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

International Nuclear Information System (INIS)

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

2008-01-01

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

Comparison of the results of the fifth dynamic AER benchmark-a benchmark for coupled thermohydraulic system/three-dimensional hexagonal kinetic core models

International Nuclear Information System (INIS)

Kliem, S.

1998-01-01

The fifth dynamic benchmark was defined at seventh AER-Symposium, held in Hoernitz, Germany in 1997. It is the first benchmark for coupled thermohydraulic system/three-dimensional hexagonal neutron kinetic core models. In this benchmark the interaction between the components of a WWER-440 NPP with the reactor core has been investigated. The initiating event is a symmetrical break of the main steam header at the end of the first fuel cycle and hot shutdown conditions with one control rod group stucking. This break causes an overcooling of the primary circuit. During this overcooling the scram reactivity is compensated and the scrammed reactor becomes re critical. The calculation was continued until the highly-borated water from the high pressure injection system terminated the power excursion. Each participant used own best-estimate nuclear cross section data. Only the initial subcriticality at the beginning of the transient was given. Solutions were received from Kurchatov Institute Russia with the code BIPR8/ATHLET, VTT Energy Finland with HEXTRAN/SMABRE, NRI Rez Czech Republic with DYN3/ATHLET, KFKI Budapest Hungary with KIKO3D/ATHLET and from FZR Germany with the code DYN3D/ATHLET.In this paper the results are compared. Beside the comparison of global results, the behaviour of several thermohydraulic and neutron kinetic parameters is presented to discuss the revealed differences between the solutions.(Authors)

Verification calculations for the WWER version of the TRANSURANUS code

International Nuclear Information System (INIS)

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

2006-01-01

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

Code package to analyse behavior of the WWER fuel rods in normal operation: TOPRA's code

International Nuclear Information System (INIS)

Scheglov, A.; Proselkov, V.

2001-01-01

This paper briefly describes the code package intended for analysis of WWER fuel rod characteristics. The package includes two computer codes: TOPRA-1 and TOPRA-2 for full-scale fuel rod analyses; MRZ and MKK codes for analyzing the separate sections of fuel rods in r-z and r-j geometry. The TOPRA's codes are developed on the base of PIN-mod2 version and verified against experimental results obtained in MR, MIR and Halden research reactors (in the framework of SOFIT, FGR-2 and FUMEX experimental programs). Comparative analysis of calculation results and results from post-reactor examination of the WWER-440 and WWER-1000 fuel rod are also made as additional verification of these codes. To avoid the enlarging of uncertainties in fuel behavior prediction as a result of simplifying of the fuel geometry, MKK and MRZ codes are developed on the basis of the finite element method with use of the three nodal finite elements. Results obtained in the course of the code verification indicate the possibility for application of the method and TOPRA's code for simplified engineering calculations of WWER fuel rods thermal-physical parameters. An analysis of maximum relative errors for predicting of the fuel rod characteristics in the range of the accepted parameter values is also presented in the paper

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

International Nuclear Information System (INIS)

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

1994-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1994-12-31

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

Code systems for effective and precise calculation of the basic neutron characteristics, core loading optimization, analysis and estimation of the operation regimes of WWER type reactors

International Nuclear Information System (INIS)

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

1993-01-01

Two directions for investigations are suggested: 1) Analysis and evaluation of the real loading patterns and operational regimes for Kozloduy NPP WWER-440 and WWER-1000 in the frame of the recent safety criteria and nuclear power plant operating limits. 2) Development of modern code system for WWER type reactor core analysis with advanced features: new design and materials for fuel and control rods, increasing the fuel enrichment, using the integral and discrete burnable absorbers etc. The fuel technology design evolution maximizes the fuel utilization efficiency, improves operation performance and enhances safety margins. By the joint efforts of specialists from INRNE, Sofia (BG) and KAB, Berlin (GE), the codes NESSEL-IV-EC, PYTHIA and DERAB have been developed and verified. In the frame of the PHARE programme the joint project ASPERCA has been proposed intended for reactor physics calculations with PHYBER-WWER code for safety enhancement and operation reliability improvement. In-core fuel management benchmarks for 4 cycles of unit 2 (WWER-440) and 2 cycles of unit 5 (WWER-1000) have been performed. The coordination of burnable absorber design implementation, low leakage loadings usage, reloading enrichment increase and steel content reduction in the core have made the reactor core analysis more demanding and the definition of loading patterns - more difficult. This complexity requires routine use of three-dimensional fast accurate core model with extended and updated cross section libraries. To meet the needs of WWER advanced loading patterns and in-core fuel management improvements the HEXANES code systems is being developed and qualified. Some test calculations have been carried out by the HEXANES code system investigating the influence of Gd in the fuel on the main reactor physics parameters. For reevaluation of the core safety-related design limits forming the basis of licensing procedure, the code DYN3D/M2 is used. 16 refs., 3 figs. (author)

Development of codes for physical calculations of WWER

International Nuclear Information System (INIS)

Novikov, A.N.

2000-01-01

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

Validation of the computer code system ATHLET / ATHLET-CD. Final report; Validierung des Rechenprogrammsystems ATHLET / ATHLET-CD. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Austregesilo, H.; Bals, C.; Erdmann, W.; Horche, W.; Krzykacz-Hausmann, B.; Pointner, W.; Schoeffel, P.; Skorek, T.; Weber, S.; Wielenberg, A.

2010-04-15

In the frame of the reactor safety project RS1173, sponsored by the German Federal Ministry of Economics and Technology, analyses of international integral and separate effects tests have been performed for the validation of the code system ATHLET/ATHLET-CD. The work mainly comprised post-test calculations of selected experiments and the contributions to the working groups accompanying the experimental programs. For the assessment of the thermal-hydraulic models in ATHLET 8 integral tests and 4 separate effect tests have been considered. Together with the corroboration of the existing models, the validation analyses were mainly dedicated to the assessment of the modelling of non-condensable gases and their influence on two-phase natural circulation and on the primary heat removal through steam generators, as well as of the simulation of multi-dimensional flow processes. The validation calculations with respect to the simulation of multi-dimensional one- and two-phase flows aimed to investigate the range of applicability and limitations of the method of parallel channels in connection with the separate momentum equations for water and steam current used in ATHLET as well as to assess the status of the coupled version ATHLET/FLUBOX-3D. The ATHLET-CD validation analyses included the post-test calculations of 9 bundle tests, and was mainly focussed on the assessment of the improved and new models for core degradation, including the models for oxidation, melt formation and relocation for BWR components, as well as of the modelling of fission products and aerosol transport within the primary circuit taking into account chemical reactions within the module SOPHAEROS. As an additional contribution to code validation, the GRS methodology of uncertainty and sensitivity analysis was applied exemplarily to two validation calculations, one with ATHLET and one with ATHLET-CD. The results of these uncertainty analyses endorse the capability of the code system to reproduce

Consistent Code Qualification Process and Application to WWER-1000 NPP

International Nuclear Information System (INIS)

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

2006-01-01

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

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

International Nuclear Information System (INIS)

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

2002-01-01

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

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.

Feasibility analysis of the modified ATHLET code for supercritical water cooled systems

Energy Technology Data Exchange (ETDEWEB)

Zhou Chong, E-mail: ch.zhou@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dongchuan Road, Shanghai 200240 (China); Institute of Fusion and Reactor Technology, Karlsruhe Institute of Technology, Vincenz-Priessnitz-Str. 3, 76131 Karlsruhe (Germany); Yang Yanhua [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dongchuan Road, Shanghai 200240 (China); Cheng Xu [Institute of Fusion and Reactor Technology, Karlsruhe Institute of Technology, Vincenz-Priessnitz-Str. 3, 76131 Karlsruhe (Germany)

2012-09-15

Highlights: Black-Right-Pointing-Pointer Modification of system code ATHLET for supercritical water application. Black-Right-Pointing-Pointer Development and assessment of a heat transfer package for supercritical water. Black-Right-Pointing-Pointer Validation of the modified code at supercritical pressures with the theoretical point-hydraulics model and the SASC code. Black-Right-Pointing-Pointer Application of the modified code to LOCA analysis of a supercritical water cooled in-pile fuel qualification test loop. - Abstract: Since the existing thermal-hydraulic computer codes for light water reactors are not applicable to supercritical water cooled reactors (SCWRs) owing to the limitation of physical models and numerical treatments, the development of a reliable thermal-hydraulic computer code is very important to design analysis and safety assessment of SCWRs. Based on earlier modification of ATHLET for SCWR, a general interface is implemented to the code, which serves as the platform for information exchange between ATHLET and the external independent physical modules. A heat transfer package containing five correlations for supercritical water is connected to the ATHLET code through the interface. The correlations are assessed with experimental data. To verify the modified ATHLET code, the Edwards-O'Brian blow-down test is simulated. As first validation at supercritical pressures, a simplified supercritical water cooled loop is modeled and its stability behavior is analyzed. Results are compared with that of the theoretical model and SASC code in the reference and show good agreement. To evaluate its feasibility, the modified ATHLET code is applied to a supercritical water cooled in-pile fuel qualification test loop. Loss of coolant accidents (LOCAs) due to break of coolant supply lines are calculated for the loop. Sensitivity analysis of some safety system parameters is performed to get further knowledge about their influence on the function of the

Review and comparison of WWER and LWR Codes and Standards

International Nuclear Information System (INIS)

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

2003-01-01

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

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

International Nuclear Information System (INIS)

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

2009-01-01

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

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

International Nuclear Information System (INIS)

Markova, L.

2001-01-01

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

Simulation of water hammer phenomena using the system code ATHLET

Energy Technology Data Exchange (ETDEWEB)

Bratfisch, Christoph; Koch, Marco K. [Bochum Univ. (Germany). Reactor Simulation and Safety Group

2017-07-15

Water Hammer Phenomena can endanger the integrity of structures leading to a possible failure of pipes in nuclear power plants as well as in many industrial applications. These phenomena can arise in nuclear power plants in the course of transients and accidents induced by the start-up of auxiliary feed water systems or emergency core cooling systems in combination with rapid acting valves and pumps. To contribute to further development and validation of the code ATHLET (Analysis of Thermalhydraulics of Leaks and Transients), an experiment performed in the test facility Pilot Plant Pipework (PPP) at Fraunhofer UMSICHT is simulated using the code version ATHLET 3.0A.

Simulation of water hammer phenomena using the system code ATHLET

International Nuclear Information System (INIS)

Bratfisch, Christoph; Koch, Marco K.

2017-01-01

Water Hammer Phenomena can endanger the integrity of structures leading to a possible failure of pipes in nuclear power plants as well as in many industrial applications. These phenomena can arise in nuclear power plants in the course of transients and accidents induced by the start-up of auxiliary feed water systems or emergency core cooling systems in combination with rapid acting valves and pumps. To contribute to further development and validation of the code ATHLET (Analysis of Thermalhydraulics of Leaks and Transients), an experiment performed in the test facility Pilot Plant Pipework (PPP) at Fraunhofer UMSICHT is simulated using the code version ATHLET 3.0A.

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

International Nuclear Information System (INIS)

Masopust, R.

1999-01-01

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

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

International Nuclear Information System (INIS)

Manolova, M.

1998-01-01

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

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

International Nuclear Information System (INIS)

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

2000-01-01

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

Validation and application of the system code ATHLET-CD for BWR severe accident analyses

Energy Technology Data Exchange (ETDEWEB)

Di Marcello, Valentino, E-mail: valentino.marcello@kit.edu; Imke, Uwe; Sanchez, Victor

2016-10-15

Highlights: • We present the application of the system code ATHLET-CD code for BWR safety analyses. • Validation of core in-vessel models is performed based on KIT CORA experiments. • A SB-LOCA scenario is simulated on a generic German BWR plant up to vessel failure. • Different core reflooding possibilities are investigated to mitigate the accident consequences. • ATHLET-CD modelling features reflect the current state of the art of severe accident codes. - Abstract: This paper is aimed at the validation and application of the system code ATHLET-CD for the simulation of severe accident phenomena in Boiling Water Reactors (BWR). The corresponding models for core degradation behaviour e.g., oxidation, melting and relocation of core structural components are validated against experimental data available from the CORA-16 and -17 bundle tests. Model weaknesses are discussed along with needs for further code improvements. With the validated ATHLET-CD code, calculations are performed to assess the code capabilities for the prediction of in-vessel late phase core behaviour and reflooding of damaged fuel rods. For this purpose, a small break LOCA scenario for a generic German BWR with postulated multiple failures of the safety systems was selected. In the analysis, accident management measures represented by cold water injection into the damaged reactor core are addressed to investigate the efficacy in avoiding or delaying the failure of the reactor pressure vessel. Results show that ATHLET-CD is applicable to the description of BWR plant behaviour with reliable physical models and numerical methods adopted for the description of key in-vessel phenomena.

Coupling of 3D neutronics models with the system code ATHLET

International Nuclear Information System (INIS)

Langenbuch, S.; Velkov, K.

1999-01-01

The system code ATHLET for plant transient and accident analysis has been coupled with 3D neutronics models, like QUABOX/CUBBOX, for the realistic evaluation of some specific safety problems under discussion. The considerations for the coupling approach and its realization are discussed. The specific features of the coupled code system established are explained and experience from first applications is presented. (author)

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

International Nuclear Information System (INIS)

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

1997-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-08-01

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

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

International Nuclear Information System (INIS)

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

1985-01-01

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

Test model of WWER core

International Nuclear Information System (INIS)

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

2007-01-01

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

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

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

International Nuclear Information System (INIS)

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

2009-01-01

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

An approach to WWER fuels with BaCo

International Nuclear Information System (INIS)

Marino, A.; Demarco, G.

2008-01-01

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

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

International Nuclear Information System (INIS)

Zalesky, K.; Markova, L.

1999-12-01

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

SCORPIO - WWER core surveillance system

International Nuclear Information System (INIS)

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

1998-01-01

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

Error analysis of supercritical water correlations using ATHLET system code under DHT conditions

Energy Technology Data Exchange (ETDEWEB)

Samuel, J., E-mail: jeffrey.samuel@uoit.ca [Univ. of Ontario Inst. of Tech., Oshawa, ON (Canada)

2014-07-01

The thermal-hydraulic computer code ATHLET (Analysis of THermal-hydraulics of LEaks and Transients) is used for analysis of anticipated and abnormal plant transients, including safety analysis of Light Water Reactors (LWRs) and Russian Graphite-Moderated High Power Channel-type Reactors (RBMKs). The range of applicability of ATHLET has been extended to supercritical water by updating the fluid-and transport-properties packages, thus enabling the code to the used in analysis of SuperCritical Water-cooled Reactors (SCWRs). Several well-known heat-transfer correlations for supercritical fluids were added to the ATHLET code and a numerical model was created to represent an experimental test section. In this work, the error in the Heat Transfer Coefficient (HTC) calculation by the ATHLET model is studied along with the ability of the various correlations to predict different heat transfer regimes. (author)

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

International Nuclear Information System (INIS)

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

2006-01-01

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

Assessment of the system code DRUFAN/ATHLET using results of LOBI tests

International Nuclear Information System (INIS)

Burwell, J.M.; Kirmse, R.E.; Kyncl, M.; Malhotra, P.K.

1989-09-01

Four post-test analyses have been performed by GRS within the Shared Cost Action Programme (SCAP) sponsored by the Commission of the European Communities (contract 3015-86-07 EL ISP D) and by the Bundesminister fuer Forschung und Technologie of the Federal Republic of Germany (Research project RS 739). The four tests were mutually selected by the contractors (CEA, GRS, IKE, Univ. Pisa) of activity No. 3 and by the project organizer. Some of the tests were selected to be analyzed by more than one participant in order to allow comparison between analytical results obtained with different codes or obtained by different code-users. DRUFAN/ATHLET verification analyses were performed by IKE too. The four tests selected for the GRS activity are: - A2-77A (Natural Circulation Test), Analysis with ATHLET - A1-76 (Steam Generator Performance Test), Analysis with DRUFAN - BL-01 (Intermediate Leak), Analysis with ATHLET - A2-81 (Small Leak), Analysis with ATHLET. This final report contains the results of the four post test analysis including the comparison between measured and calculated quantities and the description of the applied codes, the selected model of the LOBI facility and the conclusions drawn for the improvement of the codes models

Validation of thermal hydraulic computer codes for advanced light water reactor

International Nuclear Information System (INIS)

Macek, J.

2001-01-01

The Czech Republic operates 4 WWER-440 units, two WWER-1000 units are being finalised (one of them is undergoing commissioning). Thermal-hydraulics Department of the Nuclear Research Institute Rez performs accident analyses for these plants using a number of computer codes. To model the primary and secondary circuits behaviour the system codes ATHLET, CATHARE, RELAP, TRAC are applied. Containment and pressure-suppressure system are modelled with RALOC and MELCOR codes, the reactor power calculations (point and space-neutron kinetics) are made with DYN3D, NESTLE and CDF codes (FLUENT, TRIO) are used for some specific problems. An integral part of the current Czech project 'New Energy Sources' is selection of a new nuclear source. Within this and the preceding projects financed by the Czech Ministry of Industry and Trade and the EU PHARE, the Department carries and has carried out the systematic validation of thermal-hydraulic and reactor physics computer codes applying data obtained on several experimental facilities as well as the real operational data. The paper provides a concise information on these activities of the NRI and its Thermal-hydraulics Department. A detailed example of the system code validation and the consequent utilisation of the results for a real NPP purposes is included. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-09-15

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

Further development of the computer code ATHLET-CD

International Nuclear Information System (INIS)

Weber, Sebastian; Austregesilo, Henrique; Bals, Christine; Band, Sebastian; Hollands, Thorsten; Koellein, Carsten; Lovasz, Liviusz; Pandazis, Peter; Schubert, Johann-Dietrich; Sonnenkalb, Martin

2016-10-01

In the framework of the reactor safety research program sponsored by the German Federal Ministry for Economic Affairs and Energy (BMWi), the computer code system ATHLET/ATHLET-CD has been further developed as an analysis tool for the simulation of accidents in nuclear power plants with pressurized and boiling water reactors as well as for the evaluation of accident management procedures. The main objective was to provide a mechanistic analysis tool for best estimate calculations of transients, accidents, and severe accidents with core degradation in light water reactors. With the continued development, the capability of the code system has been largely improved, allowing best estimate calculations of design and beyond design base accidents, and the simulation of advanced core degradation with enhanced model extent in a reasonable calculation time. ATHLET comprises inter alia a 6-equation model, models for the simulation of non-condensable gases and tracking of boron concentration, as well as additional component and process models for the complete system simulation. Among numerous model improvements, the code application has been extended to super critical pressures. The mechanistic description of the dynamic development of flow regimes on the basis of a transport equation for the interface area has been further developed. This ATHLET version is completely integrated in ATHLET-CD. ATHLET-CD further comprises dedicated models for the simulation of fuel and control assembly degradation for both pressurized and boiling water reactors, debris bed with melting in the core region, as well as fission product and aerosol release and transport in the cooling system, inclusive of decay of nuclide inventories and of chemical reactions in the gas phase. The continued development also concerned the modelling of absorber material release, of melting, melt relocation and freezing, and the interaction with the wall of the reactor pressure vessel. The following models were newly

Further development of the computer code ATHLET-CD; Weiterentwicklung des Rechenprogramms ATHLET-CD. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Weber, Sebastian; Austregesilo, Henrique; Bals, Christine; Band, Sebastian; Hollands, Thorsten; Koellein, Carsten; Lovasz, Liviusz; Pandazis, Peter; Schubert, Johann-Dietrich; Sonnenkalb, Martin

2016-10-15

In the framework of the reactor safety research program sponsored by the German Federal Ministry for Economic Affairs and Energy (BMWi), the computer code system ATHLET/ATHLET-CD has been further developed as an analysis tool for the simulation of accidents in nuclear power plants with pressurized and boiling water reactors as well as for the evaluation of accident management procedures. The main objective was to provide a mechanistic analysis tool for best estimate calculations of transients, accidents, and severe accidents with core degradation in light water reactors. With the continued development, the capability of the code system has been largely improved, allowing best estimate calculations of design and beyond design base accidents, and the simulation of advanced core degradation with enhanced model extent in a reasonable calculation time. ATHLET comprises inter alia a 6-equation model, models for the simulation of non-condensable gases and tracking of boron concentration, as well as additional component and process models for the complete system simulation. Among numerous model improvements, the code application has been extended to super critical pressures. The mechanistic description of the dynamic development of flow regimes on the basis of a transport equation for the interface area has been further developed. This ATHLET version is completely integrated in ATHLET-CD. ATHLET-CD further comprises dedicated models for the simulation of fuel and control assembly degradation for both pressurized and boiling water reactors, debris bed with melting in the core region, as well as fission product and aerosol release and transport in the cooling system, inclusive of decay of nuclide inventories and of chemical reactions in the gas phase. The continued development also concerned the modelling of absorber material release, of melting, melt relocation and freezing, and the interaction with the wall of the reactor pressure vessel. The following models were newly

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

International Nuclear Information System (INIS)

Andreeva, Marina; Totev, Totju; Stoyanov, Stoyan

2008-01-01

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

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

International Nuclear Information System (INIS)

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

2013-01-01

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

Validation of the thermal-hydraulic system code ATHLET based on selected pressure drop and void fraction BFBT tests

Energy Technology Data Exchange (ETDEWEB)

Di Marcello, Valentino, E-mail: valentino.marcello@kit.edu; Escalante, Javier Jimenez; Espinoza, Victor Sanchez

2015-07-15

Highlights: • Simulation of BFBT-BWR steady-state and transient tests with ATHLET. • Validation of thermal-hydraulic models based on pressure drops and void fraction measurements. • TRACE system code is used for the comparative study. • Predictions result in a good agreement with the experiments. • Discrepancies are smaller or comparable with respect to the measurements uncertainty. - Abstract: Validation and qualification of thermal-hydraulic system codes based on separate effect tests are essential for the reliability of numerical tools when applied to nuclear power plant analyses. To this purpose, the Institute for Neutron Physics and Reactor Technology (INR) at the Karlsruhe Institute of Technology (KIT) is involved in various validation and qualification activities of different CFD, sub-channel and system codes. In this paper, the capabilities of the thermal-hydraulic code ATHLET are assessed based on the experimental results provided within the NUPEC BFBT benchmark related to key Boiling Water Reactors (BWR) phenomena. Void fraction and pressure drops measurements in the BFBT bundle performed under steady-state and transient conditions which are representative for e.g. turbine trip and recirculation pump trip events, are compared with the numerical results of ATHLET. The comparison of code predictions with the BFBT data has shown good agreement given the experimental uncertainty and the results are consistent with the trends obtained with similar thermal-hydraulic codes.

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

International Nuclear Information System (INIS)

Klepach, J.

1995-01-01

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

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

International Nuclear Information System (INIS)

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

2000-01-01

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

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

International Nuclear Information System (INIS)

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

2006-01-01

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

RBMK-LOCA-Analyses with the ATHLET-Code

Energy Technology Data Exchange (ETDEWEB)

Petry, A. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH Kurfuerstendamm, Berlin (Germany); Domoradov, A.; Finjakin, A. [Research and Development Institute of Power Engineering, Moscow (Russian Federation)

1995-09-01

The scientific technical cooperation between Germany and Russia includes the area of adaptation of several German codes for the Russian-designed RBMK-reactor. One point of this cooperation is the adaptation of the Thermal-Hydraulic code ATHLET (Analyses of the Thermal-Hydraulics of LEaks and Transients), for RBMK-specific safety problems. This paper contains a short description of a RBMK-1000 reactor circuit. Furthermore, the main features of the thermal-hydraulic code ATHLET are presented. The main assumptions for the ATHLET-RBMK model are discussed. As an example for the application, the results of test calculations concerning a guillotine type rupture of a distribution group header are presented and discussed, and the general analysis conditions are described. A comparison with corresponding RELAP-calculations is given. This paper gives an overview on some problems posed and experience by application of Western best-estimate codes for RBMK-calculations.

Modelling of WWER-1000 fuel: state and prospects

International Nuclear Information System (INIS)

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

1994-01-01

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

Modelling of WWER-1000 fuel: state and prospects

Energy Technology Data Exchange (ETDEWEB)

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

1994-12-31

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

Application of modified version of SPPS-1 - HEXAB-2DB computer code package for operational analyses of fuel behaviour in WWER-440 reactors at Kozloduy NPP

Energy Technology Data Exchange (ETDEWEB)

Kharalampieva, Ts; Stoyanova, I; Antonov, A; Simeonov, T [Kombinat Atomna Energetika, Kozloduj (Bulgaria); Petkov, P [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika

1994-12-31

The modified version of SPPS-1 code called SPPS-1-HEXAB-2DB was applied for the purposes of the operational analysis and power peaking factors and reactor core critical parameters predictions of WWER-440s. The results of the calculations performed by the use of SPPS-1-HEXAB-2DB code and the corresponding parameters obtained from experiments at Kozloduy NPP WWER-440s as well as the results of fuel rod power distribution are presented. The method of operation simulation of reactor core with 349 assemblies (Unit 4) and with 313 fuel assemblies and 36 dummy fuel assemblies (Unit 1) is outlined. The modified code calculates not only fuel burnup and Pm-149 and Sm-149 concentrations distributions but also the space distribution of I-135 and Xe-135 concentrations. In this way it makes possible to perform the reactor operation simulation during the immediate periods after the reactor start-up or shut-down and to predict the critical reactor core parameters during transients. The results obtained show that SPPS-1-HEXAB-2DB code describes adequately the reactor core status. The new SPPS-1 code algorithm for estimation of assembly-wise power peaking factors distribution in reactor core is also described. The new code provides an option for checking the correctness of reactor core symmetry. The experience from the use of the modified SPPS-1-HEXAB-2DB code system confirms the provision of improved availability of operational analysis, prediction of Kozloduy NPP WWER-440s safe operations and fuel behaviour estimation. 14 tabs., 4 figs., 5 refs.

Systems for noise diagnostics of WWER nuclear power plants

International Nuclear Information System (INIS)

Por, G.

1996-01-01

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

Systems for noise diagnostics of WWER nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

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

1997-12-31

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

Post-test calculation and uncertainty analysis of the experiment QUENCH-07 with the system code ATHLET-CD

International Nuclear Information System (INIS)

Austregesilo, Henrique; Bals, Christine; Trambauer, Klaus

2007-01-01

In the frame of developmental assessment and code validation, a post-test calculation of the test QUENCH-07 was performed with ATHLET-CD. The system code ATHLET-CD is being developed for best-estimate simulation of accidents with core degradation and for evaluation of accident management procedures. It applies the detailed models of the thermal-hydraulic code ATHLET in an efficient coupling with dedicated models for core degradation and fission products behaviour. The first step of the work was the simulation of the test QUENCH-07 applying the modelling options recommended in the code User's Manual (reference calculation). The global results of this calculation showed a good agreement with the measured data. This calculation was complemented by a sensitivity analysis in order to investigate the influence of a combined variation of code input parameters on the simulation of the main phenomena observed experimentally. Results of this sensitivity analysis indicate that the main experimental measurements lay within the uncertainty range of the corresponding calculated values. Among the main contributors to the uncertainty of code results are the heat transfer coefficient due to forced convection to superheated steam-argon mixture, the thermal conductivity of the shroud isolation and the external heater rod resistance. Uncertainties on modelling of B 4 C oxidation do not affect significantly the total calculated hydrogen release rates

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

International Nuclear Information System (INIS)

Hordosy, G.; Maraczy, Cs.

2000-01-01

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

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

International Nuclear Information System (INIS)

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

2006-01-01

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

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

International Nuclear Information System (INIS)

Antov, A.; Stoyanova, I.

2008-01-01

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

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

International Nuclear Information System (INIS)

Stoyanova, I.; Antov, A.

2007-01-01

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

The database 'EDUD Base' for validation of neutron-physics codes used to analyze the WWER-440 cores

International Nuclear Information System (INIS)

Rocek, J.; Belac, J.; Miasnikov, A.

2003-01-01

The program and data system EDUDBase for validation of reactor computing codes was developed at NRI. It is designed for validation and evaluation of the precision of different computer codes used for WWER core analyses. The main goal of this database is to provide data for comparison with calculation results of tested codes and tools for statistical analysis or differences between the calculation results and the test data. The benchmark data sets are based on in-core measurements performed on WWER-440 reactors of Dukovany NPP. The initial data from NPP are verified, errors and inaccuracies are eliminated and data are transferred to a form, which is suitable for comparison with results of calculations. A special reduced operating history data set is created for each operating cycle ('Benchmark Operation History') to be used as an input data for calculation. It contains values of some integral quantities for each time point: effective time, integral thermal power, boron concentration, position of working group control assemblies (group 6) and inlet coolant temperature. At present, sets are available for all completed cycles up to: (unit/cycle) 1/17, 2/16, 3/15, 4/15. Power distribution is described for approx. 40 time steps during each operating cycle. 2D-power distributions are transferred into 60-degree core symmetry sector of reactor core. At present, such data sets are available only for later cycles starting with: (unit/cycle) 1/7, 2/6, 3/5, 4/5 (in other words last II cycles for each unit) (Authors)

WWER reactor fuel performance, modelling and experimental support. Proceedings

International Nuclear Information System (INIS)

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

1994-01-01

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

WWER reactor fuel performance, modelling and experimental support. Proceedings

Energy Technology Data Exchange (ETDEWEB)

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

1994-12-31

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

Validation and uncertainty analysis of the Athlet thermal-hydraulic computer code

International Nuclear Information System (INIS)

Glaeser, H.

1995-01-01

The computer code ATHLET is being developed by GRS as an advanced best-estimate code for the simulation of breaks and transients in Pressurized Water Reactor (PWRs) and Boiling Water Reactor (BWRs) including beyond design basis accidents. A systematic validation of ATHLET is based on a well balanced set of integral and separate effects tests emphasizing the German combined Emergency Core Cooling (ECC) injection system. When using best estimate codes for predictions of reactor plant states during assumed accidents, qualification of the uncertainty in these calculations is highly desirable. A method for uncertainty and sensitivity evaluation has been developed by GRS where the computational effort is independent of the number of uncertain parameters. (author)

WWER safety investigations on LR-0 reactor

International Nuclear Information System (INIS)

Mikus, J.

2001-01-01

A set of the measurement needed for the WWER-440 and WWER-1000 reactor lifetime assessment, verification of the methods, codes and input cross section libraries for the WWER reactor pressure vessel exposure evaluation has been performed on the LR-0 experimental reactor. The WWER Mock-ups (engineering benchmarks) has been carried out on the reactor, with the aim to investigate differential neutron spectra for reactor dosimetry purposes. Critical experiments have also been performed to determine the perturbation of the fission density distribution caused by the WWER-440 control assembly. Such assembly, partially inserted in the core, has significant influence on the space power distribution. A wide research program for sub-criticality investigations of the spent nuclear fuel storage has been realized on the LR-0 reactor. A benchmark experiment is realized on the reactor in corresponding geometry for CASTOR 440/84 container for storage and transportation of spent fuel. Critical experiments with new fuel assemblies including various burnable absorbers and different enrichments are performed. A set of critical experiments is performed using the fuel assemblies with 3,6% and 4,4% enrichment, arranged in the WWER-440 type cores with various lattice pitch. The critical high of the moderator level and the moderator level coefficient of reactivity are measured and the effect of the fuel assembly, placed in a hexagonal tube of stainless steel containing boron absorber (ATABOR - STANDARD) is investigated. The obtained results are used for the validation of the codes (MCNP, KENO and SCALE) in the frame of the contract 'Burn-up credit implementation for the storage and transport containers of the spent fuel'. Combined neutron-gamma spectra measurements in the WWER-1000 Mock-up are carried out during 2001

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

International Nuclear Information System (INIS)

Adamik, V.; Tkach, A.

1982-01-01

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

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

International Nuclear Information System (INIS)

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

2007-01-01

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

Peculiarity by Modeling of the Control Rod Movement by the Kalinin-3 Benchmark

International Nuclear Information System (INIS)

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

2010-01-01

The paper presents an important part of the results of the OECD/NEA benchmark transient 'Switching off one main circulation pump at nominal power' analyzed as a boundary condition problem by the coupled system code ATHLET-BIPR-VVER. Some observations and comparisons with measured data for integral reactor parameters are discussed. Special attention is paid on the modeling and comparisons performed for the control rod movement and the reactor power history. (Authors)

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

International Nuclear Information System (INIS)

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

2006-01-01

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

Thermophysical aspects of WWER safety

International Nuclear Information System (INIS)

Kolochko, Vladimir N.

1999-01-01

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

System code improvements for modelling passive safety systems and their validation

Energy Technology Data Exchange (ETDEWEB)

Buchholz, Sebastian; Cron, Daniel von der; Schaffrath, Andreas [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany)

2016-11-15

GRS has been developing the system code ATHLET over many years. Because ATHLET, among other codes, is widely used in nuclear licensing and supervisory procedures, it has to represent the current state of science and technology. New reactor concepts such as Generation III+ and IV reactors and SMR are using passive safety systems intensively. The simulation of passive safety systems with the GRS system code ATHLET is still a big challenge, because of non-defined operation points and self-setting operation conditions. Additionally, the driving forces of passive safety systems are smaller and uncertainties of parameters have a larger impact than for active systems. This paper addresses the code validation and qualification work of ATHLET on the example of slightly inclined horizontal heat exchangers, which are e. g. used as emergency condensers (e. g. in the KERENA and the CAREM) or as heat exchanger in the passive auxiliary feed water systems (PAFS) of the APR+.

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-31

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

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

International Nuclear Information System (INIS)

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

1995-01-01

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

Interface requirements to couple thermal-hydraulic codes to severe accident codes: ATHLET-CD

Energy Technology Data Exchange (ETDEWEB)

Trambauer, K. [GRS, Garching (Germany)

1997-07-01

The system code ATHLET-CD is being developed by GRS in cooperation with IKE and IPSN. Its field of application comprises the whole spectrum of leaks and large breaks, as well as operational and abnormal transients for LWRs and VVERs. At present the analyses cover the in-vessel thermal-hydraulics, the early phases of core degradation, as well as fission products and aerosol release from the core and their transport in the Reactor Coolant System. The aim of the code development is to extend the simulation of core degradation up to failure of the reactor pressure vessel and to cover all physically reasonable accident sequences for western and eastern LWRs including RMBKs. The ATHLET-CD structure is highly modular in order to include a manifold spectrum of models and to offer an optimum basis for further development. The code consists of four general modules to describe the reactor coolant system thermal-hydraulics, the core degradation, the fission product core release, and fission product and aerosol transport. Each general module consists of some basic modules which correspond to the process to be simulated or to its specific purpose. Besides the code structure based on the physical modelling, the code follows four strictly separated steps during the course of a calculation: (1) input of structure, geometrical data, initial and boundary condition, (2) initialization of derived quantities, (3) steady state calculation or input of restart data, and (4) transient calculation. In this paper, the transient solution method is briefly presented and the coupling methods are discussed. Three aspects have to be considered for the coupling of different modules in one code system. First is the conservation of masses and energy in the different subsystems as there are fluid, structures, and fission products and aerosols. Second is the convergence of the numerical solution and stability of the calculation. The third aspect is related to the code performance, and running time.

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Modification and application of the ATHLET-SC code to trans-critical simulations

International Nuclear Information System (INIS)

Fu, S.-W.; Zhou, C.; Xu, Z.-H.; Liu, X.-J.; Yang, Y.-H.; Cheng, H.

2011-01-01

In the simulation of trans-critical transients of Supercritical water cooled reactor (SCWR), calculation will terminate because of the sudden change in void fraction across the critical point. To solve this problem, a pseudo two-phase method is proposed with a virtual region of latent heat at pseudo-critical temperatures. A smooth variation of void fraction can be realized by using liquid-field conservation equations at temperatures lower than the pseudo-critical temperature, and vapor-field conservation equations at temperatures higher than the pseudo-critical temperature. Using this method, the system code ATHLET is modified to ATHLET-SC mod 2 on the basic of the previous modified version ATHLET-SC by Shanghai Jiao Tong University. The results of tests are verified that the calculation error with the pseudo two-phase method for supercritical fluid is acceptable, when the virtual region of latent heat is kept small. Moreover, the ATHLET-SC mod 2 code is used to simulate the pressurization and depressurization process of a single flow channel with the pressure transition as well as blowdown process. The results indicate a good applicability of the modified code. (author)

Methodology, status and plans for development and assessment of the code ATHLET

Energy Technology Data Exchange (ETDEWEB)

Teschendorff, V.; Austregesilo, H.; Lerchl, G. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH Forschungsgelaende, Garching (Germany)

1997-07-01

The thermal-hydraulic computer code ATHLET (Analysis of THermal-hydraulics of LEaks and Transients) is being developed by the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) for the analysis of anticipated and abnormal plant transients, small and intermediate leaks as well as large breaks in light water reactors. The aim of the code development is to cover the whole spectrum of design basis and beyond design basis accidents (without core degradation) for PWRs and BWRs with only one code. The main code features are: advanced thermal-hydraulics; modular code architecture; separation between physical models and numerical methods; pre- and post-processing tools; portability. The code has features that are of special interest for applications to small leaks and transients with accident management, e.g. initialization by a steady-state calculation, full-range drift-flux model, dynamic mixture level tracking. The General Control Simulation Module of ATHLET is a flexible tool for the simulation of the balance-of-plant and control systems including the various operator actions in the course of accident sequences with AM measures. The code development is accompained by a systematic and comprehensive validation program. A large number of integral experiments and separate effect tests, including the major International Standard Problems, have been calculated by GRS and by independent organizations. The ATHLET validation matrix is a well balanced set of integral and separate effects tests derived from the CSNI proposal emphasizing, however, the German combined ECC injection system which was investigated in the UPTF, PKL and LOBI test facilities.

Methodology, status and plans for development and assessment of the code ATHLET

International Nuclear Information System (INIS)

Teschendorff, V.; Austregesilo, H.; Lerchl, G.

1997-01-01

The thermal-hydraulic computer code ATHLET (Analysis of THermal-hydraulics of LEaks and Transients) is being developed by the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) for the analysis of anticipated and abnormal plant transients, small and intermediate leaks as well as large breaks in light water reactors. The aim of the code development is to cover the whole spectrum of design basis and beyond design basis accidents (without core degradation) for PWRs and BWRs with only one code. The main code features are: advanced thermal-hydraulics; modular code architecture; separation between physical models and numerical methods; pre- and post-processing tools; portability. The code has features that are of special interest for applications to small leaks and transients with accident management, e.g. initialization by a steady-state calculation, full-range drift-flux model, dynamic mixture level tracking. The General Control Simulation Module of ATHLET is a flexible tool for the simulation of the balance-of-plant and control systems including the various operator actions in the course of accident sequences with AM measures. The code development is accompained by a systematic and comprehensive validation program. A large number of integral experiments and separate effect tests, including the major International Standard Problems, have been calculated by GRS and by independent organizations. The ATHLET validation matrix is a well balanced set of integral and separate effects tests derived from the CSNI proposal emphasizing, however, the German combined ECC injection system which was investigated in the UPTF, PKL and LOBI test facilities

Innovation of genetic algorithm code GenA for WWER fuel loading optimization

International Nuclear Information System (INIS)

Sustek, J.

2005-01-01

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

Thermal-Hydraulic Analysis of SWAMUP Facility Using ATHLET-SC Code

Energy Technology Data Exchange (ETDEWEB)

Wang, Zidi; Cao, Zhen; Liu, Xiaojing, E-mail: xiaojingliu@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Shanghai (China)

2015-03-16

During the loss of coolant accident (LOCA) of supercritical water-cooled reactor (SCWR), the pressure in the reactor system will undergo a rapid decrease from the supercritical pressure to the subcritical condition. This process is called trans-critical transients, which is of crucial importance for the LOCA analysis of SCWR. In order to simulate the trans-critical transient, a number of system codes for SCWR have been developed up to date. However, the validation work for the trans-critical models in these codes is still missing. The test facility Supercritical WAter MUltiPurpose loop (SWAMUP) with 2 × 2 rod bundle in Shanghai Jiao Tong University (SJTU) will be applied to provide test data for code validation. Some pre-test calculations are important and necessary to show the feasibility of the experiment. In this study, trans-critical transient analysis is performed for the SWAMUP facility with the system code ATHLET-SC, which is modified in SJTU, for supercritical water system. This paper presents the system behavior, e.g., system pressure, coolant mass flow, cladding temperature during the depressurization. The effects of some important parameters such as heating power, depressurization rate on the system characteristics are also investigated in this paper. Additionally, some sensitivities study of the code models, e.g., heat transfer coefficient, critical heat flux correlation are analyzed and discussed. The results indicate that the revised system code ATHLET-SC is capable of simulating thermal-hydraulic behavior during the trans-critical transient. According to the results, the cladding temperature during the transient is kept at a low value. However, the pressure difference of the heat exchanger after depressurization could reach 6 MPa, which should be considered in the experiment.

State of the art of computer codes and experimental investigations on safety of nuclear power plants with reactors of WWER type

International Nuclear Information System (INIS)

Asmolov, V.G.; Volkov, G.A.; Elikin, I.V.; Mysenkov, A.I.

1987-01-01

As is well-known investigations on mathematical models of accidental situations from the point of view of nuclear safety as well as their experimental investigation are of great importance in the design of reactor units for nuclear power plants. This paper gives a review of the state of the art of thermodynamic models and computer codes used for safety analysis of WWER reactors in the USSR, the experimental basis and experimental investigations of the appropriate thermal processes. The actual and future trends of theoretical and experimental investigation on safety problems of WWER type nuclear power plants are briefly described. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2001-07-01

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

Nuclear fuel utilization at the Kozloduy NPP

Energy Technology Data Exchange (ETDEWEB)

Boyadzhiev, Z [Kombinat Atomna Energetika, Kozloduj (Bulgaria); Kharalampieva, Ts; Pejchinov, Ts

1996-12-31

Data on core loading and operation conditions during past fuel cycles of the Units 1 to 6 at the Kozloduy NPP are presented. The Units 1 and 2 have reached average discharge fuel burn-up of 31 MW d/kg U, the Unit 3 - 34 MW d/kg U and the Unit 4 - 36.5 MW d/kg U. By use of dummies and low-leakage core loading patterns for WWER-440 cores an effective reduction in reactor pressure vessel irradiation is obtained. By increasing the enrichment level and improving the characteristics of the Units 3 and 4, a design fuel cycles duration has been reduced by 5-10% in number of assemblies. Core loading design has been modelled using computer codes SPPS-1, BIPR-7, ALBOM, PROROC. A 3-year fuel cycle utilizing 4.4% enriched fuel proved to be more efficient for WWER-1000 by 15% reduction in fuel cost compared to the 2-year cycle. Future developments include improvements of the in-core monitoring system and process on-line simulation based on more accurate computer codes. 7 refs., 6 figs., 10 tabs.

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Experience with the WIMS computer code at Skoda Plzen

International Nuclear Information System (INIS)

Vacek, J.; Mikolas, P.

1991-01-01

Validation of the program for neutronics analysis is described. Computational results are compared with results of experiments on critical assemblies and with results of other codes for different types of lattices. Included are the results for lattices containing Gd as burnable absorber. With minor exceptions, the results of benchmarking were quite satisfactory and justified the inclusion of WIMS in the production system of codes for WWER analysis. The first practical application was the adjustment of the WWER-440 few-group diffusion constants library of the three-dimensional diffusion code MOBY-DICK, which led to a remarkable improvement of results for operational states. Then a new library for the analysis of WWER-440 start-up was generated and tested and at present a new library for the analysis of WWER-440 operational states is being tested. Preparation of the library for WWER-1000 is in progress. (author). 19 refs

Transmutation potential of reactor WWER-440

International Nuclear Information System (INIS)

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

2001-01-01

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

Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 2B. General material: codes, standards, criteria. Working material

International Nuclear Information System (INIS)

1999-01-01

related to different structures or systems of Paks NPP and Kozloduy NPP Unit 5/6 and experience data (previous laboratory results or earthquake experience) for WWER type NPPs. The reports from the CRP participants have been compiled in volumes of IAEA Working Material grouped under five broad categories. This volume covers the Generic material (codes, standards, criteria)

Analysis of a large break LOCA in the cold leg of the WWER-440/W-213 plant Griefswald, Unit 5

International Nuclear Information System (INIS)

Horche, W.

1993-01-01

The Gessellschaft fur Anlagen und Reaktorsicherheit (GRS) has performed a safety evaluation of the nuclear power plant (NPP) Greifswald, unit 5, of the Soviet type WWER-440/W-213, in cooperation with the French Institute de Protection of de Surete Nucleaire (IPSN) and other partners. Within this project an independent accident analysis is performed by GRS in order to assess the results of existing analysis and to supplement them. In this paper the analysis of the double-ended guillotine break (DEGB) of one cold leg of the main circulation pipe is described. The major objective of the calculation was the investigation of the accident sequence with reduced availability of the emergency core cooling system (single failure criterion). In addition, the simultaneous loss of onsite and offsite power and the failure of scram were assumed. The thermal-hydraulic system code ATHLET/FLUT, developed at GRS and already applied for the safety analysis of several WWER plants, was chosen again. The pressure in the confinement, the back pressure for the discharge model, was calculated as a function of time for this accident separately with GRS-Code RALOC. Furthermore, it was necessary to model the local concentration of direct accumulator injection into the reactor vessel with the help of a special two-channel model of the core and upper plenum. For this model, results were considered obtained from the 1:1 scaled test facility UPTF. It was assumed that only 25% of the upper plenum and core volume is directly penetrated by the injected water. The DEGB was defined in that loop, which is connected with one of three low-pressure injection subsystems. This means that this injected water flows towards the leak without passing the core. As single failure the failure of one of three diesel generators was assumed. The full paper will contain nodalization schemes, which are generated by the ATHLET-Input-Grafic

Comparison of the APA-H (Westinghouse) calculations with the operational data for ZpNPP unit 3 cycles 16-19

International Nuclear Information System (INIS)

Abdullayev, A. M.; Gorbachenko, O. V.; Ignatchenko, A.I.; Maryokhin, S.V.; Zhukov, A. I.

2007-01-01

The computer simulation of ZpNPP Unit 3 (WWER-1000) Cycles 16-19 core depletion has been performed on the basis of the operational data. The changes in reactor heat rate, lead bank position and inlet temperature during the core operation have been taken into account. These calculations were performed by using Westinghouse APA-H (ALPHA/PHOENIX/ ANC-H) code system. The main objectives of the calculations were the comparison with operational data for core loading with TVS-M (Cycle 16) and transition core loading with TVS-A (Cycle 17-19). The calculation results were compared with the results of Critical boric acid concentration vs. Cycle Burnup measurements and Start up Physics Test measurements (at HZP, BOC, NoXe core conditions). Additionally, the comparison between the results of assemblies power calculation performed by ANC-H and BIPR-7A codes is presented (Authors)

Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 2. Generic material: Codes, standards, criteria. Working material

International Nuclear Information System (INIS)

1995-01-01

The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material contains reports related to generic material, namely codes, standards and criteria for benchmark analysis

Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 2. Generic material: Codes, standards, criteria. Working material

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-07-01

The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material contains reports related to generic material, namely codes, standards and criteria for benchmark analysis.

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Verification of the transuranus-WWER code version V1M2J00 by SOFIT and Kola-3 data bases: Part 2

International Nuclear Information System (INIS)

Elenkov, D.; Boneva, S.; Djourelov, N.

2002-01-01

The object of this study is standard WWER-440 fuel as described in the IFPE-OECD/IAEA-NEA database. The SOFIT and Kola-3 data base have been used the purpose, each one described in a separate paper, referred to as Part I (concerning SOFIT) and Part II (concerning Kola-3). It has been shown that the latest modifications of the programme lead to considerable improvements of the predictions on the fuel behaviour. Concerning the TRANSURANUS-WWER code calculations of different fuel behaviour characteristics of Kola-3 standard fuel rods, the following conclusions can be made: the burnup calculations are in excellent agreement with the measured values; the fission gas release calculations accounting for the high burnup structure influence are in very good agreement with the experiment for the two assemblies in the burnup region up to 40 MWd/kgU; the size changes for the cladding are reproduced in fairly good way; the gas pressure is reproduced by some 10 -15 % underestimation. From the obtained results including the discussed deviations, the present version of TRANSURANUS-WWER can be considered as applicable for the modelling fuel performance in Bulgarian nuclear power plants operating WWER reactors

Reliability of operating WWER monitoring systems

International Nuclear Information System (INIS)

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

1996-01-01

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

Reliability of operating WWER monitoring systems

Energy Technology Data Exchange (ETDEWEB)

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

1997-12-31

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

VVER-1000 main steam line break analysis using the coupled code system DYN3D/ATHLET

International Nuclear Information System (INIS)

Kliem, Soeren; Hoehne, Thomas; Rohde, Ulrich; Weiss, Frank-Peter; Kozmenkov, Yaroslav

2008-01-01

Calculations using the coupled code system DYN3D/ATHLET were performed in the frame of the OECD/NEA MSLB benchmark for a VVER-1000 reactor. The coolant mixing inside the reactor pressure vessel was treated using a validated empirical mixing model implemented into the DYN3D/ATHLET code. Using very conservative boundary conditions (reduced scram worth, two stuck rods, running MCP throughout the whole transient) a return-to-power was predicted. For the assessment of the empirical mixing model a time dependent calculation using the computational fluid dynamics code CFX-10 was performed. For that analysis, a detailed model of the reactor pressure vessel consisting of the inlets nozzles, downcomer, lower plenum and a part of the core and having 4.67 million unstructured tetra cell elements was used. For the considered case with running main coolant pumps, this calculation shows a sector formation at the core inlet with a certain amount of mixing at the edges of the sector. A core calculation using these CFX results as boundary conditions predicted also a return-to-power with a maximum value being about 200 MW lower than in the coupled code calculation. This variation calculation confirms the applicability of the empirical mixing model. The comparison shows also, that in this way results with a reasonable degree of conservatism can be obtained. (authors)

Criticality safety and shielding analysis of WWER-440 fuel configurations

International Nuclear Information System (INIS)

Christoskov, I.

2008-01-01

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

Monte-Carlo code calculation of 3D reactor core model with usage of burnt fuel isotopic compositions, obtained by engineering codes

Energy Technology Data Exchange (ETDEWEB)

Aleshin, Sergey S.; Gorodkov, Sergey S.; Shcherenko, Anna I. [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation)

2016-09-15

A burn-up calculation of large systems by Monte-Carlo code (MCU) is complex process and it requires large computational costs. Previously prepared isotopic compositions are proposed to be used for the Monte-Carlo code calculations of different system states with burnt fuel. Isotopic compositions are calculated by an approximation method. The approximation method is based on usage of a spectral functionality and reference isotopic compositions, that are calculated by the engineering codes (TVS-M, BIPR-7A and PERMAK-A). The multiplication factors and power distributions of FAs from a 3-D reactor core are calculated in this work by the Monte-Carlo code MCU using earlier prepared isotopic compositions. The separate conditions of the burnt core are observed. The results of MCU calculations were compared with those that were obtained by engineering codes.

Software in support of fuel operation in WWERS

International Nuclear Information System (INIS)

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

2013-01-01

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

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)

The time-dependent 3D discrete ordinates code TORT-TD with thermal-hydraulic feedback by ATHLET models

International Nuclear Information System (INIS)

Seubert, A.; Velkov, K.; Langenbuch, S.

2008-01-01

This paper describes the time-dependent 3D discrete ordinates transport code TORT-TD. Thermal-hydraulic feedback is considered by coupling TORT-TD with the thermal-hydraulics system code ATHLET. The coupled code TORT-TD/ATHLET allows 3D pin-by-pin analyses of transients in few energy groups and anisotropic scattering by solving the time-dependent transport equation using the unconditionally stable implicit method. The nuclear cross sections are interpolated between pre-calculated table values of fuel temperature, moderator density and boron concentration. For verification of the implementation, selected test cases have been calculated by TORT-TD/ATHLET. They include a control rod ejection transient in a small PWR fuel assembly arrangement and a local boron concentration change in a single PWR fuel assembly. In the latter, special attention has been paid to study the influence of the thermal-hydraulic feedback modelling in ATHLET. The results obtained for a control rod ejection accident in a PWR quarter core demonstrate the applicability of TORT-TD/ATHLET. (authors)

Fuel rod behaviour at high burnup WWER fuel cycles

International Nuclear Information System (INIS)

Medvedev, A.; Bogatyr, S.; Kouznetsov, V.; Khvostov, G.; Lagovsky; Korystin, L.; Poudov, V.

2003-01-01

The modernisation of WWER fuel cycles is carried out on the base of complete modelling and experimental justification of fuel rods up to 70 MWd/kgU. The modelling justification of the reliability of fuel rod and fuel rod with gadolinium is carried out with the use of certified START-3 code. START-3 code has a continuous experimental support. The thermophysical and strength reliability of WWER-440 fuel is justified for fuel rod and pellet burnups 65 MWd/kgU and 74 MWd/U, accordingly. Results of analysis are demonstrated by the example of uranium-gadolinium fuel assemblies of second generation under 5-year cycle with a portion of 6-year assemblies and by the example of successfully completed pilot operation of 5-year cycle fuel assemblies during 6 years at unit 3 of Kolskaja NPP. The thermophysical and strength reliability of WWER-1000 fuel is justified for a fuel rod burnup 66 MWd/kgU by the example of fuel operation under 4-year cycles and 6-year test operation of fuel assemblies at unit 1 of Kalininskaya NPP. By the example of 5-year cycle at Dukovany NPP Unit 2 it was demonstrated that WWER fuel rod of a burnup 58 MWd/kgU ensure reliable operation under load following conditions. The analysis has confirmed sufficient reserves of Russian fuel to implement program of JSC 'TVEL' in order to improve technical and economical parameters of WWER fuel cycles

Validation of the ATHLET-code 2.1A by calculation of the ECTHOR experiment; Validierung des ATHLET-Codes 2.1A anhand des Einzeleffekt-Tests ECTHOR

Energy Technology Data Exchange (ETDEWEB)

Jung, Andreas; Sarkadi, Peter; Schaffrath, Andreas [TUEV NORD SysTec GmbH und Co. KG, Hamburg (Germany)

2010-05-15

Before a numerical code (e.g. ATHLET) is used for simulation of physical phenomena being new or unknown for the code and/or the user, the user ensures the applicability of the code and his own experience of handling with it by means of a so-called validation. Parametric studies with the code are executed for that matter and the results have to be compared with verified experimental data. Corresponding reference values are available in terms of so-called single-effect-tests (e.g. ECTHOR). In this work the system-code ATHLET Mod. 2.1 Cycle A is validated by post test calculation of the ECTHOR experiment due to the above named aspects. With the ECTHOR-tests the clearing of a water-filled model of a loop seal by means of an air-stream was investigated including momentum exchange at the phase interface under adiabatic and atmospheric conditions. The post test calculations show that the analytical results meet the experimental data within the reproducibility of the experiments. Further findings of the parametric studies are: - The experimental results obtained with the system water-air (ECTHOR) can be assigned to a water-steam-system, if the densities of the phases are equal in both cases. - The initial water level in the loop seal has no influence on the results as long as the gas mass flow is increased moderately. - The loop seal is appropriately nodalized if the mean length of the control volumes accords approx. 1.5 tim es the hydraulic pipe diameter. (orig.)

Validation of the ATHLET-code 2.1A by calculation of the ECTHOR experiment; Validierung des ATHLET-Codes 2.1A anhand des Einzeleffekt-Tests ECTHOR

Energy Technology Data Exchange (ETDEWEB)

Jung, Andreas; Sarkadi, Peter; Schaffrath, Andreas [TUEV NORD SysTec GmbH und Co. KG, Hamburg (Germany)

2010-06-15

Before a numerical code (e.g. ATHLET) is used for simulation of physical phenomena being new or unknown for the code and/or the user, the user ensures the applicability of the code and his own experience of handling with it by means of a so-called validation. Parametric studies with the code are executed for that matter und the results have to be compared with verified experimental data. Corresponding reference values are available in terms of so-called single-effect-tests (e.g. ECTHOR). In this work the system-code ATHLET Mod. 2.1 Cycle A is validated by post test calculation of the ECTHOR experiment due to the above named aspects. With the ECTHOR-tests the clearing of a water-filled model of a loop seal by means of an air-stream was investigated including momentum exchange at the phase interface under adiabatic and atmospheric conditions. The post test calculations show that the analytical results meet the experimental data within the reproducibility of the experiments. Further findings of the parametric studies are: - The experimental results obtained with the system water-air (ECTHOR) can be assigned to a water-steam-system, if the densities of the phases are equal in both cases. - The initial water level in the loop seal has no influence on the results as long as the gas mass flow is increased moderately. - The loop seal is appropriately nodalized if the mean length of the control volumes accords approx. 1.5 times the hydraulic pipe diameter. (orig.)

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

International Nuclear Information System (INIS)

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

2011-01-01

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

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)

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

International Nuclear Information System (INIS)

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

1996-01-01

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

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

International Nuclear Information System (INIS)

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

2008-01-01

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

Modelling and simulation of process control systems for WWER

Energy Technology Data Exchange (ETDEWEB)

Pangelov, N [Energoproekt, Sofia (Bulgaria)

1996-12-31

A dynamic modelling method for simulation of process control system is developed (method for identification). It is based on the least squares method and highly efficient linear uninterrupted differential equations. The method has the following advantages: there are no significant limitations in the type of input/output signals and in the length of data time series; identification at none zero initial condition is possible; on-line identification is possible; a high accuracy is observed in case of noise. On the basis of real experiments and data time series simulated with known computer codes it is possible to construct highly efficient models of different systems for solving the following problems: real time simulation with high accuracy for training purposes; estimation of immeasurable parameters important to safety; malfunction diagnostics based on plant dynamics; prediction of dynamic behaviour; control vector estimation in regime adviser. Two real applications of this method are described: in dynamic behaviour modelling for steam generator level, and in creating of a Process Control System Simulator (PCSS) based on KASKAD-2 for WWER-1000 units of the Kozloduy NPP. 6 refs., 8 figs.

The World Anti-Doping Code: can you have asthma and still be an elite athlete?

Directory of Open Access Journals (Sweden)

Ken Fitch

2016-06-01

The World Anti-Doping Code (the Code has not prevented asthmatic individuals from becoming elite athletes. This review examines those sections of the Code that are relevant to respiratory physicians who manage elite and sub-elite athletes with asthma. The restrictions that the Code places or may place on the prescription of drugs to prevent and treat asthma in athletes are discussed. In addition, the means by which respiratory physicians are able to treat their elite asthmatic athlete patients with drugs that are prohibited in sport are outlined, along with some of the pitfalls in such management and how best to prevent or minimise them.

Modelling horizontal steam generator with ATHLET. Verification of different nodalization schemes and implementation of verified constitutive equations

Energy Technology Data Exchange (ETDEWEB)

Beliaev, J.; Trunov, N.; Tschekin, I. [OKB Gidropress (Russian Federation); Luther, W. [GRS Garching (Germany); Spolitak, S. [RNC-KI (Russian Federation)

1995-12-31

Currently the ATHLET code is widely applied for modelling of several Power Plants of WWER type with horizontal steam generators. A main drawback of all these applications is the insufficient verification of the models for the steam generator. This paper presents the nodalization schemes for the secondary side of the steam generator, the results of stationary calculations, and preliminary comparisons to experimental data. The consideration of circulation in the water inventory of the secondary side is proved to be necessary. (orig.). 3 refs.

Modelling horizontal steam generator with ATHLET. Verification of different nodalization schemes and implementation of verified constitutive equations

Energy Technology Data Exchange (ETDEWEB)

Beliaev, J; Trunov, N; Tschekin, I [OKB Gidropress (Russian Federation); Luther, W [GRS Garching (Germany); Spolitak, S [RNC-KI (Russian Federation)

1996-12-31

Currently the ATHLET code is widely applied for modelling of several Power Plants of WWER type with horizontal steam generators. A main drawback of all these applications is the insufficient verification of the models for the steam generator. This paper presents the nodalization schemes for the secondary side of the steam generator, the results of stationary calculations, and preliminary comparisons to experimental data. The consideration of circulation in the water inventory of the secondary side is proved to be necessary. (orig.). 3 refs.

Surveillance of WWER-440 fuel performance

International Nuclear Information System (INIS)

Simko, J.; Urban, P.

1999-01-01

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

Modernization of WWER-1000 radiation monitoring systems

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-31

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

Modernization of WWER-1000 radiation monitoring systems

International Nuclear Information System (INIS)

Smith, T.

1995-01-01

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

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

International Nuclear Information System (INIS)

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

2005-01-01

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

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

International Nuclear Information System (INIS)

Trgina, M.

1993-12-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-12-31

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1998-12-31

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

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

Standard model of WWER-440 fuel rod for Transuranus and its application for RELAP5 hot channel validation

International Nuclear Information System (INIS)

Hatala, B.; Cvan, M.

2001-01-01

Within the PECO European Commission project of 'Extension of the validation matrix of the TRANSURANUS code' is developed a generic model of WWER-440 fuel rod. The model is intended to be applied for both realistic and licensing, conservative analysis. For such an application the TRANSURANUS code would be complementary tool to generally used system codes, e.g. RELAP5, providing realistic, more detailed insight into processes and safety criteria, relevant to the fuel rod. The paper presents general description of the model for TRANSURANUS code, brief discussion of approaches used in TRANSURANUS and RELAP5 code safety analysis, accompanied with information about RELAP5 model (whole scope unit model, used for licensing analysis). The existing model for RELAP5 code for WWER-440/V-213 Bohunice V2 unit is checked and modified in hot channel part to allow transparent comparison with the TRANSURANUS code. The results from comparison calculations of the both codes are presented for fresh fuel and quasi steady state scenario and are in good agreement, almost identical. These results might be used as a basis for transient analysis

Simulation of the fuel rod bundle test QUENCH-03 using the system codes ASTEC and ATHLET-CD

International Nuclear Information System (INIS)

Kruse, P.; Koch, M.K.

2011-01-01

The QUENCH-03 test was performed on the 21. of January 1999 at FZK (Forschungszentrum Karlsruhe) to investigate the behaviour on reflood of PWR (Pressurized Water Reactor) fuel rods with little oxidation. This paper presents the results of the simulation of QUENCH-03 performed with the version V1.3 of the integral code ASTEC (Accident Source Term Evaluation Code) which is being developed by IRSN (France) in cooperation with GRS (Germany) and with the program version 2.1A of the mechanistic code ATHLET-CD (Analysis of Thermal-hydraulics of Leaks and Transients - Core Degradation) which is under development by GRS. At first the QUENCH test facility and the QUENCH test program in general are described. The test conduct of the test QUENCH-03 follows as well as a description of the used codes ASTEC and ATHLET-CD with the associated modeling of the test section. The results of this calculation show that during the heat-up and transient phase both codes can calculate bundle and shroud temperatures as well as the hydrogen production in good approximation to the experimental data. During the quench phase and up to the end of the test only the oxidation model PRATER of ASTEC simulates the hydrogen production very well, the other oxidation models of ASTEC cannot calculate to some extent the measured amount of hydrogen. ATHLET-CD underestimates the integral amount at the end of the test. In the ASTEC calculations the temperatures during the quench phase show qualitatively good results, only time delays on some elevations of the bundle could be noticed. ATHLET-CD reproduces the thermal behaviour up to the first temperature escalation very well, after that the temperatures are partly over-estimated. The time delay recognized in the ASTEC calculations are seen as well. The results of the integral code ASTEC emphasize that the calculation of QUENCH-03 is possible and leading to good results concerning hydrogen release and corresponding temperatures. Because the QUENCH-03 test was

Armenian earthquake WWER-440 NNPs and Turkish early warning system

International Nuclear Information System (INIS)

Bektur, Y.

1991-01-01

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

Use of the Benchmarking System for Operational Waste from WWER Reactors

International Nuclear Information System (INIS)

2017-06-01

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

WWER steam generator tube structural and leakage integrity

International Nuclear Information System (INIS)

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

1998-01-01

-wall cracks have been experimentally confirmed. An expert system of WWER steam generators has been suggested with the main goal to predict the corrosion damage rate, the residual life-time and the life-time extension. The assessment is based on the evaluation of local environment parameters by MULTEQ Code combined with SCC experiments. (author)

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

International Nuclear Information System (INIS)

Por, G.

1992-06-01

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

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

International Nuclear Information System (INIS)

2005-06-01

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

Computer modelling of the WWER fuel elements under high burnup conditions by the computer codes PIN-W and RODQ2D

Energy Technology Data Exchange (ETDEWEB)

Valach, M; Zymak, J; Svoboda, R [Nuclear Research Inst. Rez plc, Rez (Czech Republic)

1997-08-01

This paper presents the development status of the computer codes for the WWER fuel elements thermomechanical behavior modelling under high burnup conditions at the Nuclear Research Institute Rez. The accent is given on the analysis of the results from the parametric calculations, performed by the programmes PIN-W and RODQ2D, rather than on their detailed theoretical description. Several new optional correlations for the UO2 thermal conductivity with degradation effect caused by burnup were implemented into the both codes. Examples of performed calculations document differences between previous and new versions of both programmes. Some recommendations for further development of the codes are given in conclusion. (author). 6 refs, 9 figs.

Computer modelling of the WWER fuel elements under high burnup conditions by the computer codes PIN-W and RODQ2D

International Nuclear Information System (INIS)

Valach, M.; Zymak, J.; Svoboda, R.

1997-01-01

This paper presents the development status of the computer codes for the WWER fuel elements thermomechanical behavior modelling under high burnup conditions at the Nuclear Research Institute Rez. The accent is given on the analysis of the results from the parametric calculations, performed by the programmes PIN-W and RODQ2D, rather than on their detailed theoretical description. Several new optional correlations for the UO2 thermal conductivity with degradation effect caused by burnup were implemented into the both codes. Examples of performed calculations document differences between previous and new versions of both programmes. Some recommendations for further development of the codes are given in conclusion. (author). 6 refs, 9 figs

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

International Nuclear Information System (INIS)

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

1982-01-01

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

WWER in-core fuel management benchmark definition

International Nuclear Information System (INIS)

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

1994-01-01

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

Power distribution gradients in WWER type cores and fuel failure root causes

Energy Technology Data Exchange (ETDEWEB)

Mikuš, Ján M., E-mail: JanMikus.nrc@hotmail.com

2014-02-15

Highlights: • Power (fission rate) distribution gradients can represent fuel failure root causes. • Positions with above gradients were investigated in WWER type cores on reactor LR-0. • Above gradients were evaluated near core heterogeneities and construction materials. • Results can be used for code validation and fuel failure occurrence investigation. - Abstract: Neutron flux non-uniformity and gradients of neutron current resulting in corresponding power (fission rate) distribution changes can represent root causes of the fuel failure. Such situation can be expected in vicinity of some core heterogeneities and construction materials. Since needed data cannot be obtained from nuclear power plant (NPP), results of some benchmark type experiments performed on light water, zero-power research reactor LR-0 were used for investigation of the above phenomenon. Attention was focused on determination of the spatial power distribution changes in fuel assemblies (FAs): Containing fuel rods (FRs) with Gd burnable absorber in WWER-440 and WWER-1000 type cores, Neighboring the core blanket and dummy steel assembly simulators on the periphery of the WWER-440 standard and low leakage type cores, resp., Neighboring baffle in WWER-1000 type cores, and Neighboring control rod (CR) in WWER-440 type cores, namely (a) power peak in axial power distribution in periphery FRs of the adjacent FAs near the area between CR fuel part and butt joint to the CR absorbing part and (b) decrease in radial power distribution in FRs near CR absorbing part. An overview of relevant experimental results from reactor LR-0 and some information concerning leaking FAs on NPP Temelín are presented. Obtained data can be used for code validation and subsequently for the fuel failure occurrence investigation.

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

International Nuclear Information System (INIS)

Mikus, J.

2003-01-01

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

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

International Nuclear Information System (INIS)

Pappx, L.

1994-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1995-12-31

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-31

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

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

International Nuclear Information System (INIS)

Papp, L.

1995-01-01

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

Review of provisions on corrosion fatigue and stress corrosion in WWER and Western LWR Codes and Standards

International Nuclear Information System (INIS)

Buckthorpe, D.; Filatov, V.; Tashkinov, A.; Evropin, S.V.; Matocha, K.; Guinovart, J.

2003-01-01

Results are presented from a collaborative project performed on behalf of the European Commission, Working Group Codes and Standards. The work covered the contents of current codes and standards, plant experience and R and D results. Current fatigue design rules use S-N curves based on tests in air. Although WWER and LWR design curves are often similar they are derived, presented and used in different ways and it is neither convenient nor appropriate to harmonise them. Similarly the fatigue crack growth laws used in the various design and in-service inspection rules differ significantly with respect to both growth rates in air and the effects of water reactor environments. Harmonised approaches to the effects of WWER and LWR environments are possible based on results from R and D programmes carried out over the last decade. For carbon and low alloy steels a consistent approach to both crack initiation and growth can be formulated based on the superposition of environmentally assisted cracking effects on the fatigue crack development. The approach indicates that effects of the water environment are minimal given appropriate control of the oxygen content of the water and/or the sulphur content of the steel. For austenitic stainless steels a different mechanisms may apply and a harmonised approach is possible at present only for S-N curves. Although substantial progress has been made with respect to corrosion fatigue, more data and a clearer understanding are required in order to write code provisions particularly in the area of high cycle fatigue. Reactor operation experience shows stress corrosion cracking of austenitic steels is the most common cause of failure. These failures are associated with high residual stresses combined with high levels of dissolved oxygen or the presence of contaminants. For primary circuit internals there is a potential threat to integrity from irradiated assisted stress corrosion cracking. Design and in-service inspection rules do not at

Teaching WWERs at Hacettepe University Nuclear Engineering Department in Turkey

International Nuclear Information System (INIS)

Ergun, S.

2011-01-01

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

Assessment of WWER fuel condition in design basis accident

International Nuclear Information System (INIS)

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

1994-01-01

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

Two-dimensional full-core transport theory Benchmarks for the WWER reactors

International Nuclear Information System (INIS)

Petkov, P.T.

2002-01-01

Several two-dimensional full-core real geometry many-group steady-state problems for the WWER-440 and WWER-1000 reactors have been solved by the MARIKO code, based on the method of characteristics. The reference transport theory solutions include assembly-wise and pin-wise power distributions. Homogenized two-group diffusion parameters and discontinuity factors have been calculated by MARIKO for each assembly type both for the whole assembly and for each cell in the smallest sector of symmetry, using the B1 method for calculation of the critical spectrum. Accurate albedo-type boundary conditions have been calculated by MARIKO for the core-reflector and core-absorber boundaries, both for each outer assembly face and for each outer cell face. Comparison with the reference solutions of the two-group nodal diffusion code SPPS-1.6 and the few-group fine-mesh diffusion codes HEX2DA and HEX2DB are presented (Authors)

Computer code SICHTA-85/MOD 1 for thermohydraulic and mechanical modelling of WWER fuel channel behaviour during LOCA and comparison with original version of the SICHTA code

International Nuclear Information System (INIS)

Bujan, A.; Adamik, V.; Misak, J.

1986-01-01

A brief description is presented of the expansion of the SICHTA-83 computer code for the analysis of the thermal history of the fuel channel for large LOCAs by modelling the mechanical behaviour of fuel element cladding. The new version of the code has a more detailed treatment of heat transfer in the fuel-cladding gap because it also respects the mechanical (plastic) deformations of the cladding and the fuel-cladding interaction (magnitude of contact pressure). Also respected is the change in pressure of the gas filling of the fuel element, the mechanical criterion is considered of a failure of the cladding and the degree is considered of the blockage of the through-flow cross section for coolant flow in the fuel channel. The LOCA WWER-440 model computation provides a comparison of the new SICHTA-85/MOD 1 code with the results of the original 83 version of SICHTA. (author)

The international WWER fuel market

International Nuclear Information System (INIS)

Gingold, G.E.; Goldstein, L.; Strasser, A.A.

1994-01-01

The state of the world nuclear fuel market and its economic complexities are described. Currently the nuclear fuel market is oversupplied and nuclear fuel fabrication in the West far exceeds the anticipated demands. Actually the current demand is not much more than half of the capacity available to supply it. The Eastern Europe (excluding the plants in the Russian Federation) with its 20 WWER-440 and 12 WWER-1000 reactors in operation and additional 4 WWER-440 and 8 WWER-1000 units under construction is considered as a potential long-term market for the Western fuel fabricators. The following significant benefits of competition in the WWER fuel market for the operators of these reactors are : 1) lower cost; 2) more favorable contract terms and improved vendor cooperation with the customer; 3) accelerated technological development. A brief description of the main WWER fuel suppliers TVEL, ABB Atom, BNFL, EVF and Westinghouse, as well as the status of some new companies as CEZ and SEP is given. The principal differences between Western and WWER fuels are outlined. The advanced features offered by the Western vendors and Russian fuel supply organisations are discussed. 2 tabs., 1 fig

The international WWER fuel market

Energy Technology Data Exchange (ETDEWEB)

Gingold, G E; Goldstein, L; Strasser, A A [Stoller (S.M.) Corp., Pleasantville, NY (United States)

1994-12-31

The state of the world nuclear fuel market and its economic complexities are described. Currently the nuclear fuel market is oversupplied and nuclear fuel fabrication in the West far exceeds the anticipated demands. Actually the current demand is not much more than half of the capacity available to supply it. The Eastern Europe (excluding the plants in the Russian Federation) with its 20 WWER-440 and 12 WWER-1000 reactors in operation and additional 4 WWER-440 and 8 WWER-1000 units under construction is considered as a potential long-term market for the Western fuel fabricators. The following significant benefits of competition in the WWER fuel market for the operators of these reactors are : (1) lower cost; (2) more favorable contract terms and improved vendor cooperation with the customer; (3) accelerated technological development. A brief description of the main WWER fuel suppliers TVEL, ABB Atom, BNFL, EVF and Westinghouse, as well as the status of some new companies as CEZ and SEP is given. The principal differences between Western and WWER fuels are outlined. The advanced features offered by the Western vendors and Russian fuel supply organisations are discussed. 2 tabs., 1 fig.

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

International Nuclear Information System (INIS)

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

2014-01-01

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

Probabilistic analysis of strength and thermal-physic WWER fuel rod characteristics using START-3 code

International Nuclear Information System (INIS)

Medvedev, A.; Bogatyr, S.; Khramtsov; Sokolov, F.

2001-01-01

During the last years probabilistic methods for evaluation of the influence of the fuel geometry and technology parameters on fuel operational reliability are widely used. In the present work the START-3 procedure is used to calculate the thermal physics and strength characteristics of WWER fuel rods behavior. The procedure is based on the Monte-Carlo method with the application of Sobol quasi-random sequences. This technique allows to treat the fuel rod technological and operating parameters as well as its strength and thermal physics characteristics as random variables. The work deals with a series of WWER-1000 fuel rod statistical tests and verification based on the PIE results. Also preliminary calculations are implemented with the aim to determine the design schema parameters. This should ensure the accuracy of the assessment of the parameters of WWER fuel rod characteristics distribution. The probability characteristics of fuel rod strength and thermal physics are assessed via the statistical analysis of the results of probability calculations

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

International Nuclear Information System (INIS)

Mikus, J.

2004-01-01

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

Simulation of thermal fluid dynamics in parabolic trough receiver tubes with direct steam generation using the computer code ATHLET

Energy Technology Data Exchange (ETDEWEB)

Hoffmann, Alexander; Merk, Bruno [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany); Hirsch, Tobias; Pitz-Paal, Robert [DLR Deutsches Zentrum fuer Luft- und Raumfahrt e.V., Stuttgart (Germany). Inst. fuer Solarforschung

2014-06-15

In the present feasibility study the system code ATHLET, which originates from nuclear engineering, is applied to a parabolic trough test facility. A model of the DISS (DIrect Solar Steam) test facility at Plataforma Solar de Almeria in Spain is assembled and the results of the simulations are compared to measured data and the simulation results of the Modelica library 'DissDyn'. A profound comparison between ATHLET Mod 3.0 Cycle A and the 'DissDyn' library reveals the capabilities of these codes. The calculated mass and energy balance in the ATHLET simulations are in good agreement with the results of the measurements and confirm the applicability for thermodynamic simulations of DSG processes in principle. Supplementary, the capabilities of the 6-equation model with transient momentum balances in ATHLET are used to study the slip between liquid and gas phases and to investigate pressure wave oscillations after a sudden valve closure. (orig.)

Assessment of WWER fuel condition in design basis accident

Energy Technology Data Exchange (ETDEWEB)

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

1994-12-31

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

Researches of WWER fuel rods behaviour under RIA accident conditions

International Nuclear Information System (INIS)

Nechaeva, O.; Medvedev, A.; Novikov, V.; Salatov, A.

2003-01-01

Unirradiated fuel rod and refabricated fuel rod tests in the BIGR as well as acceptance criteria proving absence of fragmentation and the settlement modeling of refabricated fuel rods thermomechanical behavior in the BIGR-tests using RAPTA-5 code are discussed in this paper. The behaviour of WWER type simulators with E110 and E635 cladding was researched at the BIGR reactor under power pulse conditions simulating reactivity initiated accident. The results of the tests in four variants of experimental conditions are submitted. The behaviour of 12 WWER type refabricated fuel rods was researched in the BIGR reactor under power pulse conditions simulating reactivity initiated accident: burnup 48 and 60 MWd/kgU, pulse width 3 ms, peak fuel enthalpy 115-190 cal/g. The program of future tests in the research reactor MIR with high burnup fuel rod (up to 70 MWd/kgU) under conditions simulating design RIA in WWER-1000 is presented

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

International Nuclear Information System (INIS)

Monette, P.

1995-01-01

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

The FARC fuel archive of WWER

International Nuclear Information System (INIS)

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

1998-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-09-01

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

Radioactive waste management at WWER type reactors

International Nuclear Information System (INIS)

1993-05-01

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

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

International Nuclear Information System (INIS)

1994-01-01

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

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

International Nuclear Information System (INIS)

Georgiev, Yoto; Filipov, Kalin; Velev, Vladimir

2014-01-01

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

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

International Nuclear Information System (INIS)

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

1993-12-01

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

Calculation of an accident with delayed scram at NPP Greifswald using the coupled code DYN3D/ATHLET

Energy Technology Data Exchange (ETDEWEB)

Kliem, S

1998-10-01

Complex computer codes modeling the whole reactor system including 3D neutron kinetics in combination with advanced thermohydraulic plant models become more and more important for the safety assessment of nuclear reactors. Transients or experiments with both neutron kinetic and thermalhydraulic data are needed for the validation of such coupled codes like DYN3D/ATHLET. First of all measured results from nuclear power plant (NPP) transients should be used, because the experimental thermalhydraulic facilities do not offer the possibility to model space-dependent neutron kinetic effects and research reactors with reliably measured 3D neutron kinetic data do not allow to study thermalhydraulic feedback effects. In this paper, an accident with delayed scram which occurred in 1989 at the NPP Greifswald is analyzed. Calculations of this accident were carried out with the goal to validate the coupled code DYN3D/ATHLET. (orig.)

Calculation of an accident with delayed scram at NPP Greifswald using the coupled code DYN3D/ATHLET

International Nuclear Information System (INIS)

Kliem, S.

1998-01-01

Complex computer codes modeling the whole reactor system including 3D neutron kinetics in combination with advanced thermohydraulic plant models become more and more important for the safety assessment of nuclear reactors. Transients or experiments with both neutron kinetic and thermalhydraulic data are needed for the validation of such coupled codes like DYN3D/ATHLET. First of all measured results from nuclear power plant (NPP) transients should be used, because the experimental thermalhydraulic facilities do not offer the possibility to model space-dependent neutron kinetic effects and research reactors with reliably measured 3D neutron kinetic data do not allow to study thermalhydraulic feedback effects. In this paper, an accident with delayed scram which occurred in 1989 at the NPP Greifswald is analyzed. Calculations of this accident were carried out with the goal to validate the coupled code DYN3D/ATHLET. (orig.)

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

International Nuclear Information System (INIS)

Drab, F.; Grof, V.

1978-01-01

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

Thermal-hydraulic calculation methods for transients and accidents of the reactor cooling system under special consideration of multi-dimensional flows (ATHLET, FLUBOX, CFX). Final report; Thermohydraulische Rechenmethoden zu Transienten und Stoerfaellen im Reaktorkuehlkreislauf unter besonderer Beruecksichtigung mehrdimensionaler Stroemungen (ATHLET, FLUBOX, CFX). Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Glaeser, H.; Graf, U.; Herb, J.; and others

2012-02-15

The project RS1184 „Thermal-hydraulic Calculation Methods for Transients and Accidents of the Reactor Cooling System Under Special Consideration of Multi-Dimensional Flows (ATHLET, FLUBOX, CFX)'' consists of four work packages: 1. Further development of the computer code ATHLET 2. Termination of FLUBOX development and development of an ATHLET-internal 3D module 3. Coupling of ATHLET and CFD code ANSYS CFX as well as CFX model development to simulate three-dimensional flows in the reactor coolant system 4. Prediction capability of computer code ATHLET. One of the superior objectives of the project is to improve the prediction capability of the thermal-hydraulic system code ATHLET, including the simulation of multi-dimensional flow in the reactor vessel. The constitutive equations in ATHLET, especially the momentum equations in ATHLET, are written in one-dimensional form. It was planned to develop the 2D/3D module FLUBOX further and couple it with ATHLET. Due to reasons given in chapter 3, the FLUBOX development was terminated. Instead, the decision was made to develop a fast running internal ATHLET-module. The 2D/3D equations for ATHLET have been derived and were implemented. That strategy allows using all ATHLET models and the ATHLET code structure. An additional advantage is that different numerical schemes of different codes, and consequently a loss of efficiency, will be avoided. A second possibility is the coupling of ATHLET with the CFD code ANSYS CFX. Such a coupled code system will be used in those cases when a part of the simulation area is needed to be calculated with high resolution. Such a detailed modelling cannot be provided by ATHLET-3D. A complete representation of the cooling system by a CFD code cannot be performed due to calculation time. In order to calculate the complete system behavior still with ATHLET, that part to be investigated in more detail, will be replaced by a CFX model. Several new models and improvements of existing

Methodology for qualification of in-service inspection systems for WWER nuclear power plants. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

International Nuclear Information System (INIS)

1998-03-01

Program was initiated by IAEA in 1990 with the aim to assist the countries of Central and Eastern Europe and former Soviet Union in evaluating the safety of their first generation WWER-440/230 nuclear power plants. The main objectives were: to identify major design and operational safety issues; to establish international consensus on priorities for safety improvements; and to provide assistance in the review of the competence and and adequacy of safety improvement programs. The scope was extended in 1992 to include RBMK, WWER-440/312 and WWER-1000 plants in operation and under construction. Integrity of primary circuit is fundamental for the safe operation of any nuclear power plant. In-service inspection (ISI) in general terms and in particular, non-destructive tests (NDT) play a key role in maintaining primary circuit integrity. This report provides a methodology for qualification of ISI systems which might be used by WWER operating countries as a commonly accepted basis for further development of the necessary qualification related infrastructures. It also provides several qualification principles defining the administrative framework needed for the practical implementation of the methodology, a description of the process of qualification of an inspection system, specifying its minimum technical and documentation related requirements, as well as specific requirements with regard to the NDT procedures, equipment and personnel to be qualified and the test specimen to be used in practical trials. Finally, the report suggests an appropriate distribution of responsibilities among all the parties involved in a qualification process, based on international practice

Bounding approach in BUC implementation in pool at WWER-440

International Nuclear Information System (INIS)

Havluj, F.

2006-01-01

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

Advanced core monitoring technology for WWER reactors

International Nuclear Information System (INIS)

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

1993-01-01

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

Features of RAPTA-SFD code modelling of chemical interactions of basic materials of the WWER active zone in accident conditions with severe fuel damage

International Nuclear Information System (INIS)

Bibilashvili, Yu.K.; Sokolov, N.B.; Salatov, A.V.; Nechaeva, O.A.; Andreyeva-Andrievskaya, L.N.; Vlasov, F.Yu.

1996-01-01

A brief description of RAPTA-SFD code intended for computer simulations of WWER-type fuel elements (simulator or absorber element) in conditions of accident with severe damage of fuel. Presented are models of chemical interactions of basic materials of the active zone, emphasized are special feature of their application in carrying out of the CORA-W2 experiment within the framework of International Standard Problem ISP-36. Results obtained confirm expediency of phenomenological models application. (author). 6 refs, 7 figs, 1 tab

Experimental research on safety assurance of advanced WWER fuel cycles

International Nuclear Information System (INIS)

Krainov, Ju.; Kukushkin, Ju.

2002-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-04-01

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

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

International Nuclear Information System (INIS)

Zdarek, J.; Pecinka, L.

1997-01-01

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

A Comparison of Athletic Movement Among Talent-Identified Juniors From Different Football Codes in Australia: Implications for Talent Development.

Science.gov (United States)

Woods, Carl T; Keller, Brad S; McKeown, Ian; Robertson, Sam

2016-09-01

Woods, CT, Keller, BS, McKeown, I, and Robertson, S. A comparison of athletic movement among talent-identified juniors from different football codes in Australia: implications for talent development. J Strength Cond Res 30(9): 2440-2445, 2016-This study aimed to compare the athletic movement skill of talent-identified (TID) junior Australian Rules football (ARF) and soccer players. The athletic movement skill of 17 TID junior ARF players (17.5-18.3 years) was compared against 17 TID junior soccer players (17.9-18.7 years). Players in both groups were members of an elite junior talent development program within their respective football codes. All players performed an athletic movement assessment that included an overhead squat, double lunge, single-leg Romanian deadlift (both movements performed on right and left legs), a push-up, and a chin-up. Each movement was scored across 3 essential assessment criteria using a 3-point scale. The total score for each movement (maximum of 9) and the overall total score (maximum of 63) were used as the criterion variables for analysis. A multivariate analysis of variance tested the main effect of football code (2 levels) on the criterion variables, whereas a 1-way analysis of variance identified where differences occurred. A significant effect was noted, with the TID junior ARF players outscoring their soccer counterparts when performing the overhead squat and push-up. No other criterions significantly differed according to the main effect. Practitioners should be aware that specific sporting requirements may incur slight differences in athletic movement skill among TID juniors from different football codes. However, given the low athletic movement skill noted in both football codes, developmental coaches should address the underlying movement skill capabilities of juniors when prescribing physical training in both codes.

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

International Nuclear Information System (INIS)

1997-07-01

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

Ensuring the operational safety of nuclear power plants with WWER reactors

International Nuclear Information System (INIS)

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

1984-01-01

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

WWER core pattern enhancement using adaptive improved harmony search

International Nuclear Information System (INIS)

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

2013-01-01

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

Nuclear fuel utilization in Kozloduy NPP

Energy Technology Data Exchange (ETDEWEB)

Boyadzhiev, Z; Kharalampieva, Ts; Pejchinov, Ts [Kombinat Atomna Energetika, Kozloduj (Bulgaria)

1994-12-31

An assessment of fuel utilization in Kozloduy NPP units 1-6 is made on the basis of operational data obtained for a total of 62 fuel cycles. Basic characteristics of core loading and operation conditions are given. SPPS-1 and BIPR-7 codes are used to calculate assembly-wise power distributions for different full power days of a given cycle and unit. The data are compared with the measured values of these quantities. The analysis performed shows that the core loading option chosen has led to efficient fuel utilization without violation of the nuclear safety criteria. For WWER-440 (Units 1 - 4) this is expressed in effective reduction of the reactor vessel irradiation, maintaining the design duration of the fuel cycles at a reduced number of assemblies by a factor 5 - 5-10%, utilizing fuel with higher enrichment and implementing the 4-year fuel cycle. For WWER-1000 the improvements lead to: adoption of the 3-year fuel cycle utilizing fuel with 4.4% initial enrichment, implementation of improved fuel with a new type of absorbers and more effective low-leakage core loading patterns. 10 tabs., 6 figs., 7 refs.

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

International Nuclear Information System (INIS)

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

1997-01-01

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

WWER-1000 Burnup Credit Benchmark (CB5)

International Nuclear Information System (INIS)

Manolova, M.A.

2002-01-01

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

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

International Nuclear Information System (INIS)

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

1993-01-01

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

Power peak in vicinity of WWER-440 control rod

International Nuclear Information System (INIS)

Mikus, J.

2003-01-01

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

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

International Nuclear Information System (INIS)

1999-12-01

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

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

International Nuclear Information System (INIS)

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

2002-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-31

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

IAEA-NULIFE VERLIFE - Procedure for integrity and lifetime assessment of components and piping in WWER NPPs during operation - Tool for LTO

International Nuclear Information System (INIS)

Brumovsky, M.

2012-01-01

VERLIFE - 'Unified Procedure for Lifetime Assessment of Components and Piping in WWER NPPs during Operation' was developed within the 5th Framework Programme of the European Union in 2003 and later upgraded within the 6th Framework Programme 'COVERS - Safety of WWER NPPs' of the European Union in 2008. This Procedure had to fill the gap in original Soviet/Russian Codes and Rules for WWER type NPPs, as these codes were developed only for design and manufacture and were not changed since their second edition in 1989. VERLIFE Procedure is based on these Russian codes but incorporates also new developments in research, mainly in fracture mechanics, and also some principal approaches used in PWR codes. To assure that VERLIFE Procedure will remain a living document, new 3-years IAEA project (in close cooperation with another project of the 6th Framework Programme of the European Union 'NULIFE - Plant Life Management of NPPs') has started in 2009. Final document, was approved by expert groups of the IAEA and NULIFE in June 28-30, 2011, and will be issued as 'IAEA/NULIFE Guidelines for Integrity and Lifetime Assessment of Components and Piping in WWER NPPs during Operation'. This document represents a necessary part for any integrity and lifetime assessment during operation that is a bases for further decision about safe and potential long term operation. To prepare documents like TLAA, it is necessary to have a tool that is able to evaluate lifetime of the main NPP components taking into account existing past operation as well as proposal for the future. (author)

Present status of the physical-neutronic calculation system for nuclear fuel management at the Juragua Power Plant WWER-440 reactors

International Nuclear Information System (INIS)

Garcia H, Carlos R.; Milian, Daniel

1997-01-01

In this paper the most remarkable results obtained in Cuba, for the Juragua Nuclear Power Plant (JNPP) in core fuel management, are shown. The main characteristic of the codes used to solve the usual sequence of neutron-physical calculations available in our code library are reported. The codes validation was based on estimative of their accuracy by inter comparing calculated and WWER-440 NPP's operation data. A brief summary of the extensive calculations carried out for the JNPP Preliminary Safety Report elaboration is presented. This report is one of the requisites demanded by the Cuban Nuclear Agency for issuing the General Permission for restart works in the plant. (author). 12 refs., 4 figs., 6 tabs

The European source term code ESTER - basic ideas and tools for coupling of ATHLET and ESTER

International Nuclear Information System (INIS)

Schmidt, F.; Schuch, A.; Hinkelmann, M.

1993-04-01

The French software house CISI and IKE of the University of Stuttgart have developed during 1990 and 1991 in the frame of the Shared Cost Action Reactor Safety the informatic structure of the European Source TERm Evaluation System (ESTER). Due to this work tools became available which allow to unify on an European basis both code development and code application in the area of severe core accident research. The behaviour of reactor cores is determined by thermal hydraulic conditions. Therefore for the development of ESTER it was important to investigate how to integrate thermal hydraulic code systems with ESTER applications. This report describes the basic ideas of ESTER and improvements of ESTER tools in view of a possible coupling of the thermal hydraulic code system ATHLET and ESTER. Due to the work performed during this project the ESTER tools became the most modern informatic tools presently available in the area of severe accident research. A sample application is given which demonstrates the use of the new tools. (orig.) [de

Criticality safety study of dry spent fuel cask loaded with increased enrichment fuel

International Nuclear Information System (INIS)

Bznuni, S.; Baghdasaryan, N.; Amirjanyan, A.

2013-01-01

Existing Dry Spent Fuel Casks (DSC) for transporting and storing of Armenian NPP fuel was licensed for WWER-440 fuel assemblies with 3.6% enrichment. Having in mind that ANPP introduced new fuel assemblies with increased enrichment (3.82 %) re-assessment of criticality safety analysis for DSC is required. Criticality safety analysis of DSC was performed by KENO-VI program using 238-GROUP ENDF/B-VII.0 LIBRARY (V7-238). Results of analysis showed that additional 8 borated racks for fuel assemblies should be included in the design of DSC. In addition feasibility study was performed to find out level of burnup-credit approach implementation to keep current design of DSC unchanged. Burnup-credit analysis was performed by STARBUCS program using axial burnup profiles from Armenian NPP neutronics analysis carried out by BIPR code. (authors)

WWER core pattern enhancement using adaptive improved harmony search

Energy Technology Data Exchange (ETDEWEB)

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

2013-01-15

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

Analysis of operating reliability of WWER-1000 unit

International Nuclear Information System (INIS)

Bortlik, J.

1985-01-01

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

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

International Nuclear Information System (INIS)

Por, G.

1998-01-01

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

Application of the coupled code Athlet-Quabox/Cubbox for the extreme scenarios of the OECD/NRC BWR turbine trip benchmark and its performance on multi-processor computers

International Nuclear Information System (INIS)

Langenbuch, S.; Schmidt, K.D.; Velkov, K.

2003-01-01

The OECD/NRC BWR Turbine Trip (TT) Benchmark is investigated to perform code-to-code comparison of coupled codes including a comparison to measured data which are available from turbine trip experiments at Peach Bottom 2. This Benchmark problem for a BWR over-pressure transient represents a challenging application of coupled codes which integrate 3-dimensional neutron kinetics into thermal-hydraulic system codes for best-estimate simulation of plant transients. This transient represents a typical application of coupled codes which are usually performed on powerful workstations using a single CPU. Nowadays, the availability of multi-CPUs is much easier. Indeed, powerful workstations already provide 4 to 8 CPU, computer centers give access to multi-processor systems with numbers of CPUs in the order of 16 up to several 100. Therefore, the performance of the coupled code Athlet-Quabox/Cubbox on multi-processor systems is studied. Different cases of application lead to changing requirements of the code efficiency, because the amount of computer time spent in different parts of the code is varying. This paper presents main results of the coupled code Athlet-Quabox/Cubbox for the extreme scenarios of the BWR TT Benchmark together with evaluations of the code performance on multi-processor computers. (authors)

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

International Nuclear Information System (INIS)

Kostarev, V.

1993-01-01

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

The relevance of the IFPE Database to the modelling of WWER-type fuel behaviour

International Nuclear Information System (INIS)

Killeen, J.; Sartori, E.

2006-01-01

The aim of the International Fuel Performance Experimental Database (IFPE Database) is to provide, in the public domain, a comprehensive and well-qualified database on zircaloy-clad UO 2 fuel for model development and code validation. The data encompass both normal and off-normal operation and include prototypic commercial irradiations as well as experiments performed in Material Testing Reactors. To date, the Database contains over 800 individual cases, providing data on fuel centreline temperatures, dimensional changes and FGR either from in-pile pressure measurements or PIE techniques, including puncturing, Electron Probe Micro Analysis (EPMA) and X-ray Fluorescence (XRF) measurements. This work in assembling and disseminating the Database is carried out in close co-operation and co-ordination between OECD/NEA and the IAEA. The majority of data sets are dedicated to fuel behaviour under LWR irradiation, and every effort has been made to obtain data representative of BWR, PWR and WWER conditions. In each case, the data set contains information on the pre-characterisation of the fuel, cladding and fuel rod geometry, the irradiation history presented in as much detail as the source documents allow, and finally any in-pile or PIE measurements that were made. The purpose of this paper is to highlight data that are relevant specifically to WWER application. To this end, the NEA and IAEA have been successful in obtaining appropriate data for both WWER-440 and WWER-1000-type reactors. These are: 1) Twelve (12) rods from the Finnish-Russian co-operative SOFIT programme; 2) Kola-3 WWER-440 irradiation; 3) MIR ramp tests on Kola-3 rods; 4) Zaporozskaya WWER-1000 irradiation; 5) Novovoronezh WWER-1000 irradiation. Before reviewing these data sets and their usefulness, the paper touches briefly on recent, more novel additions to the Database and on progress made in the use of the Database for the current IAEA FUMEX II Project. Finally, the paper describes the Computer

Report of a consultants meeting on accidents during shutdown conditions for WWER nuclear power plants. Extrabudgetary programme on the safety of WWER NPPs

International Nuclear Information System (INIS)

1996-07-01

The main objectives of the meeting were to exchange information on the operational occurrences, studies performed and countermeasures taken for the accidents during shutdown for WWERs, and to define the necessity and directions of the further activities which may promote the improvement of WWER safety under shutdown conditions. The consultants have discussed some aspects concerning vulnerability of safety functions during shutdown conditions, several steps required to performed accident analysis and selected operational aspects for shutdown conditions. The discussion was supported by an evaluation of selected operational occurrences. The consultants have agreed that the discussion during the meeting in major parts is relevant to all the WWER designs (i.e. WWER-1000, WWER-440/213 and WWER-440/230). As for the plant conditions, the consultants have agreed to bound the discussion mainly by the cold shutdown and refuelling modes. Refs, figs, tabs

SEJV2 software package for radiation monitoring system of WWER 440 NPP

International Nuclear Information System (INIS)

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

1993-01-01

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

Use of NESTLE computer code for NPP transition process analysis

International Nuclear Information System (INIS)

Gal'chenko, V.V.

2001-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1995-12-31

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-31

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

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

International Nuclear Information System (INIS)

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

2004-01-01

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

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

International Nuclear Information System (INIS)

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

1984-01-01

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

About the application of MCNP4 code in nuclear reactor core design calculations

International Nuclear Information System (INIS)

Svarny, J.

2000-01-01

This paper provides short review about application of MCNP code for reactor physics calculations performed in SKODA JS. Problems of criticality safety analysis of spent fuel systems for storage and transport of spent fuel are discussed and relevant applications are presented. Application of standard Monte Carlo code for accelerator driven system for LWR waste destruction is shown and conclusions are reviewed. Specific heterogeneous effects in neutron balance of WWER nuclear cores are solved for adjusting standard design codes. (Authors)

Development of sub-channel/system coupled code and its application to a supercritical water-cooled test loop

International Nuclear Information System (INIS)

Liu, X.J.; Yang, T.; Cheng, X.

2014-01-01

To analyze the local thermal-hydraulic parameters in the supercritical water reactor-fuel qualification test (SCWR-FQT) fuel bundle with a flow blockage, a coupled sub-channel and system code system is developed in this paper. Both of the sub-channel code and system code are adapted to transient analysis of SCWR. Two codes are coupled by data transfer and data adaptation at the interface. In the coupled code, the whole system behavior including safety system characteristic is analyzed by system code ATHLET-SC, whereas the local thermal-hydraulic parameters are predicted by the sub-channel code COBRA-SC. Sensitivity analysis are carried out respectively in ATHLET-SC and COBRA-SC code, to identify the appropriate models for description of the flow blockage phenomenon in the test loop. Some measures to mitigate the accident consequence are also trialed to demonstrate their effectiveness. The results indicate that the new developed code has good feasibility to transient analysis of supercritical water-cooled test. And the peak cladding temperature caused by blockage in the fuel assembly can be reduced effectively by the safety measures of SCWR-FQT. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

1995-12-31

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

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

International Nuclear Information System (INIS)

Balabanov, E.; Ivanova, A.

1995-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-31

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

Analysis of Rod Withdrawal at Power (RWAP) Accident using ATHLET Mod 2.2 Cycle A and RELAP5/mod 3.3 Codes

International Nuclear Information System (INIS)

Bencik, V.; Cavlina, N.; Grgic, D.

2012-01-01

The system code ATHLET is being developed at Gesellschaft fuer Anlagen-und Reaktorsicherheit (GRS) in Germany. In 1996, the NPP Krsko (NEK) input deck for ATHLET Mod 1.1 Cycle C has been developed at Faculty of Electrical Engineering (FER), University of Zagreb. The input deck was tested by analyzing the realistic plant event 'Main Steam Isolation Valve Closure' and the results were assessed against the measured data. The input deck was established before plant modernization that took place in 2000 and included the power uprate and SG replacement. The released ATHLET version (Mod 2.2 Cycle A) is now being available at FER Zagreb. Accordingly, the NEK input deck for ATHLET Mod 2.2 Cycle A has been developed. A completely new input deck has been created taking into account the large number of changes due to power uprate and SG replacement as well as taking advantage of developmental work on NEK data base performed at FER. The new NEK input deck for ATHLET code has been tested by analyzing the Rod Withdrawal Power (RWAP) accident and the results were assessed against the analysis performed by RELAP5/mod 3.3 code. The RWAP accident can be either Departure from Nucleate Boiling (DNB) ratio or overpower limiting accident depending on initial power and reactivity insertion rate. Since the automatic rod control system is assumed unavailable, the only negative reactivity is due to Doppler and moderator feedback. Consequently, the nuclear power and the transferred heat in the steam generators (SGs) increase. Since the steam flow to the turbine and the extracted power from the SGs remain constant, the SG secondary pressure and the temperatures on the primary side increase. Unless terminated by manual or automatic action, the power mismatch between primary and secondary side and the resultant coolant temperature rise could eventually result in DNB ratio and/or fuel centreline melt. In order to avoid core damage, the reactor protection system is designed to automatically

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

International Nuclear Information System (INIS)

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

1984-01-01

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

Perspective decisions of WWER nuclear fuel: Implementation at Russian NPPs

International Nuclear Information System (INIS)

Molchanov, V.

2003-01-01

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

About reliability of WWER pressure vessel neutron fluence calculation

Energy Technology Data Exchange (ETDEWEB)

Belousov, S; Ilieva, K; Antonov, S [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika

1996-12-31

This reliability study was carried out under a Research Contracts F111 and TH-324 of the Bulgarian Ministry of Higher Education and the IAEA. The effect of geometry approximation and the choice of neutron cross-sections on the calculation model is estimated. The neutron flux onto reactor pressure vessel at locations, important for metal embrittlement surveillance, has been calculated using the codes TORT and DORT. The flux values calculated for both WWER-440 and WWER-1000 show good consistency within the limits of solution accuracy. It is concluded that the synthesis method (DORT) can be used for calculations at a reasonable cost whenever metal embrittlement surveillance is considered. Using an iron sphere benchmark measurement, a comparison of an experimental leakage spectrum with spectrum calculated using multigroup neutron cross-sections based on ENDF/B-4 and ENDF/B-6 data is performed. In the energy region above 1 MeV the best agreement with the experiment is achieved for ENDF/B-6 in VITAMIN-E group structure. 7 refs., 1 fig., 4 tabs.

About reliability of WWER pressure vessel neutron fluence calculation

International Nuclear Information System (INIS)

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

1995-01-01

This reliability study was carried out under a Research Contracts F111 and TH-324 of the Bulgarian Ministry of Higher Education and the IAEA. The effect of geometry approximation and the choice of neutron cross-sections on the calculation model is estimated. The neutron flux onto reactor pressure vessel at locations, important for metal embrittlement surveillance, has been calculated using the codes TORT and DORT. The flux values calculated for both WWER-440 and WWER-1000 show good consistency within the limits of solution accuracy. It is concluded that the synthesis method (DORT) can be used for calculations at a reasonable cost whenever metal embrittlement surveillance is considered. Using an iron sphere benchmark measurement, a comparison of an experimental leakage spectrum with spectrum calculated using multigroup neutron cross-sections based on ENDF/B-4 and ENDF/B-6 data is performed. In the energy region above 1 MeV the best agreement with the experiment is achieved for ENDF/B-6 in VITAMIN-E group structure. 7 refs., 1 fig., 4 tabs

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

Directory of Open Access Journals (Sweden)

Navid Poursalehi

2017-02-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-02-15

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-31

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

Inovation of the computer system for the WWER-440 simulator

International Nuclear Information System (INIS)

Schrumpf, L.

1988-01-01

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

Audit calculations of accidents analysis for second unit of Ignalina NPP with ATHLET code

International Nuclear Information System (INIS)

Adomavicius, A.; Belousov, A.; Ognerubov, V.

2004-01-01

Background of thermo hydraulic processes audit calculations in the frame of RSR-2 project is presented. Assumptions for the design based accident - RBMK-1500 group distributor header break analysis and modeling are presented. Audit calculations by ATHLET code and evaluation of results were provided. (author)

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

International Nuclear Information System (INIS)

Havel, R.

1995-01-01

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

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

International Nuclear Information System (INIS)

Stanc, S.; Repa, M.

2001-01-01

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

Sub-channel/system coupled code development and its application to SCWR-FQT loop

International Nuclear Information System (INIS)

Liu, X.J.; Cheng, X.

2015-01-01

Highlights: • A coupled code is developed for SCWR accident simulation. • The feasibility of the code is shown by application to SCWR-FQT loop. • Some measures are selected by sensitivity analysis. • The peak cladding temperature can be reduced effectively by the proposed measures. - Abstract: In the frame of Super-Critical Reactor In Pipe Test Preparation (SCRIPT) project in China, one of the challenge tasks is to predict the transient performance of SuperCritical Water Reactor-Fuel Qualification Test (SCWR-FQT) loop under some accident conditions. Several thermal–hydraulic codes (system code, sub-channel code) are selected to perform the safety analysis. However, the system code cannot simulate the local behavior of the test bundle, and the sub-channel code is incapable of calculating the whole system behavior of the test loop. Therefore, to combine the merits of both codes, and minimizes their shortcomings, a coupled sub-channel and system code system is developed in this paper. Both of the sub-channel code COBRA-SC and system code ATHLET-SC are adapted to transient analysis of SCWR. Two codes are coupled by data transfer and data adaptation at the interface. In the new developed coupled code, the whole system behavior including safety system characteristic is analyzed by system code ATHLET-SC, whereas the local thermal–hydraulic parameters are predicted by the sub-channel code COBRA-SC. The codes are utilized to get the local thermal–hydraulic parameters in the SCWR-FQT fuel bundle under some accident case (e.g. a flow blockage during LOCA). Some measures to mitigate the accident consequence are proposed by the sensitivity study and trialed to demonstrate their effectiveness in the coupled simulation. The results indicate that the new developed code has good feasibility to transient analysis of supercritical water-cooled test. And the peak cladding temperature caused by blockage in the fuel bundle can be reduced effectively by the safety measures

Sub-channel/system coupled code development and its application to SCWR-FQT loop

Energy Technology Data Exchange (ETDEWEB)

Liu, X.J., E-mail: xiaojingliu@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong Chuan Road, Shanghai 200240 (China); Cheng, X. [Institute of Fusion and Reactor Technology, Karlsruhe Institute of Technology, Vincenz-Prießnitz-Str. 3, 76131 Karlsruhe (Germany)

2015-04-15

Highlights: • A coupled code is developed for SCWR accident simulation. • The feasibility of the code is shown by application to SCWR-FQT loop. • Some measures are selected by sensitivity analysis. • The peak cladding temperature can be reduced effectively by the proposed measures. - Abstract: In the frame of Super-Critical Reactor In Pipe Test Preparation (SCRIPT) project in China, one of the challenge tasks is to predict the transient performance of SuperCritical Water Reactor-Fuel Qualification Test (SCWR-FQT) loop under some accident conditions. Several thermal–hydraulic codes (system code, sub-channel code) are selected to perform the safety analysis. However, the system code cannot simulate the local behavior of the test bundle, and the sub-channel code is incapable of calculating the whole system behavior of the test loop. Therefore, to combine the merits of both codes, and minimizes their shortcomings, a coupled sub-channel and system code system is developed in this paper. Both of the sub-channel code COBRA-SC and system code ATHLET-SC are adapted to transient analysis of SCWR. Two codes are coupled by data transfer and data adaptation at the interface. In the new developed coupled code, the whole system behavior including safety system characteristic is analyzed by system code ATHLET-SC, whereas the local thermal–hydraulic parameters are predicted by the sub-channel code COBRA-SC. The codes are utilized to get the local thermal–hydraulic parameters in the SCWR-FQT fuel bundle under some accident case (e.g. a flow blockage during LOCA). Some measures to mitigate the accident consequence are proposed by the sensitivity study and trialed to demonstrate their effectiveness in the coupled simulation. The results indicate that the new developed code has good feasibility to transient analysis of supercritical water-cooled test. And the peak cladding temperature caused by blockage in the fuel bundle can be reduced effectively by the safety measures

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Analysis of WWER-440 fuel performance under normal operating conditions

Energy Technology Data Exchange (ETDEWEB)

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

1994-12-31

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

Final Report for the FUMEX-III Exercise with the TRANSURANUS Code

International Nuclear Information System (INIS)

Van Uffelen, P.; Schubert, A.; Van de Laar, J.; Di Marcello, V.

2013-01-01

This report describes the results of the fourth round robin exercise organized by the IAEA for the LWR fuel behaviour codes. The previous exercise was organized by the IAEA in 2002-2007. ITU contributed to the IAEA FUMEX- II co-ordinated research project (CRP) by: - The development of a high burn-up TRANSURANUS-WWER Version - Verification of the TRANSURANUS-WWER Version for WWER-1000 reactors - Further verification of the TRANSURANUS Code by selected irradiations from the IFPE Database - Transfer of latest ITU knowledge in the following areas: high burnup effects and MOX behaviour as far as confidentiality is not concerned. Within the FUMEX-III project this work continues by: - Knowledge transfer and release of the TRANSURANUS code to safety authorities in several neighbouring countries of the European Union - Further verification of the TRANSURANUS Code by selected irradiations from the IFPE Database: extending the verification of the TRANSURANUS code on the basis of the cases related to the behaviour of high burnup UO 2 fuel, Gd-containing UO 2 fuel and MOX fuel under normal operating conditions in LWRs, including WWER. The simulation of the pellet-cladding mechanical interaction received particular attention. - Extending the TRANSURANUS code for simulation of fuel rods under accidental conditions, such as a loss of coolant accidents (LOCA) and a Reactivity Initiated Accident (RIA), for which various code improvements have been implemented and other changes are still under development. During the entire period of the CRP, ITU has carried out all priority cases, except for IFA-519 for which insufficient data have been provided. All these results are considered in this report. However, in order to limit the redundancy of work of various TRANSURANUS users involved in the FUMEX-III CRP, ITU has co-ordinated part of the work of the partners in Bulgaria, Romania and Italy. More precisely, the work on the WWER version of the code has been carried out in collaboration

Addressing the scaling issue by thermalhydraulic system codes: recent results

International Nuclear Information System (INIS)

D'auria, F.; Cherubini, M.; Galassi, G.M.; Muellner, N. . E-mail of corresponding author: dauria@ing.unipi.it; D'auria, F.)

2005-01-01

This lecture presents an introduction into the scaling issue following a 'top-down' approach. This means, recent studies which deal with a scaling analysis in LWR with special regards to the WWER Russian reactor type are presented to demonstrate important phenomena for scaling, to be more precise, the counterpart test (CT) methodology. As an example, one CT, a Small Break LOCA carried out in the PSB facility, is presented. PSB is a full height, full pressure rig that reproduces a WWER 1000, power and volume scaling factor is 1:300. The CT has been designed deriving boundary and initial condition from the same test performed in LOBI (that reproduces a PWR). The adopted scaling approach is based on the selection of a few characteristic parameters. They are chosen taking into account their relevance in the behaviour of the transient. The calculation of the SBLOCA has been performed using Relap5/Mod3.3 computer code and its accuracy has been demonstrated by qualitative and quantitative evaluation. For the quantitative evaluation the use of the FFT Based Method is foreseen and the fulfilment of its limits has been obtained. The aim of the example is to give an overview about the theoretical concepts of scaling, which is termed the s caling strategy , and comprises the steps of the selected scaling approach. At the same time interesting results from ongoing research projects will be presented. Comparing experimental data it was found that the investigated facilities show similar behaviour concerning the time trends, and are able to predict on a qualitative level the same thermal hydraulic phenomena. Main obtained results are summarized as follows: PSB and LOBI main parameters have similar trends. This is a confirmation of the validity of the adopted scaling approach and shows that PWR and WWER reactor type behaviour are very close to each other. No new phenomena occur during the CT, notwithstanding the two facilities have a different lay out, and the already known

Advances in new WWER designs to improve operation and maintenance

International Nuclear Information System (INIS)

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

2000-01-01

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

Final results of the fifth three-dimensional dynamic Atomic Energy Research benchmark problem calculations

International Nuclear Information System (INIS)

Hadek, J.

1999-01-01

The paper gives a brief survey of the fifth three-dimensional dynamic Atomic Energy Research benchmark calculation results received with the code DYN3D/ATHLET at NRI Rez. This benchmark was defined at the seventh Atomic Energy Research Symposium (Hoernitz near Zittau, 1997). Its initiating event is a symmetrical break of the main steam header at the end of the first fuel cycle and hot shutdown conditions with one stuck out control rod group. The calculations were performed with the externally coupled codes ATHLET Mod.1.1 Cycle C and DYN3DH1.1/M3. The standard WWER-440/213 input deck of ATHLET code was adopted for benchmark purposes and for coupling with the code DYN3D. The first part of paper contains a brief characteristics of NPP input deck and reactor core model. The second part shows the time dependencies of important global and local parameters. In comparison with the results published at the eighth Atomic Energy Research Symposium (Bystrice nad Pernstejnem, 1998), the results published in this paper are based on improved ATHLET descriptions of control and safety systems. (Author)

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

International Nuclear Information System (INIS)

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

1990-01-01

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

Investigation of flashing-induced instabilities at Circus test facility with the code ATHLET

Energy Technology Data Exchange (ETDEWEB)

Schafer, F.; Manera, A. [Forschungzentrum Rossendorf e.V., Institute of Safety Research, P.O. Box 510119, D-01314 Dresden (Germany)]. E-mail: F.Schaefer@fz-rossendorf.de; A.Manera@fz-rossendorf.de

2006-07-01

The test facility CIRCUS (CIRculation Under Start-up) was built to study the start-up phase of a natural-circulation BWR. During the start-up,so-called flashing-induced instabilities can arise. These instabilities are induced by flashing (i.e., steam production in adiabatic conditions) of the coolant in the long riser section, which is placed above the core to enhance the flow rate. The flashing that occurs in the riser causes an imbalance between driving force and pressure losses in the natural-circulation loop, giving rise to flow oscillations. Within the European-Union 5th Framework Programme, a project, NACUSP (Natural circulation and stability performance of BWRs), has been started in December 2000, having as one of its main aims the understanding of the physics of the phenomena involved during the start-up phase of natural-circulation-cooled BWRs, providing a large experimental database and validating state-of-the-art thermo-hydraulic codes in the low-pressure, low-power operational region of these reactors. One part of this project deals with the modelling of selected CIRCUS tests using the thermo-hydraulic code ATHLET (Analysis of THermal-hydraulics of LEaks and Transients). This paper gives an overview about experiments and simulations. The code ATHLET is used to investigate the dynamic behaviour of the CIRCUS test facility and the results of the calculations are compared with the experimental data. (author)

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Radiation embrittlement of WWER-1000 reactor vessel steels

International Nuclear Information System (INIS)

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

2001-01-01

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

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)

Databases on safety issues for WWER and RBMK reactors. Users' manual. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

International Nuclear Information System (INIS)

1996-04-01

At the beginning of the IAEA Extrabudgetary Programme on the safety of WWER reactors a great number of findings and recommendations (safety items) were collected as a result of design review and safety review missions of the WWER-440/230 type reactors. On the basis of these findings a technical database containing more than 1300 records was established to support the consolidation of the information obtained and to help in identification of safety issues. After the scope of the WWER extrabudgetary programme was extended similar data sets were prepared for the WWER-440/213, WWER-1000 and RBMK nuclear power plants. This publication describes the structure of the databases on safety issues of WWER and RBMK NPPs, the information sources used in the databases and interrogation capabilities for users to obtain the necessary information. 14 refs, 9 figs, 5 tabs

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Comparative analysis of a LOCA for a German PWR with ASTEC and ATHLET-CD

International Nuclear Information System (INIS)

Reinke, N.; Chan, H.W.; Sonnenkalb, M.

2013-01-01

This paper presents the results of a comparative analysis performed with ASTEC V2.02 and a coupled ATHLET-CD V2.2c /COCOSYS V2.4 calculation for a German 1300 MWe KONVOI type PWR. The purpose of this analysis is mainly to assess the ASTEC code behaviour in modelling of both the thermal-hydraulic phenomena in the coolant circuit arising during a hypothetical severe accident and the early phase of the core degradation versus the more mechanistic code system ATHLET-CD/COCOSYS. The performed analyses cover a loss of coolant accident sequence (LOCA). Such comparison has been done for the first time. The integral code ASTEC (Accident Source Term Evaluation Code) commonly developed since 1996 by IRSN and GRS is a fast running programme, which allows the calculation of entire sequences of severe accidents (SA) in light water reactors from the initiating event up to the release of fission products into the environment, thereby covering all important in-vessel and containment phenomena. The thermal-hydraulic mechanistic system code ATHLET (Analysis of THermal-hydraulics of LEaks and Transients) is being developed by GRS for the analysis of the whole spectrum of leaks and transients in PWRs and BWRs. For modeling of core degradation processes the CD part (Core Degradation) of ATHLET can be activated. For analyses of the containment behavior, ATHLET-CD has been coupled to the GRS code COCOSYS (COntainment COde SYStem). (orig.)

Validation of the coupled neutron kinetic thermohydraulic code ATHLET/DYN3D with help of measured data of the OECD Turbine Trip Benchmarks. Final report

International Nuclear Information System (INIS)

Grundmann, U.; Kliem, S.

2003-12-01

The project consisted in the validation of the coupled neutron kinetic/thermal hydraulic code system ATHLET/DYN3D for boiling water reactors by the participation at the OECD/NRC turbine trip benchmark. The benchmark defined by the OECD and the American NRC is based on an experiment with closure of the turbine stop valve which was carried out in 1977 in the nuclear power plant Peach Bottom 2 within the framework of a series of 3 experiments. In the experiment, the closure of the valve caused a pressure wave which propagated with attenuation into the reactor core. The condensation of steam in the reactor core caused by the increase of pressure lead to a positive reactivity insertion. The following rise of power was limited by the feedback and the insertion of the control rods. In the frame of the benchmark, the codes could be validated by comparisons with the measured results and the result of the other participants. The benchmark was divided into 3 phases or exercises. Phase I was used for checking the thermo-hydraulic model of the system using a given power release in the core. In phase II, three-dimensional core calculations were performed for given thermal-hydraulic boundary conditions. Coupled calculations were carried out for the selected experiment and four extreme scenarios in the phase III. In the frame of the project, FZR took part in phases II and III of the benchmark. The calculations for phase II were performed with DYN3D by using the assembly discontinuity factors (ADF) and 764 thermal-hydraulic channels (1 channel/assembly). The ATHLET input data set for the coolant system was obtained form the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS). It was slightly modified for the phase III calculations carried out with the parallel coupling of ATHLET and DYN3D. For spatially averaged parameters, a good agreement with the results of measurement and the results of other codes was achieved. The influence of the different models was investigated with the

Evaluation of fatigue crack growth in the primary circuit pipeline of a WWER 440/213c type nuclear power plant

International Nuclear Information System (INIS)

Samohyl, P.

1993-07-01

The fatigue damage of the primary circuit of WWER-440/213c reactors was evaluated proceeding from actual and design operating data of units 3 and 4 of the Bohunice V-2 nuclear power plant. A complex computation model was set up, encompassing the main circulation pipeline, pressurizer pipeline, emergency core aftercooling system pipeline, steam pipeline, and feedwater pipeline. The standardized STATIC code was applied to the stress analysis, and the FATLBB code was used to determine the crack increment for all operating states and primary circuit sections. The probability of fatigue failure of the pipelines was found to be low. (J.B.). 55 tabs., 3 figs., 9 refs

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

International Nuclear Information System (INIS)

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

2002-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1994-12-07

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

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

International Nuclear Information System (INIS)

1994-01-01

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

Solution of the 6th dynamic AER benchmark using the coupled core DYN3D/ATHLET

International Nuclear Information System (INIS)

Seidel, A.; Kliem, S.

2001-01-01

The 6 th dynamic benchmark is a logical continuation of the work to validate systematically coupled neutron kinetics/thermohydraulics code systems for the estimation of the transient behaviour of WWER type nuclear power plant which was started in the 5 th dynamic benchmark. This benchmark concerns a double ended break of the main steam line (asymmetrical MSLB) in a WWER plant. The core is at the end of first cycle in full power conditions. The asymmetric leak causes a different depressurization of all steam generators. New features in comparison to the 5 th dynamic benchmark were included: asymmetric operation of the feed water system, consideration of incomplete coolant mixing in the reactor vessel, and the definition of a fixed isothermal recriticality temperature for normalising the nuclear data (Authors)

Check and visualization of input geometry data using the geometrical module of the Monte Carlo code MCU: WWER-440 pressure vessel dosimetry benchmarks

International Nuclear Information System (INIS)

Gurevich, M.; Zaritsky, S.; Osmera, B.; Mikus, J.

1997-01-01

The Monte Carlo method gives the opportunity to conduct the calculations of neutron and photon flux without any simplifications of the 3-D geometry of the nuclear power and experimental devices. So, each graduated Monte Carlo code includes the combinatorial geometry module and tools for the geometry description giving a possibility to describe very complex systems with a number of hierarchy levels of the geometrical objects. Such codes as usual have special modules for the visual checking of geometry input information. These geometry opportunities could be used for all cases when the accurate 3-D description of the complex geometry becomes a necessity. The description (specification) of benchmark experiments is one of the such cases. Such accurate and uniform description detects all mistakes and ambiguities in the starting information of various kinds (drawings, reports etc.). Usually the quality of different parts of the starting information (generally produced by different persons during the different stages of the device elaboration and operation) is different. After using the above mentioned modules and tools, the resultant geometry description can be used as a standard for this device. One can automatically produce any type of the device figure. The detail geometry description can be used as input for different calculation models carrying out (not only for Monte Carlo). The application of that method to the description of the WWER-440 mock-ups is represented in the report. The mock-ups were created on the reactor LR-O (NRI) and the reactor vessel dosimetry benchmarks were developed on the basis of these mock-up experiments. The NCG-8 module of the Russian Monte Carlo code MCU was used. It is the combinatorial multilingual universal geometrical module. The MCU code was certified by Russian Nuclear Regulatory Body. Almost all figures for mentioned benchmarks specifications were made by the MCU visualization code. The problem of the automatic generation of the

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-09-15

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

Numerical effects in the neutron flux calculations into WWER-type reactor vessels using the Monte Carlo

International Nuclear Information System (INIS)

Garcia Yip, F.; Alvarez Cardona, C.M.; Rodriguez Gual, M.; Hernandez Valle, S.

2000-01-01

In the present paper, a PC version of the Monte Carlo 3-D HEXANN-EVALU system is used for the estimation of the WWER reactor pressure vessel irradiation. It was selected on the basis of its flexible options that, however, need to be quantified in connection with the desired magnitudes. The parameters that control the random walk of neutrons, as well as the efficiency increasing options included in the code, are studied in order to identify their impact on the final results for fluxes and fluence in the reactor pressure vessel. As a result, an optimal set of parameters is suggested. (authors)

Automated WWER steam generator eddy current testing and plugging control system

International Nuclear Information System (INIS)

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

2004-01-01

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

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

International Nuclear Information System (INIS)

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

1993-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-31

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

Calibration of the enigma code for Finnish reactor fuel with support from experimental irradiations

Energy Technology Data Exchange (ETDEWEB)

Kelppe, S; Ranta-Puska, K [VTT Energy, Jyvaeskylae (Finland)

1997-08-01

Assessment by VTT of the ENIGMA fuel performance code, the original version by Nuclear Electric plc of the UK amended by a set of WWER specific materials correlations, is described. The given examples of results include analyses for BWR 9 x 9 fuel, BWR fuel irradiated in the reinstrumented test of an international Riso project, pre-characterized commercial WWER fuel irradiated in Loviisa reactor in Finland, and instrumented WWER test fuel irradiations in the MR reactor in Russia. The effects of power uncertainty and some model parameters are discussed. Considering the fact that the described cases all mean prototypic application of the code, the results are well encouraging. The importance of the accuracy in temperature calculations is emphasized. (author). 2 refs, 12 figs, 1 tab.

PWR and WWER fuel performance. A comparison of major characteristics

International Nuclear Information System (INIS)

Weidinger, H.

2006-01-01

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

WWER water chemistry related to fuel cladding behaviour

Energy Technology Data Exchange (ETDEWEB)

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

1997-02-01

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

Express diagnostics of WWER fuel rods at nuclear power plants

International Nuclear Information System (INIS)

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

2009-01-01

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

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

International Nuclear Information System (INIS)

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

1983-01-01

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

Operational behaviour of WWER nuclear power units after Chernobyl accident

International Nuclear Information System (INIS)

Milivojevic, S.; Spasojevic, D.

2000-01-01

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

International Nuclear Information System (INIS)

Dultchenko, A.; Mikolaitchouk, H.

1995-01-01

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

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

International Nuclear Information System (INIS)

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

1982-01-01

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

Interface requirements to couple thermal-hydraulic codes to 3D neutronic codes

Energy Technology Data Exchange (ETDEWEB)

Langenbuch, S.; Austregesilo, H.; Velkov, K. [GRS, Garching (Germany)] [and others

1997-07-01

The present situation of thermalhydraulics codes and 3D neutronics codes is briefly described and general considerations for coupling of these codes are discussed. Two different basic approaches of coupling are identified and their relative advantages and disadvantages are discussed. The implementation of the coupling for 3D neutronics codes in the system ATHLET is presented. Meanwhile, this interface is used for coupling three different 3D neutronics codes.

Interface requirements to couple thermal-hydraulic codes to 3D neutronic codes

International Nuclear Information System (INIS)

Langenbuch, S.; Austregesilo, H.; Velkov, K.

1997-01-01

The present situation of thermalhydraulics codes and 3D neutronics codes is briefly described and general considerations for coupling of these codes are discussed. Two different basic approaches of coupling are identified and their relative advantages and disadvantages are discussed. The implementation of the coupling for 3D neutronics codes in the system ATHLET is presented. Meanwhile, this interface is used for coupling three different 3D neutronics codes

Review of WWER fuel and material tests in the Halden reactor

International Nuclear Information System (INIS)

Volkov, B.; Kolstad, E.

2006-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-12-15

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

ATHLET. Mod 3.0 Cycle A. Validation

Energy Technology Data Exchange (ETDEWEB)

Lerchl, G.; Austregesilo, H.; Glaeser, H.; Hrubisko, M.; Luther, W.

2012-09-15

ATHLET is an advanced best-estimate code which has been initially developed for the simulation of design basis and beyond design basis accidents (without core degradation) in light water reactors, including VVER and RBMK reactors. Furthermore, this program version enables the simulation of further working fluids like helium and liquid metals. The one-dimensional, two-phase fluiddynamic models are based on a five-equation model supplemented by a full-range drift-flux model, including a dynamic mixture-level tracking capability. Moreover, a two-fluid model based on six conservation equations is provided. The heat conduction and heat transfer module allows a flexible simulation of fuel rods and structures. The nuclear heat generation is calculated by a point-kinetics or by a one-dimensional kinetics model. A general control simulation module is provided for a flexible modelling of BOP- and auxiliary plant systems. Systematic code validation is performed by GRS and independent organizations. This Validation Manual is the fourth volume of the ATHLET Code Documentation comprising four volumes. This manual presents an overview about the complete ATHLET validation effort spent up to now. In addition, the results of five test cases simulated with the present ATHLET program version are compared with the experimental data.

Partially closed fuel cycle of WWER-440

International Nuclear Information System (INIS)

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

2002-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-09-01

specific plant status. The review of the safety features of 'small series' WWER-1000 plants shows that the main safety concept of these reactors is similar to that of model 320 with respect to the nuclear island arrangement, the amount of safety systems and the main process parameters of the primary and secondary circuits. However, the 'small series' WWER-1000 plants have major deficiencies such as a lack of separation of redundant safety systems and a single set of the reactor protection system for technological parameters which do not meet the current national standards and international practice. Differences in engineering design solutions, quality of manufacture and reliability of equipment have been revealed as deficiencies. About one third of the design safety issues have been identified by operational experience. The majority of safety issues have been identified as deviations from current standards and practices which have evolved since the WWER-1000 NPPs were designed. Much of the backfitting and upgrading work recognized as being required has been or is being performed. This activity was initiated by the WWER Owners Group and since the early 1990s international assistance has played an important role in the process of safety improvement of these NPPs. The current status of plant specific backfitting varies from site to site, depending on national regulatory requirements and the available financial means. The review of the plant specific status also indicates that certain safety issues have already been solved in some units. This report presents the information currently available to the IAEA on safety issues and safety improvement measures in 'small series' WWER-1000 plants. The IAEA intends to update this information regularly and make it available to the interested parties as part of the technical database developed within the framework of the Extra budgetary Programme on the Safety of WWER and RBMK NPPs.

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

International Nuclear Information System (INIS)

2000-09-01

specific plant status. The review of the safety features of 'small series' WWER-1000 plants shows that the main safety concept of these reactors is similar to that of model 320 with respect to the nuclear island arrangement, the amount of safety systems and the main process parameters of the primary and secondary circuits. However, the 'small series' WWER-1000 plants have major deficiencies such as a lack of separation of redundant safety systems and a single set of the reactor protection system for technological parameters which do not meet the current national standards and international practice. Differences in engineering design solutions, quality of manufacture and reliability of equipment have been revealed as deficiencies. About one third of the design safety issues have been identified by operational experience. The majority of safety issues have been identified as deviations from current standards and practices which have evolved since the WWER-1000 NPPs were designed. Much of the backfitting and upgrading work recognized as being required has been or is being performed. This activity was initiated by the WWER Owners Group and since the early 1990s international assistance has played an important role in the process of safety improvement of these NPPs. The current status of plant specific backfitting varies from site to site, depending on national regulatory requirements and the available financial means. The review of the plant specific status also indicates that certain safety issues have already been solved in some units. This report presents the information currently available to the IAEA on safety issues and safety improvement measures in 'small series' WWER-1000 plants. The IAEA intends to update this information regularly and make it available to the interested parties as part of the technical database developed within the framework of the Extra budgetary Programme on the Safety of WWER and RBMK NPPs

Improvements of radioactive waste management at WWER nuclear power plants

International Nuclear Information System (INIS)

2006-04-01

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

Current state of WWER SNF storage in Russia and the perspectives

International Nuclear Information System (INIS)

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

2006-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-31

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

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

Integrated computer codes for nuclear power plant severe accident analysis

Energy Technology Data Exchange (ETDEWEB)

Jordanov, I; Khristov, Y [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika

1996-12-31

This overview contains a description of the Modular Accident Analysis Program (MAAP), ICARE computer code and Source Term Code Package (STCP). STCP is used to model TMLB sample problems for Zion Unit 1 and WWER-440/V-213 reactors. Comparison is made of STCP implementation on VAX and IBM systems. In order to improve accuracy, a double precision version of MARCH-3 component of STCP is created and the overall thermal hydraulics is modelled. Results of modelling the containment pressure, debris temperature, hydrogen mass are presented. 5 refs., 10 figs., 2 tabs.

Integrated computer codes for nuclear power plant severe accident analysis

International Nuclear Information System (INIS)

Jordanov, I.; Khristov, Y.

1995-01-01

This overview contains a description of the Modular Accident Analysis Program (MAAP), ICARE computer code and Source Term Code Package (STCP). STCP is used to model TMLB sample problems for Zion Unit 1 and WWER-440/V-213 reactors. Comparison is made of STCP implementation on VAX and IBM systems. In order to improve accuracy, a double precision version of MARCH-3 component of STCP is created and the overall thermal hydraulics is modelled. Results of modelling the containment pressure, debris temperature, hydrogen mass are presented. 5 refs., 10 figs., 2 tabs

Assessment of systems codes and their coupling with CFD codes in thermal–hydraulic applications to innovative reactors

Energy Technology Data Exchange (ETDEWEB)

Bandini, G., E-mail: giacomino.bandini@enea.it [Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA) (Italy); Polidori, M. [Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA) (Italy); Gerschenfeld, A.; Pialla, D.; Li, S. [Commissariat à l’Energie Atomique (CEA) (France); Ma, W.M.; Kudinov, P.; Jeltsov, M.; Kööp, K. [Royal Institute of Technology (KTH) (Sweden); Huber, K.; Cheng, X.; Bruzzese, C.; Class, A.G.; Prill, D.P. [Karlsruhe Institute of Technology (KIT) (Germany); Papukchiev, A. [Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) (Germany); Geffray, C.; Macian-Juan, R. [Technische Universität München (TUM) (Germany); Maas, L. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN) (France)

2015-01-15

Highlights: • The assessment of RELAP5, TRACE and CATHARE system codes on integral experiments is presented. • Code benchmark of CATHARE, DYN2B, and ATHLET on PHENIX natural circulation experiment. • Grid-free pool modelling based on proper orthogonal decomposition for system codes is explained. • The code coupling methodologies are explained. • The coupling of several CFD/system codes is tested against integral experiments. - Abstract: The THINS project of the 7th Framework EU Program on nuclear fission safety is devoted to the investigation of crosscutting thermal–hydraulic issues for innovative nuclear systems. A significant effort in the project has been dedicated to the qualification and validation of system codes currently employed in thermal–hydraulic transient analysis for nuclear reactors. This assessment is based either on already available experimental data, or on the data provided by test campaigns carried out in the frame of THINS project activities. Data provided by TALL and CIRCE facilities were used in the assessment of system codes for HLM reactors, while the PHENIX ultimate natural circulation test was used as reference for a benchmark exercise among system codes for sodium-cooled reactor applications. In addition, a promising grid-free pool model based on proper orthogonal decomposition is proposed to overcome the limits shown by the thermal–hydraulic system codes in the simulation of pool-type systems. Furthermore, multi-scale system-CFD solutions have been developed and validated for innovative nuclear system applications. For this purpose, data from the PHENIX experiments have been used, and data are provided by the tests conducted with new configuration of the TALL-3D facility, which accommodates a 3D test section within the primary circuit. The TALL-3D measurements are currently used for the validation of the coupling between system and CFD codes.

Structural and leakage integrity assessment of WWER steam generator tubes

International Nuclear Information System (INIS)

Splichal, K.; Otruba, J.; Keilova, E.; Krhounek, V.; Turek, J.

1996-01-01

The leakage and plugging limits were derived for WWER steam generators based on leak and burst tests using tubes with axial part-through and through-wall defects. The following conclusions were arrived at: (i) The permissible primary-to-secondary leak rate with respect to the permissible through-wall defect size of WWER-440 and WWER-1000 steam generator tubes is 8 l/h. (ii) The primary-to-secondary leak rate is reduced by the blocking of the tube cracks by corrosion product particles and other substances. (iii) The rate of crack penetration through the tube wall is higher than the crack widening. (iv) The validity of the criterion of instability for tubes with through-wall cracks was confirmed experimentally. For the WWER-440 and WWER-1000 steam generators, the critical size of axial through-wall cracks, for the threshold primary-to-secondary pressure difference, is 13.8 and 12.0 mm, respectively. (v) The calculated leakage for the rupture of one tube and for the assumed extreme defects is two orders and one order of magnitude, respectively, higher than the proposed primary water leakage limit of 8 l/h. (vi) The experiments gave evidence that the use of the permissible thinning limit of 80% for the heat exchange tube plugging does not bring about uncontrollable leakage or unstable crack growth. This is consistent with experience gained at WWER-440 type nuclear power plants. 4 tabs., 5 figs., 9 refs

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Saturated steam turbines for power reactors of WWER-type

International Nuclear Information System (INIS)

Czwiertnia, K.

1978-01-01

The publication deals with design problems of large turbines for saturated steam and with problem of output limitations of single shaft normal speed units. The possibility of unification of conventional and nuclear turbines, which creates the economic basis for production of both types of turbines by one manufacturer based on standarized elements and assemblies is underlined. As separate problems the distribution of nuclear district heating power systems are considered. The choice of heat diagram for district heating saturated steam turbines, the advantages of different diagrams and evaluaton for further development are presented. On this basis a program of unified turbines both condensing and district heating type suitable for Soviet reactors of WWER-440 and WWER-1000 type for planned development of nuclear power in Poland is proposed. (author)

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Ammonia role in WWER primary circuit water chemistry optimization

International Nuclear Information System (INIS)

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

1999-01-01

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

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

International Nuclear Information System (INIS)

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

1996-01-01

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

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

International Nuclear Information System (INIS)

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

2006-01-01

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

Results of the recent investigations of the accuracy of C-PORCA code

International Nuclear Information System (INIS)

Pos, I.

2001-01-01

During the last year it was an essential demand to establish the C-PORCA code accuracy concerning the newly developed BNFL fuel. As it is well known the BNFL fuel assembly developed for the NPP Paks contains fuel pins with gadolinium. These pins are situated near the central tube so their unusual effects to be investigated. At the present investigation the C-PORCA code and the APA-H Westinghouse code system results were compared. During this comparison the differences between node and pin wise power distributions have been calculated as well. The fulfilled test have reinforced the applicability and capability of the C-PORCA code for WWER-440 core analysis (Authors)

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

International Nuclear Information System (INIS)

2007-12-01

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

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

International Nuclear Information System (INIS)

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

1993-01-01

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

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

International Nuclear Information System (INIS)

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

1993-01-01

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

WWER NPPs fuel handling machine control system

International Nuclear Information System (INIS)

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

2001-01-01

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

"Athletes¹ Perspectives on WADA and the Code"

DEFF Research Database (Denmark)

Gleaves, John; Christiansen, Ask Vest

2018-01-01

raised serious concerns on four specific issues, namely on a), the fair and consistent application of the rules for all athletes in a sport, b) the obligation to submit their whereabouts, c) the due processes when facing accusations of violations and d) the possibility for athletes to genuinely...

Mechanistic modelling of gaseous fission product behaviour in UO2 fuel by Rtop code

International Nuclear Information System (INIS)

Kanukova, V.D.; Khoruzhii, O.V.; Kourtchatov, S.Y.; Likhanskii, V.V.; Matveew, L.V.

2002-01-01

The current status of a mechanistic modelling by the RTOP code of the fission product behaviour in polycrystalline UO 2 fuel is described. An outline of the code and implemented physical models is presented. The general approach to code validation is discussed. It is exemplified by the results of validation of the models of fuel oxidation and grain growth. The different models of intragranular and intergranular gas bubble behaviour have been tested and the sensitivity of the code in the framework of these models has been analysed. An analysis of available models of the resolution of grain face bubbles is also presented. The possibilities of the RTOP code are presented through the example of modelling behaviour of WWER fuel over the course of a comparative WWER-PWR experiment performed at Halden and by comparison with Yanagisawa experiments. (author)

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

International Nuclear Information System (INIS)

2003-11-01

focuses on the issues of importance for the review of fuel safety cases. This report makes frequent reference to three reports which constitute the background for the workshops, two by the IAEA and one by the NEA/CSNI, these three reports have been included on a CD ROM that complements this report. The workshops were organized under the IAEA regional Technical Co-operation Project on Licensing Fuel and Fuel Modelling Codes for WWER Reactors

New designs of medium power WWER reactor plants

International Nuclear Information System (INIS)

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

2010-01-01

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

Large-scale experimental facility for emergency condition investigation of a new generation NPP WWER-640 reactor with passive safety systems

Energy Technology Data Exchange (ETDEWEB)

Aniskevich, Y.N.; Vasilenko, V.A.; Zasukha, V.K.; Migrov, Y.A.; Khabensky, V.B. [Research Inst. of Technology NITI (Russian Federation)

1997-12-31

The creation of the large-scale integral experimental facility (KMS) is specified by the programme of the experimental investigations to justify the engineering decisions on the safety of the design of the new generation NPP with the reactor WWER-640. The construction of KMS in a full volume will allow to conduct experimental investigations of all physical phenomena and processes, practically, occurring during the accidents on the NPPs with the reactor of WWER type and including the heat - mass exchange processes with low rates of the coolant, which is typical during the utilization of the passive safety systems, process during the accidents with a large leak, and also the complex intercommunicated processes in the reactor unit, passive safety systems and in the containment with the condition of long-term heat removal to the final absorber. KMS is being constructed at the Research Institute of Technology (NITI), Sosnovy Bor, Leningrad region, Russia. (orig.). 5 refs.

Large-scale experimental facility for emergency condition investigation of a new generation NPP WWER-640 reactor with passive safety systems

International Nuclear Information System (INIS)

Aniskevich, Y.N.; Vasilenko, V.A.; Zasukha, V.K.; Migrov, Y.A.; Khabensky, V.B.

1997-01-01

The creation of the large-scale integral experimental facility (KMS) is specified by the programme of the experimental investigations to justify the engineering decisions on the safety of the design of the new generation NPP with the reactor WWER-640. The construction of KMS in a full volume will allow to conduct experimental investigations of all physical phenomena and processes, practically, occurring during the accidents on the NPPs with the reactor of WWER type and including the heat - mass exchange processes with low rates of the coolant, which is typical during the utilization of the passive safety systems, process during the accidents with a large leak, and also the complex intercommunicated processes in the reactor unit, passive safety systems and in the containment with the condition of long-term heat removal to the final absorber. KMS is being constructed at the Research Institute of Technology (NITI), Sosnovy Bor, Leningrad region, Russia. (orig.)

Large-scale experimental facility for emergency condition investigation of a new generation NPP WWER-640 reactor with passive safety systems

Energy Technology Data Exchange (ETDEWEB)

Aniskevich, Y N; Vasilenko, V A; Zasukha, V K; Migrov, Y A; Khabensky, V B [Research Inst. of Technology NITI (Russian Federation)

1998-12-31

The creation of the large-scale integral experimental facility (KMS) is specified by the programme of the experimental investigations to justify the engineering decisions on the safety of the design of the new generation NPP with the reactor WWER-640. The construction of KMS in a full volume will allow to conduct experimental investigations of all physical phenomena and processes, practically, occurring during the accidents on the NPPs with the reactor of WWER type and including the heat - mass exchange processes with low rates of the coolant, which is typical during the utilization of the passive safety systems, process during the accidents with a large leak, and also the complex intercommunicated processes in the reactor unit, passive safety systems and in the containment with the condition of long-term heat removal to the final absorber. KMS is being constructed at the Research Institute of Technology (NITI), Sosnovy Bor, Leningrad region, Russia. (orig.). 5 refs.

WWER fuel: Results of post irradiation examination

International Nuclear Information System (INIS)

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

2006-01-01

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

CFD and system analysis code investigations of the multidimensional flow mixing phenomena in the reactor pressure vessel

International Nuclear Information System (INIS)

Ceuca, S.C.; Herb, J.; Schoeffel, P.J.; Hollands, T.; Austregesilo, H.; Hristov, H.V.

2017-01-01

The realistic numerical prediction of transient fluid-dynamic scenarios including the complex, three-dimensional flow mixing phenomena occurring in the reactor pressure vessel (RPV) both in normal or abnormal operation are an important issue in today's reactor safety assessment studies. Both Computational Fluid Dynamics (CFD) tools as well as fluid-dynamic system analysis codes, each with its advantages and drawbacks, are commonly used to model such transients. Simulation results obtained with the open-source CFD tool-box OpenFOAM and the German thermal-hydraulic system code ATHLET (Analysis of THermal-hydraulics of LEaks and Transients), the later developed by Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) for the analysis of the whole spectrum of operational transients, design-basis accidents and beyond design basis accidents anticipated for nuclear energy facilities, are compared against experimental data from the ROssendorf Coolant Mixing (ROCOM) test facility. In the case of the OpenFOAM CFD simulations the influence of various turbulence models and numerical schemes has been assessed while in the case of the system analysis code ATHLET a multidimensional nodalization recommended for real power plant applications has been employed. The simulation results show a good agreement with the experimental data, indicating that both OpenFOAM and ATHLET can capture the key flow features of the mixing processes in the Reactor Pressure Vessel (RPV). (author)

Modeling bubble condenser containment with computer code COCOSYS: post-test calculations of the main steam line break experiment at ELECTROGORSK BC V-213 test facility

International Nuclear Information System (INIS)

Lola, I.; Gromov, G.; Gumenyuk, D.; Pustovit, V.; Sholomitsky, S.; Wolff, H.; Arndt, S.; Blinkov, V.; Osokin, G.; Melikhov, O.; Melikhov, V.; Sokoline, A.

2005-01-01

Containment of the WWER-440 Model 213 nuclear power plant features a Bubble Condenser, a complex passive pressure suppression system, intended to limit pressure rise in the containment during accidents. Due to lack of experimental evidence of its successful operation in the original design documentation, the performance of this system under accidents with ruptures of large high-energy pipes of the primary and secondary sides remains a known safety concern for this containment type. Therefore, a number of research and analytical studies have been conducted by the countries operating WWER-440 reactors and their Western partners in the recent years to verify Bubble Condenser operation under accident conditions. Comprehensive experimental research studies at the Electrogorsk BC V-213 test facility, commissioned in 1999 in Electrogorsk Research and Engineering Centre (EREC), constitute essential part of these efforts. Nowadays this is the only operating large-scale facility enabling integral tests on investigation of the Bubble Condenser performance. Several large international research projects, conducted at this facility in 1999-2003, have covered a spectrum of pipe break accidents. These experiments have substantially improved understanding of the overall system performance and thermal hydraulic phenomena in the Bubble Condenser Containment, and provided valuable information for validating containment codes against experimental results. One of the recent experiments, denoted as SLB-G02, has simulated steam line break. The results of this experiment are of especial value for the engineers working in the area of computer code application for WWER-440 containment analyses, giving an opportunity to verify validity of the code predictions and identify possibilities for model improvement. This paper describes the results of the post-test calculations of the SLB-G02 experiment, conducted as a joint effort of GRS, Germany and Ukrainian technical support organizations for

Status and prospects of WWER in-core fuel management activities

International Nuclear Information System (INIS)

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

1994-01-01

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

Status and prospects of WWER in-core fuel management activities

Energy Technology Data Exchange (ETDEWEB)

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

1994-12-31

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

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

International Nuclear Information System (INIS)

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

1985-01-01

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

WWER-1000 fuel cycles: current situation and outlook

International Nuclear Information System (INIS)

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

2013-01-01

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

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

Science.gov (United States)

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

2016-04-01

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

Fuel improvement and WWER-1000 FA main operational results

International Nuclear Information System (INIS)

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

2003-01-01

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

Simulation of spray phenomena using the containment code system COCOSYS. 1{sup st} Technical report.Validation and interpretation of selected models and of the coupling of the system codes ATHLET-CD and COCOSYS (VAMKoS); Simulation von Spruehstrahlphaenomenen mit dem Containment Code System COCOSYS. 1. Technischer Fachbericht. Validierung und Analyse ausgewaehlter Modelle sowie der Kopplung der Systemcodes ATHLET-CD und COCOSYS (VAMKoS)

Energy Technology Data Exchange (ETDEWEB)

Risken, Tobias; Koch, Marco K.

2014-12-15

The present report is the first Technical Report within the research project ''Validation and interpretation of selected models and of the coupling of the system codes ATHLET-CD and COCOSYS'', funded by the German Federal Ministry for Economic Affairs and Energy (BMWi 1501465) and projected at the Reactor Simulation and Safety Group, Chair of Energy Systems and Energy Economics (LEE) at the Ruhr-Universitaet Bochum (RUB). This report deals with the simulation of spray phenomena with the containment code system COCOSYS, which is developed by the German Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH. First, post-test calculations of the OECD THAI-2 tests HD-30 and HD-31 are presented. The simulation results are compared to experimental values and thereby assessed. The analysis focuses an the assessment of the simultaneous use of the COCOSYS models IVO (spray model) and FRONT (combustion model) as well as a spray entrainment model developed at RUB regarding the simulation of the phenomena related to the interaction of spray and combustion processes. The simulation results show the necessity to consider the induced turbulences. The simulation of these turbulences is performed by modifying the FRONT input parameters leading to an improvement of the simulation results. The consideration of the entrainment positively influences the simulated flow pattern. Subsequently, the simulation of the entrainment of interacting sprays, as occurring in containment spray systems, is considered. The entrainment of interacting sprays is influenced by droplet collisions and changes of the drag between the droplets and the atmosphere. For the simulation an entrainment factor, which has to be determined externally, is implemented into COCOSYS. Exemplary simulations of the OECD SETH-2 ST3 tests show, that in general the use of entrainment factors enables the calculation of alternated gas distributions.

Equipment and building structures ageing management for WWER type NPPs

International Nuclear Information System (INIS)

Mayboroda, O.

2001-01-01

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

Surveillance specimen programmes for WWER reactor vessels in the Czech Republic

International Nuclear Information System (INIS)

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

2003-01-01

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

Steady State and Transient Fuel Rod Performance Analyses by Pad and Transuranus Codes

International Nuclear Information System (INIS)

Slyeptsov, O.; Slyeptsov, S.; Kulish, G.; Ostapov, A.; Chernov, I.

2013-01-01

The report performed under IAEA research contract No.15370/L2 describes the analysis results of WWER and PWR fuel rod performance at steady state operation and transients by means of PAD and TRANSURANUS codes. The code TRANSURANUS v1m1j09 developed by Institute for of Transuranium Elements (ITU) was used based on the Licensing Agreement N31302. The code PAD 4.0 developed by Westinghouse Electric Company was utilized in the frame of the Ukraine Nuclear Fuel Qualification Project for safety substantiation for the use of Westinghouse fuel assemblies in the mixed core of WWER-1000 reactor. The experimental data for the Russian fuel rod behavior obtained during the steady-state operation in the WWER-440 core of reactor Kola-3 and during the power transients in the core of MIR research reactor were taken from the IFPE database of the OECD/NEA and utilized for assessing the codes themselves during simulation of such properties as fuel burnup, fuel centerline temperature (FCT), fuel swelling, cladding strain, fission gas release (FGR) and rod internal pressure (RIP) in the rod burnup range of (41 - 60) GWD/MTU. The experimental data of fuel behavior at steady-state operation during seven reactor cycles presented by AREVA for the standard PWR fuel rod design were used to examine the code FGR model in the fuel burnup range of (37 - 81) GWD/MTU. (author)

Performance of the Westinghouse WWER-1000 fuel design

International Nuclear Information System (INIS)

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

2011-01-01

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

Application of the Severe Accident Code ATHLET-CD. Coolant injection to primary circuit of a PWR by mobile pump system in case of SBLOCA severe accident scenario

Energy Technology Data Exchange (ETDEWEB)

Jobst, Matthias; Wilhelm, Polina; Kliem, Soeren; Kozmenkov, Yaroslav [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety

2017-06-01

The improvement of the safety of nuclear power plants is a continuously on-going process. The analysis of transients and accidents is an important research topic, which significantly contributes to safety enhancements of existing power plants. In case of an accident with multiple failures of safety systems, core uncovery and heat-up can occur. In order to prevent the accident to turn into a severe one or to mitigate the consequences of severe accidents, different accident management measures can be applied. By means of numerical analyses performed with the compute code ATHLET-CD, the effectiveness of coolant injection with a mobile pump system into the primary circuit of a PWR was studied. According to the analyses, such a system can stop the melt progression if it is activated prior to 10 % of total core is molten.

Application of the Severe Accident Code ATHLET-CD. Coolant injection to primary circuit of a PWR by mobile pump system in case of SBLOCA severe accident scenario

International Nuclear Information System (INIS)

Jobst, Matthias; Wilhelm, Polina; Kliem, Soeren; Kozmenkov, Yaroslav

2017-01-01

The improvement of the safety of nuclear power plants is a continuously on-going process. The analysis of transients and accidents is an important research topic, which significantly contributes to safety enhancements of existing power plants. In case of an accident with multiple failures of safety systems, core uncovery and heat-up can occur. In order to prevent the accident to turn into a severe one or to mitigate the consequences of severe accidents, different accident management measures can be applied. By means of numerical analyses performed with the compute code ATHLET-CD, the effectiveness of coolant injection with a mobile pump system into the primary circuit of a PWR was studied. According to the analyses, such a system can stop the melt progression if it is activated prior to 10 % of total core is molten.

Improving the ALUeS diagnostic system for determining the coolant leak place from the WWER-440 primary circuit

International Nuclear Information System (INIS)

Markosyan, G.R.; Petrosyan, V.G.; Shakhverdyan, S.V.; Aslanyan, M.A.

2000-01-01

The new algorithm for localizing the leakage from the WWER-440 primary circuit, intended for operation in the Siemens ALUeS system, is proposed. The results of the algorithm realization in the leakage control system (the ALUeS system copy), installed at the Armenian NPP power unit-2, are presented. The leakage localization algorithm proposed was tested in other experiments. The leakage position in the majority of cases is determined exactly. Small (up to 5 m) deviations, the cause whereof were incorrect readings of the transducers, were observed [ru

Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 5C. Experience data. Working material

International Nuclear Information System (INIS)

1999-01-01

In August 1991, following the SMiRT-11 Conference in Tokyo, a Technical Committee Meeting was held on the 'Seismic safety issues relating to existing NPPs'. The Proceedings of this TCM was subsequently compiled in an IAEA Working Material. One of the main recommendations of this TCM, called for the harmonization of criteria and methods used in Member States in seismic reassessment and upgrading of existing NPPs. Twenty four institutions from thirteen countries participated in the CRP named 'Benchmark study for the seismic analysis and testing of WWER type NPPs'. Two types of WWER reactors (WWER-1000 and WWER-440/213) selected for benchmarking. Kozloduy NPP Units 5/6 and Paks NPP represented these respectively as prototypes. Consistent with the recommendations of the TCM and the working paper prepared by the subsequent Consultants' Meeting, the focal activity of the CRP was the benchmarking exercises. A similar methodology was followed both for Paks NPP and Kozloduy NPP Unit 5. Firstly, the NPP (mainly the reactor building) was tested using a blast loading generated by a series of explosions from buried TNT charges. Records from this test were obtained at several free field locations (both downhole and surface), foundation mat, various elevations of structures as well as some tanks and the stack. Then the benchmark participants were provided with structural drawings, soil data and the free field record of the blast experiment. Their task was to make a blind prediction of the response at preselected locations. The analytical results from these participants were then compared with the results from the test. Although the benchmarking exercises constituted the focus of the CRP, there were many other interesting problems related to the seismic safety of WWER type NPPs which were addressed by the participants. These involved generic studies, i.e. codes and standards used in original WWER designs and their comparison with current international practice; seismic analysis

ATHLET validation using accident management experiments

Energy Technology Data Exchange (ETDEWEB)

Teschendorff, V.; Glaeser, H.; Steinhoff, F. [Gasellschaft fuer Anlagen - und Reaktorsicherheit (GSR) mbH, Garching (Germany)

1995-09-01

The computer code ATHLET is being developed as an advanced best-estimate code for the simulation of leaks and transients in PWRs and BWRs including beyond design basis accidents. The code has features that are of special interest for applications to small leaks and transients with accident management, e.g. initialisation by a steady-state calculation, full-range drift-flux model, and dynamic mixture level tracking. The General Control Simulation Module of ATHLET is a flexible tool for the simulation of the balance-of-plant and control systems including the various operator actions in the course of accident sequences with AM measures. The systematic validation of ATHLET is based on a well balanced set of integral and separate effect tests derived from the CSNI proposal emphasising, however, the German combined ECC injection system which was investigated in the UPTF, PKL and LOBI test facilities. PKL-III test B 2.1 simulates a cool-down procedure during an emergency power case with three steam generators isolated. Natural circulation under these conditions was investigated in detail in a pressure range of 4 to 2 MPa. The transient was calculated over 22000 s with complicated boundary conditions including manual control actions. The calculations demonstrations the capability to model the following processes successfully: (1) variation of the natural circulation caused by steam generator isolation, (2) vapour formation in the U-tubes of the isolated steam generators, (3) break-down of circulation in the loop containing the isolated steam generator following controlled cool-down of the secondary side, (4) accumulation of vapour in the pressure vessel dome. One conclusion with respect to the suitability of experiments simulating AM procedures for code validation purposes is that complete documentation of control actions during the experiment must be available. Special attention should be given to the documentation of operator actions in the course of the experiment.

Development of the temperature field at the WWER-440 core outlet monitoring system and application of the data analyses methods

International Nuclear Information System (INIS)

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

2001-01-01

On-line internal reactor monitoring by 216 thermal couples, located at the reactor core outlet, is carried out during power operation of WWER-440 Units 1 and 2 at Kozloduy NPP. Automatic monitoring of technology process is performed by IB-500MA, which collects and performs initial data processing (discrediting and conversion of analogue signals into digital mode). The paper also presents the results and analyses of power distribution monitoring during the past 21-th and current 22-th fuel cycle at Kozloduy NPP, Unit 1 by using archiving system capacity and related software. The possibility to perform operational assessment and analysis of power distribution in the reactor core in each point of the fuel cycle is checked by comparison of the neutron-physical calculation results with reactor coolant system parameters. Paper shows that the processing and analysis of accumulated significant amount of data in the archive files increases accuracy and reliability of power distribution monitoring in the reactor core in each moment of the fuel cycle of WWER-440 reactors at Kozloduy NPP

Concentration of WWER-1000 unit power on one site

International Nuclear Information System (INIS)

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

1987-01-01

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

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

International Nuclear Information System (INIS)

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

2001-01-01

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

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

International Nuclear Information System (INIS)

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

2003-01-01

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

Fuel pin failure root causes and power distribution gradients in WWER cores

International Nuclear Information System (INIS)

Mikus, J.

2008-01-01

The purpose of this work is to investigate the influence of some core heterogeneities and reactor construction materials on space power distribution in WWER type cores, especially from viewpoint of the values and gradient occurrence that could result in static loads with some consequences, e.g., fuel pin (FP) or fuel assembly (FA) bowing and possible contribution to the FP failure root causes. Presented information were obtained by means of experiments on research reactor LR-0 concerning the: 1) Power distribution estimation on pellet surface of the FPs neighbouring a FP containing gadolinium (Gd 2 O 3 ) burnable absorber integrated into fuel in WWER-440 and -1000 type cores; 2) Power distribution measurement in periphery FAs neighbouring the baffle in WWER-1000 type cores and 3) Power distribution in FAs neighbouring the control rod absorbing part in a WWER-440 type core. (author)

Fuel behaviour calculations with version 2.0 of the code FUROM

International Nuclear Information System (INIS)

Kulacsy, K.

2011-01-01

The fuel modelling code FUROM (FUel ROd Model), suitable for calculating the normal operation condition behaviour of PWR and WWER fuels, has been developed at AEKI for several years. In 2010 the new version of the code, FUROM-2.0 was released. Calculations performed with this version and results are presented. (author)

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

International Nuclear Information System (INIS)

Titov, A.A.

1983-01-01

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

Water chemistry in WWER reactors

International Nuclear Information System (INIS)

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

1997-01-01

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

Water chemistry in WWER reactors

Energy Technology Data Exchange (ETDEWEB)

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

1997-02-01

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

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

International Nuclear Information System (INIS)

Sabotinov, L.; Cadet-Mercier, S.

2001-01-01

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

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)

Review of RIA and LOCA criteria for WWER fuel

International Nuclear Information System (INIS)

Hozer, Z.

2006-01-01

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

Post test calculation of the experiment 'small break loss-of- coolant test' SBL-22 at the Finnish integral test facility PACTEL with the thermohydraulic code ATHLET

International Nuclear Information System (INIS)

Lischke, W.; Vandreier, B.

1997-01-01

At the University for Applied Sciences Zittau/Goerlitz (FH) calculations for the verification of the ATHLET-code for reactors of type VVER are carried out since 1991, sponsored by the German Ministry for Education, Science and Technology (BMBF). The special features of these reactors in comparison to reactors of western countries are characterized by the duct route of reactor coolant pipes and the horizontal steam generators. Because of these special features, a check of validity of the ATHLET-models is necessary. For further verification of the ATHLET-code the post test calculation of the experiment SBL-22 (Small break loss-of-coolant test) realized at the finnish facility PACTEL was carried out. The experiment served for the examination of the natural circulation behaviour of the loop over a continuous range of primary side water inventory

Comparative calculations on selected two-phase flow phenomena using major PWR system codes

International Nuclear Information System (INIS)

1990-01-01

In 1988 a comparative study on important features and models in six major best estimate thermal hydraulic codes for PWR systems was implemented (Comparison of thermal hydraulic safety codes for PWR Graham, Trotman, London, EUR 11522). It was a limitation of that study that the source codes themselves were not available but the comparison had to be based on the available documentation. In the present study, the source codes were available and the capability of four system codes to predict complex two-phase flow phenomena has been assessed. Two areas of investigation were selected: (a) pressurized spray phenomena; (b) boil-up phenomena in rod bundles. As regards the first area, experimental data obtained in 1972 on the Neptunus Facility (Delft University of Technology) were compared with the results of the calculations using Athlet, Cathare, Relap 5 and TRAC-PT1 and, concerning the second area, the results of two experimental facilities obtained in 1980 and 1985 on Thetis (UKEA) and Pericles (CEA-Grenoble) were considered

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

International Nuclear Information System (INIS)

Szabo, B.K.

1988-07-01

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

Some consideration on reviewing of WWER

Energy Technology Data Exchange (ETDEWEB)

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

1993-07-01

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

Some consideration on reviewing of WWER

International Nuclear Information System (INIS)

Shibata, Heki

1993-01-01

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

Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 4F. Paks NPP: Analysis and testing. Working material

International Nuclear Information System (INIS)

1999-01-01

In August 1991, following the SMiRT-11 Conference in Tokyo, a Technical Committee Meeting was held on the 'Seismic safety issues relating to existing NPPs'. The Proceedings of this TCM was subsequently compiled in an IAEA Working Material. One of the main recommendations of this TCM, called for the harmonization of criteria and methods used in Member States in seismic reassessment and upgrading of existing NPPs. Twenty four institutions from thirteen countries participated in the CRP named 'Benchmark study for the seismic analysis and testing of WWER type NPPs'. Two types of WWER reactors (WWER-1000 and WWER-440/213) selected for benchmarking. Kozloduy NPP Units 5/6 and Paks NPP represented these respectively as prototypes. Consistent with the recommendations of the TCM and the working paper prepared by the subsequent Consultants' Meeting, the focal activity of the CRP was the benchmarking exercises. A similar methodology was followed both for Paks NPP and Kozloduy NPP Unit 5. Firstly, the NPP (mainly the reactor building) was tested using a blast loading generated by a series of explosions from buried TNT charges. Records from this test were obtained at several free field locations (both downhole and surface), foundation mat, various elevations of structures as well as some tanks and the stack. Then the benchmark participants were provided with structural drawings, soil data and the free field record of the blast experiment. Their task was to make a blind prediction of the response at preselected locations. The analytical results from these participants were then compared with the results from the test. Although the benchmarking exercises constituted the focus of the CRP, there were many other interesting problems related to the seismic safety of WWER type NPPs which were addressed by the participants. These involved generic studies, i.e. codes and standards used in original WWER designs and their comparison with current international practice; seismic analysis

Risk assessment basis for WWER-440 spent nuclear fuel

International Nuclear Information System (INIS)

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

2000-01-01

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

Advanced fuel cycles for WWER-1000 reactors

International Nuclear Information System (INIS)

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

2007-01-01

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

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

International Nuclear Information System (INIS)

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

2002-01-01

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Seismic resistance of WWER equipment (systematization and generalization)

International Nuclear Information System (INIS)

Kaznovski, S.; Chechenov, H.

1999-01-01

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

Low back pain status in elite and semi-elite Australian football codes: a cross-sectional survey of football (soccer, Australian rules, rugby league, rugby union and non-athletic controls

Directory of Open Access Journals (Sweden)

McHardy Andrew

2009-04-01

Full Text Available Abstract Background Our understanding of the effects of football code participation on low back pain (LBP is limited. It is unclear whether LBP is more prevalent in athletic populations or differs between levels of competition. Thus it was the aim of this study to document and compare the prevalence, intensity, quality and frequency of LBP between elite and semi-elite male Australian football code participants and a non-athletic group. Methods A cross-sectional survey of elite and semi-elite male Australian football code participants and a non-athletic group was performed. Participants completed a self-reported questionnaire incorporating the Quadruple Visual Analogue Scale (QVAS and McGill Pain Questionnaire (short form (MPQ-SF, along with additional questions adapted from an Australian epidemiological study. Respondents were 271 elite players (mean age 23.3, range 17–39, 360 semi-elite players (mean age 23.8, range 16–46 and 148 non-athletic controls (mean age 23.9, range 18–39. Results Groups were matched for age (p = 0.42 and experienced the same age of first onset LBP (p = 0.40. A significant linear increase in LBP from the non-athletic group, to the semi-elite and elite groups for the QVAS and the MPQ-SF was evident (p Conclusion Foolers in Australia have significantly more severe and frequent LBP than a non-athletic group and this escalates with level of competition.

Temperature etalon of WWER-440 reactor

International Nuclear Information System (INIS)

Stanc, S.; Slanina, M.

2001-01-01

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

Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 3G. Kozloduy NPP units 5/6: Analysis and testing. Working material

International Nuclear Information System (INIS)

1999-01-01

In August 1991, following the SMiRT-11 Conference in Tokyo, a Technical Committee Meeting was held on the 'Seismic safety issues relating to existing NPPs'. The Proceedings of this TCM was subsequently compiled in an IAEA Working Material. One of the main recommendations of this TCM, called for the harmonization of criteria and methods used in Member States in seismic reassessment and upgrading of existing NPPs. Twenty four institutions from thirteen countries participated in the CRP named 'Benchmark study for the seismic analysis and testing of WWER type NPPs'. Two types of WWER reactors (WWER-1000 and WWER-440/213) selected for benchmarking. Kozloduy NPP Units 5/6 and Paks NPP represented these respectively as prototypes. Consistent with the recommendations of the TCM and the working paper prepared by the subsequent Consultants' Meeting, the focal activity of the CRP was the benchmarking exercises. A similar methodology was followed both for Paks NPP and Kozloduy NPP Unit 5. Firstly, the NPP (mainly the reactor building) was tested using a blast loading generated by a series of explosions from buried TNT charges. Records from this test were obtained at several free field locations (both downhole and surface), foundation mat, various elevations of structures as well as some tanks and the stack. Then the benchmark participants were provided with structural drawings, soil data and the free field record of the blast experiment. Their task was to make a blind prediction of the response at preselected locations. The analytical results from these participants were then compared with the results from the test. Although the benchmarking exercises constituted the focus of the CRP, there were many other interesting problems related to the seismic safety of WWER type NPPs which were addressed by the participants. These involved generic studies, i.e. codes and standards used in original WWER designs and their comparison with current international practice; seismic analysis

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

International Nuclear Information System (INIS)

Nemes, Imre

2000-01-01

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

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

International Nuclear Information System (INIS)

Pandey, Y.; Chauhan, A.

2008-01-01

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

Post test calculation of the experiment `small break loss-of- coolant test` SBL-22 at the Finnish integral test facility PACTEL with the thermohydraulic code ATHLET

Energy Technology Data Exchange (ETDEWEB)

Lischke, W.; Vandreier, B. [Univ. for Applied Sciences, Zittau/Goerlitz (Germany). Dept. of Nuclear Technology

1997-12-31

At the University for Applied Sciences Zittau/Goerlitz (FH) calculations for the verification of the ATHLET-code for reactors of type VVER are carried out since 1991, sponsored by the German Ministry for Education, Science and Technology (BMBF). The special features of these reactors in comparison to reactors of western countries are characterized by the duct route of reactor coolant pipes and the horizontal steam generators. Because of these special features, a check of validity of the ATHLET-models is necessary. For further verification of the ATHLET-code the post test calculation of the experiment SBL-22 (Small break loss-of-coolant test) realized at the finnish facility PACTEL was carried out. The experiment served for the examination of the natural circulation behaviour of the loop over a continuous range of primary side water inventory. 5 refs.

Post test calculation of the experiment `small break loss-of- coolant test` SBL-22 at the Finnish integral test facility PACTEL with the thermohydraulic code ATHLET

Energy Technology Data Exchange (ETDEWEB)

Lischke, W; Vandreier, B [Univ. for Applied Sciences, Zittau/Goerlitz (Germany). Dept. of Nuclear Technology

1998-12-31

At the University for Applied Sciences Zittau/Goerlitz (FH) calculations for the verification of the ATHLET-code for reactors of type VVER are carried out since 1991, sponsored by the German Ministry for Education, Science and Technology (BMBF). The special features of these reactors in comparison to reactors of western countries are characterized by the duct route of reactor coolant pipes and the horizontal steam generators. Because of these special features, a check of validity of the ATHLET-models is necessary. For further verification of the ATHLET-code the post test calculation of the experiment SBL-22 (Small break loss-of-coolant test) realized at the finnish facility PACTEL was carried out. The experiment served for the examination of the natural circulation behaviour of the loop over a continuous range of primary side water inventory. 5 refs.

Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 3I. Kozloduy NPP units 5/6: Analysis and testing. Working material

International Nuclear Information System (INIS)

1999-01-01

In August 1991, following the SMiRT-11 Conference in Tokyo, a Technical Committee Meeting was held on the 'Seismic safety issues relating to existing NPPs'. The Proceedings of this TCM was subsequently compiled in an IAEA Working Material. One of the main recommendations of this TCM, called for the harmonization of criteria and methods used in Member States in seismic reassessment and upgrading of existing NPPs. Twenty four institutions from thirteen countries participated in the CRP named 'Benchmark study for the seismic analysis and testing of WWER type NPPs'. Two types of WWER reactors (WWER-1000 and WWER-440/213) selected for benchmarking. Kozloduy NPP Units 5/6 and Paks NPP represented these respectively as prototypes. Consistent with the recommendations of the TCM and the working paper prepared by the subsequent Consultants' Meeting, the focal activity of the CRP was the benchmarking exercises. A similar methodology was followed both for Paks NPP and Kozloduy NPP Unit 5. Firstly, the NPP (mainly the reactor building) was tested using a blast loading generated by a series of explosions from buried TNT charges. Records from this test were obtained at several free field locations (both downhole and surface), foundation mat, various elevations of structures as well as some tanks and the stack. Then the benchmark participants were provided with structural drawings, soil data and the free field record of the blast experiment. Their task was to make a blind prediction of the response at preselected locations. The analytical results from these participants were then compared with the results from the test. Although the benchmarking exercises constituted the focus of the CRP, there were many other interesting problems related to the seismic safety of WWER type NPPs which were addressed by the participants. These involved generic studies, i.e. codes and standards used in original WWER designs and their comparison with current international practice; seismic analysis

Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 4G. Paks NPP: Analysis and testing. Working material

International Nuclear Information System (INIS)

1999-01-01

In August 1991, following the SMiRT-11 Conference in Tokyo, a Technical Committee Meeting was held on the 'Seismic safety issues relating to existing NPPs'. The Proceedings of this TCM was subsequently compiled in an IAEA Working Material. One of the main recommendations of this TCM, called for the harmonization of criteria and methods used in Member States in seismic reassessment and upgrading of existing NPPs. Twenty four institutions from thirteen countries participated in the CRP named 'Benchmark study for the seismic analysis and testing of WWER type NPPs'. Two types of WWER reactors (WWER-1000 and WWER-440/213) selected for benchmarking. Kozloduy NPP Units 5/6 and Paks NPP represented these respectively as prototypes. Consistent with the recommendations of the TCM and the working paper prepared by the subsequent Consultants' Meeting, the focal activity of the CRP was the benchmarking exercises. A similar methodology was followed both for Paks NPP and Kozloduy NPP Unit 5. Firstly, the NPP (mainly the reactor building) was tested using a blast loading generated by a series of explosions from buried TNT charges. Records from this test were obtained at several free field locations (both downhole and surface), foundation mat, various elevations of structures as well as some tanks and the stack. Then the benchmark participants were provided with structural drawings, soil data and the free field record of the blast experiment. Their task was to make a blind prediction of the response at preselected locations. The analytical results from these participants were then compared with the results from the test. Although the benchmarking exercises constituted the focus of the CRP, there were many other interesting problems related to the seismic safety of WWER type NPPs which were addressed by the participants. These involved generic studies, i.e. codes and standards used in original WWER designs and their comparison with current international practice; seismic analysis

Fuel assembly outlet temperature profile influence on core by-pass flow and power distribution determination in WWER -440 reactors

International Nuclear Information System (INIS)

Petenyi, V.; Klucarova, K.; Remis, J.

2003-01-01

The in core instrumentation of the WWER-440 reactors consists of the thermocouple system and the system of self powered detectors (SPD). The thermocouple systems are positioned about 50 cm above the fuel bundle upper flow-mixing grid. The usual assumption is that, the coolant is well mixed in the Tc location, i.e. the temperature is constant through the flow cross-section area. The present evaluations by using the FLUENT 5.5.14 code reveal that, this assumption is not fulfilled. There exists a temperature profile that depends on fuel assembly geometry and on inner power profile of the fuel assembly. The paper presents the estimation of this effect and its influence on the core power distribution and the core by-pass flow determination. Comparison with measurements in Mochovce NPP will also be a part of this presentation (Authors)

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

International Nuclear Information System (INIS)

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

1997-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-09-01

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

Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 4E. Paks NPP: Analysis and testing. Working material

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-07-01

In August 1991, following the SMiRT-11 Conference in Tokyo, a Technical Committee Meeting was held on the 'Seismic safety issues relating to existing NPPs'. The Proceedings of this TCM was subsequently compiled in an IAEA Working Material. One of the main recommendations of this TCM, called for the harmonization of criteria and methods used in Member States in seismic reassessment and upgrading of existing NPPs. Twenty four institutions from thirteen countries participated in the CRP named 'Benchmark study for the seismic analysis and testing of WWER type NPPs'. Two types of WWER reactors (WWER-1000 and WWER-440/213) selected for benchmarking. Kozloduy NPP Units 5/6 and Paks NPP represented these respectively as prototypes. Consistent with the recommendations of the TCM and the working paper prepared by the subsequent Consultants' Meeting, the focal activity of the CRP was the benchmarking exercises. A similar methodology was followed both for Paks NPP and Kozloduy NPP Unit 5. Firstly, the NPP (mainly the reactor building) was tested using a blast loading generated by a series of explosions from buried TNT charges. Records from this test were obtained at several free field locations (both downhole and surface), foundation mat, various elevations of structures as well as some tanks and the stack. Then the benchmark participants were provided with structural drawings, soil data and the free field record of the blast experiment. Their task was to make a blind prediction of the response at preselected locations. The analytical results from these participants were then compared with the results from the test. Although the benchmarking exercises constituted the focus of the CRP, there were many other interesting problems related to the seismic safety of WWER type NPPs which were addressed by the participants. These involved generic studies, i.e. codes and standards used in original WWER designs and their comparison with current international practice; seismic analysis

Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 4E. Paks NPP: Analysis and testing. Working material

International Nuclear Information System (INIS)

1995-01-01

In August 1991, following the SMiRT-11 Conference in Tokyo, a Technical Committee Meeting was held on the 'Seismic safety issues relating to existing NPPs'. The Proceedings of this TCM was subsequently compiled in an IAEA Working Material. One of the main recommendations of this TCM, called for the harmonization of criteria and methods used in Member States in seismic reassessment and upgrading of existing NPPs. Twenty four institutions from thirteen countries participated in the CRP named 'Benchmark study for the seismic analysis and testing of WWER type NPPs'. Two types of WWER reactors (WWER-1000 and WWER-440/213) selected for benchmarking. Kozloduy NPP Units 5/6 and Paks NPP represented these respectively as prototypes. Consistent with the recommendations of the TCM and the working paper prepared by the subsequent Consultants' Meeting, the focal activity of the CRP was the benchmarking exercises. A similar methodology was followed both for Paks NPP and Kozloduy NPP Unit 5. Firstly, the NPP (mainly the reactor building) was tested using a blast loading generated by a series of explosions from buried TNT charges. Records from this test were obtained at several free field locations (both downhole and surface), foundation mat, various elevations of structures as well as some tanks and the stack. Then the benchmark participants were provided with structural drawings, soil data and the free field record of the blast experiment. Their task was to make a blind prediction of the response at preselected locations. The analytical results from these participants were then compared with the results from the test. Although the benchmarking exercises constituted the focus of the CRP, there were many other interesting problems related to the seismic safety of WWER type NPPs which were addressed by the participants. These involved generic studies, i.e. codes and standards used in original WWER designs and their comparison with current international practice; seismic analysis

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

International Nuclear Information System (INIS)

Sykora, D.; Sykorova, I.

1988-01-01

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

Power peak in vicinity of WWER-440 control assembly

International Nuclear Information System (INIS)

Mikus, J.

2002-01-01

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

Project margins of advanced reactor design WWER-500

International Nuclear Information System (INIS)

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

1994-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-31

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

Integration of KESS III models in ATHLET-CD and contributions to program verification. Final report

International Nuclear Information System (INIS)

Bruder, M.; Schatz, A.

1994-07-01

The development of the computer code ATHLET-CD is a contribution to the reactor safety research. ATHLET-CD is an extension of the system code ATHLET by core degradation models especially of the modular software package KESS. The aim of the ATHLET-CD development is the simulation of severe accident sequences from their initialisation to severe core degradation in a continous manner. In the framework of this project the ATHLET-CD development has been focused on the integration of KESS model like the control rod model as well as the models describing chemical interactions and material relocation along a rod and fission product release. The present ATHLET-CD version is able to describe severe accidents in a PWR up to the early core degradation (relocation of material along a rod surface in axial direction). Contributions to the verification of ATHLET-CD comprised calculations of the experiments PHEBUS AIC and PBF SFD 1-4. The PHEBUS AIC calculation was focused on the examination of the control rod model whereas the PBF SFD 1-4 claculation served to check the models describing melting, material relocation and fission product release. (orig.)

Seismic margin assessment and earthquake experience based methods for WWER-440/213 type NPPs

International Nuclear Information System (INIS)

Masopust, R.

1996-01-01

This report covers the review of the already completed studies, namely, safe shutdown system identification and classification for Bohunice NPP and the comparative study of standards and criteria. It contains a report on currently ongoing studies concerning seismic margin assessment and earthquake experience based methods in application for seismic evaluation and verification of structures and equipment components of the operating WWER-440/213 type NPPs. This is based on experiences obtained from Paks NPP. The work plan for the remaining period of Benchmark CRP and the new proposals are included. These are concerned with seismic evaluation of selected safety related mechanical equipment and pipes of Paks NPP, and the actual seismic issues of the Temelin WWER-1000 type NPP

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

International Nuclear Information System (INIS)

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

2001-01-01

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

'AER Working Group D On WWER Safety Analysis' Report Of The Meeting In Pisa, Italy, 26-27 April 2006

International Nuclear Information System (INIS)

Siltanen, P.

2006-01-01

AER Working Group D on WWER reactor safety analysis held its 15 t h meeting in Grand Hotel Duomo in Pisa, Italy during the period 26-27 April 2006. The meeting was hosted by the University of Pisa following the fourth workshop on the OECD/DOE/CEA WWER-1000 Coolant Transient Benchmark (W1000-CT) held at the same location on 24-25 April. Altogether 15 participants attended the Working Group D meeting, 11 from AER member organizations and 4 guests from non-member organizations. The coordinator for the working group, Mr. P. Siltanen (FNS) served as chairman. In addition to general information exchange on recent activities in the participating organizations, the topics of the meeting included: 1) Code development and benchmarking for reactor dynamics applications. 2) Safety analysis methodology and results. 3) Future activities. A list of participants and a list of handouts distributed at the meeting are attached to the report (Author)

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

International Nuclear Information System (INIS)

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

2013-01-01

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

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

International Nuclear Information System (INIS)

Brumovsky, M.; Polachova, H.

1992-01-01

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

Consideration of severe accidents in design of advanced WWER reactors

International Nuclear Information System (INIS)

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

1998-01-01

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

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

International Nuclear Information System (INIS)

Boehmer, Bertram

2000-01-01

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

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

International Nuclear Information System (INIS)

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

1981-01-01

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

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

International Nuclear Information System (INIS)

Afanasyev, A.

2000-01-01

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

Low back pain status in elite and semi-elite Australian football codes: a cross-sectional survey of football (soccer), Australian rules, rugby league, rugby union and non-athletic controls.

Science.gov (United States)

Hoskins, Wayne; Pollard, Henry; Daff, Chris; Odell, Andrew; Garbutt, Peter; McHardy, Andrew; Hardy, Kate; Dragasevic, George

2009-04-17

Our understanding of the effects of football code participation on low back pain (LBP) is limited. It is unclear whether LBP is more prevalent in athletic populations or differs between levels of competition. Thus it was the aim of this study to document and compare the prevalence, intensity, quality and frequency of LBP between elite and semi-elite male Australian football code participants and a non-athletic group. A cross-sectional survey of elite and semi-elite male Australian football code participants and a non-athletic group was performed. Participants completed a self-reported questionnaire incorporating the Quadruple Visual Analogue Scale (QVAS) and McGill Pain Questionnaire (short form) (MPQ-SF), along with additional questions adapted from an Australian epidemiological study. Respondents were 271 elite players (mean age 23.3, range 17-39), 360 semi-elite players (mean age 23.8, range 16-46) and 148 non-athletic controls (mean age 23.9, range 18-39). Groups were matched for age (p = 0.42) and experienced the same age of first onset LBP (p = 0.40). A significant linear increase in LBP from the non-athletic group, to the semi-elite and elite groups for the QVAS and the MPQ-SF was evident (p < 0.001). Elite subjects were more likely to experience more frequent (daily or weekly OR 1.77, 95% CI 1.29-2.42) and severe LBP (discomforting and greater OR 1.75, 95% CI 1.29-2.38). Foolers in Australia have significantly more severe and frequent LBP than a non-athletic group and this escalates with level of competition.

Development and verification of a three-dimensional core model for WWR type reactors and its coupling with the accident code ATHLET. Final report

International Nuclear Information System (INIS)

Grundmann, U.; Lucas, D.; Mittag, S.; Rohde, U.

1995-04-01

The main goal of the project was the coupling of the 3D core model DYN3D for Russian VVER-type reactors, which has been developed in the RCR, with the thermohydraulic code ATHLET. The coupling has been realized on two basically different ways: - The implementation of only the neutron kinetics model of DYN3D into ATHLET (internal coupling), - the connection of the complete DYN3D core model including neutron kinetics, thermohydraulics and fuel rod model via data interfaces at the core top and bottom (external coupling). For the test of the coupling, comparative calculations between internal and external coupling versions have been carried out for a LOCA and a reactivity transient. Complementary goals of the project were: - The development of a DYN3D version for burn-up calculations, - the verification of DYN3D on benchmark tasks and experimental data on fuel rod behaviour, - a study on the extension of the neutron-physical data base. The project contributed to the development of advanced tools for the safety analysis of VVER-type reactors. Future work is aimed to the verification of the coupled code complex DYN3D-ATHLET. (orig.) [de

Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 3H. Kozloduy NPP units 5/6: Analysis and testing. Working material

International Nuclear Information System (INIS)

1999-01-01

In August 1991, following the SMiRT-11 Conference in Tokyo, a Technical Committee Meeting was held on the 'Seismic safety issues relating to existing NPPs'. The Proceedings of this TCM was subsequently compiled in an IAEA Working Material. One of the main recommendations of this TCM, called for the harmonization of criteria and methods used in Member States in seismic reassessment and upgrading of existing NPPs. Twenty four institutions from thirteen countries participated in the CRP named 'Benchmark study for the seismic analysis and testing of WWER type NPPs'. Two types of WWER reactors (WWER-1000 and WWER-440/213) selected for benchmarking. Kozloduy NPP Units 5/6 and Paks NPP represented these respectively as prototypes. Consistent with the recommendations of the TCM and the working paper prepared by the subsequent Consultants' Meeting, the focal activity of the CRP was the benchmarking exercises. A similar methodology was followed both for Paks NPP and Kozloduy NPP Unit 5. Firstly, the NPP (mainly the reactor building) was tested using a blast loading generated by a series of explosions from buried TNT charges. Records from this test were obtained at several free field locations (both downhole and surface), foundation mat, various elevations of structures as well as some tanks and the stack. Then the benchmark participants were provided with structural drawings, soil data and the free field record of the blast experiment. Their task was to make a blind prediction of the response at preselected locations. The analytical results from these participants were then compared with the results from the test. Although the benchmarking exercises constituted the focus of the CRP, there were many other interesting problems related to the seismic safety of WWER type NPPs which were addressed by the participants. These involved generic studies, i.e. codes and standards used in original WWER designs and their comparison with current international practice; seismic analysis

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

International Nuclear Information System (INIS)

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

2003-01-01

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

Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 3F. Kozloduy NPP units 5/6: Analysis and testing. Working material

International Nuclear Information System (INIS)

1999-01-01

In August 1991, following the SMiRT-11 Conference in Tokyo, a Technical Committee Meeting was held on the 'Seismic safety issues relating to existing NPPs'. The Proceedings of this TCM was subsequently compiled in an IAEA Working Material. One of the main recommendations of this TCM, called for the harmonization of criteria and methods used in Member States in seismic reassessment and upgrading of existing NPPs. Twenty four institutions from thirteen countries participated in the CRP named 'Benchmark study for the seismic analysis and testing of WWER type NPPs'. Two types of WWER reactors (WWER-1000 and WWER-440/213) selected for benchmarking. Kozloduy NPP Units 5/6 and Paks NPP represented these respectively as prototypes. Consistent with the recommendations of the TCM and the working paper prepared by the subsequent Consultants' Meeting, the focal activity of the CRP was the benchmarking exercises. A similar methodology was followed both for Paks NPP and Kozloduy NPP Unit 5. Firstly, the NPP (mainly the reactor building) was tested using a blast loading generated by a series of explosions from buried TNT charges. Records from this test were obtained at several free field locations (both downhole and surface), foundation mat, various elevations of structures as well as some tanks and the stack. Then the benchmark participants were provided with structural drawings, soil data and the free field record of the blast experiment. Their task was to make a blind prediction of the response at preselected locations. The analytical results from these participants were then compared with the results from the test. Although the benchmarking exercises constituted the focus of the CRP, there were many other interesting problems related to the seismic safety of WWER type NPPs which were addressed by the participants. These involved generic studies, i.e. codes and standards used in original WWER designs and their comparison with current international practice; seismic analysis

Neutron physics calculation for WWER-1000 absorber element lifetime determination

International Nuclear Information System (INIS)

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

2009-01-01

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

Performance analysis of WWER-440/230 nuclear power plants

International Nuclear Information System (INIS)

1997-01-01

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

Performance analysis of WWER-440/230 nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-01-01

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

Validation of computer codes and modelling methods for giving proof of nuclear safety of transport and storage of spent VVER-type nuclear fuels. Pt. 2. Criticality safety during transport and storage of spent VVER fuel elements. Final report; Einschaetzung von Rechenprogrammen und Methoden zum Nachweis der nuklearen Sicherheit bei Transport und Lagerung von WWER-Kernbrennstoffen. T. 2. Kritikalitaetssicherheit bei Transport und Lagerung von WWER-Brennelementen. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Buechse, H.; Langowski, A.; Lein, M.; Nagel, R.; Schmidt, H.; Stammel, M.

1995-03-15

The report gives the results of investigations on the validation of computer codes used to prove nuclear safety during transport and storage of spent VVER - fuel of NPP Greifswald and Rheinsberg. Characteristics of typical spent fuel (nuclide concentration, neutron source strength, gamma spectrum, decay heat) - calculated with several codes - and dose rates (e.g. in the surrounding of a loaded spent fuel cask) - based on the different source terms - are presented. Differences and their possible reasons are discussed. The results show that despite the differences in the source terms all relevant health physics requirements are met for all cases of source term. The validation of the criticality code OMEGA was established by calculation of appr. 200 critical experiments of LWR fuel, including VVER fuel rod arrangements. The mean error of the effective multiplication factor k{sub eff} is -0,01 compared to the experiment for this area of applicability. Thus, the OMEGA error of 2% assumed in earlier works has turned out to be sufficiently conservative. (orig.) [Deutsch] Der Bericht enthaelt die Ergebnisse von Untersuchungen zur Validierung von Rechenprogrammen, welche zum Nachweis der nukelaren Sicherheit bei Transport und Lagerung von WWER-Kernbrennstoff der KKW Greifswald und Rheinsberg eingesetzt wurden. Es werden eine Charakteristik des abgebrannten Brennstoffs (Nuklidkonzentrationen, Neutronenquellstaerke, Gammaspektrum, Nachzerfallsleistung) - berechnet mit verschiedenen Programmen - und Ortsdosisleistungen (z.B. in der Umgebung eines Transportbehaelters) - basierend auf den verschiedenen Quelltermen - angegeben. Differenzen und Ursachen werden diskutiert. Die Ergebnisse zeigen, dass trotz der Differenzen in den Quelltermen alle strahlenschutztechnisch relevanten Aussagen unbeeinflusst bleiben. Fuer die Einschaetzung des Gueltigkeitsbereiches des Monte-Carlo-Programms OMEGA wurden ca. 200 kritische Experimente mit LWR-Brennstoff unter besonderer Beruecksichtigung

Application of the severe accident code ATHLET-CD. Modelling and evaluation of accident management measures (Project WASA-BOSS)

Energy Technology Data Exchange (ETDEWEB)

Wilhelm, Polina; Jobst, Matthias; Kliem, Soeren; Kozmenkov, Yaroslav; Schaefer, Frank [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Div. Reactor Safety

2016-07-01

The improvement of the safety of nuclear power plants is a continuously on-going process. The analysis of transients and accidents is an important research topic, which significantly contributes to safety enhancements of existing power plants. In case of an accident with multiple failures of safety systems core uncovery and heat-up can occur. In order to prevent the accident to turn into a severe one or to mitigate the consequences of severe accidents, different accident management measures can be applied. Numerical analyses are used to investigate the accident progression and the complex physical phenomena during the core degradation phase, as well as to evaluate the effectiveness of possible countermeasures in the preventive and mitigative domain [1, 2]. The presented analyses have been performed with the computer code ATHLET-CD developed by GRS [3, 4].

Joint research project WASA-BOSS: Further development and application of severe accident codes. Assessment and optimization of accident management measures. Project B: Accident analyses for pressurized water reactors with the application of the ATHLET-CD code; Verbundprojekt WASA-BOSS: Weiterentwicklung und Anwendung von Severe Accident Codes. Bewertung und Optimierung von Stoerfallmassnahmen. Teilprojekt B: Druckwasserreaktor-Stoerfallanalysen unter Verwendung des Severe-Accident-Codes ATHLET-CD

Energy Technology Data Exchange (ETDEWEB)

Jobst, Matthias; Kliem, Soeren; Kozmenkov, Yaroslav; Wilhelm, Polina

2017-02-15

Within the framework of the project an ATHLET-CD input deck for a generic German PWR of type KONVOI has been created. This input deck was applied to the simulation of severe accidents from the accident categories station blackout (SBO) and small-break loss-of-coolant accidents (SBLOCA). The complete accident transient from initial event at full power until the damage of reactor pressure vessel (RPV) is covered and all relevant severe accident phenomena are modelled: start of core heat up, fission product release, melting of fuel and absorber material, oxidation and release of hydrogen, relocation of molten material inside the core, relocation to the lower plenum, damage and failure of the RPV. The model has been applied to the analysis of preventive and mitigative accident management measures for SBO and SBLOCA transients. Therefore, the measures primary side depressurization (PSD), injection to the primary circuit by mobile pumps and for SBLOCA the delayed injection by the cold leg hydro-accumulators have been investigated and the assumptions and start criteria of these measures have been varied. The time evolutions of the transients and time margins for the initiation of additional measures have been assessed. An uncertainty and sensitivity study has been performed for the early phase of one SBO scenario with PSD (until the start of core melt). In addition to that, a code -to-code comparison between ATHLET-CD and the severe accident code MELCOR has been carried out.

Autonomic Nervous System in Paralympic Athletes with Spinal Cord Injury.

Science.gov (United States)

Walter, Matthias; Krassioukov, Andrei V

2018-05-01

Individuals sustaining a spinal cord injury (SCI) frequently suffer from sensorimotor and autonomic impairment. Damage to the autonomic nervous system results in cardiovascular, respiratory, bladder, bowel, and sexual dysfunctions, as well as temperature dysregulation. These complications not only impede quality of life, but also affect athletic performance of individuals with SCI. This article summarizes existing evidence on how damage to the spinal cord affects the autonomic nervous system and impacts the performance in athletes with SCI. Also discussed are frequently used performance-enhancing strategies, with a special focus on their legal aspect and implication on the athletes' health. Copyright © 2018 Elsevier Inc. All rights reserved.

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

International Nuclear Information System (INIS)

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

1997-01-01

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

MOPABA-H2 - Computer code for calculation of hydrogen generation and distribution in the equipment of power plants with WWER type reactors in design modes of operation

International Nuclear Information System (INIS)

Arkhipov, O.P.; Kharitonov, Yu.V.; Shumskiy, A.M.; Kabakchi, S.A.

2002-01-01

With the aim of ensuring the hydrogen explosive-proof situation in the reactor plant, a complex of scientific-and-research work was carried out including the following: revealing the mechanisms of generation and release of hydrogen in the primary equipment components under design operation modes of the reactor plant with WWER; development of calculation procedure and computer code MOPABA-H2 enabling to determine the hydrogen content in RP equipment components under design operation modes. In the process of procedure development it was found out that the calculation of hydrogen content in the plant equipment requires development of the following main mathematical models: radiochemical processes in the primary coolant which has impurities and added special reagents; absorption of the core ionizing radiation by the coolant; steam-zirconium reaction (during design-basis accident of LOCA type); coolant mass transfer over the reactor plant equipment including transition of the phase boundary by the components of the coolant. (author)

Student retention in athletic training education programs.

Science.gov (United States)

Dodge, Thomas M; Mitchell, Murray F; Mensch, James M

2009-01-01

The success of any academic program, including athletic training, depends upon attracting and keeping quality students. The nature of persistent students versus students who prematurely leave the athletic training major is not known. Understanding the profiles of athletic training students who persist or leave is important. To (1) explore the relationships among the following variables: anticipatory factors, academic integration, clinical integration, social integration, and motivation; (2) determine which of the aforementioned variables discriminate between senior athletic training students and major changers; and (3) identify which variable is the strongest predictor of persistence in athletic training education programs. Descriptive study using a qualitative and quantitative mixed-methods approach. Thirteen athletic training education programs located in District 3 of the National Athletic Trainers' Association. Ninety-four senior-level athletic training students and 31 college students who changed majors from athletic training to another degree option. Data were collected with the Athletic Training Education Program Student Retention Questionnaire (ATEPSRQ). Data from the ATEPSRQ were analyzed via Pearson correlations, multivariate analysis of variance, univariate analysis of variance, and a stepwise discriminant analysis. Open-ended questions were transcribed and analyzed using open, axial, and selective coding procedures. Member checks and peer debriefing techniques ensured trustworthiness of the study. Pearson correlations identified moderate relationships among motivation and clinical integration (r = 0.515, P accounting for 37.2% of the variance between groups. The theoretic model accurately classified 95.7% of the seniors and 53.8% of the major changers. A common theme emerging from the qualitative data was the presence of a strong peer-support group that surrounded many of the senior-level students. Understanding student retention in athletic training is

Potential of the MCNP computer code

International Nuclear Information System (INIS)

Kyncl, J.

1995-01-01

The MCNP code is designed for numerical solution of neutron, photon, and electron transport problems by the Monte Carlo method. The code is based on the linear transport theory of behavior of the differential flux of the particles. The code directly uses data from the cross section point data library for input. Experience is outlined, gained in the application of the code to the calculation of the effective parameters of fuel assemblies and of the entire reactor core, to the determination of the effective parameters of the elementary fuel cell, and to the numerical solution of neutron diffusion and/or transport problems of the fuel assembly. The agreement between the calculated and observed data gives evidence that the MCNP code can be used with advantage for calculations involving WWER type fuel assemblies. (J.B.). 4 figs., 6 refs

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

International Nuclear Information System (INIS)

Stemmler, G.

1985-01-01

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

Experience in WWER fuel assemblies vibration analysis

International Nuclear Information System (INIS)

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

2003-01-01

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

Safety of WWER type nuclear power plants - viewing from Hungary

International Nuclear Information System (INIS)

Voeroess, L.

1991-01-01

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

Visual and intelligent transients and accidents analyzer based on thermal-hydraulic system code

International Nuclear Information System (INIS)

Meng Lin; Rui Hu; Yun Su; Ronghua Zhang; Yanhua Yang

2005-01-01

Full text of publication follows: Many thermal-hydraulic system codes were developed in the past twenty years, such as RELAP5, RETRAN, ATHLET, etc. Because of their general and advanced features in thermal-hydraulic computation, they are widely used in the world to analyze transients and accidents. But there are following disadvantages for most of these original thermal-hydraulic system codes. Firstly, because models are built through input decks, so the input files are complex and non-figurative, and the style of input decks is various for different users and models. Secondly, results are shown in off-line data file form. It is not convenient for analysts who may pay more attention to dynamic parameters trend and changing. Thirdly, there are few interfaces with other program in these original thermal-hydraulic system codes. This restricts the codes expanding. The subject of this paper is to develop a powerful analyzer based on these thermal-hydraulic system codes to analyze transients and accidents more simply, accurately and fleetly. Firstly, modeling is visual and intelligent. Users build the thermalhydraulic system model using component objects according to their needs, and it is not necessary for them to face bald input decks. The style of input decks created automatically by the analyzer is unified and can be accepted easily by other people. Secondly, parameters concerned by analyst can be dynamically communicated to show or even change. Thirdly, the analyzer provide interface with other programs for the thermal-hydraulic system code. Thus parallel computation between thermal-hydraulic system code and other programs become possible. In conclusion, through visual and intelligent method, the analyzer based on general and advanced thermal-hydraulic system codes can be used to analysis transients and accidents more effectively. The main purpose of this paper is to present developmental activities, assessment and application results of the visual and intelligent

Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 3E. Kozloduy NPP units 5/6: Analysis/testing. Working material

International Nuclear Information System (INIS)

1996-01-01

The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material contains reports on data related to floor response spectra of Kozloduy NPP; calculational-experimental examination and ensuring of equipment and pipelines seismic resistance at starting and operating WWER-type NPPs; analysis of design floor response spectra and testing of the electrical systems; experimental investigations and seismic analysis Kozloduy NPP; testing of components on the shaking table facilities and contribution to full scale dynamic testing of Kozloduy NPP; seismic evaluation of the main steam line, piping systems, containment pre-stressing and steel ventilation chimney of Kozloduy NPP

Development of WWER-440 fuel. Use of fuel assemblies of 2-nd and 3-rd generations with increased enrichment

International Nuclear Information System (INIS)

Vasilchenko, I.; Lushin, V.; Ananev, U.; Baranov, A.; Kukushkin, U.

2009-01-01

The problem of increasing the power of units at NPPs with WWER-440 is of current importance. There are all the necessary prerequisites for the above-stated problem as a result of updating the design of fuel assemblies and codes. The decrease of power peaking factor in the core is achieved by using profiled fuel assemblies, fuel-integrated burning absorber, FAs with modernized docking unit, modern codes, which allows decreasing conservatism of RP safety substantiation. A wide range of experimental studies of fuel behaviour has been performed which has reached burn-up of (50-60) MW·day/kgU in transition and emergency conditions, post-reactor studies of fuel assemblies, fuel rods and fuel pellets with a 5-year operating period have been performed, which prove high reliability of fuel, presence of a large margin in the fuel pillar, which helps reactor operation at increased power. The results of the work performed on introduction of 5-6 fuel cycles show that the ultimate fuel state on operability in WWER-440 reactors is far from being achieved. Neutron-physical and thermal-hydraulic characteristics of the cores of working power units with RP V-213 are such that actual (design and measured) power peaking factors on fuel assemblies and fuel rods, as a rule, are smaller than the maximum design values. This factor is a real reserve for power forcing. There is experience of operating Units 1, 2, 4 of the Kola NPP and Unit 2 of the Rovno NPP at increased power. Units of the Loviisa NPP are operated at 109 % power. During transfer to work at increased power it is reasonable to use fuel assemblies with increased height of the fuel pillar, which allows decreasing medium linear power distribution. Further development of the 2-nd generation fuel assembly design and consequent transition to working fuel assemblies of the 3-rd generation provides significant improvement of fuel consumption under the conditions of WWER-440 reactors operation with more continuous fuel cycles and

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

International Nuclear Information System (INIS)

Gantchev, T.

1998-01-01

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

Results of post-irradiation examination of WWER fuel assembly structural components made of E110 and E635 alloys

International Nuclear Information System (INIS)

Smirnov, A.; Markov, D.; Smirnov, V.; Polenok, V.; Ivashchenko, A.; Strozhuk, A.

2006-01-01

The paper presents the main examination results on the condition of fuel rods claddings, guide tubes and spacer grids of the WWER FA made of E110 and E635 alloys operated under standard operating conditions. The paper is based on the data obtained during the examination of 28 WWER-1000 FA and 12 WWER-400 FA. E110 alloy is shown to be suitable material for the WWER fuel rod claddings under the normal operating conditions. E635 alloy is attractive to manufacturing of the skeleton components. The currently used combination (E110 as a material of fuel rods claddings and E635 - as a material of the skeleton components) is the optimal solution for the WWER fuel assembly because the advantages of the both alloys are used. (authors)

Frustrations among graduates of athletic training education programs.

Science.gov (United States)

Bowman, Thomas G; Dodge, Thomas M

2013-01-01

Although previous researchers have begun to identify sources of athletic training student stress, the specific reasons for student frustrations are not yet fully understood. It is important for athletic training administrators to understand sources of student frustration to provide a supportive learning environment. To determine the factors that lead to feelings of frustration while completing a professional athletic training education program (ATEP). Qualitative study. National Athletic Trainers' Association (NATA) accredited postprofessional education program. Fourteen successful graduates (12 women, 2 men) of accredited professional undergraduate ATEPs enrolled in an NATA-accredited postprofessional education program. We conducted semistructured interviews and analyzed data with a grounded theory approach using open, axial, and selective coding procedures. We negotiated over the coding scheme and performed peer debriefings and member checks to ensure trustworthiness of the results. Four themes emerged from the data: (1) Athletic training student frustrations appear to stem from the amount of stress involved in completing an ATEP, leading to anxiety and feelings of being overwhelmed. (2) The interactions students have with classmates, faculty, and preceptors can also be a source of frustration for athletic training students. (3) Monotonous clinical experiences often left students feeling disengaged. (4) Students questioned entering the athletic training profession because of the fear of work-life balance problems and low compensation. In order to reduce frustration, athletic training education programs should validate students' decisions to pursue athletic training and validate their contributions to the ATEP; provide clinical education experiences with graded autonomy; encourage positive personal interactions between students, faculty, and preceptors; and successfully model the benefits of a career in athletic training.

Verification of the computer code ATHLET in the framework of the external verification group ATHLET BETHSY test 9.3 - steam generator U-tube rupture with failure of the high pressure injection. Final report; Verifikation des ATHLET-Rechenprogramms im Rahmen der externen Verifikationsgruppe ATHLET BETHSY Test 9.3 - Heizrohrbruch mit Versagen der Hochdruck-Noteinspeisung. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Krepper, E.; Schaefer, F. [Forschungszentrum Rossendorf e.V. (FZR) (Germany). Inst. fuer Sicherheitsforschung

1998-08-01

In the framework of the external validation of the thermalhydraulic code ATHLET MOD 1.1 CYCLE D, which is being developed by the GRS, post test analyses of two experiments were done, which were performed at the French integral test facility BETHSY. During test 9.3 the consequences of a steam generator U-tube rupture with failure of the high pressure injection and of the auxiliary feedwater supply were investigated. As accident management measures, the depressurization of the secondary sides, first of the two intact steam generators, then of the damaged steam generator and finally the primary depressurization by opening of the pressurizer valve were performed. The results show, that the code ATHLET is able to describe the complex scenario in good accordance with the experiment. The safety relevant statement could be reproduced. Deviations, which did not impose the general results, occurred concerning the break mass flow during the depressurization of the damaged steam generator and the description of the failure of the heat transfer to the damaged steam generator. Reasons are hardly to find, because these processes are highly complex. (orig.) [Deutsch] Im Rahmen der externen Validierung des von der Gesellschaft fuer Anlagen- und Reaktorsicherheit entwickelten Stoerfallcodes ATHLET, der in der Version Mod 1.1 Cycle D vorlag, wurden zwei Experimente nachgerechnet und analysiert, die an der franzoesischen Versuchsanlage BETHSY durchgefuehrt wurden. Im Test 9.3 werden die Konsequenzen untersucht, wenn bei einem Heizrohrbruch die Hochdruckeinspeisung sowie die Not-Speisewasserversorgung der Dampferzeuger versagen und nur die Druckspeicher sowie die Niederdruckeinspeisung zur Verfuegung stehen. Als Accident Management Massnahmen wurde die sekundaere Druckentlastung und schliesslich die primaere Entlastung ueber den Druckhalter untersucht. Die Analyse kommt zu dem Ergebnis, dass der Code ATHLET in der Lage ist, dieses komplexe Szenario recht gut zu beschreiben. Die

Athletes and Supplements: Prevalence and Perspectives.

Science.gov (United States)

Garthe, Ina; Maughan, Ronald J

2018-03-01

In elite sport, where opponents are evenly matched, small factors can determine the outcome of sporting contests. Not all athletes know the value of making wise nutrition choices, but anything that might give a competitive edge, including dietary supplements, can seem attractive. Between 40% and 100% of athletes typically use supplements, depending on the type of sport, level of competition, and the definition of supplements. However, unless the athlete has a nutrient deficiency, supplementation may not improve performance and may have a detrimental effect on both performance and health. Dietary supplements are classified as a subcategory of food, so manufacturers are not required to provide evidence of product safety and efficacy, nor obtain approval from regulatory bodies before marketing supplements. This creates the potential for health risks, and serious adverse effects have been reported from the use of some dietary supplements. Athletes who compete in sports under an anti-doping code must also realize that supplement use exposes them to a risk of ingesting banned substances or precursors of prohibited substances. Government systems of regulations do not include specific laboratory testing for banned substances according to the WADA list, so a separate regulatory framework to evaluate supplements for their risk of provoking a failed doping test is needed. In the high-performance culture typical of elite sport, athletes may use supplements regardless of possible risks. A discussion around medical, physiological, cultural, and ethical questions may be warranted to ensure that the athlete has the information needed to make an informed choice.

Decommissioning strategy of the operating WWER type units in the Ukraine

International Nuclear Information System (INIS)

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

2002-01-01

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

Upgrade the startup test instrumentation (ANMS) with innovated I and C system at the Dukovany WWER 440/213 units

International Nuclear Information System (INIS)

Kocek, V.

2005-01-01

The paper focuses on the innovation procedures of physics startup tests instrumentation used in Dukovany NPP (ANMS) connected with I and C system refurbishment of the Dukovany WWER 440/213 units. The following matters are further discussed in this paper: a) changes in ANMS design, whose has been carried out due to reflect the situation connected with I and C innovation; b) process of consistency checking during implementation stage; c) results of the physics startup tests with upgraded ANMS in 2005 (introduction of Gd2 fuel on Dukovany NPP) (Authors)

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

International Nuclear Information System (INIS)

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

1996-01-01

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

Simulation of the BWR experiments CORA-17 and CORA-28 using ATHLET-CD and assessment of BWR modelling. 1{sup st} Technical report. Validation and interpretation of the ATHLET-CD model basis; Simulation der SWR-Versuche CORA-17 und CORA-28 mit dem Programmsystem ATHLET-CD und Bewertung der SWR-Modellbasis. 1. Technischer Fachbericht. Validierung und Interpretation der ATHLET-CD Modellbasis

Energy Technology Data Exchange (ETDEWEB)

Hoffmann, M.; Gremme, F.; Koch, M.K.

2013-08-15

The 1st Technical Report was prepared for the research project ''Validation and Interpretation of the ATHLET-CD Model Basis'' funded by the Federal Ministry of Economics and Technology (BMWi1501385) and carried out at the Chair of Energy Systems and Energy Economics at Ruhr-Universitaet Bochum (RUB). This report provides results of the simulation of the Boiling Water Reactor (BWR) experiments CORA-17 and -28 with ATHLET-CD Mod. 2.2A. The system code ATHLET-CD (Analysis of Thermal-hydraulics of Leaks and Transients - Core Degradation) is developed by the German Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH. Code results are compared to measurements in order to assess and to analyze the capabilities of the current code version with regard to the modeling of BWR components. The CORA test series was carried out between the years 1987 and 1993 at the former Kernforschungszentrum Karlsruhe (KfK), now Karlsruhe Institute of Technology (KIT). The investigations provided experimental data regarding the material behavior during the early phase of core degradation in Light Water Reactors (LWR). The tests CORA17 and -28 represented a typical BWR arrangement of the fuel rod bundle and provided insights about the bundle behavior during the quenching process (CORA-17) and regarding the influence of a preoxidized bundle (CORA-28), respectively. The simulation results are analyzed and discussed in terms of the thermal bundle behavior, the zirconium oxidation in steam and the resulting hydrogen generation as well as the material relocation. In particular, the recently extended modeling capabilities of the code in terms of the relocation of BWR components like the absorber blade and the canister wall are assessed. The analysis shows that the code captures the thermal behavior in good agreement in both experiments. An even enhanced reproduction of the test CORA-28 is obtained in comparison to a calculation using the previous code version ATHLET-CD Mod

Application of coupled codes for safety analysis and licensing issues

International Nuclear Information System (INIS)

Langenbuch, S.; Velkov, K.

2006-01-01

An overview is given on the development and the advantages of coupled codes which integrate 3D neutron kinetics into thermal-hydraulic system codes. The work performed within GRS by coupling the thermal-hydraulic system code ATHLET and the 3D neutronics code QUABOX/CUBBOX is described as an example. The application of the coupled codes as best-estimate simulation tools for safety analysis is discussed. Some examples from German licensing practices are given which demonstrate how the improved analytical methods of coupled codes have contributed to solve licensing issues related to optimized and more economical use of fuel. (authors)

The development of the code package PERMAK--3D//SC--1

International Nuclear Information System (INIS)

Bolobov, P. A.; Oleksuk, D. A.

2011-01-01

Code package PERMAK-3D//SC-1 was developed for performing pin-by-pin coupled neutronic and thermal hydraulic calculation of the core fragment of seven fuel assemblies and was designed on the basis of 3D multigroup pin-by-pin code PERMAK-3D and 3D (subchannel) thermal hydraulic code SC-1 The code package predicts axial and radial pin-by-pin power distribution and coolant parameters in stimulated region (enthalpies,, velocities,, void fractions,, boiling and DNBR margins).. The report describes some new steps in code package development. Some PERMAK-3D//SC-1 outcomes of WWER calculations are presented in the report. (Authors)

System recovery from athletes in team sports with individual use of medicinal plants

Directory of Open Access Journals (Sweden)

Kozina Zhanneta Leonidovna

2011-11-01

Full Text Available We consider the direction of optimization of adaptive systems. In research was attended by 43 athletes. We determined the content of cortisol, insulin and β-endorphins in the blood. Showing cause injuries of athletes. It is noted that as the growth rates of athletes increases the number and severity of injuries. It is established that the main causes of injury is a conflict between reducing functionality and increasing requirements of the game. Recommended application fee of medicinal plants according to individual characteristics of functional and psycho-physiological condition of athletes. Recommendations on the normalization of the adaptive systems.

Procedures for analysis of accidents in shutdown modes for WWER nuclear power plants. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

International Nuclear Information System (INIS)

1997-07-01

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

Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 4C. Paks NPP: Analysis and testing. Working material

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-07-01

The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material involves comparative analysis of the seismic analysis results of the reactor building for soft soil conditions, derivation of design response spectra for components and systems; and upper range design response spectra for soft soil site conditions at Paks NPP.

Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 4C. Paks NPP: Analysis and testing. Working material

International Nuclear Information System (INIS)

1996-01-01

The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material involves comparative analysis of the seismic analysis results of the reactor building for soft soil conditions, derivation of design response spectra for components and systems; and upper range design response spectra for soft soil site conditions at Paks NPP

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

International Nuclear Information System (INIS)

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

2008-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-07-01

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

Novice to expert practice via postprofessional athletic training education: a grounded theory.

Science.gov (United States)

Neibert, Peter J

2009-01-01

To discover the theoretic constructs that confirm, disconfirm, or extend the principles and their applications appropriate for National Athletic Trainers' Association (NATA)-accredited postprofessional athletic training education programs. Interviews at the 2003 NATA Annual Meeting & Clinical Symposia. Qualitative study using grounded theory procedures. Thirteen interviews were conducted with postprofessional graduates. Participants were purposefully selected based on theoretic sampling and availability. The transcribed interviews were analyzed using open coding, axial coding, and selective coding procedures. Member checks, reflective journaling, and triangulation were used to ensure trustworthiness. The participants' comments confirmed and extended the current principles of postprofessional athletic training education programs and offered additional suggestions for more effective practical applications. The emergence of this central category of novice to expert practice is a paramount finding. The tightly woven fabric of the 10 processes, when interlaced with one another, provides a strong tapestry supporting novice to expert practice via postprofessional athletic training education. The emergence of this theoretic position pushes postprofessional graduate athletic training education forward to the future for further investigation into the theoretic constructs of novice to expert practice.

SERPENT Monte Carlo reactor physics code

International Nuclear Information System (INIS)

Leppaenen, J.

2010-01-01

SERPENT is a three-dimensional continuous-energy Monte Carlo reactor physics burnup calculation code, developed at VTT Technical Research Centre of Finland since 2004. The code is specialized in lattice physics applications, but the universe-based geometry description allows transport simulation to be carried out in complicated three-dimensional geometries as well. The suggested applications of SERPENT include generation of homogenized multi-group constants for deterministic reactor simulator calculations, fuel cycle studies involving detailed assembly-level burnup calculations, validation of deterministic lattice transport codes, research reactor applications, educational purposes and demonstration of reactor physics phenomena. The Serpent code has been publicly distributed by the OECD/NEA Data Bank since May 2009 and RSICC in the U. S. since March 2010. The code is being used in some 35 organizations in 20 countries around the world. This paper presents an overview of the methods and capabilities of the Serpent code, with examples in the modelling of WWER-440 reactor physics. (Author)

Preliminary results of the seventh three-dimensional AER dynamic benchmark problem calculation. Solution with DYN3D and RELAP5-3D codes

International Nuclear Information System (INIS)

Bencik, M.; Hadek, J.

2011-01-01

The paper gives a brief survey of the seventh three-dimensional AER dynamic benchmark calculation results received with the codes DYN3D and RELAP5-3D at Nuclear Research Institute Rez. This benchmark was defined at the twentieth AER Symposium in Hanassari (Finland). It is focused on investigation of transient behaviour in a WWER-440 nuclear power plant. Its initiating event is opening of the main isolation valve and re-connection of the loop with its main circulation pump in operation. The WWER-440 plant is at the end of the first fuel cycle and in hot full power conditions. Stationary and burnup calculations were performed with the code DYN3D. Transient calculation was made with the system code RELAP5-3D. The two-group homogenized cross sections library HELGD05 created by HELIOS code was used for the generation of reactor core neutronic parameters. The detailed six loops model of NPP Dukovany was adopted for the seventh AER dynamic benchmark purposes. The RELAP5-3D full core neutronic model was coupled with 49 core thermal-hydraulic channels and 8 reflector channels connected with the three-dimensional model of the reactor vessel. The detailed nodalization of reactor downcomer, lower and upper plenum was used. Mixing in lower and upper plenum was simulated. The first part of paper contains a brief characteristic of RELAP5-3D system code and a short description of NPP input deck and reactor core model. The second part shows the time dependencies of important global and local parameters. (Authors)

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

International Nuclear Information System (INIS)

Masopust, R.

1999-01-01

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

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

International Nuclear Information System (INIS)

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

1998-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2001-10-01

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

Results of trial operation of the WWER advanced fuel assemblies

International Nuclear Information System (INIS)

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

2001-01-01

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Operational indices of WWER-1000 fuel assemblies and their improvements

Energy Technology Data Exchange (ETDEWEB)

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

1994-12-31

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

Operational indices of WWER-1000 fuel assemblies and their improvements

International Nuclear Information System (INIS)

Vasilchenko, I.; Demin, E.

1994-01-01

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

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

International Nuclear Information System (INIS)

Shishkov, L.; Tsyganov, S.

2008-01-01

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

Protection of WWER type primary loops against extreme effects

International Nuclear Information System (INIS)

Podrouzek, J.; Rejent, B.

1985-01-01

Dynamic analyses of the WWER-440 primary loops for the Mochovce nuclear power plant showed that the unprotected primary loop is very soft with a first eigenfrequency of 0.38 Hz. Protection with amortisseurs and viscous shock absorbers was compared and the viscous shock absorber in all cases proved to be more suitable. GERB viscous absorbers will be installed at the Mochovce nuclear power plant. First calculations of the dynamic resistance of the WWER-1000 primary loops for the Temelin nuclear power plant to extreme events were also made. It was shown that the unprotected primary loop is rather soft with a first eigenfrequency of 0.9 Hz, or 0.6 Hz at the pressurizer branch. It will therefore be necessary to protect the primary loops with viscous shock absorbers. (Z.M.)

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

International Nuclear Information System (INIS)

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

2006-01-01

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

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

International Nuclear Information System (INIS)

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

1977-01-01

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

Extended analysis of WWER-1000 Charpy test data

International Nuclear Information System (INIS)

Vodenicharov, St.; Kamenova, Tz.

2001-01-01

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

NOKIA - nuclear power plant monitoring system

International Nuclear Information System (INIS)

Anon.

The monitoring system is described developed specially for the LOVIISA-1 and -2 nuclear power plants with two WWER-440 units. The multiprocessor system of the WWER-440 contains 3 identical main computers. The in core instrumentation is based on stationary self-powered neutron detectors and on thermocouples for measuring the coolant temperature. The system has equipment for the automatic control of the insulation resistance of the self-powered detectors. It is also equipped with a wide range of standard and special programmes. The standard programmes permit the recording of analog and digital data at different frequencies depending on the pre-set requirements. These data are processed and form data files which are accessible from all programmes. The heart of the special programme is a code for the determination of the power distribution in the core of the WWER-440 reactor. The main part of the programme is the algorithm for computing measured neutron fluxes derived from the signals of the self-powered detectors and the algorithm for deriving the global distribution of the neutron flux in the core. The computed power distribution is used for the determination of instantaneous thermal loads and the distribution of burnup in the core. The production programme of the FINNATOM company for nuclear power plants is listed. (B.S.)

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

International Nuclear Information System (INIS)