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Sample records for csfr czechoslovakian vver-440

  1. Safety of VVER-440 reactors

    CERN Document Server

    Slugen, Vladimir

    2011-01-01

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

  2. 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)

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

    International Nuclear Information System (INIS)

    Burclova, J.

    1992-01-01

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

  4. 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)

  5. Feasibility of VVER-440 type SFAT

    International Nuclear Information System (INIS)

    Kaartinen, J.; Tarvainen, M.

    1995-05-01

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

  6. 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

  7. 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)

  8. 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

  9. 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)

  10. 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

  11. 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

  12. 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

  13. 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)

  14. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-08-01

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

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

    International Nuclear Information System (INIS)

    Markova, Ludmila

    2003-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-11-15

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

  18. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-15

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

  20. 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.)

  1. 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.)

  2. 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.)

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

    International Nuclear Information System (INIS)

    Zajac, R.; Chrapciak, V.

    2010-01-01

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

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

    International Nuclear Information System (INIS)

    Zmitko, M.

    1999-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

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

  6. 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.)

  7. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-08-01

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

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    Andres, E.; Garcia Ruiz, R.

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-31

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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  14. 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.

  15. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-06-15

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

  17. 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)

  18. 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)

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

    International Nuclear Information System (INIS)

    Bartak, J.; Fallon, B.

    2006-01-01

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

  20. 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.

  1. 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

  2. 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

  3. 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

  4. 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)

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

    International Nuclear Information System (INIS)

    Tuunanen, J.

    1993-01-01

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

  6. 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.

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  8. 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

  9. 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.

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

    International Nuclear Information System (INIS)

    Kalinenko, V.

    1994-06-01

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

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

    International Nuclear Information System (INIS)

    Charbonneau, S.; Eckert, G.

    1996-01-01

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

  12. 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.

  13. 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.

  14. 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.

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

    International Nuclear Information System (INIS)

    Tuunanen, J.

    1994-01-01

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

  16. 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

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

    International Nuclear Information System (INIS)

    Husak, S.; Patrik, M.

    2000-01-01

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

  18. 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)

  19. 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)

  20. CFD evaluation of hydrogen risk mitigation measures in a VVER-440/213 containment

    Energy Technology Data Exchange (ETDEWEB)

    Heitsch, Matthias, E-mail: Matthias.Heitsch@ec.europa.e [Institute for Energy, Joint Research Centre, PO Box 2, 1755 ZG Petten (Netherlands); Huhtanen, Risto [VTT Technical Research Centre of Finland, PO Box 1000, FI-02044 VTT (Finland); Techy, Zsolt [VEIKI Institute for Electric Power Research Co., PO Box 80, H-1251 Budapest (Hungary); Fry, Chris [Serco, Winfrith Technology Centre, Dorchester, Dorset DT2 8DH (United Kingdom); Kostka, Pal [VEIKI Institute for Electric Power Research Co., PO Box 80, H-1251 Budapest (Hungary); Niemi, Jarto [VTT Technical Research Centre of Finland, PO Box 1000, FI-02044 VTT (Finland); Schramm, Berthold [Gesellschaft fuer Anlagen- und Reaktorsicherheit, GRS mbH, Schwertnergasse 1, 50667 Koeln (Germany)

    2010-02-15

    In the PHARE project 'Hydrogen Management for the VVER440/213' (HU2002/000-632-04-01), CFD (Computational Fluid Dynamics) calculations using GASFLOW, FLUENT and CFX were performed for the Paks NPP (Nuclear Power Plant), modelling a defined severe accident scenario which involves the release of hydrogen. The purpose of this work is to demonstrate that CFD codes can be used to model gas movement inside a containment during a severe accident. With growing experience in performing such analyses, the results encourage the use of CFD in assessing the risk of losing containment integrity as a result of hydrogen deflagrations. As an effective mitigation measure in such a situation, the implementation of catalytic recombiners is planned in the Paks NPP. In order to support these plans both unmitigated and recombiner-mitigated simulations were performed. These are described and selected results are compared. The codes CFX and FLUENT needed refinement to their models of wall and bulk steam condensation in order to be able to fully simulate the severe accident under consideration. Several CFD codes were used in parallel to model the same accident scenario in order to reduce uncertainties in the results. Previously it was considered impractical to use CFD codes to simulate a full containment subject to a severe accident extending over many hours. This was because of the expected prohibitive computing times and missing physical capabilities of the codes. This work demonstrates that, because of developments in the capabilities of CFD codes and improvements in computer power, these calculations have now become feasible.

  1. 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)

  2. 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

  3. Development of fuel cycles with new fuel with 8.9 mm external diameter for VVER-440. Preliminary assessment of operating efficiency

    Energy Technology Data Exchange (ETDEWEB)

    Gagarinskiy, Alexey [National Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

    2017-09-15

    Since the introduction of VVERs-440, their fuel assemblies are subject to ongoing improvements. Until now, the basic structural parameters of fuel, such as rod diameter of 9.1 mm, have never changed. This paper focuses on computational estimates of basic neutronic parameters of the fuel cycle that involves assemblies consisting of fuel rods with diameter reduced to 8.9 mm.

  4. 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

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

    International Nuclear Information System (INIS)

    Martin, M.

    2000-01-01

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

  6. Measures for ensuring hydrogen fire and explosion safety for VVER-440/230

    International Nuclear Information System (INIS)

    Bezlepkin, V.; Semashko, S.; Svetlov, S.; Sidorov, V.; Ivkov, I.; Krylov, Yu.; Kukhtevich, V.

    2004-01-01

    This paper deals with the findings of calculation analysis as regards the release of mass, energy and hydrogen during beyond-design-basis accident (BDBA) at Kola NPP equipped with VVER-440 reactor (B-230 design) and in respect of distribution of hydrogen throughout NPP tight compartments. The analysis figures out the number and locations of passive catalytic hydrogen recombiners and of the sensors of the hydrogen concentration monitoring system. In order to prove the hydrogen safety of the design, it has been necessary to review accidents accompanied by maximum emissions (both peak and integral ones) of hydrogen into the tight area. During design-basis accident (DBA), no steam/zirconium reactions occur in the reactor core. Out of BDBA, the severe accidents with damage to the core accompanied oxidative reactions between zirconium and steel with emission of hydrogen are regarded as the most dangerous ones. Assessment of additional hydrogen sources shows that the contribution of such sources to the total amount of hydrogen that may emit during a severe accident is insignificant. Calculations have been made for the following scenarios of severe accidents, which seem to be the most important in terms of hydrogen safety analysis: - 20 mm leak from the primary circuit in combination with a failure of the emergency makeup system; - 500 mm PCP rupture in the vicinity of reactor inlet branch with bi-lateral leakage of coolant. Releases of mass and energy during the aforesaid scenarios, changes of medium parameters within the tight compartments and analysis of possible fire conditions have been analyzed by means of Russian computer codes RATEG/SVECHA/HEFEST, KUPOL-M and LIMITS. The said analysis shows that the large break accident (500 mm), i.e. PCP rupture in the vicinity of the reactor branch with bi-lateral leakage of coolant is of the keen interest in terms of hydrogen safety. This accident typifies powerful short-term release of hydrogen at a significantly lesser

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

    International Nuclear Information System (INIS)

    Mikolas, P.

    2009-01-01

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

  8. Time versus frequency domain calculation of the main building complex of the VVER 440/213 NPP PAKS

    International Nuclear Information System (INIS)

    Katona, T.; Ratkai, S.; Halbritter, A.; Krutzik, N.J.; Schuetz, W.

    1995-01-01

    Various dynamic analyses were conducted for the main building complex of the VVER 440/213 PAKS in order to determine the dynamic response and assess the aseismic capacity of this nuclear power plant. Different types of mathematical models for idealizing the soil and the building structures were used. The main goal of the study presented here was to demonstrate the effects of different procedures for consideration of soil-structure interaction on the dynamic response of the structures mentioned above. The analyses were based on appropriate mathematical models of the coupled vibration structures (reactor building, turbine hall, intermediate building structures) and the layered soil. On the basis of this study, it can be concluded that substructure models using frequency-independent impedances and cut-off of modal damping usually provide conservative results. Complex models which allow the soil-soil and the structure or by frequency-dependent impedances) provide more accurate results. The latter approach results in more efficient designs which are not only safe but also economical. (author). 7 refs., 15 figs

  9. 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)

  10. 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)

  11. 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.

  12. 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

  13. 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.)

  14. Analysis of core damage frequency: Nuclear power plant Dukovany, VVER/440 V-213 Unit 1, internal events. Volume 1: Main report

    International Nuclear Information System (INIS)

    Pugila, W.J.

    1994-01-01

    This report presents the final results from the Level 1 probabilistic safety assessment (PSA) for the Dukovany VVER/440 V-213 nuclear power plant, Unit 1. Section 1.1 describes the objectives of this study. Section 1.2 discusses the approach that was used for completing the Dukovany PSA. Section 1.3 summarizes the results of the PSA. Section 1.4 provides a comparison of the results of the Dukovany PSA with the results of other PSAs for different types of reactors worldwide. Section 1.5 summarizes the conclusions of the Dukovany PSA

  15. 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.

  16. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-15

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

  18. The System of Higher Education in CSFR

    OpenAIRE

    Kopp, Botho von

    1991-01-01

    By dividing his article in two chapters ("1. From the founding of Charles University to the modern higher education system" and "2. The higher education system 1948-1989") the author gives an historical overview over the sytem of higher education in CSFR, whereas he covers the following aspects in the second chapter: "Basic data on higher education", "Organization and structure of the course of studies" and "Developments after 1989 and future trends". (DIPF/ ssch.)

  19. Analysis of noncondensable effect during small break transient in VVER-440 geometry with CATHARE V1.3L. Preliminary results

    International Nuclear Information System (INIS)

    Sarrette, C.

    1996-11-01

    The report presents a study of the transport and dissolution-release of non-condensable gas into the fluid of the primary loop for the VVER-440 geometry. The analysis has been done using a new model developed for the CATHARE thermal hydraulic code. Results are presented, obtained from calculations of small break loss-of-coolant (SBLOCA) accidents for the Loviisa nuclear power plant (NPP) geometry. The influence of nitrogen dissolved in the water of the accumulators of the emergency core coolant system (ECCS) on natural circulation is discussed. Possibilities of formation of nitrogen bubbles in the main vessels upper plenum, top of the downcomer, steam generators collectors, and upper structures of RCP's are investigated. First results show that there is potentiality for interruption, mainly due to the presence of nitrogen in the top of the downcomer and the upper parts of the RCP's. These preliminary results should be confirmed by carrying out calculations now prematurely stopped for numerical reasons. (8 refs.)

  20. Start-up of a cold loop in a VVER-440, the 7th AER benchmark calculation with HEXTRAN-SMABRE-PORFLO

    International Nuclear Information System (INIS)

    Hovi, Ville; Taivassalo, Veikko; Haemaelaeinen, Anitta; Raety, Hanna; Syrjaelahti, Elina

    2017-01-01

    The 7 th dynamic AER benchmark is the first in which three-dimensional thermal hydraulics codes are supposed to be applied. The aim is to get a more precise core inlet temperature profile than the sector temperatures available typically with system codes. The benchmark consists of a start-up of the sixth, isolated loop in a VVER-440 plant. The isolated loop initially contains cold water without boric acid and the start-up leads to a somewhat asymmetrical core power increase due to feedbacks in the core. In this study, the 7 th AER benchmark is calculated with the three-dimensional nodal reactor dynamics code HEXTRAN-SMABRE coupled with the porous computational fluid dynamics code PORFLO. These three codes are developed at VTT. A novel two-way coupled simulation of the 7 th AER benchmark was performed successfully demonstrating the feasibility and advantages of the new reactor analysis framework. The modelling issues for this benchmark are reported and some evaluation against the previously reported comparisons between the system codes is provided.

  1. Start-up of a cold loop in a VVER-440, the 7{sup th} AER benchmark calculation with HEXTRAN-SMABRE-PORFLO

    Energy Technology Data Exchange (ETDEWEB)

    Hovi, Ville; Taivassalo, Veikko; Haemaelaeinen, Anitta; Raety, Hanna; Syrjaelahti, Elina [VTT Technical Research Centre of Finland Ltd, VTT (Finland)

    2017-09-15

    The 7{sup th} dynamic AER benchmark is the first in which three-dimensional thermal hydraulics codes are supposed to be applied. The aim is to get a more precise core inlet temperature profile than the sector temperatures available typically with system codes. The benchmark consists of a start-up of the sixth, isolated loop in a VVER-440 plant. The isolated loop initially contains cold water without boric acid and the start-up leads to a somewhat asymmetrical core power increase due to feedbacks in the core. In this study, the 7{sup th} AER benchmark is calculated with the three-dimensional nodal reactor dynamics code HEXTRAN-SMABRE coupled with the porous computational fluid dynamics code PORFLO. These three codes are developed at VTT. A novel two-way coupled simulation of the 7{sup th} AER benchmark was performed successfully demonstrating the feasibility and advantages of the new reactor analysis framework. The modelling issues for this benchmark are reported and some evaluation against the previously reported comparisons between the system codes is provided.

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

    Directory of Open Access Journals (Sweden)

    Gy. Ézsöl

    2012-01-01

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

  3. 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.

  4. 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)

  5. Application of the coupled code system KIKO3D/ATHLET to the boron dilution transients in VVER-440 Type NPP

    International Nuclear Information System (INIS)

    Hegyi, G.; Kereszturi, A.; Trosztel, I.

    2003-01-01

    The transient caused by a perturbation of boron concentration and coolant temperature at the inlet of a Russian developed reactor (VVER-440) is reanalysed as a part of the modernisation (introduction of a new type profiled fuel) and power upgrading (up to 108 %) project. This task is one of the basis cases to be investigated in the safety analysis of the pressurized water reactor (PWR) and the criteria of Anticipated Operational Occurrences (AOO) have to be fulfilled for it. First detailed planning calculations were performed with the thermal hydraulic system code ATHLET and neutron physical code system KARATE-440 to find out the appropriate initial parameter set taking into account the active safety system of the NPP. Finally the most reactive case is analysed by the KIKO3D/ATHLET coupled system code. Whereas the investigation is done for safety analysis, conservative assumptions are imposed on reactivity characteristics. Moreover at the core inlet no-mixing is supposed from the unaffected loops. The presented calculations show, how the coupled code system with a detailed description of plant functions and core behaviour can help to understand better the local phenomena in the study of a potential risk of dilution accident, as it offers the possibility to evaluate the plant safety in a more realistic and versatile manner. (author)

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  7. Severe accident experiments on PLINIUS platform. Results of first experiments on COLIMA facility related to VVER-440. Presentation of planned VULCANO and KROTOS tests

    International Nuclear Information System (INIS)

    Piluso, P.; Boccaccio, E.; Bonnet, J.-M.; Journeau, C.; Fouquart, P.; Magallon, D.; Ivanov, I.; Mladenov, I.; Kalchev, S.; Grudev, P.; Alsmeyer, H.; Fluhrer, B.; Leskovar, M.

    2005-01-01

    In the hypothetical case of a nuclear reactor severe accident, the reactor core could melt and form a mixture of nuclear fuel (UO 2 + Fission Products), metallic or oxidized cladding + steel, called c orium , of highly refractory oxides (UO 2 , ZrO 2 ) and metallic or oxidized steel, that could eventually flow out of the vessel and mix with the substrate decomposition products (generally oxides such as SiO 2 , Al 2 O 3 , CaO, Fe 2 O 3 ). The French Atomic Energy Commission (CEA) has launched a R and D programme aimed at providing the tools for improving the mastering of severe accidents. It encompasses the development of models and codes, performance of experiments in simulant and prototypic materials and the analysis of international experiments. The experiments with prototypic corium (i.e. material containing depleted UO 2 ) are performed in the PLINIUS experimental platform at CEA Cadarache. It comprises the VULCANO facility for 50-100 kg tests (corium-material interactions, corium solidification etc.), the COLIMA facility for smaller scale (∼1 kg) experiments, the VITI facility for corium properties measurement and the KROTOS facility for corium-water interaction (a few kg). In the framework of the 5 th European Framework Programme, free trans-national access to these facilities has been offered to EU and Associated States researchers. For the first PLINIUS access, COLIMA experiments have been conducted with a Bulgarian Team (TU/SOFIA, BAS/INRNE and NPP/KOZLODUY). This series of tests was devoted to experimental studies on fission products release and corium behaviour in the late phase in a hypothetic case of severe accident in a PWR type VVER-440. The COLIMA experimental results are consistent with previous experiments on irradiated fuels (VERCORS, PHEBUS) with small differences for some fission products and show new results for the remaining corium. For the second visit, scientific users from FZK in Germany were selected to validate the COMET core

  8. 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)

  9. 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)

  10. Rolls-Royce successful modernization of safety-critical Instrumentation and Control (I and C) equipment at the Dukovany VVER 440/213 Nuclear Power Plant, based on SPINLINE 3 platform

    International Nuclear Information System (INIS)

    Rebreyend, P.; Burel, J.P.; Spoc, J.; Karasek, A.

    2010-01-01

    Rolls-Royce has provided on-time delivery of a substantial safety-critical I and C overhaul for four Nuclear reactors operated by Czech Republic utility, CEZ a.s. This nine-year project is considered to be one of the largest I and C modernization projects in the world. The Dukovany VVER 440 I and C modernization project and its key success factors are profiled in this paper. The project is in the final stages with the last unit to be completed in 2009. Beginning in September 2000, the project is in compliance with the initial schedule. Rolls-Royce has been designing and manufacturing I and C solutions dedicated to the implementation of safety and safety-related functions in nuclear power plants (NPPs) for more than 30 years. Though the early solutions were non-software-based, since 1984 software-based solutions for safety I and C functions have been deployed in operating NPPs across France and 15 other countries. The Rolls-Royce platform is suitable for implementation of safety I and C functions in new NPPs, as well as in the modernization of safety equipment in existing plants. CEZ a.s. is a major electricity supplier for the national grid. At Dukovany, CEZ a.s. operates four units of VVER-440/213-type reactors producing one quarter of CEZ a.s. electricity production. The first of these units was connected to the grid in 1985. Since the year 2000, the nine-year modernization program has been underway at Dukovany, at a cost of more than 200 million Euros. The equipment replacement was implemented during regular, planned outages of the original equipment and systems. After an international bidding phase, CEZ a.s. awarded a contract to Skoda JS for general engineering and project management. Individual subcontracts were then signed between Skoda JS and a consortium between Rolls-Royce and Areva for modernization of the safety systems, including the Reactor Protection System (RPS), the Reactor Control System (RCS), and the Post-Accident Monitoring System (PAMS). Two

  11. The threat to Austria from Czechoslovakian nuclear power plants not yet headed off

    International Nuclear Information System (INIS)

    Tollmann, A.

    1990-01-01

    The author is dissapointed that the recent political changes in the neighbouring Czechoslovakia did not result in a construction stop of Temelin. The Czechoslovakian nuclear lobby fought back. With one exception all members of the present Czechoslovakian government are nuclear proponents. The way to success will be stonier than in Austria because in Czechoslovakia - in contrast to Austria - the public is unenlightened

  12. Pituitary dwarfism in Saarloos and Czechoslovakian wolfdogs is associated with a mutation in LHX3.

    Science.gov (United States)

    Voorbij, A M W Y; Leegwater, P A; Kooistra, H S

    2014-01-01

    Pituitary dwarfism in German Shepherd Dogs is associated with autosomal recessive inheritance and a mutation in LHX3, resulting in combined pituitary hormone deficiency. Congenital dwarfism also is encountered in breeds related to German Shepherd Dogs, such as Saarloos and Czechoslovakian wolfdogs. To investigate whether Saarloos and Czechoslovakian wolfdog dwarfs have the same LHX3 mutation as do Germans Shepherd Dog dwarfs. A specific aim was to determine the carrier frequency among Saarloos and Czechoslovakian wolfdogs used for breeding. Two client-owned Saarloos wolfdogs and 4 client-owned Czechoslovakian wolfdogs with pituitary dwarfism, 239 clinically healthy client-owned Saarloos wolfdogs, and 200 client-owned clinically healthy Czechoslovakian wolfdogs. Genomic DNA was amplified using polymerase chain reaction (PCR). In the Saarloos and Czechoslovakian wolfdog dwarfs, PCR products were analyzed by sequencing. DNA fragment length analysis was performed on the samples from the clinically healthy dogs. Saarloos and Czechoslovakian wolfdog dwarfs have the same 7 bp deletion in intron 5 of LHX3 as do German Shepherd Dog dwarfs. The frequency of carriers of this mutation among clinically healthy Saarloos and Czechoslovakian wolfdogs used for breeding was 31% and 21%, respectively. An LHX3 mutation is associated with pituitary dwarfism in Saarloos and Czechoslovakian wolfdogs. The rather high frequency of carriers of the mutated gene in the 2 breeds emphasizes the need for screening before breeding. If all breeding animals were genetically tested for the presence of the LHX3 mutation and a correct breeding policy would be implemented, this disease could be eradicated completely. Copyright © 2014 by the American College of Veterinary Internal Medicine.

  13. From Wolves to Dogs, and Back: Genetic Composition of the Czechoslovakian Wolfdog.

    Science.gov (United States)

    Smetanová, Milena; Černá Bolfíková, Barbora; Randi, Ettore; Caniglia, Romolo; Fabbri, Elena; Galaverni, Marco; Kutal, Miroslav; Hulva, Pavel

    2015-01-01

    The Czechoslovakian Wolfdog is a unique dog breed that originated from hybridization between German Shepherds and wild Carpathian wolves in the 1950s as a military experiment. This breed was used for guarding the Czechoslovakian borders during the cold war and is currently kept by civilian breeders all round the world. The aim of our study was to characterize, for the first time, the genetic composition of this breed in relation to its known source populations. We sequenced the hypervariable part of the mtDNA control region and genotyped the Amelogenin gene, four sex-linked microsatellites and 39 autosomal microsatellites in 79 Czechoslovakian Wolfdogs, 20 German Shepherds and 28 Carpathian wolves. We performed a range of population genetic analyses based on both empirical and simulated data. Only two mtDNA and two Y-linked haplotypes were found in Czechoslovakian Wolfdogs. Both mtDNA haplotypes were of domestic origin, while only one of the Y-haplotypes was shared with German Shepherds and the other was unique to Czechoslovakian Wolfdogs. The observed inbreeding coefficient was low despite the small effective population size of the breed, possibly due to heterozygote advantages determined by introgression of wolf alleles. Moreover, Czechoslovakian Wolfdog genotypes were distinct from both parental populations, indicating the role of founder effect, drift and/or genetic hitchhiking. The results revealed the peculiar genetic composition of the Czechoslovakian Wolfdog, showing a limited introgression of wolf alleles within a higher proportion of the dog genome, consistent with the reiterated backcrossing used in the pedigree. Artificial selection aiming to keep wolf-like phenotypes but dog-like behavior resulted in a distinctive genetic composition of Czechoslovakian Wolfdogs, which provides a unique example to study the interactions between dog and wolf genomes.

  14. Pituitary dwarfism in Saarloos and Czechoslovakian wolfdogs is associated with a mutation in LHX3

    NARCIS (Netherlands)

    Voorbij, AMWY; Leegwater, Peter; Kooistra, Hans

    2014-01-01

    Background Pituitary dwarfism in German Shepherd Dogs is associated with autosomal recessive inheritance and a mutation in LHX3, resulting in combined pituitary hormone deficiency. Congenital dwarfism also is encountered in breeds related to German Shepherd Dogs, such as Saarloos and Czechoslovakian

  15. Upgrading the safety of VVER-440/V-230

    International Nuclear Information System (INIS)

    Kelm, P.; Wenk, W.

    1995-01-01

    Besides measures seeking to restore the status as laid down in the project design, especially backfitting measures must be mentioned which serve to ensure component and pipe integrity. Ensuring component integrity is a problem not only of RPV embrittlement, but also of failure prevention. This aspect was not always taken into account properly. Further activities in the field of component integrity will focus on backing the brittle fracture evaluation of the RPV; qualifying the leak-before-breack criterion for the main pipes and in areas with screwed connections; qualifying the program of in-service inspections. Several operators are currently in the process of drafting backfitting programs. The upgrading measures envisaged must be checked as to their balanced nature. In certain plants, the integrity of the RPV coud turn out to be the weak spot in upgrading measures. As a consquence, concepts seeking to achieve upgrading for long periods of time may differ from one location to the next and even between units. Extensive modifications in systems engineering and building structures generally must be evaluated against the expected improvement in safety of the whole plant. (orig.) [de

  16. Corrosion particles in the primary coolant of VVER-440 reactors

    International Nuclear Information System (INIS)

    Vajda, N.; Molnar, Z.; Macsik, Z.; Szeles, E.; Hargittai, P.; Csordas, A.; Pinter, T.; Pinter, T.

    2010-01-01

    Corrosion and activity build-up processes are of major concern in ageing and life-extension of nuclear power reactors. Researches to study the migration of radioactive corrosion particles have been initiated at Paks Nuclear Power Plant (NPP), Hungary in order to better understand the corrosion of the primary circuit surfaces, the transport and activation of the particles of corrosion origin and their deposition on in-core and out-of-core surfaces. Radioactive corrosion particles were collected from the primary coolant and the steam generator surfaces of the 4 reactor units and subjected to detailed microanalytical and radioanalytical investigations. Scanning electron microscopy and energy dispersive X-ray microanalysis (SEM-EDX) were used to study the morphology and the composition of the matrix elements in the particles and the deposited corrosion layers. Particles identified by SEM-EDX were re-located under optical microscope by means of a coordinate transformation algorithm and were separated with a micromanipulator for further studies. Activities of γ emitting radionuclides were determined by high resolution γ spectrometry, and those of β decaying isotopes were measured by liquid scintillation (LS) spectrometry after radiochemical processing. High sensitivity of the nuclear measuring techniques allowed us to determine the low activity concentrations of the long-lived radionuclides, i.e. 60 Co, 54 Mn, 63 Ni, 55 Fe in the individual particles. Finally, high resolution sector-field inductively coupled plasma mass spectrometry (SF-ICP-MS) was applied to determine the ultralow concentrations of Co, Fe, Ni in the same particles. Specific activities of 60 Co/Co, 54 Mn/Fe, 55 Fe/Fe and 63 Ni/Ni were derived from the measured activity and concentration data. Specific activities of the radioactive corrosion products reveal the history of activity buildup processes in the particle. Typically, Fe-Cr-Ni oxide particles formed as a result of corrosion of the steel surfaces are released into the coolant and activated during the residence time (τ) in the reactor core that varies in a wide range. Smallest τ values are obtained if the particle circulates without deposition in the core and high values reflect a long deposition time in the core, probably on the fuel cladding surfaces. Residence times (τ) were calculated from the measured specific activities under various migration-activation conditions. Conclusions about particle transport-activation processes in case of the 4 reactor units have been drawn. The residence time of the migrating corrosion particles is usually small, less than 50 days. Periods of stable operation, shutdown and startup were comparatively evaluated. The 4 reactor units of the plant which have different operation history (including steam generator decontamination in certain cases) were also compared with each other. Significant differences in the Fe/Cr/Ni ratio of the oxide layers of the steam generators were observed. Correlation between the behavior of the migrating particles and particles deposited on the steam generator surfaces were also studied. (author)

  17. VVER-440 loading patterns optimization using ATHENA code

    International Nuclear Information System (INIS)

    Katovsky, K.; Sustek, J.; Bajgl, J.; Cada, R.

    2009-01-01

    In this paper the Czech optimization state-of-the-art, new code system development goals and OPAL optimization system are briefly mentioned. The algorithms, maths, present status and future developments of the ATHENA code are described. A calculation exercise of the Dukovany NPP cycles, on increased power using ATHENA, starting with on-coming 24th cycle (303 FPD) continuing with 25th (322 FPD), and 26th (336 FPD); for all cycles K R ≤1.54 is presented

  18. Nuclear safety evaluation of the VVER 440, Type 213

    International Nuclear Information System (INIS)

    Urbancik, L.

    1997-01-01

    The supervisory activities of the State Office for Nuclear Safety at the Dukovany nuclear power plant are described. No event resulting in an inpermissible radioactivity leak into the environment occurred at the plant in 1996. From among the 76 failures and events having occurred, only 4 were classified as level 1 on the International Nuclear Event Scale. Changes in the technology of radioactive waste bituminization were proposed. The Interim Spent Fuel Storage Facility at the Dukovany site was in test operation in 1996. Selected physical parameters of this facility were monitored. Seven international transports of spent fuel were accomplished in 1996. The dose rates in the surroundings of the Dukovany plant are monitored constantly by a teledosimetric system operated by the nuclear power plant. Periodical sampling and radionuclide activity measurements in the environment are also performed. (M.D.)

  19. VVER-440 fuel cycles possibilities using modified FA design

    International Nuclear Information System (INIS)

    Mikolas, P.; Svarny, J.; Razym, V.; Dostal, M.; Jenik, J.; Krupar, P.

    2009-01-01

    A nearly equilibrium five-year cycle has been achieved at Dukovany NPP over the last years. This means that working fuel assemblies (WFA) with an average enrichment of 4.25 w% (control assemblies (CA) with an average enrichment of 3.82 w%) are normally loaded and reloaded for five years. Operation at uprated thermal power (105% of the original one, increase from 1375 MW t to 1444 MW t ) is being prepared by use of WFA with an average enrichment of 4.38 w% (CA with an average enrichment of 4.25 w%). With the aim of fuel cycle economy improvement, the fuel residence time in the core has to be prolonged up to six years with one cycle duration time up to 18 months and preserving loadings with very low leakage. In order to achieve this goal, at least neutron-physical characteristics of FA must be improved and such changes should be evaluated from other viewpoints. Some particular changes have already been analyzed earlier. Designs of new fuel assemblies with higher (and in the central part of a FA the highest possible, i.e. 4.95 w%) enrichment with preserving low pin power non-uniformity are described in the presented paper. An FA with an average enrichment of 4.66 w% (lower than originally evaluated) containing six fuel pins with 3.35 w% Gd 2 O 3 content was selected in the end. Fuel pins have bigger pellet diameter, bigger pin pitch and thinner FA shroud. A newly designed FA was evaluated from the viewpoint of physics (pin power non-uniformity, criticality of fuel at transport and storage and determination of basic quantities for spent fuel storage purposes by ORIGEN code), thermo-hydraulics (comparison of subchannel output temperatures and the departure from nucleate boiling ratio - DNBR) and mechanical properties. The purpose of this study was to simulate an FA subject to the loads during its six- year lifetime whereas normal working conditions were taken into account. There are presented two models with different shroud thickness undergoing these analyses. Both models also undergo buckling shroud analyses. Next the maximum inner excessive pressure limits (as a consequence of accident conditions) were determined. Furthermore, the low shroud thickness limit for loads representing normal working conditions was assessed. The model, whose shroud thickness was reduced down to 1.0 mm, was subjected to a low cycle fatigue analysis. Possibilities of fuel cycles are evaluated on model loadings with the newly designed FA, where the base are loadings for 27 th - 34 th cycles of the third unit of Dukovany NPP for uprated power. These cycles were prolonged (from approx 330 FPD to 370 FPD) using FA with higher enrichment. Moreover, newly optimized loadings of a length of up to 500 FPD (18 months) were considered. The transient process started from the last of the set of loadings (27 th - 34 th ) for uprated power. Newly designed fuel assemblies were loaded regularly in 18-month cycles. Average enrichment of CA was 4.38 w%. Transient loadings are formed by cycles 35-37 and the equilibrium cycle is created by cycles 38 and 39. Each cycle was optimized individually and fuel assemblies intended for unloading were determined for each cycle separately. The 'equilibrium' cycle is realized by three consecutive loadings with 16 fresh WFA and 2 fresh CA. Basic characteristics of a reference cycle and an 18-month cycle were compared. Optimization was performed by the OPAL - B code on the basis of 3D n-ph calculations of the MOBY-DICK code with the target function F dh < 1.51. Consecutive thermo-hydraulic calculations were executed following the core neutron-physical analysis that had been carried out by the MOBY-DICK code in the 1/6 core symmetry. These thermo-hydraulic calculations were executed for loadings of both existing and newly designed fuel assemblies. Fast neutron fluences (calculated by TORT transport code based on neutron sources calculated by MOBY-DICK code) onto the reactor pressure vessel for proposed the 12-month and 18-month cycles were also calculated and compared. The analyses performed confirmed that fuel cycles using newly designed FA and fulfilling basic safety criteria can be designed together with fuel cycle economy improvement. (authors)

  20. Implementation of safety parameter display system at VVER-440 NPPs

    International Nuclear Information System (INIS)

    Manninen, T.

    1997-01-01

    Furnishing WWER-440 nuclear power plant units with a safety parameter display system (SPDS) fulfilling the requirements of internationally recognized standards and guidelines has been ranked high on the lists of proposed safety improvement projects. Technically such an SPDS system can be implemented either as a separate stand-alone system or as a more or less closely integrated part of a process information system of the plant unit. In the paper examples of these approaches are presented. Functionally all these examples include the well proven SPDS concept developed by IVO Power Engineering Ltd, Finland. The functional design basis, the general requirements for the system platform, experience with implementation and expansion possibilities of the systems are discussed. (author)

  1. Containment leak-tightness enhancement at VVER 440 NPPs

    International Nuclear Information System (INIS)

    Prandorfy, M.

    2000-01-01

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

  2. Highly Expressed Granulocyte Colony-Stimulating Factor (G-CSF) and Granulocyte Colony-Stimulating Factor Receptor (G-CSFR) in Human Gastric Cancer Leads to Poor Survival.

    Science.gov (United States)

    Fan, Zhisong; Li, Yong; Zhao, Qun; Fan, Liqiao; Tan, Bibo; Zuo, Jing; Hua, Kelei; Ji, Qiang

    2018-03-23

    BACKGROUND Chemotherapy for advanced gastric cancer (GC) patients has been the mainstay of therapy for many years. Although adding anti-angiogenic drugs to chemotherapy improves patient survival slightly, identifying anti-angiogenic therapy-sensitive patients remains challenging for oncologists. Granulocyte colony-stimulating factor (G-CSF) promotes tumor growth and angiogenesis, which can be minimized with the anti-G-CSF antibody. Thus, G-CSF might be a potential tumor marker. However, the effects of G-CSF and G-CSFR expression on GC patient survival remain unclear. MATERIAL AND METHODS Seventy GC tissue samples were collected for G-CSF and G-CSFR detection by immunohistochemistry. A total of 40 paired GC tissues and matched adjacent mucosa were used to measure the G-CSF and G-CSFR levels by ELISA. Correlations between G-CSF/G-CSFR and clinical characteristics, VEGF-A levels and overall survival were analyzed. Biological function and underlying mechanistic investigations were carried out using SGC7901 cell lines, and the effects of G-CSF on tumor proliferation, migration, and tube formation were examined. RESULTS The levels of G-CSFR were upregulated in GC tissues compared to normal mucosa tissues. Higher G-CSF expression was associated with later tumor stages and higher tumor VEGF-A and serum CA724 levels, whereas higher G-CSFR expression was associated with lymph node metastasis. Patients with higher G-CSF expression had shorter overall survival times. In vitro, G-CSF stimulated SGC7901 proliferation and migration through the JAK2/STAT3 pathway and accelerated HUVEC tube formation. CONCLUSIONS These data suggest that increased G-CSF and G-CSFR in tumors leads to unfavorable outcomes for GC patients by stimulating tumor proliferation, migration, and angiogenesis, indicating that these factors are potential tumor targets for cancer treatment.

  3. Comparison of physiological load tolerances between the Siberian husky and the Czechoslovakian wolfdog, during sport training

    Directory of Open Access Journals (Sweden)

    Dominika Gulda

    2018-03-01

    Full Text Available The purpose of the study was to determine the physiological load tolerances of two breeds of dogs used for sports, namely the Siberian husky and the Czechoslovakian wolfdog, on the basis of measurements of surface temperature and blood lactic acid levels.Two breeds - Czechoslovakian wolfdog (10 individuals and Siberian husky (10 individuals, 20 dogs - male, 4-6 years old, were selected for the study. All the qualified animals were previously examined by a veterinarian and considered to be healthy. The dogs tested were used in dogtrekking sport competitions. For both breeds, an attempt was made to test the dogtrekking harness for 5 km of non-stop track running. The animals trotted while being led by a guide. Three attempts were made for each dog at 48 hour intervals. All dogs were tested for two parameters, first before and then after the exercise – measuring surface temperature at selected points of the body as well as lactic acid concentration. A higher and statistically significant level of lactic acid was recorded in the case of Siberian husky. Before the run, the level of lactic acid was comparable in both breeds. The second parameter was the surface temperature measured at the selected measuring points. Significant statistical differences were noted for the wolfdog breed at P≤0.05 before the exercise and 10 minutes after resting, at the neck, rump and abdominal points. In addition, the same level of statistical significance was measured by surface thermography at the abdominal point, both before and immediately after the run. The high statistically significant increase (P≤0.01 in surface temperature was noted for muscles of the so-called rump, both before and after the exertion. There was no statistically significant difference in the back thermography in the wolfdog breed. In the Siberian husky breed, statistically significant differences (P≤0.05 were observed at the neck, rump and chest points, in confrontation with the

  4. Secular Trends and Latitude Gradients in Sex Ratios at Birth in Czechoslovakia and the Post-Czechoslovakian States

    Directory of Open Access Journals (Sweden)

    Victor Grech

    2012-01-01

    Full Text Available Latitude gradients and secular trends in Europe and North America have been found in the male-female ratio at birth (M/F: male births divided by total births which is expected to be 0.515. Annual national data for Czechoslovakia and the post-Czechoslovakian (Czech Republic and Slovakia countries for male and female live births were obtained from the World Health Organisation and analysed with contingency tables. This study analysed 13,123,538 live births. An overall decreasing trend in M/F was found (p < 00001. No latitude gradient was noted. There was an overall deficit of 15,232 male births based on an M/F of 0.515. M/F is declining in this region, despite well developing economies that have resisted the worldwide slowdown. An interplay of several poorly understood factors is likely.

  5. 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)

  6. VVER-440 control rod follower induced local power peaking computational benchmark: MCNP and KARATE solutions

    International Nuclear Information System (INIS)

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

    2009-01-01

    In this paper the linear pin power calculation in the KARATE-440 code system are presented. The calculation results show that: 1) The coupler mathematical benchmark was solved by the KARATE code system using the same methods and approximations as in case of NPP applications. 2) Without Hf plate in the fuel pin row next to the problematic coupler section pronounced local power peak arises. 4) Though the underprediction of the peak by KARATE-SADR is 4%, it is in accordance with the applied uncertainty of KARATE-SADR. 5) The application of Hf plate solves the problem.

  7. Seismic assessment of Kozloduy VVER 440, Model 230 nuclear power plant

    International Nuclear Information System (INIS)

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

    1991-01-01

    Excluding system design deficiency relative to US and Western Europe standards, it was found that the plant has many seismic vulnerabilities similar to those that existed in many of the US plants prior to about 1979 when the Systematic Evaluation Program was initiated. The primary coolant system has been substantially upgraded after the 1977 Vrancea earthquake. Other upgrades have been made to weak elements in the ECCS and electrical systems. There are still a number of components that could likely survive the currently defined Safe Shutdown Earthquake of 0.1 g but which would not meet current design standards. Many of the weakest components could be upgraded at a moderate cost to withstand a seismic event exceeding 0.1 g. Current studies of the site seismicity lean toward a higher peak ground acceleration and increased amplification of building motion, thus backfits that have been accomplished may become marginal for newly defined loads. However the proper consideration of soil structure interaction and detailed structural analysis using less conservative modeling assumptions, could mitigate the impact of increasing the seismic input and limit the amount of reinforcement required. In the interim, substantial improvements to seismic safety could be accomplished by simple, inexpensive modifications to equipment anchorage and some achievable improvements to connection detail of the precast concrete structures. (author)

  8. Accident localization system with jet condensers for VVER 440-V 230 NPP at Jaslovske Bohunice

    International Nuclear Information System (INIS)

    Murani, J.

    1995-01-01

    The operational safety of the V1 nuclear power plant (NPP) is unsatisfactory and does not correspond to present requirements as to nuclear safety. Further NPP operation after 1995 is conditional on nuclear safety enhancement to a level comparable with that in West European countries. This aim should be achieved by a principal reconstruction involving in addition to others also backfitting the V1 NPP with technical facilities aimed at coping with a design basis accident (DBA).To cope with such an accident the Power Equipment Research Institute (VUEZ) designed an accident localization system with jet condensers. This system consists of (a) an air trap (one for each unit, mutually interconnected) with an expansion bell enclosed within, placed on a plate with 200 pipes of jet condensers passing through, and (b) a connecting duct between the hermetic zone and the air trap. The vertical jet condenser is an essential element of the system designed for steam condensation. Apart from condensation it serves as a water seal separating units 1 and 2.Demonstration tests of the jet condenser (model 1:1) condensing function were carried out at the testing unit of the All-Union Research Institute for NPP Operation (VNIIAES), Moscow in Kashir, 11-22 September 1992. These experiments proved the jet condenser ability to ensure complete condensation of the steam produced. Experimental verification of the sealing function (model 1:1) was carried out at the testing unit of the VUEZ Tlmace. These experiments concerning the dynamics and overpressure in the free space above the pool were close to the conditions in the air trap during DBA. The jet condenser height was proved to be sufficient to ensure the sealing function. Design and experimental work has been implemented in close cooperation with Russian experts Mr. V.N. Bulynin from the VNIIAES, Moscow, and Mr. M.V. Kuznecov from the Scientific and Engineering Center for Nuclear and Radiological Safety, Moscow. (orig.)

  9. Radioactive sludge and wastewater analysis and treatment in the Hungarian VVER-440/213-type NPP

    International Nuclear Information System (INIS)

    Patzay, G.; Weiser, L.; Feil, F.; Schunk, J.; Patek, G.; Pinter, T.

    2010-01-01

    It is well known that in the Hungarian VVER-type nuclear power plant Paks the radioactive waste waters are collected in common tanks. These water streams contain radioactive isotopes in ultra-low concentration and inactive compounds as major components (borate 1.7 g/dm 3 , sodium-nitrate 0.4 g/dm 3 , sodium-hydroxide 0.16 g/dm 3 , and oxalate 0.25 g/dm 3 ). These low salinity solutions were evaporated by adding sodium-hydroxide, until 400 g/dm 3 salt content is reached. There is about 6000 m 3 concentrated evaporator bottom residues in the tanks of the reactor. There are some tanks at the power plant containing sludge type radioactive waste containing more or less liquid phase too. The general physical and chemical characteristics (density, pH, total solid, dissolved solid etc.) and chemical and radiochemical composition are important information for volume reduction and solidification treatment of these wastes. We have investigated and constructed a complex analysis system for the radioactive sludge and supernatant analysis, including the physical, as well as the chemical and radiochemical analysis methods. Using well known analysis techniques as ion chromatography, ICP-MS, AAS, gamma-and alpha-spectrometry and chemical alkaline fusion digestion and acidic dissolution methods we could analyze the main inorganic, organic and radioactive components of the sludges and supernatants. Determination of the mass and charge balance for the sludge samples were more difficult then for the supernatant samples. Not only are there assumptions required about the chemical form and the oxidation state of the species present in the sludge, but many of the compounds in the sludge are mixed oxides which are not directly measured. Also, the sludge is actually a slurry with a high water content. The interstitial liquid is in close contact with the sludge, and there are many ionic solubility equilibriums. The anion data for the sludge samples are based on the water soluble anions that would be available to a water wash. The water wash would not account for the insoluble hydroxides, carbonates, and mixed oxides present. The insoluble species do not contribute to the charge balance, and the cation charge is not used in the calculation. Most of the nitrate reported for the sludge is due to the interstitial liquid. Considering the limitations of these calculations, the mass balance was within the analytical error (±20%) for the sludge samples. There were three sample preparation methods used to investigate the total anion content of the sludge samples, which included water leach, potassium-hydroxide and/or sodium peroxide/sodium hydroxide fusion and acidic dissolution. (author)

  10. TACIS 91: Application of leak-before-break concept in VVER 440-230

    Energy Technology Data Exchange (ETDEWEB)

    Bartholome, G.; Faidy, C.; Franco, C. [and others

    1997-04-01

    The applicability of the leak-before-break (LBB) concept for primary piping in the first generation of WWER type plants in Russia is investigated. The procedures for LBB behavior used in France and Germany are applied, and the evaluation is discussed within the framework of the European Technical Assistance for the Community of Independent States (TACIS) project. Emphasis is placed on experimental validation of national and international engineering practice for evaluating and optimizing existing installations. Design criteria of WWER plants are compared to western standard design.

  11. 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.)

  12. The results of postirradiation examinations of VVER-1000 and VVER-440 fuel rods

    Science.gov (United States)

    Dubrovin, K. P.; Ivanov, E. G.; Strijov, P. N.; Yakovlev, V. V.

    1991-02-01

    The paper presents the results of postirradiation examination of the fuel rods having different fuel-cladding gaps, pellet densities, pellet inner diameters and so on. The fuel rods were irradiated in the material science reactor (MR) of the Kurchatov Institute of Atomic Energy and at 4 unit of the Novo-Voronezh nuclear powerplant. Some data on fission gas release and rod geometry and compared with computer code predictions.

  13. Innovation of blow-down system in steam generators of a VVER 440 unit

    International Nuclear Information System (INIS)

    Matal, O.; Simo, T.; Mancev, M.D.

    1997-01-01

    The impurities getting into the steam generator with the feedwater are continually removed by the blowdown and unit sludge system. The mostly non-symmetrical type of pipe branches under steam generators at WWER-440 units causes nonuniform blowdown flow rates at the halves of the steam generator; this often leads to a blocking of the pipe with the lower flow rate. The most simple way of hydraulically equalizing the blowdown pipes is to implement symmetric blowdown pipes and to install efficient throttling elements in the pipe. The proposed innovation will make it possible to re-distribute the blowdown flow rates and to reduce more effectively the concentrations of impurities in steam generators. (M.D.)

  14. 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 accomplished rewetting of the core. Pre and post-test analyses were performed by RELAP5/mod3.3 - used for accident analysis of the Paks NPP - with the aim to validate the capability of the code, especially in the low primary pressure domain. Accumulator injection took place in several steps both in the tests and in the calculations, but step size and timing were different that led to differences in primary inventory distribution. In the calculations core heat-up consequently occurred at lower core collapsed level than in the tests. Prediction of break flow, especially transition from single phase to two-phase conditions, remains a major cause of uncertainty. (authors)

  15. Relap5/Mod3.1 analysis of main steam header rupture in VVER- 440/213 NPP

    Energy Technology Data Exchange (ETDEWEB)

    Kral, P. [Nuclear Research Inst. Rez (Switzerland)

    1995-12-31

    The presentation is focused on two main topics. First the applied modelling of PGV-4 steam generator for RELAP5 code are described. The results of steady-state calculation under reference conditions are compared against measured data. The problem of longitudinal subdivision of SG tubes is analysed and evaluated. Secondly, a best-estimate analysis of main steam header (MSH) rupture accident in WWER-440/213 NPP is presented. The low reliability of initiation of ESFAS signal `MSH Rupture` leads in this accident to big loss of secondary coolant, full depressurization of main steam system, extremely fast cool-down of both secondary and primary system, opening of PRZ SV-bypass valve with later liquid outflow, potential reaching of secondary criticality by failure of HPIS. 7 refs.

  16. Relap5/Mod3.1 analysis of main steam header rupture in VVER- 440/213 NPP

    Energy Technology Data Exchange (ETDEWEB)

    Kral, P [Nuclear Research Inst. Rez (Switzerland)

    1996-12-31

    The presentation is focused on two main topics. First the applied modelling of PGV-4 steam generator for RELAP5 code are described. The results of steady-state calculation under reference conditions are compared against measured data. The problem of longitudinal subdivision of SG tubes is analysed and evaluated. Secondly, a best-estimate analysis of main steam header (MSH) rupture accident in WWER-440/213 NPP is presented. The low reliability of initiation of ESFAS signal `MSH Rupture` leads in this accident to big loss of secondary coolant, full depressurization of main steam system, extremely fast cool-down of both secondary and primary system, opening of PRZ SV-bypass valve with later liquid outflow, potential reaching of secondary criticality by failure of HPIS. 7 refs.

  17. Development and installation of a new on-line plant safety monitoring system for the Paks VVER-440 units

    International Nuclear Information System (INIS)

    Vegh, J.; Major, C.; Buerger, L.; Lipcsei, S.; Horvath, C.; Kapocs, G.; Eiler, J.; Hornaes, A.; Hulsund, J.E.

    2000-01-01

    The paper describes the architecture, modules, algorithms and human-machine interface of a new operator support system (OSS), which is integrated into the new, reconstructed Paks NPP plant computers. The main task of the new OSS is to perform continuous plant safety monitoring and assessment, it has the following basic functions: on-line evaluation and presentation of critical safety function (CSF) status trees, continuous evaluation and presentation of the actual safety status of the plant, displaying and browsing the new symptom-oriented EOPs, automatic displaying of those process signals which are quoted in the EOPs. The first version of the new operator support system was connected to the Paks NPP full scope simulator in October 1999. This configuration was later successfully applied for the simulator testing of the new symptom-oriented EOP set for the Paks NPP in November 1999. The installation process was continued in 2000: the new system started its operation on Unit 2 (June) and on Unit 1 (August), together with the reconstructed, new PCS. (author)

  18. Organization and management of the plant safety evaluation of the VVER-440/230 units at Novovoronezh

    International Nuclear Information System (INIS)

    Afshar, C. M.; Pizzica, P.; Puglia, W. J.; Rozin, V.

    1999-01-01

    As part of the Soviet-Designed Reactor Safety (SDRS) element of the International Nuclear Safety Program (INSP), the US Department of Energy (US DOE) is funding a plant safety evaluation (PSE) project for the Novovoronezh Nuclear Power Plant (NvNPP). The Novovoronezh PSE Project is a multi-faceted project with participants from sixteen different international organizations from five different countries scattered across eleven time zones. The purpose of this project is to provide a thorough Probabilistic Risk Analysis (PRA) and Deterministic Safety Analysis (DSA) for Units 3 and 4 of the NvNPP. In addition, this project provides assistance to the operation organizations in meeting their international commitments in support of safety upgrades, and their regulatory requirements for the conduct of safety analyses. Managing this project is a complex process requiring numerous management tools, constant monitoring, and effective communication skills. Employing management tools to resolve unanticipated problems one of the keys to project success. The overall scope, programmatic context, objectives, project interactions, communications, practical hindrances, and lessons learned from the challenging performance of the PSE project are summarized in this paper

  19. Relap5/Mod3.1 analysis of main steam header rupture in VVER- 440/213 NPP

    International Nuclear Information System (INIS)

    Kral, P.

    1995-01-01

    The presentation is focused on two main topics. First the applied modelling of PGV-4 steam generator for RELAP5 code are described. The results of steady-state calculation under reference conditions are compared against measured data. The problem of longitudinal subdivision of SG tubes is analysed and evaluated. Secondly, a best-estimate analysis of main steam header (MSH) rupture accident in WWER-440/213 NPP is presented. The low reliability of initiation of ESFAS signal 'MSH Rupture' leads in this accident to big loss of secondary coolant, full depressurization of main steam system, extremely fast cool-down of both secondary and primary system, opening of PRZ SV-bypass valve with later liquid outflow, potential reaching of secondary criticality by failure of HPIS

  20. Coal graders in Czechoslovakian mines

    Energy Technology Data Exchange (ETDEWEB)

    Vasek, J.; Klimek, M.

    1980-01-01

    Problems related to sections of the area of application of graders depending on different mining and geological mining-engineering factors are examined. The principal factors are selected from a general group of influencing factors: dip angle of a formation, separability (shear ability) of coal, characteristics of country rocks, adhesion of coal to rock, tectonic fracturing of a seam, and thickness of a formation. Based on practical and theoretical studies all of the principal factors have been categorized. This allows one to obtain an objective picture of the possibility of using graders under specific conditions by comparing different factors.

  1. Energy savings in CSFR - building sector

    International Nuclear Information System (INIS)

    Jacobsen, F.R.

    1993-01-01

    The Czechoslovak/Danish project on energy savings in buildings proves that it is possible to save up to 30% of the energy in buildings. 10% can be saved at an investment of 27 bill KCS. The total investment that is needed to save 30% is 140 bill KCS. Further energy savings can be obtained through more energy efficient supply systems. Information dissemination is important for the energy saving programme as are economic incentives. Investments in energy savings should be profitable for the investor, but this is not the case in the Czech and Slovak republics today. Changes are needed. Energy prices are still to low, compared to investment costs. Financial possibilities are not satisfactory for private investors. Price systems are not favourable to investment in energy savings. Training is needed for boiler men and energy consultants. Legislation is essential for the support of the full range of activities in the energy sector. Research and Development activities must back up the development of the sector. Pilot projects can illuminate the savings potential. The production of technical equipment for control and metering and production of insulation materials must be promoted. (AB)

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  3. AER working group A on improvement extension and validation of parametrized few-group libraries for VVER-440 and VVER-1000

    International Nuclear Information System (INIS)

    Svarny, J.

    1998-01-01

    The AER Working Groups A and B held its sixth meeting at SKODA JS, Plzen in April 28 and 29, 1998. There were altogether 13 participants from 6 member organizations. The list of participants and the list of papers are attached. Main topics of the meeting were: A few-group cross-section library preparation methodology (standard few-group libraries, kinetics parameters, SPND signal interpretation parametrization) and its validation; Participation on intercomparisons of spectral codes (spectral codes benchmark); of kinetics parameters calculations (kinetics parameters benchmark). (author)

  4. 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

  5. Calculation methodology validation. Pt. 2/01-R. Calculation of the multiplication factor for eight experiments with a critical set of nineteen VVER-440 fuel assemblies

    International Nuclear Information System (INIS)

    Kyncl, J.

    2001-04-01

    Comparison calculations were performed for 8 experiments accomplished in 2000 on the LR-0 reactor. The MCNP4a code was applied using effective cross section data in the continuous representation as per the ENDF/B-VI library. (P.A.)

  6. Experimental Studies for the VVER-440/213 Bubble Condenser System for Kola NPP at the Integral Test Facility BC V-213

    International Nuclear Information System (INIS)

    Blinkov, V.N.; Melikhov, O.I.; Melikhov, V.I.; Davydov, M.V.; Wolff, H.; Arndt, S.

    2012-01-01

    In the frame of Tacis Project R2.01/99, which was running from 2003 to 2005, the bubble condenser system of Kola NPP (unit 3) was qualified at the integral test facility BC V-213. Three LB LOCA tests, two MSLB tests, and one SB LOCA test were performed. The appropriate test scenarios for BC V-213 test facility, modeling accidents in the Kola NPP unit 3, were determined with pretest calculations. Analysis of test results has shown that calculated initial conditions and test scenarios were properly reproduced in the tests. The detailed posttest analysis of the tests performed at BC V-213 test facility was aimed to validate the COCOSYS code for the calculation of thermohydraulic processes in the hermetic compartments and bubble condenser. After that the validated COCOSYS code was applied to NPP calculations for Kola NPP (unit 3). Results of Tacis R2.01/99 Project confirmed the bubble condenser functionality during large and small break LOCAs and MSLB accidents. Maximum loads were reached in the LB LOCA case. No condensation oscillations were observed.

  7. Safety provisions for steam generator in Mochovce nuclear power plant. BO CI 04 Integrity of primary collectors of VVER 440 steam generators

    International Nuclear Information System (INIS)

    Cikryt, F.; Bednarek, L.; Matocha, K.; Vejvoda, S.

    1997-01-01

    This paper dealt with the identification of possible damaging mechanism of the collector of the WWER 440 steam generator, cracking of primary collectors, corrosion damage of the protective coat of the primary collector circumferential weld, cracking of breathing space in the region of blinding effect by corrosion and strain, leaking of disassembling joint of the primary collector lid and with the integrity of heat exchanging tubes

  8. 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.

  9. 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.

  10. 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)

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

    International Nuclear Information System (INIS)

    Mikolas, P.

    2010-01-01

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

  12. Selected examples for safety analysis in VVER-440 type reactors simulated by the coupled ATHLET/KIKO3D code system

    International Nuclear Information System (INIS)

    Hegyi, Gy.; Kereszturi, A.; Trosztel, I.

    2005-01-01

    Recently several projects have been initiated in Hungary aiming at the introduction of new fuel type, increased maximum allowed power and economic fuel cycle. The planned upgraded power and parallel application of new fuel type require the renewal of the relevant chapter of the Final Safety Analysis Report (FSAR). One of the main tools used for analyzing transient scenarios initiating by reactivity and power distribution anomalies was the ATHLET/KIKO3D coupled neutron kinetic / thermal-hydraulic code. This paper gives an overview of two analyses, which was prepared in the frame of the revision of Paks FSAR, namely the ''withdrawal of one control rod'' and ''initial phase of main steam line break'' events. (author)

  13. Radioecological state of the Czechoslovakian Danube section

    International Nuclear Information System (INIS)

    Kortus, J.; Mayer, J.; Kopuncova, T.; Vondra, M.

    1984-01-01

    As to be able to determine the radioactivity of the Czechoslovak section of the Danube river monthly water sampling and quarterly sediment, aquatic flora, and fauna sampling was performed. The results of this investigation have shown that the radioactivity level of our Danube section is low, consisting mostly of natural radionuclides. No relationships among respective seasons and radioactivity level were observed, neither any significant dependencies between flow rate and radionuclide level. In spite of the low radioactivity of the Czechoslovak Danube section we assume it necessary to continue in carrying out radiological analyses of water and sediment samples, as well as aquatic flora and fauna, since there are some nuclear power plants already in operation in the Danube basin and further are to be constructed or are already under construction. (author)

  14. State and perspectives of Czechoslovakian nuclear law

    International Nuclear Information System (INIS)

    Bezdek, R.

    1992-01-01

    In Czechoslovakia, the peaceful utilization of nuclear energy is governed by a series of legislative norms of varied character and legal power. The most important are the Act No. 194/1988 and the Act No. 28/1984. The former defines the competence of the Czechoslovak Atomic Energy Commission (CAEC), which is the central authority of state administration in the field of utilization of nuclear energy. The latter deals with the State inspection for the nuclear safety of nuclear facilities. In accordance with this Act, the CAEC is the competent authority for the licensing and inspection of nuclear safety. In addition to the two main Acts, a series of CAEC Regulations govern nuclear activities (accounting and control of nuclear materials, radioactive waste management, physical protection, qualifications of personnel in nuclear facilities, quality assurance, etc.). There is no specific legislation governing nuclear third liability. The solution for the various shortcomings of the contemporary codification lies primarily in change of the present codification. This change, however, should not mean a general and indiscriminate ''destruction'' of the legal norms in force at present, but in gradual and purposive creation of an integral, legal system capable of reacting flexibly, the core of which would consist of an Act concerning the peaceful utilization of nuclear energy and on liability for nuclear damage. (author)

  15. Technical-economic evaluation of uranium reserves in the DIAMO company, Straz pod Ralskem (Czech Republic)

    International Nuclear Information System (INIS)

    Hradek, J.

    1998-01-01

    Changing economic conditions in the Czechoslovakian (CSFR) economy (in 1992) made it necessary to realign exploitation of uranium in the country, taking into consideration economical, and environmental factors. This was done partly through mathematical-geological modelling. This analysis, which take into account mining practice and costs, involved reevaluating the uranium resources. The report describes how this was accomplished. It also describes how the uranium classification system used in the CSFR, which is based on the categories A,B,C1,C2,PI and P2, compares to the IAEA system. (author)

  16. Front against the Temelin nuclear power station

    International Nuclear Information System (INIS)

    Anon.

    1990-01-01

    Though the main concern of the author is the Czechoslovakian Temelin power station, the main target of his attacks are the Austrian proponents of nuclear energy i.e. the Reactor Safty Commission and the Austrian Chancellor. The newly opened possibility of anti-nuclear propaganda in the CSFR, by Greenpeace and the author's organisation is welcomed. The number of signatures collected against Temelin is given as 350.000. 815 signatures come from Japan: a facsimile of some Japanese signatures is presented

  17. Quality assurance of gamma spectrometry in monitoring network of CSFR

    International Nuclear Information System (INIS)

    Malatova, I.; Drabova, D.; Bucina, I.

    2004-01-01

    On the basis of the Chernobyl experience the Czechoslovak government decided in July 1986 to set up Czechoslovak Monitoring Network and to assign the Centre of Radiation Hygiene of the Institute of Hygiene and Epidemiology to be its headquarters (Centre of Czechoslovak Monitoring Network). The requirements for emergency monitoring are stated in the document The principles of Monitoring for Protection of Public Health in case of a Radiation Accident approved by the Czechoslovak government in April 1987. Assignments of components of the Network, equipment and technical support required, aims of their activities and chronological order of their activation are stated in the document Requirements on Monitoring, Setting up and Equipment of the Czechoslovak Monitoring Network drawn up by the Centre and approved by the Czechoslovak Governmental Commission for Coordination of the Measures in Case of a Radiation Accident in April 1988. It should be noted, however, that basic principles of environmental monitoring aimed at obtaining the complete information of radiation situation, discharges and releases of radionuclides both during the normal operation and in case of an accident were worked on since putting the first PWR-type NPP in Czechoslovakia into operation in 1979. In March 1986 the Instruction for emergency monitoring was approved by the commission. The existence of this instruction and corresponding professional, technical and organizational preparedness of organizations departments responsible for monitoring manifested its positive impact especially in the situation after the Chernobyl accident. This fact refers especially to institutions of hygienic service and nuclear power engineering. National and international experience gained after the Chernobyl accident led to some elaboration in the organization of monitoring and to more precise definition of its conception

  18. Uranium industry in the CSFR - Present and future

    International Nuclear Information System (INIS)

    Kubant, J.; Bezdek, Z.; Marek, J.

    1990-01-01

    Czechoslovak uranium industry is at present going through principal reorganization and reconstruction. The supplies to the USSR have finished and the principal partners to the Czechoslovak Uranium Industry (CSUP) have become the Czechoslovak utilities. CSUP is gradually decreasing the output of mined uranium by closing the old mines with the aim to enhance the effectiveness of its activity. It is entering the uranium world market and there is an interest under acceptable conditions to increase the purchase of Czechoslovak uranium aboard to some extent, in the first place on the basis of the long-term contracts. On the other side together with the purchase of Czechoslovak uranium on the world market also the area of Czechoslovak nuclear energy opens to the world including the nuclear fuel cycle

  19. Planning for the next CSFR nuclear power plants. Empresarios Agrupados' view

    International Nuclear Information System (INIS)

    Garcia Rodriquez, A.

    1993-01-01

    A description of the evolution in the nuclear industry in Western countries is given. Spanish experience in nuclear program is presented and the future of the nuclear industry and the alternatives open to Czechoslovakia are discussed

  20. 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

  1. 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.

  2. RISKAUDIT Report no. 7, Vol. 2: 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 second part of the document covers the following aspects of the report: accident analysis; systems analysis; plant operation; operating experience feedback; radio protection and health; probabilistic safety assessment; summary and future plans.

  3. RISKAUDIT Report no. 7, Vol. 2: 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 second part of the document covers the following aspects of the report: accident analysis; systems analysis; plant operation; operating experience feedback; radio protection and health; probabilistic safety assessment; summary and future plans

  4. Incidents in Czechoslovakian "Socialist Management" between 1956 and 1989. Conflict and Reconciliation

    Czech Academy of Sciences Publication Activity Database

    Vilímek, Tomáš; Tůma, Oldřich

    42/43, Spring-Summer (2016), s. 187-240 ISSN 1310-9456 Institutional support: RVO:68378114 Keywords : socialist management * centrally planned economy * Czechoslovakia 1956-1989 Subject RIV: AB - History

  5. VVER fuel cycle development at Slovakia

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  6. The integration of fast reactor to the fuel cycle in Slovakia

    International Nuclear Information System (INIS)

    Zajac, R.; Darilek, P.; Necas, V.

    2009-01-01

    A very topical problem of nuclear power is the fuel cycle back-end. One of the options is a LWR spent fuel reprocessing and a fissile nuclides re-use in the fast reactor. A large amount of spent fuel has been stored in the power plant intermediate storage during the operation of VVER-440 reactors in Slovakia. This paper is based on an analysis of Pu and minor actinides content in actual VVER-440 spent fuel stored in Slovakia. The next part presents the possibilities of reprocessing and Pu re-use in fast reactor under Slovak conditions. The fuel cycle consisting of the VVER-440 reactor, PUREX reprocessing plant and a sodium fast reactor was designed. The last section compares two parts of this fuel cycle: one is UOX cycle in VVER-440 reactor and the other is cycle in the fast reactor - SUPER PHENIX loaded with MOX fuel (Pu + Minor Actinides). The starting point is a single recycling of Pu from VVER-440 in the FR. The next step is multirecycling of Pu in the FR to obtain equilibrium cycle. This article is dealing with the solution of power production and fuel cycle indicators. All kinds of calculations were performed by computer code HELIOS 1.10. (authors)

  7. Lifelong Teacher Education in the Czechoslovakian Socialist Republic with Especial Reference to the Training of Teachers in Special Education.

    Science.gov (United States)

    Cerna, Marie

    1987-01-01

    The structure of the Czechoslovak system of preservice and inservice teacher education introduced by Education Act No. 29 of 1984 is described. The preparation of special education teachers is given particular emphasis. (Author/MT)

  8. 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

  9. Nuclear safety at the Paks Plant

    International Nuclear Information System (INIS)

    Bajsz, Jozsef; Vamos, Gabor

    1991-01-01

    The Paks Nuclear Power Plant is located on the Danube river 114 km south of Budapest. It consists of four PWR units of the Soviet VVER-440 design. These are of the second generation design VVER 440 (model 213) with safety features as of 1975. It should be emphasized that these are two different generations of VVER 440 units. This is not always clear, not only to the general public, but sometimes even to people working in the nuclear industry. The widespread criticism of the first generation type 230 reactors is often extended to model 213 reactors, as the differences between the two models are often not sufficiently emphasized. In this situation it is very important to provide balanced information about the advantages and disadvantages of this reactor type. This paper attempts to do that. (author)

  10. The further development of WWER-440 fuel design performance

    International Nuclear Information System (INIS)

    Lushin, V.; Vasilchenko, I.; Ananjev, J.; Abashina, G.

    2011-01-01

    The most distinguished stages in VVER-440 fuel development of the latest ten years are: designing of second generation FA complex; and designing of sheathless working fuel assembly of the third generation (RK-3) which are presented in this report. Designing of fuel assemblies of the second generation and RK-3 is characterized by the tendency to power increase of VVER-440 operating units with V-213-type reactor, that, in turn, has given a stimulus to further design enhancement of fuel assemblies specified. The further development of the second generation fuel assembly design and the change-over to the third generation working assemblies will allow for fuel utilization to be considerably increased under the conditions of application the more long-term fuel cycles for VVER-440 reactors and operation of the Units at the increased power

  11. Fuel leak testing performance at NPP Jaslovske Bohunice

    International Nuclear Information System (INIS)

    Slugen, V.; Krnac, S.; Smiesko, I.

    1995-01-01

    The NPP Bohunice VVER-440 fuel leak testing experience are relatively extensive in comparison with other VVER-440 users. As the first Europe NPP was adapted Siemens (KWU) in core-sipping equipment to VVER-440 units and since this time were have done these tests also for NPP Paks (Hungary) and NPP Dukovany (Czech Republic). The occurrence of leaking fuel assemblies in NPP is in the last 5 years relatively stabilised and low. A significant difference can be observed between type V-230 (31 leaks) and type V-213 (1 leak). None of of the indicated leaking fuel assemblies has been investigated in the hot cell. Therefore cannot be confirm the effective causes of leak occurrence. Nevertheless, the fuel failure rate and the performance of leak testing in NPP Bohunice are comparable to the world standard at PWR's. 1 tab., 2 figs., 3 refs

  12. Fuel leak testing performance at NPP Jaslovske Bohunice

    Energy Technology Data Exchange (ETDEWEB)

    Slugen, V; Krnac, S [Slovak Technical Univ., Bratislava (Slovakia); Smiesko, I [Nuclear Powr Plant EBO, Jaslovske Bohuce (Slovakia)

    1996-12-31

    The NPP Bohunice VVER-440 fuel leak testing experience are relatively extensive in comparison with other VVER-440 users. As the first Europe NPP was adapted Siemens (KWU) in core-sipping equipment to VVER-440 units and since this time were have done these tests also for NPP Paks (Hungary) and NPP Dukovany (Czech Republic). The occurrence of leaking fuel assemblies in NPP is in the last 5 years relatively stabilised and low. A significant difference can be observed between type V-230 (31 leaks) and type V-213 (1 leak). None of of the indicated leaking fuel assemblies has been investigated in the hot cell. Therefore cannot be confirm the effective causes of leak occurrence. Nevertheless, the fuel failure rate and the performance of leak testing in NPP Bohunice are comparable to the world standard at PWR`s. 1 tab., 2 figs., 3 refs.

  13. Advanced nondestructive examination of the reactor vessel head penetration tube welds

    International Nuclear Information System (INIS)

    Cvitanovic, M.; Zado, V.

    1996-01-01

    Beside a referent code examination requirements, appearance of the service induced flaws on the Reactor Vessel Head (RVH) penetration tube welds forced development of remotely operated examination tools and techniques. Several systems were developed for examination of RVH PWR type while only one system for examination of VVER - 440 type RVH has been developed by Inetec. In this article the most advanced RVH VVER - 440 type examination techniques such as ultrasonic, eddy current and visual testing techniques as well as remotely operated tool are described. (author)

  14. Validation experience with the core calculation program karate

    International Nuclear Information System (INIS)

    Hegyi, Gy.; Hordosy, G.; Kereszturi, A.; Makai, M.; Maraczy, Cs.

    1995-01-01

    A relatively fast and easy-to-handle modular code system named KARATE-440 has been elaborated for steady-state operational calculations of VVER-440 type reactors. It is built up from cell, assembly and global calculations. In the frame of the program neutron physical and thermohydraulic process of the core at normal startup, steady and slow transient can be simulated. The verification and validation of the global code have been prepared recently. The test cases include mathematical benchmark and measurements on operating VVER-440 units. Summary of the results, such as startup parameters, boron letdown curves, radial and axial power distributions of some cycles of Paks NPP is presented. (author)

  15. 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.

  16. Safety improvements made at the Loviisa nuclear power plant to reduce fire risks originating from the turbine generators

    International Nuclear Information System (INIS)

    Virolainen, T.; Marttila, J.; Aulamo, H.

    1998-01-01

    Comprehensive upgrading measures have been completed for the Loviisa Nuclear Power Plant (modified VVER440/V213). These were carried out from the start of the design phase and during operation to ensure safe plant shutdown in the event of a large turbine generator oil fire. These modifications were made mainly on a deterministic basis according to specific risk studies and fire analyses. As part of the probabilistic safety assessment, a fire risk analysis was made that confirmed the importance of these upgrading measures. In fact, they should be considered as design basis modifications for all VVER440 plants. (author)

  17. Third international seminar on horizontal steam generators

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-12-31

    The Third International Seminar on Horizontal Steam Generators held on October 18-20, 1994 in Lappeenranta, consisted of six sessions dealing with the topics: thermal hydraulic experiments and analyses, primary collector integrity, management of primary-to-secondary leakage accidents, feedwater collector replacement and discussion of VVER-440 steam generator safety issues.

  18. Third international seminar on horizontal steam generators

    International Nuclear Information System (INIS)

    1995-01-01

    The Third International Seminar on Horizontal Steam Generators held on October 18-20, 1994 in Lappeenranta, consisted of six sessions dealing with the topics: thermal hydraulic experiments and analyses, primary collector integrity, management of primary-to-secondary leakage accidents, feedwater collector replacement and discussion of VVER-440 steam generator safety issues

  19. Acoustic leak detection at complicated topologies using neural netwoks

    International Nuclear Information System (INIS)

    Hessel, G.; Schmitt, W.; Weiss, F.P.

    1994-01-01

    Considering the shortcomings of all the existing leak detecting principles, a new method again based on the measurement of the leak induced sound but also applying pattern recognition is being developed. The capability of neural networks to localize leaks at the reactor pressure vessel (RPV) head of VVER-440 reactors is discussed. (orig./DG)

  20. EBO feed water distribution system, experience gained from operation

    Energy Technology Data Exchange (ETDEWEB)

    Matal, O. [Energovyzkum, Brno (Switzerland); Schmidt, S.; Mihalik, M. [Atomove Elektrarne Bohunice, Jaslovske Bohunice (Switzerland)

    1997-12-31

    Advanced feed water distribution systems of the EBO design have been installed into steam generators at Units 3 and 4 of the NPP Jaslovske Bohunice (VVER 440). Experiences gained from the operation of steam generators with the advanced feed water distribution systems are discussed in the paper. (orig.). 4 refs.

  1. Expert evaluation in NPP safety important systems licensing process

    International Nuclear Information System (INIS)

    Mikhail, A Yastrebenetsky; Vasilchenko, V.N.

    2001-01-01

    Expert evaluation of nuclear power plant safety important systems modernization is an integral part of these systems licensing process. The paper contains some aspects of this evaluation which are based on Ukrainian experience of VVER-1000 and VVER-440 modernization. (authors)

  2. Third international seminar on horizontal steam generators

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-12-31

    The Third International Seminar on Horizontal Steam Generators held on October 18-20, 1994 in Lappeenranta, consisted of six sessions dealing with the topics: thermal hydraulic experiments and analyses, primary collector integrity, management of primary-to-secondary leakage accidents, feedwater collector replacement and discussion of VVER-440 steam generator safety issues.

  3. Press travel to the Mochovce power plant in Slovakia

    International Nuclear Information System (INIS)

    1998-07-01

    This press dossier gives, first, a general presentation of the Mochovce power plant of the republic of Slovakia (history, technical characteristics, safety aspects, technical particularities of VVER 440/213 reactors (safety analysis, recent improvements), financing of the modernization project). Then, the activities of the Framatome group in central and eastern Europe are presented (partnerships, cooperation agreements and contracts). (J.S.)

  4. EBO feed water distribution system, experience gained from operation

    Energy Technology Data Exchange (ETDEWEB)

    Matal, O [Energovyzkum, Brno (Switzerland); Schmidt, S; Mihalik, M [Atomove Elektrarne Bohunice, Jaslovske Bohunice (Switzerland)

    1998-12-31

    Advanced feed water distribution systems of the EBO design have been installed into steam generators at Units 3 and 4 of the NPP Jaslovske Bohunice (VVER 440). Experiences gained from the operation of steam generators with the advanced feed water distribution systems are discussed in the paper. (orig.). 4 refs.

  5. Expert evaluation in NPP safety important systems licensing process

    Energy Technology Data Exchange (ETDEWEB)

    Mikhail, A Yastrebenetsky; Vasilchenko, V.N. [Ukrainian State Scientific Technical Center of Nuclear and Radiation Safety (Ukraine)

    2001-07-01

    Expert evaluation of nuclear power plant safety important systems modernization is an integral part of these systems licensing process. The paper contains some aspects of this evaluation which are based on Ukrainian experience of VVER-1000 and VVER-440 modernization. (authors)

  6. Experimental and numerical investigation of the coolant mixing during fast deboration transients

    International Nuclear Information System (INIS)

    Hoehne, T.; Rohde, U.; Weiss, F.P.

    1999-01-01

    For the analysis of boron dilution transients and main steam line break scenarios the modeling of the coolant mixing inside the reactor vessel is important, because the reactivity insertion strongly depends on boron acid concentration or the coolant temperature distribution. Calculations for steady state flow conditions for the VVER-440 were performed with a CFD code (CFX-4). The comparison with experimental data and an analytical mixing model which is implemented in the neutron-kinetic code DYN3D showed a good agreement for near-nominal conditions. First experiments at the Rossendorf Mixing Test Facility ROCOM were performed simulating the start-up of the first main coolant pump. The reference reactor for the geometrically 1:5 scaled Plexiglas model is the German Konvoi type PWR. After demonstrating the capability of the CFD code to simulate these complicated flow transients, calculations were performed for the start-up of the first pump in a VVER-440 type reactor. These calculations are a first step of understanding the coolant mixing in the RPV of a VVER-440 type reactor under transient conditions. The results of the calculation show a very complex flow in the downcomer. A high downcomer of VVER-440 and the existence of the lower control rod chamber support coolant mixing is concluded. (author)

  7. Appraisal of the implementation status of appendix general design criteria, of the 10CFR50 in the design of the Juragua NPP

    International Nuclear Information System (INIS)

    Rodriguez Aleman, Carlos; Mitjans Sanchez, Guillermo

    1996-01-01

    The work intends to reflect in a general manner the state of accomplishment of the general design requirements for NPP in the US and other parts of the world contained in 10CFR50 appendix A in the design solutions adopted for the Juragua plant with reactors VVER-440 V-318

  8. 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

  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. Condensation driven water hammer studies for feedwater distribution pipe

    International Nuclear Information System (INIS)

    Savolainen, S.; Katajala, S.; Elsing, B.; Nurkkala, P.; Hoikkanen, J.; Pullinen, J.; Logvinov, S.A.; Trunov, N.B.; Sitnik, J.K.

    1997-01-01

    Imatran Voima Oy, IVO, operates two VVER 440 reactors. Unit 1 has been operating since 1977 and unit 2 since 1981. First damages of the feed water distribution (FWD) pipes were observed in 1989. In closer examinations FWD-pipe T-connection turned out to suffer from severe erosion corrosion damages. Similar damages have been found also in other VVER 440 type NPPs. In 1994 the first new FWD-pipe was replaced and in 1996 extensive water hammer experiments were carried out together with EDO Gidropress in Podolsk. After the first phase of the experiments some fundamental changes were made to the construction of the FWD-pipe. The object of this paper is to give short insight to the design of the new FWD-pipe concentrating on water hammer experiments. (orig.)

  11. 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.

  12. Condensation driven water hammer studies for feedwater distribution pipe

    Energy Technology Data Exchange (ETDEWEB)

    Savolainen, S.; Katajala, S.; Elsing, B.; Nurkkala, P.; Hoikkanen, J. [Imatran Voima Oy, Vantaa (Finland); Pullinen, J. [IVO Power Engineering Ltd., Vantaa (Finland); Logvinov, S.A.; Trunov, N.B.; Sitnik, J.K. [EDO Gidropress (Russian Federation)

    1997-12-31

    Imatran Voima Oy, IVO, operates two VVER 440 reactors. Unit 1 has been operating since 1977 and unit 2 since 1981. First damages of the feed water distribution (FWD) pipes were observed in 1989. In closer examinations FWD-pipe T-connection turned out to suffer from severe erosion corrosion damages. Similar damages have been found also in other VVER 440 type NPPs. In 1994 the first new FWD-pipe was replaced and in 1996 extensive water hammer experiments were carried out together with EDO Gidropress in Podolsk. After the first phase of the experiments some fundamental changes were made to the construction of the FWD-pipe. The object of this paper is to give short insight to the design of the new FWD-pipe concentrating on water hammer experiments. (orig.).

  13. 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)

  14. The effect of the volumetric heat source distribution of the fuel pellet on the minimum DNBR ratio

    International Nuclear Information System (INIS)

    Hordosy, G.; Kereszturi, A.; Maroti, L.; Trosztel, I.

    1995-01-01

    The radial power distribution in a VVER-440 type fuel assembly is strongly non-uniform as a result of the water-gap between the shrouds and the moderator filled central tube. Consequently, it can be expected that the power density inside a single fuel rod is inhomogeneous, as well. In the paper the methodology and the results of coupled thermohydraulic and neutronic calculations are presented. The objective of the analysis was the investigation of the heat source distribution and the determination of the possible extent of the power non-uniformity in a corner rod which has always the highest peaking factor in a VVER-440 type assembly. The results of the analysis revealed that there can be a strong non-uniformity of power distribution inside a fuel pellet, and the effect depends first of all on the general assembly conditions, while the local subchannel parameters have only a slight influence on the pellet heat source distribution. (author)

  15. On detection of the possible use of VVERs for unreported production of plutonium. Final report for the period July 1988 - December 1989

    International Nuclear Information System (INIS)

    Simov, R.; Nelov, N.; Stoyanova, I.; Kovachev, N.; Yonchev, P.

    1989-01-01

    The study includes an analysis of the feasibility of unreported production of plutonium-239 in VVER-440 reactors. It is shown that for VVER-440 reactors 36 natural uranium oxide fuel assemblies in the peripheral region of the core need to be loaded to produce 8 kg of extra plutonium in one cycle. Substituting the peripheral fuel assemblies with natural uranium oxide fuel assemblies, the changes in the power peaking are negligible and do not affect reactor safety. Unreported production outside the core is not practical due to physical and mechanical constraints, low flux level, etc. The feasibility of unreported removal of irradiated material in spent fuel cask has been also assessed. After about a month cooling time, still within the refueling period, the irradiated natural uranium fuel assemblies could be removed off-site without significant health hazard to the workers. To improve the effectiveness of the safeguards objectives, additional inspection activities are suggested. 10 figs

  16. Fatigue flaw growth assessment and inclusion of stratification to the LBB assessment

    Energy Technology Data Exchange (ETDEWEB)

    Samohyl, P.

    1997-04-01

    The application of the LBB requires also fatigue flaw growth assessment. This analysis was performed for PWR nuclear power plants types VVER 440/230, VVER 440/213c, VVER 1000/320. Respecting that these NPP`s were designed according to Russian codes that differ from US codes it was needed to compare these approaches. Comparison with our experimental data was accomplished, too. Margins of applicability of the US methods and their modifications for the materials used for construction of Czech and Slovak NPP`s are shown. Computer code accomplishing the analysis according to described method is presented. Some measurement and calculations show that thermal stratifications in horizontal pipelines can lead to additive loads that are not negligible and can be dangerous. An attempt to include these loads induced by steady-state stratification was made.

  17. Condensation driven water hammer studies for feedwater distribution pipe

    Energy Technology Data Exchange (ETDEWEB)

    Savolainen, S; Katajala, S; Elsing, B; Nurkkala, P; Hoikkanen, J [Imatran Voima Oy, Vantaa (Finland); Pullinen, J [IVO Power Engineering Ltd., Vantaa (Finland); Logvinov, S A; Trunov, N B; Sitnik, J K [EDO Gidropress (Russian Federation)

    1998-12-31

    Imatran Voima Oy, IVO, operates two VVER 440 reactors. Unit 1 has been operating since 1977 and unit 2 since 1981. First damages of the feed water distribution (FWD) pipes were observed in 1989. In closer examinations FWD-pipe T-connection turned out to suffer from severe erosion corrosion damages. Similar damages have been found also in other VVER 440 type NPPs. In 1994 the first new FWD-pipe was replaced and in 1996 extensive water hammer experiments were carried out together with EDO Gidropress in Podolsk. After the first phase of the experiments some fundamental changes were made to the construction of the FWD-pipe. The object of this paper is to give short insight to the design of the new FWD-pipe concentrating on water hammer experiments. (orig.).

  18. Formulae for thermal feedback of group constants in digital reactor simulation

    International Nuclear Information System (INIS)

    Perneczky, L.; Toth, I.; Vigassy, J.

    1976-01-01

    The problem, how the feedback of the thermohydraulic field to the neutron density in a reactor can be calculated is analysed. After a brief survey of the digital models in reactor simulation the applied model based on the time-dependent two-group diffusion equations is described. Using the reactor physical code system THERESA numerical results for the VVER-440 reactor are presented. (Sz.Z.)

  19. AER Working Group B activities in 2009

    International Nuclear Information System (INIS)

    Darilek, P.

    2009-01-01

    Regular meeting of Core Design Group was organized by SKODA JS a.s. in Plzen (Czech Republic) during the period of 4 to 6 May 2009. There were presented altogether 17 participants from 7 member organizations and 7 presentations were read. Presented papers covered topics as follows: 1) Two presentations dealt with upgrade of calculation and display tools. 2) Three presentations were devoted to benchmark calculations. 3) Two presentations informed about gradual improvement of fuel assembly and cycle for VVER-440 reactors

  20. Evaluation gives the activity inventory the nuclear fuel irradiated and its radioactive waste

    International Nuclear Information System (INIS)

    Rodriguez Gual, Maritza

    1998-01-01

    The present work has as objectives to give a quantitative evaluation to the activity that possesses the nuclear fuel for 3,6% enrichment with a burnt one the 33 000 NWd/Tu proposed for the Juragua Nuclear Power Plant . In this work the method is used I calculate ORIGEN2. Obtained results are presented and they are compared with other calculations carried out in reactors type VVER-440

  1. Safety upgrading of the PAKS Nuclear Plant

    International Nuclear Information System (INIS)

    Vamos, G.; Vigassy, J.

    1993-01-01

    In the last several years the net electricity from the Paks NPP represents almost half of the Hungarian total. The 4 units of Paks belong to the latest generation of the VVER-440 units, the small-sized Russian designed PWRs. Reviewing the main design features of them, the safety merits and safety concerns are summarized. Due to the conservative design and the extensive operating experience the safety merits appear to be more significant than generally believed. The VVER-440 type has two models, the 230 and 213, which have a large number of distinctive safety features. These are highlighted in the section comparisons. A quality assurance program was initiated in Paks very early. A long-term safety upgrading program was also initiated, originating from vendor recommendations, regulatory decisions, in-house operating experience and safety concerns, and independent reviews. The main areas and some examples of the measures are described. This program, like all other activities related to nuclear safety, has been under regulatory control. The specific features of the Hungarian regulatory system are described. For advanced, general and new evaluation of the safety of the units in Paks in accordance with the internationally recommended criteria of the 90's, the project AGNES has been launched with international participation. The scope of this project is summarized. International efforts as the IAEA Regional Project on safety assessment of VVER-440/213 and VVER-440/230 units are underway. Since safety is not only a question of design, but it can be significantly influenced by operations and maintenance practices, the Paks NPP has invited LAEA's OSART and ASSET missions, WANO's Pilot Peer Review

  2. Assessment of the event at Rovno NPP owing to the unexpected opening and obstruction of the safety valve lock mechanism of the pressure compensator

    International Nuclear Information System (INIS)

    Alonso, C.

    1993-01-01

    The main objective of this analysis is to be able to identify the sequence of the accident and evaluate the real frequency observed in similar nuclear power plants, according to the experience registered in the data bases of the NPP accident information system (ISI-AES) as well as the quality assurance information system (ISKO). This work describes how the analysis of events in a VVER-440 reactor NPP was performed

  3. Structural seismic upgrading of NPPs in Czech and Slovak republics

    Energy Technology Data Exchange (ETDEWEB)

    David, M [DAVID Consulting, Engineering and Design Office, Prague (Czech Republic)

    1997-03-01

    Several Nuclear Power Plants of the VVER type has been constructed during the past years in former Czechoslovak Republic. Some of them has been already put in operation and some of them are under construction. Nuclear Power Plants V1(2 units of VVER 440/230), V2(2 units of VVER 440/213) in Slovak and NPP Dukovany (4 units of VVER 440/213) in Czech republic are in operation. NPP Mochovce (4 units of VVER 440/213) in Slovak and NPP Temelin (4 units reduced now to 2 units VVER 1000) have been already almost completed, but still under construction. All above cited NPPs have not been either explicitly designed against earthquake or the design against earthquake or its input data must be upgraded to be compatible with present requirements. The upgrading of seismic input as well the seismic upgrading of all structures and technological equipments for so many NPPs has involved a lot of comprehensive work in Czech as well as in Slovak republics. The upgrading cannot be completed in a short time and as a rule the seismic upgrading has been usually performed in several steps, beginning with the most important arrangements against seismic hazard. The basic principles and requirements for seismic upgrading has been defined in accordance with the international and particularly with the IAEA recommendations. About the requirements for seismic upgrading of NPPs in Czech and Slovak republics will be reported in other contribution. This contribution is dealing with the problems of seismic upgrading of NNPs civil engineering structures. The aim of this contribution is to point out some specific problems connected firstly with very complicated concept of Versa structures and secondly with the difficult task to increase the structural capacity to the required seismic level. (J.P.N.)

  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. Assessment of financial expenditure for Rivne NPP power units decommissioning

    International Nuclear Information System (INIS)

    Nosovskij, A.V.; Salij, L.M.

    2007-01-01

    The article covers some financial aspects of developing a decommissioning concept for Rivne NPP power units with reactor VVER-440 and VVER-1000. Possible methodological approaches to costs estimate have been analyzed. Preliminary results of cost estimation are presented for two decommissioning options: deferred and immediate dismantling. Principally possible options for accumulating assets have been analyzed to finance measures related to Rivne NPP decommissioning. A mathematical model has been proposed for creating decommissioning financial reserve

  6. Structural seismic upgrading of NPPs in Czech and Slovak republics

    International Nuclear Information System (INIS)

    David, M.

    1997-01-01

    Several Nuclear Power Plants of the VVER type has been constructed during the past years in former Czechoslovak Republic. Some of them has been already put in operation and some of them are under construction. Nuclear Power Plants V1(2 units of VVER 440/230), V2(2 units of VVER 440/213) in Slovak and NPP Dukovany (4 units of VVER 440/213) in Czech republic are in operation. NPP Mochovce (4 units of VVER 440/213) in Slovak and NPP Temelin (4 units reduced now to 2 units VVER 1000) have been already almost completed, but still under construction. All above cited NPPs have not been either explicitly designed against earthquake or the design against earthquake or its input data must be upgraded to be compatible with present requirements. The upgrading of seismic input as well the seismic upgrading of all structures and technological equipments for so many NPPs has involved a lot of comprehensive work in Czech as well as in Slovak republics. The upgrading cannot be completed in a short time and as a rule the seismic upgrading has been usually performed in several steps, beginning with the most important arrangements against seismic hazard. The basic principles and requirements for seismic upgrading has been defined in accordance with the international and particularly with the IAEA recommendations. About the requirements for seismic upgrading of NPPs in Czech and Slovak republics will be reported in other contribution. This contribution is dealing with the problems of seismic upgrading of NNPs civil engineering structures. The aim of this contribution is to point out some specific problems connected firstly with very complicated concept of Versa structures and secondly with the difficult task to increase the structural capacity to the required seismic level. (J.P.N.)

  7. 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

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

    International Nuclear Information System (INIS)

    Masopust, R.

    2003-01-01

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

  9. The next 20 years operation of the 36 years old Hungarian training reactor

    International Nuclear Information System (INIS)

    Aszodi, A.

    2007-01-01

    Hungary prepares for extending the design lifetime of the four VVER-440/213 type units; in that case they will finish operation between 2032 and 2037. Discussion on possible new nuclear units in Hungary was recently commenced. The paper describes actions in human resource management and knowledge management, and also the new safety analysis methods which were applied during the recent Periodic Safety Review of the Hungarian Training Reactor

  10. Shallow crack effect on brittle fracture of RPV during pressurised thermal shock

    International Nuclear Information System (INIS)

    Ikonen, K.

    1995-12-01

    This report describes the study on behaviour of postulated shallow surface cracks in embrittled reactor pressure vessel subjected to pressurised thermal shock loading in an emergency core cooling. The study is related to the pressure vessel of a VVER-440 type reactor. Instead of a conventional fracture parameter like stress intensity factor or J integral the maximum principal stress distribution on a crack tip area is used as a fracture criteria. The postulated cracks locate circumferentially at the inner surface of the reactor pressure wall and they penetrate the cladding layer and open to the inner surface. Axisymmetric and semielliptical crack shapes were studied. Load is formed of an internal pressure acting also on crack faces and of a thermal gradient in the pressure vessel wall. Physical properties of material and loading data correspond real conditions in VVER-440 RPV. The study was carried out by making lot of 2D- and 3D- finite element calculations. Analysing principles and computer programs are explained. Except of studying the shallow crack effect, one objective of the study has also been to develop further expertise and the in-house developed computing system to make effectively elastic-plastic fracture mechanical analyses for real structures under complicated loads. Though the study concerns VVER-440 RPV, the results are of more general interest especially related to thermal loads. (orig.) (11 refs.)

  11. MEASUREMENTS OF THE CONFINEMENT LEAKTIGHTNESS AT THE KOLA NUCLEAR POWER STATION (UNIT 2) IN RUSSIA

    International Nuclear Information System (INIS)

    GREENE, G.A.; GUPPY, J.G.

    1998-01-01

    This is the final report on the INSP project entitled, ''Kola Confinement Leaktightness'' conducted by BNL under the authorization of Project Work Plan WBS 1.2.2.1. This project was initiated in February 1993 to assist the Russians to reduce risks associated with the continued operation of older Soviet-designed nuclear power plants, specifically the Kola VVER-440/230 Units 1 and 2, through upgrades in the confinement performance to reduce the uncontrolled leakage rate. The major technical objective of this-project was to improve the leaktightness of the Kola NPP VVER confinement boundaries, through the application of a variety of sealants to penetrations, doors and hatches, seams and surfaces, to the extent that current technology permitted. A related objective was the transfer, through training of Russian staff, of the materials application procedures to the staff of the Kola NPP. This project was part of an overall approach to minimizing uncontrolled releases from the Kola NPP VVER440/230s in the event of a serious accident, and to thereby significantly mitigate the consequences of such an accident. The US provided materials, application technology, and applications equipment for application of sealant materials, surface coatings, potting materials and gaskets, to improve the confinement leaktightness of the Kola VVER-440/23Os. The US provided for training of Russian personnel in the applications technology

  12. Large eddy simulation of a fuel rod subchannel

    International Nuclear Information System (INIS)

    Mayer, Gusztav

    2007-01-01

    In a VVER-440 reactor the measured outlet temperature is related to fuel limit parameters and the power upgrading plans of VVER-440 reactors motivated us to obtain more information on the mixing process of the fuel assemblies. In a VVER-440 rod bundle the fuel rods are arranged in triangular array. Measurement shows (Krauss and Meyer, 1998) that the classical engineering approach, which tries to trace the characterization of such systems back to equivalent (hydraulic diameter) pipe flows, does not give reasonable results. Due to the different turbulence characteristics, the mixing is more intensive in rod bundles than it would be expected based on equivalent pipe flow correlations. As a possible explanation of the high mixing, secondary flow was deduced from measurements by several experimentalists (Trupp and Azad, 1975). Another candidate to explain the high mixing is the so-called flow pulsation phenomenon (Krauss and Meyer, 1998). In this paper we present subchannel simulations (Mayer et al. 2007) using large eddy simulation (LES) methodology and the lattice Boltzmann method (LBM) without the spacers at Reynolds number 21000. The simulation results are compared with the measurements of Trupp and Azad (1975). The mean axial velocity profile shows good agreement with the measurement data. Secondary flow has been observed directly in the simulation results. Reasonable agreement has been achieved for most Reynolds stresses. Nevertheless, the calculated normal stresses show small, but systematic deviation from the measurement data. (author)

  13. Numerical Boron mixing studies for Loviisa Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Gango, P. [IVO International Ltd. (Finland)

    1995-09-01

    A program has been started for studying numerically boron mixing in the downcomer of Loviisa NPP (VVER-440). Mixing during the transport of a diluted slug from the loop to the core might serve as an inherent protection mechanism against severe reactivity accidents in inhomogenous boron dilution scenarios for PWRs. The commercial general purpose Computational Fluid Dynamics (CFD) core PHOENICS is used for solving the governing fluid flow equations in the downcomer geometry of VVER-440. So far numerical analyses have been performed for steady state operation conditions and two different pump driven transients. The steady state analyses focused on model development and validation against existing experimental data. The two pump driven transient scenarios reported are based on slug transport during the start of the sixth and first loop respectively. The results from the two transients show that mixing is case and plant specific; the high and open downcomer geometry of VVER-440 seems to be advantageous from mixing point of view. In addition the analyzing work for the {open_quotes}first pump start{close_quotes} scenario brought up some considerations about flow distribution in the existing experimental facilities.

  14. Implementation of the neutronics model of HEXTRAN/HEXBU-3D into APROS for WWER calculations

    International Nuclear Information System (INIS)

    Rintala, J.

    2008-01-01

    A new three-dimensional nodal model for neutronics calculation is currently under implementation into APROS - Advanced PROcess Simulation environment - to conform the increasing accuracy requirements. The new model is based on an advanced nodal code HEXTRAN and its static version HEXBU-3D by VTT, Technical Research Centre of Finland. Currently the new APROS is under a testing programme. Later a systematic validation will be performed. In the first phase, a goal is to obtain a fully validated model for VVER-440 calculations. Thus, all the current test calculations are performed by using Loviisa NPP's VVER-440 model of APROS. In future, the model is planned to be applied for the calculations of VVER-1000 type reactors as well as in rectangular fuel geometry. The paper outlines first the general aspects of the method, and then the current situation of the implementation. Because of the identical model with the models of HEXTRAN and HEXBU-3D, the results in the test calculations are compared to the results of those. In the paper, results of two static test calculations are shown. Currently the model works well already in static analyses. Only minor problems with the control assemblies of VVER-440 type reactor still exist but the reasons are known and will be corrected in near future. Dynamical characteristics of the model are up to now tested only by some empirical tests. (author)

  15. MEASUREMENTS OF THE CONFINEMENT LEAKTIGHTNESS AT THE KOLA NUCLEAR POWER STATION (UNIT 2) IN RUSSIA

    Energy Technology Data Exchange (ETDEWEB)

    GREENE,G.A.; GUPPY,J.G.

    1998-08-01

    This is the final report on the INSP project entitled, ``Kola Confinement Leaktightness'' conducted by BNL under the authorization of Project Work Plan WBS 1.2.2.1. This project was initiated in February 1993 to assist the Russians to reduce risks associated with the continued operation of older Soviet-designed nuclear power plants, specifically the Kola VVER-440/230 Units 1 and 2, through upgrades in the confinement performance to reduce the uncontrolled leakage rate. The major technical objective of this-project was to improve the leaktightness of the Kola NPP VVER confinement boundaries, through the application of a variety of sealants to penetrations, doors and hatches, seams and surfaces, to the extent that current technology permitted. A related objective was the transfer, through training of Russian staff, of the materials application procedures to the staff of the Kola NPP. This project was part of an overall approach to minimizing uncontrolled releases from the Kola NPP VVER440/230s in the event of a serious accident, and to thereby significantly mitigate the consequences of such an accident. The US provided materials, application technology, and applications equipment for application of sealant materials, surface coatings, potting materials and gaskets, to improve the confinement leaktightness of the Kola VVER-440/23Os. The US provided for training of Russian personnel in the applications technology.

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

    International Nuclear Information System (INIS)

    Horacek, L.

    2002-01-01

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

  17. Slovak Republic Act of 11 February 1998 on the energetics and on alterations to Act No. 455/1991 Collection of Acts of CSFR on small business (trade Act) in version of posterior regulations

    International Nuclear Information System (INIS)

    1998-01-01

    This act constitute: (a) conditions of undertaking in electro-energetic, gas industry, and heat supply (in next only 'energetic' branches) ; (b) rights and responsibility of physical and act person undertaking in energetic branches; (c) rights and responsibility of customers of electricity, gas, and heat; counteract measures in the need situations, (d) and at prevention before need situations in energetic branches; (e) state regulation in energetic; (f) authority on keep of this act. The act is divided into for parts: (1) General constitutions, (2) Energetic branches; (3) The state authority; (4) Common, transient and invalidation constitutions.This act deals with the specific conditions for undertaking in nuclear power plants, too (licensing, security). This act shall into effect on 1 July 1998

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

    International Nuclear Information System (INIS)

    Evans, M.; Petri, M. C.

    2003-01-01

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

  19. La política exterior republicana en los informes diplomáticos checoslovacos (1931-1936 = The Republican Foreign Policy in the Czechoslovakian Diplomatic Reports (1931-1936.

    Directory of Open Access Journals (Sweden)

    Luis Montilla Amador

    2016-05-01

    Full Text Available Este artículo es una aproximación al estudio de las relaciones bilaterales entre la II República española y Checoslovaquia. La joven nación centroeuropea saludó con entusiasmo la llegada del nuevo régimen español, sensación que se vio confirmada por las primeras actuaciones exteriores republicanas, muy próximas al espíritu con el que Praga afrontaba su política internacional. Esta cercanía de intereses se vio materializada en Ginebra con la formación del Grupo de los Ocho. Los tres responsables de la diplomacia checoslovaca en Madrid durante este periodo (Vlastimil Kybal, Zdeněk Formánek y Robert Flieder reflejaron en los informes remitidos a Praga los temas más importantes de la política exterior republicana. This research is an approach to the study of the bilateral relations between the Second Spanish Republic and Czechoslovakia.The young Central European nation greeted enthusiastically the arrival of the new Spanish regime. This good feeling was confirmed by the new Spanish way of dealing with the foreign affairs, very close to the spirit in which Prague was facing its international policy. As a result of this common interests was created in Geneva the Group of Eight, which included both countries.The three responsible persons for the Czecoslovakian diplomacy in Madrid during this period (Vlastimil Kybal, Zdeněk Formánek and Robert Flieder reflected in the reports submitted to Prague the most important issues of the Republican foreign policy.

  20. Nuclear plant analyzer program for Bulgaria

    International Nuclear Information System (INIS)

    Shier, W.; Kennett, R.

    1993-01-01

    An interactive nuclear plant analyzer(NPA) has been developed for use by the Bulgarian technical community 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. The current NPA includes models for a VVER-440 Model 230 and a VVER-1000 Model 320 and is operational on an IBM RISC6000 workstation. The RELAP5/MOD2 computer code has been used for the calculation of the reactor responses to the interactive commands initiated by the NPA operator. The interactive capabilities of the NPA have been developed to provide considerable flexibility in the plant actions that can be initiated by the operator. The current capabilities for both the VVER-440 and VVER-1000 models include: (1) scram initiation; (2) reactor coolant pump trip; (3) high pressure safety injection system initiation; (4) low pressure safety injection system initiation; (5) pressurizer safety valve opening; (6) steam generator relief/safety valve opening; (7) feedwater system initiation and trip; (8) turbine trip; and (9) emergency feedwater initiation. The NPA has the capability to display the results of the simulations in various forms that are determined by the model developer. Results displayed on the reactor mask are shown 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 temperature of other metal structures. In addition, changes in the status of various components and systems can be initiated and/or displayed both numerically and graphically on the mask. This paper provides a description of the structure of the NPA, a discussion of the simulation models used for the VVER-440 and the VVER-1000, and an overview of the NPA capabilities. Typical results obtained using both simulation models will be discussed

  1. Expression of granulocyte colony-stimulating factor receptor correlates with prognosis in oral and mesopharyngeal carcinoma.

    Science.gov (United States)

    Tsuzuki, H; Fujieda, S; Sunaga, H; Noda, I; Saito, H

    1998-02-15

    Granulocyte colony-stimulating factor receptors (G-CSFRs) have been observed on the surface of not only hematopoietic cells but also several cancer cells. The stimulation of G-CSF has been demonstrated to induce proliferation and activation of G-CSFR-positive cells. In this study, we investigated the expression of G-CSFR on the surface of tumor cells and G-CSF production in oral and mesopharyngeal squamous cell carcinoma (SCC) by an immunohistochemical approach. Of 58 oral and mesopharyngeal SCCs, 31 cases (53.4%) and 36 cases (62.1%) were positive for G-CSFR and G-CSF, respectively. There was no association between G-CSFR expression and G-CSF staining. In the group positive for G-CSFR expression, relapse was significantly more likely after primary treatment (P = 0.0069), whereas there was no association between G-CSFR expression and age, sex, tumor size, lymph node metastasis, and clinical stage. Also, the G-CSFR-positive groups had a significantly lower disease-free and overall survival rate than the G-CSFR-negative groups (P = 0.0172 and 0.0188, respectively). However, none of the clinical markers correlated significantly with G-CSF staining, nor did the status of G-CSF production influence the overall survival. The results imply that assessment of G-CSFR may prove valuable in selecting patients with oral and mesopharyngeal SCC for aggressive therapy.

  2. Experiences from Loviisa Nuclear Power Station concerning the decontamination of steam generators and primary system components

    International Nuclear Information System (INIS)

    Jaernstroem, R.

    1989-01-01

    Loviisa 1 and 2 are 465 MWe PWR units of the Soviet type VVER-440. Loviisa 1 has been in commercial operation since spring 1977 and Loviisa 2 from the beginning of 1980. Decontamination of primary circuit components - even big ones as steam generators - can be performed in an efficient and quick way with good results and resonable expences. Total costs for decontamination of the two steam generators including planning, construction, documentation, operation, chemicals etc. did not rise above 100,000.00 dollars. (author) 6 figs., 2 tabs

  3. Analysis of the COLIMA CA-U3 test using the ELSA module of ASTEC

    International Nuclear Information System (INIS)

    Godin-Jacqmin, L.; Journeau, C.; Piluso, P.

    2006-01-01

    The main purpose of this study is to calculate the COLIMA CA-U3 experimental test with the ELSA module of the ASTEC code. This experimental test was performed to represent the fission product and structural material releases from a VVER-440 magma type configuration. Thus, some additional work is also done on test result analyses and corresponding ASTEC parameter usage to model as closely as possible the test configuration. Code results on fission product releases are compared to experimental results in a qualitative way for all elements that can be evaluated by the ASTEC code. Sensitivity cases are also performed on the gas flow rate carrying the fission product. (author)

  4. Thermal-hydraulics of PGV-4 water volume during damage of the feedwater collector nozzles

    Energy Technology Data Exchange (ETDEWEB)

    Logvinov, S.A.; Titov, V.F. [OKB Gidropress (Russian Federation); Notaros, U.; Lenkei, I. [NPP Paks (Hungary)

    1995-12-31

    A number of VVER-440 plants has experienced the distributing nozzles of feedwater collector being damaged due to corrosion-erosion wearing. Such phenomenon could result in feedwater redistribution within the SG inventory with undesirable consequences. The collector with damaged nozzles has to be replaced but a certain time is needed for the preparatory works. The main objective of the investigation conducted is to assess if the safe operation of SG is possible before collector replacement. It was shown that the nozzle damage as observed did not result in the dangerous disturbances of thermobydraulics as compared with the conditions existing at the initial period of operation. (orig.).

  5. Design report of the canister for nuclear fuel disposal

    International Nuclear Information System (INIS)

    Raiko, H.; Salo, J.P.

    1996-12-01

    The report provides a summary of the design of the canister for final disposal of nuclear fuel. The canister structure consists of a cylindrical massive nodular graphite cast iron insert covered by a 50 mm thick copper overlay. The capacity of the canister is 11 assemblies of BWR or VVER 440 fuel. The canister shall be tight with a high probability for about 100 000 years. The design basis of the canister is set, the performed analyses are summarised and the results are assessed and discussed in the report. (26 refs.)

  6. 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.)

  7. SKODA JS in Central and Eastern European markets

    International Nuclear Information System (INIS)

    Luptacik, P.

    2006-01-01

    During the past decade, SKODA JS a.s. gained a leading position in the manufacture of equipment for the nuclear power sector in Central and Eastern Europe. The company benefited from its experience accumulated during the construction of VVER 440 and VVER 1000 NPPs at a time that saw a global decrease in popularity of the nuclear power sector. The company is building its success on a comprehensive understanding of the function of the primary coolant circuit and efforts to upgrade the equipment and processes so as to reflect new requirements in terms of safety and reactor unit service life. (orig.)

  8. Criticality safety calculations for the nuclear waste disposal canisters

    International Nuclear Information System (INIS)

    Anttila, M.

    1996-12-01

    The criticality safety of the copper/iron canisters developed for the final disposal of the Finnish spent fuel has been studied with the MCNP4A code based on the Monte Carlo technique and with the fuel assembly burnup programs CASMO-HEX and CASMO-4. Two rather similar types of spent fuel disposal canisters have been studied. One canister type has been designed for hexagonal VVER-440 fuel assemblies used at the Loviisa nuclear power plant (IVO canister) and the other one for square BWR fuel bundles used at the Olkiluoto nuclear power plant (TVO canister). (10 refs.)

  9. Probabilistic assessment of Juragua Nuclear Power Plant response under station blackout conditions

    International Nuclear Information System (INIS)

    Valhuerdi, C.; Vilaragut, J.J.; Perdomo, M.; Torres, A.

    1995-01-01

    The preliminary results concerning the response of station blackout are shown in this paper. These results have been obtained in the framework of initiator lass of external electrical supply as a aport of the revision o of the current probabilistic safety analysis. The work is also based on the results reported in the thermohydraulic calculations of VVER 440 plants responses under these conditions and the experience of this type of notified incidents. Finally, a comparative analysis with the results obtained for other reactor technologies is presented

  10. Space dependence of reactivity parameters on reactor dynamic perturbation measurements

    International Nuclear Information System (INIS)

    Maletti, R.; Ziegenbein, D.

    1985-01-01

    Practical application of reactor-dynamic perturbation measurements for on-power determination of differential reactivity weight of control rods and power coefficients of reactivity has shown a significant dependence of parameters on the position of outcore detectors. The space dependence of neutron flux signal in the core of a VVER-440-type reactor was measured by means of 60 self-powered neutron detectors. The greatest neutron flux alterations are located close to moved control rods and in height of the perturbation position. By means of computations, detector positions can be found in the core in which the one-point model is almost valid. (author)

  11. ISP 33. OECD/NEA/CSNI International Standard Problem n. 33. Pactel natural circulation stepwise coolant inventory reduction experiment. Comparison report. Volume 1 + 2

    International Nuclear Information System (INIS)

    Purhonen, H.; Kouhia, J.; Holmstrom, H.

    1994-12-01

    This is the comparison report of the CSNI ISP n.33, which is based on a natural circulation experiment with various coolant inventories conducted in Pactel facility (Finland), a 1/305 volumetrically scaled, full-height simulator of a Russian type VVER-440 pressurized water reactor. It presents all submitted blind calculational results from different countries, using various codes (Athlet, Cathare2, etc.) and compares them with the experimental data. The Pactel facility and the ISP 33 experiment are described, and the summaries of the participants, the computer codes and the nodalizations used for the blind calculations are presented

  12. Thermal-hydraulics of PGV-4 water volume during damage of the feedwater collector nozzles

    Energy Technology Data Exchange (ETDEWEB)

    Logvinov, S A; Titov, V F [OKB Gidropress (Russian Federation); Notaros, U; Lenkei, I [NPP Paks (Hungary)

    1996-12-31

    A number of VVER-440 plants has experienced the distributing nozzles of feedwater collector being damaged due to corrosion-erosion wearing. Such phenomenon could result in feedwater redistribution within the SG inventory with undesirable consequences. The collector with damaged nozzles has to be replaced but a certain time is needed for the preparatory works. The main objective of the investigation conducted is to assess if the safe operation of SG is possible before collector replacement. It was shown that the nozzle damage as observed did not result in the dangerous disturbances of thermobydraulics as compared with the conditions existing at the initial period of operation. (orig.).

  13. Flow in rod bundles

    International Nuclear Information System (INIS)

    Hazi, G.; Mayer, G.

    2005-01-01

    For power upgrading VVER-440 reactors we need to know exactly how the temperature measured by the thermocouples is related to the average outlet temperature of the fuel assemblies. Accordingly, detailed knowledge on mixing process in the rod bundles and in the fuel assembly head have great importance. Here we study the hydrodynamics of rod bundles based on the results of direct numerical and large eddy simulation of flows in subchannels. It is shown that secondary flow and flow pulsation phenomena can be observed using both methodologies. Some consequences of these observations are briefly discussed. (author)

  14. Core design experience of WWER-440 reactors when they working on increased power level

    International Nuclear Information System (INIS)

    Adeev, V.; Panov, A.; Melenchuk, I.

    2015-01-01

    The Kola NPP continues commercial operation of 2nd generation fuel (FA-2) and trial operation of 3rd generation fuel (FA-3), which has a number of design features providing the best operational characteristics. This report gives the results of VVER-440 core operation with FA-2 and FA-3 with enrichment increased up to 4.87%, and at the power level uprated to 107% of nominal power level. Brief analysis of obtained data is carried out. Peculiarities and techniques of developing loading patterns with new types of nuclear fuel for operation at the uprated power level are reviewed. (authors)

  15. Calculation of the 5th AER dynamic benchmark with APROS

    International Nuclear Information System (INIS)

    Puska, E.K.; Kontio, H.

    1998-01-01

    The model used for calculation of the 5th AER dynamic benchmark with APROS code is presented. In the calculation of the 5th AER dynamic benchmark the three-dimensional neutronics model of APROS was used. The core was divided axially into 20 nodes according to the specifications of the benchmark and each six identical fuel assemblies were placed into one one-dimensional thermal hydraulic channel. The five-equation thermal hydraulic model was used in the benchmark. The plant process and automation was described with a generic VVER-440 plant model created by IVO PE. (author)

  16. Mechanical fragmentation of nuclear reactor fuel assemblies by the double cutting method

    International Nuclear Information System (INIS)

    Voitsekhovskii, B.V.; Istomin, V.L.; Mitrofanov, V.V.

    1995-01-01

    A method is described for cutting a spent fuel assembly with straight shears into pieces of a prescribed size. The method does not require separation of the casing and the lattices. The double cutting method is briefly described, and experiments designed for cutting BN-350 and VVER-440 fuel assemblies are outlined. The testing showed that the cutting method was suitable for mechanical polarization of fuel assemblies. The investigations led to the development of turnkey industrial equipment for cutting spent fuel assemblies of different geometries with a maximum size up to 170 mm. 6 refs., 8 figs., 1 tab

  17. BNFL Springfields Fuel Division

    International Nuclear Information System (INIS)

    Tarkiainen, S.; Plit, H.

    1998-01-01

    The Fuel Division of British Nuclear Fuels Ltd (BNFL) manufactures nuclear fuel elements for British Magnox and AGR power plants as well as for LWR plants. The new fuel factory - Oxide Fuel Complex (OFC), located in Springfields, is equipped with modern technology and the automation level of the factory is very high. With their quality products, BNFL aims for the new business areas. A recent example of this expansion was shown, when BNFL signed a contract to design and license new VVER-440 fuel for Finnish Loviisa and Hungarian Paks power plants. (author)

  18. Contribution of the Slovak University of Technology Bratislava to the Education of NPP Operation Staff in Slovakia

    International Nuclear Information System (INIS)

    Hascik, J.; Slugen, V.; Hinca, R.; Miglierini, M.

    2006-01-01

    Paper is focused on the preparation of NPP VVER -440 staff in Slovak conditions. The realisation is managed via special technical courses, seminars, workshops, and trainings on selected experimental facilities at domestic as well as international level. Post-gradual re-qualification study: Safety aspects of NPP operation is discussed in detail. Six-year experience with NPP operating staff education can be shared and recommended also at international level. Based on these courses, special training for optimal preparation of NPP supervising physicists was started in 2002. In addition to all our activities, the international course: Safety aspects of NPP operation for subcontractors was prepared and realised in 2005.(author)

  19. A proposal of a benchmark for calculation of the power distribution next to the absorber

    International Nuclear Information System (INIS)

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

    1999-01-01

    A proposal of a new benchmark problem was formulated to consider the characteristics of the VVER-440 fuel assembly with enrichment zoning, i. e. to study the space dependence of the power distribution near to a control assembly. A quite detailed geometry and the material composition of the fuel and the control assemblies were modeled by the help of MCNP calculations in AEKI. The results of the MCNP calculations were built in the KARATE code system as the new albedo matrices. The comparison of the KARATE calculation results and the MCNP calculations for this benchmark is presented. (Authors)

  20. Comparison of PWR-IMF and FR fuel cycles

    International Nuclear Information System (INIS)

    Darilek, Petr; Zajac, Radoslav; Breza, Juraj; Necas, Vladimir

    2007-01-01

    The paper gives a comparison of PWR (Russia origin VVER-440) cycle with improved micro-heterogeneous inert matrix fuel assemblies and FR cycle. Micro-heterogeneous combined assembly contains transmutation pins with Pu and MAs from burned uranium reprocessing and standard uranium pins. Cycle analyses were performed by HELIOS spectral code and SCALE code system. Comparison is based on fuel cycle indicators, used in the project RED-IMPACT - part of EU FP6. Advantages of both closed cycles are pointed out. (authors)

  1. 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

  2. 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

  3. Severe accident phenomena

    International Nuclear Information System (INIS)

    Boneham, P.

    2003-01-01

    In this paper the following issues are considered: containment failure modes; induced RCS ruptures; overview of containment challenges; hydrogen detonations; direct containment heating; long term containment overpressure; cavity issues in VVER-440/213; phenomenology related to in-vessel retention. The paper also presents the measured and calculated detonation cell width of the H 2 -air-steam mixtures for 100 0 C and constant air density (41.6 mole/m 3 ); dependence of result class on mixture and geometric classes; experimental data on turbulent jet initiation and Deflagration to Detonation Transition (DDT) in confined volumes are also presented

  4. Installation and testing of the ERANOS computer code for fast reactor calculations

    International Nuclear Information System (INIS)

    Gren, Milan

    2010-12-01

    The French ERANOS computer code was acquired and tested by solving benchmark problems. Five problems were calculated: 1D XZ Model, 1D RZ Model, 3D HEX SNR 300 reactor, 2S HEX and 3D HEX VVER 440 reactor. The multi-group diffuse approximation was used. The multiplication coefficients were compared within the first problem, neutron flux density in the calculation points was obtained within the second problem, and powers in the various reactor areas and in the assemblies were calculated within the remaining problems. (P.A.)

  5. Mochovce NPP safety improvement and completion

    International Nuclear Information System (INIS)

    1997-01-01

    6th Nuclear society information meeting dealt with the completion of the Mochovce NPP with regard to implementation of safety measures. It was aimed to next problems: I. 'Survey' presentation on the situation of the nuclear power industry in partner countries; II. Basic technical presentations; III. Presentations of operators of the other VVER 440/213 NPPs on their activities in the field of safety improvement in relation to IAEA recommendations; IV. Technical solutions of safety improvements ranked with IAEA degree 3 (Report SC 108 VVER); V: Technical solutions of selected Safety Measures ranked with IAEA degree 2 and 1 (Report SC 108 VVER)

  6. Problems of the Murmansk Region and adjacent areas. Problemy energetiki Murmanskoy oblasti i sosednykh rayonov

    Energy Technology Data Exchange (ETDEWEB)

    Stepanov, I.R.

    1980-01-01

    The anthology contains articles devoted to complex energy problems in the Murmansk region and neighboring areas belonging to the European USSR. The first group of problems includes centralized heating of industrial centers on the basis of atomic energy sources, study of the feasibility of improving the management of atomic power stations with VVER-440 reactors, as well as the use of thorium as a nuclear fuel. The second group of articles deals with problems of operating power installations in the Murmansk region until 2000. Five articles discuss actual problems of hydroelectric power and one -- designing the Lumbovsk Bay Power Station.

  7. Project for qualification of the Kozloduy confinement system; Proyecto de cualificacion del sistema de confinamiento de Kozloduy

    Energy Technology Data Exchange (ETDEWEB)

    Montes Rodriguez, J L [Empresarios Agrupados, A.I.E., Madrid (Spain)

    1993-12-15

    One of the projects awarded to Empresarios Agrupados within the Six-Month WANO Programme for Kozloduy NPP, financed through the European Community's PHARE Programme, relates to the first tasks of plant confinement system qualification. Development of this project - the results of which will serve as a reference for other power plants with VVER-440/230 models - aroused considerable interest in the Bulgarian nuclear community as well as in international entities which render assistance to Eastern power plants. In fact, this is one of the few projects in the programme which takes into account hardware-oriented activities to be carried out urgently in the plant. The VVER-440/230 confinement system performs functions parallel to the containment system of Western PWR reactors. However, it differs significantly in its criteria and operation details. These important differences form the basis for a rational and prudent application of the essence of Western codes and standards relating to the qualification of these systems. The criteria which are developed to make this application viable constitute the most challenging and, at the same time, most risky part of this job. The project will be in its final stages when the 18th Annual Meeting of the Spanish Nuclear Society is held; it is therefore likely that this paper will advance the results and conclusions expected. (author)

  8. Project for qualification of the Kozloduy confinement system

    International Nuclear Information System (INIS)

    Montes Rodriguez, J.L.

    1993-01-01

    One of the projects awarded to Empresarios Agrupados within the Six-Month WANO Programme for Kozloduy NPP, financed through the European Community's PHARE Programme, relates to the first tasks of plant confinement system qualification. Development of this project - the results of which will serve as a reference for other power plants with VVER-440/230 models - aroused considerable interest in the Bulgarian nuclear community as well as in international entities which render assistance to Eastern power plants. In fact, this is one of the few projects in the programme which takes into account hardware-oriented activities to be carried out urgently in the plant. The VVER-440/230 confinement system performs functions parallel to the containment system of Western PWR reactors. However, it differs significantly in its criteria and operation details. These important differences form the basis for a rational and prudent application of the essence of Western codes and standards relating to the qualification of these systems. The criteria which are developed to make this application viable constitute the most challenging and, at the same time, most risky part of this job. The project will be in its final stages when the 18th Annual Meeting of the Spanish Nuclear Society is held; it is therefore likely that this paper will advance the results and conclusions expected. (author)

  9. Management of primary-to-secondary leaks at Loviisa nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Mohnsen, B.; Jaenkaelae, K. [IVO International Ltd, Vantaa (Finland)

    1995-12-31

    The Loviisa Nuclear power plant consisting of two VVER-440 type press water reactor units has been in commercial operation since the late 1970`s. Specific features for VVER-440 reactors are six primary loops with horizontal steam generators and main gate valves. The structure of the horizontal steam generators construction may cause a large primary to secondary leak in case of a break in the cover of the primary collector. An accident where two primary collector covers opened totally and two covers opened partly took place in Rovno, Ukraine January 1982. Primary to secondary leaks are one of the main contributors to the core melt frequency in VVER reactors according to the Loviisa 1 Probabilistic Safety Assessment. The high core damage contribution has set requirements for the development of effective means to cope with all sizes of primary to secondary leaks in the steam generator. A concept for all leak sizes has been developed for Loviisa 1 and 2. The solution includes four main areas which are a new steam generator leakage monitoring system based on nitrogen-16 measurement, an upgraded pressurizer spray system, an increased emergency cooling water reserve and an automated isolation of the defected steam generator.

  10. 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.

  11. The Plinius/Colima CA-U3 test on fission-product aerosol release over a VVER-type corium pool

    International Nuclear Information System (INIS)

    Journeau, Ch.; Piluso, P.; Correggio, P.; Godin-Jacqmin, L.

    2007-01-01

    In a hypothetical case of severe accident in a PWR type VVER-440, a complex corium pool could be formed and fission products could be released. In order to study aerosols release in terms of mechanisms, kinetics, nature or quantity, and to better precise the source term of VVER-440, a series of experiments have been performed in the Colima facility and the test Colima CA-U3 has been successfully performed thanks to technological modifications to melt a prototypical corium at 2760 C degrees. Specific instrumentation has allowed us to follow the evolution of the corium melt and the release, transport and deposition of the fission products. The main conclusions are: -) there is a large release of Cr, Te, Sr, Pr and Rh (>95%w), -) there is a significant release of Fe (50%w), -) there is a small release of Ba, Ce, La, Nb, Nd and Y (<90%w), -) there is a very small release of U in proportion (<5%w) but it is one of the major released species in mass, and -) there is no release of Zr. The Colima experimental results are consistent with previous experiments on irradiated fuels except for Ba, Fe and U releases. (A.C.)

  12. Level 1 shutdown and low power operation of Mochovce NPP, Unit 1, Slovakia

    International Nuclear Information System (INIS)

    Halada, P.; Cillik, I.; Stojka, T.; Kuzma, M.; Prochaska, J.; Vrtik, L.

    2004-01-01

    The paper presents general approach, used methods and form of documentation of the results that have been applied within the shutdown and low power PSA (SPSA) study for Mochovce NPP, Unit 1, Slovakia. The SPSA project was realized by VUJE Trnava Inc., Slovakia in 2001-2002 years. The Level 1 SPSA study for Mochovce NPP Unit 1 covers internal events as well as internal (fires, floods and heavy load drop) and external (aircraft crash, extreme meteorological conditions, seismic event and influence of surrounding industry) hazards. Mochovce NPP consists of two operating units equipped with VVER 440/V213 reactors safety upgraded before construction finishing and operation start. 87 safety measures based on VVER 440 operational experience and international mission insights were implemented to enhance its operational and nuclear safety. The SPSA relates to full power PSA (FPSA) as a continuation of the effort to create a harmonized level 1 PSA model for all operational modes of the plant with the goal to use it for further purposes as follows: Real Time Risk Monitor, Maintenance Optimization, Technical Specifications Optimization, Living PSA. (author)

  13. Management of primary-to-secondary leaks at Loviisa nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Mohnsen, B; Jaenkaelae, K [IVO International Ltd, Vantaa (Finland)

    1996-12-31

    The Loviisa Nuclear power plant consisting of two VVER-440 type press water reactor units has been in commercial operation since the late 1970`s. Specific features for VVER-440 reactors are six primary loops with horizontal steam generators and main gate valves. The structure of the horizontal steam generators construction may cause a large primary to secondary leak in case of a break in the cover of the primary collector. An accident where two primary collector covers opened totally and two covers opened partly took place in Rovno, Ukraine January 1982. Primary to secondary leaks are one of the main contributors to the core melt frequency in VVER reactors according to the Loviisa 1 Probabilistic Safety Assessment. The high core damage contribution has set requirements for the development of effective means to cope with all sizes of primary to secondary leaks in the steam generator. A concept for all leak sizes has been developed for Loviisa 1 and 2. The solution includes four main areas which are a new steam generator leakage monitoring system based on nitrogen-16 measurement, an upgraded pressurizer spray system, an increased emergency cooling water reserve and an automated isolation of the defected steam generator.

  14. Application of the combined cycle LWR-gas turbine to PWR for NPP life extension, safety upgrade and improving economy

    International Nuclear Information System (INIS)

    Kuznetsov, Yu.N.; Gabaraev, B.A.

    2002-01-01

    Full text: The unconventional technology to extend the lifetime for the NPPs now in operation and make a construction of new NPPs cheaper - erection of steam-gas toppings to the nuclear power units - is considered in the paper. Application of the steam-gas toppings permits through reducing power of ageing reactors to extend lifetime of nuclear power unit, enhance its safety and at the same time to keep full load operation of NPP turbine and other balance-of-plant equipment. Proposed technology is examined for Russian VVER-440 reactor as an example and, also, as a pilot project, for Russian boiling VK-50 reactor now in operation Application of the steam-gas topping permits: extend the service life of ageing VVER-440 reactor by 10...15 years; use the turbine and other NPP balance-of-plant equipment at full power; increase the efficiency of combined cycle up to 48% and more; enhance the safety of NPP operation; utilize NPP balance-of-plant equipment after reactor decommissioning; perform the cost-effective operation in maneuvering modes; increase capacity factor of the plant. The construction of pilot project on the basis of the VK-50 reactor will allow not only to demonstrate new technology but also to attain appreciable economic effect including that obtained due to using the available reserves of the NPP turbine. (author)

  15. Regeneration and localization of radioactive waste in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Egorov, N.N.; Kudryavtsev, E.G.; Nikipelov, B.V.; Polyakov, A.S.; Zakharkin, B.S.; Mamaev, L.A.

    1993-01-01

    Normal functioning of the nuclear-power industry is only possible with a closed fuel cycle, including regeneration of the spent fuel from atomic power plants, the production and recycling of the secondary fuel, and localization of the radioactive waste. Despite the diversity of contemporary attitudes toward the structure of the nuclear fuel cycle around the world, the closure of the fuel cycle has been fundamental to the atomic-power industry in the USSR since the very beginning, and has taken on even greater significance in Russia today. From the beginning, the idea of a closed fuel cycle has been based essentially on one fundamental criterion: the concept of expanded productivity on the basis of fuel regeneration, i.e., the economic factor. Important as economic factors are, safety issues have taken on great significance in recent years: not only power-station reactors but all the ancillary stages of the fuel cycle must meet fundamentally new reliability, safety, and environmental hazards. The RT-1 plant is a versatile operation, regenerating spent fuel from VVER-440, BN-350, and BN-600 reactors, nuclear icebreakers and submarines, research reactors, and other power units. The plant can reprocess 400 ton/year of basic VVER-440 fuel. World-class modern processes have been introduced at the plant, meeting the necessary quality standards: zonal planning, remote operation to eliminate direct contact of the staff with radioactive material, extensive monitoring and control systems, multistage gas-purification systems, and new waste-treatment methods

  16. ISP33 standard problem on the PACTEL facility

    Energy Technology Data Exchange (ETDEWEB)

    Purhonen, H.; Kouhia, J. [VTT Energy, Lappeenranta (Finland); Kalli, H. [Lappeenranta Univ. of Technology (Finland)

    1995-09-01

    ISP33 is the first OECD/NEA/CSNI standard problem related to VVER type of pressurized water reactors. The reference reactor of the PACTEL test facility, which was used to carry out the ISP33 experiment, is the VVER-440 reactor, two of which are located near the Finnish city of Loviisa. The objective of the ISP33 test was to study the natural circulation behaviour of VVER-440 reactors at different coolant inventories. Natural circulation was considered as a suitable phenomenon to focus on by the first VVER related ISP due to its importance in most accidents and transients. The behaviour of the natural circulation was expected to be different compared to Western type of PWRs as a result of the effect of horizontal steam generators and the hot leg loop seals. This ISP was conducted as a blind problem. The experiment was started at full coolant inventory. Single-phase natural circulation transported the energy from the core to the steam generators. The inventory was then reduced stepwise at about 900 s intervals draining 60 kg each time from the bottom of the downcomer. the core power was about 3.7% of the nominal value. The test was terminated after the cladding temperatures began to rise. ATHLET, CATHARE, RELAP5 (MODs 3, 2.5 and 2), RELAP4/MOD6, DINAMIKA and TECH-M4 codes were used in 21 pre- and 20 posttest calculations submitted for the ISP33.

  17. The AMES network in the 6th Framework Programme

    International Nuclear Information System (INIS)

    Sevini, F.; Debarberis, L.; Taylor, N.; Gerard, R.; English, C.; Brumovsky, M.

    2003-01-01

    The AMES (Ageing Materials European Strategy) European network started its activity in 1993 with the aim of studying ageing mechanisms and remedial procedures for structural materials used for nuclear reactor components. Operated by JRC-IE, it has been supporting the co-ordination of the project cluster throughout the 4th and 5th EURATOM Framework Programs, carrying out projects on with plant life management implications. Among them we can list the development of non-destructive techniques applied to thermal ageing and neutron embrittlement monitoring (AMES-NDT and GRETE), improved surveillance for VVER 440 reactors (COBRA), dosimetry (AMESDOSIMETRY, MADAM and REDOS), chemical composition effects on neutron embrittlement (PISA) and advanced fracture mechanics for integrity assessment (FRAME). Main frame of the network in the 5th Framework Programme is the ATHENA project, which is aimed at summarizing the obtained achievements and edit guidelines on important issues like the Master Curve, Effect of chemical composition on embrittlement rate in RPV steels, Re-embrittlement models validation after VVER-440 annealing and open issues in embrittlement of VVER type reactors. In the 6th EURATOM Framework Programme started in 2003 the network will be part of a broader initiative on PLIM including in a more integrated way NESC, ENIQ, NET and AMALIA networks. This paper shows an overview of the concluded projects, achievements of the running ones and open issues tackled in the 6th EURATOM FWP and a summary of the plans for a new broader network on NPP Plant Life management (SAFELIFE). (author)

  18. The Plinius/Colima CA-U3 test on fission-product aerosol release over a VVER-type corium pool; L'essai Plinius/Colima CA-U3 sur le relachement des aerosols de produits de fission au-dessus d'un bain de corium de type VVER

    Energy Technology Data Exchange (ETDEWEB)

    Journeau, Ch.; Piluso, P.; Correggio, P.; Godin-Jacqmin, L

    2007-07-01

    In a hypothetical case of severe accident in a PWR type VVER-440, a complex corium pool could be formed and fission products could be released. In order to study aerosols release in terms of mechanisms, kinetics, nature or quantity, and to better precise the source term of VVER-440, a series of experiments have been performed in the Colima facility and the test Colima CA-U3 has been successfully performed thanks to technological modifications to melt a prototypical corium at 2760 C degrees. Specific instrumentation has allowed us to follow the evolution of the corium melt and the release, transport and deposition of the fission products. The main conclusions are: -) there is a large release of Cr, Te, Sr, Pr and Rh (>95%w), -) there is a significant release of Fe (50%w), -) there is a small release of Ba, Ce, La, Nb, Nd and Y (<90%w), -) there is a very small release of U in proportion (<5%w) but it is one of the major released species in mass, and -) there is no release of Zr. The Colima experimental results are consistent with previous experiments on irradiated fuels except for Ba, Fe and U releases. (A.C.)

  19. Management of primary-to-secondary leaks at Loviisa nuclear power plant

    International Nuclear Information System (INIS)

    Mohnsen, B.; Jaenkaelae, K.

    1995-01-01

    The Loviisa Nuclear power plant consisting of two VVER-440 type press water reactor units has been in commercial operation since the late 1970's. Specific features for VVER-440 reactors are six primary loops with horizontal steam generators and main gate valves. The structure of the horizontal steam generators construction may cause a large primary to secondary leak in case of a break in the cover of the primary collector. An accident where two primary collector covers opened totally and two covers opened partly took place in Rovno, Ukraine January 1982. Primary to secondary leaks are one of the main contributors to the core melt frequency in VVER reactors according to the Loviisa 1 Probabilistic Safety Assessment. The high core damage contribution has set requirements for the development of effective means to cope with all sizes of primary to secondary leaks in the steam generator. A concept for all leak sizes has been developed for Loviisa 1 and 2. The solution includes four main areas which are a new steam generator leakage monitoring system based on nitrogen-16 measurement, an upgraded pressurizer spray system, an increased emergency cooling water reserve and an automated isolation of the defected steam generator

  20. Fuel reliability experience in Finland

    International Nuclear Information System (INIS)

    Kekkonen, L.

    2015-01-01

    Four nuclear reactors have operated in Finland now for 35-38 years. The two VVER-440 units at Loviisa Nuclear Power Plant are operated by Fortum and two BWR’s in Olkiluoto are operated by Teollisuuden Voima Oyj (TVO). The fuel reliability experience of the four reactors operating currently in Finland has been very good and the fuel failure rates have been very low. Systematic inspection of spent fuel assemblies, and especially all failed assemblies, is a good practice that is employed in Finland in order to improve fuel reliability and operational safety. Investigation of the root cause of fuel failures is important in developing ways to prevent similar failures in the future. The operational and fuel reliability experience at the Loviisa Nuclear Power Plant has been reported also earlier in the international seminars on WWER Fuel Performance, Modelling and Experimental Support. In this paper the information on fuel reliability experience at Loviisa NPP is updated and also a short summary of the fuel reliability experience at Olkiluoto NPP is given. Keywords: VVER-440, fuel reliability, operational experience, poolside inspections, fuel failure identification. (author)

  1. 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.)

  2. Skoda sets out its stall in the world market

    International Nuclear Information System (INIS)

    Kralovec, J.

    1992-01-01

    Through joint ventures and new products, the Czechoslovakian heavy-machinery enterprise Skoda is planning a future beyond the former Comecon and the nuclear industry. This article reports on future projects in nuclear engineering. (author)

  3. Idea národa československého na stránkách týdeníku Přítomnost (1924-1938)

    Czech Academy of Sciences Publication Activity Database

    Harna, Josef

    2009-01-01

    Roč. 17, č. 1 (2009), s. 169-193 ISSN 1210-6860 Institutional research plan: CEZ:AV0Z80150510 Keywords : 20th Century * Czechoslovak history * Czecho-Slovak relations * czechoslovakianism * journalism Subject RIV: AB - History

  4. Regulation of granulocyte colony-stimulating factor receptor-mediated granulocytic differentiation by C-mannosylation.

    Science.gov (United States)

    Otani, Kei; Niwa, Yuki; Suzuki, Takehiro; Sato, Natsumi; Sasazawa, Yukiko; Dohmae, Naoshi; Simizu, Siro

    2018-04-06

    Granulocyte colony-stimulating factor (G-CSF) receptor (G-CSFR) is a type I cytokine receptor which is involved in hematopoietic cell maturation. G-CSFR has three putative C-mannosylation sites at W253, W318, and W446; however, it is not elucidated whether G-CSFR is C-mannosylated or not. In this study, we first demonstrated that G-CSFR was C-mannosylated at only W318. We also revealed that C-mannosylation of G-CSFR affects G-CSF-dependent downstream signaling through changing ligand binding capability but not cell surface localization. Moreover, C-mannosylation of G-CSFR was functional and regulated granulocytic differentiation in myeloid 32D cells. In conclusion, we found that G-CSFR is C-mannosylated at W318 and that this C-mannosylation has role(s) for myeloid cell differentiation through regulating downstream signaling. Copyright © 2018 Elsevier Inc. All rights reserved.

  5. Politický život Výmarské republiky 1932-1933 očima československého tisku

    OpenAIRE

    Juranka, Lukáš

    2015-01-01

    The master's thesis focuses on analysis of selected Czechoslovakian party - controlled press a through it seeks to analyse attitude of Czechoslovakian political parties (belonging to to various ends of political spectrum) towards foreign and domestic policy of German government representatives and events that were taking place on the political scene of the Weimar Republic on the brink of its existence from January 1932 to Nazi seizure of power in the end of January 1933. Main emphasis of this...

  6. 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

  7. 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

  8. Evaluation of pharmaceutical and chemical equivalence of selected ...

    African Journals Online (AJOL)

    Personal

    Evaluation of Pharmaceutical and Chemical Equivalence of Selected Brands of Diclofenac Sodium .... strength- friability. /disintegration time ratio. (CSFR/DT). Drug content. % w/w ... Table 3: Parameters obtained from Kitazawa analysis. Brand.

  9. Note to the Secretariat from the Permanent Mission of the Czech and Slovak Federal Republic to the International Organizations in Vienna

    International Nuclear Information System (INIS)

    1993-01-01

    The document reproduces the Note received by the Director General from the Permanent Mission of the Czech and Slovak Federal Republic to the International Organizations in Vienna in connection with the dissolution of the CSFR on 31 December 1992

  10. 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)

  11. 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.)

  12. Construction gets underway on Hungary's Modern Vault Dry Store

    International Nuclear Information System (INIS)

    Anon.

    1995-01-01

    A construction licence has recently been granted for a Modular Vault Dry Store (MVDS) for spent fuel at the Paks reactor site in Hungary. The store will be used for medium term (50 years) storage of spent fuel from four VVER-440 reactors. It is anticipated that storage capacity for 1350 fuel assemblies will be available by 1996. Two further construction phases will take the capacity to 4950, covering the first ten years of reactor operation. The design provides for further extension to accommodate a total 15000 assemblies, corresponding to 30 years of reactor operation. The MVDS has developed out of the first application of dry store technology to spent Magnox reactor fuel at the Wylfa power station in the United Kingdom 25 years ago. (UK)

  13. Dry spent fuel storage facility at Kozloduy Nuclear Power Plant

    International Nuclear Information System (INIS)

    Goehring, R.; Stoev, M.; Davis, N.; Thomas, E.

    2004-01-01

    The Dry Spent Fuel Storage Facility (DSF) is financed by the Kozloduy International Decommissioning Support Fund (KIDSF) which is managed by European Bank for Reconstruction and Development (EBRD). On behalf of the Employer, the Kozloduy Nuclear Power Plant, a Project Management Unit (KPMU) under lead of British Nuclear Group is managing the contract with a Joint Venture Consortium under lead of RWE NUKEM mbH. The scope of the contract includes design, manufacturing and construction, testing and commissioning of the new storage facility for 2800 VVER-440 spent fuel assemblies at the KNPP site (turn-key contract). The storage technology will be cask storage of CONSTOR type, a steel-concrete-steel container. The licensing process complies with the national Bulgarian regulations and international rules. (authors)

  14. The Paks Nuclear Power Station

    International Nuclear Information System (INIS)

    Erdosi, N.; Szabo, L.

    1978-01-01

    As the first stage in the construction of the Paks Nuclear Power Station, two units of 440 MW(e) each will be built. They are operated with two coolant loops each. The reactor units are VVER 440 type water-moderated PWR type heterogeneous power reactors designed in the Soviet Union and manufactured in Czechoslovakia. Each unit operates two Soviet-made K-220-44 steam turbines and Hungarian-made generators of an effective output of 220 MW. The output of the transformer units - also of Hungarian made - is 270 MVA. The radiation protection system of the nuclear power station is described. Protection against system failures is accomplished by specially designed equipment and security measures especially within the primary circuit. Some data on the power station under construction are given. (R.P.)

  15. PACTEL: Experiments on the behaviour of the new horizontal steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Kouhia, J.; Riikonen, V.; Purhonen, H. [VTT Energy, Lappeenranta (Finland)

    1995-12-31

    Experiments were performed to study the behaviour of the PACTEL facility, a medium scale integral test loop simulating VVER 440 pressurized water reactors. The study focused on the operation of the new horizontal steam generator model installed in PACTEL. Three experiments were conducted: a small-break test to observe the steam generator behaviour over a range of primary coolant inventories, a hot leg loop seal experiment to study the cyclic behaviour of a loop seal and a loss of secondary side feedwater test to examine the effect of uncovered tubes in the steam generator. A reverse flow was observed in the lower part of the U-tube bundle of the steam generator during natural circulation. The flow reversal point dropped when the tubes uncovered, during secondary inventory reduction. (orig.). 5 refs.

  16. Effect of high temperature filtration on out-core corrosion product activity

    International Nuclear Information System (INIS)

    Horvath, G.L.; Bogancs, J.

    1983-01-01

    Investigation of the effect of high temperature filtration on corrosion product transport and out-core corrosion product activity has been carried out for VVER-440 plants. In the physico-chemical model applied particulate and dissolved corrosion products were taken into account. We supposed 100% effectivity for the particulate filter. It was found that about 0,5% 160 t/h/ of the main flow would result in an approx.50% reduction of the out-core corrosion product activity. Investigation of the details of the physico-chemical model in Nuclear Power Plant Paks showed a particle deposition rate measured during power transients fairly agreeing with other measurements and data used in the calculations. (author)

  17. East/West cooperation on the safety of USSR-designed nuclear power stations

    International Nuclear Information System (INIS)

    Spencer, P.H.

    1991-01-01

    In the aftermath of the accident at the Chernobyl nuclear power station in the Soviet Union, nuclear power plant operators throughout the world came together in May 1989 to form the World Association of Nuclear Operators (WANO). When it became clear that the operators of plants of an early design supplied by the USSR needed assistance in the upgrading of the safety of these units, WANO was uniquely placed to assist and facilitate in this. In July 1990, WANO took the decision to form a special project to assist the operators of the VVER 440/230 plants in their efforts to increase the safety standards for these units. The work performed by this special project team is described

  18. Accident sequences simulated at the Juragua nuclear power plant

    International Nuclear Information System (INIS)

    Carbajo, J.J.

    1998-01-01

    Different hypothetical accident sequences have been simulated at Unit 1 of the Juragua nuclear power plant in Cuba, a plant with two VVER-440 V213 units under construction. The computer code MELCOR was employed for these simulations. The sequences simulated are: (1) a design-basis accident (DBA) large loss of coolant accident (LOCA) with the emergency core coolant system (ECCS) on, (2) a station blackout (SBO), (3) a small LOCA (SLOCA) concurrent with SBO, (4) a large LOCA (LLOCA) concurrent with SBO, and (5) a LLOCA concurrent with SBO and with the containment breached at time zero. Timings of important events and source term releases have been calculated for the different sequences analyzed. Under certain weather conditions, the fission products released from the severe accident sequences may travel to southern Florida

  19. Computation cluster for Monte Carlo calculations

    International Nuclear Information System (INIS)

    Petriska, M.; Vitazek, K.; Farkas, G.; Stacho, M.; Michalek, S.

    2010-01-01

    Two computation clusters based on Rocks Clusters 5.1 Linux distribution with Intel Core Duo and Intel Core Quad based computers were made at the Department of the Nuclear Physics and Technology. Clusters were used for Monte Carlo calculations, specifically for MCNP calculations applied in Nuclear reactor core simulations. Optimization for computation speed was made on hardware and software basis. Hardware cluster parameters, such as size of the memory, network speed, CPU speed, number of processors per computation, number of processors in one computer were tested for shortening the calculation time. For software optimization, different Fortran compilers, MPI implementations and CPU multi-core libraries were tested. Finally computer cluster was used in finding the weighting functions of neutron ex-core detectors of VVER-440. (authors)

  20. Compact simulators for WWER-440 type nuclear power plants

    International Nuclear Information System (INIS)

    Vegh, E.; Janosy, J.S.

    1991-09-01

    This paper describes a compact simulator for VVER-440 type plants. Up till now three simulators have been delivered: to Paks Nuclear Power Plant (Hungary), and to Kola and Rovno NPPs (Soviet Union). In this compact simulator the modelling complexity of the plant is almost similar to that of a full-scope one apart from the Control Room being replaced by a Control Desk and four colour graphic display units. The simulation of the plant covers the whole operating range: from cold shut-down state to nominal power level. The simulator contains up to 32 different initial conditions. Moreover, every 5 minutes the simulator produces so called 'snapshots', i.e. disc images of the Data Base. Using these snapshots the instructor can go back in time (called backtracking) in order to repeat some previously passed events

  1. Commercial production of metal hafnium and hafnium-based products

    International Nuclear Information System (INIS)

    Negodin, D.A.; Shtutsa, M.G.; Akhtonov, S.G.; Il'enko, E.V.; Kobyzev, A.M.

    2012-01-01

    Hafnium possesses a unique complex of physical and chemical properties which allow the application of products on its basis in various industries. Joint Stock Company 'Chepetsky Mechanical Plant' is the single enterprise which produces hafnium on the territory of Russia. The manufacture of metal hafnium with the total content of zirconium and hafnium, at least, 99,8 % of weights is developed at the present time at Joint Stock Company CHMZ. The weight of melted hafnium ingots is up to 1 ton. Manufacture of wide range of products from hafnium is implemented. The plates from a hafnium with thickness of 0.60 mm which are used for emergency control cartridges of VVER-440 reactors are the most critical product. It is shown that ingots and products obtained from metal hafnium correspond to the Russian and international standards for reactor materials in chemical composition, mechanical and corrosion properties.

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  3. Backfitting of Nuclear Power Plant Bohunice V1 in Slovakia

    International Nuclear Information System (INIS)

    Ferenc, M.

    1999-01-01

    Nuclear power plants in the Slovak Republic generate almost 55 % of electricity. The operating organization and the Nuclear Regulatory Authority of the Slovak Republic pay a great attention to safe and reliable operation of four units with VVER 440 reactors at Bohunices site and one in Mochovce side. Engineering and design organizations in cooperation with well known international companies prepare evaluation of safety conditions, safety analyses and projects for the implementation of modifications to upgrade the nuclear safety of the units in operation. A gradual safety upgrading (reconstruction) of the V-1 Bohunice plant has been in progress, a modernization of the V-2 Bohunice plant is being prepared. Simultaneously the commissioning of Unit 2 at the Mochovce plant is being implemented.(author)

  4. Support calculations for management of PRISE leakage accidents

    Energy Technology Data Exchange (ETDEWEB)

    Matejovic, P.; Vranka, L. [Nuclear Power Plants Research Inst. Vuje, Trnava (Slovakia)

    1997-12-31

    Accidents involving primary-to-secondary leakage (PRISE) caused by rupture of one or a few tubes are well known design basis events in both, western and VVER NPPs. Operating experience and in-service inspections of VVER-440 units have demonstrated also the potential for large PRISE leaks in the case of the steam generator (SG) primary collector cover lift-up (Rovno NPP). Without performing any countermeasure for limitation of SG collector cover lift-up, a full opening results in PRISE leak with an equivalent diameter 107 mm. Although this accident was not considered in the original design, this event is usually analysed as DBA too. Different means are available for detection and mitigation of PRISE leakage in NPPs currently in operation (J.Bohunice V-1 and V-2) or under construction (Mochovce) in Slovakia. 8 refs.

  5. Application of the combined cycle LWR-gas turbine to PWR for NPP life extension, safety upgrade and improving economy

    International Nuclear Information System (INIS)

    Kuznetsov, Yu. N.; Lisitsa, F. D.; Smirnov, V. G.

    2000-01-01

    The unconventional technology to extend the lifetime for the NPPs now in operation and make a construction of new NPPs cheaper three-quarter erection of steam-gas toppings to the nuclear power units three-quarter is considered in the paper. Application of the steam-gas toppings permits through reducing power of aging reactors to extend lifetime of nuclear power unit, enhance its safety and at the same time to keep full load operation of NPP turbine and other balance-of-plant equipment. Proposed technology is examined for Russian VVER-440 reactor as an example and, also, as a pilot project. for Russian boiling VK-50 reactor now in operation. Heat flow sheets of the power plants, their parameters and economic problems are discussed. (author)

  6. Computation cluster for Monte Carlo calculations

    Energy Technology Data Exchange (ETDEWEB)

    Petriska, M.; Vitazek, K.; Farkas, G.; Stacho, M.; Michalek, S. [Dep. Of Nuclear Physics and Technology, Faculty of Electrical Engineering and Information, Technology, Slovak Technical University, Ilkovicova 3, 81219 Bratislava (Slovakia)

    2010-07-01

    Two computation clusters based on Rocks Clusters 5.1 Linux distribution with Intel Core Duo and Intel Core Quad based computers were made at the Department of the Nuclear Physics and Technology. Clusters were used for Monte Carlo calculations, specifically for MCNP calculations applied in Nuclear reactor core simulations. Optimization for computation speed was made on hardware and software basis. Hardware cluster parameters, such as size of the memory, network speed, CPU speed, number of processors per computation, number of processors in one computer were tested for shortening the calculation time. For software optimization, different Fortran compilers, MPI implementations and CPU multi-core libraries were tested. Finally computer cluster was used in finding the weighting functions of neutron ex-core detectors of VVER-440. (authors)

  7. Probabilistic assessments of fuel performance

    International Nuclear Information System (INIS)

    Kelppe, S.; Ranta-Puska, K.

    1998-01-01

    The probabilistic Monte Carlo Method, coupled with quasi-random sampling, is applied for the fuel performance analyses. By using known distributions of fabrication parameters and real power histories with their randomly selected combinations, and by making a large number of ENIGMA code calculations, one expects to find out the state of the whole reactor fuel. Good statistics requires thousands of runs. A sample case representing VVER-440 reactor fuel indicates relatively low fuel temperatures and mainly athermal fission gas release if any. The rod internal pressure remains typically below 2.5 MPa, which leaves a large margin to the system pressure of 12 MPa Gap conductance, an essential parameter in the accident evaluations, shows no decrease from its start-of-life value. (orig.)

  8. An analysis of reactivity prediction during the reactor start-up process

    International Nuclear Information System (INIS)

    Bajgl, Josef; Krysl, Vaclav; Svarny, Jiri

    2015-01-01

    The different VVER-440 core fuel loadings subcriticality evaluations are performed during the start-up process by boron dilution or control assembly withdrawn by macrocode MOBY-DICK calculations. The dynamic reactivity and quasicritical reactivity are compared and sensitivity of reactivity prediction at the low boundary of start-up interval (ρ = -0,01) has been provided on the basis of different modelling of ionization chamber (IC) response calculation. Special attention is paid to the impact of power distribution and spontaneous fission distribution form factor on IC response correction during control assembly movement. Precision and robustness of different corrections of IC signal processing in real core start-up processed IC signals was evaluated.

  9. PACTEL: Experiments on the behaviour of the new horizontal steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Kouhia, J; Riikonen, V; Purhonen, H [VTT Energy, Lappeenranta (Finland)

    1996-12-31

    Experiments were performed to study the behaviour of the PACTEL facility, a medium scale integral test loop simulating VVER 440 pressurized water reactors. The study focused on the operation of the new horizontal steam generator model installed in PACTEL. Three experiments were conducted: a small-break test to observe the steam generator behaviour over a range of primary coolant inventories, a hot leg loop seal experiment to study the cyclic behaviour of a loop seal and a loss of secondary side feedwater test to examine the effect of uncovered tubes in the steam generator. A reverse flow was observed in the lower part of the U-tube bundle of the steam generator during natural circulation. The flow reversal point dropped when the tubes uncovered, during secondary inventory reduction. (orig.). 5 refs.

  10. Benchmark calculations by KENO-Va using the JEF 2.2 library

    Energy Technology Data Exchange (ETDEWEB)

    Markova, L.

    1994-12-01

    This work has to be a contribution to the validation of the JEF2.2 neutron cross-section libarary, following the earlier published benchmark calculations having been performed to validate the previous version JEF1.1 of the libarary. Several simple calculational problems and one experimental problem were chosen for a criticality calculations. In addition also a realistic hexagonal arrangement of the VVER-440 fuel assemblies in a spent fuel cask were analyzed in a partly cylindrized model. All criticality calculations, carried out by the KENO-Va code using the JEF2.2 neutron cross-section library in 172 energy groups, resulted in multiplication factors (k{sub eff}) which were tabulated and compared with the results of other available calculations of the same problems. (orig.).

  11. Innovated feed water distributing system of VVER steam generators

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  12. Quantitative code accuracy evaluation of ISP33

    Energy Technology Data Exchange (ETDEWEB)

    Kalli, H.; Miwrrin, A. [Lappeenranta Univ. of Technology (Finland); Purhonen, H. [VTT Energy, Lappeenranta (Finland)] [and others

    1995-09-01

    Aiming at quantifying code accuracy, a methodology based on the Fast Fourier Transform has been developed at the University of Pisa, Italy. The paper deals with a short presentation of the methodology and its application to pre-test and post-test calculations submitted to the International Standard Problem ISP33. This was a double-blind natural circulation exercise with a stepwise reduced primary coolant inventory, performed in PACTEL facility in Finland. PACTEL is a 1/305 volumetrically scaled, full-height simulator of the Russian type VVER-440 pressurized water reactor, with horizontal steam generators and loop seals in both cold and hot legs. Fifteen foreign organizations participated in ISP33, with 21 blind calculations and 20 post-test calculations, altogether 10 different thermal hydraulic codes and code versions were used. The results of the application of the methodology to nine selected measured quantities are summarized.

  13. Structural response of rectilinear containment to overpressurization

    International Nuclear Information System (INIS)

    Pfeiffer, P.A.; Kulak, R.F.

    1995-01-01

    Containment structures for nuclear reactors are the final barrier between released radionuclides and the public. Containment structures are constructed from steel, reinforced concrete, or prestressed concrete. US nuclear reactor containment geometries tend to be cylindrical with elliptical or hemispherical heads. The older Soviet designed reactors do not use a containment building to mitigate the effects of accidents. Instead, they employ a sealed set of rectilinear, interconnected compartments, collectively called the accident localization system (ALS), to reduce the release of radionuclides to the atmosphere during accidents. The purpose of this paper is to present a methodology that can be used to find the structural capacity of reinforced concrete structures. The method is applicable to both cylindrical and rectilinear geometries. As an illustrative example, the methodology is applied to a generic VVER-440/V213 design

  14. Nuclear power and Imatran Voima in the future

    International Nuclear Information System (INIS)

    Numminen, K.

    1995-01-01

    As the owner of the Loviisa NPS with two VVER-440 units, Imatran Voima (IVO) has worked with nuclear power for more than twenty years. After the negative decision of the Finnish Parliament in 1993 there are no possibilities to build nuclear power in Finland in the near future. However, the preparation work for increasing the produced power of all four operating NPP's of Finland is going on. The emphasis in the work with new nuclear energy is on the supporting programs in Eastern Europe and the preparation of a building contract of a new NPS to China together with the Russians. With a new decision of the Finnish Parliament, the nuclear option could still be an important part of the future energy strategy of Finland. (orig.)

  15. Computer modelling the potential benefits of amines in NPP Bohunice secondary circuit

    International Nuclear Information System (INIS)

    Fountain, M.J.; Smiesko, I.

    1998-01-01

    The use of computer modelling of PWR and WWER secondary circuit chemistry was already demonstrated in the past. The model was used to illustrate the technical and economic advantages, compared with ammonia, of using an 'advanced', high basicity, low volatility amines to raise the liquid phase pH(T) in the moisture separator and other areas swept by wet steam. Since the 1995, this technique has been successfully applied to a number of power plants and the computer model has been progressively developed. This paper describes the preliminary results of an ongoing assessment being carried out for the VVER 440 plants at Bohunice. The work for Bohunice is being funded by the 'Know How Fund', a department in the British Government's Foreign and Commonwealth Office. (J.P.N.)

  16. Project Management Unit for decommissioning of NPP Bohunice VI (2003-2014)

    International Nuclear Information System (INIS)

    Gonzalez Fernandez-conde, A.; Brochet, I.; Ferreira, A.

    2015-01-01

    From October 2003 until december 2014 the Consortium consisting of Iberdrola Engineering and Construction (leader). Empresarios Agrupados Internacional, and Indra Sistemas has carried out the project Project Management Unit ((PMU) for the decommissioning of Bohunice V1 NPP (units 1 and 2), type VVER-440/V-230 in Slovakia. during the first phase (2003-2007) EdF was also part of the Consortium. The project is funded by the Bohunice International Decommissioning Support Fund (BIDSF) administered by the RBRD. The main objective of the project is to provide the necessary engineering and resources of project management for planning, execution, management, coordination and monitoring of all tasks in support of the decommissioning. (Author)

  17. Investigation to determine the absolute sensitivity of Rh SPNDs

    International Nuclear Information System (INIS)

    Adorian, F.; Patai Szabo, S.; Pos, I.

    1998-01-01

    The goal of the work was to find an empirical sensitivity function of the Rh SPNDs used in VVER-440 reactors and to investigate the accuracy and adequateness of the detector signal predicting capability of the associated model. In our case the model was based on the HELIOS transport code and the C-PORCA nodal code. A statistical sensitivity analysis versus some selected parameters (e.g. enrichment, burn-up) has been carried out by using a substantial amount of measured data. We also investigated the stability of the electron collecting probability of the detectors versus their burn-up and other parameters with the aim of obtaining a tuned semi-empirical formula for the detector burnup correction. (Authors)

  18. Improvement of MSLB transient analysis for VVER by the coupled code system KIKO3D/ATHLET

    International Nuclear Information System (INIS)

    Hegyi, Gy.; Kereszturi, A.; Trosztel, I.

    2001-01-01

    An overview is given on the investigations of the Main Steam Line Break transient in a VVER- 440 NPP by using the KIKO3D/ATHLET 1.2.A coupled code system. Special attention was paid for the influence of modeling the outcore detector signals and the malfunctioning of the emergency control system (scram with stuck rod). The conservatism of the calculations was assured even in the case of application of the 3D best estimate KIKO3D code. The consequence of MSLB accident is investigated at the end of cycle (EOC), at full power (FP) and shut down initial conditions. Even if very strong conservative assumptions were applied, dangerous hot spots were not found in the supposed scenarios.(author)

  19. Most significant preliminary results of the probabilistic safety analysis on the Juragua nuclear power plant

    International Nuclear Information System (INIS)

    Perdomo, Manuel

    1995-01-01

    Since 1990 the Group for PSA Development and Applications (GDA/APS) is working on the Level-1 PSA for the Juragua-1 NPP, as a part of an IAEA Technical Assistance Project. The main objective of this study, which is still under way, is to assess, in a preliminary way, the Reactor design safety to find its potential 'weak points' at the construction stage, using a eneric data base. At the same time, the study allows the PSA team to familiarize with the plant design and analysis techniques for the future operational PSA of the plant. This paper presents the most significant preliminary results of the study, which reveal some advantages of the safety characteristics of the plant design in comparison with the homologous VVER-440 reactors and some areas, where including slight modifications would improve the plant safety, considering the level of detail at which the study is carried out. (author). 13 refs, 1 fig, 2 tabs

  20. 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.)

  1. REKO - Bohunice V-1. Experience with instrumentation and control system

    International Nuclear Information System (INIS)

    Arbet, L.; Ziska, D.; Golan, P.; Karaba, P.; Krupa, S.; Wiening, K.-H.

    2000-01-01

    In this paper and in presentation some results of upgrading of the NPP Bohunice V-1 are presented. For the first time, extensive upgrades are performed in all safety-related areas of both units with VVER 440/230 reactors. These upgrades focused on: - Expansion and upgrading of the process safety systems; - Replacement of the safety I and C system with a TELEPERM XS-based system; - Spatial separation of safety equipment; - Modernisation of the electrical auxiliary power systems; - Seismic upgrading and fire protection; - Improvement of the man-machine interface. This upgrade is considered exemplary around the world. The most extensive stage of gradual reconstruction of Unit 2 was completed according to the schedule in January 1999. For the first time, a reactor which incorporates state-of-the-art digital I and C in its reactor protection system is on-line. (author)

  2. Preparation of Long Term Operation in Dukovany NPP, Czech Republic

    International Nuclear Information System (INIS)

    Krivanek, R.; Sabata, M.

    2012-01-01

    Dukovany NPP in the south-east of the Czech Republic operates four VVER 440/213 type units. The first unit was commissioned in 1985 and the last one in 1987. The operational results of the whole NPP have been excellent and NPP permanently belongs between the first quartile of the best operated NPPs in the world in accordance with WANO factors. Large safety improvement programme have been implemented in last 15 years. The original design lifetime of main components is 30 years which means till 2015 and it is understandable that NPP is preparing for long-term operation (LTO). The paper is describing activities carried out and planned for safe and successful LTO. (author)

  3. Advanced fuel cycles options for LWRs and IMF benchmark definition

    International Nuclear Information System (INIS)

    Breza, J.; Darilek, P.; Necas, V.

    2008-01-01

    In the paper, different advanced nuclear fuel cycles including thorium-based fuel and inert-matrix fuel are examined under light water reactor conditions, especially VVER-440, and compared. Two investigated thorium based fuels include one solely plutonium-thorium based fuel and the second one plutonium-thorium based fuel with initial uranium content. Both of them are used to carry and burn or transmute plutonium created in the classical UOX cycle. The inert-matrix fuel consist of plutonium and minor actinides separated from spent UOX fuel fixed in Yttria-stabilised zirconia matrix. The article shows analysed fuel cycles and their short description. The conclusion is concentrated on the rate of Pu transmutation and Pu with minor actinides cumulating in the spent advanced thorium fuel and its comparison to UOX open fuel cycle. Definition of IMF benchmark based on presented scenario is given. (authors)

  4. ATHLET calculations of the pressurizer surge line break (PH-SLB test) at the PMK-2 test facility

    International Nuclear Information System (INIS)

    Krepper, E.; Schaefer, F.

    2000-01-01

    At the Hungarian integral test facility PMK-2 a pressurizer surge line break experiment (PH-SLB test) was carried out with the PHARE 4.2.6b project. The primary objective of the test was to provide experimental data for a surge line break transient at VVER-440 reactors with reduced injection from the emergency core cooling systems (ECC). At the Institute of Safety Research calculations of the experiment were performed with the thermohydraulic computer code ATHLET, which was developed by GRS (Gesellschaft fuer Anlagen- und Reaktorsicherheit) mbH. In the context of the PHARE 4.2.6b project the Institute of Safety Research has also supplied the void fraction measurement system for the PMK-2 test facility and was involved in the evaluation of the experimental results. (orig.)

  5. Support calculations for management of PRISE leakage accidents

    Energy Technology Data Exchange (ETDEWEB)

    Matejovic, P; Vranka, L [Nuclear Power Plants Research Inst. Vuje, Trnava (Slovakia)

    1998-12-31

    Accidents involving primary-to-secondary leakage (PRISE) caused by rupture of one or a few tubes are well known design basis events in both, western and VVER NPPs. Operating experience and in-service inspections of VVER-440 units have demonstrated also the potential for large PRISE leaks in the case of the steam generator (SG) primary collector cover lift-up (Rovno NPP). Without performing any countermeasure for limitation of SG collector cover lift-up, a full opening results in PRISE leak with an equivalent diameter 107 mm. Although this accident was not considered in the original design, this event is usually analysed as DBA too. Different means are available for detection and mitigation of PRISE leakage in NPPs currently in operation (J.Bohunice V-1 and V-2) or under construction (Mochovce) in Slovakia. 8 refs.

  6. Knowledge Management Aspects of Decommissioning. Case Study

    International Nuclear Information System (INIS)

    Pironkov, Lyubomir

    2017-01-01

    Kozloduy NPP: Units 5&6, type VVER-1000 - in operation. SE RAW SD “Decommissioning of units 1-4” (type VVER-440); SD “Radioactive Waste – Kozloduy”; SD “National Repository for Radioactive Waste“; SD “Permanent Repository for Radioactive Waste – Novi Han”. Decommissioning Strategy: Strategy Target: “Brown Field”. Initial Version: Safe enclosure – completing by 2050. Updated Version: Continuous dismantling of equipment; Completing the process of Decommissioning of Units 1-4 by 2030. Major Phases: 1.Pre-decommissioning activities; 2.Facility shutdown activities; 3.Procurement of equipment; 4.Dismantling activities; 5.Treatment of RAM and RAW and delivery for disposal; 6.Site management and support; 7.Project management and engineering; 8.Management of SNF and activated materials

  7. Most significant preliminary results of the probabilistic safety analysis on the Juragua nuclear power plant; Resultados preliminares mas significativos del analysis probabilista de seguridad de la Central Nuclear de Juragua

    Energy Technology Data Exchange (ETDEWEB)

    Perdomo, Manuel [Instituto Superior de Ciencia y Tecnologia Nuclear (ISCTN), La Habana (Cuba)

    1995-12-31

    Since 1990 the Group for PSA Development and Applications (GDA/APS) is working on the Level-1 PSA for the Juragua-1 NPP, as a part of an IAEA Technical Assistance Project. The main objective of this study, which is still under way, is to assess, in a preliminary way, the Reactor design safety to find its potential `weak points` at the construction stage, using a eneric data base. At the same time, the study allows the PSA team to familiarize with the plant design and analysis techniques for the future operational PSA of the plant. This paper presents the most significant preliminary results of the study, which reveal some advantages of the safety characteristics of the plant design in comparison with the homologous VVER-440 reactors and some areas, where including slight modifications would improve the plant safety, considering the level of detail at which the study is carried out. (author). 13 refs, 1 fig, 2 tabs.

  8. TEM study of radiation induced defects in baffle-former-barrel assembly from decommissioned NPP Greifswald

    International Nuclear Information System (INIS)

    Srba, O.; Michalicka, J.; Keilova, E.; Kocik, K.

    2013-06-01

    A complex transmission electron microscopy (TEM) study of reactor vessel internal (RVI) materials from the baffle-former-barrel assembly from NPP Greifswald (VVER 440), Unit 1 decommissioned after 15 service cycles has been undertaken. All parts of the baffle-former-barrel assembly are made from Ti-stabilized austenitic stainless steel 08Ch18N10T. The materials were exposed to different dose of neutron radiation (2.4 - 11.4 dpa) at temperatures 267 - 398 deg. C depending on position in the core. Three types of radiation induced defects were identified and quantified, namely: dislocations, cavities (voids) and fine-scaled precipitated particles of Ni-Si rich phases. Black-dot type defects were observed too. Operation conditions are around ≅ 300 deg. C that is why we have observed defect typical for both low and high regions of irradiation temperatures. (authors)

  9. Investigation of bubble-condenser operation under large break LOCA conditions

    International Nuclear Information System (INIS)

    Blinkov, V.; Melikhov, O.; Melikhov, V.; Davydov, M.; Sokolin, A.; Hoffmann, D.; Simon, U.; Bajsz, J.

    2000-01-01

    In the framework of the PHARE/TACIS project, the experimental test facility for bubble condenser experimental qualification was built at Electrogorsk Research and Engineering Centre. The test facility contains high pressure system, compartments upstream of the bubble condenser and a section of the bubble condenser system. The scaling of the test facility is 1:100. The high pressure system consists of five vessels to appropriately model the leak functions (mass flow rate and enthalpy) during the loss of coolant accidents postulated in the design of VVER-440/V213. Design basis accident (LB LOCA) was experimentally and analytically considered. Results of pre-test analysis with ATHLET and DRASYS codes for determination of necessary test parameters and post-test analysis of three tests are presented. (author)

  10. Steam condensation induced water hammer simulations for different pipelines

    International Nuclear Information System (INIS)

    Barna, I.F.; Ezsol, G.

    2011-01-01

    We investigate steam condensation induced water hammer (CIWH) phenomena and present theoretical results for different kind of pipelines. We analyze the process with the WAHA3 model based on two-phase flow six first-order partial differential equations that present one dimensional, surface averaged mass, momentum and energy balances. A second order accurate high-resolution shock-capturing numerical scheme was applied with different kind of limiters in the numerical calculations. At first, we present calculations for various pipelines in the VVER-440-312 type nuclear reactor. Our recent calculation clearly shows that the six conditions of Griffith are only necessary conditions for CIWH but not sufficient. As second results we performed calculations for various geometries and compare with the theory of Chun. (author)

  11. Effect of spacer grid mixing vanes on coolant outlet temperature distribution

    Energy Technology Data Exchange (ETDEWEB)

    Raemae, Tommi; Lahtinen, Tuukka; Brandt, Tellervo; Toppila, Timo [Fortum Power and Heat, Fortum (Finland). Nuclear Competence Center

    2012-08-15

    In Loviisa VVER-440-type NPP the coolant outlet temperature of the hot subchannel is constantly monitored during the operation. According to the authority requirement the maximum subchannel outlet temperature must not exceed the saturation temperature. Coolant temperature distribution inside the fuel assembly is affected by the efficiency of the coolant mixing. In order to enhance the coolant mixing the fuel manufacturer is introducing the additional mixing vanes on the fuel bundle spacer grids. In the paper the effect of the different mixing vane modifications is studied with computational fluid dynamics (CFD) simulation. Goal of the modelling is to find vane modifications with which sufficient mixing is reached with acceptable increase in the spacer grid pressure loss. The results of the studies are discussed in the paper. (orig.)

  12. CATHARE Assessment of PACTEL LOCA Experiments with Accident Management

    Directory of Open Access Journals (Sweden)

    Luben Sabotinov

    2010-01-01

    Full Text Available This paper summarizes the analysis results of three PACTEL experiments, carried out with the advanced thermal-hydraulic system computer CATHARE 2 code as a part of the second work package WP2 (analytical work of the EC project “Improved Accident Management of VVER nuclear power plants” (IMPAM-VVER. The three LOCA experiments, conducted on the Finnish test facility PACTEL (VVER-440 model, represent 7.4% cold leg breaks with combination of secondary bleed and primary bleed and feed and different actuation modes of the passive safety injection. The code was used for both defining and analyzing the experiments, and to assess its capabilities in predicting the associated complex VVER-related phenomena. The code results are in reasonable agreement with the measurements, and the important physical phenomena are well predicted, although still further improvement and validation might be necessary.

  13. Stress and fatigue analyses of primary circuit components of NPP using FEM

    International Nuclear Information System (INIS)

    Gal, P.

    2015-01-01

    This poster is a short illustration of the numerical assessment of the VVER-440 reactor pressure vessel (RPV) main flange. RPV main flange consists in free flange, pressure ring, flange bolts, nut and nickel gasket. Operating temperature transient modes, like heat up regime can lead to serious tension in bolts. So temperature fields have to be calculated. The fatigue assessment of the main flange bolt requires the determination of the coefficient of stress concentrators in bolt thread. Stress concentrators can be computed through FEM or given by norms (PNAEG). The most significant value of fatigue usage factor is in the first thread connection between bolt and nut. A finite element method (FEM) is used for calculation stress and temperature distribution in the reactor flange. The reassessment was performed according Czech normative document NTD-A.S.I. and VERLIFE

  14. A Prototype for Passive Gamma Emission Tomography

    International Nuclear Information System (INIS)

    Honkamaa, T.; Levai, F.; Berndt, R.; Schwalbach, P.; Vaccaro, S.; ); Turunen, A.

    2015-01-01

    Combined efforts of multiple stakeholders of the IAEA Support Programme task JNT 1510: ''Prototype of passive gamma emission tomograph (PGET)'', resulted in the design, manufacturing and extensive testing of an advanced verification tool for partial defect testing on light water reactor spent fuel. The PGET has now reached a proven capability of detecting a single missing or substituted pin inside a BWR and VVER-440 fuel assemblies. The task started in 2004 and it is planned to be finished this year. The PGET head consists of two banks of 104 CdTe detectors each with integrated data acquisition electronics. The CdTe detectors are embedded in tungsten collimators which can be rotated around the fuel element using an integrated stepping motor mounted on a rotating table. All components are packed inside a toroid watertight enclosure. Control, data acquisition and image reconstruction analysis is fully computerized and automated. The design of the system is transportable and suitable for safeguards verifications in spent fuel ponds anywhere. Four test campaigns have been conducted. In 2009, the first test in Ringhals NPP failed collecting data but demonstrated suitability of the PGET for field deployments. Subsequent tests on fuel with increasing complexity were all successful (Ispra, Italy (2012), Olkiluoto, Finland (2013) and Loviisa, Finland (2014)). The paper will present the PGET design, results obtained from the test campaigns and mention also drawbacks that were experienced in the project. The paper also describes further tests which would allow evaluating the capabilities and limitations of the method and the algorithm used. Currently, the main technical shortcoming is long acquisition time, due to serial control and readout of detectors. With redesigned electronics it can be expected that the system would be able to verify a VVER-440 assembly in five minutes, which meets the IAEA user requirements. (author)

  15. Heat transfer characteristics of horizontal steam generators under natural circulation conditions

    International Nuclear Information System (INIS)

    Hyvaerinen, J.

    1996-01-01

    This paper deals with the heat transfer characteristics of horizontal steam generators, particularly under natural circulation (decay heat removal) conditions on the primary side. Special emphasis is on the inherent features of horizontal steam generator behaviour. A mathematical model of the horizontal steam generator primary side is developed and qualitative results are obtained analytically. A computer code, called HSG, is developed to solve the model numerically, and its predictions are compared with experimental data. The code is employed to obtain for VVER 440 steam generators quantitative results concerning the dependence of primary-to-secondary heat transfer efficiency on the primary side flow rate, temperature and secondary level. It turns out that the depletion of the secondary inventory leads to an inherent limitation of the decay energy removal in VVER steam generators. The limitation arises as a consequence of the steam generator tube bundle geometry. As an example, it is shown that the grace period associated with pressurizer safety valve opening during a station black-out is 2 1/2-3 hours instead of the 5-6 hours reported in several earlier studies. (However, the change in core heat-up timing is much less-about 1 h at most.) The heat transfer limitation explains the fact that, in the Greifswald VVER 440 station black-out accident in 1975, the steam generators never boiled dry. In addition, the stability of single-phase natural circulation is discussed and insights on the modelling of horizontal steam generators with general-purpose thermal-hydraulic system codes are also presented. (orig.)

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-01

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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

  20. RELAP5 simulation of surge line break accident using combined and best estimate plus uncertainty approaches

    International Nuclear Information System (INIS)

    Kristof, Marian; Kliment, Tomas; Petruzzi, Alessandro; Lipka, Jozef

    2009-01-01

    Licensing calculations in a majority of countries worldwide still rely on the application of combined approach using best estimate computer code without evaluation of the code models uncertainty and conservative assumptions on initial and boundary, availability of systems and components and additional conservative assumptions. However best estimate plus uncertainty (BEPU) approach representing the state-of-the-art in the area of safety analysis has a clear potential to replace currently used combined approach. There are several applications of BEPU approach in the area of licensing calculations, but some questions are discussed, namely from the regulatory point of view. In order to find a proper solution to these questions and to support the BEPU approach to become a standard approach for licensing calculations, a broad comparison of both approaches for various transients is necessary. Results of one of such comparisons on the example of the VVER-440/213 NPP pressurizer surge line break event are described in this paper. A Kv-scaled simulation based on PH4-SLB experiment from PMK-2 integral test facility applying its volume and power scaling factor is performed for qualitative assessment of the RELAP5 computer code calculation using the VVER-440/213 plant model. Existing hardware differences are identified and explained. The CIAU method is adopted for performing the uncertainty evaluation. Results using combined and BEPU approaches are in agreement with the experimental values in PMK-2 facility. Only minimal difference between combined and BEPU approached has been observed in the evaluation of the safety margins for the peak cladding temperature. Benefits of the CIAU uncertainty method are highlighted.

  1. Granulocyte-Colony Stimulating Factor Receptor, Tissue Factor, and VEGF-R Bound VEGF in Human Breast Cancer In Loco.

    Science.gov (United States)

    Wojtukiewicz, Marek Z; Sierko, Ewa; Skalij, Piotr; Kamińska, Magda; Zimnoch, Lech; Brekken, Ralf A; Thorpe, Philip E

    2016-01-01

    Doxorubicin and docetaxel-based chemotherapy regimens used in breast cancer patients are associated with high risk of febrile neutropenia (FN). Granulocyte colony-stimulating factors (G-CSF) are recommended for both treating and preventing chemotherapy-induced neutropenia. Increased thrombosis incidence in G-CSF treated patients was reported; however, the underlying mechanisms remain unclear. The principal activator of blood coagulation in cancer is tissue factor (TF). It additionally contributes to cancer progression and stimulates angiogenesis. The main proangiogenic factor is vascular endothelial growth factor (VEGF). The aim of the study was to evaluate granulocyte-colony stimulating factor receptor (G-CSFR), tissue factor (TF) expression and vascular endothelial growth factor receptor (VEGF-R) bound VEGF in human breast cancer in loco. G-CSFR, TF and VEGFR bound VEGF (VEGF: VEGFR) were assessed in 28 breast cancer tissue samples. Immunohistochemical (IHC) methodologies according to ABC technique and double staining IHC procedure were employed utilizing antibodies against G-CSFR, TF and VEGF associated with VEGFR (VEGF: VEGFR). Expression of G-CSFR was demonstrated in 20 breast cancer tissue specimens (71%). In 6 cases (21%) the expression was strong (IRS 9-12). Strong expression of TF was observed in all investigated cases (100%). Moreover, expression of VEGF: VEGFR was visualized in cancer cells (IRS 5-8). No presence of G-CSFR, TF or VEGF: VEGFR was detected on healthy breast cells. Double staining IHC studies revealed co-localization of G-CSFR and TF, G-CSFR and VEGF: VEGFR, as well as TF and VEGF: VEGFR on breast cancer cells and ECs. The results of the study indicate that GCSFR, TF and VEGF: VEGFR expression as well as their co-expression might influence breast cancer biology, and may increase thromboembolic adverse events incidence.

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

    International Nuclear Information System (INIS)

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

    1999-01-01

    and metal masses of VVER reactors results for some accident sequences in later core degradation. The unique construction of VVER-440 control assemblies plays a special role during accident progression, having several fuel assemblies below the core and creating a heterogeneous core structure with absorber assemblies. Some events (e.g. B 4 C melting in the VVER-1000) are more similar to processes in BWRs. The early phase of core degradation seems to be similar in VVERs and PWRs, however the role of boron steel absorber assembly melting can change the sequence for the VVER-440. Accident progression during the late phase of core degradation can be influenced by the interaction of molten material with lower plenum structures in VVER-440. The mechanism of vessel failure can show some differences due to the VVER bottom heads being elliptical and without penetrations, compared with those of PWRs which are hemispherical and where penetrations are present. The severe accident phenomena were compared with the help of categories used for PWRs and BWRs. Most of the phenomena were found not to be VVER-specific, which means that the phenomena takes place in a similar way in the reactor types compared. Very few phenomena were found not relevant to VVER (e.g. AIC control rod related ones). When the phenomena was given a VVER-specific character the role of design features was discussed. In some cases there was experimental evidence, however in most of the cases the effect of VVER design could only be estimated. Some phenomena are not known in detail even for Western LWRs, so the comparison can show only some likelihood of differences. The review of related experiments showed, that the VVER experimental database is not as extensive as that for PWRs and BWRs. A number of separate-effect and integral tests indicated that the behaviour of materials used in VVERs is generally similar to that of PWRs. Some specific areas are not covered by experiments at all and their investigation should be

  3. Mechanical analysis of cylindrical part of canisters for spent nuclear fuel

    International Nuclear Information System (INIS)

    Ikonen, K.

    2005-06-01

    This report describes mechanical analyses of cylindrical part of the VVER 440-, BWR and EPR-type canisters for spent nuclear fuel. The task was first to evaluate the stresses at maximum design pressure and further by increasing pressure load to determine the limit collapse load and corresponding safety factor. Maximum design pressure 44 MPa is a sum of the hydrostatic pressure 30 MPa caused by 3 km ice layer, 7 MPa caused by ground water pressure at the deepest disposal depth of 700 m and 7 MPa from bentonite swelling pressure. The analysis presented in this report concern the middle area of the canisters, where the cast iron insert is considered to be more critical than in the ends of the canister. For the model a piece from the middle area of the canister was separated by two planes perpendicular to the axis of the canister. This piece was studied first by two-dimensional plane strain model, where the planes are constrained and no elongation of the canister takes place. In the second model one of the planes was constrained and the other plane was allowed to displace in axial direction, which remains as a plane during deformation and to which axial pressure force is directed. This analysis, which corresponds better the real condition in the canister, was performed as threedimensional. The analyses gave however practically equal results due to plastic deformation. Thus the analysis can be done by two-dimensional plane strain model leading to same accuracy with less computation effort. Analyses were performed as large displacement and large strain analyses by the PASULA computing package, which has been developed at VTT for a variety of structural analysis and for heat conduction calculations. A special routine was developed for automatic mesh generation. Before the analysis of the VVER 440-, BWR- and EPR-type canisters the calculation methodology was validated with test results, which were received from pressure tests performed with a short BWR canister in Germany

  4. Research into the behaviour and transport of radionuclides in waters. Methods of testing sorption properties of materials present in aqueous environment and evaluation of results

    International Nuclear Information System (INIS)

    Mansfeld, A.; Kortus, J.; Mayer, J.; Hanslik, E.

    1979-01-01

    The study deals with the occurrence of Ra-226, uranium, Cr-51, Co-60, Sr-89, J-131 and Cs-137 in the Czechoslovakian reach of the Danube and the accumulation of nuclides in bottom sediments in function of the various physico-mechanical properties and their adsorptive characteristics. (author)

  5. Study tour of the Czech and Slovak Federal Republic (formerly Czechoslovakia)

    International Nuclear Information System (INIS)

    1993-05-01

    The radioactive Waste Management Advisory Committee (RWMAC) is the independent body that advises the Secretaries of State for the Environment, Scotland and Wales on civil radioactive waste management issues. In September 1992, a RWMAC Study Group visited the Czech and Slovak Federal Republic (CSFR - formerly Czechoslovakia) to learn about the radioactive waste management practices there. This publication reports on the Group's findings. The rapid political change, social conflicts over energy options, growing environmental concern, and lack of financial resources, being experienced by the CSFR, would point to the need for a body similar to RWMAC to advise on an overall policy. (Author)

  6. Bohunice Simulator Data Collection Project

    International Nuclear Information System (INIS)

    Cillik, Ivan; Prochaska, Jan

    2002-01-01

    The paper describes the way and results of human reliability data analysis collected as a part of the Bohunice Simulator Data Collection Project (BSDCP), which was performed by VUJE Trnava, Inc. with funding support from the U.S. DOE, National Nuclear Security Administration. The goal of the project was to create a methodology for simulator data collection and analysis to support activities in probabilistic safety assessment (PSA) and human reliability assessment for Jaslovske Bohunice nuclear power plant consisting of two sets of twin units: two VVER 440/V-230 (V1) and two VVER 440/V-213 (V2) reactors. During the project training of V-2 control room crews was performed at VUJE-Trnava simulator. The simulator training and the data collection were done in parallel. The main goal of BSDCP was to collect suitable data of human errors under simulated conditions requiring the use of symptom-based emergency operating procedures (SBEOPs). The subjects of the data collection were scenario progress time data, operator errors, and real-time technological parameters. The paper contains three main parts. The first part presents preparatory work and semi-automatic computer-based methods used to collect data and to check technological parameters in order to find hidden errors of operators, to be able to retrace the course of each scenario for purposes of further analysis, and to document the whole training process. The first part gives also an overview of collected data scope, human error taxonomy, and state classifications for SBEOP instructions coding. The second part describes analytical work undertaken to describe time distribution necessary for execution of various kinds of instructions performed by operators according to the classification for coding of SBEOP instructions. It also presents the methods used for determination of probability distribution for different operator errors. Results from the data evaluation are presented in the last part of the paper. An overview of

  7. Non-linear triangle-based polynomial expansion nodal method for hexagonal core analysis

    International Nuclear Information System (INIS)

    Cho, Jin Young; Cho, Byung Oh; Joo, Han Gyu; Zee, Sung Qunn; Park, Sang Yong

    2000-09-01

    This report is for the implementation of triangle-based polynomial expansion nodal (TPEN) method to MASTER code in conjunction with the coarse mesh finite difference(CMFD) framework for hexagonal core design and analysis. The TPEN method is a variation of the higher order polynomial expansion nodal (HOPEN) method that solves the multi-group neutron diffusion equation in the hexagonal-z geometry. In contrast with the HOPEN method, only two-dimensional intranodal expansion is considered in the TPEN method for a triangular domain. The axial dependence of the intranodal flux is incorporated separately here and it is determined by the nodal expansion method (NEM) for a hexagonal node. For the consistency of node geometry of the MASTER code which is based on hexagon, TPEN solver is coded to solve one hexagonal node which is composed of 6 triangular nodes directly with Gauss elimination scheme. To solve the CMFD linear system efficiently, stabilized bi-conjugate gradient(BiCG) algorithm and Wielandt eigenvalue shift method are adopted. And for the construction of the efficient preconditioner of BiCG algorithm, the incomplete LU(ILU) factorization scheme which has been widely used in two-dimensional problems is used. To apply the ILU factorization scheme to three-dimensional problem, a symmetric Gauss-Seidel Factorization scheme is used. In order to examine the accuracy of the TPEN solution, several eigenvalue benchmark problems and two transient problems, i.e., a realistic VVER1000 and VVER440 rod ejection benchmark problems, were solved and compared with respective references. The results of eigenvalue benchmark problems indicate that non-linear TPEN method is very accurate showing less than 15 pcm of eigenvalue errors and 1% of maximum power errors, and fast enough to solve the three-dimensional VVER-440 problem within 5 seconds on 733MHz PENTIUM-III. In the case of the transient problems, the non-linear TPEN method also shows good results within a few minute of

  8. Bohunice Simulator Data Collection Project

    International Nuclear Information System (INIS)

    Cillik, I.; Prochaska, J.

    2002-01-01

    The paper describes the way and results of human reliability data analysis collected as a part of the Bohunice Simulator Data Collection Project (BSDCP), which was performed by VUJE Trnava, Inc. with funding support from the U.S. DOE, National Nuclear Security Administration. The goal of the project was to create a methodology for simulator data collection and analysis to support activities in probabilistic safety assessment (PSA) and human reliability assessment for Jaslovske Bohunice nuclear power plant consisting of two sets of twin units: two VVER 440/V-230 (V1) and two VVER 440/V-213 (V2) reactors. During the project, training of V-2 control room crews was performed at VUJE Trnava simulator. The simulator training and the data collection were done in parallel. The main goal of BSDCP was to collect suitable data of human errors under simulated conditions requiring the use of symptom-based emergency operating procedures (SBEOPs). The subjects of the data collection were scenario progress time data, operator errors, and real-time technological parameters. The paper contains three main parts. The first part presents preparatory work and semi-automatic computer-based methods used to collect data and to check technological parameters in order to find hidden errors of operators, to be able to retrace the course of each scenario for purposes of further analysis, and to document the whole training process. The first part gives also an overview of collected data scope, human error taxonomy, and state classifications for SBEOP instructions coding. The second part describes analytical work undertaken to describe time distribution necessary for execution of various kinds of instructions performed by operators according to the classification for coding of SBEOP instructions. It also presents the methods used for determination of probability distribution for various operator errors. Results from the data evaluation are presented in the last part of the paper. An overview of

  9. Water in the March river radioactively contaminated

    International Nuclear Information System (INIS)

    Englander, A.G.

    1990-01-01

    A curve of the tritium contamination of the March river measured in Austria from 1976 to 1989 is shown. The conjecture is put forward that this contamination is caused by the Dukovani power plant in the neighbouring CSFR. Further measurements are called for

  10. Effect of Formulation Methods on the Mechanical and Release ...

    African Journals Online (AJOL)

    The mechanical properties of the tablets were assessed using the crushing strength (CS), friability(F) and crushing strength-friability ratio (CSFR) of the tablets, while drug release properties were assessed using disintegration time and dissolution profile. The granules possessed better flow properties than the powder ...

  11. Manufacture of disposal canisters

    International Nuclear Information System (INIS)

    Nolvi, L.

    2009-12-01

    The report summarizes the development work carried out in the manufacturing of disposal canister components, and present status, in readiness for manufacturing, of the components for use in assembly of spent nuclear fuel disposal canister. The disposal canister consist of two major components: the nodular graphite cast iron insert and overpack of oxygen-free copper. The manufacturing process for copper components begins with a cylindrical cast copper billet. Three different manufacturing processes i.e. pierce and draw, extrusion and forging are being developed, which produce a seamless copper tube or a tube with an integrated bottom. The pierce and draw process, Posiva's reference method, makes an integrated bottom possible and only the lid requires welding. Inserts for BWR-element are cast with 12 square channels and inserts for VVER 440-element with 12 round channels. Inserts for EPR-elements have four square channels. Casting of BWR insert type has been studied so far. Experience of casting inserts for PWR, which is similar to the EPR-type, has been got in co-operation with SKB. The report describes the processes being developed for manufacture of disposal canister components and some results of the manufacturing experiments are presented. Quality assurance and quality control in manufacture of canister component is described. (orig.)

  12. Restart of the Armenia-2 Nuclear Power Station: Radiological emergency preparedness considerations for the nearby American community

    International Nuclear Information System (INIS)

    Vargo, G.J.; Sherwood, G.L.

    1996-01-01

    The Armenia Nuclear Power Station is located at Metsamor, approximately 30 km NW of the capital, Yerevan. The station, a two-unit, first-generation Soviet-designed VVER-440/270 pressurized water reactor plant was closed following the 1988 earthquake near Spitak. Because of a severe energy shortage the Government of Armenia has undertaken a program to recommission Unit 2. The plant design and circumstances surrounding its closure caused members of the U.S. Embassy staff and the American community in Armenia to express concerns for their safety in the event of a radiological emergency. In response, two representatives from the U.S. Department of Energy's International Nuclear Safety Program traveled to Armenia to review the Status of radiological emergency preparedness, meet with the American community, and make protective action recommendations. In this presentation we examine the major issues associated with recommissioning of Armenia-2, the challenges involved with developing a radiological emergency preparedness program for the American community, and our recommendations for protective actions in the absence of a strong communications and radiological monitoring infrastructure

  13. Building Up an On-Line Plant Information System for the Emergency Response Center of the Hungarian Nuclear Safety Directorate

    International Nuclear Information System (INIS)

    Vegh, Janos; Major, Csaba; Horvath, Csaba; Hozer, Zoltan; Adorjan, Ferenc; Lux, Ivan; Horvath, Kristof

    2002-01-01

    The main design features, services, and human-machine interface characteristics are described of the CERTA VITA on-line plant information system developed and installed by KFKI AEKI at the Nuclear Safety Directorate (NSD) of the Hungarian Atomic Energy Authority (HAEA) in cooperation with experts from the NSD. The Center for Emergency Response, Training, and Analysis (CERTA) located at the headquarters of NSD, Budapest, Hungary, was established in 1997. The center supports the NSD installation, radiological monitoring, and advisory team in case of nuclear emergencies, with appropriate hardware and software for communication, diagnosis, prognosis, and prediction. The vital information transfer and analysis (VITA) system represents an important part of the CERTA, as it provides for the continuous remote inspection of the four VVER-440/V213 units of the Hungarian Paks nuclear power plant (NPP). The on-line information system maintains a continuous data link with the NPP through a managed leased line that connects CERTA to a gateway computer located at the Paks NPP. The present scope of the system is a result of a 4-yr development project: In addition to the basic safety parameter display functions, the VITA system now includes an on-line break parameter estimation module, an extensive training package based on simulated transients, and on-line data transfer capabilities to feed accident diagnosis/analysis codes

  14. Verification of the AZNHEX code v.1.4 with MCNP6 for different reference cases

    International Nuclear Information System (INIS)

    Galicia A, J.; Francois L, J. L.; Bastida O, G. E.; Del Valle G, E.

    2017-09-01

    The codes that make up the AZTLAN platform (AZTHECA, AZTRAN, AZKIND and AZNHEX) are currently in the testing phase simulating a variety of nuclear reactor assemblies and cores to compare and validate the results obtained for a particular case, with codes globally used in the nuclear area such as CASMO, Serpent and MCNP. The objective of this work is to continue improving the future versions of the codes of the AZTLAN platform so that accurate and reliable results can be obtained for the user. To test the current version of the AZNHEX code, 3 cases were taken into account, the first being the simulation of a VVER-440 reactor assembly; for the second case, the assembly of a fast reactor cooled with helium was simulated and for the third case it was decided to take up the case of the core of a fast reactor cooled with sodium, this because the previous versions of AZNHEX did not show adequate results and, in addition, they presented a considerable amount of limitations. The comparison and validation of the results (neutron multiplication factor, radial power, radial flow, axial power) for these three cases were made using the code MCNP6. The results obtained show that this version of AZNHEX produces values of the neutron multiplication factor and the neutron and power flow distributions very close to those of MCNP6. (Author)

  15. Verification of the AZNHEX code v.1.4 with MCNP6 for different reference cases; Verificacion del codigo AZNHEX v.1.4 con MCNP6 para diferentes casos de referencia

    Energy Technology Data Exchange (ETDEWEB)

    Galicia A, J.; Francois L, J. L.; Bastida O, G. E. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, Circuito Exterior s/n, 04510 Ciudad de Mexico (Mexico); Del Valle G, E., E-mail: jgaliciaa87@gmail.com [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, 07738 Ciudad de Mexico (Mexico)

    2017-09-15

    The codes that make up the AZTLAN platform (AZTHECA, AZTRAN, AZKIND and AZNHEX) are currently in the testing phase simulating a variety of nuclear reactor assemblies and cores to compare and validate the results obtained for a particular case, with codes globally used in the nuclear area such as CASMO, Serpent and MCNP. The objective of this work is to continue improving the future versions of the codes of the AZTLAN platform so that accurate and reliable results can be obtained for the user. To test the current version of the AZNHEX code, 3 cases were taken into account, the first being the simulation of a VVER-440 reactor assembly; for the second case, the assembly of a fast reactor cooled with helium was simulated and for the third case it was decided to take up the case of the core of a fast reactor cooled with sodium, this because the previous versions of AZNHEX did not show adequate results and, in addition, they presented a considerable amount of limitations. The comparison and validation of the results (neutron multiplication factor, radial power, radial flow, axial power) for these three cases were made using the code MCNP6. The results obtained show that this version of AZNHEX produces values of the neutron multiplication factor and the neutron and power flow distributions very close to those of MCNP6. (Author)

  16. Results of questionnaire for the needs of measured data for the steady-state calculations

    Energy Technology Data Exchange (ETDEWEB)

    Yrjoelae, V. [VTT Energy, Espoo (Finland)

    1995-12-31

    In the First International Seminar on the Modelling of Horizontal Steam Generators arranged in March 1991 was agreed to arrange a common calculational exercise to calculate the secondary side flow conditions during normal plant operation. OKB Gidropress of Russia supplied the experimental results for the exercise. They included some measured data of the local velocities and void fractions for the steam generators of the VVER-440 and VVER-1000 type reactors. The results of the common calculational exercise presented in the Second International Seminar in September 1992 were still mainly preliminary and it was felt necessary to continue these efforts. It was concluded that the given experimental results were not sufficient for a real code assessment - still too many quantities have to be guessed. It was pointed out that it is advisable to define a minimum set of necessary data. For this reason it was decided that VTT should made a query among the participants of the seminar, where they can give their opinion of the essential data. In this presentation the results of the questionnaire are given.

  17. Proposal of In-vessel corium retention concept for Paks NPP

    International Nuclear Information System (INIS)

    Elter, J.; Toth, E.; Matejovic, P.

    2011-01-01

    The in-vessel corium retention (IVR) via external reactor vessel cooling (ERVC) seems to be a promising severe accident management strategy not only for new generation of advanced PWRs, but also for VVER-440/V213 reactors, which were designed several years ago. The basic idea of in-vessel retention of corium is to prevent RPV failure by flooding the reactor cavity so that the reactor pressure vessel is submerged in water up to its support structures, and thus the decay heat can be transferred from the corium pool through the vessel wall and into the water surrounding the vessel. An IVR concept with simple ECVR loop based only on minor modifications of existing plant technology was proposed for the Paks Nuclear Power Plant. 2 severe accident (LB and SB LOCA) without availability of HP and LP safety injection in power upgrade (108%) conditions were simulated using the ASTEC code. The analyses show that the proposed solution is effective in preserving RPV integrity in the case of severe accident. Possible uncertainties in code predictions are covered by the applied conservative assumptions

  18. 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)

  19. 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.)

  20. 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)

  1. 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)

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

    International Nuclear Information System (INIS)

    Corsini, G.; Yelfimov, S.

    1995-01-01

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

  3. Modelling of thermal mechanical behaviour of high burn-Up VVER fuel at power transients with special emphasis on the impact of fission gas induced swelling of fuel pellets

    International Nuclear Information System (INIS)

    Novikov, V.; Medvedev, A.; Khvostov, G.; Bogatyr, S.; Kuzetsov, V.; Korystin, L.

    2005-01-01

    This paper is devoted to the modelling of unsteady state mechanical and thermo-physical behaviour of high burn-up VVER fuel at a power ramp. The contribution of the processes related to the kinetics of fission gas to the consequences of pellet-clad mechanical interaction is analysed by the example of integral VVER-440 rod 9 from the R7 experimental series, with a pellet burn-up in the active part at around 60 MWd/kgU. This fuel rod incurred ramp testing with a ramp value ΔW 1 ∼ 250 W/cm in the MIR research reactor. The experimentally revealed residual deformation of the clad by 30-40 microns in the 'hottest' portion of the rod, reaching a maximum linear power of up to 430 W/cm, is numerically justified on the basis of accounting for the unsteady state swelling and additional degradation of fuel thermal conductivity due to temperature-induced formation and development of gaseous porosity within the grains and on the grain boundaries. The good prediction capability of the START-3 code, coupled with the advanced model of fission gas related processes, with regard to the important mechanical (residual deformation of clad, pellet-clad gap size, central hole filling), thermal physical (fission gas release) and micro-structural (profiles of intra-granular concentration of the retained fission gas and fuel porosity across a pellet) consequences of the R7 test is shown. (authors)

  4. 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)

  5. Resolving the issues arisen during the development of License Renewal Application at PAKS NPP

    International Nuclear Information System (INIS)

    Ratkai, Sandor; Katona, Tamas Janos

    2012-01-01

    Operational license of the four VVER-440/213 units at Paks NPP, Hungary is limited to the design lifetime of 30 years. Extension by an additional 20 years of the original license is one of the main goals of the plant owner. In 2008 a programme for long-term operation (LTO) was developed and submitted to the Hungarian Atomic Energy Authority. The LTO Programme defines the activities for ensuring the extension of the operational lifetime and contains the justification of the safe LTO. The LTO Programme has to be implemented and comprehensive justification of the safe LTO has to be provided in the formal License Renewal Applications unit by unit. This work has been completed by the end of 2011, and the application has been submitted for approval for the Unit 1. The authority review and the approval of the License Renewal Application for Unit 1 should be finished before the expiration of the original design lifetime in 2012. In line with the regulations and supporting the License Renewal Application, a large number of engineering tasks have been performed. In this paper the entire project will be reported. The issues will be discussed, which have been arisen during the development of the application. The difficulties caused by the Hungarian technical and regulatory peculiarities will be presented. (author)

  6. Some uncertainty results obtained by the statistical version of the KARATE code system related to core design and safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Panka, Istvan; Hegyi, Gyoergy; Maraczy, Csaba; Temesvari, Emese [Hungarian Academy of Sciences, Budapest (Hungary). Reactor Analysis Dept.

    2017-11-15

    The best-estimate KARATE code system has been widely used for core design calculations and simulations of slow transients of VVER reactors. Recently there has been an increasing need for assessing the uncertainties of such calculations by propagating the basic input uncertainties of the models through the full calculation chain. In order to determine the uncertainties of quantities of interest during the burnup, the statistical version of the KARATE code system has been elaborated. In the first part of the paper, the main features of the new code system are discussed. The applied statistical method is based on Monte-Carlo sampling of the considered input data taking into account mainly the covariance matrices of the cross sections and/or the technological uncertainties. In the second part of the paper, only the uncertainties of cross sections are considered and an equilibrium cycle related to a VVER-440 type reactor is investigated. The burnup dependence of the uncertainties of some safety related parameters (e.g. critical boron concentration, rod worth, feedback coefficients, assembly-wise radial power and burnup distribution) are discussed and compared to the recently used limits.

  7. Results of questionnaire for the needs of measured data for the steady-state calculations

    International Nuclear Information System (INIS)

    Yrjoelae, V.

    1995-01-01

    In the First International Seminar on the Modelling of Horizontal Steam Generators arranged in March 1991 was agreed to arrange a common calculational exercise to calculate the secondary side flow conditions during normal plant operation. OKB Gidropress of Russia supplied the experimental results for the exercise. They included some measured data of the local velocities and void fractions for the steam generators of the VVER-440 and VVER-1000 type reactors. The results of the common calculational exercise presented in the Second International Seminar in September 1992 were still mainly preliminary and it was felt necessary to continue these efforts. It was concluded that the given experimental results were not sufficient for a real code assessment - still too many quantities have to be guessed. It was pointed out that it is advisable to define a minimum set of necessary data. For this reason it was decided that VTT should made a query among the participants of the seminar, where they can give their opinion of the essential data. In this presentation the results of the questionnaire are given

  8. 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

  9. Investigation and evaluation of erosion-corrosion status of the secondary side steam-water system at Paks Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Schunk, J.; Patek, G.; Pinter, T.; Baracska Varju, I.; Doma, A.; Kovacs, A.; Nemeth, P.; Tilky, P. [Paks Nuclear Power Plant Ltd (Hungary); Osz, J. [Budapest Univ. of Science and Economy (Hungary)

    2009-07-01

    There were four identical VVER-440 type units installed at Paks Nuclear Power Plant with 440 MW electrical capacity of each, between 1982 and 1987. Based on inherent reserve capacity of these units, the operational and maintenance experiences and examples of power upgrading abroad it was decided to increase the power of our units in a stepwise way. Presently, this upgrading project is close to finalization resulting in 500 MW electrical capacity of each unit. To realize the planned electrical power increase it was necessary to increase the primary and secondary heat power, which requested the increase of mass flow, temperature and pressure of secondary coolant flowing through the steam generators. According to the preliminary expert evaluations these increased parameters would not cause such an increase in humidity of steam leaving the turbine houses that could adversely impair our long term power upgrading plans. The steam humidity was determined on our units and the values did not even exceed the original design values. In spite of these preliminary investigations, significant accumulation of erosion-corrosion products at different places of secondary circuit was found and erosion damages of some secondary side equipment were discovered as a probable consequence of power upgrading. An extensive evaluation programme has been started involving experts and institutes of material testing, chemistry and hydrodynamics. Results and details of that work are given in our presentation. (authors)

  10. Support of the Ukrainian supervisory authority in establishing a modern nuclear power plant monitoring

    International Nuclear Information System (INIS)

    Beyer, M.; Carl, H.; Schumann, P.; Seidel, A.; Weiss, F.P.; Zschau, J.; Nowak, K.

    2000-01-01

    The type of monitoring of nuclear power plants in Ukraine practiced in the early nineties provided the supervisory authority with only inadequate access to information about the current safety status of plants. For the Zaporozhye nuclear power plant, unit 5, a technical system to improve operational monitoring has been designed, installed and commissioned for trial operation at the end of 1995 as a pilot project. The system complements existing operational checking and monitoring facilities by including modern means of information technology. It enables concentration on a continuous monitoring of the state of unit 5 in normal operation and in cases of anormalies or incidents so that when recognisable deviations from the regular plant operation occur, the authority can immediately inquire and if necessary impose conditions on the operator. In 1997, the Information and Crisis Centre of the Ukraninian supervisory authority in Kiev was equipped with the most essential technical means necessary for quasi-simultaneous transfer of data and voice and for monitoring purposes and connected to the Centre to the Zaporozhye system. A similar monitoring system for both VVER-440 units of the Rovno nuclear power plant by analogy with the pilot project was specified and put into operation and connected to the ICC in 1998. (orig.) [de

  11. Simulation of the SPE-4 small-break loss-of-coolant accident

    International Nuclear Information System (INIS)

    Cebull, P.; Hassan, Y.A.

    1993-01-01

    A small-break loss of coolant accident (SBLOCA) conducted at the PMK-2 integral test facility was analyzed using RELAP5/MOD3. 1. The experiment simulated a 7.4% break in the cold leg of a VVER-440/213-type nuclear power plant as part of the International Atomic Energy Agency's Fourth Standard Problem Exercise (SPE-4). The VVER design differs from pressurized water reactors (PWRS) of western origin, primarily in its use of horizontal steam generators, hot- and cold-leg loop seals, and safety injection tanks. Because of these differences, it will exhibit somewhat different transient behavior than most PWRS. The PMK-2 test facility, located at the KFKI Atomic Energy Research Institute (AEKI), is a scale model of the Paks nuclear power plant in Hungary with scaling factors of 1:2070 in power and volume and 1:1 in elevation. Primarily used to study SBLOCAs and natural circulation behavior of VVER reactors, it has been used in three previous SPEs

  12. PACTEL and PWR PACTEL Test Facilities for Versatile LWR Applications

    Directory of Open Access Journals (Sweden)

    Virpi Kouhia

    2012-01-01

    Full Text Available This paper describes construction and experimental research activities with two test facilities, PACTEL and PWR PACTEL. The PACTEL facility, comprising of reactor pressure vessel parts, three loops with horizontal steam generators, a pressurizer, and emergency core cooling systems, was designed to model the thermal-hydraulic behaviour of VVER-440-type reactors. The facility has been utilized in miscellaneous applications and experiments, for example, in the OECD International Standard Problem ISP-33. PACTEL has been upgraded and modified on a case-by-case basis. The latest facility configuration, the PWR PACTEL facility, was constructed for research activities associated with the EPR-type reactor. A significant design basis is to utilize certain parts of PACTEL, and at the same time, to focus on a proper construction of two new loops and vertical steam generators with an extensive instrumentation. The PWR PACTEL benchmark exercise was launched in 2010 with a small break loss-of-coolant accident test as the chosen transient. Both facilities, PACTEL and PWR PACTEL, are maintained fully operational side by side.

  13. PACTEL and PWR PACTEL Test Facilities for Versatile LWR Applications

    International Nuclear Information System (INIS)

    Virpi Kouhia, V.; Purhonen, H.; Riikonen, V.; Puustinen, M.; Kyrki-Rajamaki, R.; Vihavainen, J.

    2012-01-01

    This paper describes construction and experimental research activities with two test facilities, PACTEL and PWR PACTEL. The PACTEL facility, comprising of reactor pressure vessel parts, three loops with horizontal steam generators, a pressurizer, and emergency core cooling systems, was designed to model the thermal-hydraulic behaviour of VVER-440-type reactors. The facility has been utilized in miscellaneous applications and experiments, for example, in the OECD International Standard Problem ISP-33. PACTEL has been upgraded and modified on a case-by-case basis. The latest facility configuration, the PWR PACTEL facility, was constructed for research activities associated with the EPR-type reactor. A significant design basis is to utilize certain parts of PACTEL, and at the same time, to focus on a proper construction of two new loops and vertical steam generators with an extensive instrumentation. The PWR PACTEL benchmark exercise was launched in 2010 with a small break loss-of-coolant accident test as the chosen transient. Both facilities, PACTEL and PWR PACTEL, are maintained fully operational side by side.

  14. Integrated experimental test program on waterhammer pressure pulses and associated structural responses within a feedwater sparger

    International Nuclear Information System (INIS)

    Nurkkala, P.; Hoikkanen, J.

    1997-01-01

    This paper describes the methods and systems as utilized in an integrated experimental thermohydraulic/mechanics analysis test program on waterhammer pressure pulses within a revised feedwater sparger of a Loviisa generation VVER-440-type reactor. This program was carried out in two stages: (1) measurements with a strictly limited set of operating parameters at Loviisa NPP, and (2) measurements with the full set of operating parameters on a test article simulating the revised feedwater sparger. The experiments at Loviisa NPS served as an invaluable source of information on the nature of waterhammer pressure pulses and structural responses. These tests thus helped to set the objectives and formulate the concept for series of tests on a test article to study the water hammer phenomena. The heavily instrumented full size test article of a steam generator feedwater sparger was placed within a pressure vessel simulating the steam generator. The feedwater sparger was subjected to the full range of operating parameters which were to result in waterhammer pressure pulse trains of various magnitudes and duration. Two different designs of revised feedwater sparger were investigated (i.e. 'grounded' and 'with goose neck'). The following objects were to be met within this program: (1) establish the thermohydraulic parameters that facilitate the occurrence of water hammer pressure pulses, (2) provide a database for further analysis of the pressure pulse phenomena, (3) establish location and severity of these water hammer pressure pulses, (4) establish the structural response due to these pressure pulses, (5) provide input data for structural integrity analysis. (orig.)

  15. Development of 3D multi-group neutron diffusion code for hexagonal geometry

    International Nuclear Information System (INIS)

    Sun Wei; Wang Kan; Ni Dongyang; Li Qing

    2013-01-01

    Based on the theory of new flux expansion nodal method to solve the neutron diffusion equations, the intra-nodal fluence rate distribution was expanded in a series of analytic basic functions for each group. In order to improve the accuracy of calculation result, continuities of neutron fluence rate and current were utilized across the nodal surfaces. According to the boundary conditions, the iteration method was adopted to solve the diffusion equation, where inner iteration speedup method is Gauss-Seidel method and outer is Lyusternik-Wagner. A new speedup method (one-outer-iteration and multi-inner-iteration method) was proposed according to the characteristic that the convergence speed of multiplication factor is faster than that of neutron fluence rate and the update of inner iteration matrix is slow. Based on the proposed model, the code HANDF-D was developed and tested by 3D two-group vver440 benchmark, experiment 2 of HFETR, 3D four-group thermal reactor benchmark, and 3D seven-group fast reactor benchmark. The numerical results show that HANDF-D can predict accurately the multiplication factor and nodal powers. (authors)

  16. Experiments with the HORUS-II test facility

    Energy Technology Data Exchange (ETDEWEB)

    Alt, S.; Lischke, W. [Univ. for Applied Sciences Zittau/Goerlitz, Zittau (Germany). Dept. of Nuclear Engineering

    1997-12-31

    Within the scope of the German reactor safety research the thermohydraulic computer code ATHLET which was developed for accident analyses of western nuclear power plants is more and more used for the accident analysis of VVER-plants particularly for VVER-440,V-213. The experiments with the HORUS-facilities and the analyses with the ATHLET-code have been realized at the Technical University Zittau/Goerlitz since 1991. The aim of the investigations was to improve and verify the condensation model particularly the correlations for the calculation of the heat transfer coefficients in the ATHLET-code for pure steam and steam-noncondensing gas mixtures in horizontal tubes. About 130 condensation experiments have been performed at the HORUS-II facility. The experiments have been carried out with pure steam as well as with noncondensing gas injections into the steam mass flow. The experimental simulations are characterized as accident simulation tests for SBLOCA for VVER-conditions. The simulation conditions had been adjusted correspondingly to the parameters of a postulated SBLOCA`s fourth phase at the original plant. 4 refs.

  17. Replacement of nickel sealing rings by expanded graphite sealing rings -upgrading of SG primary collector flange connection

    Energy Technology Data Exchange (ETDEWEB)

    Cikryt, F.; Bednarek, L.; Kusyn, L. [Vitkovice, Ostrava (Switzerland)

    1997-12-31

    One of the most loaded parts of a steam generator of VVER 440 MW type are the bolts and thread holes of the primary collector cover sealing set. The strength calculations and tensometric measurings performed during operation proved the high degree of a load on the bolts. The conditions of the stress limitation are not met in some cases according to the pertinent standards. The untightnesses at nickel rings occurred during putting the units of Jaslovske Bohunice and Dukovany nuclear power stations into operation. With regard to improve the reliability, the producer has taken measures to improve the quality of the rings and users have introduced more strict regulations for bolts tightening. Due to these measures the high reliability of the set has been obtained from point of view of the tightness, but substantial reduction of bolts and holes threads loading have not been obtained. Several years operation experience proved relatively low service of bolts, damage of thread holes and sealing grooves. As the degree of mechanical load is one of the vital parameters influencing the damage of sealing set, in 1996 we started with the works aimed at a possibility of nickel sealing rings replacement for a more modern type of sealing which assure the higher reliability and service life of the individual part of sealing set under the reduced load.

  18. Progress in implementation of the neutronics model of HEXTRAN into APROS

    International Nuclear Information System (INIS)

    Rintala, J.

    2009-01-01

    A new three-dimensional nodal model for neutronics calculation is currently under implementation into APROS - Advanced PROcess Simulation environment - to conform the increasing accuracy requirements. The new model is based on an advanced nodal code HEXTRAN and its static version HEXBU-3D by VTT, Technical Research Centre of Finland. However, several improvements for the model are made and the whole model has been reprogrammed. They don't change the theory basement of the method, but rather makes the implementation more flexible. Currently the computational part of the program is ready and current work concentrates on testing and validation. User interface details and usability issues need also work in the future. In this paper, general information about the improvements of the theory is explained first. Then the latest validation results are given. Currently the dynamical characteristics are tested by calculating the AER's kinetic benchmarks for VVER-440 reactors. In this paper, the results for the first benchmark are shown for two version of the code. The first version is fully HEXTRAN-comparable code to test that the basic structure works as wanted. The second version is the actual improved model for APROS. (author)

  19. Introducing advanced ISI requirements at Paks NPP for supporting the LTO

    International Nuclear Information System (INIS)

    Trampus, P.; Ratkai, S.

    2012-01-01

    The four VVER-440 model 213 units in operation at Paks NPP, Hungary, are facing to approach their licensed term of operation, which is 30 years. To extend the safe operation of the units beyond the original licensed term by additional 20 years belongs to the highest priorities of the owners/operator of MVM Paks NPP. According to the nuclear legislation, a formal license renewal application for the extended period has to be submitted to the Hungarian Atomic Energy Authority. A significant feature of the license renewal process is the demonstration of the effectiveness of the currently applied ageing management program. ISI is an essential part of the ageing management program thus the adequate ISI techniques and the tailor made requirements have to be incorporated in it. To cope with the expectations originating from the LTO at Paks NPP, it was decided to replace the original Soviet based ISI system by the widely applied ASME BPVC Section XI requirements. Additionally, in 2011 a new nuclear regulation was issued in Hungary, in which the ISI requirements have also been changed. This paper intends to present the entire structure of the new Hungarian regulation related to the ISI but mainly focusing on the deviation to the ASME Section XI with the perspective of the licence renewal. (author)

  20. Results of questionnaire for the needs of measured data for the steady-state calculations

    Energy Technology Data Exchange (ETDEWEB)

    Yrjoelae, V [VTT Energy, Espoo (Finland)

    1996-12-31

    In the First International Seminar on the Modelling of Horizontal Steam Generators arranged in March 1991 was agreed to arrange a common calculational exercise to calculate the secondary side flow conditions during normal plant operation. OKB Gidropress of Russia supplied the experimental results for the exercise. They included some measured data of the local velocities and void fractions for the steam generators of the VVER-440 and VVER-1000 type reactors. The results of the common calculational exercise presented in the Second International Seminar in September 1992 were still mainly preliminary and it was felt necessary to continue these efforts. It was concluded that the given experimental results were not sufficient for a real code assessment - still too many quantities have to be guessed. It was pointed out that it is advisable to define a minimum set of necessary data. For this reason it was decided that VTT should made a query among the participants of the seminar, where they can give their opinion of the essential data. In this presentation the results of the questionnaire are given.

  1. Estimation of Externalities for Juragua Nuclear Project

    International Nuclear Information System (INIS)

    Mora, H. R.; Carbonell, L. T.

    2002-01-01

    Estimation of externalities allows taking into account environmental impacts due to any activity in total costs calculation. In the present work, the external costs of electricity generation from nuclear energy were calculated considering three scenarios: normal operation (routine releases), accident situation and solid waste disposal. A comparison between these results and those obtained for electricity generation from fossil fuels was made. IAEA proposals of Simplified methodologies were used for externality calculations. The Juragua project was selected as a study case; it is based in two energetic blocks both PWR, VVER 440/318 type with a plant capacity of 417 MWe each. Four impact ways were considered for all scenarios: (1) Inhalation of radionuclides in the air, (2) External irradiation from radionuclides immersed in clouds, (3) External irradiation from deposited radionuclides and (4) Ingestion of radionuclides in agricultural products. Besides, two impact categories (local and regional) for all scenarios were considered. The total cost of externalities was 0.01425 c/kWh, value smaller than the one obtained for electricity generation from fossil fuel (0.256 c/kWh). For the normal operation scenario, the external cost calculated was 0.00112 c/kWh, for accident situation 0.01103 c/kWh, and for the solid wastes management scenario 0.0021 c/kWh. The high value obtained for solid waste disposal scenario is due to repository placement features. (author)

  2. CATHARE2 calculation of SPE-3 test small break loca on PMK facility

    Energy Technology Data Exchange (ETDEWEB)

    Laugier, E.; Radet, J. [Institut de Protection et de Surete Nucleaire, Cadarache (France)

    1995-09-01

    Bind and post test calculations with CATHARE2 have been performed concerning the SPE-4 exercise organized under the auspices of IAEA on the hungarian PMK-2 facility, a one loop scaled model of VVER 440/213 Nuclear Power Plant. The SPE-4 test is a cold leg SBLOCA associated to a {open_quotes}bleed and feed{close_quotes} procedure applied in the secondary circuit. The present paper is devoted to the analysis of the post test calculation. For the first part of the transient (until the end of the SIT activations), the primary and secondary pressures are rather well predicted, leading to a good agreement with the experimental trips, as scram, flow coast down, SIT beginning and end of activation. Nevertheless, some discrepancy with the experiment may be due to an over prediction of the thermal exchanges from the primary to the secondary circuits. For the second part of the transient, the predicted primary circuit repressurization is shifted after the SITs are off, while in the experiment this event immediately follows the end of SIT activation. The delay in the calculation leads to underpredict primary and secondary pressures, thus anticipating the timing of events, such as LPIS and emergency feedwater activation.

  3. PLEX at Paks: making a virtue out of necessity

    International Nuclear Information System (INIS)

    Katona, T.; Bajsz, J.

    1992-01-01

    There are four VVER-440 units in operation at Paks Nuclear plant in Hungary. The units are of the V-213 type, ie the design which approaches the current, demanding Western standards in many aspects. During construction a number of innovations were adopted in these units. In particular, the instrumentation and control system was thoroughly improved, but there were also some changes to the main components (eg the pressure vessel). Although the quality assurance carried out during construction and commissioning was not the same as in Western practice, it was effective and resulted in relatively well constructed and tested units. Due to this, Pak has a good foundation on which to build a life extension programme-high availability, quality upgrading, a high integrity pressure vessel and a careful operating policy. Although life extension, economically speaking, is a necessity and not an option for Paks, the programme in itself should bring other benefits which would pay for themselves within the plant's design lifetime. (author)

  4. INSTALLATION OF A POST-ACCIDENT CONFINEMENT HIGH-LEVEL RADIATION MONITORING SYSTEM IN THE KOLA NUCLEAR POWER STATION (UNIT 2) IN RUSSIA

    Energy Technology Data Exchange (ETDEWEB)

    GREENE,G.A.; GUPPY,J.G.

    1998-09-01

    This is the final report on the INSP project entitled, ``Post-Accident Confinement High-Level Radiation Monitoring System'' conducted by BNL under the authorization of Project Work Plan WBS 1.2.2.6 (Attachment 1). This project was initiated in February 1993 to assist the Russians in reducing risks associated with the continued operation of older Soviet-designed nuclear power plants, specifically the Kola VVER-440/230 Unit 2, through improved accident detection capability, specifically by the installation of a dual train high-level radiation detection system in the confinement of Unit 2 of the Kola NPP. The major technical objective of this project was to provide, install and make operational the necessary hardware inside the confinement of the Kola NPP Unit 2 to provide early and reliable warning of the release of radionuclides from the reactor into the confinement air space as an indication of the occurrence of a severe accident at the plant. In addition, it was intended to provide hands-on experience and training to the Russian plant workers in the installation, operation, calibration and maintenance of the equipment in order that they may use the equipment without continued US assistance as an effective measure to improve reactor safety at the plant.

  5. BUC implementation in Slovakia

    International Nuclear Information System (INIS)

    Chrapciak, V.; Vaclav, J.

    2009-01-01

    Improved calculation methods allow one to take credit for the reactivity reduction associated with fuel burnup. This means reducing the analysis conservatism while maintaining an adequate criticality safety margin. Application of burnup credit (BUC) requires knowledge of the reactivity state of the irradiated fuel for which BUC is taken. The isotopic inventory and reactivity has to be calculated with validated codes. We use in Slovakia Gd 2 fuel with maximal enrichment of fuel pins 4.4%. Our transport and storage basket KZ-48 with boron steel is licensed for fresh fuel with enrichment 4.4%. In near future (2011 or 2012) we will use a new fuel with maximal enrichment of fuel pins 4.9%. For this fuel we plan to use existing KZ-48 with BUC application. In cooperation with Slovak Nuclear Regulatory Authority (UJD) we have started several years ago process of BUC implementation in Slovakia for VVER-440 reactors. We have already prepared methodology according IAEA methodology. We have validated computational systems (SCALE 5.1 already, SCALE 6 in progress). UJD will prepare regulation about BUC application in Slovakia. Last item is preparation of safety reports (for transport and storage) for the new fuel with average enrichment 4.87% in basket KZ-48 with BUC application.

  6. Prediction of iodine activity peak during refuelling

    International Nuclear Information System (INIS)

    Hozer, Z.; Vajda, N.

    2001-01-01

    The increase of fission product activities in the primary circuit of a nuclear power plant indicates the existence of defects in some fuel rods. The power change leads to the cooling down of the fuel and results in the fragmentation of the UO 2 pellets, which facilitates the release of fission products from the intergranular regions. Furthermore the injection of boric acid after shutdown will increase the primary activity, due to the solution of deposited fission products from the surface of the core components. The calculation of these phenomena usually is based on the evaluation of activity measurements and power plant data. The estimation of iodine spiking peak during reactor transients is based on correlation with operating parameters, such as reactor power and primary pressure. The approach used in the present method was applied for CANDU reactors. The VVER-440 specific correlations were determined using the activity measurements of the Paks NPP and the data provided by the Russian fuel supplier. The present method is used for the evaluation of the iodine isotopes, as well as the noble gases. A numerical model has been developed for iodine spiking simulation and has been validated against several shutdown transients, measured at Paks NPP. (R.P.)

  7. In-Vessel Retention via External Reactor Cooling

    Energy Technology Data Exchange (ETDEWEB)

    Bachrata, Andrea [CTU in Prague, Faculty of nuclear sciences and physical engineering, V Holesovickach 2 180 00, Prague 8 (Czech republic)

    2008-07-01

    In-vessel (corium) retention (IVR) via external reactor pressure vessel (RPV) cooling is considered to be an effective severe accident management strategy for corium localisation and stabilisation. The main idea of IVR strategy consists in flooding the reactor cavity and transferring the decay heat through the wall of RPV to the recirculating water and than to the atmosphere of the containment of nuclear power plant. The aim of this strategy is to localise and to stabilise the corium inside the RPV. Not using this procedure could destroy the integrity of RPV and might cause the interaction of the corium with the concrete at the bed of the reactor cavity. Several experimental facilities and computer codes (MVITA, ASTEC module DIVA and CFD codes) were applied to simulate the IVR strategy for concrete reactor designs. The necessary technical modifications concerning the implementation of IVR concept were applied at the Loviisa NPP (VVER-440/V213). This strategy is also an important part of the advanced reactor designs AP600 and AP1000. (authors)

  8. Simulation of a beyond design-basis-accident with RELAP5/MOD3.1

    Energy Technology Data Exchange (ETDEWEB)

    Banati, J. [Lappeenranta Univ. of Technology, Lappeenranta (Finland)

    1995-09-01

    This paper summarizes the results of analyses, parametric and sensitivity studies, performed using the RELAP5/MOD3.1 computer code for the 4th IAEA Standard Problem Exercise (SPE-4). The test, conducted on the PMK-2 facility in Budapest, involved simulation of a Small Break Loss Of Coolant Accident (SBLOCA) with a 7.4% break in the cold leg of a VVER-440 type pressurized water reactor. According to the scenario, the unavailability of the high pressure injection system led to a beyond design basis accident. For prevention of core damage, secondary side bleed-and-feed accident management measures were applied. A brief description of the PMK-2 integral type test facility is presented, together with the profile and some key phenomenological aspects of this particular experiment. Emphasis is placed on the ability of the code to predict the main trends observed in the test and thus, an assessment is given for the code capabilities to represent the system transient.

  9. Multiple condensation induced water hammer events, experiments and theoretical investigations

    International Nuclear Information System (INIS)

    Barna, Imre Ferenc; Ezsoel, Gyoergy

    2011-01-01

    We investigate steam condensation induced water hammer (CIWH) phenomena and present experimental and theoretical results. Some of the experiments were performed in the PMK-2 facility, which is a full-pressure thermalhydraulic model of the nuclear power plant of VVER-440/312 type and located in the Atomic Energy Research Institute Budapest, Hungary. Other experiments were done in the ROSA facility in Japan. On the theoretical side CIWH is studied and analyzed with the WAHA3 model based on two-phase flow six first-order partial differential equations that present one dimensional, surface averaged mass, momentum and energy balances. A second order accurate high-resolution shockcapturing numerical scheme was applied with different kind of limiters in the numerical calculations. The applied two-fluid model shows some similarities to RELAP5 which is widely used in the nuclear industry to simulate nuclear power plant accidents. New features are the existence of multiple, independent CIWH pressure peaks both in experiments and in simulations. Experimentally measured and theoretically calculated CIWH pressure peaks are in qualitative agreement. However, the computational results are very sensitive against flow velocity. (orig.)

  10. Experimental and theoretical study of steam condensation induced water hammer phenomena

    International Nuclear Information System (INIS)

    Barna, Imre Ferenc; Baranyai, Gabor; Ezsoel, Gyoergy

    2009-01-01

    We investigate steam condensation induced water hammer (waha) phenomena and present experimental and theoretical results. Some of the experiments were performed in the PMK-2 facility, which is a full-pressure thermohydraulic model of the nuclear power plant of VVER-440/312 type and located in the Atomic Energy Research Institute Budapest, Hungary. Other experiments were done in the ROSA facility in Japan. On the theoretical side waha is studied and analyzed with the WAHA3 model based on two-phase flow six first-order partial differential equations that present one dimensional, surface averaged mass, momentum and energy balances. A second order accurate high-resolution shock-capturing numerical scheme was applied with different kind of limiters in the numerical calculations. The applied two-fluid model shows some similarities to Relap5 which is widely used in the nuclear industry to simulate nuclear power plant accidents. Experimentally measured and theoretically calculated waha pressure peaks are in qualitative agreement. (author)

  11. 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

  12. Studies on the assessment and validation of reactor dynamics models used in Finland

    International Nuclear Information System (INIS)

    Vanttola, T.

    1993-10-01

    Two reactor dynamics related computer codes of the calculation system at the Technical Research Centre of Finland have been assessed. The codes TRAB and SMATRA, have been examined from two points of view. First, models of some critical phenomena determining the worst fuel rod conditions during reactor transients have been evaluated on the basis of experimental information. Second, the the overall behaviour of the codes describing the dynamics of the reactor core and its cooling system has been studied on the basis of simulation of real transients and of performed safety analyses of selected accidents. The emphasis is on the VVER-440 reactors, but the generality of the methods has been demonstrated by showing that the key phenomena of the Chernobyl accident can be reproduced and analysed using the same calculation system. In the study the separate phenomena examined are single- and two-phase friction, post DNB heat transfer and critical heat flux in the VVER rod bundle. (60 refs., 11 figs., 4 tabs.)

  13. Experimental investigation of the coolability of blocked hexagonal bundles

    Energy Technology Data Exchange (ETDEWEB)

    Hózer, Zoltán, E-mail: zoltan.hozer@energia.mta.hu; Nagy, Imre; Kunstár, Mihály; Szabó, Péter; Vér, Nóra; Farkas, Róbert; Trosztel, István; Vimi, András

    2017-06-15

    Highlights: • Experiments were performed with electrically heated hexagonal fuel bundles. • Coolability of ballooned VVER-440 type bundle was confirmed up to high blockage rate. • Pellet relocation effect causes delay in the cool-down of the bundle. • The bypass line does not prevent the reflood of ballooned fuel rods. - Abstract: The CODEX-COOL experimental series was carried out in order to evaluate the effect of ballooning and pellet relocation in hexagonal bundles on the coolability of fuel rods after a LOCA event. The effects of blockage geometry, coolant flowrate, initial temperature and axial profile were investigated. The experimental results confirmed that a VVER bundle up to 80% blockage rate remains coolable after a LOCA event under design basis conditions. The ballooned section creates some obstacles for the cooling water during reflood of the bundle, but this effect causes only a short delay in the cooling down of the hot fuel rods. The accumulation of fuel pellet debris in the ballooned volume results in a local power peak, which leads to further slowing down of quench front.

  14. Theoretical study of steam condensation induced water hammer phenomena in horizontal pipelines

    International Nuclear Information System (INIS)

    Barna, Imre Ferenc; Pocsai, Mihaly Andras; Pecs Univ.; Guba, Attila; Imre, Attila Rikard; Budapest University of Technology and Economics

    2015-01-01

    Steam condensation induced water hammer (CIWH) phenomena are investigated and new theoretical results are presented. We use the WAHA3 model based on two-phase flow six first-order partial differential equations that present one dimensional, surface averaged mass, momentum and energy balances. A second order accurate high-resolution shock-capturing numerical scheme was applied with different kind of limiters in the numerical calculations. The applied two-fluid model shows some similarities to RELAP5 which is widely used in the nuclear industry to simulate nuclear power plant accidents. This model was validated with different CIWH experiments which were performed in the PMK-2 facility, which is a full-pressure thermohydraulic model of the nuclear power plant of VVER-440/312 type in the Energy Research Center of the Hungarian Academy of Sciences and in the Rosa facility of the Japan Atomic Energy Agency. In our present study we show the first part of a planned large database which will give us the upper and lower flooding mass flow rates for various pipe geometries where CIWH can happen. Such a reliable database would be a great help for future reactor constructions and scheming.

  15. Recent improvements in on-line core supervision at Loviisa NPP

    International Nuclear Information System (INIS)

    Antila, M.; Kuusisto, J.

    1999-01-01

    On-line core supervision system (RESU) based on monitoring of local fuel limits has been in use at the Loviisa VVER-440 reactors for more than twenty years. Minor modifications were made ten years ago when the computer hardware was upgraded. In April 1998 Loviisa got the licence for 1500 MW power. Power uprating and introduction of new fuel types gave rise to the latest improvements in the core supervision system, which is called RESU-98. In August 1999 the Finnish Safety Authority (STUK) has given approval for RESU-98, which is now in use at the Loviisa NPP. RESU-98 includes essentially the same computer codes, which are used in reload planning. The extensive in-core instrumentation is utilised to adjust the theoretical 3D-power distribution to get a best-estimate results. In this paper a general review of the RESU-98 system is given including instrumentation, methods, core monitoring, predictive functions and validation. Special attention is paid on the recent improvements. (Authors)

  16. Simulation of the IAEA's fourth Standard Problem Exercise small-break loss-of-coolant accident using RELAP5/MOD.3.1

    International Nuclear Information System (INIS)

    Cebull, P.P.; Hassan, Y.A.

    1995-01-01

    A small-break loss-of-coolant accident experiment conducted at the PMK-2 integral test facility in Hungary is analyzed using the RELAP5/MOD3.1 thermal-hydraulic code. The experiment simulated a 7.4% break in the cold leg of a VVER-440/213-type nuclear power plant as part of the International Atomic Energy Agency's Fourth Standard Problem Exercise (SPE-4). Blind calculations of the exercise are presented, and the timing of various events throughout the transient is discussed. A posttest analysis is performed in which the sensitivity of the calculated results is investigated. The code RELAP5 predicts most of the transient events well, although a few problems are noted, particularly the failure of RELAP5 to predict dryout in the core even through the collapsed liquid level fell below the top of the heated portion. A discrepancy between the predicted primary mass inventory distribution and the experimental data is identified. Finally, the primary and secondary pressures calculated by RELAP5 fell too rapidly during the latter part of the transient, resulting in rather large errors in the predicted timing of some pressure-actuated events

  17. Recent improvements and new features in the Westinghouse lattice physics codes

    International Nuclear Information System (INIS)

    Huria, H.C.; Buechel, R.J.

    1995-01-01

    Westinghouse has been using the ANC three-dimensional, two-energy-group nodal model for nuclear analysis and fuel management calculations for standard pressurized water reactor (PWR) reload design analysis since 1988. The cross sections are obtained from PHOENIX-P, a modified version of the PHOENIX lattice physics code for all square-assembly PWR cores. The PHOENIX-H code was developed for modeling both the VVER-1000 and VVER-440 fuel lattice configurations. The PHOENIX-H code has evolved from PHOENIX-P, the primary difference being in the neutronic solution modules. The PHOENIX-P code determines the assembly flux distribution using integral transport theory-based pin-cell nodal coupling followed by two-dimensional discrete ordinates solution in x-y geometry. The PHOENIX-H code uses the two-dimensional heterogeneous response method. The other infrastructure is identical in both the codes, and they share the same 42-group cross-section library

  18. PMK-2 the Hungarian integral type test facility. Documentations, publications and archivations of experiments

    International Nuclear Information System (INIS)

    Perneczky, L.; Guba, A.; Ezsoel, G.; Toth, I.; Szabados, L.

    2002-01-01

    The PMK-2 experimental facility at the KFKI-AEKI, Budapest, is a full pressure, scaled down model of the primary and partly the secondary circuit of the Paks NPP, which is equipped with four VVER-440/213-type reactors. Since the start-up of the facility altogether 48 experiments have been performed for groups of transients as follows: one- and two-phase natural circulation, loss of coolant accidents, special plant transients and experiments in support of accident management procedures. The results have been used for the validation of thermal-hydraulic system codes for VVER applications. Following the experiments a detailed documentation and archiving activity - using an optimised data storage - was required to preserve the essential information and to assure these for a widely utilisation for the international nuclear community. In the publication list related to the facility and the experiments for the moment altogether 280 items - documents, articles in periodicals, papers in proceedings and research reports - in six languages were collected. The paper gives an overview on this activity including the participation in the EU CERTA-TN programme, where AEKI introduced representative databases of two PMK-2 tests in the STRESA Network.(author)

  19. Supply of appropriate nuclear technology for the developing world: small power reactors for electricity generation

    International Nuclear Information System (INIS)

    Heising-Goodman, C.D.

    1981-01-01

    This paper reviews the supply of small nuclear power plants (200 to 500 MWe electrical generating capacity) available on today's market, including the pre-fabricated designs of the United Kingdom's Rolls Royce Ltd and the French Alsthom-Atlantique Company. Also, the Russian VVER-440 conventionally built light-water reactor design is reviewed, including information on the Soviet Union's plans for expansion of its reactor-building capacity. A section of the paper also explores the characteristics of LDC electricity grids, reviewing methods available for incorporating larger plants into smaller grids as the Israelis are planning. Future trends in reactor supply and effects on proliferation rates are also discussed, reviewing the potential of the Indian 220 MWe pressurised heavy-water reactor, South Korean and Jananese potential for reactor exports in the Far East, and the Argentine-Brazilian nuclear programme in Latin America. This study suggests that small reactor designs for electrical power production and other applications, such as seawater desalination, can be made economical relative to diesel technology if traditional scaling laws can be altered by adopting and standardising a pre-fabricated nuclear power plant design. Also, economy can be gained if sufficient attention is concentrated on the design, construction and operating experience of suitably sized conventionally built reactor systems. (author)

  20. Theoretical study of steam condensation induced water hammer phenomena in horizontal pipelines

    Energy Technology Data Exchange (ETDEWEB)

    Barna, Imre Ferenc [Hungarian Academy of Sciences, Budapest (Hungary). Wigner Research Center; ELI-HU Nonprofit Kft., Szeged (Hungary); Pocsai, Mihaly Andras [Hungarian Academy of Sciences, Budapest (Hungary). Wigner Research Center; Pecs Univ. (Hungary). Inst. of Physics; Guba, Attila [Hungarian Academy of Sciences, Budapest (Hungary). Energy Research Center; Imre, Attila Rikard [Hungarian Academy of Sciences, Budapest (Hungary). Energy Research Center; Budapest University of Technology and Economics (Hungary). Dept. of Energy Engineering

    2015-11-15

    Steam condensation induced water hammer (CIWH) phenomena are investigated and new theoretical results are presented. We use the WAHA3 model based on two-phase flow six first-order partial differential equations that present one dimensional, surface averaged mass, momentum and energy balances. A second order accurate high-resolution shock-capturing numerical scheme was applied with different kind of limiters in the numerical calculations. The applied two-fluid model shows some similarities to RELAP5 which is widely used in the nuclear industry to simulate nuclear power plant accidents. This model was validated with different CIWH experiments which were performed in the PMK-2 facility, which is a full-pressure thermohydraulic model of the nuclear power plant of VVER-440/312 type in the Energy Research Center of the Hungarian Academy of Sciences and in the Rosa facility of the Japan Atomic Energy Agency. In our present study we show the first part of a planned large database which will give us the upper and lower flooding mass flow rates for various pipe geometries where CIWH can happen. Such a reliable database would be a great help for future reactor constructions and scheming.

  1. Multifunctional optimised scope simulators in Central and Eastern Europe

    International Nuclear Information System (INIS)

    Bartak, J.; Hauesberger, P.; Dalleur, J.P.; Houard, J.

    1999-01-01

    In the field of operator training, multiple functions have to be covered such as basic principles training, training on specific systems, operations training addressing operating procedures in normal, incidental and accidental situations, plant physical phenomena analysis. Training simulators are appropriate tools to meet theses needs. Optimisation of the scope of simulation is required to meet specific training objectives and produce cost-effective solutions that allow for possible future extensions. Training needs and training programs have to be identified with the participation of final users, leading to the development of appropriate training materials: 'multifunctional' (also called analytical) optimised scope simulators are a concrete solution to meeting this challenge. For these simulators, the quality of physical models used is equivalent to that used in the full-scope replica-type simulators. Moreover, all state-of-the-art technical requirements in terms of development of training simulators, must be satisfied: realism of modelling, tolerances, simulated incidents and accidents. Examples of this concept will be illustrated in the paper through the presentation of recent developments of simulators in Central and Eastern European NPPs (VVER-1000, VVER-440, RBMK, BN600, PWR 600). A brief presentation of the software workshop used to develop these simulators concludes the paper. (author)

  2. Acoustic leak detection at complicated geometrical structures using fuzzy logic and neural networks

    International Nuclear Information System (INIS)

    Hessel, G.; Schmitt, W.; Weiss, F.P.

    1993-10-01

    An acoustic method based on pattern recognition is being developed. During the learning phase, the localization classifier is trained with sound patterns that are generated with simulated leaks at all locations endangered by leak. The patterns are extracted from the signals of an appropriate sensor array. After training unknown leak positions can be recognized through comparison with the training patterns. The experimental part is performed at an acoustic 1:3 model of the reactor vessel and head and at an original VVER-440 reactor in the former NPP Greifswald. The leaks were simulated at the vessel head using mobile sound sources driven either by compressed air, a piezoelectric transmitter or by a thin metal blade excited through a jet of compressed air. The sound patterns of the simulated leaks are simultaneously detected with an AE-sensor array and with high frequency microphones measuring structure-borne sound and airborne sound, respectively. Pattern classifiers based on Fuzzy Pattern Classification (FPC) and Artificial Neural Networks (ANN) are currently tested for validation of the acoustic emission-sensor array (FPC), leak localization via structure-borne sound (FPC) and the leak localization using microphones (ANN). The initial results show the used classifiers principally to be capable of detecting and locating leaks, but they also show that further investigations are necessary to develop a reliable method applicable at NPPs. (orig./HP)

  3. INSTALLATION OF A POST-ACCIDENT CONFINEMENT HIGH-LEVEL RADIATION MONITORING SYSTEM IN THE KOLA NUCLEAR POWER STATION (UNIT 2) IN RUSSIA

    International Nuclear Information System (INIS)

    GREENE, G.A.; GUPPY, J.G.

    1998-01-01

    This is the final report on the INSP project entitled, ''Post-Accident Confinement High-Level Radiation Monitoring System'' conducted by BNL under the authorization of Project Work Plan WBS 1.2.2.6 (Attachment 1). This project was initiated in February 1993 to assist the Russians in reducing risks associated with the continued operation of older Soviet-designed nuclear power plants, specifically the Kola VVER-440/230 Unit 2, through improved accident detection capability, specifically by the installation of a dual train high-level radiation detection system in the confinement of Unit 2 of the Kola NPP. The major technical objective of this project was to provide, install and make operational the necessary hardware inside the confinement of the Kola NPP Unit 2 to provide early and reliable warning of the release of radionuclides from the reactor into the confinement air space as an indication of the occurrence of a severe accident at the plant. In addition, it was intended to provide hands-on experience and training to the Russian plant workers in the installation, operation, calibration and maintenance of the equipment in order that they may use the equipment without continued US assistance as an effective measure to improve reactor safety at the plant

  4. The Role of Nuclear Power in Hungary, a Regional Comparison

    International Nuclear Information System (INIS)

    Cserhati, A.

    2016-01-01

    The presentation and paper are giving the broad picture on the country's nuclear competence and the atomic electricity production in international appraisal. The Central and Eastern Europe (CEE) region nuclear statistics were compiled by the author and appear in compact, easily understandable graphical form: Worldwide and regional figures of nuclear share for electricity generation. Types of operating nuclear power plant units. Age and performance (cumulative load factors) of units. Power uprate history of affected units. Status of the long term operation (or by earlier name: lifetime extension) projects. The cycle length expansion is a transition from 12 to 15 month fuel cycle, as a pioneering initiative of Paks NPP for efficiency increase of VVER-440 type. New build trends will be outlined in general for the world, Europe and the CEE region. The last topic is the Hungarian path to Paks-5 and Paks-6 completion. Preparatory projects (Teller and Levai), establishment of the Paks II company. Potential suppliers and types for the new build (AREVA EPR, ATMEA Atmea1, KEPCO/KHNP APR1400, Rosatom AES-2006, Toshiba-Westinghouse AP1000). Intergovernmental agreement with Russian Federation for building of two 1200 MW units. Financing of 80 percent of the investment by Russian loan with moderate interest rates. Planned schedule of the new build. EC infringement procedures. Public acceptance. (author).

  5. The EOP Visualization Module Integrated into the Plasma On-Line Nuclear Power Plant Safety Monitoring and Assessment System

    International Nuclear Information System (INIS)

    Hornaes, Arne; Hulsund, John Einar; Vegh, Janos; Major, Csaba; Horvath, Csaba; Lipcsei, Sandor; Kapocs, Gyoergy

    2001-01-01

    An ambitious project to replace the unit information systems (UISs) at the Hungarian Paks nuclear power plant was started in 1998-99. The basic aim of the reconstruction project is to install a modern, distributed UIS architecture on all four Paks VVER-440 units. The new UIS includes an on-line plant safety monitoring and assessment system (PLASMA), which contains a critical safety functions monitoring module and provides extensive operator support during the execution of the new, symptom-oriented emergency operating procedures (EOPs). PLASMA includes a comprehensive EOP visualization module, based on the COPMA-III procedure-handling software developed by the Organization for Economic Cooperation and Development, Halden Reactor Project. Intranet technology is applied for the presentation of the EOPs with the use of a standard hypertext markup language (HTML) browser as a visualization tool. The basic design characteristics of the system, with a detailed description of its user interface and functions of the new EOP display module, are presented

  6. Current state of spent fuel management in the Russian Federation

    International Nuclear Information System (INIS)

    Makarchuk, T.F.; Spichev, V.V.; Tikhonov, N.S.; Simanovsky, V.M.; Tokarenko, A.I.; Bespalov, V.N.

    1998-01-01

    Twenty nine power units of nine nuclear power plants of total installed capacity 22 GW(e) are now in operation in the Russian Federation. They produce approximately 12% of electric power in the country. The annual spent fuel arising is about 790 tU. The spent fuel from VVER-440 and BN-600 is reprocessed at the RT-1 plant near Chelyabinsk. The VVER-1000 spent fuel is planned to be reprocessed at the reprocessing plant RT-2 which is under construction near Krasnoyarsk. The RBMK-1000 spent fuel is not reprocessed because of its low fissile content. It is meant to be stored in intermediate storage facilities at the NPP sites and in a centralized storage facility during a period not less than 50 years and then to be disposed of in geological formations. State of the art of spent fuel reprocessing, storage and transportation is considered in the paper. Problems of nuclear fuel cycle back-end in Russia are taken into account. (author)

  7. Used mixed oxide fuel reprocessing at RT-1 plant

    Energy Technology Data Exchange (ETDEWEB)

    Kolupaev, D.; Logunov, M.; Mashkin, A.; Bugrov, K.; Korchenkin, K. [FSUE PA ' Mayak' , 30, Lenins str, Ozersk, 460065 (Russian Federation); Shadrin, A.; Dvoeglazov, K. [ITCP ' PRORYV' , 2/8 Malaya Krasmoselskay str, Moscow, 107140 (Russian Federation)

    2016-07-01

    Reprocessing of the mixed uranium-plutonium spent nuclear fuel of the BN-600 reactor was performed at the RT-1 plant twice, in 2012 and 2014. In total, 8 fuel assemblies with a burn-up from 73 to 89 GW day/t and the cooling time from 17 to 21 years were reprocessed. The reprocessing included the stages of dissolution, clarification, extraction separation of U and Pu with purification from the fission products, refining of uranium and plutonium at the relevant refining cycles. Dissolution of the fuel composition of MOX used nuclear fuel (UNF) in nitric acid solutions in the presence of fluoride ion has occurred with the full transfer of actinides into solution. Due to the high content of Pu extraction separation of U and Pu was carried out on a nuclear-safe equipment designed for the reprocessing of highly enriched U spent nuclear fuel and Pu refining. Technological processes of extraction, separation and refining of actinides proceeded without deviations from the normal mode. The output flow of the extraction outlets in their compositions corresponded to the regulatory norms and remained at the level of the compositions of the streams resulting from the reprocessing of fuel types typical for the RT-1 plant. No increased losses of Pu into waste have been registered during the reprocessing of BN-600 MOX UNF an compare with VVER-440 uranium UNF reprocessing. (authors)

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  9. 3D optical measuring technologies for dimensional inspection

    International Nuclear Information System (INIS)

    Chugui, Yu V

    2005-01-01

    The results of the R and D activity of TDI SIE SB RAS in the field of the 3D optical measuring technologies and systems for noncontact 3D optical dimensional inspection applied to atomic and railway industry safety problems are presented. This activity includes investigations of diffraction phenomena on some 3D objects, using the original constructive calculation method, development of hole inspection method on the base of diffractive optical elements. Ensuring the safety of nuclear reactors and running trains as well as their high exploitation reliability takes a noncontact inspection of geometrical parameters of their components. For this tasks we have developed methods and produced the technical vision measuring systems LMM, CONTROL, PROFILE, and technologies for non-contact 3D dimensional inspection of grid spacers and fuel elements for the nuclear reactor VVER-1000 and VVER-440, as well as automatic laser diagnostic system COMPLEX for noncontact inspection of geometrical parameters of running freight car wheel pairs. The performances of these systems and the results of the industrial testing at atomic and railway companies are presented

  10. Design report of the disposal canister for twelve fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Raiko, H. [VTT Energy, Espoo (Finland); Salo, J.P. [Posiva Oy, Helsinki (Finland)

    1999-05-01

    The report provides a summary of the design of the canister for final disposal of spent nuclear fuel. The canister structure consists of a cylindrical massive nodular graphite cast iron insert covered by a 50 mm thick copper overlay. The capacity of the canister is 12 assemblies of BWR or VVER 440 fuel. The canister shall be tight with a high probability for about 100 000 years. The good and long lasting tightness requires: (1) The good initial tightness that is achieved by high quality requirements and extensive quality control, (2) The good corrosion resistance, which is obtained by the overpack of oxygen free copper, and (3) Mechanical strength of the canister, that is ensured by analyses (the following loads are considered: hydrostatic pressure, even and uneven swelling pressure of bentonite, thermal effects, and elevated hydrostatic pressure during glaciation. The allowed stresses and strains are set in such a way that reasonable engineering safety factors are obtained in all assessed design base loading cases). The canister shall limit the radiation dose rate outside the canister to minimise the radiolysis of the water in the vicinity of the canister. The canister insert shall keep the fuel assemblies in a subcritical configuration even if the void in the canister is filled with water due to postulated leakage. The design basis of the canister is set, the performed analyses are summarised and the results are assessed and discussed in the report. (orig.) 35 refs.

  11. Nuclear power in the Ukraine: Problems and prospects

    International Nuclear Information System (INIS)

    Nigmatullin, N.R.

    1995-01-01

    Nuclear power production in the Ukraine started in 1977 with the startup of the first 1000-MW power-generating unit at the Chernobyl nuclear power plant. During the period from 1977 to 1989 sixteen power-generating units with a total electric capacity of 14,880 MW were put into operation at five nuclear power plants: ten VVER-1000, two VVER-440, and four RBMK-1000. As a result of the accident in 1986 in the fourth power-generating unit and the fire in 1991 in the second power-generating unit of the Chernobyl nuclear power plant, these units are no longer operating. Therefore the total installed nuclear power plant capacity is 12,880 MW. Moreover, the construction of three more power-generating units with VVER-1000 reactors is almost completed at three nuclear power plants - Zaporozh'e, Roven, and Khmel'nitsk. These units are not in operation because of the moratorium announced by the Supreme Council of Ukraine. In connection with the Council's decision, the Chernobyl nuclear power plant should have been shut down in 1993

  12. Experiments with the HORUS-II test facility

    Energy Technology Data Exchange (ETDEWEB)

    Alt, S; Lischke, W [Univ. for Applied Sciences Zittau/Goerlitz, Zittau (Germany). Dept. of Nuclear Engineering

    1998-12-31

    Within the scope of the German reactor safety research the thermohydraulic computer code ATHLET which was developed for accident analyses of western nuclear power plants is more and more used for the accident analysis of VVER-plants particularly for VVER-440,V-213. The experiments with the HORUS-facilities and the analyses with the ATHLET-code have been realized at the Technical University Zittau/Goerlitz since 1991. The aim of the investigations was to improve and verify the condensation model particularly the correlations for the calculation of the heat transfer coefficients in the ATHLET-code for pure steam and steam-noncondensing gas mixtures in horizontal tubes. About 130 condensation experiments have been performed at the HORUS-II facility. The experiments have been carried out with pure steam as well as with noncondensing gas injections into the steam mass flow. The experimental simulations are characterized as accident simulation tests for SBLOCA for VVER-conditions. The simulation conditions had been adjusted correspondingly to the parameters of a postulated SBLOCA`s fourth phase at the original plant. 4 refs.

  13. Replacement of nickel sealing rings by expanded graphite sealing rings -upgrading of SG primary collector flange connection

    Energy Technology Data Exchange (ETDEWEB)

    Cikryt, F; Bednarek, L; Kusyn, L [Vitkovice, Ostrava (Switzerland)

    1998-12-31

    One of the most loaded parts of a steam generator of VVER 440 MW type are the bolts and thread holes of the primary collector cover sealing set. The strength calculations and tensometric measurings performed during operation proved the high degree of a load on the bolts. The conditions of the stress limitation are not met in some cases according to the pertinent standards. The untightnesses at nickel rings occurred during putting the units of Jaslovske Bohunice and Dukovany nuclear power stations into operation. With regard to improve the reliability, the producer has taken measures to improve the quality of the rings and users have introduced more strict regulations for bolts tightening. Due to these measures the high reliability of the set has been obtained from point of view of the tightness, but substantial reduction of bolts and holes threads loading have not been obtained. Several years operation experience proved relatively low service of bolts, damage of thread holes and sealing grooves. As the degree of mechanical load is one of the vital parameters influencing the damage of sealing set, in 1996 we started with the works aimed at a possibility of nickel sealing rings replacement for a more modern type of sealing which assure the higher reliability and service life of the individual part of sealing set under the reduced load.

  14. Bulgaria: INIS Center - 45 years experience

    International Nuclear Information System (INIS)

    Georgieva, Albena

    2015-01-01

    Bulgaria is one of 35 countries in the world operating nuclear power plants. Bulgaria's nuclear program was launched in 1956 with the construction of an IRT-2000 research reactor at the Institute for Nuclear Research and Nuclear Energy (INRNE), which was commissioned in 1961. The reactor is now under reconstruction. In 1960, construction of the first Bulgarian nuclear power plant started. At the moment, there are 6 power units at the Kozloduy NPP site; 4 of them (VVER-440/B-230) under decommissioning and 2 (VVER-1000/B-320) in operation. Several storage facilities for radioactive waste, mainly from the Kozloduy NPP and from various sources of ionizing radiation in medicine and industry are also in operation. The Kozloduy NPP, INRNE, Sofia University, the Technical University, and the State Enterprise Radioactive Waste are the main generators of nuclear information in Bulgaria and the main consumers of INIS products. The Bulgarian INIS Center, therefore, maintains continuous and effective cooperation with these Institutions

  15. Design report of the disposal canister for twelve fuel assemblies

    International Nuclear Information System (INIS)

    Raiko, H.; Salo, J.P.

    1999-05-01

    The report provides a summary of the design of the canister for final disposal of spent nuclear fuel. The canister structure consists of a cylindrical massive nodular graphite cast iron insert covered by a 50 mm thick copper overlay. The capacity of the canister is 12 assemblies of BWR or VVER 440 fuel. The canister shall be tight with a high probability for about 100 000 years. The good and long lasting tightness requires: (1) The good initial tightness that is achieved by high quality requirements and extensive quality control, (2) The good corrosion resistance, which is obtained by the overpack of oxygen free copper, and (3) Mechanical strength of the canister, that is ensured by analyses (the following loads are considered: hydrostatic pressure, even and uneven swelling pressure of bentonite, thermal effects, and elevated hydrostatic pressure during glaciation. The allowed stresses and strains are set in such a way that reasonable engineering safety factors are obtained in all assessed design base loading cases). The canister shall limit the radiation dose rate outside the canister to minimise the radiolysis of the water in the vicinity of the canister. The canister insert shall keep the fuel assemblies in a subcritical configuration even if the void in the canister is filled with water due to postulated leakage. The design basis of the canister is set, the performed analyses are summarised and the results are assessed and discussed in the report. (orig.)

  16. 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)

  17. The evolution of CANDU containment design

    International Nuclear Information System (INIS)

    Pendergast, Duane R.; Meneley, Daniel A.

    1995-01-01

    This paper reviews Canadian containment design, reflects on forces which have shaped the evolving designs and contemplates future containment design subject to existing constraints of Canadian and International nuclear power regulations. The discussion mentions modifications which could play a role in easing customary restrictions on siting while meeting the intent and literal definition of Canadian licensing requirements. Many different containment concepts have been considered and deployed over the years. Professor Birkhofer points out that the earliest systems installed were 'dry containment'. These rely on large containment buildings designed to contain all the the steam from a loss of coolant. It was soon realized the large water inventory of boiling water reactors would require very large and expensive containment buildings. This led to the development of pressure suppression concepts. Various pressure suppression schemes have been devised to condense steam and thus reduce design pressure and volume requirements. Examples include the forcing of steam into condensing water pools of BWR's steam condensation by passage into arrays of ice, dousing spray systems in CANDU reactors, and the spray condenser systems of late model Soviet designed VVER - 440 reactors. The fundamental concept of energy removal by steam condensation can result in a small containment volume requiring low design pressure capability

  18. Recent improvements in on-line core supervision at Loviisa NPP

    International Nuclear Information System (INIS)

    Antila, M.; Kuusisto, J.

    2000-01-01

    AN on-line core supervision system (RESU) based on monitoring of local fuel limits has been in use at the Loviisa VVER-440 reactors for more than twenty years. Minor modifications were made ten years ago to upgrade the computer hardware. In April 1998 Loviisa obtained a licence for 1500 MW th power. Power up-rating and introduction of new fuel types gave rise to the latest improvements in the core supervision system, which is called RESU-98. In August 1999 the Finnish Safety Authority (STUCK) officially approved RESU-98, which is now in use at the Loviisa NPP. RESU-98 includes essentially the same computer codes, which are used in reload planning. The extensive in-core instrumentation is utilised to adjust the theoretical 3-D power distribution to get a best-estimate result. In this paper a general review of the RESU-98 system is given including instrumentation, methods, core monitoring, predictive functions and validation. Special attention is paid to recent improvements. (author)

  19. Experiments with the HORUS-II test facility

    International Nuclear Information System (INIS)

    Alt, S.; Lischke, W.

    1997-01-01

    Within the scope of the German reactor safety research the thermohydraulic computer code ATHLET which was developed for accident analyses of western nuclear power plants is more and more used for the accident analysis of VVER-plants particularly for VVER-440,V-213. The experiments with the HORUS-facilities and the analyses with the ATHLET-code have been realized at the Technical University Zittau/Goerlitz since 1991. The aim of the investigations was to improve and verify the condensation model particularly the correlations for the calculation of the heat transfer coefficients in the ATHLET-code for pure steam and steam-noncondensing gas mixtures in horizontal tubes. About 130 condensation experiments have been performed at the HORUS-II facility. The experiments have been carried out with pure steam as well as with noncondensing gas injections into the steam mass flow. The experimental simulations are characterized as accident simulation tests for SBLOCA for VVER-conditions. The simulation conditions had been adjusted correspondingly to the parameters of a postulated SBLOCA's fourth phase at the original plant

  20. AER benchmark specification sheet

    International Nuclear Information System (INIS)

    Aszodi, A.; Toth, S.

    2009-01-01

    In the VVER-440/213 type reactors, the core outlet temperature field is monitored with in-core thermocouples, which are installed above 210 fuel assemblies. These measured temperatures are used in determination of the fuel assembly powers and they have important role in the reactor power limitation. For these reasons, correct interpretation of the thermocouple signals is an important question. In order to interpret the signals in correct way, knowledge of the coolant mixing in the assembly heads is necessary. Computational Fluid Dynamics (CFD) codes and experiments can help to understand better these mixing processes and they can provide information which can support the more adequate interpretation of the thermocouple signals. This benchmark deals with the 3D CFD modeling of the coolant mixing in the heads of the profiled fuel assemblies with 12.2 mm rod pitch. Two assemblies of the 23rd cycle of the Paks NPP's Unit 3 are investigated. One of them has symmetrical pin power profile and another possesses inclined profile. (authors)

  1. The fifth AER dynamic benchmark calculation with hextran-smabre

    International Nuclear Information System (INIS)

    Haemaelaeinen, A.; Kyrki-Rajamaeki, R.

    1998-01-01

    The first AER benchmark for coupling of the thermohydraulic codes and three-dimensional reactordynamic core models is discussed. HEXTRAN 2.7 is used for the core dynamics and SMABRE 4.6 as a thermohydraulic model for the primary and secondary loops. The plant model for SMABRE is based mainly on two input models, the Loviisa model and standard VVER-440/213 plant model. The primary circuit includes six separate loops, totally 505 nodes and 652 junctions. The reactor pressure vessel is divided into six parallel channels. In HEXTRAN calculation 1/6 symmetry is used in the core. In the calculations nuclear data is based on the ENDF/B-IV library and it has been evaluated with the CASMO-HEX code. The importance of the nuclear data was illustrated by repeating the benchmark calculation with using three different data sets. Optimal extensive data valid from hot to cold conditions were not available for all types of fuel enrichments needed in this benchmark. (author)

  2. Improving the raw management in Ukraine

    Energy Technology Data Exchange (ETDEWEB)

    Jung, H.G.; Koever, M. [NUKEM Technologies GmbH, Alzenau (Germany)

    2013-07-01

    In Ukraine 13 VVER-1000 and 2 VVER-440 reactor units generate about 50 % of the country's energy production. By contrast, when looking at the large volume of radioactive waste of all categories, which is produced by these nuclear power plants (NPP), the radioactive waste (RAW) management - onsite of NPP, towards interim storage as well as towards final disposal - is still not adequately developed. Currently all operational waste of Ukrainian NPP is incompletely treated/conditioned and stored onsite, though insufficient storage capacity is available at the NPP. As no effective Ukraine-wide strategy is yet established to manage RAW beyond NPP a serious situation could arise, which even leads to constrained temporary shut-down of reactor units, threatening the energy supply of whole the country. In addition large quantities of RAW, partly badly sorted, derive from decommissioning of the shut-down Chernobyl NPP. Also other sources of RAW contribute to the whole volume, which has to be managed in Ukraine, as research reactors, reprocessing of spent nuclear fuel and other nuclear facilities and applications. Nevertheless, operational and shut-down NPP in Ukraine are by far the largest producers. To support Ukraine in managing their radioactive waste NUKEM Technologies GmbH was appointed to provide technical support for the improvement of the current situation in cooperation with responsible Ukrainian experts towards a targeted, comprehensive and effective RAW management. (orig.)

  3. Early detection and identification of anomalies in chemical regime based on computational intelligence techniques

    International Nuclear Information System (INIS)

    Figedy, Stefan; Smiesko, Ivan

    2012-01-01

    This article provides brief information about the fundamental features of a newly-developed diagnostic system for early detection and identification of anomalies being generated in water chemistry regime of the primary and secondary circuit of the VVER-440 reactor. This system, which is called SACHER (System of Analysis of CHEmical Regime), was installed within the major modernization project at the NPP-V2 Bohunice in the Slovak Republic. The SACHER system has been fully developed on MATLAB environment. It is based on computational intelligence techniques and inserts various elements of intelligent data processing modules for clustering, diagnosing, future prediction, signal validation, etc, into the overall chemical information system. The application of SACHER would essentially assist chemists to identify the current situation regarding anomalies being generated in the primary and secondary circuit water chemistry. This system is to be used for diagnostics and data handling, however it is not intended to fully replace the presence of experienced chemists to decide upon corrective actions. (author)

  4. System for measuring of air concentration in air-steam mixture during the transients

    International Nuclear Information System (INIS)

    Gorbenko, Gennady A.; Gakal, Pavlo G.; Epifanov, Konstantin S.; Osokin, Gennady V.; Smirnov, Sergey V.

    2006-01-01

    Description of system for air concentration measuring in air-steam mixture during the transients is represented. Air concentration measuring is based on discrete sampling method. The measuring system consists of sampler, transport pipeline, distributor and six measuring vessels. From the sampler air-steam mixture comes to distributor through transport pipeline and fills consecutively the measuring vessels. The true air concentration in place of measurement was defined based on measured air concentration in samples taken from measuring vessels. For this purpose, the mathematical model of transients in measuring system was developed. Air concentration transient in air-steam mixture in place of measurement was described in mathematical model by air concentration time-dependent function. The function parameters were defined based on air concentration measured in samples taken from measuring vessels. Estimated error of air concentration identification was about 10%. Measuring system was used in experiments on EREC BKV-213 test facility intended for testing of VVER-440/V-213 reactor barbotage-vacuum system

  5. In-pile loop experiments in water chemistry and corrosion

    International Nuclear Information System (INIS)

    Kysela, J.

    1986-09-01

    Results on the study of Zr-1% Nb alloy corrosion, in out-of and in-pile loops simulating the working conditions of the VVER-440 reactor (Soviet, PWR type), covered the time period May 1982-April 1986 were reported, as well as, results on transport and filtration of corrosion products. Methods and techniques used in the study included remote measurement of corrosion rate by polarizing resistance, out-of-pile loop at the temperature 350 deg. C, pressure 19 MPa, circulation 20 kgs/h and in-pile water loop with constant flow rate 10,000 kgs/h, pressure 16 MPa, temperature 330 deg. C and neutron flux 7x10 13 n/cm 2 .s. It was shown that solid suspended particles with chemical composition corresponding most frequently to magnetite or nickelous ferrite, though with non-stoichiometric composition Me x 2+ Fe 3- x 3+ O 4 were found. Continuous filtration of water by means of electromagnetic filter leads to a decrease of radioactivity of the outer epitactic layer only. Effect of filtration on the inner topotactic layer is negligible. The corrosion rates for the above-mentioned parameters are given

  6. Water regime of steam power plants

    International Nuclear Information System (INIS)

    Oesz, Janos

    2011-01-01

    The water regime of water-steam thermal power plants (secondary side of pressurized water reactors (PWR); fossil-fired thermal power plants - referred to as steam power plants) has changed in the past 30 years, due to a shift from water chemistry to water regime approach. The article summarizes measures (that have been realised by chemists of NPP Paks) on which the secondary side of NPP Paks has become a high purity water-steam power plant and by which the water chemistry stress corrosion risk of heat transfer tubes in the VVER-440 steam generators was minimized. The measures can also be applied to the water regime of fossil-fired thermal power plants with super- and subcritical steam pressure. Based on the reliability analogue of PWR steam generators, water regime can be defined as the harmony of construction, material(s) and water chemistry, which needs to be provided in not only the steam generators (boiler) but in each heat exchanger of steam power plant: - Construction determines the processes of flow, heat and mass transfer and their local inequalities; - Material(s) determines the minimal rate of general corrosion and the sensitivity for local corrosion damage; - Water chemistry influences the general corrosion of material(s) and the corrosion products transport, as well as the formation of local corrosion environment. (orig.)

  7. 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

  8. Risks and benefits of energy systems in Czechoslovakia

    International Nuclear Information System (INIS)

    Bohal, L.; Erban, P.; Kadlec, J.; Kraus, V.; Trcka, V.

    1984-01-01

    The paper describes the fundamental philosophy of an approach to risk and benefit assessment in the fuel and energy complex in Czechoslovakia. The first part analyses the need to solve the risk and benefit problems stemming from structural changes occurring in the Czechoslovakian fuel and energy complex. The second part describes main features of risk and benefit research with special respect to the fuel and energy complex defined within the framework of the national economy with interfaces to the relevant environment. Furthermore, a glimpse is given of how to assess, using the general philosophy, the risks and benefits of various developing variants of the fuel and energy complex. The third part deals with methodological aspects of such risk and benefit evaluation research with special consideration of the methods of long-term prediction in structural analysis and multi-measure assessment. Finally, further progress in solving these problems in VUPEK and some other Czechoslovakian scientific institutions is briefly noted. (author)

  9. Důsledky světové hospodářské krize pro Československo ve 30. letech 20. století

    OpenAIRE

    PAŘIL, Aleš

    2013-01-01

    This Bachelor's work deals with the consequences of the Great Depression for Czechoslovakia in the 1930's. To better understand the situation before the Great Depression the work begins with characteristic of Czechoslovakia in the 1920's. The course of this crisis is also shortly outlined there. This bachelor's work focuses primarily on the consequences of the Great Depression on the Czechoslovakian economy connected with the social security of the inhabitants. In this work we can also read a...

  10. Success of opposition against Temelin

    International Nuclear Information System (INIS)

    Tollmann, A.

    1989-01-01

    A first success for the anti-nuclear movement emerges: the Czechoslovakian government renounces of blocks 3 and 4, 1000 MW each, in Temelin. Although lack of money is admitted as a partial cause, the main cause is the broad opposition of the population, especially in Austria, says the author. Therefore the author appeals to the coworkers to double their efforts in the signature collection against Temelin. The slogan is: we shall make Temelin too. (qui)

  11. Literatura faktu na českém knižním trhu: portrét Grigorije Rasputina

    OpenAIRE

    Štollová, Jitka

    2012-01-01

    This bachelor thesis explores the development of non-fiction literature in the Czech and Czechoslovakian book markets by means of the analysis of publications focused on the character of Russian mystic Grigory Efimovich Rasputin. The main attention is paid to the question in which manner the heretofore released titles show the versatility of the genre of non-fiction literature. The thesis at first examines the history of non-fiction literature in the world and in Czechoslovakia and the Czech ...

  12. Srovnání československé II. a III. republiky z hlediska teorie autoritativních režimů Juana J. Linze

    OpenAIRE

    Svoboda, Ladislav

    2013-01-01

    The diploma thesis deals with pair of domestic regimes known as the second and third | Czechoslovakian republic. The choice of these periods of the national history was conditioned by a relative absence of comparative political scientific works on these periods and by timelessness and topicality of the studying of domestic regimes, that refused the idea of a liberal democracy twice in ten years. The goal of the diploma thesis is to analyze, compare and categorize the second and the third Czec...

  13. Reactor Dosimetry State of the Art 2008

    Science.gov (United States)

    Voorbraak, Wim; Debarberis, Luigi; D'Hondt, Pierre; Wagemans, Jan

    2009-08-01

    Oral session 1: Retrospective dosimetry. Retrospective dosimetry of VVER 440 reactor pressure vessel at the 3rd unit of Dukovany NPP / M. Marek ... [et al.]. Retrospective dosimetry study at the RPV of NPP Greifswald unit 1 / J. Konheiser ... [et al.]. Test of prototype detector for retrospective neutron dosimetry of reactor internals and vessel / K. Hayashi ... [et al.]. Neutron doses to the concrete vessel and tendons of a magnox reactor using retrospective dosimetry / D. A. Allen ... [et al.]. A retrospective dosimetry feasibility study for Atucha I / J. Wagemans ... [et al.]. Retrospective reactor dosimetry with zirconium alloy samples in a PWR / L. R. Greenwood and J. P. Foster -- Oral session 2: Experimental techniques. Characterizing the Time-dependent components of reactor n/y environments / P. J. Griffin, S. M. Luker and A. J. Suo-Anttila. Measurements of the recoil-ion response of silicon carbide detectors to fast neutrons / F. H. Ruddy, J. G. Seidel and F. Franceschini. Measurement of the neutron spectrum of the HB-4 cold source at the high flux isotope reactor at Oak Ridge National Laboratory / J. L. Robertson and E. B. Iverson. Feasibility of cavity ring-down laser spectroscopy for dose rate monitoring on nuclear reactor / H. Tomita ... [et al.]. Measuring transistor damage factors in a non-stable defect environment / D. B. King ... [et al.]. Neutron-detection based monitoring of void effects in boiling water reactors / J. Loberg ... [et al.] -- Poster session 1: Power reactor surveillance, retrospective dosimetry, benchmarks and inter-comparisons, adjustment methods, experimental techniques, transport calculations. Improved diagnostics for analysis of a reactor pulse radiation environment / S. M. Luker ... [et al.]. Simulation of the response of silicon carbide fast neutron detectors / F. Franceschini, F. H. Ruddy and B. Petrović. NSV A-3: a computer code for least-squares adjustment of neutron spectra and measured dosimeter responses / J. G

  14. 15 years of The Hungarian integral type test facility: horizontal SG related PMK-2 experiments

    International Nuclear Information System (INIS)

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

    2001-01-01

    The PMK-2 experimental facility at the KFKI-AEKI, Budapest, is a full pressure, scaled down model of the primary and partly the secondary circuit of the Paks Nuclear Power Plant. This NPP is equipped with four VVER-440/213-type reactors. Such plants are slightly different from PWRs of usual design and have a number of special features as 6-loop primary circuit, horizontal steam generators, loop seal in hot and cold legs, setpoint pressure of passive safety injection tanks (SIT) higher than secondary pressure, etc. The PMK-2 was primarily designed for investigating operational and off-normal transient processes, as well as small-break loss of coolant accidents of Paks NPP. The volume and power scaling ratios are 1:2070. Due to the importance of gravitational forces in both single- and two-phase flow the elevation ratio is 1:1 except for the lower plenum and pressuriser. The six loops of the plant are modelled by a single active loop. Transients can be started from nominal operating conditions. The pressuriser (PRZ) is connected to the lower part of the hot leg as in the reference system. The core model consists of 19 electrically heated rods. The main circulating pump of the PMK-2 serves to produce the nominal operating conditions and to simulate the flow coast-down following pump trip. The horizontal design of the VVER-440 steam generator is modelled by horizontal heat transfer tubes between hot and cold vertical collectors in the primary side. The emergency core cooling systems including the SITs. High and low pressure injection systems of the Paks NPP are also modelled. The first design of the PMK-NVH facility only modelled the primary circuit of plant. This version was used until 1990. The PMK-2 facility is an upgraded version (first of all by addition of a controlled secondary heat removal system) extending the capability of the test loop to modelling transient processes evoked by initiating events in the secondary circuit or including accident sequences in

  15. Database for the OECD-IAEA Paks Fuel Project

    International Nuclear Information System (INIS)

    Szabo, Emese; Hozer, Zoltan; Gyori, Csaba; Hegyi, Gyoergy

    2010-01-01

    really necessary for the analytical work. The present version of the database can be extended according to the requests of project participants and considering the availability of requested data. The present version of the database was collected by the experts of AEKI in close cooperation with Paks NPP. The work was supported by the International Atomic Energy Agency (IAEA) and the Hungarian Atomic Energy Authority (HAEA). The database has been prepared to support three types of calculations: - Thermal-hydraulic calculations to describe how the inadequate cooling conditions could have established during the incident. - Fuel behaviour simulation to describe the oxidation and degradation mechanisms of fuel assemblies. - Activity release and transport calculations to simulate the release of fission products from the failed fuel rods. The database includes the following main parts: - Design characteristics of VVER-440 fuel assemblies (main geometrical data, some mechanical properties, oxidation kinetics of Zr1%Nb cladding, and integral data of assemblies). - Operational data of damaged fuel assemblies (power histories of fuel assemblies, burnup, fuel rod internal pressure, isotope inventories, decay heat and axial power distribution). - Design characteristic of the cleaning tank (main geometrical data). - Measured data during the incident: (temperature, water level measurements, cleaning tank outlet flowrates). - Activity measurements (measured coolant activity concentrations, activity release through the chimney, flowrate of water make-up system, released activities). - Reports (describing results of investigations, chronology). - Status of fuel (results of visuals observations). The database items were collected from different sources. Some of them were calculated by Paks NPP and AEKI using VVER-440 and Paks specific input data. The details of the present version of the database including the main information on calculational work will be described in the following

  16. Bubbler condenser related research work. Present situation

    International Nuclear Information System (INIS)

    2001-02-01

    Intensive discussions within the OECD Support Group on 'VVER-440 Bubbler Condenser Containment Research Work' between 1991 and 1994 demonstrated the need for supplementary research work to achieve an adequate level of basic knowledge. In 1994, the European Commission (EC) asked for a specific 'VVER-440/213 Bubble Condenser Qualification Feasibility Study', which was finished early in 1996, confirming the need for additional research in this field. The Feasibility study formed the basis for the Bubble Condenser Experimental Qualification Project (BCEQ) with two separate experimental activities to be executed within the frame of the PHARE/TACIS 2.13/95 project of the European Commission. A first activity served to study the thermal-hydraulic phenomena and the associated structure dynamic interactions. This part of the project was performed at EREC, in Elektrogorsk, Russia. The design of the test facility was based on the prototypical bubbler condenser configuration for the Hungarian Paks nuclear power plant. A second activity addressed the structural integrity of certain components of the bubbler condenser steel structures under DBA-typical conditions. This part of the project was performed at VUEZ, in Levice, Slovak Republic. The design of the components of this facility was based on the structural properties of the Dukovany and/or Bohunice nuclear power plants. A third component of the BCEQ project was specified later asking for analytical studies, which should be supported by a number of small-scale separate effects tests to be performed at SVUSS, in Bechovice, Czech Republic. The main experimental and analytical results of the BCEQ test campaigns have been presented and discussed within the frame of the 4. meeting of the Technical Advisory Committee to the BCEQ (Bubble Condenser Experimental Qualification) Project in Brussels in December 1999 and on occasion of the 11. OECD Support Group Meeting in Berlin in April 2000. The discussions had evidenced several

  17. The MAPK ERK5, but not ERK1/2, inhibits the progression of monocytic phenotype to the functioning macrophage

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Xuening [Department of Pathology and Laboratory Medicine, Rutgers, NJ Medical School, 185 South Orange Ave, Newark, NJ 07103 (United States); Pesakhov, Stella [Department of Clinical Biochemistry and Pharmacology, Faculty of Health Sciences, Ben-Gurion University of the Negev, PO Box 653, 84105 Beer-Sheva (Israel); Harrison, Jonathan S [Department of Medicine, Rutgers, Robert Wood Johnson Medical School, New Brunswick, NJ 08903 (United States); Kafka, Michael; Danilenko, Michael [Department of Clinical Biochemistry and Pharmacology, Faculty of Health Sciences, Ben-Gurion University of the Negev, PO Box 653, 84105 Beer-Sheva (Israel); Studzinski, George P, E-mail: studzins@njms.rutgers.edu [Department of Pathology and Laboratory Medicine, Rutgers, NJ Medical School, 185 South Orange Ave, Newark, NJ 07103 (United States)

    2015-01-01

    Intracellular signaling pathways present targets for pharmacological agents with potential for treatment of neoplastic diseases, with some disease remissions already recorded. However, cellular compensatory mechanisms usually negate the initial success. For instance, attempts to interrupt aberrant signaling downstream of the frequently mutated ras by inhibiting ERK1/2 has shown only limited usefulness for cancer therapy. Here, we examined how ERK5, that overlaps the functions of ERK1/2 in cell proliferation and survival, functions in a manner distinct from ERK1/2 in human AML cells induced to differentiate by 1,25D-dihydroxyvitamin D{sub 3} (1,25D). Using inhibitors of ERK1/2 and of MEK5/ERK5 at concentrations specific for each kinase in HL60 and U937 cells, we observed that selective inhibition of the kinase activity of ERK5, but not of ERK1/2, in the presence of 1,25D resulted in macrophage-like cell morphology and enhancement of phagocytic activity. Importantly, this was associated with increased expression of the macrophage colony stimulating factor receptor (M-CSFR), but was not seen when M-CSFR expression was knocked down. Interestingly, inhibition of ERK1/2 led to activation of ERK5 in these cells. Our results support the hypothesis that ERK5 negatively regulates the expression of M-CSFR, and thus has a restraining function on macrophage differentiation. The addition of pharmacological inhibitors of ERK5 may influence trials of differentiation therapy of AML. - Highlights: • ERK5 has at least some functions in AML cells which are distinct from those of ERK1/2. • ERK5 activity negatively controls the expression of M-CSFR. • ERK5 retards the progression of differentiation from monocyte to functional macrophage.

  18. Measurement of coronary flow response to cold pressor stress in asymptomatic women with cardiovascular risk factors using spiral velocity-encoded cine MRI at 3 Tesla

    International Nuclear Information System (INIS)

    Maroules, Christopher D.; Peshock, Ronald M.; Chang, Alice Y.; Kontak, Andrew; Dimitrov, Ivan; Kotys, Melanie

    2010-01-01

    Background: Coronary sinus (CS) flow in response to a provocative stress has been used as a surrogate measure of coronary flow reserve, and velocity-encoded cine (VEC) magnetic resonance imaging (MRI) is an established technique for measuring CS flow. In this study, the cold pressor test (CPT) was used to measure CS flow response because it elicits an endothelium-dependent coronary vasodilation that may afford greater sensitivity for detecting early changes in coronary endothelial function. Purpose: To investigate the feasibility and reproducibility of CS flow reactivity (CSFR) to CPT using spiral VEC MRI at 3 Tesla in a sample of asymptomatic women with cardiovascular risk factors. Material and Methods: Fourteen asymptomatic women (age 38 years ± 10) with cardiovascular risk factors were studied using 3D spiral VEC MRI of the CS at 3 T. The CPT was utilized as a provocative stress to measure changes in CS flow. CSFR to CPT was calculated from the ratio of CS flow during peak stress to baseline CS flow. Results: CPT induced a significant hemodynamic response as measured by a 45% increase in rate-pressure product (P<0.01). A significant increase in CS volume flow was also observed (baseline, 116 ± 26 ml/min; peak stress, 152 ± 34 ml/min, P=0.01). CSFR to CPT was 1.31 ± 0.20. Test-retest variability of CS volume flow was 5% at baseline and 6% during peak stress. Conclusion: Spiral CS VEC MRI at 3 T is a feasible and reproducible technique for measuring CS flow in asymptomatic women at risk for cardiovascular disease. Significant changes in CSFR to CPT are detectable, without demanding pharmacologic stress

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

    International Nuclear Information System (INIS)

    Arffman, A.; Cazalis, B.

    2010-01-01

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

  20. Early Detection and Identification of Anomalies in Chemical Regime

    International Nuclear Information System (INIS)

    Figedy, Stefan; Smiesko, Ivan

    2011-01-01

    This paper provides a brief information about the basic features of a newly developed diagnostic system for early detection and identification of anomalies incoming in the water chemistry regime of the primary and secondary circuit of VVER-440 reactor. This system, called SACHER (System of Analysis of CHEmical Regime) is being installed within the major modernization project at the NPP-V2 Bohunice in the Slovak Republic. System SACHER has been developed fully in MATLAB environment. The availability of prompt information about the chemical conditions of the primary and secondary circuit is very important to prevent the undue corrosion and deposit build-up. The typical chemical information systems that exist and work at the NPPs give the user values of the measured quantities together with their time trends and other derived values. It is then the experienced user's role to recognize the situation the monitored process is in and make the subsequent decisions and take the measures. The SACHER system, based on the computational intelligence techniques, inserts the elements of intelligence into the overall chemical information system. It has the modular structure with the following most important modules: normality module- its aim is to recognize that the process starts to deviate from the normal one and serves as the early warning to the staff to take the adequate measures, fuzzy identification module- its aim is to identify the anomaly on the basis of a set of fuzzy rules, time-prediction module- its aim is to predict the behavior/trend of selected chemical quantities 8 hours ahead in 15 min step from the moment of request, validation module- its aim is to validate the measured quantities, trend module- this module serves for showing the trends of the acquired quantities

  1. Basic principles of creating a new generation of high- temperature brazing filler alloys

    Science.gov (United States)

    Kalin, B. A.; Suchkov, A. N.

    2016-04-01

    The development of new materials is based on the formation of a structural-phase state providing the desired properties by selecting the base and the complex of alloying elements. The development of amorphous filler alloys for a high-temperature brazing has its own features that are due to the limited life cycle and the production method of brazing filler alloys. The work presents a cycle of analytical and experimental materials science investigations including justification of the composition of a new amorphous filler alloy for brazing the products from zirconium alloys at the temperature of no more than 800 °C and at the unbrazing temperature of permanent joints of more than 1200 °C. The experimental alloys have been used for manufacture of amorphous ribbons by rapid quenching, of which the certification has been made by X-ray investigations and a differential-thermal analysis. These ribbons were used to obtain permanent joints from the spacer grid cells (made from the alloy Zr-1% Nb) of fuel assemblies of the thermal nuclear reactor VVER-440. The brazed samples in the form of a pair of cells have been exposed to corrosion tests in autoclaves in superheated water at a temperature of 350 °C, a pressure of 160 MPa and duration of up to 6,000 h. They have been also exposed to destructive tests using a tensile machine. The experimental results obtained have made it possible to propose and patent a brazing filler alloy of the following composition: Zr-5.5Fe-(2.5-3.5)Be-1Nb-(5-8)Cu-2Sn-0.4Cr-(0.5-1.0)Ge. Its melting point is 780 °C and the recommended brazing temperature is 800°C.

  2. Improving Research Reactor Accident Response Capability at the Hungarian Nuclear Safety Authority

    International Nuclear Information System (INIS)

    Vegh, J.; Gajdos, F.; Horvath, Cs.; Matisz, A.; Nyisztor, D.

    2013-06-01

    The paper describes the design and implementation of an on-line operation monitoring and accident response support system to be used at the CERTA emergency response centre of Hungarian Atomic Energy Authority (HAEA). The monitored facility is the Budapest Research Reactor (BRR), which is a tank-type thermal reactor having 10 MW thermal power. The basic reason for the development of the on-line safety information system is to extend the emergency response capability of the CERTA crisis centre to include emergencies related to BRR, as well. CERTA is operated by HAEA at its Budapest headquarters and the centre already has an on-line system for monitoring the state of the four units of Paks NPP, Hungary. The system is called CERTA VITA and it is able to monitor the four VVER-440/V213 units during their normal operation, and during emergencies (including severe accidents). Ensuring appropriate emergency response capabilities, as well as improving the presently available systems and tools was one of the important recommendations resulting from the analyses following the severe accident at Fukushima. This task is valid not only for the operators of the nuclear facilities but also for the nuclear safety authorities, therefore HAEA decided to launch a project - together with the Centre for Energy Research, the operator of BRR - to establish an on-line information system similar to the CERTA VITA used for monitoring the four units of the Paks NPP. It is believed that by the introduction of this new on-line system the accident response capabilities of HAEA will be further enhanced and the BRR emergencies will be handled at the same professional level as potential emergencies at Paks NPP. (authors)

  3. Application of Moessbauer spectroscopy on corrosion products of NPP

    Energy Technology Data Exchange (ETDEWEB)

    Dekan, J., E-mail: julius.dekan@stuba.sk; Lipka, J.; Slugen, V. [Institute of Nuclear and Physical Engineering, Faculty of Electrical Engineering and Information Technology, SUT (Slovakia)

    2013-04-15

    Steam generator (SG) is generally one of the most important components at all nuclear power plants (NPP) with close impact to safe and long-term operation. Material degradation and corrosion/erosion processes are serious risks for long-term reliable operation. Steam generators of four VVER-440 units at nuclear power plants V-1 and V-2 in Jaslovske Bohunice (Slovakia) were gradually changed by new original 'Bohunice' design in period 1994-1998, in order to improve corrosion resistance of SGs. Corrosion processes before and after these design and material changes in Bohunice secondary circuit were studied using Moessbauer spectroscopy during last 25 years. Innovations in the feed water pipeline design as well as material composition improvements were evaluated positively. Moessbauer spectroscopy studies of phase composition of corrosion products were performed on real specimens scrapped from water pipelines or in form of filters deposits. Newest results in our long-term corrosion study confirm good operational experiences and suitable chemical regimes (reduction environment) which results mostly in creation of magnetite (on the level 70 % or higher) and small portions of hematite, goethite or hydrooxides. Regular observation of corrosion/erosion processes is essential for keeping NPP operation on high safety level. The output from performed material analyses influences the optimisation of operating chemical regimes and it can be used in optimisation of regimes at decontamination and passivation of pipelines or secondary circuit components. It can be concluded that a longer passivation time leads more to magnetite fraction in the corrosion products composition.

  4. Conditional release of materials from decommissioning process into the environment in the form of steel railway tracks

    International Nuclear Information System (INIS)

    Tatransky, Peter; Necas, Vladimir

    2009-01-01

    This work points to the possibilities of conditional release of materials from the process of decommissioning the nuclear unit from the operation. According to the valid legislation, materials which do not meet the condition of direct-unconditional release into the environment, should be modified and processed into the matrix designed for the final disposal in the storing place. However there exists a group of materials which activity is on the borderline of the limit of releasing into the environment and it is possible to release them conditionally. The matter of conditional release is that notable amount of materials, mainly metals, is usually contaminated only by radionuclids with relatively short time of half decay. These materials are suitable to use for a specific industrial purpose where the longtime fixation of shortly living radionuclids is expected in one place. This work deals with the conditional release of metals into the form of steel railway tracks. It describes the (working) groups of workers working with the steel railway tracks and defines in the numbers the critical group and its critical individual. For critical individual it dimensions the amounts of materials, which are possible to be released conditionally from one double-unit of the plant of the type VVER 440 V-230 which operation was ended on the regular basis. According to the calculations in the software VISIPLAN and OMEGA there is defined a number of released steel in such way that the internationally recommended rate of maximal effective dose for critical individual-10 μSv/year [IAEA, 2008. Managing low radioactivity material from the decommissioning of nuclear facility. Technical reports series no. 462] is not extended. In the final part there are compared the estimates of the costs of the decommissioning process with the application of conditional release and without it, which is directly reflected in the amount of saved costs and number of containers for surface disposal.

  5. Test of large-scale specimens and models as applied to NPP equipment materials

    International Nuclear Information System (INIS)

    Timofeev, B.T.; Karzov, G.P.

    1993-01-01

    The paper presents the test results on low-cycle fatigue, crack growth rate and fracture toughness of large-scale specimens and structures, manufactured from steel, widely applied in power engineering industry and used for the production of NPP equipment with VVER-440 and VVER-1000 reactors. The obtained results are compared with available test results of standard specimens and calculation relations, accepted in open-quotes Calculation Norms on Strength.close quotes At the fatigue crack initiation stage the experiments were performed on large-scale specimens of various geometry and configuration, which permitted to define 15X2MFA steel fracture initiation resistance by elastic-plastic deformation of large material volume by homogeneous and inhomogeneous state. Besides the above mentioned specimen tests in the regime of low-cycle loading, the test of models with nozzles were performed and a good correlation of the results on fatigue crack initiation criterium was obtained both with calculated data and standard low-cycle fatigue tests. It was noted that on the Paris part of the fatigue fracture diagram a specimen thickness increase does not influence fatigue crack growth resistance by tests in air both at 20 and 350 degrees C. The estimation of the comparability of the results, obtained on specimens and models was also carried out for this stage of fracture. At the stage of unstable crack growth by static loading the experiments were conducted on specimens of various thickness for 15X2MFA and 15X2NMFA steels and their welded joints, produced by submerged arc welding, in as-produced state (the beginning of service) and after embrittling heat treatment, simulating neutron fluence attack (the end of service). The obtained results give evidence of the possibility of the reliable prediction of structure elements brittle fracture using fracture toughness test results on relatively small standard specimens. 35 refs., 23 figs

  6. Physical model of the nuclear fuel cycle simulation code SITON

    International Nuclear Information System (INIS)

    Brolly, Á.; Halász, M.; Szieberth, M.; Nagy, L.; Fehér, S.

    2017-01-01

    Finding answers to main challenges of nuclear energy, like resource utilisation or waste minimisation, calls for transient fuel cycle modelling. This motivation led to the development of SITON v2.0 a dynamic, discrete facilities/discrete materials and also discrete events fuel cycle simulation code. The physical model of the code includes the most important fuel cycle facilities. Facilities can be connected flexibly; their number is not limited. Material transfer between facilities is tracked by taking into account 52 nuclides. Composition of discharged fuel is determined using burnup tables except for the 2400 MW thermal power design of the Gas-Cooled Fast Reactor (GFR2400). For the GFR2400 the FITXS method is used, which fits one-group microscopic cross-sections as polynomial functions of the fuel composition. This method is accurate and fast enough to be used in fuel cycle simulations. Operation of the fuel cycle, i.e. material requests and transfers, is described by discrete events. In advance of the simulation reactors and plants formulate their requests as events; triggered requests are tracked. After that, the events are simulated, i.e. the requests are fulfilled and composition of the material flow between facilities is calculated. To demonstrate capabilities of SITON v2.0, a hypothetical transient fuel cycle is presented in which a 4-unit VVER-440 reactor park was replaced by one GFR2400 that recycled its own spent fuel. It is found that the GFR2400 can be started if the cooling time of its spent fuel is 2 years. However, if the cooling time is 5 years it needs an additional plutonium feed, which can be covered from the spent fuel of a Generation III light water reactor.

  7. Some reactor properties of the new designed nuclear fuels after neutron irradiation

    International Nuclear Information System (INIS)

    Bajan, M.; Necas, V.

    2013-01-01

    The main goal of this paper was perform the optimisation of the fuel assemblies from the profiling point of view as well as the enrichment of individual rods in such a way that the power peaking factor is steady as possible and also the stock of reactivity for six year fuel cycle. For this reason the limit for maximum fuel rod enrichment was increased to 5.95%. The power in the individual rods is the factor, which can limit the total reactor's power, it is very important to minimise the power peaking factor as possible. At the first the power peaking factor of selected fuel assemblies used in VVER-440 reactor were investigated and from results was based perspective designs which was divided into four parts according to the position of pins with gadolinium burnable absorber and according to the shroudless design. From every part the most perspective fuel assembly was chosen. The results are shown in the Fig. 7. The best result is using the shroudless design. As the second best design is fuel assembly with three gadolinium rods in the middle of the assembly. The power peaking factor unsteadiness is much lower as the reference fuel assembly Gd-2. Also it was demonstrate that the increase of enrichment to 5.95% is perspective, because in several designs the difference in enrichment in individual pins was 1% "2'3"5U. Considering only the present allowed value (max 5%) it would not be possible to reach such good power peaking factor and the reactivity sufficient for 6-years fuel cycle. Profiling optimisation together with modernization of structural changes of assembly was achieved the low power peaking factor unsteadiness in individual pins and higher average enrichment of "2"3"5U. So the optimisation can be summarized as very prosperous and perspective. (authors)

  8. Experimental approach to investigate the dynamics of mixing coolant flow in complex geometry using PIV and PLIF techniques

    Directory of Open Access Journals (Sweden)

    Hutli Ezddin

    2015-01-01

    Full Text Available The aim of this work is to investigate experimentally the increase of mixing phenomenon in a coolant flow in order to improve the heat transfer, the economical operation and the structural integrity of Light Water Reactors-Pressurized Water Reactors (LWRs-PWRs. Thus the parameters related to the heat transfer process in the system will be investigated. Data from a set of experiments, obtained by using high precision measurement techniques, Particle Image Velocimetry and Planar Laser-Induced Fluorescence (PIV and PLIF, respectively are to improve the basic understanding of turbulent mixing phenomenon and to provide data for CFD code validation. The coolant mixing phenomenon in the head part of a fuel assembly which includes spacer grids has been investigated (the fuel simulator has half-length of a VVER 440 reactor fuel. The two-dimensional velocity vector and temperature fields in the area of interest are obtained by PIV and PLIF technique, respectively. The measurements of the turbulent flow in the regular tube channel around the thermocouple proved that there is rotation and asymmetry in the coolant flow caused by the mixing grid and the geometrical asymmetry of the fuel bundle. Both PIV and PLIF results showed that at the level of the core exit thermocouple the coolant is homogeneous. The discrepancies that could exist between the outlet average temperature of the coolant and the temperature at in-core thermocouple were clarified. Results of the applied techniques showed that both of them can be used as good provider for data base and to validate CFD results.

  9. Validation of the VTT's reactor physics code system

    International Nuclear Information System (INIS)

    Tanskanen, A.

    1998-01-01

    At VTT Energy several international reactor physics codes and nuclear data libraries are used in a variety of applications. The codes and libraries are under constant development and every now and then new updated versions are released, which are taken in use as soon as they have been validated at VTT Energy. The primary aim of the validation is to ensure that the code works properly, and that it can be used correctly. Moreover, the applicability of the codes and libraries are studied in order to establish their advantages and weak points. The capability of generating program-specific nuclear data for different reactor physics codes starting from the same evaluated data is sometimes of great benefit. VTT Energy has acquired a nuclear data processing system based on the NJOY-94.105 and TRANSX-2.15 processing codes. The validity of the processing system has been demonstrated by generating pointwise (MCNP) and groupwise (ANISN) temperature-dependent cross section sets for the benchmark calculations of the Doppler coefficient of reactivity. At VTT Energy the KENO-VI three-dimensional Monte Carlo code is used in criticality safety analyses. The KENO-VI code and the 44GROUPNDF5 data library have been validated at VTT Energy against the ZR-6 and LR-0 critical experiments. Burnup Credit refers to the reduction in reactivity of burned nuclear fuel due to the change in composition during irradiation. VTT Energy has participated in the calculational VVER-440 burnup credit benchmark in order to validate criticality safety calculation tools. (orig.)

  10. Status report on severe accident material property measurements

    International Nuclear Information System (INIS)

    Farmer, M.T.; McUmber, L.; Spencer, B.W.; Aeschlimann, R.W.

    1997-06-01

    Measurements of selected material properties of molten reactor core material (corium) were made. The corium used was a mixture of UO 2 , ZrO 2 and Zr, with oxygen content being a parameter to reflect different stages of zirconium oxidation. The mixtures used were representative of typical in-vessel melt sequences. For most measurements, the UO 2 /ZrO 2 mass ratio was 1.51, representative of VVER/440 melt compositions and melt compositions of most US BWRs. Measurements were made of the solidus/liquidus temperatures of corium compositions using a Differential Thermal Analysis technique. Observation of the solubility of unoxidized Zr in the oxide phase was made by metallographic analysis of solidus/liquidus melt samples. The results of laminar flow corium spreading tests in one dimension were used to estimate the viscosity of corium compositions. Measured solidus and liquidus temperatures for compositions representative of Zr oxidation of 30, 50 and 70% were compared with those obtained form a phase diagram provided by Kurchatov Institute. It was found that experimental measurements agreed well with the phase diagram values at 70% oxidation, but the measured solidus temperatures were higher than those on the phase diagram and the measured liquidus temperatures were lower than those on the phase diagram at 30 and 50% oxidation. From a microstructure examination it was determined that there was no global segregation into distinct metal and oxide phases during the cooldown of a sample in which there was initially 70% Zr oxidation. Therefore it is concluded that Zr metal is soluble in the oxide phase under molten conditions. Viscosity estimates were made for compositions representative of Zr oxidation of 30, 50 and 70% by fitting the results of spreading tests to Huppert's equation. It was found that, at a temperature of 2500 C, the viscosity varied by three orders of magnitude over this range of compositions. 10 refs., 39 figs., 16 tabs

  11. Integrated experimental test program on waterhammer pressure pulses and associated structural responses within a feedwater sparger

    Energy Technology Data Exchange (ETDEWEB)

    Nurkkala, P.; Hoikkanen, J. [Imatran Voima Oy, Vantaa (Finland)

    1997-12-31

    This paper describes the methods and systems as utilized in an integrated experimental thermohydraulic/mechanics analysis test program on waterhammer pressure pulses within a revised feedwater sparger of a Loviisa generation VVER-440-type reactor. This program was carried out in two stages: (1) measurements with a strictly limited set of operating parameters at Loviisa NPP, and (2) measurements with the full set of operating parameters on a test article simulating the revised feedwater sparger. The experiments at Loviisa NPS served as an invaluable source of information on the nature of waterhammer pressure pulses and structural responses. These tests thus helped to set the objectives and formulate the concept for series of tests on a test article to study the water hammer phenomena. The heavily instrumented full size test article of a steam generator feedwater sparger was placed within a pressure vessel simulating the steam generator. The feedwater sparger was subjected to the full range of operating parameters which were to result in waterhammer pressure pulse trains of various magnitudes and duration. Two different designs of revised feedwater sparger were investigated (i.e. `grounded` and `with goose neck`). The following objects were to be met within this program: (1) establish the thermohydraulic parameters that facilitate the occurrence of water hammer pressure pulses, (2) provide a database for further analysis of the pressure pulse phenomena, (3) establish location and severity of these water hammer pressure pulses, (4) establish the structural response due to these pressure pulses, (5) provide input data for structural integrity analysis. (orig.). 3 refs.

  12. Radioactive waste management and plutonium recovery within the context of the development of nuclear energy in Russia

    Energy Technology Data Exchange (ETDEWEB)

    Kushnikov, V. [V.G. Khlopin Radium Institute, St. Petersburg (Russian Federation)

    1996-05-01

    The Russian strategy for radioactive waste and plutonium management is based on the concept of the closed fuel cycle that has been adopted in Russia, and, to a great degree, falls under the jurisdiction of the existing Russian nuclear energy structures. From its very beginning, Russian atomic energy policy was based on finding the most effective method of developing the new fuel direction with the maximum possible utilization of the energy potential from the fission of heavy atoms and the achievement of fuel self-sufficiency through the recycling of secondary fuel. Although there can be no doubt about the importance of economic considerations (for the future), concerns for the safety of the environment are currently of the utmost importance. In this context, spent NPP fuel can be viewed as a waste to be buried only if there is persuasive evidence that such an approach is both economically and environmentally sound. The production of I GW of energy per year is accompanied by the accumulation of up to 800-1000 kg of highly radioactive fission products and approximately 250 kg of plutonium. Currently, spent fuel from the VVER 100 and the RBNK reactors contains approximately 25 tons of plutonium. There is an additional 30 tons of fuel-grade plutonium in the form of purified oxide, separated from spent fuels used in VVER440 reactors and other power production facilities, as well as approximately 100 tons of weapons-grade plutonium from dismantled warheads. The spent fuel accumulates significant amounts of small actinoids - neptunium americium, and curium. Science and technology have not yet found technical solutions for safe and secure burial of non-reprocessed spent fuel with such a broad range of products, which are typically highly radioactive and will continue to pose a threat for hundreds of thousands of years.

  13. RETU The Finnish research programme on reactor safety 1995-1998. Final Symposium

    International Nuclear Information System (INIS)

    Vanttola, T.

    1998-01-01

    The Reactor Safety (RETU, 1995-1998) research programme concentrated on search of safe limits for nuclear fuel and the reactor core, accident management methods and risk management of nuclear power plants. The total volume of the programme was 100 person years and funding FIM 58 million. This symposium report summarises the research fields, the objectives and the main results obtained. In the field of operational margins of a nuclear reactor, the behaviour of high burnup nuclear fuel was studied both in normal operation and during power transients. The static and dynamic reactor analysis codes were developed and validated to cope with new fuel designs and complicated three-dimensional reactivity transients. Advanced flow models and numerical solution methods for the dynamics codes were developed and tested. Research on accident management developed and validated calculation methods needed to plan preventive measures and to train the personnel to severe accident mitigation. Efforts were made to reduce uncertainties in phenomena important in severe accidents and to study actions planned for accident management. The programme included experimental work, but also participation in large international tests. The Finnish thermal-hydraulic test facility PACTEL was used extensively for the evaluation of the VVER-440 plant accident behaviour, for the validation of the accident analysis computer codes and for the testing of passive safety system concepts for future plant designs. In risk management probabilistic methods were developed for safety related decision making and for complex event sequences. Effects of maintenance on safety were studied and effective methods for assessment of human reliability and safety critical organisations were searched. To enhance human competencies in control of complex environments, practical tools for training and continuous learning were worked out, and methods to evaluate appropriateness of control room design were developed. (orig)

  14. Definition of the seventh dynamic AER benchmark-WWER-440 pressure vessel coolant mixing by re-connection of an isolated loop

    International Nuclear Information System (INIS)

    Kotsarev, A.; Lizorkin, M.; Petrin, R.

    2010-01-01

    The seventh dynamic benchmark is a continuation of the efforts to validate systematically codes for the estimation of the transient behavior of VVER type nuclear power plants. This benchmark is a continuation of the work in the sixth dynamic benchmark. It is proposed to be simulated the transient - re-connection of an isolated circulating loop with low temperature or low boron concentration in a VVER-440 plant. It is supposed to expand the benchmark to other cases when a different number of loops are in operation leading to different symmetric and asymmetric core boundary conditions. The purposes of the proposed benchmark are: 1) Best-estimate simulations of an transient with a coolant flow mixing in the Reactor Pressure Vessel of WWER-440 plant by re-connection of one coolant loop to the several ones on operation, 2) Performing of code-to-code comparisons. The core is at the end of its first cycle with a power of 1196.25 MWt. The basic additional difference of the 7-seventh benchmark is in the detailed description of the downcomer and bottom part of the reactor vessel that allow describing the effects of coolant mixing in the Reactor Pressure Vessel without any additional conservative assumptions. The burn-up and the power distributions at this reactor state have to be calculated by the participants. The thermohydraulic conditions of the core in the beginning of the transient are specified. Participants self-generated best estimate nuclear data is to be used. The main geometrical parameters of the plant and the characteristics of the control and safety systems are also specified. Use generated input data decks developed for a WWER-440 plant and for the applied codes should be used. The behaviour of the plant should be studied applying coupled system codes, which combine a three-dimensional neutron kinetics description of the core with a pseudo or real 3D thermohydraulics system code. (Authors)

  15. Assessment of reactor vessel integrity (ARVI)

    Energy Technology Data Exchange (ETDEWEB)

    Sehgal, B.R. [Division of Nuclear Power Safety (NPS), Royal Institute of Technology (KTH), Drottning Kristinas Vaeg 33A, 10044 Stockholm (Sweden)]. E-mail: sehgal@ne.kth.se; Karbojian, A. [Division of Nuclear Power Safety (NPS), Royal Institute of Technology (KTH), Drottning Kristinas Vaeg 33A, 10044 Stockholm (Sweden); Giri, A. [Division of Nuclear Power Safety (NPS), Royal Institute of Technology (KTH), Drottning Kristinas Vaeg 33A, 10044 Stockholm (Sweden); Kymaelaeinen, O. [FortumEngNP (Finland); Bonnet, J.M. [CEA (France); Ikkonen, K. [Division of Nuclear Power Safety (NPS), Royal Institute of Technology (KTH), Drottning Kristinas Vaeg 33A, 10044 Stockholm (Sweden); Sairanen, R. [VTT (Finland); Bhandari, S. [FRAMATOME (France); Buerger, M. [USTUTT (Germany); Dienstbier, J. [NRI Rez (Czech Republic); Techy, Z. [VEIKI (Hungary); Theofanous, T. [UCSB (United States)

    2005-02-01

    The assessment of reactor vessel integrity (ARVI) project involved a total of nine organizations from Europe and USA. The work consisted of experiments and analysis development. The modeling activities in the area of structural analyses were focused on the support of EC-FOREVER experiments as well as on the exploitation of the data obtained from those experiments for modeling of creep deformation and the validation of the industry structural codes. Work was also performed for extension of melt natural convection analyses to consideration of stratification, and mixing (in the CFD codes). Other modeling activities were for (1) gap cooling CHF and (2) developing simple models for system code. Finally, the methodology and data was applied for the design of IVMR severe accident management scheme for VVER-440/213 plants. The work was broken up into five packages. They were divided into tasks, which were performed by different partners. The major experimental project continued was EC-FOREVER in which data was obtained on in-vessel melt pool coolability. In previous EC-FOREVER experiments data was obtained on melt pool natural convection and lower head creep failure and rupture. Those results obtained were related to the following issues: (1) multiaxial creep laws for different vessel steels (2) effects of penetrations, and (3) mode and location of lower head failure. The two EC-FOREVER tests reported here are related to (a) the effectiveness of gap cooling and (b) water ingression for in vessel melt coolability. Two other experimental projects were also conducted. One was the COPO experiments, which was concerned with the effects of stratification and metal layer on the thermal loads on the lower head wall during melt pool convection. The second experimental project was conducted at ULPU facility, which provided data and correlations of CHF due to the external cooling of the lower head.

  16. Experimental data on heat flux distribution from a volumetrically heated pool with frozen boundaries

    International Nuclear Information System (INIS)

    Helle, Maria; Kymaelaeinen, Olli; Tuomisto, Harri

    1999-01-01

    The COPO II experiments are confirmatory experiments and a continuation project to the earlier COPO I experiments. As in COPO 1, a molten corium pool on the lower head of a RPV is simulated by a two - dimensional slice of it in linear scale 1:2. The corium is simulated by water-zinc sulfate solution with volumetric Joule heating. The heat flux distribution on the boundaries and the temperature distribution in the pool are measured. The major new feature in COPO II is the cooling arrangement which is based on circulation of liquid nitrogen on the outside of the pool boundaries. The use of liquid nitrogen leads to formation of ice on the inside of boundaries. Two geometrically different versions of the COPO II facility have been constructed: one with a tori-spherical bottom shape, simulating the RPV of a VVER-440 reactor as COPO I, and another one with semicircular bottom simulating a western PWR such as AP600. The modified Rayleigh number in the COPO II experiments corresponds to the one in a prototypic corium pool (∼ 10 15 ). This paper reports results from the COPO II-Lo and COPO II-AP experiments with homogenous pool. Results indicate that the upward heat fluxes are in agreement with the results of the COPO I experiments. Also, as expected, the time averaged upward heat flux profile was relatively flat. On the other hand, the heat fluxes at the side and bottom boundaries of the pool were slightly higher in COPO II-Lo than in COPO I. In COPO II-AP, the average heat transfer coefficients to the curved boundary were higher than predicted by Jahn's and Mayinger's correlation, but slightly lower than in BALI experiments. (authors)

  17. Results of operation and current safety performance of nuclear facilities located in the Russian Federation

    Science.gov (United States)

    Kuznetsov, V. M.; Khvostova, M. S.

    2016-12-01

    After the NPP radiation accidents in Russia and Japan, a safety statu of Russian nuclear power plants causes concern. A repeated life time extension of power unit reactor plants, designed at the dawn of the nuclear power engineering in the Soviet Union, power augmentation of the plants to 104-109%, operation of power units in a daily power mode in the range of 100-70-100%, the use of untypical for NPP remixed nuclear fuel without a careful study of the results of its application (at least after two operating periods of the research nuclear installations), the aging of operating personnel, and many other management actions of the State Corporation "Rosatom", should attract the attention of the Federal Service for Ecological, Technical and Atomic Supervision (RosTekhNadzor), but this doesn't happen. The paper considers safety issues of nuclear power plants operating in the Russian Federation. The authors collected statistical information on violations in NPP operation over the past 25 years, which shows that even after repeated relaxation over this period of time of safety regulation requirements in nuclear industry and highly expensive NPP modernization, the latter have not become more safe, and the statistics confirms this. At a lower utilization factor high-power pressure-tube reactors RBMK-1000, compared to light water reactors VVER-440 and 1000, have a greater number of violations and that after annual overhauls. A number of direct and root causes of NPP mulfunctions is still high and remains stable for decades. The paper reveals bottlenecks in ensuring nuclear and radiation safety of nuclear facilities. Main outstanding issues on the storage of spent nuclear fuel are defined. Information on emissions and discharges of radioactive substances, as well as fullness of storages of solid and liquid radioactive waste, located at the NPP sites are presented. Russian NPPs stress test results are submitted, as well as data on the coming removal from operation of NPP

  18. Seismic assessment and upgrading of nuclear power plants in Eastern Europe

    Energy Technology Data Exchange (ETDEWEB)

    Katona, T; Kostov, M

    1997-03-01

    The basic findings of the seismic re-qualification programmes going on recently at all VVER plants in Eastern Europe can be summarised. The problems of the seismic safety have to be solved taking into account the general concept of the nuclear safety enhancement of the units. There are cases where the system improvements lead to better and more effective solution of the problem than the structural upgrading. The equipment and piping of the primary system have sufficient capacity. The viscous dampers are considered usually for the upgrading. The equipment anchorage especially the electrical and I and C equipment anchorage have to be upgraded. There are general consideration for replacement of the hydraulic snubbers by viscous dampers in the primary circuit of the VVER 440/V230. The considerations are not only because of the better seismic behaviour but mainly because of the better operational performance. There is relatively good seismic instrumentation at the plants considered. The definition of the scram level of the units not designed for an OBE is an essential problem. More effort needed for the definition of this level on the basis of re-evaluation experience of the plant equipment and after the proper definition of post-earthquake activities. The seismic re-evaluation and re-qualification of the VVER units is a general safety issue in Easter European countries. This rather complex problem can be solved adopting the experience, methods and requirements of western countries and taking into account the design features of the VVER units as well as the as built and as it is conditions. (J.P.N.)

  19. Low cycle thermomechanical fatigue of reactor steels: Microstructural and fractographic investigations

    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); Kasl, Josef; Jandova, Dagmar [Výzkumný a zkušební ústav Plzeň s.r.o., Tylova 1581/46, 316 00 Plzen (Czech Republic); Jóni, Bertalan [College of Dunaujvaros, Tancsics 1A, Dunaujvaros H-2400 (Hungary); Eötvös Loránd University, Egyetem tér 1-3, Budapest H-1053 (Hungary); Misják, Fanni [Centre for Energy Research, Institute of Technical Physics and Materials Science, Konkoly-Thege M. 29-33, Budapest H-1121 (Hungary); Trampus, Peter [College of Dunaujvaros, Tancsics 1A, Dunaujvaros H-2400 (Hungary)

    2015-07-29

    The fatigue life of the structural materials 15Ch2MFA (CrMoV-alloyed ferritic steel) and 08Ch18N10T (CrNi-alloyed austenitic steel) of a VVER-440 reactor pressure vessel were investigated under fully reversed total strain controlled low cycle fatigue tests. The measurements were carried out in isothermal conditions at 260 °C and with thermal-mechanical conditions in the range 150–270 °C using a GLEEBLE-3800 servo-hydraulic thermal-mechanical simulator. The low cycle fatigue results were evaluated with the Coffin–Manson law, and the parameters of the Ramberg–Osgood stress–strain relation were investigated. Fracture mechanics behavior was observed using scanning electron microscopic analysis of the crack shapes and fracture surfaces. Crack propagation was assessed in relation to the actual crack size and the loading level. Interrupted fatigue tests were also carried out to investigate the kinetics of the fatigue evolution of the materials. Microstructural evaluation of the samples was performed using light, scanning and transmission electron microscopy as well as X-ray diffraction, and measurement of dislocations was completed using TEM and XRD. The course of dislocation density in relation to cumulative usage factor was similar for both steels. However, the nature and distribution of dislocations were different in the individual steels and this resulted in different mechanical behaviors. The nature of the fracture surfaces of both steels appeared similar despite differences in dislocation arrangement. The distances between striation lines initially increased with increasing crack length and then became saturated. The low cycle fatigue behavior investigated can provide a reference for the remaining life assessment and lifetime extension analysis of nuclear power plant components.

  20. YKAe Research programme on nuclear power plant systems behaviour and operational aspects of safety 1990-1994, Final report

    International Nuclear Information System (INIS)

    Mattila, L.; Vanttola, T.

    1995-04-01

    The research programme on Nuclear Power Plant Systems Behaviour and Operational Aspects of Safety was carried out between 1990 and 1994. In the field of Safe operational margins of nuclear fuel and reactor core, an up-to-date steady-state fuel performance model was validated for higher burn-ups and well-characterized VVER fuel experiments were carried out. A comprehensive reactor analysis code system was extended and validated for complex 3-D phenomena, such as ATWS and boron dilution transients. Advanced hydraulics methods were added to the dynamics codes. Experiments were carried out with PACTEL, the most comprehensive thermal-hydraulic test facility for VVER-440-type reactors worldwide. For example, a series of natural circulation tests were provided for the first VVER-related international standard problem of the OECD/NEA. Advanced foreign computer codes for severe accidents were evaluated and modified for the needs of Finnish power plants. Specific progress was made in modelling the reflooding of an overheated core and in the structural analysis of a pressure vessel under corium load, as well as in experimental and theoretical studies of aerosol and hydrogen behaviour. Fire modelling was improved by implementing advanced calculation methods and by validating them against our own experiments and international test data. Techniques in living probabilistic safety assessment and risk decision-making were refined in pilot applications for continuous monitoring, follow-up and management of risks of an operating power plant. In the area of human reliability and organizational performance, factors important for the continuous development of the competence of control room operator teams and plant maintenance staff were identified. (237 refs., 75 figs., 13 tabs.)

  1. A Few Examples of ISPs Addressing Specific Reactor Safety Problems

    International Nuclear Information System (INIS)

    Reocreux, M.

    2008-01-01

    Four International Standard Problems which were related to safety reactor problems are briefly discussed. ISP-20 (Steam Generator Tube Rupture in DOEL 2) is a unique ISP as it is based on a real incident which occurred in a commercial Power Plant. This ISP clearly illustrated the special problems of an ISP based on a real plant, namely limited access to precise plant data, some lack in the detailed knowledge of sensor behaviour, etc. ISP-26 (ROSA IV-LSTF small break test) was an open ISP. A qualitatively good prediction of the measured events was obtained even if some modelling deficiencies were identified. ISP-27 (BETHSY Exp. 9.1 B) was a blind ISP. All important trends observed during the test were qualitatively calculated by most computer codes. However, some deficiencies in calculating some variables were evident. ISP-33 (PACTEL Natural Circulation) was an exercise with a test facility modelled on the basis of a Russian VVER 440 and with participations from Eastern and Western organisations. ISP-33 was a double-blind exercise. The simulation of some variables caused some problems although they were in principle not too complicated. Post-test calculations demonstrated significant improvements. For all the four ISPs, the influence of the code user was evident and caused some scatter in the results. A specific study was performed in ISP-26 to clarify from where those user effects were coming. The reactor safety problems related to those ISPs are detailed and the specific contribution of the ISPs to bring solutions is discussed.

  2. Principal trends in ensuring safety in nuclear power plant operation in the CSSR

    International Nuclear Information System (INIS)

    Beranek, J.; Kriz, Z.; Kovar, P.; Macoun, J.

    1984-01-01

    At present two reactor units of the VVER-440 type industrial nuclear power plant are in operation in Czechoslovakia and another ten units are planned to be commissioned and put in operation by 1990. The operation of these units is carried out in compliance with licences and regulations issued by the State Regulatory Body for Nuclear Safety, a body established within the framework of the Czechoslovak Atomic Energy Commission. Operational nuclear safety assurance is based primarily on compliance with the basic safety concept as conceived in the plant design and on compliance with the requirements and terms stipulated in the course of the licensing process. On this basis, the State supervisory activity concentrates on the quality assurance of components and installations important for nuclear safety, on the quality of operating personnel and on compliance with limits and conditions for safe operation. The paper presents the main requirements stipulated in Regulation No.5 on quality assurance issued by the Czechoslovak Atomic Energy Commission and shows how the regulation is being applied. The conditions and modes of proving compliance with quality assurance programmes during plant implementation (design, fabrication, assembly, commissioning) and plant operation are described. The qualification prerequisites and capability requirements for selected categories of operating personnel as stipulated in the existing regulations are outlined. The experience accumulated by the regulatory body in preparing, examining and supervising the activity of the personnel is described. Consideration is given to the question of operational management, with the emphasis on compliance with the limits and conditions for safe operation and on the procedures for their alteration and for reporting infringements. (author)

  3. Gamma-Ray Emission Tomography: Modeling and Evaluation of Partial-Defect Testing Capabilities

    International Nuclear Information System (INIS)

    Jacobsson Svard, S.; Jansson, P.; Davour, A.; Grape, S.; White, T.A.; Smith, L.E.; Deshmukh, N.; Wittman, R.S.; Mozin, V.; Trellue, H.

    2015-01-01

    Gamma emission tomography (GET) for spent nuclear fuel verification is the subject for IAEA MSP project JNT1955. In line with IAEA Safeguards R&D plan 2012-2023, the aim of this effort is to ''develop more sensitive and less intrusive alternatives to existing NDA instruments to perform partial defect test on spent fuel assembly prior to transfer to difficult to access storage''. The current viability study constitutes the first phase of three, with evaluation and decision points between each phase. Two verification objectives have been identified; (1) counting of fuel pins in tomographic images without any a priori knowledge of the fuel assembly under study, and (2) quantitative measurements of pinby- pin properties, e.g., burnup, for the detection of anomalies and/or verification of operator-declared data. Previous measurements performed in Sweden and Finland have proven GET highly promising for detecting removed or substituted fuel rods in BWR and VVER-440 fuel assemblies even down to the individual fuel rod level. The current project adds to previous experiences by pursuing a quantitative assessment of the capabilities of GET for partial defect detection, across a broad range of potential IAEA applications, fuel types and fuel parameters. A modelling and performance-evaluation framework has been developed to provide quantitative GET performance predictions, incorporating burn-up and cooling-time calculations, Monte Carlo radiation-transport and detector-response modelling, GET instrument definitions (existing and notional) and tomographic reconstruction algorithms, which use recorded gamma-ray intensities to produce images of the fuel's internal source distribution or conclusive rod-by-rod data. The framework also comprises image-processing algorithms and performance metrics that recognize the inherent tradeoff between the probability of detecting missing pins and the false-alarm rate. Here, the modelling and analysis framework is

  4. The full stories on Armenia and Beloyarsk

    International Nuclear Information System (INIS)

    Aulamo, H.; Marttila, J.; Reponen, H.

    1995-01-01

    Details are presented of the fires which occurred at the Armenia 1 reactor in 1982 and the Beloyarsk 2 reactor in 1978. Armenia 1 is a variant of the VVER 440 (V-230) known as the V-270 which started commercial operation in 1976. The fire started as a consequence of a short circuit in a 6 KV power cable of a large boron make-up pump and the failure of electrical protection. It resulted in the destruction of many power and control cables and several malfunctions leading to a fire in the turbine hall and the start up transformer. Control of the plant was endangered because of smoke in the control room and the total lack of emergency control provisions. Inadequacies in the fire fighting arrangements were revealed. After the fire, emergency controls were installed, cables were given a fire resistant coating, and fire fighting procedures were improved. Beloyarsk 2 is a 200 MWe RBMK operating from 1967. The cause of the fire was a lubrication oil pipe break in the second turbine generator which ignited when it came into contact with hot surfaces. The fire spread rapidly in the vertical cable shafts damaging power and control cables. Control of the plant was endangered, again due to flames and smoke and lack of emergency provisions. Cable damage rendered fire extinguishing systems inactive and a poor communications system delayed the arrival of external fire brigades. The reactor was saved mainly by good luck. Subsequently, cables were coated with fire resistant paste, and fire-fighting procedures, communications system and training were improved. (author)

  5. The response-matrix based AFEN method for the hexagonal geometry

    International Nuclear Information System (INIS)

    Noh, Jae Man; Kim, Keung Koo; Zee, Sung Quun; Joo, Hyung Kook; Cho, Byng Oh; Jeong, Hyung Guk; Cho, Jin Young

    1998-03-01

    The analytic function expansion nodal (AFEN) method, developed to overcome the limitations caused by the transverse integration, has been successfully to predict the neutron behavior in the hexagonal core as well as rectangular core. In the hexagonal node, the transverse leakage resulted from the transverse integration has some singular terms such as delta-function and step-functions near the node center line. In most nodal methods using the transverse integration, the accuracy of nodal method is degraded because the transverse leakage is approximated as a smooth function across the node center line by ignoring singular terms. However, the AFEN method in which there is no transverse leakage term in deriving nodal coupling equations keeps good accuracy for hexagonal node. In this study, the AFEN method which shows excellent accuracy in the hexagonal core analyses is reformulated as a response matrix form. This form of the AFEN method can be implemented easily to nodal codes based on the response matrix method. Therefore, the Coarse Mesh Rebalance (CMR) acceleration technique which is one of main advantages of the response matrix method can be utilized for the AFEN method. The response matrix based AFEN method has been successfully implemented into the MASTER code and its accuracy and computational efficiency were examined by analyzing the two- and three- dimensional benchmark problem of VVER-440. Based on the results, it can be concluded that the newly formulated AFEN method predicts accurately the assembly powers (within 0.2% average error) as well as the effective multiplication factor (within 0.2% average error) as well as the effective multiplication factor (within 20 pcm error). In addition, the CMR acceleration technique is quite efficient in reducing the computation time of the AFEN method by 8 to 10 times. (author). 22 refs., 1 tab., 4 figs

  6. Corrosion of steam generator pipelines analysed using Moessbauer spectroscopy

    International Nuclear Information System (INIS)

    Slugen, V.; Lipka, J.; Toth, I.; Hascik, J.; Hinca, R.; Lehota, M.

    2003-01-01

    The variability of the properties and the composition of the corrosion products of the stainless CrNi and mild steels in dependence on the conditions (temperature, acidity, etc.) is of such range that, in practice, it is impossible to determine the properties of the corrosion products for an actual case from the theoretical data only. Since the decontamination processes for the materials of the water-cooled reactor (VVER-440) secondary circuits are in the progress of development, it is necessary to draw the needed information by the measurement and analysis of the real specimens. The corrosion layers was separated by scraping the rust off the surface and the powder samples were studied by transmission Moessbauer spectroscopy. It should be noted that the gamma spectroscopic measurements give no evidence of the presence of low-energy gamma radiation emitted from the samples. The scraped specimen powder was homogenised (using a 50 μm sieve) and fixed into a special holder. The 57 Co in Rh matrix was used as the radioactive Moessbauer source. Measured spectra were fitted using program NORMOS SITE. According to the results obtained from Moessbauer spectra, it is possible to establish that the main component of secondary circuit's corrosion products is magnetite Fe 3 O 4 . Next components are hematite α-Fe 2 O 3 and hydroxide akagenite β-FeOOH, which is characterised by significant paramagnetic doublet in the middle of spectra. The sextets corresponding to base materials (martensite and austenite steels) were identified in all measured spectra. (author)

  7. Phase Analysis of Corrosion Products from Nuclear Power Plants

    International Nuclear Information System (INIS)

    Lipka, J.; Slugen, V.; Toth, I.; Hascik, J.; Lehota, M.

    2002-01-01

    The variability of the properties and the composition of the corrosion products of the stainless CrNi and mild steels in dependence on the conditions (temperature, acidity, etc.) is of such a range that, in practice, it is impossible to determine the properties of the corrosion products for an actual case from the theoretical data only. Since the decontamination processes for the materials of the water-cooled reactor (VVER-440) secondary circuits are in a process of development, it is necessary to draw the needed information by the measurement and analysis of the real specimens. The corrosion layer was separated by scraping the rust off the surface and the powder samples were studied by transmission Moessbauer spectroscopy. It should be noted that the gamma spectroscopic measurements give no evidence of the presence of low-energy gamma radiation emitted from the samples. The scrapped specimen powder was homogenised (using the 50 μm sieve) and fixed into the special holder. The 57 Co in Rh matrix was used as the radioactive Moessbauer source. Measured spectra were fitted using program NORMOS SITE. According to the results obtained from Moessbauer spectra, it is possible to establish that the main component of secondary circuit's corrosion products is magnetite Fe 3 O 4 . Next components are hematite α-Fe 2 O 3 and hydroxide akagenite β-FeOOH, which is characterised by a significant paramagnetic doublet in the middle of the spectra. The sextets corresponding to base materials (martensite and austenite steels) were identified in all measured spectra.

  8. Acoustic events during fatigue test of structural steels

    Energy Technology Data Exchange (ETDEWEB)

    Por, Gabor; Fekete, Balazs; Csicso, Gabor; Trampus, Peter [College of Dunaujvaros (Hungary)

    2014-11-01

    Acoustic emission sensors were applied recording noises during low cycle fatigue tests in steel materials. The test specimens were machined from the base metal (15H2MFA) and the anticorrosive cladding metal (08H18N10T) of the VVER-440/V-213 (Russian designed PWR) reactor pressure vessel. During the first period, the measurements were carried out with isothermal condition at 260 C on GLEEBLE 3800 servo-hydraulic thermal-mechanical simulator. The tests were run under uniaxial tension-compression loading with total strain control. The programmed waveform was triangular for all the fatigue tests with the frequency of 0.08 Hz. The cyclic loading was started from the compressed side. It was observed that besides rare acoustic emission events regular 10 msec Acoustic Barkhausen Noise (ABN) burst were recorded due to 50Hz AC current drive for heating and maintaining the constant temperature. The amplitude of MABN was higher under pressure than during relaxing and drawing-out by a factor of 2-5. We have carried out also thermo-mechanical fatigue experiment with the same strain-controlled mechanical cycle and simultaneous thermal cycle between 150 C and 270 C. The total number of cycles was terminated, when the force level necessary for the original elongation had been reduced to 75% of its original value. Visual examination showed always some at least surface cracks after stopping the fatigue test. ABN events registered during the beginning cycle exhibited different spectra from the middle and especially from the last cycles before the end of the test, where also double ABN bursts could be recorded. At the end of the test explicit AE events could be found by a new technique. The most interesting result is the possibility to use ABN for testing reactor materials, which could have practical application for fatigue testing.

  9. Integrated experimental test program on waterhammer pressure pulses and associated structural responses within a feedwater sparger

    Energy Technology Data Exchange (ETDEWEB)

    Nurkkala, P; Hoikkanen, J [Imatran Voima Oy, Vantaa (Finland)

    1998-12-31

    This paper describes the methods and systems as utilized in an integrated experimental thermohydraulic/mechanics analysis test program on waterhammer pressure pulses within a revised feedwater sparger of a Loviisa generation VVER-440-type reactor. This program was carried out in two stages: (1) measurements with a strictly limited set of operating parameters at Loviisa NPP, and (2) measurements with the full set of operating parameters on a test article simulating the revised feedwater sparger. The experiments at Loviisa NPS served as an invaluable source of information on the nature of waterhammer pressure pulses and structural responses. These tests thus helped to set the objectives and formulate the concept for series of tests on a test article to study the water hammer phenomena. The heavily instrumented full size test article of a steam generator feedwater sparger was placed within a pressure vessel simulating the steam generator. The feedwater sparger was subjected to the full range of operating parameters which were to result in waterhammer pressure pulse trains of various magnitudes and duration. Two different designs of revised feedwater sparger were investigated (i.e. `grounded` and `with goose neck`). The following objects were to be met within this program: (1) establish the thermohydraulic parameters that facilitate the occurrence of water hammer pressure pulses, (2) provide a database for further analysis of the pressure pulse phenomena, (3) establish location and severity of these water hammer pressure pulses, (4) establish the structural response due to these pressure pulses, (5) provide input data for structural integrity analysis. (orig.). 3 refs.

  10. 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

  11. Seismic assessment and upgrading of nuclear power plants in Eastern Europe

    International Nuclear Information System (INIS)

    Katona, T.; Kostov, M.

    1997-01-01

    The basic findings of the seismic re-qualification programmes going on recently at all VVER plants in Eastern Europe can be summarised. The problems of the seismic safety have to be solved taking into account the general concept of the nuclear safety enhancement of the units. There are cases where the system improvements lead to better and more effective solution of the problem than the structural upgrading. The equipment and piping of the primary system have sufficient capacity. The viscous dampers are considered usually for the upgrading. The equipment anchorage especially the electrical and I and C equipment anchorage have to be upgraded. There are general consideration for replacement of the hydraulic snubbers by viscous dampers in the primary circuit of the VVER 440/V230. The considerations are not only because of the better seismic behaviour but mainly because of the better operational performance. There is relatively good seismic instrumentation at the plants considered. The definition of the scram level of the units not designed for an OBE is an essential problem. More effort needed for the definition of this level on the basis of re-evaluation experience of the plant equipment and after the proper definition of post-earthquake activities. The seismic re-evaluation and re-qualification of the VVER units is a general safety issue in Easter European countries. This rather complex problem can be solved adopting the experience, methods and requirements of western countries and taking into account the design features of the VVER units as well as the as built and as it is conditions. (J.P.N.)

  12. Characterization and evaluation of acid-modified starch of Dioscorea oppositifolia (Chinese yam as a binder in chloroquine phosphate tablets

    Directory of Open Access Journals (Sweden)

    Adenike Okunlola

    2013-12-01

    Full Text Available Chinese yam (Dioscorea oppositifolia starch modified by acid hydrolysis was characterized and compared with native starch as a binder in chloroquine phosphate tablet formulations. The physicochemical and compressional properties (using density measurements and the Heckel and Kawakita equations of modified Chinese yam starch were determined, and its quantitative effects as a binder on the mechanical and release properties of chloroquine phosphate were analyzed using a 2³ full factorial design. The nature (X1, concentration of starch (X2 and packing fraction (X3 were taken as independent variables and the crushing strength-friability ratio (CSFR, disintegration time (DT and dissolution time (t80 as dependent variables. Acid-modified Chinese yam starch showed a marked reduction (p<0.05 in amylose content and viscosity but increased swelling and water-binding properties. The modified starch had a faster onset and greater amount of plastic flow. Changing the binder from native to acid-modified form led to significant increases (p<0.05 in CSFR and DT but a decrease in t80. An increase in binder concentration and packing fraction gave similar results for CSFR and DT only. These results suggest that acid-modified Chinese yam starches may be useful as tablet binders when high bond strength and fast dissolution are required.

  13. [Effect of G-CSF in vitro Stimulation on Distribution of Peripheral Lymphocyte Subsets in the Healthy Persons].

    Science.gov (United States)

    Zhao, Sha-Sha; Fang, Shu; Zhu, Cheng-Ying; Wang, Li-Li; Gao, Chun-Ji

    2018-02-01

    To investigate the effect of granulocyte-colony stimulating factor (G-CSF) in vitro stimulation on the distribution of lymphocyte subset in healthy human. Peripheral blood mononuclear cells (PBMNCs) were collected from 8 healthy volunteers by density gradient centrifugation on Ficoll-Paque TM . In vitro 200 ng/ml G-CSF or 200 ng/ml G-CSF plus 10 µg/ml ConA directly act on PBMNCs, then the colleted cells were cultivated for 3 days. Lymphocyte subsets were stained with the corresponding fluoresce labeled antibodies and detected by flow cytometry. The levels of T cells in G-CSF group and G-CSF+ConA group were both higher than that in the control group (PCSF on T cell subsets indicated that the levels of CD4 + T cells and CD8 + T cells in G-CSF group were both significantly higher than those in control group (PCSF and control group. Compared with the control group, the level of CD4 + T cells, CD8 + T cells and Treg cells in G-CSF+ConA group significantly increased (PCSF receptor (G-CSFR) expression showed that G-CSFR expression on T cells in G-CSF+ConA group dramatically increased, as compared with control group (PCSF stimulation. ConA can enhance the level of T cells and induce G-CSFR expression on T cells.

  14. RADIOACTIVE WASTE MANAGEMENT IN THE USSR: A REVIEW OF UNCLASSIFIED SOURCES, 1963-1990

    Energy Technology Data Exchange (ETDEWEB)

    Bradley, D. J.; Schneider, K. J.

    1990-03-01

    The Soviet Union operates a vast and growing radioactive waste management system. Detailed information on this system is rare and a general overall picture only emerges after a review of a great deal of literature. Poor waste management practices and slow implementation of environmental restoration activities have caused a great deal of national concern. The release of information on the cause and extent of an accident involving high-level waste at the Kyshtym production reactor site in 1957, as well as other contamination at the site, serve to highlight past Soviet waste management practices. As a result, the area of waste management is now receiving greater emphasis, and more public disclosures. Little is known about Soviet waste management practices related to uranium mining, conversion, and fuel fabrication processes. However, releases of radioactive material to the environment from uranium mining and milling operations, such as from mill tailings piles, are causing public concern. Official Soviet policy calls for a closed fuel cycle, with reprocessing of power reactor fuel that has been cooled for five years. For power reactors, only VVER-440 reactor fuel has been reprocessed in any significant amount, and a decision on the disposition of RBMK reactor fuel has been postponed indefinitely. Soviet reprocessing efforts are falling behind schedule; thus longer storage times for spent fuel will be required, primarily at multiple reactor stations. Information on reprocessing in the Soviet Union has been severely limited until 1989, when two reprocessing sites were acknowledged by the Soviets. A 400-metric ton (MT) per year reprocessing facility, located at Kyshtym, has been operational since 1949 for reprocessing production reactor fuel. This facility is reported to have been reprocessing VVER-440 and naval reactor fuel since 1978, with about 2000 MT of VVER-440 fuel being reprocessed by July 1989. A second facility, located near Krasnoyarsk and having a 1500 MT per

  15. Spalled, aerodynamically modified moldavite from Slavice, Moravia, Czechoslovakia

    Science.gov (United States)

    Chao, E.C.T.

    1964-01-01

    A Czechoslovakian tektite or moldavite shows clear, indirect evidence of aerodynamic ablation. This large tektite has the shape of a teardrop, with a strongly convex, deeply corroded, but clearly identifiable front and a planoconvex, relatively smooth, posterior surface. In spite of much erosion and corrosion, demarcation of the posterior and the anterior part of the specimen (the keel) is clearly preserved locally. This specimen provides the first tangible evidence that moldavites entered the atmosphere cold, probably at a velocity exceeding 5 kilometers per second; the result was selective heating of the anterior face and perhaps ablation during the second melting. This provides evidence of the extraterrestial origin of moldavites.

  16. One year after Chernobyl - the world has changed

    International Nuclear Information System (INIS)

    1987-06-01

    The importance of the Chernobyl accidents for the antiatomic movement and nuclear power in Austria and other European countries is outlined. In the same number there several other very short contributions (without authors) whose content is indicated by the headings: The mentality of the proponents (of nuclear power). The callousness of the proponents. The feigned play of the atomic lobby shocks the Austrian public. West Germany on the march to an atomic state. First success against Wackersdorf (fuel reprocessing plant in West Germany). Temelin -the czechoslovakian Chernobyl/Cattenom- on the Austrian border. 5 figs., 1 tab. (qui)

  17. Application of MOS structures to gamma dosimetry

    International Nuclear Information System (INIS)

    Frank, H.

    1978-01-01

    Lattice disorders induced in SiO 2 layers by irradiation are described, and the possibility of using MOS transistors for gamma dosimetry is discussed. Furthermore, experimental results are given for Czechoslovakian MOS transistors of MH 2009 type after gamma irradiation. Reference measurements with other irradiation sources have shown that the transistors respond only to those types of radiation which induce space charges in the oxide layer. They are, therefore, insensitive to neutrons and thus in contrast to dosimetric silicon diodes. Circuitry, sensitivity, and fading of MOS transistors are given, and a physical functional model is compared with the experimental results. (author)

  18. Application of the Combined Cycle LWR-Gas Turbine to PWR for NPP Life Extension Safety Upgrade and Improving Economy

    International Nuclear Information System (INIS)

    Kuznetsov, Yu. N.

    2006-01-01

    further reduce neutron flux to the vessel. This leads to decreasing the steam flow from the SG, but the capacity of the nuclear turbine (IP and LP turbines) remains the same due to the steam produced in the heat recovery steam generator using the waste gas heat. The efficiency of the IP turbine increases due to decreasing moisture content in the turbine flow path, thus increasing the plant total electrical efficiency from 33.5 to 45 percent. The analysis of conceptual design, heat balance, efficiency and economics of VVER-440 with different gas turbine combined topping cycle is presented for illustration. (author)

  19. FP7 Project LONGLIFE: Overview of results and implications

    International Nuclear Information System (INIS)

    Altstadt, Eberhard; Keim, Elisabeth; Hein, Hieronymus; Serrano, Marta; Bergner, Frank; Viehrig, Hans-Werner; Ballesteros, Antonio; Chaouadi, Rachid; Wilford, Keith

    2014-01-01

    Highlights: • Radiation effects in reactor pressure vessel steels under long term operation. • Indications of late blooming effects were found in some cases. • Significant flux effect on the size of defect clusters in high-Cu weld materials. • Guideline for monitoring radiation embrittlement during life extension. - Abstract: LONGLIFE (“Treatment of long term irradiation embrittlement effects in RPV safety assessment”) was a collaborative project of the 7th Framework Programme of EURATOM under the umbrella of NULIFE/NUGENIA, aiming at an improved understanding of irradiation effects in reactor pressure vessel steels under conditions representative of long term operation. The LONGLIFE project was completed by the end of January 2014. The paper gives an overview of the main project results and their implications for future research, as discussed at the final project workshop. The microstructural database for neutron-irradiated RPV steels was extended considerably and existing gaps on mechanical property data were closed. Indications of late blooming effects (LBE) were found in some cases, but clear criteria for the occurrence/exclusion in terms of irradiation conditions and chemical composition have still to be developed. The commonly accepted trend, that low flux and low irradiation temperature promotes LBE, is supported. A significant flux effect on the size of defect clusters was observed in two high Cu weld materials, while the changes of mechanical properties are not affected by the neutron flux. The database requires completion in particular for low-Cu RPV steels. The shift of reference temperature T 0 over the thickness location of a VVER-440 welding seam does not follow the prediction Russian code, because of the strong variation of the intrinsic weld bead structure. Therefore, the effect of the initial microstructure and of the heterogeneity on the radiation behaviour has to be addressed in future works. Existing embrittlement trend curves models

  20. ASTEC application to in-vessel corium retention

    International Nuclear Information System (INIS)

    Tarabelli, D.; Ratel, G.; Pelisson, R.; Guillard, G.; Barnak, M.; Matejovic, P.

    2009-01-01

    This paper summarizes the work done in the SARNET European Network of Excellence on Severe Accidents (6th Framework Programme of the European Commission) on the capability of the ASTEC code to simulate in-vessel corium retention (IVR). This code, jointly developed by the French Institut de Radioprotection et de Surete Nucleaire (IRSN) and the German Gesellschaft fuer Anlagen und Reaktorsicherheit mbH (GRS) for simulation of severe accidents, is now considered as the European reference simulation tool. First, the DIVA module of ASTEC code is briefly introduced. This module treats the core degradation and corium thermal behaviour, when relocated in the reactor lower head. Former ASTEC V1.2 version assumed a predefined stratified molten pool configuration with a metallic layer on the top of the volumetrically heated oxide pool. In order to reflect the results of the MASCA project, improved models that enable modelling of more general corium pool configurations were implemented by the CEA (France) into the DIVA module of the ASTEC V1.3 code. In parallel, the CEA was working on ASTEC modelling of the external reactor vessel cooling (ERVC). The capability of the ASTEC CESAR circuit thermal-hydraulics to simulate the ERVC was tested. The conclusions were that the CESAR module is capable of simulating this system although some numerical and physical instabilities can occur. Developments were then made on the coupling between both DIVA and CESAR modules in close collaboration with IRSN. In specific conditions, code oscillations remain and an analysis was made to reduce the numerical part of these oscillations. A comparison of CESAR results of the SULTAN experiments (CEA) showed an agreement on the pressure differences. The ASTEC V1.2 code version was applied to IVR simulation for VVER-440/V213 reactors assuming defined corium mass, composition and decay heat. The external cooling of reactor wall was simulated by applying imposed coolant temperature and heat transfer

  1. Development and application of Siton, a new fuel cycle simulation code

    International Nuclear Information System (INIS)

    Brolly, Aron; Szieberth, Mate; Halasz, Mate; Nagy, Lajos; Feher, Sandor

    2015-01-01

    As the result of the co-operation between the Centre for Energy Research (EK) and the Institute of Nuclear Techniques (NTI) a new fuel cycle simulation code called SITON was developed. Physical model of the code takes into account six facilities of the nuclear fuel cycle namely material stocks, spent fuel interim storages, plants for uranium enrichment, fuel fabrication, spent fuel reprocessing and reactors. Facilities can be linked in a flexible manner and their number is not limited. Lag time of the facilities and cooling time of the spent fuel, which are the two main parameters to introduce lag time into the fuel cycle, are taken into account. Material transfer between the facilities is modelled in a discrete manner tracking 52 nuclides and their short-lived decay daughters. Composition of the discharged fuel is determined by means of burn-up tables except for the 2400 MWth design of gas cooled fast reactor (GFR2400) which has a separate burn-up module developed at the NTI. To demonstrate the capabilities of SITON introduction of a GFR2400 into the Hungarian reactor park using the legacy spent fuel of the four presently operating VVER-440 units was simulated. 2040 was assumed as the commissioning date of the GFR2400 and recycling of its fuel was started as soon as possible. It was found that the plutonium content of the legacy spent fuel is sufficient to the start-up of only one GFR2400. There is an intermediate period between the commissioning of the reactor and the recycling of its first discharged fuel. Plutonium need of this period can be covered by the legacy spent fuel if the cooling time of the spent GFR2400 fuel is 2 years. If the cooling time is 5 years there will be a lack of plutonium in this period. To counterbalance this lack an EPR was started before the GFR2400 and its spent fuel was accumulated and reprocessed. Cooling time of the spent EPR fuel was also varied. Finally, an EPR only scenario is presented using two EPRs as a reference case

  2. Utilization of the simulators in I and C renewal project of Loviisa NPP

    International Nuclear Information System (INIS)

    Porkholm, K.; Ahonen, A.; Tiihonen, O.

    2006-01-01

    There are two VVER-440 type reactors in Loviisa Nuclear Power Plant. The first unit has been in operation since 1977 and the second since 1980. The availability of the plant as well as the operational experiences of the I and C systems are good. However it is obvious that the lifetime of the original I and C systems is not sufficient to guarantee the good availability of the plant in the future. Due to this fact a project for the renewal of the existing I and C systems has been started at Loviisa Nuclear Power Plant. In the project the analogue I and C systems will be renewed by digital I and C systems in four phases during 2005...2014. Simulators will be utilized extensively in the project to assure that the renewal of I and C systems can be realized safely and economically. An engineering simulator will be used in the design and validation of the modifications of the renewal I and C systems. A development simulator is aimed for the design, testing and acceptance of the new Man Machine Interface. A testing simulator will be used for the testing of the new I and C systems and retuning of the controllers mainly during the Factory Acceptance Tests. A training simulator will be used in training the operators and the other technical personnel in the operation of the new monitor-based control room facilities. All the simulators in the renewal project are based on APROS (Advanced PROcess Simulator) Simulation Software. Fortum Nuclear Services Ltd and the Technical Research Centre of Finland have developed APROS Simulation Software since 1986. APROS is a good example of the real multifunctional simulation software; i.e. it can be used in process and automation design, safety analysis and training simulator applications. APROS has been used extensively for various analysis and simulation tasks of the Loviisa Nuclear Power Plant in the past years. It has also been applied to various nuclear and thermal power plants elsewhere. First a short overview of Loviisa Nuclear Power

  3. 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

  4. 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

  5. Plant modeling as a key tool for nuclear I and C design and V and V

    International Nuclear Information System (INIS)

    Krasnov, V.; Sokolov, O.; Symkin, B.

    2006-01-01

    This paper summarizes an intensive experience of LvivORGRES in the design and implementation of the digital control systems at VVER-1000 and VVER-440 nuclear power plants in Ukraine and Bulgaria. This experience is applicable to the digital I and C upgrade projects for other types of reactor equipment as well as to the design and testing of new I and C systems for new constructions. LvivORGRES was recently involved in several modernization projects as a functional designer and, also, provided technical support and supervision during the factory and site acceptance testing. It is widely accepted and proved by the industry's practice that a level and quality of system validation at all design and implementation phases are key to the successful future operation of I and C systems. The plant control systems have some additional validation requirements in comparing with the information and monitoring systems. According to the Ukrainian nuclear regulation standards, the scope of the control system projects should include the close loop stability analysis at all unit modes of operation. Besides the control system algorithms verification and validation, it was necessary to determine the tuning parameters for the system and use them initially during the system commissioning. LvivORGRES has developed the Adaptive Plant Modeling process that was used as a key tool in all design stages of control system upgrade projects: Software engineering tests, Integrated system validation tests, Factory acceptance tests. The Plant Model was developed on a modular basis which allowed the testing of all primary and secondary side regulators for all unit modes of operation including transients and unit start-up and shutdown. The Plant Model has been adapted to each project's requirements. The use of the plant simulation provided technical bases for important project decisions and documents including among others: system test strategy, initial tuning parameters, training plan, etc. The Plant

  6. Market conditions in Hungary, central europe

    International Nuclear Information System (INIS)

    Nagy, S.

    2000-01-01

    Paks Nuclear Power Plant is the only nuclear power plant in Hungary covering 38% of electricity consumption in the country. The nuclear electricity production of the four VVER440/213 type units in the year of 1999 was 14096 GWh, the second best result in the history of the company. After very detailed safety analyses Paks NPP started a safety upgrading program in 1996, and today the CDF values of the reactor units reached an internationally accepted value. The operational and safety culture and the level of safety was evaluated and reviewed by different international organizations like the IAEA, WENRA, WANO. Based on the conclusions of these international organizations the Paks plant w111 be ready for EU accession after the completion of the ongoing safety upgrading program. Capacity enhancement as a part of the preparation for the market conditions resulted a power upgrade around 20 MW for all four units. This way a relatively small investment in comparison with new installations resulted more efficient to market nuclear capacity. Last year the cost of a generated 1 kWh electricity was 5.98 Ft, which is still to be decreased in 2000. To upgrade the competitiveness of nuclear generation lots of efforts were done in the areas of plant management cost reduction, man-power efficiency upgrade. Un-bundling of activities not directly related to electricity generation in one hand, and more efficient cooperation with other VVER operators and the Hungarian Power Companies Ltd. in certain areas on the other hand are good examples for efficient steps in economic improvement. The company as one of the electricity producers also should follow the capabilities of producers in the neighboring countries like Slovakia and Ukraine, where electricity production is with government subsidiary. To find the right balance between the necessary investments and the market induced cost reduction is one of the most important task in Hungary in Eastern-Europe especially when the nuclear energy

  7. Methodology and measures for preventing unacceptable flow-accelerated corrosion thinning of pipelines and equipment of NPP power generating units

    Science.gov (United States)

    Tomarov, G. V.; Shipkov, A. A.; Lovchev, V. N.; Gutsev, D. F.

    2016-10-01

    Problems of metal flow-accelerated corrosion (FAC) in the pipelines and equipment of the condensate- feeding and wet-steam paths of NPP power-generating units (PGU) are examined. Goals, objectives, and main principles of the methodology for the implementation of an integrated program of AO Concern Rosenergoatom for the prevention of unacceptable FAC thinning and for increasing operational flow-accelerated corrosion resistance of NPP EaP are worded (further the Program). A role is determined and potentialities are shown for the use of Russian software packages in the evaluation and prediction of FAC rate upon solving practical problems for the timely detection of unacceptable FAC thinning in the elements of pipelines and equipment (EaP) of the secondary circuit of NPP PGU. Information is given concerning the structure, properties, and functions of the software systems for plant personnel support in the monitoring and planning of the inservice inspection of FAC thinning elements of pipelines and equipment of the secondary circuit of NPP PGUs, which are created and implemented at some Russian NPPs equipped with VVER-1000, VVER-440, and BN-600 reactors. It is noted that one of the most important practical results of software packages for supporting NPP personnel concerning the issue of flow-accelerated corrosion consists in revealing elements under a hazard of intense local FAC thinning. Examples are given for successful practice at some Russian NPP concerning the use of software systems for supporting the personnel in early detection of secondary-circuit pipeline elements with FAC thinning close to an unacceptable level. Intermediate results of working on the Program are presented and new tasks set in 2012 as a part of the updated program are denoted. The prospects of the developed methods and tools in the scope of the Program measures at the stages of design and construction of NPP PGU are discussed. The main directions of the work on solving the problems of flow

  8. 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

  9. Chemistry monitoring and diagnostic system at NPP Jaslovske Bohunice

    International Nuclear Information System (INIS)

    Smiesko, Ivan; Figedy, Stefan

    2012-09-01

    This paper provides a description of water chemistry monitoring and diagnostic system installed at Slovak NPP Jaslovske Bohunice. System has complex architecture and covers laboratory data, chemistry and radiochemistry on-line monitoring data, process data acquisition and processing and diagnostics. Pre-filtered data from process computer and chemistry on-line monitors are recorded together with laboratory data in the ORACLE-based information system CHEMIS with many presentation and processing features. Brief information is given about the basic features of a newly developed diagnostic system for early detection and identification of anomalies incoming in the water chemistry regime of the primary and secondary circuit of VVER-440 type unit. This system, called SACHER (System of Analysis of Chemical Regime) has been installed within the major modernization project at the NPP Bohunice in the Slovak Republic. System SACHER has been developed fully in MATLAB environment. Diagnostic system works exclusively with available on-line data as an operation personnel support application allowing effective response to adverse chemistry events/trends. The availability of prompt information about the chemical conditions of the primary and secondary circuit is very important in order to prevent the undue corrosion and deposit build-up processes within the plant systems. The typical chemical information systems that exist and work at the NPPs give the user values of the measured quantities together with their time trends and other derived values. It is then the experienced user's role to recognize the situation the monitored process is in and make the subsequent decisions and take the measures. The SACHER system, based on the computational intelligence techniques, inserts the elements of intelligence into the overall chemical information system. It has the modular structure with the following most important modules: - normality module- its aim is to recognize that the process

  10. Chapter No.4. Safety analyses

    International Nuclear Information System (INIS)

    2002-01-01

    for NPP V-1 Bohunice and on review of the impact of the modelling of selected components to the results of calculation safety analysis (a sensitivity study for NPP Mochovce). In 2001 UJD joined a new European project Alternative Approaches to the Safety Performance Indicators. The project is aimed at the information collecting and determining of approaches and recommendations for implementation of the risk oriented indicators, identification of the impact of the safety culture level and organisational culture on safety and applying of indicators to the needs of regulators and operators. In frame of the PHARE project UJD participated in the task focused on severe accident mitigation for nuclear power plants with VVER-440/V213 units. The main results of the analyses of nuclear power plants responses to severe accidents were summarised and the state of their analytical base performed in the past was evaluated within the project. Possible severe accident mitigation and preventative measures were proposed and their applicability for the nuclear power plants with VVER-440/V213 was investigated. The obtained results will be used in assessment activities and accident management of UJD. UJD has been involved also in EVITA project which makes a part of the 5 th EC Framework Programme. The project aims at validation of the European computer code ASTEC dedicated for severe accidents modelling. In 2001 the ASTEC computer code was tested on different platforms. The results of the testing are summarised in the technical report of EC issued in September 2001. Further activities within this project were focused on performing of selected accident scenarios analyses and comparison of the obtained results with the analyses realised with the help of other computer codes. The work on the project will continue in 2002. In 2001 a groundwork on establishing the Centre for Nuclear Safety in Central and Eastern Europe (CENS), the seat of which is going to be in Bratislava, has continued. The

  11. Completion of Mochovce 1 and 2 nuclear power plant project

    International Nuclear Information System (INIS)

    Valach, J.

    2000-01-01

    Commissioning of the Mochovce Nuclear Power Plant unit 2 fully completed the Mochovce unit 1 and 2 project. The power plant construction passed through a complicated evolution since early 80's. The genesis started with the change of the plant location due to a seismic resistance requirement raised by the former Czechoslovak Atomic Energy Commission, change of the design and contractor of the Instrumentation and Control system in early 90's, halt of construction in 1993 till 1996 due to a lack of funding, as well as a series of international reviews of safety aspects. All the obstacles had been overcome in the end, and completion works were re-started in 1996. Looking back to the unit 1 and 2 completion project, I would like to highlight some aspects of the project, particularly the level of safety, technical and organisational conditions, and terms of funding. The most important argument for completion of the plant was that VVER 440/213 reactor units were distinguished with very high level of passive safety, even overcoming Western-type PWR's in some aspects. The replacement of the original I and C system was another important contribution to the initial level of safety. International safety reviews performed at Mochovce NPP between 1990 and 1995 (Siemens, EdF, IAEA, RlSKAUDIT) confirmed the high level of safety of VVER 440/213 reactors, however, pointed out at some safety issues which had not been dealt with in the design. Those included e.g. the 'external risks', primary-to-secondary leaks, qualification of pressurizer safety valves, and others. It is necessary to stress that the Safety Improvement Programme defined for Mochovce NPP to comply with international standards did not become the most important part of the completion project due to external pressures, but based on natural requirements on nuclear power plant safety level generally accepted in 90's. The good example could be the seismic resistance of the plant which, though designed as seismic resistant

  12. Health effects of radon in air

    International Nuclear Information System (INIS)

    Cohen, B.L.

    1988-01-01

    Widely accepted risk estimates for exposure to radon in homes are derived largely from studies of miners. These include large groups of US Czechoslovakian, and Canadian uranium miners, Newfoundland fluorspar miners, and Swedish iron, lead, and zinc miners, all of which give roughly consistent results, with the excess risk of lung cancer increasing linearly with the exposure to radon. The authors have studied correlations between average radon levels and lung cancer rates in counties of the US. One study based on 50,000 purchased measurements in the main living areas of houses in which there have been no previous measurements involves 310 counties. It gives a weak but statistically significant negative correlation between mean radon levels and lung cancer rates for both females and males, whereas the usual risk estimates predict a large positive correlation

  13. Cestování Čechů na jihoslovanské pobřeží ve 20. letech 20. století

    Directory of Open Access Journals (Sweden)

    Jiří Šoukal

    2014-12-01

    Full Text Available This article centres around the Czechoslovakian perception of holiday travel to Yugoslavia in the 1920s with particular attention to the typology of Czech tourists. It has been shown that travel to Yugoslavia was very popular among the middle classes who had enough time and money. The wealthier classes preferred France. The main selling points travel agents and hotel owners used to promote travel to Yugoslavia were affordability, service targeted to Czechs and Pan-slavism. The idea of a mutual Slavonic tradition had been in existence since the 19th century. Evidence would seem to show that the most significant factor for repeat travel was affordability. Conservative Czech tourists remained loyal guests of Yugoslavia during the 1920s and 1930s.

  14. Forest decline research in Eastern Central Europe and Bavaria

    International Nuclear Information System (INIS)

    Reuther, M.; Kirchner, M.; Kirchinger, E.; Reiter, H.; Roesel, K.; Pfeifer, U.

    1991-07-01

    In 26 conference contributions, the condition of the forest in eastern central Europe (new Federal German laender, CSFR, Hungary, Romania, Bulgaria, Estonia, Poland) and in Bavaria and Austria is described. The methodics of the countries' comprehensive monitoring and ecosystem analyzes in selected sites with their results are presented, mostly for the 80s. Possibilities and advantages of the modelling of forest ecosystems by computer are indicated as well as the gain of knowledge from extensive screening. For some regions, especially the Sudeten, maps showing the spatial distribution of airborne pollutants are presented. Pollutant concentrations are, in part, related to emittors. In almost all cases, indirect effects of acidic gaseous pollutants via changes in soil chemism are blamed for tree disease jointly with other factors or their outcome (silvicultural mistakes, drought, insect infestation). A striking fact is that in Hungary and Romania, unlike other European countries, oak-trees not conifers are most seriously affected. (UWA) [de

  15. Modification of T cell responses by stem cell mobilization requires direct signaling of the T cell by G-CSF and IL-10

    DEFF Research Database (Denmark)

    MacDonald, Kelli P.A.; Le Texier, Laetitia; Zhang, Ping

    2014-01-01

    The majority of allogeneic stem cell transplants are currently undertaken using G-CSF mobilized peripheral blood stem cells. G-CSF has diverse biological effects on a broad range of cells and IL-10 is a key regulator of many of these effects. Using mixed radiation chimeras in which...... the hematopoietic or nonhematopoietic compartments were wild-type, IL-10(-/-), G-CSFR(-/-), or combinations thereof we demonstrated that the attenuation of alloreactive T cell responses after G-CSF mobilization required direct signaling of the T cell by both G-CSF and IL-10. IL-10 was generated principally by radio......-resistant tissue, and was not required to be produced by T cells. G-CSF mobilization significantly modulated the transcription profile of CD4(+)CD25(+) regulatory T cells, promoted their expansion in the donor and recipient and their depletion significantly increased graft-versus-host disease (GVHD). In contrast...

  16. Primary system hydraulic characteristics after modification of reactor coolant pumps' impeller wheels at Bohunice NPP executed in 2012 and 2013

    International Nuclear Information System (INIS)

    Hermansky, Jozef; Zavodsky, Martin

    2014-01-01

    A coolant flow through the reactor is usually determined after annual outages at Slovak NPP (VVER 440) in two distinct ways. First method is determination on the basis of the secondary system parameters - measurement of thermal balances. The value achieved by this method is used as the input parameter in the Table of allowed reactor operation modes. The second method draws from the primary system parameters - measurement of primary system hydraulic characteristics. Flow nozzles used for the measurement of feed water flow behind high pressure heaters were replaced at both Bohunice Units during outages in 2008. The feed water flow behind high pressure heaters is one of the main parameters used for the determination of coolant flow through the reactor by the first method. Compared to the measurement executed during previous fuel cycles, the calculated coolant flow through the reactor decreased considerably after the change of flow nozzles. The imaginary change of coolant flow through the reactor at Unit 3 was -1,6 %; and at Unit 4 -2,6 %. This change was not proved by the parallel measurement of primary system hydraulic characteristics. Later it was found out that the original flow nozzles used for 25 years were substantially deposited (original inner diameter of the nozzles was reduced by about 0,6-0,9 mm). Therefore feed water flow measurement was untrustworthy within the recent years. On the findings stated above, Bohunice NPP has decided to increase coolant flow through the reactor by changing the reactor coolant pumps impeller wheels. The modification of impellers wheels is planned within years 2012 to 2014. During the outages in 2013 two impeller wheels were replaced at both units. Nowadays Unit 4 is operated with all 6 new impeller wheels and Unit 3 with four new impeller wheels. Modification of last two impeller wheels at Unit 3 will be performed during the outage in 2014. On account of impeller wheels modification, non-standard measurement of PS hydraulic

  17. Uncertainty and sensitivity analysis for the modeling of transients with interaction of thermal hydraulics and neutron kinetics

    International Nuclear Information System (INIS)

    Soeren Kliem; Siegfried Mittag; Siegfried Langenbuch

    2005-01-01

    Full text of publication follows: The transition from the application of conservative models to the use of best-estimate models raises the question about the uncertainty of the obtained results. This question becomes especially important, if the best-estimate models should be used for safety analyses in the field of nuclear engineering. Different methodologies were developed to assess the uncertainty of the calculation results of computer simulation codes. One of them is the methodology developed by Gesellschaft fuer Anlagenund Reaktorsicherheit (GRS) which uses the statistical code package SUSA. In the past, this methodology was applied to the calculation results of the advanced thermal hydraulic system code ATHLET. In the frame of the recently finished EU FP5 funded research project VALCO, that methodology was extended and successfully applied to different coupled code systems, including the uncertainty analysis for neutronics. These code systems consist of a thermal hydraulic system code and a 3D neutron kinetic core model. One of the code systems applied was ATHLET coupled with the Rossendorf kinetics code DYN3D. Two real transients at NPPs with VVER-type reactors documented within the VALCO project were selected for analyses. One was the load drop of one of two turbines to house load level at the Loviisa-1 NPP (VVER-440), the second was a test with the switching-off of one of two main feed water pumps at the VVER-1000 Balakovo-4 NPP. The current paper is dedicated to the different steps of the use and implementation of the GRS methodology to coupled code systems and to the assessment of the results obtained by the DYN3D/ATHLET code. Based on the relevant physical processes in both transients, lists of possible sources of uncertainties were compiled. They are specific for the two transients. Besides control parameters like control rod movement and thermal hydraulic parameters like secondary side pressure, mass flow rates, pressurizer sprayer and heater

  18. Final Treatment Center Project for Liquid and Wet Radioactive Waste in Slovakia

    International Nuclear Information System (INIS)

    Kravarik, K.; Stubna, M.; Pekar, A.; Krajc, T.; Zatkulak, M.; Holicka, Z.; Slezak, M.

    2006-01-01

    The Final Treatment Center (FTC) for Mochovce nuclear power plant (NPP) is designed for treatment and final conditioning of radioactive liquid and wet waste produced from plant operation. Mochovce NNP uses a Russian VVER-440 type reactor. Treated wastes comprise radioactive concentrates, spent resin and sludge. VUJE Inc. as an experienced company in field of treatment of radioactive waste in Slovakia has been chosen as main contractor for technological part of FTC. This paper describes the capacity, flow chart, overall waste flow and parameters of the main components in the FTC. The initial project was submitted for approval to the Slovak Electric plc. in 2003. The design and manufacture of main components were performed in 2004 and 2005. FTC construction work started early in 2004. Initial non-radioactive testing of the system is planned for summer 2006 and then radioactive tests are to be followed. A one-year trial operation of facility is planned for completion in 2007. SE - VYZ will be operates the FTC during trial operation and after its completion. SE - VYZ is subsidiary company of Slovak Electric plc. and it is responsible for treatment with radioactive waste and spent fuel in the Slovak republic. SE - VYZ has, besides of other significant experience with operation of Jaslovske Bohunice Treatment Centre. The overall capacity of the FTC is 870 m 3 /year of concentrates and 40 m 3 /year of spent resin and sludge. Bituminization and cementation were provided as main technologies for treatment of these wastes. Treatment of concentrate is performed by bituminization. Concentrate and bitumen are metered into a thin film evaporator with rotating wiping blades. Surplus water is evaporated and concentrate salts are embedded in bitumen. Bitumen product is discharged into 200 l steel drums. Spent resin and sludge are decanted, dried and mixed with bitumen. These mixtures are also discharged into 200 l steel drums. Drums are moved along bituminization line on a roller

  19. Overview on technological and operational aspects of PLIM

    International Nuclear Information System (INIS)

    Nevander, Olli; Hytoenen, Yrjoe; Raitanen, Raimo

    2002-01-01

    Full text: Loviisa NPP with two VVER 440 units is owned and operated by Fortum. In year 1997 after the power uprating and relicensing project a comprehensive life management strategy with a target lifetime of 50 years was selected as a key issue for the remaining lifetime. Two licensed fuel suppliers as well as the successful power uprating assist with long-term economy of production. The plant has recently established a life management programme (LMP), which should improve and unify the present programmes and separate ageing projects. The scope of the new LMP of the Loviisa NPP comprises the scientific and analytic research of the ageing phenomena as well as routine repairs and preventive maintenance at the site. On one hand, the work emphasises practical repair and replacement of active components, and, on the other hand, the structural and safety review of the passive nonreplaceable main components as well. The goal of the development is that all maintenance and inspection activities should support plant life management. In general, an effective approach of ageing management of systems, structures and components (SSC) is a mixture of four elements: careful operation and maintenance, replacement strategy, technical modifications of SSCs and mitigation of analysed ageing effects. Therefore, at Loviisa the organisation and functions of LMP are included as far as possible in the normal operating organisation. Thus the practical LMP program includes effective links from operating experiences to the ageing database and also to the long-term decision making. The difficulty in the implementation of the LMP is to apply an adequate and balanced approach through all areas and important components in long-term. Therefore, such assessment tools as cost-benefit analysis and classification of components according to their impact on safety and operation is needed. During the past few years, the most important ageing attributes of main passive components and safety systems of

  20. 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

  1. C30 Support Plate for Replacing Function of Service Pool 1 at Unit 2

    International Nuclear Information System (INIS)

    Zsoldos, F.

    2006-01-01

    the refuelling work. When filling into the spent fuel pool all fuel assemblies we needed to set up the new core, we can start to put them into the reactor, of course the usage of C30 support plate is ready at this moment, it was needed just to carry out spent fuel and fill into the spent fuel pool the fresh fuel. The plant made many of stress calculations, lot of scientific institutions were involved to the problem solution from Hungary and abroad also (from abroad mainly SKODA, the manufacturer of the reactor). The period from finding out and getting permission to manufacture it and to use it at Unit 2 took more than 1 year. A lot of test and probing was done before the usage of this C30 support plate. All in all in 2005, January we used successfully this C30 support plate. This support plate can be used at any VVER 440 reactors, it should be adjusted to the container used for fresh or spent fuel carrying. (author)

  2. The use of the acoustic emission for the components of the primary circuit of the nuclear power plants

    International Nuclear Information System (INIS)

    Svoboda, V.

    1992-01-01

    Full text: The Modrany Engineering Works (Modranske strojirny) is a producer and a final supplier of the main connecting piping circuit systems and valves for the nuclear power plants (type VVER 440 and VVER 1000) built in Czechoslovakia. Besides the delivery and assembly of valves and components methods there were developed for a monitoring of the stated equipment ability of a service in the Material and Diagnostic Laboratory, which is a part of the company. An important object of this work is to obtain a sufficient set of data and to work out suitable methods, on the basis of which it would be possible to perform a serious estimation of residual service life of the main piping components after certain service operation of the nuclear power plant. During the operation of a nuclear power station a failure of the main piping circuit could happen in either of two possible modes: 1.) A sudden break - by an unstable defect propagation leading to a. final fracture of the piping; 2) A fatigue failure - which is characterised by a gradual subcritical growth of defect in relation to the loading parameters. This process is frequently accelerated by further processes, e.g. corrosion. It is therefore suitable to use such physical and mechanical quantities, which characterize the material damage. Acoustic emission signals belongs to these quantities. A knowledge of the response of these signals in relation to the damage of the material gives us the possibility to evaluate the residual life of the piping containing defects. The importance of this is increasing mainly after a long period of service. She paper deals in details with experience gained in application of acoustic emission, during pressure tests of primary circuit components (elbow, welds, T- junction etc) in laboratory conditions which imitate those in service. There are shown some results of cyclic fatigue tests by internal pressure on prototypes models and specimen. Acoustic emission method represents the

  3. Statement of participants at the International Conference on Can Slovakia secure energy supply and sustainable development without nuclear?

    International Nuclear Information System (INIS)

    Mikus, T.; Suchomel, J.

    2004-01-01

    completion as possible. The Slovak participants at the Conference stated with deep concern that the commitment of the Slovak government to close V1 Bohunice, accepted during EU pre-entry negotiations and reminding e nergy imperialism , as warned by the former Finnish premier Paavo Lipponen, is not fair as it is based on a political appraisal from G7 summit in Munich in 1992 that VVER-440/V230 reactors cannot be upgraded with reasonable costs, which had been disproved by the Slovak evidence. The participants called on the Slovak government and the future Slovak members of the European Parliament to revive negotiations on a revision of this groundless commitment. The participants called on the European nuclear community to support the Slovak demand to revise the commitment to close the two V1 Bohunice units and to complete the construction of the Mochovce units 3 and 4. So the major message from the Conference is: Go Nuke Slovakia!

  4. 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

  5. Progress and perspectives of ASTEC applications in the European Network SARNET

    International Nuclear Information System (INIS)

    Van Dorsselaere, J.P.; Allelein, H.J.; Neu, K.

    2006-01-01

    The ASTEC integral code is jointly developed by IRSN (France) and GRS (Germany) for LWR source term Severe Accident (SA) evaluation, PSA level 2 studies and SA management evaluation. ASTEC constitutes now the reference European integral code through its role in the Network SARNET (Severe Accident Network of Excellence) in the EC 6. Framework Program. The models of next version V1.3, released before end of 2006, represent the current State of the Art, its validation is very extensive (in particular on Phebus FP) and, after next implementation of a model for reflooding of degraded cores, it will cover all needs for SA evaluation in PWR and VVER. It will be the reference code for the IRSN PSA level 2 (Probabilistic Safety Analysis) on French PWR 1300 MWe that starts in 2006. In the frame of SARNET, IRSN coordinates the ASTEC Topic gathering 30 partners that assess the code through validation against experiments and benchmarks with reference codes like CATHARE or RELAP5 for the reactor coolant circuit and COCOSYS for the containment. Plant application calculations are compared with MELCOR and MAAP4 results for a series of different SA sequences. Besides, the knowledge generated by SARNET Topics (Corium, Source Term and Containment) will be progressively integrated into the code through improved or new models. The 2. Users' Club organized at Aix-en-Provence in June 06, with 45 participants from 27 organizations, allowed fruitful discussions with the Maintenance Team. After 2 years of work, code validation shows good overall results, often close to results of reference codes. Some results reach the limits of present knowledge, for instance on Molten-Corium-Concrete-Interaction (MCCI) and Direct Containment Heating (DCH). Benchmarks on plant applications have been performed on diverse reactor types: PWR 900, Konvoi 1300, Westinghouse 1000, VVER-1000 and VVER-440. The main trends of results are similar to MELCOR or MAAP4 results. The objective of the quantitative

  6. Results of the Czech-Austrian calculations of BDBA radiological consequences

    International Nuclear Information System (INIS)

    Carny, P.; Hohenberg, J.-K.

    2003-01-01

    Full text: Common Czech - Austrian comparisons of codes and calculations of BDBA radiological consequences have been performed. Background of these comparisons is described in the paper presented at this symposium. Results of single steps are summarized and discussed in this poster presentation. From the Czech side calculations have been performed with computer codes PC Cosyma, este, RTARC, HAVAR, HERALD, PTM, RODOS/MATCH and long range code MEDIA used by the Czech meteorological institute (CHMI). Code PC Cosyma is taken as main comparable code in this inter-comparisons as it is used by the Czech and the Austrian side. For every accident scenario and for deterministic as well as probabilistic assessment of accident consequences results of both sides have been practically identical. Computer code 'este' is instrument for projection of release and evaluation of real release under real VVER 440 and WER 1000 emergency conditions. The code can be operated with real radiological, meteorological and technological data from the plant. The code calculates projection of avertable doses and simulates movement of radioactive clouds in the vicinity (up to 40-50 km) of the plant. The code participates in these comparisons as it serves as a support instrument for the staff at the emergency centre of the Czech nuclear regulatory body. Code RTARC (Real Time Accident Release Consequences) serves as an instrument for evaluation of radiation situation in the vicinity of the plant (up to 40 km) during the early phase of an accident. The code participates in these comparisons as it was used in the process of the Czech nuclear power plants protective action planning zone determination. Codes HERALD and HAVAR have been used by Skoda and Energoprojekt for analyses of consequences of design bases accidents in Temelin safety report. They were compared with PC Cosyma in one step of these common calculations by the Czech side. The code HAVAR enables to calculate ingestion doses, too, and

  7. Development and application of the ultrasonic technologies in nuclear engineering

    International Nuclear Information System (INIS)

    Lebedev, Nikolay; Krasilnikov, Dmitry; Vasiliev, Albert; Dubinin, Gennady; Yurmanov, Viktor

    2012-09-01

    Efficiency of some traditional chemical technologies in different areas could be significantly increased by adding ultrasonic treatment. For example, ultrasonic treatment was found to improve make-up water systems, decontamination procedures, etc. Improvement of traditional chemical technologies with implementation of ultrasonic treatment has allowed to significantly reducing water waste, including harmful species and radioactive products. The report shows the examples of the recent ultrasonic technology development and application in Russian nuclear engineering. They are as follows: - Preliminary cleaning of surfaces of in-pile parts (e.g. control sensors) prior to their assemblage and welding - Decontamination of grounds and metal surfaces of components with a complex structure -Decrease in sliding friction between fuel rods and grids during VVER reactor fuel assembly manufacturing -Removal of deposits from reactor fuel surfaces in VVER-440s -Increasing the density and strength of pressed sintered items while making fuel pellets and fuel elements, especially mixed-oxide fuel Surface cleanness is very important for the fuel assembly manufacturing, especially prior to welding. An ultrasonic technology for surface cleaning (from graphite and other lubricants, oxides etc.) was developed and implemented. The ultrasonic cleaning is applicable to the parts having both simple shape and different holes. Ultrasonic technology has allowed to improve the surface quality and environmental safety. Ultrasonic treatment appears to be expedient to intensify the chemical decontamination of solid radioactive waste from grounds of different fractions to metallic components. Ultrasonic treatment reduces the decontamination process duration up to 100 times as much. Excellent decontamination factor was received even for the ground fractions below 1 mm. It should be noted that alternative decontamination techniques (e.g. hydraulic separation) are poorly applicable for such ground

  8. Macrophage colony-stimulating factor induces prolactin expression in rat pituitary gland.

    Science.gov (United States)

    Hoshino, Satoya; Kurotani, Reiko; Miyano, Yuki; Sakahara, Satoshi; Koike, Kanako; Maruyama, Minoru; Ishikawa, Fumio; Sakatai, Ichiro; Abe, Hiroyuki; Sakai, Takafumi

    2014-06-01

    We investigated the role of macrophage colony-stimulating factor (M-CSF) in the pituitary gland to understand the effect of M-CSF on pituitary hormones and the relationship between the endocrine and immune systems. When we attempted to establish pituitary cell lines from a thyrotropic pituitary tumor (TtT), a macrophage cell line, TtT/M-87, was established. We evaluated M-CSF-like activity in conditioned media (CM) from seven pituitary cell lines using TtT/M-87 cells. TtT/M-87 proliferation significantly increased in the presence of CM from TtT/GF cells, a pituitary folliculostellate (FS) cell line. M-CSF mRNA was detected in TtT/GF and MtT/E cells by reverse transcriptase-polymerase chain reaction (RT-PCR), and its expression in TtT/GF cells was increased in a lipopolysaccharide (LPS) dose-dependent manner. M-CSF mRNA expression was also increased in rat anterior pituitary glands by LPS. M-CSF receptor (M-CSFR) mRNA was only detected in TtT/ M-87 cells and increased in the LPS-stimulated rat pituitary glands. In rat pituitary glands, M-CSF and M-CSFR were found to be localized in FS cells and prolactin (PRL)-secreting cells, respectively, by immunohistochemistry. The PRL concentration in rat sera was significantly increased at 24 h after M-CSF administration, and mRNA levels significantly increased in primary culture cells of rat anterior pituitary glands. In addition, TNF-α mRNA was increased in the primary culture cells by M-CSF. These results revealed that M-CSF was secreted from FS cells and M-CSF regulated PRL expression in rat pituitary glands.

  9. Granulocyte colony-stimulating factor decreases the Th1/Th2 ratio in peripheral blood mononuclear cells from patients with chronic immune thrombocytopenic purpura in vitro.

    Science.gov (United States)

    Ge, Fei; Zhang, Zhuo; Hou, Jinxiao; Cao, Fenglin; Zhang, Yingmei; Wang, Ping; Wei, Hong; Zhou, Jin

    2016-12-01

    Chronic immune thrombocytopenia purpura (ITP) is an autoimmune disease that exhibits an abnormally high Th1/Th2 ratio. Granulocyte colony-stimulating factor (G-CSF) has been shown to decrease the Th1/Th2 ratio in healthy donors. In this study, we investigated the effects of G-CSF treatment on the Th1/Th2 cells and the underlying mechanisms in patients with ITP in vitro. Peripheral blood mononuclear cells (PBMCs) isolated from patients with ITP and healthy controls were treated with G-CSF. Expression levels of interferon (IFN)-γ, interleukin (IL)-2, IL-4, and IL-13 in supernatants were measured by enzyme-linked immunosorbent assays. The expression of IFN-γ, IL-4, and G-CSF receptor (G-CSFR) on Th1 and Th2 cells were examined by flow cytometry and confocal microscopy. The mRNA expression of IFN-γ, IL-2, IL-4, IL-13, and T-box expressed in T cells (T-bet) and GATA-binding protein 3 (GATA-3) in PBMCs was evaluated by reverse transcription polymerase chain reaction. The results showed that G-CSF could significantly reduce the Th1/Th2 ratio in PBMCs from patients with ITP in vitro. As the concentration of G-CSF increased, Th1/Th2 ([IFN-γ+IL-2]/[IL-4+IL-13]) cytokine ratios and T-bet/GATA-3 mRNA ratios decreased in a concentration-dependent manner. Th1 cells and Th2 cells both expressed G-CSFR. These results suggest that G-CSF could decrease the Th1/Th2 ratio in the context of ITP, and elucidate the direct and indirect immunomodulatory mechanisms underlying G-CSF functions in Th1/Th2 cells, thus supporting the therapeutic potential of G-CSF in the treatment of patients with ITP. Copyright © 2016 Elsevier Ltd. All rights reserved.

  10. Granulocyte macrophage colony stimulating factor (GM-CSF biological actions on human dermal fibroblasts

    Directory of Open Access Journals (Sweden)

    S Montagnani

    2009-12-01

    Full Text Available Fibroblasts are involved in all pathologies characterized by increased ExtraCellularMatrix synthesis, from wound healing to fibrosis. Granulocyte Macrophage-Colony Stimulating Factor (GM-CSF is a cytokine isolated as an hemopoietic growth factor but recently indicated as a differentiative agent on endothelial cells. In this work we demonstrated the expression of the receptor for GM-CSF (GMCSFR on human normal skin fibroblasts from healthy subjects (NFPC and on a human normal fibroblast cell line (NHDF and we try to investigate the biological effects of this cytokine. Human normal fibroblasts were cultured with different doses of GM-CSF to study the effects of this factor on GMCSFR expression, on cell proliferation and adhesion structures. In addition we studied the production of some Extra-Cellular Matrix (ECM components such as Fibronectin, Tenascin and Collagen I. The growth rate of fibroblasts from healthy donors (NFPC is not augmented by GM-CSF stimulation in spite of increased expression of the GM-CSFR. On the contrary, the proliferation of normal human dermal fibroblasts (NHDF cell line seems more influenced by high concentration of GM-CSF in the culture medium. The adhesion structures and the ECM components appear variously influenced by GM-CSF treatment as compared to fibroblasts cultured in basal condition, but newly only NHDF cells are really induced to increase their synthesis activity. We suggest that the in vitro treatment with GM-CSF can shift human normal fibroblasts towards a more differentiated state, due or accompanied by an increased expression of GM-CSFR and that such “differentiation” is an important event induced by such cytokine.

  11. Connecting the Cosmic Star Formation Rate with the Local Star Formation

    Science.gov (United States)

    Gribel, Carolina; Miranda, Oswaldo D.; Williams Vilas-Boas, José

    2017-11-01

    We present a model that unifies the cosmic star formation rate (CSFR), obtained through the hierarchical structure formation scenario, with the (Galactic) local star formation rate (SFR). It is possible to use the SFR to generate a CSFR mapping through the density probability distribution functions commonly used to study the role of turbulence in the star-forming regions of the Galaxy. We obtain a consistent mapping from redshift z˜ 20 up to the present (z = 0). Our results show that the turbulence exhibits a dual character, providing high values for the star formation efficiency ( ˜ 0.32) in the redshift interval z˜ 3.5{--}20 and reducing its value to =0.021 at z = 0. The value of the Mach number ({{ M }}{crit}), from which rapidly decreases, is dependent on both the polytropic index (Γ) and the minimum density contrast of the gas. We also derive Larson’s first law associated with the velocity dispersion ( ) in the local star formation regions. Our model shows good agreement with Larson’s law in the ˜ 10{--}50 {pc} range, providing typical temperatures {T}0˜ 10{--}80 {{K}} for the gas associated with star formation. As a consequence, dark matter halos of great mass could contain a number of halos of much smaller mass, and be able to form structures similar to globular clusters. Thus, Larson’s law emerges as a result of the very formation of large-scale structures, which in turn would allow the formation of galactic systems, including our Galaxy.

  12. Creutzfeldt-Jakob disease, Heidenhain variant: case report with MRI (DWI) findings; Doenca de Creutzfeldt-Jakob forma Heidenhain: relato de caso com achados de ressonancia magnetica e DWI

    Energy Technology Data Exchange (ETDEWEB)

    Arruda, Walter Oleschko; Bordignon, Kelly C; Milano, Jeronimo B; Ramina, Ricardo [Instituto de Neurologia de Curitiba, PR (Brazil)

    2004-06-01

    Creutzfeldt-Jakob disease (CJD) is a pre senile dementia characterized by rapidly progressive mental deterioration, myoclonic jerking, and other less common neurological signs. Few accentuates cases have been described in Brazil. A 54-year-old white woman, was admitted in our service with a month history of progressive, bilateral cortical blindness. After admission, she developed right partial motor seizures (right facial, upper and lower limbs), she became progressively aphasic (mixed aphasia). Seizures were controlled with phenytoine, but she developed choreoathetotic movements on her right dimidium, with partial control after introduction of chlorpromazine 25 mg q/d. She could no longer stand up or walk due to severe ataxia. The first EEG (October, 2001) showed left hemisphere severe seizure activity (status epilepticus partial is). She was delivered home with enteral nutrition, phenytoine, chlorpromazine and mepacrine 100 mg q d. The following laboratory tests were negative or normal: blood series, platelets, ESR, kidney and liver function, copper, ceruloplasmin, Vedril, HIV, HTLV-1, lactate, and cerebral Dsa (performed in other service). A spinal tap with normal opening pressure was perform and CSFR examination was normal. CSFR 14-3-3 protein was positive, CSF specific neuronal enolase 7.5 ng/ml(normal). Genetic study of PRNP gene did not disclosed any known mutation. A MRI (October, 2001) showed areas of hyperintense signal (T 2 and FLAIR) without Gd-enhancement on T1, in the left temporal lobe and in both occipital lobes; basal ganglia have a normal appearance. DWI imaging showed bright areas at the same sites. An EEG (March, 2002) disclosed a periodical sharp triphasic waves pattern, suggestive of CJD. A second MRI (April, 2002) showed mild generalized atrophy, no ventricular dilatation, and the hyperintense sites disappeared. She remained clinically stable and under use of chlorpromazine and mepacrine until she died due to pulmonary complications on April

  13. Creutzfeldt-Jakob disease, Heidenhain variant: case report with MRI (DWI) findings

    International Nuclear Information System (INIS)

    Arruda, Walter Oleschko; Bordignon, Kelly C.; Milano, Jeronimo B.; Ramina, Ricardo

    2004-01-01

    Creutzfeldt-Jakob disease (CJD) is a pre senile dementia characterized by rapidly progressive mental deterioration, myoclonic jerking, and other less common neurological signs. Few accentuates cases have been described in Brazil. A 54-year-old white woman, was admitted in our service with a month history of progressive, bilateral cortical blindness. After admission, she developed right partial motor seizures (right facial, upper and lower limbs), she became progressively aphasic (mixed aphasia). Seizures were controlled with phenytoine, but she developed choreoathetotic movements on her right dimidium, with partial control after introduction of chlorpromazine 25 mg q/d. She could no longer stand up or walk due to severe ataxia. The first EEG (October, 2001) showed left hemisphere severe seizure activity (status epilepticus partial is). She was delivered home with enteral nutrition, phenytoine, chlorpromazine and mepacrine 100 mg q d. The following laboratory tests were negative or normal: blood series, platelets, ESR, kidney and liver function, copper, ceruloplasmin, Vedril, HIV, HTLV-1, lactate, and cerebral Dsa (performed in other service). A spinal tap with normal opening pressure was perform and CSFR examination was normal. CSFR 14-3-3 protein was positive, CSF specific neuronal enolase 7.5 ng/ml(normal). Genetic study of PRNP gene did not disclosed any known mutation. A MRI (October, 2001) showed areas of hyperintense signal (T 2 and FLAIR) without Gd-enhancement on T1, in the left temporal lobe and in both occipital lobes; basal ganglia have a normal appearance. DWI imaging showed bright areas at the same sites. An EEG (March, 2002) disclosed a periodical sharp triphasic waves pattern, suggestive of CJD. A second MRI (April, 2002) showed mild generalized atrophy, no ventricular dilatation, and the hyperintense sites disappeared. She remained clinically stable and under use of chlorpromazine and mepacrine until she died due to pulmonary complications on April

  14. Estimating population health risk from low-level environmental radon

    International Nuclear Information System (INIS)

    Fisher, D.R.

    1980-01-01

    Although incidence of respiratory cancer is directly related to inhalation of radon and radon daughters, the magnitude of the actual risk is uncertain for members of the general population exposed for long periods to low-level concentrations. Currently, any such estimate of the risk must rely on data obtained through previous studies of underground-miner populations. Several methods of risk analysis have resulted from these studies. Since the breathing atmospheres, smoking patterns, and physiology are different between miners and the general public, overestimates of lung cancer risk to the latter may have resulted. Strong evidence exists to support the theory of synergistic action between alpha radiation and other agents, and therefore a modified relative risk model was developed to predict lung cancer risks to the general public. The model considers latent period, observation period, age dependency, and inherent risks from smoking or geographical location. A test of the model showed excellent agreement with results of the study of Czechoslovakian uranium miners, for which the necessary time factors were available. The risk model was also used to predict lung cancer incidence among residents of homes on reclaimed Florida phosphate lands, and results of this analysis indicate that over the space of many years, the increased incidence of lung cancer due to elevated radon levels may be indisgtinguishable from those due to other causes

  15. Economic Transformation in Democratic Czechoslovakia (1990–1992

    Directory of Open Access Journals (Sweden)

    V. Baka

    2016-07-01

    Full Text Available The article deals with the problem of the implementation of economic reforms in the Czech and Slovak Federal Republic, which took place in the aftermath of the Velvet Revolution, from 1990 to 1992. The author examines the positive and negative effects of economic transformation in the Czech Republic and the Slovak Republic. Particular attention is paid to influence of siting of the defense industry in the growth of unemployment in Slovakia. The article also touches upon the issue of privatization process, defined its main elements and stages. It is spoken in detail about illegal machinations, which are primarily engaged in by privatization investment funds. According to the article, Czechoslovakian politicians and economists agreed on a scenario of rapid market reforms, and Czech part of the federation was been better prepare for this step. Much attention is given to the problem of restitution of property that was nationalized in the period of communism. The researcher concludes that the economic transformation have contributed to the collapse of the federal state of Czechs and Slovaks.

  16. Jaromir Krejcar, technique’s apology. Czechoslovak pavilion at the International Exhibition, Paris 1937

    Directory of Open Access Journals (Sweden)

    María Pura Moreno Moreno

    2018-05-01

    Full Text Available L'Exposition Internationale des Arts et Techniques dans les Temps Modernes in Paris (1937 offered, across the construction of their pavilions to each country the opportunity to materialize their technological progress and also to show socio-political factors referred to the art’s integration or aesthetic formalization of their ideologies. In the Czechoslovakian pavilion, made by the architect Jaromir Krejcar, crystallized architectural postulates from the purism, cubism, and new objectivity close to technical aspects driven by the russian constructivism that had been developed during the first decades of the twentieth century in the architecture of the new Czech Republic. Its geographical location, at the crossroads of Europe, had favored the knowledge of avant-garde movements of its two flanks: the Eastern Europe –France, Germany, Austria and Netherlands— on one hand, and the other the Soviet Union. This article will recognize the architectural thought of Krejcar’s generation across the decisions of the pavilion’s project. The characteristics of its emplacement, combined with the moral obligation of showing the high technological level of his country, managed to place it as one of the best examples of that exhibition. Its materialization represent the culmination of previous architectural researches in Czechoslovakia that were truncated from 1939 by the political circumstances.

  17. Operator support system for power unit control in abnormal modes of operation

    International Nuclear Information System (INIS)

    Kurka, J.

    1993-01-01

    I and C system technology, partly Soviet and partly Czechoslovakian, used on the NPP Dukovany units represents the control technology standard of late 70-ties and it becomes the weak part of the whole system. The modernization of the system, therefore, is necessary and it is already in preparation. The specification of both the scope and the depth of upgrading/replacement is being carried out within the framework of the PHARE program. The second phase of the program aimed at the final specifications of requirements on new I and C system is in progress. The output will serve as detailed specification for bid invitation for control system supplier. Parallely, the preparation of specification for WWER-440 full-scope plant simulator for operator training is in progress as well. In the case of two units with WWER-1000 MW reactors, the completion of construction of which was even threaten for a certain period of time, essential changes have taken place in the design of both the I and C systems and the reactor core. 7 figs

  18. Il monumento praghese a Stalin: un’ombra ingombrante sul ‘disgelo’

    Directory of Open Access Journals (Sweden)

    Massimo Tria

    2006-12-01

    Full Text Available The Monument to Stalin in Prague. A Shadow that loomed over the 'Thaw' The author of the article focuses on the historical moment immediately preceding the death of Stalin and on the following period of partial political and cultural liberalization in the European countries under Soviet influence. He then concentrates on some guidelines of the so-called “destalinization” process, in particular regarding Czechoslovakia from 1953 to 1963. He points out how, for a few years, the leaders of the Czech Communist Party managed to stem the tide of positive reform and anti-dogmatic protests that instead spread faster and wider in countries like Hungary and Poland, as well as in the Soviet Union itself. The essay effectively highlights a symbolic episode of particular importance: the construction of the gigantic statue of Stalin, that dominated the landscape of Prague until 1962. Due to its particular chronology and its emblematic significance, this event summarizes the paradoxical features of the Czechoslovakian political context; in the historical moment in question, Czech politicians were particularly reluctant to welcome the liberal stimuli coming directly from Moscow and the Soviet leaders. Finally, he notes how the statue of Stalin has featured in several texts of Czech literature, in particular amongst dissenters and emigrants, but also in a classic such as Bohumil Hrabal, a testimony to the strong emotional and cultural impact that the statue had on the society and culture of Prague and of Czechoslovakia as a whole.

  19. Nuclear power in Czechoslovakia and its development within the CMEA collaboration framework

    International Nuclear Information System (INIS)

    Gavel, S.

    1984-01-01

    The state of and prospects for the development of nuclear pwer and energy machine building in Czechoslovakia are considered. In 1980 the NPPs generated 1.5% of the country's energy. It is envisaged that by 1990 the share of NPPs will rise to 9.2% and by 2000 to 15%. Since 1974, when an agreement with the Soviet Union was signed on cooperation and specialization in producing equipment for NPPs, a new industry, nuclear power machine building, has been successfully developed in Czechoslovakia. The production of steam generators, volume compensators, main gate valves, pipelines, special fittings, special pumps and auxiliary systems of the primary circuit has been mastered. Presently, the Czechoslovakian industry is capable of manufacturing above 80% of the whole servicing equipment of WWER-440 power units. Alongside with the USSR, Czechoslovakia has become the second greatest supplier of equipment for NPPs among the COMECON member-states. By 2000 it is planned to put into operation 12 WWER-440 power unit and 6 WWER-1000 power units with the total capacity of 11280 MW. An experimental agro-nuclear complex will soon be constructed in which NPP heat will be utilized for growing vegetables, fruit, mushrooms and for breeding fish

  20. [Molecular characterization of atypical chronic myeloid leukemia and chronic neutrophilic leukemia].

    Science.gov (United States)

    Senín, Alicia; Arenillas, Leonor; Martínez-Avilés, Luz; Fernández-Rodríguez, Concepción; Bellosillo, Beatriz; Florensa, Lourdes; Besses, Carles; Álvarez-Larrán, Alberto

    2015-06-08

    Atypical chronic myeloid leukemia (aCML) and chronic neutrophilic leukemia (CNL) display similar clinical and hematological characteristics. The objective of the present study was to determine the mutational status of SETBP1 and CSF3R in these diseases. The mutational status of SETBP1 and CSF3R was studied in 7 patients with aCML (n = 3), CNL (n = 1) and unclassifiable myeloproliferative neoplasms (MPN-u) (n = 3). Additionally, mutations in ASXL1, SRSF2, IDH1/2, DNMT3A, and RUNX1 were also analyzed. SETBP1 mutations (G870S and G872R) were detected in 2 patients with MPN-u, and one of them also presented mutations in SRSF2 (P95H) and ASXL1 (E635fs). The CNL case showed mutations in CSFR3 (T618I), SETBP1 (G870S) and SRSF2 (P95H). No patient classified as aCML had mutations in SETBP1 or CSF3R. One of the patients with mutations evolved to acute myeloid leukemia, while the other 2 had disease progression without transformation to overt leukemia. The knowledge of the molecular alterations involved in these rare diseases is useful in the diagnosis and may have an impact on both prognosis and therapy. Copyright © 2014 Elsevier España, S.L.U. All rights reserved.

  1. Inhibition of lignifying processes by sulfur dioxide

    International Nuclear Information System (INIS)

    Pfanz, H.; Oppmann, B.

    1991-01-01

    Intercellular washing fluids (IWF) from spruce needles (Picea abies L. Karst.) contain peroxidases 1-2% of total IWF protein. These apoplastic enzymes show the ability to polymerize monophenols or phenylpropanes to form lignin precursors in vitro. In the presence of potentially acidic air pollutants like NO 2 , HF(20 mM of salts in solution), and in the presence of Pb-, Cd- (0.5 mM) or Al-salts (8 mM) no inhibitory effect on the polymerization reactions examined was detectable. In contrast, the anions of SO 2 (sulfite and bisulfite) revealed a strong inhibition on the dimerization of ferulic and caffeic acid (Ki ca. 1 mM), and on the dehydration of syringaldazine (Ki ca. 8 μM). Polymerization of coniferyl alcohol, on the other hand, seemed to be enhanced. Maier-Maercker and Koch (1986) demonstrated that the cell walls of guard cells from undamaged spruce needles are properly lignified, whereas those of damaged needles seem to be affected. It is therefore assumed that cell wall lignification, and concomitantly stomatal regulation of coniferous needles are disturbed in regions with high atmospheric SO 2 pollution (e.g. Ore Mountains in CSFR)

  2. Compressional, mechanical and release properties of a novel gum in paracetamol tablet formulations

    Directory of Open Access Journals (Sweden)

    Adedokun Musiliu O.

    2014-09-01

    Full Text Available The binding properties of Eucalyptus gum obtained from the incised trunk of Eucalyptus tereticornis, were evaluated in paracetamol tablet formulations, in comparison with that of Gelatin B.P. In so doing, the compression properties were analyzed using density measurements and the compression equations of Heckel, Kawakita and Gurham. In our work, the mechanical properties of the tablets were assessed using the crushing strength and friability of the tablets, while the drug release properties of the tablets were assessed using disintegration and dissolution times. The results of the study reveal that tablet formulations incorporating Eucalyptus gum as binder, exhibited faster onset and higher amount of plastic deformation during compression than those containing gelatin. What is more, the Gurnham equation could be used as a substitute for the Kawakita equation in describing the compression properties of pharmaceutical tablets. Furthermore, the crushing strength, disintegration and dissolution times of the tablets increased with binder concentration, while friability values decreased. We noted that no significant differences in properties exist between formulations derived from the two binders (p > 0.05 exist. While tablets incorporating gelatin exhibited higher values for mechanical properties, Eucalyptus gum tablets had better balance between mechanical and release properties - as seen from the CSFR/Dt values. Tablets of good mechanical and release properties were prepared using Eucalyptus gum as a binder, and, therefore, it could serve as an alternative binder in producing tablets with good mechanical strength and fast drug release.

  3. Hypereosinophilia associated with genital sarcomas

    International Nuclear Information System (INIS)

    Terzieff, V.; Alonso, I.; V ázquez, A.

    2004-01-01

    Eosinophils are phagocytic leukocytes, regulators reactions Mast cell mediated hypersensitivity. toxicity and primarily responsible antiparasitic. Predominate in epithelial tissues near the interface surface (skin, digestive tract) .The cytotoxic reaction exerted by deposit cell surface substances from the granules themselves: peroxidases, neurotoxins, and other cationic proteins. Hypereosinophilia is defined as the increase in eosinophils above 1500 / m m3. The most common causes are parasitic infections and reactions allergic. About 60% of tumors may be associated with an elevation eosinophil discrete but marked eosinophilia in these patients is little frequent. Tumors are most associated lung cancer and tumors hematology. There are few reports of this entity associated with uterine sarcomas. Although the pathophysiologic mechanism is unclear, it is assumed that the base is the increased secretion of cytokines eosinofilopoiétics: interleukins (Il) I L-3, Il-5 and G M-CSFR among other possible. Self-morbidity is primarily maintained hypereosinophilic heart, and is derived from the cytotoxic action, with endomyocardial fibrosis and thrombosis. Treatment should be directed at the control of the underlying disease, as good Tumor response was associated with the account and normalizaciónd eosinófiles. Los corticosteroids prednisone, 60 mg / day po) may be effective because antagonize The stimulatory effect of cytokines. In the vast majority of cases, the disease is associated with hypereosinophilia disseminated and poor overall prognosis We present a case of vaginal sarcoma with pulmonary metastases and hypereosinophilia seniors who responded to treatment with chemotherapy

  4. Large Scale EOF Analysis of Climate Data

    Science.gov (United States)

    Prabhat, M.; Gittens, A.; Kashinath, K.; Cavanaugh, N. R.; Mahoney, M.

    2016-12-01

    We present a distributed approach towards extracting EOFs from 3D climate data. We implement the method in Apache Spark, and process multi-TB sized datasets on O(1000-10,000) cores. We apply this method to latitude-weighted ocean temperature data from CSFR, a 2.2 terabyte-sized data set comprising ocean and subsurface reanalysis measurements collected at 41 levels in the ocean, at 6 hour intervals over 31 years. We extract the first 100 EOFs of this full data set and compare to the EOFs computed simply on the surface temperature field. Our analyses provide evidence of Kelvin and Rossy waves and components of large-scale modes of oscillation including the ENSO and PDO that are not visible in the usual SST EOFs. Further, they provide information on the the most influential parts of the ocean, such as the thermocline, that exist below the surface. Work is ongoing to understand the factors determining the depth-varying spatial patterns observed in the EOFs. We will experiment with weighting schemes to appropriately account for the differing depths of the observations. We also plan to apply the same distributed approach to analysis of analysis of 3D atmospheric climatic data sets, including multiple variables. Because the atmosphere changes on a quicker time-scale than the ocean, we expect that the results will demonstrate an even greater advantage to computing 3D EOFs in lieu of 2D EOFs.

  5. CO2/H2O adsorption equilibrium and rates on metal-organic frameworks: HKUST-1 and Ni/DOBDC.

    Science.gov (United States)

    Liu, Jian; Wang, Yu; Benin, Annabelle I; Jakubczak, Paulina; Willis, Richard R; LeVan, M Douglas

    2010-09-07

    Metal-organic frameworks (MOFs) have recently attracted intense research interest because of their permanent porous structures, huge surface areas, and potential applications as novel adsorbents and catalysts. In order to provide a basis for consideration of MOFs for removal of carbon dioxide from gases containing water vapor, such as flue gas, we have studied adsorption equilibrium of CO(2), H(2)O vapor, and their mixtures and also rates of CO(2) adsorption in two MOFs: HKUST-1 (CuBTC) and Ni/DOBDC (CPO-27-Ni or Ni/MOF-74). The MOFs were synthesized via solvothermal methods, and the as-synthesized products were solvent exchanged and regenerated before experiments. Pure component adsorption equilibria and CO(2)/H(2)O binary adsorption equilibria were studied using a volumetric system. The effects of H(2)O adsorption on CO(2) adsorption for both MOF samples were determined, and the results for 5A and NaX zeolites were included for comparison. The hydrothermal stabilities for the two MOFs over the course of repetitive measurements of H(2)O and CO(2)/H(2)O mixture equilibria were also studied. CO(2) adsorption rates from helium for the MOF samples were investigated by using a unique concentration-swing frequency response (CSFR) system. Mass transfer into the MOFs is rapid with the controlling resistance found to be macropore diffusion, and rate parameters were established for the mechanism.

  6. Trophic Effects and Regenerative Potential of Mobilized Mesenchymal Stem Cells From Bone Marrow and Adipose Tissue as Alternative Cell Sources for Pulp/Dentin Regeneration.

    Science.gov (United States)

    Murakami, Masashi; Hayashi, Yuki; Iohara, Koichiro; Osako, Yohei; Hirose, Yujiro; Nakashima, Misako

    2015-01-01

    Dental pulp stem cell (DPSC) subsets mobilized by granulocyte-colony-stimulating factor (G-CSF) are safe and efficacious for complete pulp regeneration. The supply of autologous pulp tissue, however, is very limited in the aged. Therefore, alternative sources of mesenchymal stem/progenitor cells (MSCs) are needed for the cell therapy. In this study, DPSCs, bone marrow (BM), and adipose tissue (AD)-derived stem cells of the same individual dog were isolated using G-CSF-induced mobilization (MDPSCs, MBMSCs, and MADSCs). The positive rates of CXCR4 and G-CSFR in MDPSCs were similar to MADSCs and were significantly higher than those in MBMSCs. Trophic effects of MDPSCs on angiogenesis, neurite extension, migration, and antiapoptosis were higher than those of MBMSCs and MADSCs. Pulp-like loose connective tissues were regenerated in all three MSC transplantations. Significantly higher volume of regenerated pulp and higher density of vascularization and innervation were observed in response to MDPSCs compared to MBMSC and MADSC transplantation. Collagenous matrix containing dentin sialophosphoprotein (DSPP)-positive odontoblast-like cells was the highest in MBMSCs and significantly higher in MADSCs compared to MDPSCs. MBMSCs and MADSCs, therefore, have potential for pulp regeneration, although the volume of regenerated pulp tissue, angiogenesis, and reinnervation, were less. Thus, in conclusion, an alternative cell source for dental pulp/dentin regeneration are stem cells from BM and AD tissue.

  7. Lifetime-management and lifetime-extension at PAKS nuclear power plant

    International Nuclear Information System (INIS)

    Katona, Tamas; Ratkai, Sandor; Janosi, Agnes Biro

    2002-01-01

    core might be necessary with 50% probability. Annealing is not a cost critical method and it has been successfully applied in the practice of VVER plants (i.e. Finland and Slovakia). The study indicates that the steam generators are very important and critical equipment at VVER/440/213 type units. Stress corrosion cracking of heat-transfer tubes shall be considered also in case of steam generators. Considering the effects of already implemented measures, e.g. main turbine-condenser replacement, copper removing, protecting the secondary side of steam generators, replacement of steam generators can be excluded also in case of 50 years of operation. However, local corrosion effects appearing on the secondary side shall be monitored even in case of high PH water regime, transport of erosion products into the steam generator shall be minimised, for example by means of correct selection of the structural materials during replacement of high pressure preheaters. Business model of the lifetime extension covered incomes, expenses and financing. The costs of plant life management and lifetime extension have been estimated conservatively. The market conditions, including liberalisation of the electrical power market have been taken into account. The business analysis shows that the plant lifetime extension is a reasonable business decision. Based on the results of the feasibility study and the public acceptance of extended operation NPP Paks launched a project to elaborate on the plant lifetime management programme and preparation of the necessary licence renewal. The current ageing-management programme of the Paks NPP shall be reviewed and extended to the required scope of SSC (passive long living items). Programme for maintaining the environmental qualification of equipment has to be elaborated on and implemented. Simultaneous to the preparation of the license renewal new requirements shall be implemented in the field of maintenance, also. The plant lifetime extension might

  8. Severe accident management at the Loviisa NPP - Application of integrated ROAAM and PSA level 2

    International Nuclear Information System (INIS)

    Siltanen, S.; Routamo, T.; Tuomisto, H.; Lundstrom, P.

    2007-01-01

    The Risk Oriented Accident Analysis Methodology (ROAAM) was developed for assessment and management of rare, high consequence hazards. The purpose of most ROAAM applications has been to solve major, isolated severe accident issues related to early containment failure such as Mark-I Liner Attack and Direct Containment Heating. In addition to ROAAM in the issue resolution context, the so called Integrated ROAAM approach can be used to provide an overall frame of safety evaluation that allows determination of whether an adequate level of safety has been achieved for a plant. Integrated ROAAM approach brings together quantifications of probabilistic elements based on statistical inference and treatment of deterministic elements based on identification of dominant physics, for severe accident phenomenology, in a well defined and clearly structured way. Fortum, as an owner of the Loviisa NPP, used the Integrated ROAAM approach when developing and implementing a comprehensive severe accident management (SAM) strategy for the Loviisa NPP. The SAM strategy is based on unique features of this VVER-440 plant with ice condenser containment and it includes hardware modifications at the plant, substantial new I and C qualified for severe accident conditions, new SAM guidelines, a SAM Handbook, revision of emergency preparedness organization, and versatile training approaches. It could be argued that the resolution of individual severe accident issues is not sufficient for assessing the overall safety of a nuclear power plant, and thus the ROAAM (in an issue resolution context) is not performing the same function as a PSA study (level 2 included). Actually the Integrated ROAAM approach takes on even a more ambitious task than the PSA, since it determines how a balance can be achieved between accident prevention and mitigation of containment-threatening physical phenomena. Thus it provides a tool for implementing a sound diverse defence-in-depth strategy at a plant. Integrated

  9. ŠIMÁNĚ 2016

    Directory of Open Access Journals (Sweden)

    Editorial Foreword

    2016-12-01

    Full Text Available Czecho-Slovak Student Conference on Nuclear EngineeringPrague, Czech Republic, 12. – 13. May 2016Edited byVojtěch Caha, Martin Cesnek, Adam Kecek, Martin Ševeček and Milan ŠtefánikOrganizing InstitutionFaculty of Nuclear Sciences and Physical Engineering, Department of Nuclear Reactors (FNSPE CTU, Czech Technical University in PragueOrganizing CommitteeVojtěch Caha, FNSPE CTU, Czech RepublicMartin Cesnek, FNSPE CTU, MagyarországAdam Kecek, FNSPE CTU, Czech RepublicMartin Ševeček, FNSPE CTU, Czech RepublicMilan Štefánik, FNSPE CTU, Czech RepublicForeword:Nuclear science and technology have a long history within the Czechoslovakia. Even though the Czech Republic and Slovakia have been separated since 1993, wide cooperation on different levels including education and research still exists. In 2015, the 60th anniversary of the Faculty of Nuclear Sciences and Physical Engineering and also several other institutions and companies in the Czech Republic was celebrated. In the light of these events, the organizing committee decided to hold a student conference on nuclear engineering. The only question was whom should be the conference dedicated. The right person seemed to be professor Čestmír Šimáně, one of the Czechoslovakian nuclear pioneers.Professor Šimáně graduated at the Dr. Edvard Beneš Technical university in Brno. During his early career, he had the opportunity to study at Fréderic Joliot-Curie at College de France. Later in Czechoslovakia, he significantly contributed the foundation of the nuclear science and industry in the Czechoslovakia. He became the first employee of the Institute of Nuclear Physics at the Czechoslovakian Academy of Science, later he became the first director of the Nuclear Research Institute in Řež. Professor Šimáně was well established even on the international level. He was in charge as a director assign at the Joint Institute of Nuclear Research in Dubna, Russia or as a division director at

  10. The LA-10 Project for complex testing of nuclear incineration systems

    International Nuclear Information System (INIS)

    Hron, M.; Chochlovsky, I.; Lelek, V.; Peka, I.; Valenta, V.

    1997-01-01

    The current proposals for high-active long-lived (more then 10''4 years) waste from spent nuclear fuel disposal call forth an increasing mistrust of society towards nuclear power at all. These problems are highly topical even in the Czech Republic. In 1993, the Czech Republic and the Slovak Republic were constituted instead of the former Czechoslovakia. The first of two operated Czechoslovakian NNPs, the NNP Bohunice together with the only one in that time intermediate storage of spent fuel, remains on the territory of Slovakia and the NPP Dukovany as well as the NPP Temelian (designed with totally 4 x VVER-1000, 2 x VVER-1000 of which are ready for start up in the next few years) are located in the Czech Republic. As a consequence of this political step, the decision was accepted to build up an intermediate storage in the Czech Republic (to put it into operation in 2005) and to store all Dukovany's fuel in transport casks (84 complete fuel assemblies per one cask) on the Dukovany's site territory up to the 600 tons of heavy metal and the move it into the new intermediate storage, which should be designed for the lifetime of about 50-60 years. At about 2030 as a main variant, the final disposal of all spent fuel should be prepared. There is an opinion of a rational part of the Czech nuclear community that the final disposal of spent fuel is only something like fairy-tale for public and that a real technical solution should be found definitely. (Author) 13 refs

  11. Atlanto-axial malformation and instability in dogs with pituitary dwarfism due to an LHX3 mutation.

    Science.gov (United States)

    Voorbij, A M W Y; Meij, B P; van Bruggen, L W L; Grinwis, G C M; Stassen, Q E M; Kooistra, H S

    2015-01-01

    Canine pituitary dwarfism or combined pituitary hormone deficiency (CPHD) in shepherd dogs is associated with an LHX3 mutation and can lead to a wide range of clinical manifestations. Some dogs with CPHD have neurological signs that are localized to the cervical spine. In human CPHD, caused by an LHX3 mutation, anatomical abnormalities in the atlanto-axial (C1-C2) joint have been described. To evaluate the presence of atlanto-axial malformations in dogs with pituitary dwarfism associated with an LHX3 mutation and to investigate the degree of similarity between the atlanto-axial anomalies found in canine and human CPHD patients with an LHX3 mutation. Three client-owned Czechoslovakian wolfdogs and 1 client-owned German shepherd dog, previously diagnosed with pituitary dwarfism caused by an LHX3 mutation, with neurological signs indicating a cervical spinal disorder. Radiography, computed tomography, and magnetic resonance imaging of the cranial neck and skull, necropsy, and histology. Diagnostic imaging identified abnormal positioning of the dens axis and incomplete ossification of the suture lines between the ossification centers of the atlas with concurrent atlanto-axial instability and dynamic compression of the spinal cord by the dens axis. The malformations and aberrant motion at C1-C2 were confirmed at necropsy and histology. The atlanto-axial abnormalities of the dwarf dogs resemble those encountered in human CPHD patients with an LHX3 mutation. These findings suggest an association between the LHX3 mutation in dogs with CPHD and atlanto-axial malformations. Consequently, pituitary dwarfs should be monitored closely for neurological signs. Copyright © 2015 The Authors. Journal of Veterinary Internal Medicine published by Wiley Periodicals, Inc. on behalf of American College of Veterinary Internal Medicine.

  12. Solution of safety problems for nuclear power plants with WWER-440 reactors

    International Nuclear Information System (INIS)

    Krett, V.; Pernitsa, R.; Pfann, Ya.; Zbeglik, J.

    1982-01-01

    Institute of nuclear research (INR) of Czechoslovakian Atomic Energy Commission isto fulfil the supervision functions within the field of nuclear power research and development. The problems of safe operation ensurance for the nuclear power plants (NPP) with WWER-440 reactors are studied within the frame of sever major issues: code standardization and devolopment of guiding materials for the state supervision; neutronic and thermohydrolic data processing for the accident analysis; operation reliability studies of the safety systems and estimates of separate component failure importance; assessment of the accidents resulting from the equipment misfunctioning and component failures; development of a controlled reliability program; evaluation of the atomic installationsimpact on the environment; ensurence of the reactor vessel reliability and durability under irradiation. The NPP safety analysis incorporates the calculations of transient and accidental regimes for the core, the primary loop and the entire plant. A number of codes has been produced which allow to determine the state of fuel elements during operation just before the accidents assessed, thermohydrolic conditions in the coolant and the temperature distribution within the fuel both for the stationary reactor conditions and for transient regimes. A mathematical model has been deveoloped, including the description of all the primary loop major components. The Soviet code DYNAMIKA has been adopted and adjusted for EC-1040 computer, there by the accident analysis for the entire NPP has been made possible. On the basis of american SAFTE code a faster SAFEDO-2 code has been developed employing the Monte Carlo method for the accident analysis of a complex system described by means of a failure tree. The discussed codes are used at the data assessment for the accident analysis part of the safety reports as well as for the reliability evaluation of the emergency core cooling system [ru

  13. Characteristics of the aerial regime of an aridisol at the central high plateau of Bolivia

    International Nuclear Information System (INIS)

    Orsag, V.

    1989-01-01

    In this work, the aerial regime of an aridisol with a loam clay-sandy texture of the central area of the high plateau of Bolivia, was determined as a function of the monthly volumetric moisture (using mean values) and the total porosity of the horizons Ap (0-19 cm), Bt (19-49 cm) and Cca (49-70 cm), Bt (19-49 cm) and Cca (49-70 cm). Values of moisture were determined during three year: 1983-84, 1984-5 and 1985-86 with the neutron probe. Fluctuations of the soil air content during the research period in the Ap and Bt horizons were strictly dependant on the precipitations and their seasonal variations due to porosity which was considered constant. Because this work was made in a soil with natural vegetation cover and without physical changes produced by tillage. The soil air content in the sandy Ap horizon showed higher values and greater difference (4-27%) than inferior horizons rich in clay (2-14%). The results were compared with quotes from the Czechoslovakian literature on the requirement of soil air content for certain crops: 15-35% for potatoes, 10-15% for cereals and 5-10% for pastures. This aridisol showing certain properties as clay horizons from 20 cm depth, unstable structure and insufficient air contents, can only be used for cereals and pastures. The use of these soils could be extended to crops with greater air requirements, if the water and aerial regimes are improved with the help of subsoilers, deepening the top soil and incorporating manure and green manure for improving the soil structure and porosity

  14. Genetic mapping of a new race specific resistance allele effective to Puccinia hordei at the Rph9/Rph12 locus on chromosome 5HL in barley.

    Science.gov (United States)

    Dracatos, Peter M; Khatkar, Mehar S; Singh, Davinder; Park, Robert F

    2014-12-20

    Barley is an important cereal crop cultivated for malt and ruminant feed and in certain regions it is used for human consumption. It is vulnerable to numerous foliar diseases including barley leaf rust caused by the pathogen Puccinia hordei. A temporarily designated resistance locus RphCantala (RphC) identified in the Australian Hordeum vulgare L. cultivar 'Cantala' displayed an intermediate to low infection type (";12 = N") against the P. hordei pathotype 253P- (virulent on Rph1, Rph2, Rph4, Rph6, Rph8 and RphQ). Phenotypic assessment of a 'CI 9214' (susceptible) x 'Stirling' (RphC) (CI 9214/Stirling) doubled haploid (DH) population at the seedling stage using P. hordei pathotype 253P-, confirmed that RphC was monogenically inherited. Marker-trait association analysis of RphC in the CI 9214/Stirling DH population using 4,500 DArT-seq markers identified a highly significant (-log10Pvalue > 17) single peak on the long arm of chromosome 5H (5HL). Further tests of allelism determined that RphC was genetically independent of Rph3, Rph7, Rph11, Rph13 and Rph14, and was an allele of Rph12 (Rph9.z), which also maps to 5HL. Multipathotype tests and subsequent pedigree analysis determined that 14 related Australian barley varieties (including 'Stirling' and 'Cantala') carry RphC and that the likely source of this resistance is via a Czechoslovakian landrace LV-Kvasice-NA-Morave transferred through common ancestral cultivars 'Hanna' and 'Abed Binder'. RphC is an allele of Rph12 (Rph9.z) and is therefore designated Rph9.am. Bioinformatic analysis using sequence arrays from DArT-seq markers in linkage disequilibrium with Rph9.am identified possible candidates for further gene cloning efforts and marker development at the Rph9/Rph12/Rph9.am locus.

  15. Indoor air radon

    International Nuclear Information System (INIS)

    Cothern, C.R.

    1990-01-01

    This review concerns primarily the health effects that result from indoor air exposure to radon gas and its progeny. Radon enters homes mainly from the soil through cracks in the foundation and other holes to the geologic deposits beneath these structures. Once inside the home the gas decays (half-life 3.8 d) and the ionized atoms adsorb to dust particles and are inhaled. These particles lodge in the lung and can cause lung cancer. The introduction to this review gives some background properties of radon and its progeny that are important to understanding this public health problem as well as a discussion of the units used to describe its concentrations. The data describing the health effects of inhaled radon and its progeny come both from epidemiological and animal studies. The estimates of risk from these two data bases are consistent within a factor of two. The epidemiological studies are primarily for hard rock miners, although some data exist for environmental exposures. The most complete studies are those of the US, Canadian, and Czechoslovakian uranium miners. Although all studies have some deficiencies, those of major importance include uranium miners in Saskatchewan, Canada, Swedish iron miners, and Newfoundland fluorspar miners. These six studies provide varying degrees of detail in the form of dose-response curves. Other epidemiological studies that do not provide quantitative dose-response information, but are useful in describing the health effects, include coal, iron ore and tin miners in the UK, iron ore miners in the Grangesburg and Kiruna, Sweden, metal miners in the US, Navajo uranium miners in the US, Norwegian niobian and magnitite miners, South African gold and uranium miners, French uranium miners, zinc-lead miners in Sweden and a variety of small studies of environmental exposure. An analysis of the epidemiological studies reveals a variety of interpretation problem areas.172 references

  16. Changes in the state territory, systemic changes, the development of ecological thinking and the issues of the utilisation of hydroelectric power in Hungary

    Directory of Open Access Journals (Sweden)

    HAJDÚ Zoltán

    2012-09-01

    Full Text Available Each big river, including the Danube, makes a single system from their spring to their delta or estuary. Wherever we intervene into the system, this will be necessarily affect the whole of that. Thespillover effects are especially striking in the case of the lower reaches, in the Danube River in the Delta area. The utilisation of the hydroelectric power of the Danube River has been continuously on the agenda in the Hungarian areas since the turn of the 19th and 20th centuries, as an economic, planning, development and thenpolitical and environmental issue. The plans made before the World War I featured hydroelectric plants as the carriers of technical and scientific progress. World War I and the subsequent changes of the state borders swept these plans aside. Between the two world wars, the planning of hydroelectric power stations on the Danube wascontinued within the new state borders, two Hungarian governments and prime ministers of the time were fully committed to the construction of three power plants on the Danube River. After World War II, the hydroelectric plant planned on the Hungarian-Czechoslovakian border and its construction became a basic economic and, from the 1980s, an environmental issue. Finally, only a part of the planned hydroelectric power plant system was constructed, and the issue was not finally closed by specific interpretation of the international law. The Danube Strategy of the European Union, the programmes until 2020, the increase of the role of renewable anergies, the nuclear disaster in Japan, and the interests of international navigation filled the debate with new content. Thepossibility of the construction of hydroelectric power plants on the Danube River was “accidentally” put among the official documents of the government in power since 2010. The government declared it had nothing to do with the plans, and the debate has been going on since then.

  17. Imatinib mesylate exerts anti-proliferative effects on osteosarcoma cells and inhibits the tumour growth in immunocompetent murine models.

    Directory of Open Access Journals (Sweden)

    Bérengère Gobin

    Full Text Available Osteosarcoma is the most common primary malignant bone tumour characterized by osteoid production and/or osteolytic lesions of bone. A lack of response to chemotherapeutic treatments shows the importance of exploring new therapeutic methods. Imatinib mesylate (Gleevec, Novartis Pharma, a tyrosine kinase inhibitor, was originally developed for the treatment of chronic myeloid leukemia. Several studies revealed that imatinib mesylate inhibits osteoclast differentiation through the M-CSFR pathway and activates osteoblast differentiation through PDGFR pathway, two key cells involved in the vicious cycle controlling the tumour development. The present study investigated the in vitro effects of imatinib mesylate on the proliferation, apoptosis, cell cycle, and migration ability of five osteosarcoma cell lines (human: MG-63, HOS; rat: OSRGA; mice: MOS-J, POS-1. Imatinib mesylate was also assessed as a curative and preventive treatment in two syngenic osteosarcoma models: MOS-J (mixed osteoblastic/osteolytic osteosarcoma and POS-1 (undifferentiated osteosarcoma. Imatinib mesylate exhibited a dose-dependent anti-proliferative effect in all cell lines studied. The drug induced a G0/G1 cell cycle arrest in most cell lines, except for POS-1 and HOS cells that were blocked in the S phase. In addition, imatinib mesylate induced cell death and strongly inhibited osteosarcoma cell migration. In the MOS-J osteosarcoma model, oral administration of imatinib mesylate significantly inhibited the tumour development in both preventive and curative approaches. A phospho-receptor tyrosine kinase array kit revealed that PDGFRα, among 7 other receptors (PDFGFRβ, Axl, RYK, EGFR, EphA2 and 10, IGF1R, appears as one of the main molecular targets for imatinib mesylate. In the light of the present study and the literature, it would be particularly interesting to revisit therapeutic evaluation of imatinib mesylate in osteosarcoma according to the tyrosine-kinase receptor

  18. STAT3-Activated GM-CSFRα Translocates to the Nucleus and Protects CLL Cells from Apoptosis

    Science.gov (United States)

    Li, Ping; Harris, David; Liu, Zhiming; Rozovski, Uri; Ferrajoli, Alessandra; Wang, Yongtao; Bueso-Ramos, Carlos; Hazan-Halevy, Inbal; Grgurevic, Srdana; Wierda, William; Burger, Jan; O'Brien, Susan; Faderl, Stefan; Keating, Michael; Estrov, Zeev

    2014-01-01

    Here it was determined that Chronic Lymphocytic Leukemia (CLL) cells express the α-subunit but not the β-subunit of the granulocyte-macrophage colony-stimulating factor receptor (GM-CSFR/CSF3R). GM-CSFRα was detected on the surface, in the cytosol, and the nucleus of CLL cells via confocal microscopy, cell fractionation, and GM-CSFRα antibody epitope mapping. Because STAT3 is frequently activated in CLL and the GM-CSFRα promoter harbors putative STAT3 consensus binding sites, MM1 cells were transfected with truncated forms of the GM-CSFRα promoter, then stimulated with IL-6 to activate STAT3 to identify STAT3 binding sites. Chromatin immunoprecipitation (ChIP) and an electoromobility shift assay (EMSA) confirmed STAT3 occupancy to those promoter regions in both IL-6 stimulated MM1 and CLL cells. Transfection of MM1 cells with STAT3 siRNA or CLL cells with STAT3 shRNA significantly down-regulated GM-CSFRα mRNA and protein levels. RNA transcripts, involved in regulating cell-survival pathways, and the proteins KAP1 (TRIM28) and ISG15 co-immunoprecipitated with GM-CSFRα. GM-CSFRα-bound KAP1 enhanced the transcriptional activity of STAT3, whereas ISG15 inhibited the NF-κB pathway. Nevertheless, overexpression of GM-CSFRα protected MM1 cells from dexamethasone-induced apoptosis, and GM-CSFRα knockdown induced apoptosis in CLL cells, suggesting that GM-CSFRα provides a ligand-independent survival advantage. PMID:24836891

  19. Evaluation of a High-Resolution Regional Reanalysis for Europe

    Science.gov (United States)

    Ohlwein, C.; Wahl, S.; Keller, J. D.; Bollmeyer, C.

    2014-12-01

    Reanalyses gain more and more importance as a source of meteorological information for many purposes and applications. Several global reanalyses projects (e.g., ERA, MERRA, CSFR, JMA9) produce and verify these data sets to provide time series as long as possible combined with a high data quality. Due to a spatial resolution down to 50-70km and 3-hourly temporal output, they are not suitable for small scale problems (e.g., regional climate assessment, meso-scale NWP verification, input for subsequent models such as river runoff simulations). The implementation of regional reanalyses based on a limited area model along with a data assimilation scheme is able to generate reanalysis data sets with high spatio-temporal resolution. Within the Hans-Ertel-Centre for Weather Research (HErZ), the climate monitoring branch concentrates efforts on the assessment and analysis of regional climate in Germany and Europe. In joint cooperation with DWD (German Meteorological Service), a high-resolution reanalysis system based on the COSMO model has been developed. The regional reanalysis for Europe matches the domain of the CORDEX EURO-11 specifications, albeit at a higher spatial resolution, i.e., 0.055° (6km) instead of 0.11° (12km) and comprises the assimilation of observational data using the existing nudging scheme of COSMO complemented by a special soil moisture analysis with boundary conditions provided by ERA-Interim data. The reanalysis data set covers 6 years (2007-2012) and is currently extended to 16 years. Extensive evaluation of the reanalysis is performed using independent observations with special emphasis on precipitation and high-impact weather situations indicating a better representation of small scale variability. Further, the evaluation shows an added value of the regional reanalysis with respect to the forcing ERA Interim reanalysis and compared to a pure high-resolution dynamical downscaling approach without data assimilation.

  20. The high-resolution regional reanalysis COSMO-REA6

    Science.gov (United States)

    Ohlwein, C.

    2016-12-01

    Reanalyses gain more and more importance as a source of meteorological information for many purposes and applications. Several global reanalyses projects (e.g., ERA, MERRA, CSFR, JMA9) produce and verify these data sets to provide time series as long as possible combined with a high data quality. Due to a spatial resolution down to 50-70km and 3-hourly temporal output, they are not suitable for small scale problems (e.g., regional climate assessment, meso-scale NWP verification, input for subsequent models such as river runoff simulations). The implementation of regional reanalyses based on a limited area model along with a data assimilation scheme is able to generate reanalysis data sets with high spatio-temporal resolution. Within the Hans-Ertel-Centre for Weather Research (HErZ), the climate monitoring branch concentrates efforts on the assessment and analysis of regional climate in Germany and Europe. In joint cooperation with DWD (German Meteorological Service), a high-resolution reanalysis system based on the COSMO model has been developed. The regional reanalysis for Europe matches the domain of the CORDEX EURO-11 specifications, albeit at a higher spatial resolution, i.e., 0.055° (6km) instead of 0.11° (12km) and comprises the assimilation of observational data using the existing nudging scheme of COSMO complemented by a special soil moisture analysis with boundary conditions provided by ERA-Interim data. The reanalysis data set covers the past 20 years. Extensive evaluation of the reanalysis is performed using independent observations with special emphasis on precipitation and high-impact weather situations indicating a better representation of small scale variability. Further, the evaluation shows an added value of the regional reanalysis with respect to the forcing ERA Interim reanalysis and compared to a pure high-resolution dynamical downscaling approach without data assimilation.