authorities and lives up to the expectations of electricity utilities. This document presents the main characteristics of the EPR, and in particular the additional measures to prevent the occurrence of events likely to damage the core, the leak-tight containment, the measures to reduce the exposure of operating and maintenance personnel, the solutions for an even greater protection of the environment. The foreseen development of the EPR in France and abroad (Finland, China, the United States) is summarized

  12. Tritium issues in commercial pressurized water reactors

    International Nuclear Information System (INIS)

    Jones, G.

    2008-01-01

    Tritium has become an important radionuclide in commercial Pressurized Water Reactors because of its mobility and tendency to concentrate in plant systems as tritiated water during the recycling of reactor coolant. Small quantities of tritium are released in routine regulated effluents as liquid water and as water vapor. Tritium has become a focus of attention at commercial nuclear power plants in recent years due to inadvertent, low-level, chronic releases arising from routine maintenance operations and from component failures. Tritium has been observed in groundwater in the vicinity of stations. The nuclear industry has undertaken strong proactive corrective measures to prevent recurrence, and continues to eliminate emission sources through its singular focus on public safety and environmental stewardship. This paper will discuss: production mechanisms for tritium, transport mechanisms from the reactor through plant, systems to the environment, examples of routine effluent releases, offsite doses, basic groundwater transport and geological issues, and recent nuclear industry environmental and legal ramifications. (authors)

  13. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1980-01-01

    This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the 13 N content in the containment atmosphere. 13 N is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/ 13 N+ 4 He. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium 13 N concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

  14. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1980-01-01

    This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/Nl3+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

  15. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1979-08-01

    The present paper deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process H1+016 → N13+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m -3 and 7 kBq m -3 for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge (Li) flow detector assembly operated at elevated pressure. (Auth.)

  16. The safety of pressurized water reactors

    International Nuclear Information System (INIS)

    Panossian, J.; Tanguy, P.

    1991-01-01

    In this paper we present a review of the status of the safety level of modern pressurized water reactors, that is to say those that meet the safety criteria accepted today by the international nuclear community. We will mainly rely on the operating experience and the Probabilistic Safety Assessments concerning French reactors. We will not back over the basic safety concepts of these reactors, which are well known. We begin with a brief review of some of the lessons learned from the two main accidents discussed in the present meeting. Three Mile Island and Chernobyl, without entering into details presented in previous papers. The presentation ends with a rather lengthy conclusion, aimed more at those not directly involved in the technical details of nuclear safety matters

  17. Reactor Pressure Vessel (RPV) Acquisition Strategy

    Energy Technology Data Exchange (ETDEWEB)

    Mizia, Ronald Eugene [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2008-04-01

    The Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed reactor and use low-enriched uranium, TRISO-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development (R&D) Program is responsible for performing R&D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while at the same time setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. The purpose of this report is to address the acquisition strategy for the NGNP Reactor Pressure Vessel (RPV). This component will be larger than any nuclear reactor pressure vessel presently in service in the United States. The RPV will be taller, larger in diameter, thicker walled, heavier and most likely fabricated at the Idaho National Laboratory (INL) site of multiple subcomponent pieces. The pressure vessel steel can either be a conventional materials already used in the nuclear industry such as listed within ASME A508/A533 specifications or it will be fabricated from newer pressure vessel materials never before used for a nuclear reactor in the US. Each of these characteristics will present a

  18. 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)

  19. Coolant mixing in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, T; Grunwald, G

    1998-10-01

    The behavior of PWRs during cold water or boron dilution transients is strongly influenced by the distribution of coolant temperature and boron concentration at the core inlet. This distribution is the needed input to 3-dimensional neutron kinetics to calculate the power distribution in the core. It mainly depends on how the plugs of cold or unborated water formed in a single loop are mixed in the downcomer and in the lower plenum. To simulate such mixture phenomena requires the application of 3-dimensional CFD (computational fluid dynamics) codes. The results of the simulation have to be validated against mixture experiments at scaled facilities. Therefore, in the framework of a research project funded by BMBF, the institute creates a 1:5 mixture facility representing first the geometry of a German pressurized water reactor and later the European Pressurized Water Reactor (EPR) geometry. The calculations are based on the CFD Code CFX-4. (orig.)

  20. Pumps for German pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dernedde, R.

    1984-01-01

    The article describes the development of a selection of pumps which are used in the primary coolant system and the high-pressure safety injection system and feed water system during the past 2 decades. The modifications were caused by the step-wise increasing power output of the plants from 300 MW up to 1300 MW. Additional important influences were given be the increased requirements for quality assurance and final-documentation. The good operating results of the delivered pumps proved that the reliability is independent of the volume of the software-package. The outlook expects that consolidation will be followed by additional steps for the order processing of components for the convoy pumps. KW: main coolant pump; primary system; boiler feed pump; reactor pump; secondary system; barrel insert pump; pressure water reactor; convoy pump; state of the art.

  1. Radiation effects on reactor pressure vessel supports

    International Nuclear Information System (INIS)

    Johnson, R.E.

    1996-05-01

    The purpose of this report is to present the findings from the work done in accordance with the Task Action Plan developed to resolve the Nuclear Regulatory Commission (NRC) Generic Safety Issue No. 15, (GSI-15). GSI-15 was established to evaluate the potential for low-temperature, low-flux-level neutron irradiation to embrittle reactor pressure vessel (RPV) supports to the point of compromising plant safety. An evaluation of surveillance samples from the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) had suggested that some materials used for RPV supports in pressurized-water reactors could exhibit higher than expected embrittlement rates. However, further tests designed to evaluate the applicability of the HFIR data to reactor RPV supports under operating conditions led to the conclusion that RPV supports could be evaluated using traditional method. It was found that the unique HFIR radiation environment allowed the gamma radiation to contribute significantly to the embrittlement. The shielding provided by the thick steel RPV shell ensures that degradation of RPV supports from gamma irradiation is improbable or minimal. The findings reported herein were used, in part, as the basis for technical resolution of the issue

  2. 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)

  3. Pressure vessel for a BWR type reactor

    International Nuclear Information System (INIS)

    Shimamoto, Yoshiharu.

    1980-01-01

    Purpose: To prevent the retention of low temperature water and also prevent the thermal fatigue of the pressure vessel by making large the curvature radius of a pressure vessel of a feed water sparger fitting portion and accelerating the mixing of low-temperature water at the feed water sparger base and in-pile hot water. Constitution: The curvature radius of the corner of the feed water sparger fitting portion in a pressure vessel is formed largely. In-pile circulating water infiltrates up to the base portion of the feed water sparger to carry outside low-temperature water at the base part, which is mixed with in-pile hot water. Accordingly, low temperature water does not stay at the base portion of the feed water sparger and generation of thermal fatigue in the pressure vessel can be prevented and the safety of the BWR type reactor can be improved. (Yoshino, Y.)

  4. Integral Pressurized Water Reactor Simulator Manual

    International Nuclear Information System (INIS)

    2017-01-01

    This publication provides detailed explanations of the theoretical concepts that the simulator users have to know to gain a comprehensive understanding of the physics and technology of integral pressurized water reactors. It provides explanations of each of the simulator screens and various controls that a user can monitor and modify. A complete description of all the simulator features is also provided. A detailed set of exercises is provided in the Exercise Handbook accompanying this publication.

  5. Analysis of nuclear reactor pressure vessel flanges

    International Nuclear Information System (INIS)

    Oliveira, C.A.N. de; Augusto, O.B.

    1985-01-01

    This work proposes a methodology for the structural analysis of high diameter nuclear reactor pressure vessel flanges. In the analysis the vessel is divided into shell-of-revolution elements, the flanges are represented by rigid rings, and the bolts are treated as beams. The flexibility method is used for solving the problem, and the results are compared with results obtained by the finite element method. (Author) [pt

  6. Pressure suppression system for a nuclear reactor

    International Nuclear Information System (INIS)

    Jost, N.

    1977-01-01

    The invention pertains to a pressure suppression system for PWR reactors where the parts enclosing the primary coolant are contained in two pressure-tight separate chambers. According to the invention, these chambers are partly filled with water and are connected with each other below the water surface. This way, gases cannot escape from the containment, not even if a valve and a line are damaged at the same time, as the vapours released condensate in the water of at least one of the other chambers. (HP) [de

  7. Head spray nozzle in reactor pressure vessel

    International Nuclear Information System (INIS)

    Hatano, Shun-ichi.

    1990-01-01

    In a reactor pressure vessel of a BWR type reactor, a head spray nozzle is used for cooling the head of the pressure vessel and, in view of the thermal stresses, it is desirable that cooling is applied as uniformly as possible. A conventional head spray is constituted by combining full cone type nozzles. Since the sprayed water is flown down upon water spraying and the sprayed water in the vertical direction is overlapped, the flow rate distribution has a high sharpness to form a shape as having a maximum value near the center and it is difficult to obtain a uniform flow rate distribution in the circumferential direction. Then, in the present invention, flat nozzles each having a spray water cross section of laterally long shape, having less sharpness in the circumferential distribution upon spraying water to the inner wall of the pressure vessel and having a wide angle of water spray are combined, to make the flow rate distribution of spray water uniform in the inner wall of the pressure vessel. Accordingly, the pressure vessel can be cooled uniformly and thermal stresses upon cooling can be decreased. (N.H.)

  8. 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

  9. 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.

  10. Pressure thermal shock analysis for nuclear reactor pressure vessel

    International Nuclear Information System (INIS)

    Galik, G.; Kutis, V.; Jakubec, J.; Paulech, J.; Murin, J.

    2015-01-01

    The appearance of structural weaknesses within the reactor pressure vessel or its structural failure caused by crack formation during pressure thermal shock processes pose as a severe environmental hazard. Coolant mixing during ECC cold water injection was simulated in a detailed CFD analysis. The temperature distribution acting on the pipe wall internal surface was calculated. Although, the results show the formation of high temperature differences and intense gradients, an additional structural analysis is required to determine the possibility of structural damage from PTS. Such an analysis will be the subject of follow-up research. (authors)

  11. 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)

  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. 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)

  14. 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)

  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. Status of advanced small pressurized water reactors

    International Nuclear Information System (INIS)

    Chen Peipei; Zhou Yun

    2012-01-01

    In order to expand the nuclear power in energy and desalination, increase competitiveness in global nuclear power market, many developed countries with strong nuclear energy technology have realized the importance of Small Modular Reactor (SMR) and initiated heavy R and D programs in SMR. The Advanced Small Pressurized Water Reactor (ASPWR) is characterized by great advantages in safety and economy and can be used in remote power grid and replace mid/small size fossil plant economically. This paper reviews the history and current status of SMR and ASPWR, and also discusses the design concept, safety features and other advantages of ASPWR. The purpose of this paper is to provide an overall review of ASPWR technology in western countries, and to promote the R and D in ASPWR in China. (authors)

  17. 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

  18. 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.

  19. Startup and commissioning of pressurized water reactors

    International Nuclear Information System (INIS)

    Albert, L.J.; Gilbert, C.F.

    1983-05-01

    A critical phase of plant development is the test, startup, and commissioning period. The effort expended prior to commissioning has a definite effect on the reliability and continuing availability of the plant during its life. This paper describes a test, startup, and commissioning program for a pressurized water reactor (PWR) plant. This program commences with the completion of construction and continues through the turnover of equipment/systems to the owner's startup/ commissioning group. The paper addresses the organization of the test/startup group, planning and scheduling, test procedures and initial testing, staffing and certification of the test group, training of operators, and turnover to the owner

  20. Reliability analysis of reactor pressure vessel intensity

    International Nuclear Information System (INIS)

    Zheng Liangang; Lu Yongbo

    2012-01-01

    This paper performs the reliability analysis of reactor pressure vessel (RPV) with ANSYS. The analysis method include direct Monte Carlo Simulation method, Latin Hypercube Sampling, central composite design and Box-Behnken Matrix design. The RPV integrity reliability under given input condition is proposed. The result shows that the effects on the RPV base material reliability are internal press, allowable basic stress and elasticity modulus of base material in descending order, and the effects on the bolt reliability are allowable basic stress of bolt material, preload of bolt and internal press in descending order. (authors)

  1. 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.

  2. Crack growth behaviour of low alloy steels for pressure boundary components under transient light water reactor operating conditions (CASTOC)

    International Nuclear Information System (INIS)

    Foehl, J.; Weissenberg, T.; Gomez-Briceno, D.; Lapena, J.; Ernestova, M.; Zamboch, M.; Seifert, H.P.; Ritter, S.; Roth, A.; Devrient, B.; Ehrnsten, U.

    2004-01-01

    The CASTOC project addresses environmentally assisted cracking (EAC) phenomena in low alloy steels used for pressure boundary components in both Western type boiling water reactors (BWR) and Russian type pressurised water reactors (VVER). It comprises the four work packages (WP): inter-laboratory comparison test (WP1); EAC behaviour under static load (WP2), EAC behaviour under cyclic load and load transients (WP3); evaluation of the results with regard to their relevance for components in practice (WP4). The use of sophisticated test facilities and measurement techniques for the on-line detection of crack advances have provided a more detailed understanding of the mechanisms of environmentally assisted cracking and provided quantitative data of crack growth rates as a function of loading events and time, respectively. The effect of several major parameters controlling EAC was investigated with particular emphasis on the transferability of the results to components in service. The obtained crack growth rate data were reflected on literature data and on commonly applied prediction curves as presented in the appropriate Code. At relevant stress intensity factors it could be shown that immediate cessation of growing cracks occurs after changing from cyclic to static load in high purity oxygenated BWR water and oxygen-free VVER water corresponding to steady state operation conditions. Susceptibility to environmentally assisted cracking under static load was observed for a heat affected zone material in oxygenated high purity water and also in base materials during a chloride transient representing BWR water condition below Action Level 1 of the EPRI Water Chemistry Guidelines according to the lectrical conductivity of the water but in the range of Action Level 2 according to the content of chlorides. Time based crack growth was also observed in one Russian type base material in oxygenated VVER water and in one Western type base material in oxygenated high purity BWR

  3. Seal analysis technology for reactor pressure vessel

    International Nuclear Information System (INIS)

    Zheng Liangang; Zhang Liping; Yang Yu; Zang Fenggang

    2009-01-01

    There is the coolant with radiation, high temperature and high pressure in the reactor pressure vessel (RPV). It is closely correlated to RPV sealing capability whether the whole nuclear system work well or not. The aim of this paper is to study the seal analysis method and technology, such as the pre-tensioning of the bolt, elastoplastic contact and coupled technology of thermal and structure. The 3 D elastoplastic seal analysis method really and generally consider the loads and model the contact problem with friction between the contact plates. This method is easier than the specialized seal program and used widely. And it is more really than the 2 D seal analysis method. This 3 D elastoplastic seal analysis method has been successfully used in the design and analysis of RPV. (authors)

  4. Basic conceptions for reactor pressure vessel manipulators and their evaluation

    International Nuclear Information System (INIS)

    Popp, P.

    1987-01-01

    The study deals with application fields and basic design conceptions of manipulators in reactor pressure vessels as well as their evaluation. It is shown that manipulators supported at the reactor flange have essential advantages

  5. Pressure suppression device for nuclear reactor building

    International Nuclear Information System (INIS)

    Ikegame, Noboru.

    1992-01-01

    In a nuclear reactor building, there are disposed cooling coils connected to an air supply duct at the outside of the building, an air supply blower, an air supply duct having the top end opened, an exhaustion duct having the top end opened and a bypassing pipeline interposed between the exhaustion duct and the air supply duct on the side of the inlet of the cooling coils. In the reactor building, when a radioactive material leakage accident should occur, an isolation valve is closed to isolate the building from the outside. Further, bypassing isolation valve is opened to form a closed cooling circuit by the cooling coils, the air supply blower and the air supply duct, the exhaustion duct and the bypassing pipeline in the reactor building. With such a constitution, since air as the atmosphere in the building is circulated through the closed cooling circuit and cooled by the cooling coils, the temperature is not elevated. Accordingly, since the pressure elevation of the atmosphere in the building is suppressed, the atmosphere containing radioactive materials do not flow out of the building. (I.N.)

  6. Nuclear reactor pressure vessel flaw distribution development

    International Nuclear Information System (INIS)

    Kennedy, E.L.; Foulds, J.R.; Basin, S.L.

    1991-12-01

    Previous attempts to develop flaw distributions for probabilistic fracture mechanics analyses of pressurized water reactor (PWR) vessels have aimed at the estimation of a ''generic'' distribution applicable to all PWR vessels. In contrast, this report describes (1) a new flaw distribution development analytic methodology that can be applied to the analysis of vessel-specific inservice inspection (ISI) data, and (2) results of the application of the methodology to the analysis of flaw data for each vessel case (ISI data on three PWR vessels and laboratory inspection data on sections of the Midland reactor vessel). Results of this study show significant variation among the flaw distributions derived from the various data sets analyzed, strongly suggesting than a vessel-specific flaw distribution (for vessel integrity prediction under pressurized thermal shock) is preferred over a ''generic'' distribution. In addition, quantitative inspection system flaw sizing accuracy requirements have been identified for developing a flaw distribution from vessel ISI data. The new flaw data analysis methodology also permits quantifying the reliability of the flaw distribution estimate. Included in the report are identified needs for further development of several aspects of ISI data acquisition and vessel integrity prediction practice

  7. Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

    International Nuclear Information System (INIS)

    Greneche, D.

    2014-01-01

    This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

  8. 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

  9. 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

  10. Performance of pressure tubes in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rodgers, D.; Griffiths, M.; Bickel, G.; Buyers, A.; Coleman, C.; Nordin, H.; St Lawrence, S. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    The pressure tubes in CANDU reactors typically operate for times up to about 30 years prior to refurbishment. The in-reactor performance of Zr-2.5Nb pressure tubes has been evaluated by sampling and periodic inspection. This paper describes the behavior and discusses the factors controlling the behaviour of these components. The Zr–2.5Nb pressure tubes are nominally extruded at 815{sup o}C, cold worked nominally 27%, and stress relieved at 400 {sup o}C for 24 hours, resulting in a structure consisting of elongated grains of hexagonal close-packed alpha-Zr, partially surrounded by a thin network of filaments of body-centred-cubic beta-Zr. These beta-Zr filaments are meta-stable and contain about 20% Nb after extrusion. The stress-relief treatment results in partial decomposition of the beta-Zr filaments with the formation of hexagonal close-packed alpha-phase particles that are low in Nb, surrounded by a Nb-enriched beta-Zr matrix. The material properties of pressure tubes are determined by variations in alpha-phase texture, alpha-phase grain structure, network dislocation density, beta-phase decomposition, and impurity concentration that are a function of manufacturing variables. The pressure tubes operate at temperatures between 250 {sup o}C and 310 {sup o}C with coolant pressures up to about 11 MPa in fast neutron fluxes up to 4 x 10{sup 17} n·m{sup -2}·s{sup -1} (E > 1 MeV) and the properties are modified by these conditions. The properties of the pressure tubes in an operating reactor are therefore a function of both manufacturing and operating condition variables. The ultimate tensile strength, fracture toughness, and delayed hydride-cracking properties (velocity (V) and threshold stress intensity factor (K{sub IH})) change with irradiation, but all reach a nearly limiting value at a fluence of less than 10{sup 25} n·m{sup -2} (E > 1 MeV). At this point the ultimate tensile strength is raised about 200 MPa, toughness is reduced by about 50%, V increases

  11. Methods and tools for the validation of neutron instrumentation; methods for the detection of loose VVER-1000 reactor internals. Technical report

    International Nuclear Information System (INIS)

    Stulik, P.; Sipek, B.; Pecinka, L.

    2004-12-01

    The following topics are addressed: (1) Development, tuning and laboratory testing of the proposed DMTS distributed system; (2) Testing of selected technological equipment and software within the technology of the Temelin NPP; (3) Proposal for basic performance testing of the temperature measurement dynamics on Temelin primary circuit loops; (4) Data for the design and manufacture of 2 measuring chains for the processing of operating signals from internal reactor detectors at the Dukovany-4 reactor unit using a modified experimental AMV set and the DMTS system being developed; (5) Trial measurement with the DMTS system; (6) Evaluation of the usability of signals from the ionization chambers of the innovated instrumentation and control system within the in-service diagnosis system of the Dukovany NPP using the DMTS system being developed; and (7) Calculation of acoustic frequencies of the Temelin primary circuit by means of electromechanical analogy for loop configurations including the effects of the pressurizer and idle coolant loops. (P.A.)

  12. 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

  13. 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.

  14. Pressurized-water-reactor station blackout

    International Nuclear Information System (INIS)

    Dobbe, C.A.

    1983-01-01

    The purpose of the Severe Accident Sequence Analysis (SASA) Program was to investigate accident scenarios beyond the design basis. The primary objective of SASA was to analyze nuclear plant transients that could lead to partial or total core melt and evaluate potential mitigating actions. The following summarizes the pressurized water reactor (PWR) SASA effort at the Idaho National Engineering Laboratory (INEL). The INEL is presently evaluating Unresolved Safety Issue A-44 - Station Blackout from initiation of the transient to core uncovery. The balance of the analysis from core uncovery until fission product release is being performed at Sandia National Laboratory (SNL). The current analyses involve the Bellefonte Nuclear Steam Supply System (NSSS), a Babcock and Wilcox (B and W) 205 Fuel Assembly (205-FA) raised loop design to be operated by the Tennessee Valley Authority

  15. Reactor pressure vessel stud management automation strategies

    International Nuclear Information System (INIS)

    Biach, W.L.; Hill, R.; Hung, K.

    1992-01-01

    The adoption of hydraulic tensioner technology as the standard for bolting and unbolting the reactor pressure vessel (RPV) head 35 yr ago represented an incredible commitment to new technology, but the existing technology was so primitive as to be clearly unacceptable. Today, a variety of approaches for improvement make the decision more difficult. Automation in existing installations must meet complex physical, logistic, and financial parameters while addressing the demands of reduced exposure, reduced critical path, and extended plant life. There are two generic approaches to providing automated RPV stud engagement and disengagement: the multiple stud tensioner and automated individual tools. A variation of the latter would include the handling system. Each has its benefits and liabilities

  16. Operator Support System for Pressurized Water Reactor

    International Nuclear Information System (INIS)

    Wei Renjie; Shen Shifei

    1996-01-01

    Operator Support System for Pressurized Water Reactor (OSSPWR) has been developed under the sponsorship of IAEA from August 1994. The project is being carried out by the Department of Engineering Physics, Tsinghua University, Beijing, China. The Design concepts of the operator support functions have been established. The prototype systems of OSSPWR has been developed as well. The primary goal of the project is to create an advanced operator support system by applying new technologies such as artificial intelligence (AI) techniques, advanced communication technologies, etc. Recently, the advanced man-machine interface for nuclear power plant operators has been developed. It is connected to the modern computer systems and utilizes new high performance graphic displays. (author). 6 refs, 4 figs

  17. The nuclear reactor strategy between fast breeder reactors and advanced pressurized water reactors

    International Nuclear Information System (INIS)

    Seifritz, W.

    1983-01-01

    A nuclear reactor strategy between fast breeder reactors (FBRs) and advanced pressurized water reactors (APWRs) is being studied. The principal idea of this strategy is that the discharged plutonium from light water reactors (LWRs) provides the inventories of the FBRs and the high-converter APWRs, whereby the LWRs are installed according to the derivative of a logistical S curve. Special emphasis is given to the dynamics of reaching an asymptotic symbiosis between FBRs and APWRs. The main conclusion is that if a symbiotic APWR-FBR family with an asymptotic total power level in the terawatt range is to exist in about half a century from now, we need a large number of FBRs already in an early phase

  18. Verification of computer code for calculation of coolant radiolysis in the VVER reactor core with regard for boiling in its upper part

    Energy Technology Data Exchange (ETDEWEB)

    Arkhipov, O.P.; Kabakchi, S.A. [OKB Gidropress, Podolsk, Moscow (Russian Federation)

    2010-07-01

    Code Bora for WWER coolant radiolysis calculation considering single jets boiling in the reactor core top part is developed on the basis of computer codes MOPABA-H2 (radiolysis of aqueous solutions) and SteamRad (radiolysis of vapor). Physico-chemical processes taking place in boiling core coolant are complex and diversified. Still, for the solution of certain problems their simulation can be simplified. The approach of reasonable simplification was used for development of code Bora: mathematical model assumed is purposed for simulation of phenomena only in the area of interest; the number of simulated chemical reactions and particles shall be reasonably minimum; complexity of interphase mass transfer calculation procedure shall be adequate to actually available accuracy of modeling. The analysis of new experimental initial yields of water radiolysis products data and kinetic parameters of elementary chemical reactions with their participation has been carried out. Some changes have been introduced in the mechanism of liquid water and aqueous solutions of ammonia radiolysis have been significantly revised on the basis of this analysis. Examples of the calculations provided for code Bora verification are presented. Despite of very simple simulation of interphase mass transfer, Bora allows to obtain average chemical composition of two-phase coolant at BWR core outlet with the accuracy sufficient for engineering calculations. The report also presents the results of two-phase coolant chemical composition test calculation for reactor core top part coolant boiling in pressurized water reactor. (author)

  19. Reactor pressure vessel structural integrity research

    International Nuclear Information System (INIS)

    Pennell, W.E.; Corwin, W.R.

    1994-01-01

    Development continues on the technology used to assess the safety of irradiation-embrittled nuclear reactor pressure vessels (RPVs) containing flaws. Fracture mechanics tests on RPV steel, coupled with detailed elastic-plastic finite-element analyses of the crack-tip stress fields, have shown that (1) constraint relaxation at the crack tip of shallow surface flaws results in increased data scatter but no increase in the lower-bound fracture toughness, (2) the nil ductility temperature (NDT) performs better than the reference temperature for nil ductility transition (RT NDT ) as a normalizing parameter for shallow-flaw fracture toughness data, (3) biaxial loading can reduce the shallow-flaw fracture toughness, (4) stress-based dual-parameter fracture toughness correlations cannot predict the effect of biaxial loading on shallow-flaw fracture toughness because in-plane stresses at the crack tip are not influenced by biaxial loading, and (5) an implicit strain-based dual-parameter fracture toughness correlation can predict the effect of biaxial loading on shallow-flaw fracture toughness. Experimental irradiation investigations have shown that (1) the irradiation-induced shift in Charpy V-notch vs temperature behavior may not be adequate to conservatively assess fracture toughness shifts due to embrittlement, and (2) the wide global variations of initial chemistry and fracture properties of a nominally uniform material within a pressure vessel may confound accurate integrity assessments that require baseline properties

  20. Prestressed concrete pressure vessels for nuclear reactors - 1973

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    This standard deals with the design, construction, inspection and testing of prestressed concrete pressure vessels for nuclear reactors. Such pressure vessels serve the dual purpose of shielding and containing gas cooled nuclear reactors and are a form of civil engineering structure requiring particularly high integrity, and ensured leak tightness. (Metric)

  1. Cascading pressure reactor and method for solar-thermochemical reactions

    Science.gov (United States)

    Ermanoski, Ivan

    2017-11-14

    Reactors and methods for solar thermochemical reactions are disclosed. The reactors and methods include a cascade of reduction chambers at successively lower pressures that leads to over an order of magnitude pressure decrease compared to a single-chambered design. The resulting efficiency gains are substantial, and represent an important step toward practical and efficient solar fuel production on a large scale.

  2. Pressure vessel for nuclear reactor plant consisting of several pre-stressed cast pressure vessels

    International Nuclear Information System (INIS)

    Bodmann, E.

    1984-01-01

    Several cylindrical pressure vessel components made of pressure castings are arranged on a sector of a circle around the cylindrical cast pressure vessel for accommodating the helium cooled HTR. Each component pressure vessel is connected to the reactor vessel by a horizontal gas duct. The contact surfaces between reactor and component pressure vessel are in one plane. In the spaces between the individual component pressure vessels, there are supporting blocks made of cast iron, which are hollow and also have flat surfaces. With the reactor vessel and the component pressure vessels they form a disc-shaped connecting part below and above the gas ducts. (orig./PW)

  3. 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)

  4. 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.

  5. Evaluation of Pressure Changes in HANARO Reactor Hall after a Reactor Shutdown

    International Nuclear Information System (INIS)

    Han, Geeyang; Han, Jaesam; Ahn, Gukhoon; Jung, Hoansung

    2013-01-01

    The major objective of this work is intended to evaluate the characteristics of the thermal behavior regarding how the decay heat will be affected by the reactor hall pressure change and the increase of pool water temperature induced in the primary coolant after a reactor shutdown. The particular reactor pool water temperature at the surface where it is evaporated owing to the decay heat resulting in the local heat transfer rate is related to the pressure change response in the reactor hall associated with the primary cooling system because of the reduction of the heat exchanger to remove the heat. The increase in the pool water temperature is proportional to the heat transfer rate in the reactor pool. Consequently, any limit on the reactor pool water temperature imposes a corresponding limit on the reactor hall pressure. At HANARO, the decay heat after a reactor shutdown is mainly removed by the natural circulation cooling in the reactor pool. This paper is written for the safety feature of the pressure change related leakage rate from the reactor hall. The calculation results show that the increase of pressure in the reactor hall will not cause any serious problems to the safety limits although the reactor hall pressure is slightly increased. Therefore, it was concluded that the pool water temperature increase is not so rapid as to cause the pressure to vary significantly in the reactor hall. Furthermore, the mathematical model developed in this work can be a useful analytical tool for scoping and parametric studies in the area of thermal transient analysis, with its proper representation of the interaction between the temperature and pressure in the reactor hall

  6. 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

  7. 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)

  8. Cooling of pressurized water nuclear reactor vessels

    International Nuclear Information System (INIS)

    Curet, H.D.

    1978-01-01

    The improvement of pressurized water nuclear reactor vessels comprising flow dividers providing separate and distinct passages for the flow of core coolant water from each coolant water inlet, the flow dividers being vertically disposed in the annular flow areas provided by the walls of the vessel, the thermal shield (if present), and the core barrel is described. In the event of rupture of one of the coolant water inlet lines, water, especially emergency core coolant water, in the intact lines is thus prevented from by-passing the core by circumferential flow around the outermost surface of the core barrel and is instead directed so as to flow vertically downward through the annulus area between the vessel wall and the core barrel in a more normal manner to increase the probability of cooling of the core by the available cooling water in the lower plenum, thus preventing or delaying thermal damage to the core, and providing time for other appropriate remedial or damage preventing action by the operator

  9. Midland reactor pressure vessel flaw distribution

    International Nuclear Information System (INIS)

    Foulds, J.R.; Kennedy, E.L.; Rosinski, S.T.

    1993-12-01

    The results of laboratory nondestructive examination (NDE), and destructive cross-sectioning of selected weldment sections of the Midland reactor pressure vessel were analyzed per a previously developed methodology in order to develop a flaw distribution. The flaw distributions developed from the NDE results obtained by two different ultrasonic test (UT) inspections (Electric Power Research Institute NDE Center and Pacific Northwest Laboratories) were not statistically significantly different. However, the distribution developed from the NDE Center's (destructive) cross-sectioning-based data was found to be significantly different than those obtained through the UT inspections. A fracture mechanics-based comparison of the flaw distributions showed that the cross-sectioning-based data, conservatively interpreted (all defects considered as flaws), gave a significantly lower vessel failure probability when compared with the failure probability values obtained using the UT-based distributions. Given that the cross-sectioning data were reportedly biased toward larger, more significant-appearing (by UT) indications, it is concluded that the nondestructive examinations produced definitively conservative results. In addition to the Midland vessel inspection-related analyses, a set of twenty-seven numerical simulations, designed to provide a preliminary quantitative assessment of the accuracy of the flaw distribution method used here, were conducted. The calculations showed that, in more than half the cases, the analysis produced reasonably accurate predictions

  10. Computerized cost model for pressurized water reactors

    International Nuclear Information System (INIS)

    Meneely, T.K.; Tabata, Hiroaki; Labourey, P.

    1999-01-01

    A computerized cost model has been developed in order to allow utility users to improve their familiarity with pressurized water reactor overnight capital costs and the various factors which influence them. This model organizes its cost data in the standard format of the Energy Economic Data Base (EEDB), and encapsulates simplified relationships between physical plant design information and capital cost information in a computer code. Model calculations are initiated from a base case, which was established using traditional cost calculation techniques. The user enters a set of plant design parameters, selected to allow consideration of plant models throughout the typical three- and four-loop PWR power range, and for plant sites in Japan, Europe, and the United States. Calculation of the new capital cost is then performed in a very brief time. The presentation of the program's output allows comparison of various cases with each other or with separately calculated baseline data. The user can start at a high level summary, and by selecting values of interest on a display grid show progressively more and more detailed information, including links to background information such as individual cost driver accounts and physical plant variables for each case. Graphical presentation of the comparison summaries is provided, and the numerical results may be exported to a spreadsheet for further processing. (author)

  11. Inductive testing of reactor pressure vessels

    International Nuclear Information System (INIS)

    Bergh, H.

    1987-01-01

    In Service Inspection of Reactor Pressure Vessels is mostly done with ultrasonics. Using special 2 crystal-probes good detectability is achieved for near surface defects. The problem is to detect closely spaced cracks, to decide if the defects are surface braking and, if not, to decide the remaining ligament. The purpose of this study is to investigate to what extent Eddy Current can solve these problems. Detecting surfacebreaking cracks and fields of cracks can be done using conventional Eddy Current techniques. Mapping of closely spaced cracks requires a small probe and a high frequency. Measurement of depths a larger probe, a lower frequency and knowledge of the crackfield since 2 closely spaced shallow cracks might be mistaken for one deep crack. Depths of singel cracks can be measured down to 7-8 mm. In closely spaced crackfields the depths can not be measured. The measurement is mostly based on amplitude. For not surface breaking defects the problem is to decide the ligament, i.e. the distance between surface and cracktip. To achieve good penetration a large probe, low frequency and high energy or pulsed energy is used. Ligament up to 4 mm can be measured with good accuracy. The measurements is mostly based on phase. Noise, which originates from rough surface, varied material structure and lift off, can be reduced using multi frequency mix, probe design and scanning pattern. (author)

  12. Calandria cooling structure in pressure tube reactor

    International Nuclear Information System (INIS)

    Hyugaji, Takenori; Sasada, Yasuhiro.

    1976-01-01

    Purpose: To contrive the structure of a heavy water distributing device in a pressure tube reactor thereby to reduce the variation in the cooling function thereof due to the welding deformation and installation error. Constitution: A heating water distributing plate is provided at the lower part of the upper tubular plate of a calandria tank to form a heavy water distributing chamber between both plates and a plurality of calandria tubes. Heavy water which has flowed in the upper part of the heavy water distributing plate from the heavy water inlet nozzle flows down through gaps formed around the calandria tubes, whereby the cooling of the calandria tank and the calandria tubes is carried out. In the above described calandria cooling structure, a heavy water distributing plate support is provided to secure the heavy water distributing plate and torus-shaped heavy water distributing rings are fixed to holes formed in the heavy water distributing plate penetrating through the calandria tubes thereby to form torus-shaped heavy water outlet ports each having a space. (Seki, T.)

  13. Fuel assembly for pressure loss variable PWR type reactor

    International Nuclear Information System (INIS)

    Yoshikuni, Masaaki.

    1993-01-01

    In a PWR type reactor, a pressure loss control plate is attached detachably to a securing screw holes on the lower surface of a lower nozzle to reduce a water channel cross section and increase a pressure loss. If a fuel assembly attached with the pressure loss control plate is disposed at a periphery of the reactor core where the power is low and heat removal causes no significant problem, a flowrate at the periphery of the reactor core is reduced. Since this flowrate is utilized for removal of heat from fuel assemblies of high powder at the center of the reactor core where a pressure loss control plate is not attached, a thermal limit margin of the whole reactor core is increased. Thus, a limit of power peaking can be moderated, to obtain a fuel loading pattern improved with neutron economy. (N.H.)

  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. Fracture risk assessment for the pressurized water reactor pressure vessel under pressurized thermal shock events

    International Nuclear Information System (INIS)

    Chou, Hsoung-Wei; Huang, Chin-Cheng

    2016-01-01

    Highlight: • The PTS loading conditions consistent with the USNRC's new PTS rule are applied as the loading condition for a Taiwan domestic PWR. • The state-of-the-art PFM technique is employed to analyze a reactor pressure vessel. • Novel flaw model and embrittlement correlation are considered in the study. • The RT-based regression formula of NUREG-1874 was also utilized to evaluate the failure risks of RPV. • For slightly embrittled RPV, the SO-1 type PTSs play more important role than other types of PTS. - Abstract: The fracture risk of the pressurized water reactor pressure vessel of a Taiwan domestic nuclear power plant has been evaluated according to the technical basis of the U.S.NRC's new pressurized thermal shock (PTS) screening criteria. The ORNL's FAVOR code and the PNNL's flaw models were employed to perform the probabilistic fracture mechanics analysis associated with plant specific parameters of the domestic reactor pressure vessel. Meanwhile, the PTS thermal hydraulic and probabilistic risk assessment data analyzed from a similar nuclear power plant in the United States for establishing the new PTS rule were applied as the loading conditions. Besides, an RT-based regression formula derived by the U.S.NRC was also utilized to verify the through-wall cracking frequencies. It is found that the through-wall cracking of the analyzed reactor pressure vessel only occurs during the PTS events resulted from the stuck-open primary safety relief valves that later reclose, but with only an insignificant failure risk. The results indicate that the Taiwan domestic PWR pressure vessel has sufficient structural margin for the PTS attack until either the current license expiration dates or during the proposed extended operation periods.

  16. Instrumentation and control strategies for an integral pressurized water reactor

    Directory of Open Access Journals (Sweden)

    Belle R. Upadhyaya

    2015-03-01

    Full Text Available Several vendors have recently been actively pursuing the development of integral pressurized water reactors (iPWRs that range in power levels from small to large reactors. Integral reactors have the features of minimum vessel penetrations, passive heat removal after reactor shutdown, and modular construction that allow fast plant integration and a secure fuel cycle. The features of an integral reactor limit the options for placing control and safety system instruments. The development of instrumentation and control (I&C strategies for a large 1,000 MWe iPWR is described. Reactor system modeling—which includes reactor core dynamics, primary heat exchanger, and the steam flashing drum—is an important part of I&C development and validation, and thereby consolidates the overall implementation for a large iPWR. The results of simulation models, control development, and instrumentation features illustrate the systematic approach that is applicable to integral light water reactors.

  17. Pressurizer pump reliability analysis high flux isotope reactor

    International Nuclear Information System (INIS)

    Merryman, L.; Christie, B.

    1993-01-01

    During a prolonged outage from November 1986 to May 1990, numerous changes were made at the High Flux Isotope Reactor (HFIR). Some of these changes involved the pressurizer pumps. An analysis was performed to calculate the impact of these changes on the pressurizer system availability. The analysis showed that the availability of the pressurizer system dropped from essentially 100% to approximately 96%. The primary reason for the decrease in availability comes because off-site power grid disturbances sometimes result in a reactor trip with the present pressurizer pump configuration. Changes are being made to the present pressurizer pump configuration to regain some of the lost availability

  18. 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)

  19. The inner containment of an EPR trademark pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ostermann, Dirk; Krumb, Christian; Wienand, Burkhard [AREVA GmbH, Offenbach (Germany)

    2014-08-15

    On February 12, 2014 the containment pressure and subsequent leak tightness tests on the containment of the Finnish Olkiluoto 3 EPR trademark reactor building were completed successfully. The containment of an EPR trademark pressurized water reactor consists of an outer containment to protect the reactor building against external hazards (such as airplane crash) and of an inner containment that is subjected to internal overpressure and high temperature in case of internal accidents. The current paper gives an overview of the containment structure, the design criteria, the validation by analyses and experiments and the containment pressure test.

  20. Radiation embrittlement of Spanish nuclear reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Bros, J.; Ballesteros, A.; Lopez, A.

    1993-01-01

    Commercial pressurized water reactor (PWR) and boiling water reactor (BWR) nuclear power plants contain a series of pressure vessel steel surveillance capsules as the principal means of monitoring radiation effects on the pressure vessel. Changes in fracture toughness are more severe in surveillance capsules than in reactor vessel materials because of their proximity of the reactor core. Therefore, it is possible to predict changes in fracture toughness of the reactor vessel materials. This paper describes the status of the reactor vessel surveillance program relating to Spanish nuclear power plants. To date, twelve capsules have been removed and analyzed from seven of the nine Spanish reactors in operation. The results obtained from the analysis of these capsules are compared with the predictions of the Nuclear Regulatory Commission (NRC) Regulatory Guide 1.99, Rev. 2, by means of measured and expected increase of the nil-ductility transition reference temperature (RT NDT ). The comparison is made considering the different variables normally included in the studies of radiation response of reactor pressure vessel materials, such as copper content of steel, level of neutron fluence above 1 MeV, base metal or weld metal, and so forth. The surveillance data have been used for determining the adjusted reference temperatures and upper shelf energies at any time. The results have shown that the seven pressure vessels are in excellent condition to continue operating with safety against brittle fracture beyond the design life, without the need to recuperate the degraded properties of the materials by annealing of the vessel

  1. 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

  2. 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.

  3. 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.

  4. Fuel rod bundles proposed for advanced pressure tube nuclear reactors

    International Nuclear Information System (INIS)

    Prodea, Iosif; Catana, Alexandru

    2010-01-01

    The paper aims to be a general presentation for fuel bundles to be used in Advanced Pressure Tube Nuclear Reactors (APTNR). The characteristics of such a nuclear reactor resemble those of known advanced pressure tube nuclear reactors like: Advanced CANDU Reactor (ACR TM -1000, pertaining to AECL) and Indian Advanced Heavy Water Reactor (AHWR). We have also developed a fuel bundle proposal which will be referred as ASEU-43 (Advanced Slightly Enriched Uranium with 43 rods). The ASEU-43 main design along with a few neutronic and thermalhydraulic characteristics are presented in the paper versus similar ones from INR Pitesti SEU-43 and CANDU-37 standard fuel bundles. General remarks regarding the advantages of each fuel bundle and their suitability to be burned in an APTNR reactor are also revealed. (authors)

  5. 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.

  6. 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

  7. Radiation embrittlement in pressure vessels of power reactors

    International Nuclear Information System (INIS)

    Kempf, Rodolfo; Fortis, Ana M.

    2007-01-01

    It is presented the project to study the effect of lead factors on the mechanical behavior of Reactor Pressure Vessel steels. It is described the facility designed to irradiate Charpy specimens with V notch of SA-508 type 3 steel at power reactor temperature, installed in the RA-1 reactor. The objective is to obtain the fracture behavior of irradiated specimens with different lead factors and to know their dependence with the diffusion of alloy elements. (author) [es

  8. 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)

  9. Surveillance of irradiation embrittlement of nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Najzer, M.

    1982-01-01

    Surveillance of irradiation embrittlement of nuclear reactor pressure vessels is briefly discussed. The experimental techniques and computer programs available for this work at the J. Stefan Institute are described. (author)

  10. TORT application in reactor pressure vessel neutron flux calculations

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  11. Specific Features of Structural-Phase State and Properties of Reactor Pressure Vessel Steel at Elevated Irradiation Temperature

    Directory of Open Access Journals (Sweden)

    E. A. Kuleshova

    2017-01-01

    Full Text Available This paper considers influence of elevated irradiation temperature on structure and properties of 15Kh2NMFAA reactor pressure vessel (RPV steel. The steel is investigated after accelerated irradiation at 300°C (operating temperature of VVER-1000-type RPV and 400°C supposed to be the operating temperature of advanced RPVs. Irradiation at 300°C leads to formation of radiation-induced precipitates and radiation defects-dislocation loops, while no carbide phase transformation is observed. Irradiation at a higher temperature (400°C neither causes formation of radiation-induced precipitates nor provides formation of dislocation loops, but it does increase the number density of the main initial hardening phase—of the carbonitrides. Increase of phosphorus concentration in grain boundaries is more pronounced for irradiation at 400°C as compared to irradiation at 300°C due to influence of thermally enhanced diffusion at a higher temperature. The structural-phase changes determine the changes of mechanical properties: at both irradiation temperatures irradiation embrittlement is mainly due to the hardening mechanism with some contribution of the nonhardening one for irradiation at 400°C. Lack of formation of radiation-induced precipitates at T = 400°C provides a small ΔTK shift (17°C. The obtained results demonstrate that the investigated 15Kh2NMFAA steel may be a promising material for advanced reactors with an elevated operating temperature.

  12. Dismantling method for reactor pressure vessel and system therefor

    International Nuclear Information System (INIS)

    Hayashi, Makoto; Enomoto, Kunio; Kurosawa, Koichi; Saito, Hideyo.

    1994-01-01

    Upon dismantling of a reactor pressure vessel, a containment building made of concretes is disposed underground and a spent pressure vessel is contained therein, and incore structures are contained in the spent pressure vessel. Further, a plasma-welder and a pressing machine are disposed to a pool for provisionally placing reactor equipments in the reactor building for devoluming the incore structures by welding and compression. An overhead-running crane and rails therefor are disposed on the roof and the outer side of the reactor building for transporting the pressure vessel from the reactor building to the containment building. They may be contained in the containment building after incorporation of the incore structures into the pressure vessel at the outside of the reactor building. For the devoluming treatment, a combination of cutting, welding, pressing and the like are optically conducted. A nuclear power plant can be installed by using a newly manufactured nuclear reactor, with no requirement for a new site and it is unnecessary to provide a new radioactive waste containing facility. (N.H.)

  13. Plant life management strategies for pressurized heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Suk; Kwon, Sang Chul; Choo, Ki Nam; Ahn, Sang Bok; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-06-01

    This technical report reviewed aging mechanism of the major components of CANDU 6 reactor such as pressure tubes, calandria tube, end fitting, fuel channel spacer and calandria. Furthermore, the surveillance methodology was described for monitoring and inspection of these core components. Based on the in-reactor performances data such as delayed hydride cracking, leak-before-break, enhanced deformation-creep and growth, the life management of pressure tubes was illustrated in this report. (author). 19 refs., 11 figs., 2 tabs.

  14. Innovative inspection system for reactor pressure vessels

    International Nuclear Information System (INIS)

    Mertens, K.; Trautmann, H.

    1999-01-01

    The versatile, compact and modern underwater systems described, the DELPHIN manipulators and MIDAS submarines, are innovative systems enabling RPV inspections at considerably reduced efforts and time, thus reducing the total time required for ISI of reactors. (orig./CB) [de

  15. Application of the REMIX thermal mixing calculation program for the Loviisa reactor

    International Nuclear Information System (INIS)

    Kokkonen, I.; Tuomisto, H.

    1987-08-01

    The REMIX computer program has been validated to be used in the pressurized thermal shock study of the Loviisa reactor pressure vessel. The program has been verified against the data from the thermal and fluid mixing experiments. These experiments have been carried out in Imatran voima Oy to study thermal mixing of the high-pressure safety injection water in the Loviisa VVER-440 type pressurized water reactor. The verified REMIX-versions were applied to reactor calculations in the probabilistic pressurized thermal shock study of the Loviisa Plant

  16. Neutron fluence determination for light water reactor pressure vessels

    International Nuclear Information System (INIS)

    Gold, R.

    1994-01-01

    A general description of limitations that exist in pressure vessel neutron fluence determinations for commercial light water reactors is presented. Complexity factors that arise in light water reactor pressure vessel neutron fluence calculations are identified and used to analyze calculational limitations. Two broad categories of calculational limitations are introduced, namely benchmark field limitations and deep penetration limitations. Explicit examples of limitations that can arise in each of these two broad categories are presented. These limitations are used to show that the recent draft regulatory guide for the determination of pressure vessel neutron fluence, developed by the Nuclear Regulatory Commission, is based upon procedures and assumptions that are not valid. To eliminate the complexity and limitations of calculational methods, it is recommended that the determination of light water reactor pressure vessel neutron fluence be based upon experiment. Recommendations for improved methods of pressure vessel surveillance neutron dosimetry are advanced

  17. Excess-pressure suppression device in a reactor container

    International Nuclear Information System (INIS)

    Nishio, Masahide

    1985-01-01

    Purpose: To reliably decrease the radioactivity of radioactive gases when they are released externally. Constitution: The exit of a gas exhaust pipe for discharging gases in a reactor container, on generation of an excess pressure in the reactor container upon loss of coolant accident, is adapted to be always fluided in the cooling tank. Then, the exhaust gases discharged in the cooling tank is realeased to the atmosphere. In this way, the excess pressure in the reactor container can be prevented previously and the radioactivity of the gases released externally is significantly reduced by the scrubbing effect. (Kamimura, M.)

  18. Standard Technical Specifications for Westinghouse pressurized water reactors

    International Nuclear Information System (INIS)

    Virgilio, M.J.

    1980-09-01

    The Standard Technical Specifications for Westinghouse Pressurized Water Reactors (W-STS) is a generic document prepared by the U.S. NRC for use in the licensing process of current Westinghouse Pressurized Water Reactors. The W-STS sets forth the Limits, Operating Conditions and other requirements applicable to nuclear reactor facility operation as set forth in by Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public. This document is revised periodically to reflect current licensing requirements

  19. Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Vito, D.J.

    1980-12-01

    The Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors (CE-STS) is a generic document prepared by the US NRC for use in the licensing process of current Combustion Engineering Pressurized Water Reactors. The CE-STS sets forth the limits, operating conditions, and other requirements applicable to nuclear reactor facility operation as set forth by Section 50.36 of 10 CFR 50 for the protection of the health and safety of the public. The document is revised periodically to reflect current licensing requirements

  20. The future 700 MWe pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Bhardwaj, S.A.

    2006-01-01

    The design of a 700 MWe pressurized heavy water reactor has been developed. The design is based on the twin 540 MWe reactors at Tarapur of which the first unit has been made critical in less than 5 years from construction commencement. In the 700 MWe design boiling of the coolant, to a limited extent, has been allowed near the channel exit. While making the plant layout more compact, emphasis has been on constructability. Saving in capital cost of about 15%, over the present units, is expected. The paper describes salient design features of 700 MWe pressurized heavy water reactor

  1. Some local dilution transient in a pressurized water reactor

    International Nuclear Information System (INIS)

    Jacobson, S.

    1989-01-01

    Reactivity accidents are important in the safety analysis of a pressurized water reactor. In this anlysis ejected control rod, steam line break, start of in-active loop and boron dilution accidents are usually dealt with. However, in the analysis is not included what reactivity excursions might happen when a zone,depleted of boron passes the reactor core. This thesis investigates during what operation and emergency conditions diluted zones might exist in a pressurized water reactor and what should be the maximum volumes for then. The limiting transport means are also established in terms of reactivty addition, for the depleted zones. In order to describe the complicated mixing process in the reactor vessel during such a transportation, a typical 3-loop reactor vessel has been modulated by means of TRAC-PF1's VESSEL component. Three cases have been analysed. In the first case the reactor is in a cold condition and the ractor coolant has boron concentration of 2000 ppm. To the reactor vessel is injected an clean water colume of 14 m 3 . In the two other cases the reactor is close to hot shutdown and borated to 850 ppm. To the reactor vessel is added 41 and 13 m 3 clean water, respectively. In the thesis is shown what spatial distribution the depleted zone gets when passing through the reactor vessel in the three cases. The boron concentration in the first case did not decrease the values which would bring the reactor to critical condition. For case two was shown by means of TRAC's point kinetics model that the reactor reaches prompt criticality after 16.03 seconds after starting of the reactor coolant pump. Another prompt criticality occured two seconds later. The total energy developed during the two power escalations were about 55 GJ. A comparision with the criteria used to evaluate the ejected control rod reactivity transient showed that none of these criteria were exceeded. (64 figs.)

  2. Plastic reactor suitable for high pressure and supercritical fluid electrochemistry

    DEFF Research Database (Denmark)

    Branch, Jack; Alibouri, Mehrdad; Cook, David A.

    2017-01-01

    The paper describes a reactor suitable for high pressure, particularly supercritical fluid, electrochemistry and electrodeposition at pressures up to 30 MPa at 115◦C. The reactor incorporates two key, new design concepts; a plastic reactor vessel and the use of o-ring sealed brittle electrodes...... by the deposition of Bi. The application of the reactor to the production of nanostructures is demonstrated by the electrodeposition of ∼80 nm diameter Te nanowires into an anodic alumina on silicon template. Key advantages of the new reactor design include reduction of the number of wetted materials, particularly...... glues used for insulating electrodes, compatability with reagents incompatible with steel, compatability with microfabricated planar multiple electrodes, small volume which brings safety advantages and reduced reagent useage, and a significant reduction in experimental time....

  3. Use of superheated steam to anneal the reactor pressure vessel

    International Nuclear Information System (INIS)

    Porowski, J.S.

    1994-01-01

    Thermal annealing of an embrittled Reactor Pressure Shell is the only recognized means for recovering material properties lost due to long-term exposure of the reactor walls to radiation. Reduced toughness of the material during operation is a major concern in evaluations of structural integrity of older reactors. Extensive studies performed within programs related to life extension of nuclear plants have confirmed that the thermal treatment of 850 degrees F for 168 hours on irradiated material essentially recovers material properties lost due to neutron exposure. Dry and wet annealing methods have been considered. Wet annealing involves operating the reactor at near design temperatures and pressures. Since the temperature of wet annealing must be limited to vessel design temperature of 650 degrees F, only partial recovery of the lost properties is achieved. Thus dry annealing was selected as an alternative for future development and industrial implementation to extend the safe life of reactors

  4. Light Water Reactor-Pressure Vessel Surveillance project computer system

    International Nuclear Information System (INIS)

    Merriman, S.H.

    1980-10-01

    A dedicated process control computer has been implemented for regulating the metallurgical Pressure Vessel Wall Benchmark Facility (PSF) at the Oak Ridge Research Reactor. The purpose of the PSF is to provide reliable standards and methods by which to judge the radiation damage to reactor pressure vessel specimens. Benchmark data gathered from the PSF will be used to improve and standardize procedures for assessing the remaining safe operating lifetime of aging reactors. The computer system controls the pressure vessel specimen environment in the presence of gamma heating so that in-vessel conditions are simulated. Instrumented irradiation capsules, in which the specimens are housed, contain temperature sensors and electrical heaters. The computer system regulates the amount of power delivered to the electrical heaters based on the temperature distribution within the capsules. Time-temperature profiles are recorded along with reactor conditions for later correlation with specimen metallurgical changes

  5. Reactor plant for Belene NPP completion

    International Nuclear Information System (INIS)

    Dragunov, Yu. G.; Ryzhov, S. B.; Ermakov, D. N.; Repin, A. I.

    2004-01-01

    Construction of 'Belene' NPP was started at the end of 80-ties using project U-87 with V-320 reactor plant, general designer of this plant is OKB 'Gidropress'. At the beginning of 90-ties, on completing the considerable number of deliveries and performance of civil engineering work at the site the NPP construction was suspended. Nowadays, considering the state of affairs at the site and the work performed by Bulgarian Party on preservation of the equipment delivered, the most perspective is supposed to be implementation of the following versions in completing 'Belene' NPP: for completion of Unit 1 - reactor plant VVER-1000 on the basis of V-320 reactor with the maximum use of the delivered equipment (V-320M) having the extended service life and safety improvement; for Unit 2 - advanced reactor plant VVER-1000. For the upgraded reactor plant V-230M the basic solutions and characteristics are presented, as well as the calculated justification of strength and safety analyses, design of the reactor core and fuel cycle, instrumentation and control systems, application of the 'leak-before break' in the project and implementation of safety measures. For the modernised reactor plant V-392M the main characteristics and basic changes are presented, concerning reactor pressure vessel, steam generator, reactor coolant pump set. Design of NPP with the modernized reactor plant V-320M meets the up-to-date requirements and can be licensed for completion and operation. In the design of NPP with the advanced reactor plant the basic solutions and the equipment are used that are similar to those used in standard reactor plant V-320 and new one with VVER-1000 under construction and completion in Russia, and abroad. Compliance of reactor design with the up-to-date international requirements, considering the extended service life of the main equipment, shows its rather high potential for implementation during completion of 'Belene' NPP

  6. Introduction to reactor internal materials for pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Woo Suk; Hong, Joon Hwa; Jee, Se Hwan; Lee, Bong Sang; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-06-01

    This report reviewed the R and D states of reactor internal materials in order to be a reference for researches and engineers who are concerning on localization of the materials in the field or laboratory. General structure of PWR internals and material specification for YGN 3 and 4 were reviewed. States-of-arts on R and D of stainless steel and Alloy X-750 were reviewed, and degradation mechanisms of the components were analyzed. In order to develop the good domestic materials for reactor internal, following studies would be carried out: microstructure, sensitization behavior, fatigue property, irradiation-induced stress corrosion cracking/radiation-induced segregation, radiation embrittlement. (Author) 7 refs., 14 figs., 5 tabs.,.

  7. Introduction to reactor internal materials for pressurized water reactor

    International Nuclear Information System (INIS)

    Ryu, Woo Suk; Hong, Joon Hwa; Jee, Se Hwan; Lee, Bong Sang; Kuk, Il Hyun

    1994-06-01

    This report reviewed the R and D states of reactor internal materials in order to be a reference for researches and engineers who are concerning on localization of the materials in the field or laboratory. General structure of PWR internals and material specification for YGN 3 and 4 were reviewed. States-of-arts on R and D of stainless steel and Alloy X-750 were reviewed, and degradation mechanisms of the components were analyzed. In order to develop the good domestic materials for reactor internal, following studies would be carried out: microstructure, sensitization behavior, fatigue property, irradiation-induced stress corrosion cracking/radiation-induced segregation, radiation embrittlement. (Author) 7 refs., 14 figs., 5 tabs.,

  8. Methodology of fuel rod design for pressurized light water reactors

    International Nuclear Information System (INIS)

    Teixeira e Silva, A.; Esteves, A.M.

    1988-01-01

    The fuel performance program FRAPCON-1 and the structural finite element program SAP-IV are applied in a pressurized water reactor fuel rod design methodology. The applied calculation procedure allows to dimension the fuel rod components and characterize its internal pressure. (author) [pt

  9. Containment for small pressurized water reactors

    International Nuclear Information System (INIS)

    Siler, W.C.; Marda, R.S.; Smith, W.R.

    1977-01-01

    Babcock and Wilcox Company has prepared studies under ERDA contract of small and intermediate size (313, 365 and 1200 MWt) PWR reactor plants, for industrial cogeneration or electric power generation. Studies and experience with nuclear plants in this size range indicate unfavorable economics. To offset this disadvantage, modular characteristics of an integral reactor and close-coupled vapor suppression containment have been exploited to shorten construction schedules and reduce construction costs. The resulting compact reactor/containment complex is illustrated. Economic studies to date indicate that the containment design and the innovative construction techniques developed to shorten erection schedules have been important factors in reducing estimated project costs, thus potentially making such smaller plants competetive with competing energy sources

  10. Pressurized heavy-water reactor safety

    International Nuclear Information System (INIS)

    Pease, L.; Wilson, R.

    1977-09-01

    CANDU-PWR type reactors routinely release small amounts of radioactive liquids and gases and large quantities of low-grade waste heat. Radioactive emissions are usually below 1% of the derived release limits based on ICRP limits. Waste heat is common to all power plants and is not foreseen as a problem in Canadian conditions. Risk analysis shows a very low accident probability for CANDU type reactors. Multiple barriers to release of radionuclides, quality assurance, control, and inspection, containment systems, the shutdown system, the ECCS, and leak-before-break design, would all combine to mitigate the effects of an accident. (E.C.B.)

  11. Pressurized water reactor simulator. Workshop material. 2. ed

    International Nuclear Information System (INIS)

    2005-01-01

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development. And the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21, 2nd edition, 'WWER-1000 Reactor Simulator' (2005). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23, 2nd edition, 'Boiling Water Reactor Simulator' (2005). This report consists of course material for workshops using a pressurized water reactor simulator

  12. Evolution of Framatome pressurized water reactor systems

    International Nuclear Information System (INIS)

    Leroy, C.; Bitsch, D.; Millot, J.P.

    1985-10-01

    FRAMATOME's PWR experience covers a total of 63 units, 36 of which are operating by end of 1984. More than 10 units were operated in load follow mode. Progress features, resulting from the feedback of construction and operating experience, and from the returns of a vast research and development program, were incorporated in their design through subsequent series of standard units. The last four loop standard, the N4 model, integrates in a rational way all those progress features, together with a significant design effort. The core design is based on the new Advanced Fuel Assemblies. The reactor control implements the ''Reactor Maximum Flexibility Package'' (R-MAX) which provides a high level of automatic reactor control. The steam generator incorporates an axial-mixed flow economizer design. The triangular-pitch tube bundle, together with modular steam/water separators and a rearrangement of the dryers resulted in a compact design. The reactor coolant pump benefits of higher performances over that of previous models due to an optimal hydraulic design, and of mechanical features which increase margins and facilitate the maintenance work. Following the N4 project, design work on advanced concepts is pursued by FRAMATOME. A main way of research is focused on the optimal use of fissile materials. These concepts are based on tight pitch fuel arrays, associated with a mechanical spectral shift device

  13. Reactor core and passive safety systems descriptions of a next generation pressure tube reactor - mechanical aspects

    Energy Technology Data Exchange (ETDEWEB)

    Yetisir, M.; Gaudet, M.; Rhodes, D.; Hamilton, H.; Pencer, J. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    Canada has been developing a channel-type supercritical water-cooled nuclear reactor concept, often called the Canadian SCWR. The objective of this reactor concept is to meet the technology goals of the Generation IV International Forum (GIF) for the next generation nuclear reactor development, which include enhanced safety features (inherent safe operation and deploying passive safety features), improved resource utilization, sustainable fuel cycle, and greater proliferation resistance than Generation III nuclear reactors. The Canadian SCWR core concept consists of a high-pressure inlet plenum, a separate low-pressure heavy water moderator contained in a calandria vessel, and 336 pressure tubes surrounded by the moderator. The reactor uses supercritical water as a coolant, and a direct steam power cycle to generate electricity. The reactor concept incorporates advanced safety features such as passive core cooling, long-term decay heat rejection to the environment and fuel melt prevention via passive moderator cooling. These features significantly reduce core damage frequency relative to existing nuclear reactors. This paper presents a description of the design concepts for the Canadian SCWR core, reactor building layout and the plant layout. Passive safety concepts are also described that address containment and core cooling following a loss-of coolant accident, as well as long term reactor heat removal at station blackout conditions. (author)

  14. Electrochemical noise measurements under pressurized water reactor conditions

    International Nuclear Information System (INIS)

    Van Nieuwenhove, R.

    2000-01-01

    Electrochemical potential noise measurements on sensitized stainless steel pressure tubes under pressurized water reactor (PWR) conditions were performed for the first time. Very short potential spikes, believed to be associated to crack initiation events, were detected when stressing the sample above the yield strength and increased in magnitude until the sample broke. Sudden increases of plastic deformation, as induced by an increased tube pressure, resulted in slower, high-amplitude potential transients, often accompanied by a reduction in noise level

  15. SCW Pressure-Channel Nuclear Reactor Some Design Features

    Science.gov (United States)

    Pioro, Igor L.; Khan, Mosin; Hopps, Victory; Jacobs, Chris; Patkunam, Ruban; Gopaul, Sandeep; Bakan, Kurtulus

    Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950s and 1960s in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with SuperCritical Water (SCW) became attractive again as the ultimate development path for water cooling. The main objectives of using SCW in nuclear reactors are: 1) to increase the thermal efficiency of modern Nuclear Power Plants (NPPs) from 30-35% to about 45-48%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs (˜1000 US/kW or even less). SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625°C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. Pressure-tube or pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia for some time. Some design features of the Canadian concept related to fuel channels are discussed in this paper. The main conclusion is that the development of SCW pressure-tube nuclear reactors is feasible and significant benefits can be expected over other thermal-energy systems.

  16. Pressure equalization systems in pressurized water reactor fuel rods

    International Nuclear Information System (INIS)

    Steven, J.; Wunderlich, F.

    1979-01-01

    For the development of a pressure reduction system in PWR fuel rods the capability of charcoal to adsorb Helium, Xenon and Krypton at temperatures of about 300 0 C was investigated. The influence of the adsorption on fuel rod internal pressure and in creep strain on the tube was evaluated in a design study. (orig.) [de

  17. Limit regulation system for pressurized water nuclear reactors

    International Nuclear Information System (INIS)

    Aleite, W.; Bock, H.W.

    1976-01-01

    Described is a limit regulation system for a pressurized water nuclear reactor in combination with a steam generating system connected to a turbine, the nuclear reactor having control rods as well as an operational regulation system and a protective system, which includes reactor power limiting means operatively associated with the control rods for positioning the same and having response values between operating ranges of the operational regulation system, on the one hand, and response values of the protective system, on the other hand, and a live steam-minimal pressure regulation system cooperating with the reactor power limiting means and operatively connected to a steam inlet valve to the turbine for controlling the same

  18. Lessons Learned From Developing Reactor Pressure Vessel Steel Embrittlement Database

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL

    2010-08-01

    Materials behaviors caused by neutron irradiation under fission and/or fusion environments can be little understood without practical examination. Easily accessible material information system with large material database using effective computers is necessary for design of nuclear materials and analyses or simulations of the phenomena. The developed Embrittlement Data Base (EDB) at ORNL is this comprehensive collection of data. EDB database contains power reactor pressure vessel surveillance data, the material test reactor data, foreign reactor data (through bilateral agreements authorized by NRC), and the fracture toughness data. The lessons learned from building EDB program and the associated database management activity regarding Material Database Design Methodology, Architecture and the Embedded QA Protocol are described in this report. The development of IAEA International Database on Reactor Pressure Vessel Materials (IDRPVM) and the comparison of EDB database and IAEA IDRPVM database are provided in the report. The recommended database QA protocol and database infrastructure are also stated in the report.

  19. Assessment of a small pressurized water reactor for industrial energy

    International Nuclear Information System (INIS)

    Klepper, O.H.; Fuller, L.C.; Myers, M.L.

    1977-01-01

    An evaluation of several recent ERDA/ORNL sponsored studies on the application of a small, 365 MW(t) pressurized water reactor for industrial energy is presented. Preliminary studies have investigated technical and reliability requirements; costs for nuclear and fossil based steam were compared, including consideration of economic inflation and financing methods. For base-load industrial steam production, small reactors appear economically attractive relative to coal fired boilers that use coal priced at $30/ton

  20. Enriched uranium cycles in pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Mazzola, A.

    1994-01-01

    A study was made on the substitution of natural uranium with enriched and on plutonium recycle in unmodified PHWRs (pressure vessel reactor). Results clearly show the usefulness of enriched fuel utilisation for both uranium ore consumption (savings of 30% around 1.3% enrichment) and decreasing fuel cycle coasts. This is also due to a better plutonium exploitation during the cycle. On the other hand plutonium recycle in these reactors via MOX-type fuel appears economically unfavourable under any condition

  1. An expert system for pressurized water reactor load maneuvers

    International Nuclear Information System (INIS)

    Chaung Lin; Jungping Chen; Yihjiunn Lin; Lianshin Lin

    1993-01-01

    Restartup after reactor shutdown and load-follow operations are the important tasks in operating pressurized water reactors. Generally, the most efficient method is to apply constant axial offset control (CAOC) strategy during load maneuvers. An expert system using CAOC strategy, fuzzy reasoning, a two-node core model, and operational constraints has been developed. The computation time is so short that this system, which leads to an approximate closed-loop control, could be useful for on-site calculation

  2. Thermal annealing of an embrittled reactor pressure vessel

    International Nuclear Information System (INIS)

    Mager, T.R.; Dragunov, Y.G.; Leitz, C.

    1998-01-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. Chapter 11 deals with thermal annealing of an embrittled reactor pressure vessel. Anneal procedures for vessels from both the US and the former USSR are mentioned schematically, wet anneals at lower temperature and dry anneals above RPV design temperatures are investigated. It is shown that heat treatment is a means of recovering mechanical properties which were degraded by neutron radiation exposure, thus assuring reactor pressure vessel compliance with regulatory requirements

  3. Design of virtual SCADA simulation system for pressurized water reactor

    International Nuclear Information System (INIS)

    Wijaksono, Umar; Abdullah, Ade Gafar; Hakim, Dadang Lukman

    2016-01-01

    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor

  4. Design of virtual SCADA simulation system for pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wijaksono, Umar, E-mail: umar.wijaksono@student.upi.edu; Abdullah, Ade Gafar; Hakim, Dadang Lukman [Electrical Power System Research Group, Department of Electrical Engineering Education, Jl. Dr. Setiabudi No. 207 Bandung, Indonesia 40154 (Indonesia)

    2016-02-08

    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

  5. 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

  6. 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)

  7. Heat insulation device for reactor pressure vessel in water

    International Nuclear Information System (INIS)

    Nakamura, Heiichiro; Tanaka, Yoshimi.

    1993-01-01

    Outer walls of a reactor pressure vessel are covered with water-tight walls made of metals. A heat insulation metal material is disposed between them. The water tight walls are joined by welding and flanges. A supply pipeline for filling gases and a discharge pipeline are in communication with the inside of the water tight walls. Further, a water detector is disposed in the midway of the gas discharge pipeline. With such a constitution, the following advantages can be attained. (1) Heat transfer from the reactor pressure vessel to water of a reactor container can be suppressed by filled gases and heat insulation metal material. (2) Since the pressure at the inside of the water tight walls can be equalized with the pressure of the inside of the reactor container, the thickness of the water-tight walls can be reduced. (3) Since intrusion of water to the inside of the walls due to rupture of the water tight walls is detected by the water detector, reactor scram can be conducted rapidly. (4) The sealing property of the flange joint portion is sufficient and detaching operation thereof is easy. (I.S.)

  8. Decay ratio estimation in pressurized water reactor

    International Nuclear Information System (INIS)

    Por, G.; Runkel, J.

    1990-11-01

    The well known decay ratio (DR) from stability analysis of boiling water reactors (BWR) is estimated from the impulse response function which was evaluated using a simplified univariate autoregression method. This simplified DR called modified DR (mDR) was applied on neutron noise measurements carried out during five fuel cycles of a 1300 MWe PWR. Results show that this fast evaluation method can be used for monitoring of the growing oscillation of the neutron flux during the fuel cycles which is a major concern of utilities in PWRs, thus it can be used for estimating safety margins. (author) 17 refs.; 10 figs

  9. 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)

  10. Design of an integral missile shield in integrated head assembly for pressurized water reactor at commercial nuclear plants

    International Nuclear Information System (INIS)

    Baliga, Ravi; Watts, Tom Neal; Kamath, Harish

    2015-01-01

    containment and damaging other safety components in the containment that are required to be operated for safe shutdown of the reactor. The missile shield in the IHA is designed to absorb missile energy due to an impact from missiles associated with a postulated CRDM housing break. This paper provides details of the CRDM missile shield design in the IHA for Westinghouse Pressurized Water Reactors (PWR) and it can be extended to other PWRs such as VVERs. (author)

  11. Assessment of the integrity of WWER type reactor pressure vessels

    International Nuclear Information System (INIS)

    Brumovsky, M.

    1995-01-01

    Procedures are given for the assessment of the residual lifetime of reactor pressure vessels with respect to a sudden failure, the lifetime of vessels with defects disclosed during in-service inspections, and the fatigue or corrosion-mechanical lifetime. Also outlined are the ways of assessing the effects of major degradation mechanisms, i.e. radiation embrittlement, thermal aging, and fatigue damage, including the use of calculated values and experimental examination, by means of surveillance specimens in particular. All results of assessment performed so far indicate that the life of reactor pressure vessels at the Dukovany, Jaslovske Bohunice, and Temelin nuclear power plants is well secured. 7 figs., 3 refs

  12. Acoustic Emission for on-line reactor pressure boundary monitoring

    International Nuclear Information System (INIS)

    Hutton, P.H.; Kurtz, R.J.; Pappas, R.A.

    1985-01-01

    The program objective is to develop AE for continuous surveillance to assess flaw growth in reactor pressure boundaries. Technology in the laboratory is being evaluated on structures. Results have demonstrated basic feasibility of the program objective. AE monitoring a long term fatigue test of a pressure vessel demonstrated an instrument system, and the ability to detect unexpected as well as well as known fatigue cracks. Monitoring a nuclear reactor system shows that the coolant flow noise problem is manageable and AE can be detected under simulated operating conditions

  13. Microstructural evolution in neutron irradiated reactor pressure vessel steels

    International Nuclear Information System (INIS)

    English, C.A.; Phythian, W.J.

    1998-01-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. The microstructural evolution in neutron irradiated reactor pressure vessel steels is described. The damage mechanisms are elaborated and techniques for examining the microstructure are suggested. The importance of the initial damage event is analysed, and the microstructural evolution in RPV steels is examined

  14. 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

  15. PWR [pressurized water reactor] pressurizer transient response: Final report

    International Nuclear Information System (INIS)

    Murphy, S.I.

    1987-08-01

    To predict PWR pressurizer transients, Ahl proposed a three region model with a universal coefficient to represent condensation on the water surface. Specifically, this work checks the need for three regions and the modeling of the interfacial condensation coefficient. A computer model has been formulated using the basic mass and energy conservation laws. A two region vapor and liquid model was first used to predict transients run on a one-eleventh scale Freon pressurizer. These predictions verified the need for a second liquid region. As a result, a three region model was developed and used to predict full-scale pressurizer transients at TMI-2, Shippingport, and Stade. Full-scale pressurizer predictions verified the three region model and pointed out the shortcomings of Ahl's universal condensation coefficient. In addition, experiments were run using water at low pressure to study interface condensation. These experiments showed interface condensation to be significant only when spray flow is turned on; this result was incorporated in the final three region model

  16. Assessment of capability for modeling the core degradation in 2D geometry with ASTEC V2 integral code for VVER type of reactor

    International Nuclear Information System (INIS)

    Dimov, D.

    2011-01-01

    The ASTEC code is progressively becoming the reference European severe accident integral code through in particular the intensification of research activities carried out since 2004. The purpose of this analysis is to assess ASTEC code modelling of main phenomena arising during hypothetical severe accidents and particularly in-vessel degradation in 2D geometry. The investigation covers both early and late phase of degradation of reactor core as well as determination of corium which will enter the reactor cavity. The initial event is station back-out. In order to receive severe accident condition, failure of all active component of emergency core cooling system is apply. The analysis is focus on ICARE module of ASTEC code and particularly on so call MAGMA model. The aim of study is to determine the capability of the integral code to simulate core degradation and to determine the corium composition entering the reactor cavity. (author)

  17. Introduction of advanced pressurized water reactors in France

    International Nuclear Information System (INIS)

    Millot, J.P.; Nigon, M.; Vitton, M.

    1988-01-01

    Designed >30 yr ago, pressurized water reactors (PWRs) have evolved well to match the current safety, operating, and economic requirements. The first advanced PWR generation, the N4 reactor, is under construction with 1992 as a target date for commercial operation. The N4 may be considered to be a technological outcome of PWR evolution, providing advances in the fields of safety, man/machine interfaces, and load flexibility. As a step beyond N4, a second advanced PWR generation is presently under definition with, as a main objective, a greater ability to cope with the possible deterioration of the natural uranium market. In 1986, Electricite de France (EdF) launched investigations into the possible characteristics of this advanced PWR, called REP-2000 (PWR-2000: the reactor for the next century). Framatome joined EdF in 1987 but had been working on a new tight-lattice reactor. Main options are due by 1988; preliminary studies will begin and, by 1990, detailed design will proceed with the intent of firm commitments for the first unit by 1995. Commissioning is planned in the early years of the next century. This reactor type should be either an improved version of the N4 reactor or a spectral shift convertible reactor (RCVS). Through research and development efforts, Framatome, Commissariat a l'Energie Atomique (CEA), and EdF are investigating the physics of fuel rod tight lattices including neutronics, thermohydraulics, fuel behavior, and reactor mechanics

  18. 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)

  19. In-service ultrasonic inspection of nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Prepechal, J.; Sulc, J.

    1982-01-01

    Ultrasonic tests of pressure vessels for WWER 440 reactors, type 213 V, are carried out partly manually and partly by test equipment. The inner surface of the pressure vessel is tested using device REACTORTEST TRC which is fully mobile. The outer surface of the cylindrical parts and bottoms of the body is tested using handling equipment permanently in-built under the pressure vessel and dismountable testing heads. A set of these heads may be used for two reactor units. The testing equipment REACTORTEST TRC is equipped with a TRC 800 ultrasound device. The equipment for testing the outer surface of the vessel operates with the UDAR 16 ultrasound apparatus to which may be simultaneously connected 10 ultrasound probes and six probes for acoustic feedback. The whole system of ultrasonic tests makes possible a first-rate and reliable volume control of the whole pressure vessel and all points where cracks may originate and grow. (Z.M.)

  20. Absorber rod bundle actuator in a pressurized water nuclear reactor

    International Nuclear Information System (INIS)

    Martin, J.; Peletan, R.

    1984-01-01

    The invention concerns an absorber rod bundle actuator in a pressurized water reactor with spectral shift control. The device comprises two coaxial control bars. The inner bar is integral with the absorber rod bundle; it has an enlarged zone which acts as a proton under pressure difference across an annular seal which can be radially expanded, the pressure difference allowing to the absorber rod bundles actuating on the piston. When a pressure difference is applied, the seal expands radially by a sufficient amount to make sealing contact with the zone of larger diameter in the outer bar. The invention applies more particularly to reactors with spectral shift control using bundles of fertile rods [fr

  1. 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)

  2. 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)

  3. High pressure sealing systems for nuclear reactors

    International Nuclear Information System (INIS)

    Garam, E. de

    1993-01-01

    TIA is the FRAMATOME Division in charge of design, manufacture maintenance and improvement of reactor core instrumentation. In the course of its activities, TIA was rapidly confronted with problems of leakage occurring in PWR in-core instrumentation, both in the neutron flux measurement system (flux thimbles and thimble guide tubes) and in the equipment used for core temperature sensing. TIA has likewise placed emphasis, in setting objectives for its operations, on improving instrumentation reliability, reducing maintenance costs and limiting the radiation doses sustained during maintenance. The very satisfactory results achieved by TIA in all of these areas have led us to look to the future with confidence. The purpose of this presentation is to describe the various improvements devised by TIA over the years and to take inventory of the experience gained by the Division with instrumentation for all types of nuclear power plants. (author)

  4. A powerful methodology for reactor vessel pressurized thermal shock analysis

    International Nuclear Information System (INIS)

    Boucau, J.; Mager, T.

    1994-01-01

    The recent operating experience of the Pressurized Water Reactor (PWR) Industry has focused increasing attention on the issue of reactor vessel pressurized thermal shock (PTS). More specifically, the review of the old WWER-type of reactors (WWER 440/230) has indicated a sensitive behaviour to neutron embrittlement. This led already to some remedial actions including safety injection water preheating or vessel annealing. Such measures are usually taken based on the analysis of a selected number of conservative PTS events. Consideration of all postulated cooldown events would draw attention to the impact of operator action and control system effects on reactor vessel PTS. Westinghouse has developed a methodology which couples event sequence analysis with probabilistic fracture mechanics analyses, to identify those events that are of primary concern for reactor vessel integrity. Operating experience is utilized to aid in defining the appropriate event sequences and event frequencies of occurrence for the evaluation. Once the event sequences of concern are identified, detailed deterministic thermal-hydraulic and structural evaluations can be performed to determine the conditions required to minimize the extension of postulated flaws or enhance flaw arrest in the reactor vessel. The results of these analyses can then be used to better define further modifications in vessel and plant system design and to operating procedures. The purpose of the present paper will be to describe this methodology and to show its benefits for decision making. (author). 1 ref., 3 figs

  5. Pressure suppression device for a reactor container

    International Nuclear Information System (INIS)

    Shimizu, Toshiaki

    1982-01-01

    Purpose: To prevent damages in drain pipes or the likes upon the water level increase due to blowing of incompressible gases. Constitution: An exhaust pipe for guiding escaping steams is connected to a main steam releaf valve. The exhaust pipe is guided into pressure-suppression-chamber water through the inside of a dry-well and by way of a vent pipe, a vent header and a drain pipe or a downcomer. Since the exhaust pipe is not exposed to the water surface inside the pressure suppression chamber, even if steams blow out into the dry-well by the rapture of pipeways or the likes to rapidly increase the water level, the water surface does not hit on the exhaust pipe, whereby the damages for the exhaust pipe and support members can be prevented to improve the reliability. (Seki, T.)

  6. 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.

  7. Development of alternative fuel for pressurized water reactors

    International Nuclear Information System (INIS)

    Cardoso, P.E.; Ferreira, R.A.N.; Ferraz, W.B.; Lameiras, F.S.; Santos, A.; Assis, G. de; Doerr, W.O.; Wehner, E.L.

    1984-01-01

    The utilization of alternative fuel cycles in Pressurized Water Reactors (PWR) such as Th/U and Th/Pu cycles can permit a better utilization of uranium reserves without the necessity of developing new power reactor concepts. The development of the technology of alternative fuels for PWR is one of the objectives of the 'Program on Thorium Utilization in Pressurized Water Reactors' carried out jointly by Empresas Nucleares Brasileiras S.A. (NUCLEBRAS), through its Centro de Desenvolvimento da Tecnologia Nuclear (CDTN) and by German institutions, the Julich Nuclear Research Center (KFA), the Kraftwerk Union A.G. (KWU) and NUKEM GmbH. This paper summarizes the results so far obtained in the fuel technology. The development of a fabrication process for PWR fuel pellets from gel-microspheres is reported as well as the design, the specification, and the fabrication of prototype fuel rods for irradiation tests. (Author) [pt

  8. Pressure vessel codes: Their application to nuclear reactor systems

    International Nuclear Information System (INIS)

    1966-01-01

    A survey has been made by the International Atomic Energy Agency of how the problems of applying national pressure vessel codes to nuclear reactor systems have been treated in those Member States that had pressurized reactors in operation or under construction at the beginning of 1963. Fifteen answers received to an official inquiry form the basis of this report, which also takes into account some recently published material. Although the answers to the inquiry in some cases data back to 1963 and also reflect the difficulty of describing local situations in answer to standard questions, it is hoped that the report will be of interest to reactor engineers. 21 refs, 1 fig., 2 tabs

  9. Structural integrity evaluation of PWR nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Cruz, Julio R.B.; Mattar Neto, Miguel

    1999-01-01

    The reactor pressure vessel (RPV) is the most important structural component of a PWR nuclear power plant. It contains the reactor core and is the main component of the primary system pressure boundary, the system responsible for removing the heat generated by the nuclear reactions. It is considered not replaceable and, therefore, its lifetime is a key element to define the plant life as a whole. Three critical issues related to the reliability of the RPV structural integrity come out by reason of the radiation damage imposed to the vessel material during operation. These issues concern the definition of pressure versus temperature limits for reactor heatup and cooldown, pressurized thermal shock evaluation and assessment of reactor vessels with low upper shelf Charpy impact energy levels. This work aims to present the major aspects related to these topics. The requirements for preventing fracture of the RPV are reviewed as well as the available technology for assessing the safety margins. For each mentioned problem, the several steps for structural integrity evaluation are described and the analysis methods are discussed. (author)

  10. The installation welding of pressure water reactor coolant piping

    International Nuclear Information System (INIS)

    Deng Feng

    2010-01-01

    Large pressure water reactor nuclear power plants are constructing in our country. There are three symmetry standard loops in reactor coolant system. Each loop possesses a steam generator and a primary poop, in which one of the loops is equipped with a pressurizer. These components are connected with reactor pressure vessel by installation welding of the coolant piping. The integrity of reactor coolant pressure boundary is the second barrier to protect the radioactive substance from release to outside, so the safe operation of nuclear power plant is closely related to the quality of coolant piping installation welding. The heavy tube with super low carbon content austenitic stainless steel is selected for coolant piping. This kind of material has good welding behavior, but the poor thermal conductivity, the big liner expansion coefficient and the big welding deformation will cause bigger welding stress. To reduce the welding deformation, to control the dimension precision, to reduce the residual stress and to ensure the welding quality the installation sequence should be properly designed and the welding technology should be properly controlled. (authors)

  11. Marine reactor pressure vessels dumped in the Kara Sea

    International Nuclear Information System (INIS)

    Mount, M.E.

    1997-01-01

    Between 1965 and 1988, 16 marine reactors from seven Russian submarines and the icebreaker Lenin, each of which suffered some form of reactor accident, were dumped in a variety of containments, using a number of sealing methods, at five sites in the Kara Sea. All reactors were dumped at sites that varied in depth from 12 to 300 m and six contained their spent nuclear fuel (SNF). This paper examines the breakdown of the reactor pressure vessel (RPV) barriers due to corrosion, with specific emphasis on those RPVs containing SNF. Included are discussions of the structural aspects of the steam generating installations and their associated RPVs, a summary of the disposal operations, assumptions on corrosion rates of structural and filler materials, and an estimate of the structural integrity of the RPVs at the present time (1996) and in the year 2015

  12. 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

  13. Aging considerations for PWR [pressurized water reactor] control rod drive mechanisms and reactor internals

    International Nuclear Information System (INIS)

    Ware, A.G.

    1988-01-01

    This paper describes age-related degradation mechanisms affecting life extension of pressurized water reactor control rod drive mechanisms and reactor internals. The major sources of age-related degradation for control rod drive mechanisms are thermal transients such as plant heatups and cooldowns, latchings and unlatchings, long-term aging effects on electrical insulation, and the high temperature corrosive environment. Flow induced loads, the high-temperature corrosive environment, radiation exposure, and high tensile stresses in bolts all contribute to aging related degradation of reactor internals. Another problem has been wear and fretting of instrument guide tubes. The paper also discusses age-related failures that have occurred to date in pressurized water reactors

  14. Press kit. EPR (European pressurized water reactor). The advanced nuclear reactor

    International Nuclear Information System (INIS)

    2004-10-01

    Nuclear energy, which provides a steady supply of electricity at low cost, has its rightful place in the energy mix of the 21 century, which puts the emphasis on sustainable development. In this framework, this document presents the advantages of the EPR (European Pressurized water Reactor). The EPR is the only third generation reactor under construction today. It is an evolutionary reactor that represents a new generation of pressurized water reactors with no break in the technology used for the most recent models. The EPR can guarantee a safe, inexpensive electricity supply, without adding to the greenhouse effect. It meets the requirements of the safety authorities and lives up to the expectations of electricity utilities. (A.L.B.)

  15. Leak detection device for nuclear reactor pressure vessel

    International Nuclear Information System (INIS)

    Ikeda, Jun.

    1988-01-01

    Purpose: To test the leakage of a nuclear reactor pressure vessel during stopping for a short period of time with no change to the pressure vessel itself. Constitution: The device of the present invention comprises two O-rings disposed on the flange surface that connects a pressure vessel main body and an upper cover, a leak-off pipeway derived from the gap of the O-rings at the flange surface to the outside of the pressure vessel, a pressure detection means connected to the end of the pipeway, a humidity detection means disposed to the lead-off pipeway, a humidity detection means disposed to the lead-off pipeway, and gas supply means and gas suction means disposed each by way of a check valve to a side pipe branched from the pipeway. After stopping the operation of the nuclear reactor and pressurizing the pressure vessel by filling water, gases supplied to the gap between the O-rings at the flange surface by opening the check valve. In a case where water in the pressure vessel should leak to the flange surface, when gas suction is applied by properly opening the check valve, increase in the humidity due to the steams of leaked water diffused into the gas is detected to recognize the occurrence of leakage. (Kamimura, M.)

  16. Dual pressurized light water reactor producing 2000 M We

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-10-15

    The dual unit optimizer 2000 M We (Duo2000) is proposed as a new design concept for large nuclear power plant. Duo is being designed to meet economic and safety challenges facing the 21 century green and sustainable energy industry. Duo2000 has two nuclear steam supply systems (NSSS) of the unit nuclear optimizer (Uno) pressurized water reactor (PWR) in a single containment so as to double the capacity of the plant. Uno is anchored to the optimized power reactor 1000 M We (OPR1000) of the Korea Hydro and Nuclear Power Co., Ltd. The concept of Duo can be extended to any number of PWRs or pressurized heavy water reactors (PHWR s), or even boiling water reactor (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. In particular, since it is required that the small and medium sized reactors (SMRs) be built as units, the concept of Duo2000 will apply to SMRs as well. With its in-vessel retention as severe accident management strategy, Duo can not only put the single most querulous PWR safety issue to end, but also pave ways to most promising large power capacity dispensing with huge redesigning cost for generation III + nuclear systems. The strengths of Duo2000 include reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting NSSS. The technology can further be extended to coupling modular reactors as dual, triple, or quadruple units to increase their economics, thus accelerating the commercialization as well as the customization of SMRs. (Author)

  17. Decontamination and recycle of zirconium pressure tubes from Pressurized Heavy Water Reactor

    International Nuclear Information System (INIS)

    Gantayet, L.M.; Verma, R.; Remya Devi, P.S.; Banerjee, S.; Kotak, V.; Raha, A.; Sandeep, K.C.; Joshi, Shreeram W.; Lali, A.M.

    2009-01-01

    An ion exchange process has been developed for decontamination of zirconium pressure tubes from Pressurized Heavy Water Reactor and recycling of neutronically improved zirconium. Distribution coefficient, equilibrium isotherm, kinetic and breakthrough data were used to develop the separation process. Effect of gamma radiation on indigenous resins was also studied to assess their suitability in high radiation field. (author)

  18. 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)

  19. 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

  20. 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.

  1. Advanced Approach of Reactor Pressure Vessel In-service Inspection

    International Nuclear Information System (INIS)

    Matokovic, A.; Picek, E.; Pajnic, M.

    2006-01-01

    The most important task of every utility operating a nuclear power plant is the continuously keeping of the desired safety and reliability level. This is achieved by the performance of numerous inspections of the components, equipment and system of the nuclear power plant in operation and in particular during the scheduled maintenance periods at re-fueling time. Periodic non-destructive in-service inspections provide most relevant criteria of the integrity of primary circuit pressure components. The task is to reliably detect defects and realistically size and characterize them. One of most important and the most extensive examination is a reactor pressure vessel in-service inspection. That inspection demand high standards of technology and quality and continual innovation in the field of non-destructive testing (NDT) advanced technology as well as regarding reactor pressure vessel tool and control systems. A remote underwater contact ultrasonic technique is employed for the examination of the defined sections (reactor welds), whence eddy current method is applied for clad surface examinations. Visual inspection is used for examination of the vessel inner surface. The movement of probes and data positioning are assured by using new reactor pressure vessel tool concept that is fully integrated with NDT systems. The successful performance is attributed thorough pre-outage planning, training and successful performance demonstration qualification of chosen NDT 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. The advanced approach as presented in this paper offer more flexibility of application (non-destructive tests, local grinding action as well as taking of boat samples

  2. 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

  3. Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

    Energy Technology Data Exchange (ETDEWEB)

    Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

    1997-04-01

    This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications.

  4. Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

    International Nuclear Information System (INIS)

    Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

    1997-01-01

    This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications

  5. Light-water reactor pressure vessel surveillance standards

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    The master matrix standard describes a series of standard practices, guides, and methods for the prediction of neutron-induced changes in light-water reactor (LWR) pressure vessel steels throughout a pressure vessel's service life. Some of these are existing American Society for Testing and Materials (ASTM) standards, some are ASTM standards that have been modified, and some are newly proposed ASTM standards. The current (1) scope, (2) areas of application, (3) interrelationships, and (4) status and time table of development, improvement, validation, and calibration for a series of 16 ASTM standards are defined. The standard also includes a discussion of LWR pressure vessel surveillance - justification, requirements, and status of work

  6. 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

  7. Manipulator for testing a top-opened reactor pressure vessel

    International Nuclear Information System (INIS)

    Bauer, R.; Kastl, H.

    1991-01-01

    The design is described of a manipulator to be inserted into the inside of reactor pressure vessels opened at the top. The main components of the manipulator include a fixed column protruding into the pressure vessel and a support which is slidable on the column and carries the bearing component for the measuring, testing, inspection and repair instruments. The device includes a driving equipment for the support as well as the power supply for the sets accommodated on the support, with the aim to reduce the failure rate of the manipulator as a whole, shorten the time necessary for its assembling and thus the time of staying in the reactor pressure vessel and, at the same time, make its maintenance and operation easier. (Z.S.). 13 figs

  8. Welding of the A1 reactor pressure vessel

    International Nuclear Information System (INIS)

    Becka, J.

    1975-01-01

    As concerns welding, the A-1 reactor pressure vessel represents a geometrically complex unit containing 1492 welded joints. The length of welded sections varies between 10 and 620 mm. At an operating temperature of 120 degC and a pressure of 650 N/cm 2 the welded joints in the reactor core are exposed to an integral dose of 3x10 18 n/cm 2 . The chemical composition is shown for pressure vessel steel as specified by CSN 413090.9 modified by Ni, Ti and Al additions, and for the welding electrodes used. The requirements are also shown for the mechanical properties of the base and the weld metals. The technique and conditions of welding are described. No defects were found in ultrasonic testing of welded joints. (J.B.)

  9. 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

  10. 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)

  11. 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)

  12. Nuclear power - replacement of pressure tubes in CANDU reactors

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    The CANDU pressure tube reactor is an effective electricity generator. While most units have been built in Canada, units are successfully operated in Argentina and Korea as well as India and Pakistan, which have early versions of the same concept. Units are also under construction in Korea and Romania. The main constructional components of a CANDU core are the calandria vessel, the fuel channels and the reactivity control mechanisms. The fuel channel, in particular the pressure tubes, see an environment comprising high flux, high temperature water at high pressures, which induces changes in the properties and dimensions of the channel components. From the first, fuel channels were designed to be replaced because of the difficulty in predicting the behaviour of zirconium alloys in such service over a long period of time. In fact some phenomena, that were not known at the time of the earliest designs, have led to unacceptable changes in the properties of the channels and these early reactors have had to be retubed at half their intended life. These deficiencies have been corrected in the latest designs and fuel channels in reactors that have commenced operation over the last 10 years, are predicted to reach the intended 30 years life before replacement is necessary. The changing of fuel channels, the details and experience of which are explained, has been shown to be an effective way of refurbishing the CANDU reactor, extending its lifetime a further 25-30 years. (author)

  13. The Westinghouse Advanced Passive Pressurized Water Reactor, AP1000

    International Nuclear Information System (INIS)

    Schene, R.

    2009-01-01

    Featuring proven technology and innovative passive safety systems, the Westinghouse AP1000 pressurized water reactor can achieve competitive generation costs in the current electricity market without emitting harmful greenhouse gases and further harming the environment. Westinghouse Electric Company, the pioneer in nuclear energy once again sets a new industry standard with the AP1000. The AP1000 is a two-loop pressurized water reactor that uses simplified, innovative and effective approach to safety. With a gross power rating of 3415 megawatt thermal and a nominal net electrical output of 1117 megawatt electric, the AP1000 is ideal for new base load generation. The AP1000 is the safest and most economical nuclear power plant available in the worldwide commercial marketplace, and is the only Generation III+ reactor to receive a design certification from the U.S. Nuclear Regulatory Commission (NRC). Based on nearly 20 years of research and development, the AP1000 builds and improves upon the established technology of major components used in current Westinghouse designed plants. These components, including steam generators, digital instrumentation and controls, fuel, pressurizers, and reactor vessels, are currently in use around the world and have years of proven, reliable operating experience. Historically, Westinghouse plant designs and technology have forged the cutting edge technology of nuclear plant around the world. Today, nearly 50 percent of the world's 440 nuclear plants are based on Westinghouse technology. Westinghouse continues to be the nuclear industry's global leader. (author)

  14. Technological status of reactor coolant pumps in generation III+ pressurized nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Brecht, Bernhard; Bross, Stephan [KSB Aktiengesellschaft, Frankenthal (Germany)

    2016-05-15

    KSB has been developing and producing pumps for thermal power plants for nearly 90 years. Consequently, KSB also started to develop and manufacture pumps for all kinds of nuclear power plants from the very beginning of the civil use of nuclear energy. This is especially true for reactor coolant pumps for pressurized water reactors. For the generation of advanced evolutionary reactors (Generation III+ reactors), KSB developed an advanced shaft seal system which is also able to fulfill the requirements of station blackout conditions. The tests in the KSB test rigs, which were successfully completed in December 2015, proved the full functionality of the new design. For generation III+ passive plant reactors KSB developed a new reactor coolant pump type called RUV, which is based on the experience of classic reactor coolant pumps and reactor internal pumps. It is a very compact, hermetically sealed vertical pump-motor unit with a wet winding motor. A full scale prototype successfully passed the 1st stage qualification test program in October 2015.

  15. Neutron physics of a high converting advanced pressurized water reactor

    International Nuclear Information System (INIS)

    Berger, H.D.

    1985-01-01

    The neutron physics of an APWR are analysed by single pin-cell calculations as well as two-dimensional whole-reactor computations. The calculational methods of the two codes employed for this study, viz. the cell code SPEKTRA and the diffusion-burnup code DIBU, are presented in detail. The APWR-investigations carried out concentrate on the void coefficient characteristics of tight UO 2 /PuO 2 -lattices, control rod worths, burnup behaviour and spatial power distributions in APWR cores. The principal physics design differences between advanced pressurized water reactors and present-day PWRs are identified and discussed. (orig./HP) [de

  16. Safety systems and features of boiling and pressurized water reactors

    International Nuclear Information System (INIS)

    Khair, H. O. M.

    2012-06-01

    The safe operation of nuclear power plants (NPP) requires a deep understanding of the functioning of physical processes and systems involved. This study was carried out to present an overview of the features of safety systems of boiling and pressurized water reactors that are available commercially. Brief description of purposes and functions of the various safety systems that are employed in these reactors was discussed and a brief comparison between the safety systems of BWRs and PWRs was made in an effort to emphasize of safety in NPPs.(Author)

  17. Computing radiation dose to reactor pressure vessel and internals

    International Nuclear Information System (INIS)

    1996-01-01

    Within the next twenty years many of the nuclear reactors currently in service will reach their design lifetime. One of the key factors affecting decisions on license extensions will be the ability to confidently predict the integrity of the reactor pressure vessel and core structural components which have been subjected to many years of cumulative radiation exposure. This report gives an overview of the most recent scientific literature and current methodologies for computational dosimetry in the OECD/NEA Member countries. Discussion is extended to consider some related issues of materials science, such as the metals, and limitations of the models in current use. Proposals are made for further work. (author)

  18. Modeling irradiation embrittlement in reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Odette, G.R.

    1998-01-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. In chapter 10, numerical modeling of irradiation embrittlement in reactor vessel steels are introduced. Physically-based models are developed and their role in advancing the state-of-the-art of predicting irradiation embrittlement of RPV steels is stressed

  19. State of the art of the advanced pressurized water reactor

    International Nuclear Information System (INIS)

    Seifritz, W.; Chawla, R.

    1987-01-01

    A review is given of the present status of the works concerned with an advanced pressurized water reactor (APWR). It includes the following items: reactor physics, thermal and hydraulic investigations and other engineering aspects as well as an analysis of electricity generation cost and long-term problems of embedding the APWR in a plutonium economy. As a summary it can be stated that there are discernible no principal obstacles of technically accomplishing an APWR, but there will be necessary considerable expenses in research and development works if it should be intended to start commercial service of an APWR up to the end of this century. (author)

  20. Rupture tests with reactor pressure vessel head models

    International Nuclear Information System (INIS)

    Talja, H.; Keinaenen, H.; Hosio, E.; Pankakoski, P.H.; Rahka, K.

    2003-01-01

    In the LISSAC project (LImit Strains in Severe ACcidents), partly funded by the EC Nuclear Fission and Safety Programme within the 5th Framework programme, an extensive experimental and computational research programme is conducted to study the stress state and size dependence of ultimate failure strains. The results are aimed especially to make the assessment of severe accident cases more realistic. For the experiments in the LISSAC project a block of material of the German Biblis C reactor pressure vessel was available. As part of the project, eight reactor pressure vessel head models from this material (22 NiMoCr 3 7) were tested up to rupture at VTT. The specimens were provided by Forschungszentrum Karlsruhe (FzK). These tests were performed under quasistatic pressure load at room temperature. Two specimens sizes were tested and in half of the tests the specimens contain holes describing the control rod penetrations of an actual reactor pressure vessel head. These specimens were equipped with an aluminium liner. All six tests with the smaller specimen size were conducted successfully. In the test with the large specimen with holes, the behaviour of the aluminium liner material proved to differ from those of the smaller ones. As a consequence the experiment ended at the failure of the liner. The specimen without holes yielded results that were in very good agreement with those from the small specimens. (author)

  1. Simulation of coolant mixing in pressure vessel reactors

    International Nuclear Information System (INIS)

    Hoehne, T.

    2003-06-01

    The work was aimed at the experimental investigation and numerical simulation of coolant mixing in the downcomer and the lower plenum of PWRs. Generally, the coolant mixing is of relevance for two classes of accident scenarios - boron dilution and cold water transients. For the investigation of the relevant mixing phenomena, the Rossendorf test facility ROCOM has been designed. ROCOM is a 1:5 scaled Plexiglas trademark model of the PWR Konvoi allowing conductivity measurements by wire mesh sensors and velocity measurements by the LDA technique. The CFD calculations were carried out with the CFD-code CFX-4. For the design of the facility, calculations were performed to analyze the scaling of the model. It was found, that the scaling of 1:5 to the prototype meets both: physical and economical demands. Flow measurements and the corresponding CFD calculations in the ROCOM downcomer under steady state conditions showed a Re number independency at nominal flow rates. The flow field is dominated by recirculation areas below the inlet nozzles. Transient flow measurements with high performance LDA-technique showed in agreement with CFX-4 results, that in the case of the start up of a pump after a laminar stage large vortices dominate the flow. In the case of stationary mixing, the maximum value of the averaged mixing scalar at the core inlet was found in the sector below the inlet nozzle, where the tracer was injected. At the start-up case of one pump due to a strong impulse driven flow at the inlet nozzle the horizontal part of the flow dominates in the downcomer. The injection is distributed into two main jets, the maximum of the tracer concentration at the core inlet appears at the opposite part of the loop where the tracer was injected. Additionally, the stationary three-dimensional flow distribution in the downcomer and the lower plenum of a VVER-440/V-230 reactor was calculated with CFX-4. The comparison with experimental data and an analytical mixing model showed a

  2. Superheated steam annealing of pressurized water reactor vessel

    International Nuclear Information System (INIS)

    Porowski, J.S.

    1993-01-01

    Thermal annealing of an embrittled Reactor Pressure Shell is the only recognized means for recovering material properties lost due to long-term exposure of the reactor walls to radiation. Reduced toughness of the material during operation is a major concern in evaluations of structural integrity of older reactors. Extensive studies performed within programs related to life extension of nuclear plants have confirmed that the thermal treatment of 850 deg. F for 168 hours on irradiated material essentially recovers material properties lost due to neutron exposure. Dry and wet annealing methods have been considered. Wet annealing involves operating the reactor at near design temperatures and pressures. Since the temperature of wet annealing must be limited to vessel design temperature of 650 deg. F, only partial recovery of the lost properties is achieved. Thus dry annealing was selected as an alternative for future development and industrial implementation to extend the safe life of reactors. Dry thermal annealing consists of heating portions of the reactor vessel at a specific temperature for a given period of time using a high temperature heat source. The use of spent fuel assemblies, induction heating and resistance heating elements as well as the circulation of heated fluid were investigated as potential candidate methods. To date the use of resistance heating elements which are lowered into a dry empty reactor was considered to be the preferred method. In-depth research in the United States and practical applications of such a method in Russia have confirmed feasibility of the method. The method of using circulating superheated steam to anneal the vessel at 850 deg. F without complete removal of the reactor internals is described herein. After removing the reactor head and fuel, the core barrel along with the upper and lower core in PWRs is lifted to open an annular space between the reactor shell flange and the core barrel flange. The thermal shield can remain

  3. 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

  4. Stability analysis of supercritical-pressure light water-cooled reactor in constant pressure operation

    International Nuclear Information System (INIS)

    Suhwan, JI; Shirahama, H.; Koshizuka, S.; Oka, Y.

    2001-01-01

    The purpose of this study is to evaluate the thermal-hydraulic and the thermal-nuclear coupled stabilities of a supercritical pressure light water-cooled reactor. A stability analysis code at supercritical pressure is developed. Using this code, stabilities of full and partial-power reactor operating at supercritical pressure are investigated by the frequency-domain analysis. Two types of SCRs are analyzed; a supercritical light water reactor (SCLWR) and a supercritical water-cooled fast reactor (SCFR). The same stability criteria as Boiling Water Reactor are applied. The thermal-hydraulic stability of SCLWR and SCFR satisfies the criteria with a reasonable orifice loss coefficient. The decay ratio of the thermal-nuclear coupled stability in SCFR is almost zero because of a small coolant density coefficient of the fast reactor. The evaluated decay ratio of the thermal-nuclear coupled stability is 3,41 ∼ 10 -V at 100% power in SCFR and 0,028 at 100% power in SCLWR. The sensitivity is investigated. It is found that the thermal-hydraulic stability is sensitive to the mass flow rate strongly and the thermal-nuclear coupled stability to the coolant density coefficient. The bottom power peak distribution makes the thermal-nuclear stability worse and the thermal-nuclear stability better. (author)

  5. N13 - based reactor coolant pressure boundary leakage system

    International Nuclear Information System (INIS)

    Dissing, E.; Marbaeck, L.; Sandell, S.; Svansson, L.

    1980-05-01

    A system for the monitoring of leakage of coolant from the reactor coolant pressure boundary and auxiliary systems to the reactor containment, based on the detection of the N13 content in the atmosphere, has been tested. N13 is produced from the oxyegen of the reactor water via the recoil photon nuclear process H1 + 016 + He4. The generation of N13 is therefore independent of fuel element leakage and of the corrosion product content in the water. In the US AEC regulatory guide 1.45 has a leakage increase of 4 liter/ min been suggested as the response limit. The experiments carried out in Ringhals indicate, that with the accomplishment of minor improvements in the installation, a 4 liter/min leakage to the containment will give rise to a signal with a random error range of +- 0.25 liter/min, 99.7 % confidence level. (author)

  6. 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

  7. 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.

  8. EDF's nuclear safety approach for pressurized water reactors

    International Nuclear Information System (INIS)

    Tanguy, P.; Kus, J.P.

    1987-01-01

    The realization of the important French program fifty-four units equipped with pressurized water reactors in service, or under construction-had led to the progressive implementation of an original approach in the field of nuclear safety. From an initial core consisting of the deterministic approach to safety devised on the other side of the Atlantic, which has been entirely preserved and often specified, further extras have been added which overall increase the level of safety of the installations, without any particular complications. This paper aims at presenting succinctly the outcome of the deliberation, which constitutes now the approach adopted by Electricite de France for the safety of nuclear units equipped with pressurized water reactors. This approach is explained in more detail in EDF's 'with book' on nuclear safety. (author)

  9. Needs for evaluated covariance data for reactor pressure vessel dosimetry

    International Nuclear Information System (INIS)

    Maerker, R.E.; Broadhead, B.L.; Wagschal, J.J.

    1992-01-01

    This report discusses new methodology for quantifying and then reducing uncertainties in the calculated pressure vessel fluences of a pressurized water reactor (PWR). The technique involves combining the integral results of the calculated and measured PWR surveillance dosimetry activities with the differential data used in the calculations, along with covariances of all the quantities, into a generalized linear least-squares adjustment procedure. Based on analysis of both PWRs and test reactor benchmarks, substantial evidence now exists to support the conclusion that, of all the nuclear as well as non-nuclear differential data considered, ENDF/B-VI values of the total inelastic iron cross sections and their covariances are the most important data controlling the outcome of the adjustment procedure. Predicted adjustments in these cross sections provided the stimulus for new measurements, the results of which impacted the ENDF/B-VI evaluation of iron 56

  10. Development of pressure boundaries leak detection technology for nuclear reactor

    International Nuclear Information System (INIS)

    Zhang Yao; Zhang Dafa; Chen Dengke; Zhang Liming

    2008-01-01

    The leak detection for the pressure boundaries is an important safeguard in nuclear reactor operation. In the paper, the status and the characters on the development of the pressure boundaries leak detection technology for the nuclear reactor were reviewed, especially, and the advance of the radiation leak detection technology and the acoustic emission leak detection technology were analyzed. The new advance trend of the leak detection technology was primarily explored. According to the analysis results, it is point out that the advancing target of the leak detection technology is to enhance its response speed, sensitivity, and reliability, and to provide effective information for operator and decision-maker. The realization of the global leak detection and the whole life cycle health monitoring for the nuclear boundaries is a significant advancing tendency of the leak detection technology. (authors)

  11. Sources of radioiodine at pressurized water reactors. Final report

    International Nuclear Information System (INIS)

    Pelletier, C.A.; Cline, J.E.; Barefoot, E.D.; Hemphill, R.T.; Voilleque, P.G.; Emel, W.A.

    1978-11-01

    The report determines specific components and operations at operating pressurized water reactors that have a potential for being significant emission sources of radioactive iodine. The relative magnitudes of these specific sources in terms of the chemical forms of the radioiodine and the resultant annual averages from major components are established. The data are generalized for broad industry use for predictive purposes. The conclusions of this study indicate that the majority of radioiodine emanating from the primary side of pressurized water reactors comes from a few major areas; in some cases these sources are locally treatable; the interaction of radioiodine with plant interior surfaces is an important phenomenon mediating the source and affecting its release to the atmosphere; the chemical form varies depending on the circumstances of the release

  12. Liquid radioactive waste processing system for pressurized water reactor plants

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    This Standard sets forth design, construction, and performance requirements, with due consideration for operation, of the Liquid Radioactive Waste Processing System for pressurized water reactor plants for design basis inputs. For the purpose of this Standard, the Liquid Radioactive Waste Processing System begins at the interfaces with the reactor coolant pressure boundary and the interface valve(s) in lines from other systems, or at those sumps and floor drains provided for liquid waste with the potential of containing radioactive material; and it terminates at the point of controlled discharge to the environment, at the point of interface with the waste solidification system, and at the point of recycle back to storage for reuse

  13. Allowable minimum upper shelf toughness for nuclear reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Zahoor, A.

    1988-05-01

    The paper develops methodology and procedure for determining the allowable minimum upper shelf toughness for continued safe operation of nuclear reactor pressure vessels. Elastic-plastic fracture mechanics analysis method based on the J-integral tearing modulus (J/T) approach is used. Closed from expressions for the applied J and tearing modulus are presented for finite length, part-throughwall axial flaw with aspect ratio of 1/6. Solutions are then presented for Section III, Appendix G flaw. A simple flaw evaluation procedure that can be applied quickly by utility engineers is presented. An attractive feature of the simple procedure is that tearing modulus calculations are not required by the user, and a solution for the slope of the applied J/T line is provided. Results for the allowable minimum upper shelf toughness are presented for a range of reactor pressure vessel thickness and heatup/cooldown rates.

  14. The coolability limits of a reactor pressure vessel lower head

    Energy Technology Data Exchange (ETDEWEB)

    Theofanous, T.G.; Syri, S. [Univ. of California, Santa Barbara, CA (United States)

    1995-09-01

    Configuration II of the ULPU experimental facility is described, and from a comprehensive set of experiments are provided. The facility affords full-scale simulations of the boiling crisis phenomenon on the hemispherical lower head of a reactor pressure vessel submerged in water, and heated internally. Whereas Configuration I experiments (published previously) established the lower limits of coolability under low submergence, pool-boiling conditions, with Configuration II we investigate coolability under conditions more appropriate to practical interest in severe accident management; that is, heat flux shapes (as functions of angular position) representative of a core melt contained by the lower head, full submergence of the reactor pressure vessel, and natural circulation. Critical heat fluxes as a function of the angular position on the lower head are reported and related the observed two-phase flow regimes.

  15. Allowable minimum upper shelf toughness for nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Zahoor, A.

    1988-01-01

    The paper develops methodology and procedure for determining the allowable minimum upper shelf toughness for continued safe operation of nuclear reactor pressure vessels. Elastic-plastic fracture mechanics analysis method based on the J-integral tearing modulus (J/T) approach is used. Closed from expressions for the applied J and tearing modulus are presented for finite length, part-throughwall axial flaw with aspect ratio of 1/6. Solutions are then presented for Section III, Appendix G flaw. A simple flaw evaluation procedure that can be applied quickly by utility engineers is presented. An attractive feature of the simple procedure is that tearing modulus calculations are not required by the user, and a solution for the slope of the applied J/T line is provided. Results for the allowable minimum upper shelf toughness are presented for a range of reactor pressure vessel thickness and heatup/cooldown rates. (orig.)

  16. EDF'S nuclear safety approach for pressurized water reactors

    International Nuclear Information System (INIS)

    Tanguy, P.; Kus, J.P.

    1988-01-01

    The realization of the important French program fifty-four units equipped with pressurized water reactors in service, or under construction - had led to the progressive implementation of an original approach in the field of nuclear safety. From an initial core consisting of the deterministic approach to safety devised on the other side of the Atlantic, which has been entirely preserved and often specified, further extras have been added which overall increase the level of safety of the installations, without any particular complications. This paper aims at presenting succinctly the outcome of the deliberation, which constitutes now the approach adopted by Electricite de France for the safety of nuclear units equipped with pressurized water reactors. This approach is explained in more detail in EDF's white book on nuclear safety

  17. Thermohydraulic feedbacks in self-pressurized reactor systems

    International Nuclear Information System (INIS)

    Fiebig, R.

    1977-01-01

    The impact on the dynamic behaviour of a self-pressurized reactor by the thermodynamic properties of the steam dome is investigated. For self-stabilization of the system the water of the primary circuit must be coupled thermodynamically to the steam in the steam dome, or alternatively the water in the reactor core must be subcooled sufficiently. Ways of thermodynamically coupling the water to the steam are heat conduction, boiling and condensation. A heat sink within the steam dome forces thermodynamic equilibrium between water and steam. This condition yields excellent self-control. Without heat sink thermal coupling is suspended at transients resulting in pressure rises. However, the reactor is still controlable as long as circuit and steam dome have direct contact. At the reactor of the NCS-80 a buffer volume of water separates primary circuit and steam volume. Stability is achieved by a heat sink in the steam dome and a shift of the core temperature into the subcooled domain effected by steam bubbles rising into the steam dome. (orig.) [de

  18. Thermohydraulic feedbacks in self-pressurized reactor systems

    International Nuclear Information System (INIS)

    Fiebig, R.

    1977-01-01

    The impact on the dynamic behaviour of a self-pressurized reactor by the thermodynamic properties of the steam dome is investigated. For self-stabilization of the system the water of the primary circuit must be coupled thermodynamically to the steam in the steam dome, or alternatively the water in the reactor core must be subcooled sufficiently. Ways of thermodynamically coupling the water to the steam are heat conduction, boiling and condensation. A heat sink within the steam dome forces thermodynamic equilibrium between water and steam. This condition yields excellent self-control. Without heat sink thermal coupling is suspended at transients resulting in pressure rises. However, the reactor is still controllable as long as circuit and steam dome have direct contact. At the reactor of the NCS-80 a buffer volume of water separates primary circuit and steam volume. Stability is achieved by a heat sink in the steam dome and a shift of the core temperature into the subcooled domain effected by steam bubbles rising into the steam dome. (orig.) [de

  19. Detection of steam generator tube leaks in pressurized water reactors

    International Nuclear Information System (INIS)

    Roach, W.H.

    1984-11-01

    This report addresses the early detection of small steam generator tube leaks in pressurized water reactors. It identifies physical parameters, establishes instrumentation performance goals, and specifies sensor types and locations. It presents a simple algorithm that yields the leak rate as a function of known or measurable quantities. Leak rates of less than one-tenth gram per second should be detectable with existing instrumentation

  20. Decommissioning of the BR3 pressurized-water reactor

    International Nuclear Information System (INIS)

    Massaut, V.

    1996-01-01

    The dismantling and the decommissioning of nuclear installations at the end of their life-cycle is a new challenge to the nuclear industry. Different techniques and procedures for the dismantling of a nuclear power plant on an existing installation, the BR-3 pressurized-water reactor, are described. The scientific programme, objectives, achievements in this research area at the Belgian Nuclear Research Centre SCK-CEN for 1995 are summarized

  1. Welding in repair of nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Pilous, V.; Kovarik, R.

    1987-01-01

    Specific welding conditions are described in repair of the pressure vessels of nuclear reactors in operation and the effect is pointed out to of neutrons on changes in steel properties. Some of the special regulations are discussed to be observed in welding jobs. The welding methods are briefly described; the half-bead method is most frequently used. It is stressed that the defect must first be identified using a nondestructive method and the stages must be defined of the welding repair of the pressure vessel. (J.B.). 4 figs., 1 tab., 16 refs

  2. Annealing of the BR3 reactor pressure vessel

    International Nuclear Information System (INIS)

    Fabry, A.; Motte, F.; Stiennon, G.; Debrue, J.; Gubel, P.; Van de Velde, J.; Minsart, G.; Van Asbroeck, P.

    1985-01-01

    The pressure vessel of the Belgian BR-3 plant, a small (11 MWe) PWR presently used for fuel testing programs and operated since 1962, was annealed during March, 1984. The anneal was performed under wet conditions for 168 hours at 650 0 F with core removal and within plant design margins justification for the anneal, summary of plant characteristics, description of materials sampling, summary of reactor physics and dosimetry, development of embrittlement trend curves, hypothesized pressurized and overcooling thermal shock accidents, and conclusions are provided in detail

  3. Prestressed concrete pressure vessels for boiling water reactors

    International Nuclear Information System (INIS)

    Menon, S.

    1979-12-01

    Following a general description of the Scandinavian cooperative project on prestressed concrete pressure vessels for boiling water reactors, detailed discussion is given in four appendices of the following aspects: the verification programme of tests and studies, the development and testing of a liner venting system, a preliminary safety philosophy and comparative assessment of cold and hot liners. Vessel failure probability is briefly discussed and some figures presented. The pressure gradients in the vessel wall resulting from various stipulated linear cracks, with a liner venting system are presented graphically. (JIW)

  4. Steam generator for a pressurized-water coolant nuclear reactor

    International Nuclear Information System (INIS)

    Schroeder, H.J.; Berger, W.

    1975-01-01

    A description is given of a steam generator which has a vertical cylindrical housing having a steam output outlet, a horizontal tube sheet closing the lower end of this housing, and an inverted U-shaped tube bundle inside of the housing and having vertical inlet and outlet legs with their ends mounted in the tube sheet. Beneath the tube sheet there are inlet and outlet manifolds for the respective ends of the tube bundle so that pressurized-water coolant from a pressurized-water coolant nuclear reactor can be circulated through the tube bundle

  5. 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)

  6. 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

  7. A dual pressurized water reactor producing 2000 MWe

    International Nuclear Information System (INIS)

    Kang, K. M.; Suh, K. Y.

    2010-01-01

    The Dual Unit Optimizer 2000 MWe (DUO2000) is proposed as a new design concept for large nuclear power plant. DUO is being designed to meet economic and safety challenges facing the 21. century green and sustainable energy industry. DUO2000 has two nuclear steam supply systems (NSSSs) of the Unit Nuclear Optimizer (UNO) pressurized water reactor (PWR) in a single containment so as to double the capacity of the plant. UNO is anchored to the Optimized Power Reactor 1000 MWe (OPR1000). The concept of DUO can be extended to any number of PWRs or pressurized heavy water reactors (PHWRs), or even boiling water reactor (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. In particular, since it is required that the Small and Medium sized Reactors (SMRs) be built as units, the concept of DUO2000 will apply to SMRs as well. With its in-vessel retention external reactor vessel cooling (IVR-ERVC) as severe accident management strategy, DUO can not only put the single most querulous PWR safety issue to end, but also pave ways to most promising large power capacity dispensing with huge redesigning cost for Generation III+ nuclear systems. Also, the strengths of DUO2000 include reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS. Two prototypes are presented for the DUO2000, and their respective advantages and drawbacks are considered. The strengths include, but are not necessarily limited to, reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS, just to name the few. The Coolant Unit Branching Apparatus (CUBA) is proposed

  8. Dual shell pressure balanced reactor vessel. Final project report

    International Nuclear Information System (INIS)

    Robertus, R.J.; Fassbender, A.G.

    1994-10-01

    The Department of Energy's Office of Energy Research (OER) has previously provided support for the development of several chemical processes, including supercritical water oxidation, liquefaction, and aqueous hazardous waste destruction, where chemical and phase transformations are conducted at high pressure and temperature. These and many other commercial processes require a pressure vessel capable of operating in a corrosive environment where safety and economy are important requirements. Pacific Northwest Laboratory (PNL) engineers have recently developed and patented (U.S. patent 5,167,930 December 1, 1992) a concept for a novel Dual Shell Pressure Balanced Vessel (DSPBV) which could solve a number of these problems. The technology could be immediately useful in continuing commercialization of an R ampersand D 100 award-winning technology, Sludge-to-oil Reactor System (STORS), originally developed through funding by OER. Innotek Corporation is a small business that would be one logical end-user of the DSPBV reactor technology. Innotek is working with several major U.S. engineering firms to evaluate the potential of this technology in the disposal of wastes from sewage treatment plants. PNL entered into a CRADA with Innotek to build a bench-scale demonstration reactor and test the system to advance the economic feasibility of a variety of high pressure chemical processes. Hydrothermal processing of corrosive substances on a large scale can now be made significantly safer and more economical through use of the DSPBV. Hydrothermal chemical reactions such as wet-air oxidation and supercritical water oxidation occur in a highly corrosive environment inside a pressure vessel. Average corrosion rates from 23 to 80 miles per year have been reported by Rice (1994) and Latanision (1993)

  9. Application-specific integrated circuit design for a typical pressurized water reactor pressure channel trip

    International Nuclear Information System (INIS)

    Battle, R.E.; Manges, W.W.; Emery, M.S.; Vendermolen, R.I.; Bhatt, S.

    1994-01-01

    This article discusses the use of application-specific integrated circuits (ASICs) in nuclear plant safety systems. ASICs have certain advantages over software-based systems because they can be simple enough to be thoroughly tested, and they can be tailored to replace existing equipment. An architecture to replace a pressurized water reactor pressure channel trip is presented. Methods of implementing digital algorithms are also discussed

  10. Heat treatment device for extending the life of a pressure vessel, particularly a reactor pressure vessel

    International Nuclear Information System (INIS)

    Krauss, P.; Mueller, E.; Poerner, H.; Weber, R.

    1979-01-01

    A support body in the form of an insulating cylinder is tightly sealed by connected surfaces at its outer circumference to the inner wall of the pressure vessel. It forms an annular heating space. The heat treatment or tempering of the pressure vessel takes place with the reactor space empty and screened from the outside by ceiling bolts. Heating gas or an induction winding can be used as the means of heating. (DG) [de

  11. BWR [boiling-water reactor] and PWR [pressurized-water reactor] off-normal event descriptions

    International Nuclear Information System (INIS)

    1987-11-01

    This document chronicles a total of 87 reactor event descriptions for use by operator licensing examiners in the construction of simulator scenarios. Events are organized into four categories: (1) boiling-water reactor abnormal events; (2) boiling-water reactor emergency events; (3) pressurized-water reactor abnormal events; and (4) pressurized-water reactor emergency events. Each event described includes a cover sheet and a progression of operator actions flow chart. The cover sheet contains the following general information: initial plant state, sequence initiator, important plant parameters, major plant systems affected, tolerance ranges, final plant state, and competencies tested. The progression of operator actions flow chart depicts, in a flow chart manner, the representative sequence(s) of expected immediate and subsequent candidate actions, including communications, that can be observed during the event. These descriptions are intended to provide examiners with a reliable, performance-based source of information from which to design simulator scenarios that will provide a valid test of the candidates' ability to safely and competently perform all licensed duties and responsibilities

  12. RISKAUDIT Report no. 7, Vol. 1: Safety evaluation of VVER 440/213 and VVER 1000/320 reactors in Rovno NPP Units 1, 2 and 3. Final Report by AIB-Vincotte Nuclear, CIEMAT, ANPA, GRS, IPSN, AEA-T

    International Nuclear Information System (INIS)

    1994-07-01

    The Riskaudit 7 report has been made by a group of independent experts from Western European countries representing technical organizations specialized in the support of regulatory bodies of those countries. It represents a preliminary estimation of the Ukrainian WWER B 213 and B 320 reactors, based on the example of Rovno NPP, analysed with a Western practice. The first part of the document covers the following aspects of the report: core design and fuel management; pressurized components; electrical supply; instrumentation and control; containment; internal events; site conditions and external events

  13. RISKAUDIT Report no. 7, Vol. 1: Safety evaluation of VVER 440/213 and VVER 1000/320 reactors in Rovno NPP Units 1, 2 and 3. Final Report by AIB-Vincotte Nuclear, CIEMAT, ANPA, GRS, IPSN, AEA-T

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-07-15

    The Riskaudit 7 report has been made by a group of independent experts from Western European countries representing technical organizations specialized in the support of regulatory bodies of those countries. It represents a preliminary estimation of the Ukrainian WWER B 213 and B 320 reactors, based on the example of Rovno NPP, analysed with a Western practice. The first part of the document covers the following aspects of the report: core design and fuel management; pressurized components; electrical supply; instrumentation and control; containment; internal events; site conditions and external events.

  14. Structural analysis of fuel rod applied to pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Faria, Danilo P.; Pinheiro, Andre Ricardo M.; Lotto, André A., E-mail: danilo.pinheiro@marinha.mil.br [Centro Tecnológico da Marinha em São Paulo (CTMSP), São Paulo, SP (Brazil)

    2017-07-01

    The design of fuel assemblies applied to Pressurized Water Reactors (PWR) has several requirements and acceptance criteria that must be attended for licensing. In the case of PWR fuel rods, an important mechanical structural requirement is to keep the radial stability when submitted to the coolant external pressure. In the framework of the Accident Tolerant Fuel (ATF) program new materials have been studied to replace zirconium based alloys as cladding, including iron-based alloys. In this sense, efforts have been made to evaluate the behavior of these materials under PWR conditions. The present work aims to evaluate the collapse cold pressure of a stainless steel thin-walled tube similar to that used as cladding material of fuel rods by means of the comparison of numeric data, and experimental results. As a result of the simulations, it was observed that the collapse pressure has a value intermediate value between those found by regulatory requirements and analytical calculations. The experiment was carried out for the validation of the computational model using test specimens of thin-walled tubes considering empty tube. The test specimens were sealed at both ends by means of welding. They were subjected to a high pressure device until the collapse of the tubes. Preliminary results obtained from experiments with the empty test specimens indicate that the computational model can be validated for stainless steel cladding, considering the difference between collapse pressure indicated in the regulatory document and the actual limit pressure concerning to radial instability of tubes with the studied characteristics. (author)

  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. Aging study of boiling water reactor high pressure injection systems

    International Nuclear Information System (INIS)

    Conley, D.A.; Edson, J.L.; Fineman, C.F.

    1995-03-01

    The purpose of high pressure injection systems is to maintain an adequate coolant level in reactor pressure vessels, so that the fuel cladding temperature does not exceed 1,200 degrees C (2,200 degrees F), and to permit plant shutdown during a variety of design basis loss-of-coolant accidents. This report presents the results of a study on aging performed for high pressure injection systems of boiling water reactor plants in the United States. The purpose of the study was to identify and evaluate the effects of aging and the effectiveness of testing and maintenance in detecting and mitigating aging degradation. Guidelines from the United States Nuclear Regulatory Commission's Nuclear Plant Aging Research Program were used in performing the aging study. Review and analysis of the failures reported in databases such as Nuclear Power Experience, Licensee Event Reports, and the Nuclear Plant Reliability Data System, along with plant-specific maintenance records databases, are included in this report to provide the information required to identify aging stressors, failure modes, and failure causes. Several probabilistic risk assessments were reviewed to identify risk-significant components in high pressure injection systems. Testing, maintenance, specific safety issues, and codes and standards are also discussed

  17. Analysis of aging mechanism and management for HTR-PM reactor pressure vessel

    International Nuclear Information System (INIS)

    Sun Yunxue; Shao Jin

    2015-01-01

    Reactor pressure vessel is an important part of the reactor pressure boundary, its important degree ranks high in ageing management and life assessment of nuclear power plant. Carrying out systematic aging management to ensure reactor pressure vessel keeping enough safety margins and executing design functions is one of the key factors to guarantee security and stability operation for nuclear power plant during the whole lifetime and prolong life. This paper briefly introduces the structure and aging mechanism of reactor pressure vessel in pressurized water reactor nuclear power plant, and introduces the design principle and structure characteristics of HTR-PM. At the same time, this paper carries out preliminary analysis and exploration. and discusses aging management of HTR-PM reactor pressure vessel. Finally, the advice of carring out aging management for HTR-PM reactor pressure vessel is proposed. (authors)

  18. Probabilistic integrity assessment of pressure tubes in an operating pressurized heavy water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Young-Jin; Park, Heung-Bae [KEPCO E and C, 188 Gumi-dong, Bundang-gu, Seongnam-si, Gyeonggi-do 463-870 (Korea, Republic of); Lee, Jung-Min; Kim, Young-Jin [School of Mechanical Engineering, Sungkyunkwan University, 300 Chunchun-dong, Jangan-gu, Suwon-si, Gyeonggi-do 440-746 (Korea, Republic of); Ko, Han-Ok [Korea Institute of Nuclear Safety, 34 Gwahak-ro, Yuseong-gu, Daejeon-si 305-338 (Korea, Republic of); Chang, Yoon-Suk, E-mail: yschang@khu.ac.kr [Department of Nuclear Engineering, Kyung Hee University, 1 Seocheon-dong, Giheung-gu, Yongin-si, Gyeonggi-do 446-701 (Korea, Republic of)

    2012-02-15

    Even though pressure tubes are major components of a pressurized heavy water reactor (PHWR), only small proportions of pressure tubes are sampled for inspection due to limited inspection time and costs. Since the inspection scope and integrity evaluation have been treated by using a deterministic approach in general, a set of conservative data was used instead of all known information related to in-service degradation mechanisms because of inherent uncertainties in the examination. Recently, in order that pressure tube degradations identified in a sample of inspected pressure tubes are taken into account to address the balance of the uninspected ones in the reactor core, a probabilistic approach has been introduced. In the present paper, probabilistic integrity assessments of PHWR pressure tubes were carried out based on accumulated operating experiences and enhanced technology. Parametric analyses on key variables were conducted, which were periodically measured by in-service inspection program, such as deuterium uptake rate, dimensional change rate of pressure tube and flaw size distribution. Subsequently, a methodology to decide optimum statistical distribution by using a robust method adopting a genetic algorithm was proposed and applied to the most influential variable to verify the reliability of the proposed method. Finally, pros and cons of the alternative distributions comparing with corresponding ones derived from the traditional method as well as technical findings from the statistical assessment were discussed to show applicability to the probabilistic assessment of pressure tubes.

  19. Extended fuel cycle operation for pressurized water reactor plants

    International Nuclear Information System (INIS)

    Silvestri, G.J. Jr.

    1978-01-01

    A nuclear steam turbine power plant system having an arrangement therein for extended fuel cycle operation is described. The power plant includes a turbine connected at its inlet to a source of motive fluid having a predetermined pressure associated therewith. The turbine has also connected thereto an extraction conduit which extracts steam from a predetermined location therein for use in an associated apparatus. A bypass conduit is provided between a point upstream of the inlet and the extraction conduit. A flow control device is provided within the bypass conduit and opens when the pressure of the motive steam supply drops beneath the predetermined pressure as a result of reactivity loss within the nuclear reactor. Opening of the bypass conduit provides flow to the associated apparatus and at the same time provides an increased flow orifice to maintain fluid flow rate at a predetermined level

  20. Automatic power control for a pressurized water reactor

    International Nuclear Information System (INIS)

    Hah, Yung Joon

    1994-02-01

    During a normal operation of a pressurized water reactor (PWR), the reactivity is controlled by control rods, boron, and the average temperature of the primary coolant. Especially in load follow operation, the reactivity change is induced by changes in power level and effects of xenon concentration. The control of the core power distribution is concerned, mainly, with the axial power distribution which depends on insertion and withdrawal of the control rods resulting in additional reactivity compensation. The utilization of part strength control element assemblies (PSCEAs) is quite appropriate for a control of the power distribution in the case of Yonggwang Nuclear Unit 3 (YGN Unit 3). However, control of the PSCEAs is not automatic, and changes in the boron concentration by dilution/boration are done manually. Thus, manual control of the PSCEAs and the boron concentration require the operator's experience and knowledge for a successful load follow operation. In this thesis, the new concepts have been proposed to adapt for an automatic power control in a PWR. One of the new concepts is the mode K control, another is a fuzzy power control. The system in mode K control implements a heavy-worth bank dedicated to axial shape control, independent of the existing regulating banks. The heavy bank provides a monotonic relationship between its motion and the axial power shape change, which allows automatic control of the axial power distribution. And the mode K enables precise regulation, by using double closed-loop control of the reactor coolant temperature and the axial power difference. Automatic reactor power control permits the nuclear power plant to accommodate the load follow operations, including frequency control, to respond to the grid requirements. The mode K reactor control concepts were tested using simulation responses of a Korean standardized 1000-MWe PWR which is a reference plant for the YGN Unit 3. The simulation results illustrate that the mode K would be

  1. 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)

  2. Performance demonstration experience for reactor pressure vessel shell ultrasonic testing

    International Nuclear Information System (INIS)

    Zado, V.

    1998-01-01

    The most ultrasonic testing techniques used by many vendors for pressurized water reactor (PWR) examinations were based on American Society of Mechanical Engineers 'Boiler and Pressurized Vessel Code' (ASME B and PV Code) Sections XI and V. The Addenda of ASME B and PV Code Section XI, Edition 1989 introduced Appendix VIII - 'Performance Demonstration for Ultrasonic Examination Systems'. In an effort to increase confidence in performance of ultrasonic testing of the operating nuclear power plants in United States, the ultrasonic testing performance demonstration examination of reactor vessel welds is performed in accordance with Performance Demonstration Initiative (PDI) program which is based on ASME Code Section XI, Appendix VIII requirements. This article provides information regarding extensive qualification preparation works performed prior EPRI guided performance demonstration exam of reactor vessel shell welds accomplished in January 1997 for the scope of Appendix VIII, Supplements IV and VI. Additionally, an overview of the procedures based on requirements of ASME Code Section XI and V in comparison to procedure prepared for Appendix VIII examination is given and discussed. The samples of ultrasonic signals obtained from artificial flaws implanted in vessel material are presented and results of ultrasonic testing are compared to actual flaw sizes. (author)

  3. Stresses in reactor pressure vessel nozzles -- Calculations and experiments

    International Nuclear Information System (INIS)

    Brumovsky, M.; Polachova, H.

    1995-01-01

    Reactor pressure vessel nozzles are characterized by a high stress concentration which is critical in their low-cycle fatigue assessment. Program of experimental verification of stress/strain field distribution during elastic-plastic loading of a reactor pressure vessel WWER-1000 primary nozzle model in scale 1:3 is presented. While primary nozzle has an ID equal to 850 mm, the model nozzle has ID equal to 280 mm, and was made from 15Kh2NMFA type of steel. Calculation using analytical methods was performed. Comparison of results using different analytical methods -- Neuber's, Hardrath-Ohman's as well as equivalent energy ones, used in different reactor Codes -- is shown. Experimental verification was carried out on model nozzles loaded statically as well as by repeated loading, both in elastic-plastic region. Strain fields were measured using high-strain gauges, which were located in different distances from center of nozzle radius, thus different stress concentration values were reached. Comparison of calculated and experimental data are shown and compared

  4. In-place thermal annealing of nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Server, W.L.

    1985-04-01

    Radiation embrittlement of ferritic pressure vessel steels increases the ductile-brittle transition temperature and decreases the upper shelf level of toughness as measured by Charpy impact tests. A thermal anneal cycle well above the normal operating temperature of the vessel can restore most of the original Charpy V-notch energy properties. The Amry SM-1A test reactor vessel was wet annealed in 1967 at less than 343 0 C (650 0 F), and wet annealing of the Belgian BR-3 reactor vessel at 343 0 C (650 0 F) has recently taken place. An industry survey indicates that dry annealing a reactor vessel in-place at temperatures as high as 454 0 C (850 0 F) is feasible, but solvable engineering problems do exist. Economic considerations have not been totally evaluated in assessing the cost-effectiveness of in-place annealing of commercial nuclear vessels. An American Society for Testing and Materials (ASTM) task group is upgrading and revising guide ASTM E 509-74 with emphasis on the materials and surveillance aspects of annealing rather than system engineering problems. System safety issues are the province of organizations other than ASTM (e.g., the American Society of Mechanical Engineers Boiler and Pressure Vessel Code body)

  5. Detection of steam generator tube leaks in pressurized water reactors

    International Nuclear Information System (INIS)

    Roach, W.H.

    1985-01-01

    This report addresses the early detection of small steam generator tube leaks in pressurized water reactors. It discusses the third, and final, year's work on an NRC-funded project examining diagnostic instrumentation in water reactors. The first two years were broad in coverage, concentrating on anticipatory measurements for detection of potential problems in both pressurized- and boiling-water reactors, with recommendations for areas of further study. One of these areas, the early detection of small steam tube leaks in PWRs, formed the basis of study for the last year of the project. Four tasks are addressed in this study of the detection of steam tube leaks. (1) Determination of which physical parameters indicate the onset of steam generator tube leaks. (2) Establishing performance goals for diagnostic instruments which could be used for early detection of steam generator tube leaks. (3) Defining the diagnostic instrumentation and their location which satisfy Items 1 and 2 above. (4) Assessing the need for diagnostic data processing and display. Parameters are identified, performance goals established, and sensor types and locations are specified in the report, with emphasis on the use of existing instrumentation with a minimum of retrofitting. A simple algorithm is developed which yields the leak rate as a function of known or measurable quantities. The conclusion is that leak rates of less than one-tenth gram per second should be detectable with existing instrumentation. (orig./HP)

  6. Dynamic loads on reactor vessel components by low pressure waves

    International Nuclear Information System (INIS)

    Benkert, J.; Mika, C.; Stegemann, D.; Valero, M.

    1978-01-01

    Starting from the conservation theorems for mass and impulses the code DRUWE has been developed enabling the calculation of dynamic loads of the reactor shell on the basis of simplified assumptions for the first period shortly after rupture. According to the RSK-guidelines it can be assumed that the whole weld size is opened within 15 msec. This time-dependent opening of the fractured plane can be taken into account in the computer program. The calculation is composed in a way that for a reactor shell devided into cross and angle sections the local, chronological pressure and strength curves, the total dynamic load as well as the moments acting on the fastenings of the reactor shell can be calculated. As input data only geometrical details concerning the concept of the pressure vessel and its components as well as the effective subcooling of the fluid are needed. By means of several parameters the program can be operated in a way that the results are available in form of listings or diagrams, respectively, but also as card pile for further examinations, e.g. strength analysis. (orig./RW) [de

  7. Thermal shield support degradation in pressurized water reactors

    International Nuclear Information System (INIS)

    Sweeney, F.J.; Fry, D.N.

    1986-01-01

    Damage to the thermal shield support structures of three pressurized water reactors (PWRs) due to flow-induced vibrations was recently discovered during refueling. In two of the reactors, severe damage occurred to the thermal shield, and in one reactor the core support barrel (CSB) was damaged, necessitating extended outages for repairs. In all three reactors, several of the thermal shield supports were either loose, damaged, or missing. The three plants had been in operation for approximately 10 years before the damage was apparent by visual inspection. Because each of the three US PWR manufacturers have experienced thermal shield support degradation, the Nuclear Regulatory Commission requested that Oak Ridge National Laboratory analyze ex-core neutron detector noise data to determine the feasibility of detecting incipient thermal shield support degradation. Results of the noise data analysis indicate that thermal shield support degradation probably began early in the life of both severely damaged plants. The degradation was characterized by shifts in the resonant frequencies of core internal structures and the appearance of new resonances in the ex-core neutron detector noise. Both the data analyses and the finite element calculations indicate that these changes in resonant frequencies are less than 3 Hz. 11 refs., 16 figs

  8. Current state of research on pressurized water reactor safety

    International Nuclear Information System (INIS)

    Couturier, Jean; Schwarz, Michel; Roubaud, Sebastien; Lavarenne, Caroline; Mattei, Jean-Marie; Rigollet, Laurence; Scotti, Oona; Clement, Christophe; Lancieri, Maria; Gelis, Celine; Jacquemain, Didier; Bentaib, Ahmed; Nahas, Georges; Tarallo, Francois; Guilhem, Gilbert; Cattiaux, Gerard; Durville, Benoit; Mun, Christian; Delaval, Christine; Sollier, Thierry; Stelmaszyk, Jean-Marc; Jeffroy, Francois; Dechy, Nicolas; Chanton, Olivier; Tasset, Daniel; Pichancourt, Isabelle; Barre, Francois; Bruna, Gianni; Evrard, Jean-Michel; Gonzalez, Richard; Loiseau, Olivier; Queniart, Daniel; Vola, Didier; Goue, Georges; Lefevre, Odile

    2018-03-01

    For more than 40 years, IPSN then IRSN has conducted research and development on nuclear safety, specifically concerning pressurized water reactors, which are the reactor type used in France. This publication reports on the progress of this research and development in each area of study - loss-of-coolant accidents, core melt accidents, fires and external hazards, component aging, etc. -, the remaining uncertainties and, in some cases, new measures that should be developed to consolidate the safety of today's reactors and also those of tomorrow. A chapter of this report is also devoted to research into human and organizational factors, and the human and social sciences more generally. All of the work is reviewed in the light of the safety issues raised by feedback from major accidents such as Chernobyl and Fukushima Daiichi, as well as the issues raised by assessments conducted, for example, as part of the ten-year reviews of safety at French nuclear reactors. Finally, through the subjects it discusses, this report illustrates the many partnerships and exchanges forged by IRSN with public, industrial and academic bodies both within Europe and internationally

  9. Consequence evaluation of hypothetical reactor pressure vessel support failure

    International Nuclear Information System (INIS)

    Lu, S.C.; Holman, G.S.; Lambert, H.E.

    1991-01-01

    This paper describes a consequence evaluation to address safety concerns raised by the radiation embrittlement of the reactor pressure vessel (RPV) supports for the Trojan nuclear power plant. The study comprises a structural evaluation and an effects evaluation and assumes that all four reactor vessel supports have completely lost the load carrying capability. The structural evaluation concludes that the Trojan reactor coolant loop (RCL) piping is capable of transferring loads to the steam generator (SG) supports and the reactor coolant pump (RCP) supports and that the SG supports and the RCP supports have sufficient design margins to accommodate additional loads transferred to them through the RCL piping. The effects evaluation, employing a systems analysis approach, investigates initiating events and the reliability of the engineered safeguard systems as the RPV is subject to movements caused by the RPV support failure. The evaluation identifies a number of areas for further investigation and concludes that a hypothetical failure of the Trojan RPV supports due to radiation embrittlement will not result in consequences of significant safety concerns. (author)

  10. 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

  11. Behavior of a corium jet in high pressure melt ejection from a reactor pressure vessel

    International Nuclear Information System (INIS)

    Frid, W.

    1988-04-01

    Discharge of the molten core debris from a pressurized reactor vessel has been recognized as an important accident scenario for pressurized water reactors. Recent high-pressure melt streaming experiments conducted at Sandia National Laboratories, designed to study cavity and containment events related to melt ejection, have resulted in two important observations: (1) Expansion and breakup of the ejected molten jet. (2) Significant aerosol generation during the ejection process. The expansion and breakup of the jet in the experiments are attributed to rapid evolution of the pressurizing gas (nitrogen or hydrogen) dissolved in the melt. It has been concluded that aerosol particles may be formed by condensation of melt vapor and mechanical breakup of the melt and generation. It was also shown that the above stated phenomena are likely to occur in reactor accidents. This report provides results from analytical and experimental investigations on the behavior of a gas supersaturated molten jet expelled from a pressurized vessel. Aero-hydrodynamic stability of liquid jets in gas, stream degassing of molten metals, and gas bubble nucleation in molten metals are relevant problems that are addressed in this work

  12. Contemporary pressurized water reactor technology in the world

    International Nuclear Information System (INIS)

    Komarek, A.

    1991-01-01

    The recent political events enabled Czechoslovak industrial companies to come into direct contact with leading western companies involved in pressurized water ractor technology. A survey is presented of the present situation at the world market of PWR type nuclear power plant suppliers and suppliers of fuel cycle services. Information is given on the potential bids for the next Czechoslovak nuclear power plants with PWR reactors. Economic aspects of the potential bids are presented including some considerations about the participation of the Czechoslovak nuclear industry as a supplier of the reactor for the future power plants. Main technical parameters are listed of PWR units with an output about 1000 MW supplied by Westinghouse EC, ABB -Combustion Engineering and Siemens AG. At present, the bids for new Czechoslovak nuclear power plants are being evaluated. No information on terms of the bids actually coming from foreign companies is used in the article. (Z.S.). 9 figs., 5 tabs

  13. On-line fatigue monitoring system for reactor pressure vessel

    International Nuclear Information System (INIS)

    Tokunaga, K.; Sakai, A.; Aoki, T.; Ranganath, S.; Stevens, G.L.

    1994-01-01

    A workstation-based, on-line fatigue monitoring system for tracking fatigue usage applied to an operating boiling water reactor (BWR), Tsuruga Unit-1, is described. The system uses the influence function approach and determines component stresses using temperature, pressure, and flow rate data that are made available via signal taps from previously existing plant sensors. Using plant unique influence functions developed specifically for the feedwater nozzle location, the system calculates stresses as a function of time and computed fatigue usage. The analysis method used to compute fatigue usage complies with MITI Code Notification No.501. Fatigue usage results for an entire fuel cycle are presented and compared to assumed design basis events to confirm that actual plant thermal duty is significantly less severe than originally estimated in the design basis stress report. As a result, the system provides the technical basis to more accurately evaluate actual reactor conditions as well as the justification for plant life extension. (author)

  14. Natural Circulation Characteristics of an Integral Pressurized Water Reactor

    International Nuclear Information System (INIS)

    Junli Gou; Suizheng Qiu; Guanghui Su; Dounan Jia

    2006-01-01

    Natural circulation potential is of great importance to the inherent safety of a nuclear reactor. This paper presents a theoretical investigation on the natural circulation characteristics of an integrated pressurized water reactor. Through numerically solved the one-dimensional model, the steady-state single phase conservative equations for the primary circuit and the steady-state two-phase drift-flux conservative equations for the secondary side of the once-through steam generator, the natural circulation characteristics are studied. Based on the preliminary calculation analysis, it is found that natural circulation mass flow rate is proportional to the exponential function of the power, and the value of the exponent is related to working conditions of the steam generator secondary side. The higher height difference between the core center and the steam generator center is favorable to the heat removal capacity of the natural circulation. (authors)

  15. Crystal Plasticity Model of Reactor Pressure Vessel Embrittlement in GRIZZLY

    Energy Technology Data Exchange (ETDEWEB)

    Chakraborty, Pritam [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Biner, Suleyman Bulent [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Zhang, Yongfeng [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Spencer, Benjamin Whiting [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

    2015-07-01

    The integrity of reactor pressure vessels (RPVs) is of utmost importance to ensure safe operation of nuclear reactors under extended lifetime. Microstructure-scale models at various length and time scales, coupled concurrently or through homogenization methods, can play a crucial role in understanding and quantifying irradiation-induced defect production, growth and their influence on mechanical behavior of RPV steels. A multi-scale approach, involving atomistic, meso- and engineering-scale models, is currently being pursued within the GRIZZLY project to understand and quantify irradiation-induced embrittlement of RPV steels. Within this framework, a dislocation-density based crystal plasticity model has been developed in GRIZZLY that captures the effect of irradiation-induced defects on the flow stress behavior and is presented in this report. The present formulation accounts for the interaction between self-interstitial loops and matrix dislocations. The model predictions have been validated with experiments and dislocation dynamics simulation.

  16. Crystal Plasticity Model of Reactor Pressure Vessel Embrittlement in GRIZZLY

    International Nuclear Information System (INIS)

    Chakraborty, Pritam; Biner, Suleyman Bulent; Zhang, Yongfeng; Spencer, Benjamin Whiting

    2015-01-01

    The integrity of reactor pressure vessels (RPVs) is of utmost importance to ensure safe operation of nuclear reactors under extended lifetime. Microstructure-scale models at various length and time scales, coupled concurrently or through homogenization methods, can play a crucial role in understanding and quantifying irradiation-induced defect production, growth and their influence on mechanical behavior of RPV steels. A multi-scale approach, involving atomistic, meso- and engineering-scale models, is currently being pursued within the GRIZZLY project to understand and quantify irradiation-induced embrittlement of RPV steels. Within this framework, a dislocation-density based crystal plasticity model has been developed in GRIZZLY that captures the effect of irradiation-induced defects on the flow stress behavior and is presented in this report. The present formulation accounts for the interaction between self-interstitial loops and matrix dislocations. The model predictions have been validated with experiments and dislocation dynamics simulation.

  17. Dominant accident sequences in Oconee-1 pressurized water reactor

    International Nuclear Information System (INIS)

    Dearing, J.F.; Henninger, R.J.; Nassersharif, B.

    1985-04-01

    A set of dominant accident sequences in the Oconee-1 pressurized water reactor was selected using probabilistic risk analysis methods. Because some accident scenarios were similar, a subset of four accident sequences was selected to be analyzed with the Transient Reactor Analysis Code (TRAC) to further our insights into similar types of accidents. The sequences selected were loss-of-feedwater, small-small break loss-of-coolant, loss-of-feedwater-initiated transient without scram, and interfacing systems loss-of-coolant accidents. The normal plant response and the impact of equipment availability and potential operator actions were also examined. Strategies were developed for operator actions not covered in existing emergency operator guidelines and were tested using TRAC simulations to evaluate their effectiveness in preventing core uncovery and maintaining core cooling

  18. Neurocontrol of Pressurized Water Reactors in Load-Follow Operations

    International Nuclear Information System (INIS)

    Lin Chaung; Shen Chihming

    2000-01-01

    The neurocontrol technique was applied to control a pressurized water reactor (PWR) in load-follow operations. Generalized learning or direct inverse control architecture was adopted in which the neural network was trained off-line to learn the inverse model of the PWR. Two neural network controllers were designed: One provided control rod position, which controlled the axial power distribution, and the other provided the change in boron concentration, which adjusted core total power. An additional feedback controller was designed so that power tracking capability was improved. The time duration between control actions was 15 min; thus, the xenon effect is limited and can be neglected. Therefore, the xenon concentration was not considered as a controller input variable, which simplified controller design. Center target strategy and minimum boron strategy were used to operate the reactor, and the simulation results demonstrated the effectiveness and performance of the proposed controller

  19. ASTM Standards for Reactor Dosimetry and Pressure Vessel Surveillance

    International Nuclear Information System (INIS)

    GRIFFIN, PATRICK J.

    1999-01-01

    The ASTM standards provide guidance and instruction on how to field and interpret reactor dosimetry. They provide a roadmap towards understanding the current ''state-of-the-art'' in reactor dosimetry, as reflected by the technical community. The consensus basis to the ASTM standards assures the user of an unbiased presentation of technical procedures and interpretations of the measurements. Some insight into the types of standards and the way in which they are organized can assist one in using them in an expeditious manner. Two example are presented to help orient new users to the breadth and interrelationship between the ASTM nuclear metrology standards. One example involves the testing of a new ''widget'' to verify the radiation hardness. The second example involves quantifying the radiation damage at a pressure vessel critical weld location through surveillance dosimetry and calculation

  20. Microstructural evolution in reactor pressure vessel steel under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Ohno, Katsumi; Fukuya, Koji [Institute of Nuclear Safety System Inc., Seika, Kyoto (Japan)

    2000-09-01

    Understanding microstructural changes in reactor pressure vessel steels is important in order to evaluate radiation-induced embrittlement, one of the major aging phenomena affecting the extension of plant life. In this study, actual surveillance test specimens and samples of rector vessel low-alloy steel (A533B steel) irradiated in a research reactor were examined using state-of-the-art techniques to clarify the neutron flux effect on the microstructural changes. These techniques included small angle neutron scattering and atom probes. Microstructural changes which are considered to be the main factors affecting embrittlement, including the production of copper-rich precipitates and the segregation of impurity elements, were confirmed by the results of the study. In addition, the mechanical properties were predicted based on the obtained quantitative data such as the diameters of precipitates. Consequently, the hardening due to irradiation was almost simulated. (author)