PSA Level 2 activities for RBMK reactors

International Nuclear Information System (INIS)

Gubler, R.

1998-01-01

Probabilistic safety analyses (PSAs) of the boiling water graphite moderated pressure tube reactors (RBMKs) have been developed only recently and they are limited to Level 1. Activities at the IAEA were first motivated because of the difficulties to characterize core damage for RBMK reactors. Core damage probability is used in documents of the IAEA as a convenient single valued measure, for example for probabilistic safety criteria. The limited number of PSAs that have been completed for the RBMK reactors have shown that several special features of these channel type reactors necessitate revisiting of the characterization of core damage for these reactors. Furthermore, it has become increasingly evident that detailed deterministic analysis of DBAs and beyond design basis accidents reveal considerable insights into RBMK response to various accident conditions. These analyses can also help in better characterizing the outstanding phenomenological uncertainties, improved EOPs and AM strategies, including potential risk-beneficial accident negative backfits. The deterministic efforts should be focused first on elucidating accident progression processes and phenomena, and second on finding, qualifying and implementing procedures to minimize the risk of severe accident states The IAEA PSA procedures were mainly developed in New of vessel type LWRs, and would therefore require extensions to make them directly applicable. to channel type reactors. (author) (author)

Uncertainty of determination of 158Tb in the RBMK nuclear reactor waste.

Science.gov (United States)

Plukis, Artūras; Barkauskas, Vytenis; Druteikienė, Rūta; Duškesas, Grigorijus; Germanas, Darius; Gudelis, Arūnas; Juodis, Laurynas; Lagzdina, Elena; Plukienė, Rita; Remeikis, Vidmantas

2018-04-01

The activity of 158 Tb was measured in waste samples from the Ignalina NPP Unit I RBMK-1500 reactor using gamma-ray spectrometry. The origin of 158 Tb and the other observed gamma-ray emitters has been studied by using SCALE 6.1 modeling and comparing radionuclide ratios in the RBMK-1500 radioactive waste. The results of the calculation of the massic activity of gamma-ray emitters were used for interpretation of the total gamma-ray spectrum and the determination of 158 Tb massic activity uncertainty in the waste of RBMK-1500. Copyright © 2017 Elsevier Ltd. All rights reserved.

RBMK nuclear power plants: Generic safety issues. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

International Nuclear Information System (INIS)

1996-05-01

This report has been prepared on the basis of above mentioned report and it is intended to provide information on RBMK NPPs generic safety issues. As all other insights, recommendations and conclusions resulting from the IAEA Programme, this report is intended to assist national decision makers, who have sole responsibility for the regulation and safe operation of their nuclear power plants. It also serves to focus national and international projects on priority of the RBMK safety improvements. 23 refs, 10 figs, 3 tabs

RELAP5-3D code validation for RBMK phenomena

International Nuclear Information System (INIS)

Fisher, J.E.

1999-01-01

The RELAP5-3D thermal-hydraulic code was assessed against Japanese Safety Experiment Loop (SEL) and Heat Transfer Loop (HTL) tests. These tests were chosen because the phenomena present are applicable to analyses of Russian RBMK reactor designs. The assessment cases included parallel channel flow fluctuation tests at reduced and normal water levels, a channel inlet pipe rupture test, and a high power, density wave oscillation test. The results showed that RELAP5-3D has the capability to adequately represent these RBMK-related phenomena

Method of experimental and theoretical modeling for multiple pressure tube rupture for RBMK reactor

International Nuclear Information System (INIS)

Medvedeva, N.Y.; Goldstein, R.V.; Burrows, J.A.

2001-01-01

The rupture of single RBMK reactor channels has occurred at a number of stations with a variety of initiating events. It is assumed in RBMK Safety Cases that the force of the escaping fluid will not cause neighbouring channels to break. This assumption has not been justified. A chain reaction of tube breaks could over-pressurise the reactor cavity leading to catastrophic failure of the containment. To validate the claims of the RBMK Safety Cases the Electrogorsk Research and Engineering Centre, in participation with experts from the Institute of Mechanics of RAS, has developed the method of interacting multiscale physical and mathematical modelling for coupled thermophysical, hydrogasodynamic processes and deformation and break processes causing and (or) accompanying potential failures, design and beyond the design RBMK reactor accidents. To realise the method the set of rigs, physical and mathematical models and specialized computer codes are under creation. This article sets out an experimental philosophy and programme for achieving this objective to solve the problem of credibility or non-credibility for multiple fuel channel rupture in RBMK.(author)

Remote technology in RBMK-1000 spent fuel management at NPP site

International Nuclear Information System (INIS)

Makarchuk, T.F.; Kozlov, Y.V.; Tikhonov, N.S.; Tokarenko, A.I.; Spichev, V.V.; Kaljazin, N.N.

1999-01-01

The report describes the remote technologies employed in the nuclear power plant with RBMK-1000 type. Spent fuel transfer and handling operations at reactor (AR) and away from reactor (AFR) on reactor site (RS) facilities are illustrated by the example of the Leningradskaya NPP and are typical for all NPPs with RBMK-1000. The current approach to spent fuel management at NPP sites is also presented. (author)

A WIMS-NESTLE reactor physics model for an RBMK reactor

International Nuclear Information System (INIS)

Perry, R.T.; Meriwether, G.H.

1996-01-01

This work describes the static neutronic calculations made for a three-dimensional model of an RBMK (Russian) reactor. Future work will involve the use of this neutronic model and a thermal-hydraulic model in coupled calculations. The lattice code, WIMS-D, was used to obtain the cross sections for the static neutronic calculations. The static reactor neutronic calculations were made with NESTLE, a three-dimensional nodal diffusion code. The methods used to establish an RBMK reactor model for use in these codes are discussed, and the cross sections calculated are given

A WIMS-NESTLE reactor physics model for an RBMK reactor

International Nuclear Information System (INIS)

Perry, R.T.; Meriwether, G.H.

1996-01-01

This work describes the static neutronic calculations made for a three-dimensional model of an RBMK (Russian) reactor. Future work will involve the use of this neutronic model and a thermal-hydraulic model in coupled calculations. The lattice code, WIMS-D, was used to obtain the cross sections for the static neutronic calculations. The static reactor neutronic calculations were made with NESTLE, a three-dimensional nodal diffusion code. The methods used to establish an RBMK reactor model for use in these codes are discussed, and the cross sections calculated are given. (author)

Additional reactor protection system of RBMK-1500

International Nuclear Information System (INIS)

1999-01-01

Analysis of anticipated transients without scram of RBMK-1500 reactor showed that additional reactor protection system is required. Data of accident analysis in the case of loose of external electric power and loose of vacuum in condensers of turbines are provided

Calculations of a station blackout transient in a RBMK type nuclear power plant with the CATHARE code

International Nuclear Information System (INIS)

Niklaus, F.; Korteniemi, V.

1996-01-01

At the Department of Energy Technology at Lappeenranta University of Technology a CATHARE model of one unit of the St. Petersburg (RBMK) nuclear power plant has been generated. The investigations have been done in order to understand better the thermal-hydraulic behaviour of RBMK type reactors and in order to see how far the French thermal-hydraulic safety code CATHARE can predict the physical phenomena during various RBMK transients. (12 refs.)

Mitigation of intergranular stress corrosion cracking in RBMK reactors. Final report of the programme's steering committee

International Nuclear Information System (INIS)

2002-09-01

In 2000 the IAEA initiated an Extrabudgetary Programme on Mitigation of Intergranular Stress Corrosion Cracking in RBMK Reactors to assist countries operating RBMK reactors in addressing the issue in austenitic stainless steel 300 mm diameter piping. Intergranular stress corrosion cracking of austenitic stainless steel piping in BWRs has been a major safety concern since the early seventies. Similar degradation was found in RBMK reactor piping in 1997. Early in 1998 the IAEA responded to requests for assistance from RBMK operating countries on this issue through activities organized in the framework of Technical Co-operation Department regional projects and the Extrabudgetary Programme on the Safety of WWER and RBMK Nuclear Power Plants. Results of these activities were a basis for the formulation of the objective and scope of the Extrabudgetary Programme on Mitigation of Intergranular Stress Corrosion Cracking in RBMK reactors ('the Programme'). The scope of the Programme included in-service inspection, assessment, repair and mitigation, and water chemistry and decontamination. The Programme was pursued by means of exchange of experience, formulation of guidance, transfer of technology, and training, which will assist the RBMK operators to address related safety concerns. The Programme implementation relied on voluntary extrabudgetary financial contributions from Japan, Spain, the United Kingdom and the USA, and on in kind contributions from Finland, Germany and Sweden. The Programme was implemented in close co-ordination with ongoing national and bilateral activities and major inputs to the Programme were provided through the activities of the Swedish International Project Nuclear Safety and of the US DOE International Nuclear Safety Program. The RBMK nuclear power plants in Lithuania, Russian Federation and Ukraine hosted most of the Programme activities. Support of these Member States involved in the Programme was instrumental for its successful completion in

RBMK fuel channel integrity. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

International Nuclear Information System (INIS)

1999-01-01

The fuel channel integrity in the RBMK NPPs is an issue of high safety concern. To date, three single fuel channel ruptures have occurred. Fuel channel rupture results in release of radioactivity to the reactor cavity and may lead to a release of radioactivity to the environment if the confinement safety system does not function properly. A multiple fuel channel rupture exceeding the venting capacity of the reactor cavity overpressure protection system poses a major impact on the plant safety. Further, due to incorrect prediction at the design stage the gas gap between the fuel channel pressure tube and the graphite blocks closes after approximately 17 years of plant operation. There is no safety justification available for the continued plant operation in this condition and the reactors are being retubed to avoid operation in this out of design condition, which may have negative impact on the fuel channel integrity. The loss of the mechanical integrity of fuel channel pressure tubes is a major safety concern for RBMK reactors since it may lead to overpressurization of the reactor cavity and consequently develop into a severe accident. In this report, information on the main design features of the RBMK reactor related to the fuel channel integrity is given. Further, detailed information on the fuel channel pressure tube and the graphite blocks with respect to their design, manufacture, in-service inspection, operating experience, ageing behaviour including degradation mechanisms is discussed in detail. The behaviour of the system fuel channel-graphite core including the corrective actions developed and implemented is discussed. Both normal operating conditions and accident conditions are addressed, considering also the gas gap closure process and its impact. The report also covers the fuel channel ducts. It is concluded in the report that for RBMK-1000 reactors and the adopted retubing strategy, limited local gas gap closure occurs at the time of pressure tube

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

International Nuclear Information System (INIS)

Masopust, R.

2003-01-01

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

Flux stability and power control in the Soviet RBMK-1000 reactors

International Nuclear Information System (INIS)

Meriwether, G.H.; McNeece, J.P.

1993-08-01

As a result of the Chernobyl accident, the Soviets have studied and implemented various design changes to improve the safety of the RBMK reactors. The safety measurements include modifications of the control rod configuration, fuel enrichment increase from 2.0 to 2.4 weight percent U-235, and installation of additional supplemental absorbers. The purpose of this study is to investigate the effects of increased fuel enrichment, different control rod positions, and supplemental absorber loadings on reactivity control, power distribution within the large RBMK core, and relative stability against power oscillations

Analysis methodology for RBMK-1500 core safety and investigations on corium coolability during a LWR severe accident

Energy Technology Data Exchange (ETDEWEB)

Jasiulevicius, Audrius

2003-07-01

This thesis presents the work involving two broad aspects within the field of nuclear reactor analysis and safety. These are: - development of a fully independent reactor dynamics and safety analysis methodology of the RBMK-1500 core transient accidents and - experiments on the enhancement of coolability of a particulate bed or a melt pool due to heat removal through the control rod guide tubes. The first part of the thesis focuses on the development of the RBMK-1500 analysis methodology based on the CORETRAN code package. The second part investigates the issue of coolability during severe accidents in LWR type reactors: the coolability of debris bed and melt pool for in-vessel and ex-vessel conditions. The first chapter briefly presents the status of developments in both the RBMK-1500 core analysis and the corium coolability areas. The second chapter describes the generation of the RBMK-1500 neutron cross section data library with the HELIOS code. The cross section library was developed for the whole range of the reactor conditions. The results of the benchmarking with the WIMS-D4 code and validation against the RBMK Critical Facility experiments is also presented here. The HELIOS generated neutron cross section data library provides a close agreement with the WIMS-D4 code results. The validation against the data from the Critical Experiments shows that the HELIOS generated neutron cross section library provides excellent predictions for the criticality, axial and radial power distribution, control rod reactivity worths and coolant reactivity effects, etc. The reactivity effects of voiding for the system, fuel assembly and additional absorber channel are underpredicted in the calculations using the HELIOS code generated neutron cross sections. The underprediction, however, is much less than that obtained when the WIMS-D4 code generated cross sections are employed. The third chapter describes the work, performed towards the accurate prediction, assessment and

The dry spent RBMK fuel cask storage site at the Ignalina NPP in Lithuania

International Nuclear Information System (INIS)

Penkov, V.V.; Diersch, R.

1999-01-01

At present, there are about 15,000 spent RBMK fuel assemblies stored in the water pools near the reactors at the Ignalina Nuclear Power Plant (INPP). Part of them are cut in two bundles and stored in standardized baskets in the pools. Each basket is loaded with 102 bundles. For long-term interim storage of this fuel, it was decided to use dry storage in casks. For this reason, the total activity to be stored is split into individual units (casks). Each cask represents a closed and independent safety system, fulfilling all safety-relevant requirements for both normal operational and hypothetical accidental conditions. The main safety relevant features of the storage cask system are: (1) Inherent safety system; (2) Double barrier system; (3) Passive cooling by natural convection; (4) Safety against accidents. The cask dry storage system is a cost effective and multi-functional system for storage, transport after the operation time and final disposal under consideration of additional protective elements. From an economical point of view, cask storage has a number of advantages. Two cask types have been intended for the INPP storage site: (1) The CASTOR RBMK cask made of ductile cast iron; (2) The CONSTOR RBMK sandwich cask made of an inner and outer steel shell and reinforced heavy concrete. The CASTOR RBMK and the CONSTOR RBMK casks are designed to withstand severe storage site accidents and with help of impact limiters - to fulfil the IAEA test criteria for type B(U)F packages. The INPP spent RBMK fuel storage site is designed as an open air storage for an operational time of 50 years. The casks are arranged on the concrete storage pad. The site is equipped with a crane for cask handling and technological buildings and security systems. The safety analyses for fuel and cask handling and for cask handling and for cask technology at the site have been made and accepted by the Lithuanian Competent Authority. (author)

RBMK full scope simulator gets virtual refuelling machine

International Nuclear Information System (INIS)

Khoudiakov, M.; Slonimsky, V.; Mitrofanov, S.

2006-01-01

The paper describes a continuation of efforts of an international Russian-Norwegian joint team to drastically increase operational safety during the refuelling process of an RBMK-type reactor by implementing a training simulator based on an innovative Virtual Reality (VR) approach. During the preceding stage of the project a display-based simulator was extended with VR models of the real Refueling Machine (RM) and its environment in order to improve both the learning process and operation's effectiveness. The simulator's challenge is to support the performance (operational activity) of RM operational staff firstly and to take major part in developing basic knowledge and skills as well as to keep skilled staff in close touch with the complex machinery of the Refueling Machine. At the given 2nd stage the functional scope of the VR-simulator was greatly enhanced - firstly, by connecting to the RBMK-unit full-scope simulator, and, secondly, by a training program and simulator model upgrade. (author)

Analysis methodology for RBMK-1500 core safety and investigations on corium coolability during a LWR severe accident

International Nuclear Information System (INIS)

Jasiulevicius, Audrius

2003-01-01

This thesis presents the work involving two broad aspects within the field of nuclear reactor analysis and safety. These are: - development of a fully independent reactor dynamics and safety analysis methodology of the RBMK-1500 core transient accidents and - experiments on the enhancement of coolability of a particulate bed or a melt pool due to heat removal through the control rod guide tubes. The first part of the thesis focuses on the development of the RBMK-1500 analysis methodology based on the CORETRAN code package. The second part investigates the issue of coolability during severe accidents in LWR type reactors: the coolability of debris bed and melt pool for in-vessel and ex-vessel conditions. The first chapter briefly presents the status of developments in both the RBMK-1500 core analysis and the corium coolability areas. The second chapter describes the generation of the RBMK-1500 neutron cross section data library with the HELIOS code. The cross section library was developed for the whole range of the reactor conditions. The results of the benchmarking with the WIMS-D4 code and validation against the RBMK Critical Facility experiments is also presented here. The HELIOS generated neutron cross section data library provides a close agreement with the WIMS-D4 code results. The validation against the data from the Critical Experiments shows that the HELIOS generated neutron cross section library provides excellent predictions for the criticality, axial and radial power distribution, control rod reactivity worths and coolant reactivity effects, etc. The reactivity effects of voiding for the system, fuel assembly and additional absorber channel are underpredicted in the calculations using the HELIOS code generated neutron cross sections. The underprediction, however, is much less than that obtained when the WIMS-D4 code generated cross sections are employed. The third chapter describes the work, performed towards the accurate prediction, assessment and

Deterministic Safety Technology for RBMK Reactors

Directory of Open Access Journals (Sweden)

F. D'Auria

2008-01-01

The paper summarizes the activities performed at NIKIET in Moscow and at University of Pisa (UNIPI in Pisa. A top-down approach is pursued in structuring the executive summary that includes the following sections: (i the safety needed for the RBMK NPP, (ii the roadmap, (iii\tthe adopted computational tools, (iv\tkey findings, (v\tEmphasis is given to the multiple pressure tube rupture (MPTR issue and the individual channel monitoring (ICM proposal.

Experiences from the LNPP-P and DSA review. Lessons learned from RBMK safety studies

International Nuclear Information System (INIS)

Mankamo, T.; Marttila, J.; Reponen, H.

2000-09-01

RBMK is the Russian acronym for 'Channelized Large Power Reactor'. The Soviet-designed RBMK plants deviate substantially from typical Western BWR or PWR plants. The safety of the RBMK plants has raised severe concerns since the major accident at Chernobyl Unit 4 in 1986. In addition, a fire destroyed the turbine hall of Chernobyl Unit 2 in 1991 resulting in a near-accident: the reactor cooling could only be maintained through improvised measures. Another well-known fire event is the control cable room fire at Ignalina Unit 2 in 1989, which led to a partial loss of the main control room functions. After the collapse of Soviet Union several multilateral safety programs were started to evaluate and improve the safety of the RBMK plants. A Probabilistic and Deterministic Safety Assessment (P and DSA) of the Leningrad Nuclear Power Plant (LNPP) Unit 2 was started in 1996. Phase 2 of the project was completed in January 1999. A Peer Review was performed by Russian and Western experts. This report describes the insights from the RBMK risk studies, especially from the LNPP P and DSA with emphasis on the deeper understanding of the risk-important design factors and identification of possible ways to increase safety. LNPP P and DSA has meant a significant progress in this respect. Despite of its certain limitations P and DSA Phase 2 could point out short-term measures, which substantially reduced the risk of identified weaknesses, mostly related to the reliability of the emergency feedwater function and its support systems. The findings of LNPP P and DSA and the review recommendations emphasise the extensions needed to the analysis scope. The spreading and other influences of fires and floods between connected spaces should be analysed because of incomplete separation and protection in these regards in the 16st generation RBMK plants. High priority should be given to the analysis of external hazards, which were found important at the Loviisa NPP on the Northern side of the

Experiences from the LNPP-P and DSA review. Lessons learned from RBMK safety studies

Energy Technology Data Exchange (ETDEWEB)

Mankamo, T. [Avaplan Oy (Finland); Marttila, J.; Reponen, H. [Radiation and Nuclear Safety Authority, Helsinki (Finland)

2000-09-01

RBMK is the Russian acronym for 'Channelized Large Power Reactor'. The Soviet-designed RBMK plants deviate substantially from typical Western BWR or PWR plants. The safety of the RBMK plants has raised severe concerns since the major accident at Chernobyl Unit 4 in 1986. In addition, a fire destroyed the turbine hall of Chernobyl Unit 2 in 1991 resulting in a near-accident: the reactor cooling could only be maintained through improvised measures. Another well-known fire event is the control cable room fire at Ignalina Unit 2 in 1989, which led to a partial loss of the main control room functions. After the collapse of Soviet Union several multilateral safety programs were started to evaluate and improve the safety of the RBMK plants. A Probabilistic and Deterministic Safety Assessment (P and DSA) of the Leningrad Nuclear Power Plant (LNPP) Unit 2 was started in 1996. Phase 2 of the project was completed in January 1999. A Peer Review was performed by Russian and Western experts. This report describes the insights from the RBMK risk studies, especially from the LNPP P and DSA with emphasis on the deeper understanding of the risk-important design factors and identification of possible ways to increase safety. LNPP P and DSA has meant a significant progress in this respect. Despite of its certain limitations P and DSA Phase 2 could point out short-term measures, which substantially reduced the risk of identified weaknesses, mostly related to the reliability of the emergency feedwater function and its support systems. The findings of LNPP P and DSA and the review recommendations emphasise the extensions needed to the analysis scope. The spreading and other influences of fires and floods between connected spaces should be analysed because of incomplete separation and protection in these regards in the 16st generation RBMK plants. High priority should be given to the analysis of external hazards, which were found important at the Loviisa NPP on the Northern

Neutron field control cybernetics model of RBMK reactor operator

International Nuclear Information System (INIS)

Polyakov, V.V.; Postnikov, V.V.; Sviridenkov, A.N.

1992-01-01

Results on parameter optimization for cybernetics model of RBMK reactor operator by power release control function are presented. Convolutions of various criteria applied previously in algorithms of the program 'Adviser to reactor operator' formed the basis of the model. 7 refs.; 4 figs

Coordinated research programme on safety of RBMK type NPPs in relation to external events. V. 1. Working material

International Nuclear Information System (INIS)

1999-01-01

The present volume is a collection of progress reports which have been submitted within the scope of the CRP on safety of RBMK type NPPs in relation to external events including seismic related papers and man-induced events (explosions and airplane crash). It includes papers concerned with experience related to RBMK equipment testing and calculations of seismic resistance, soil-structure interactions analysis, safety assurance, aircraft impact qualification and other external events for RBMK type NPP, seismic stability of NPPs in Eastern Europe, probabilistic assessment of NPP safety under aircraft impact, dynamic analysis of NPPs, screening of external hazards for NPP

Evolution of the hafnium isotopic composition in the RBMK reactor

International Nuclear Information System (INIS)

Jurkevicius, A.; Remeikis, V.

2002-01-01

The isotopic composition of hafnium in the radial neutron flux sensor of the RBMK-1500 reactor, the rates of the neutron absorption on Hf isotopes and the neutron spectrum in the sensor were numerically modeled. The sequence SAS2 (Shielding Analysis Sequence) program from the package SCALE 4.4A and the HELIOS code system were used for calculations. It has been obtained that the overall neutron absorption rates in hafnium for the sensors located in the 2.4 % and 2.6 % enrichment uranium-erbium nuclear fuel assemblies are by 16 % and 19 % lower than in the 2.0 % enrichment uranium nuclear fuel assemblies. The overall neutron absorption rate in hafnium decreases 2.70-2.75 times due to the sensor burnup to 5800 MW d. The sensitivity of the Hf sensors to the thermal neutron flux increases twice due to the nuclear fuel assembly burnup to 3000 MW d. The corrective factors ξ d (I) at the different integral current I of the sensors and ξ td (E) at the different burnup E of the nuclear fuel assemblies were calculated. The obtained dependence ξ d (I) calculated numerically was compared to the experimental one determined by comparing signals of the fresh sensor and the sensor with the integral current I and by processing repeated calibration results of Hf sensors in RBMK-1500 reactors. The relative relationship coefficients K T (T FA ) were found for all RBMK-1500 nuclear fuel types. (author)

Validation of coupled Relap5-3D code in the analysis of RBMK-1500 specific transients

International Nuclear Information System (INIS)

Evaldas, Bubelis; Algirdas, Kaliatka; Eugenijus, Uspuras

2003-01-01

This paper deals with the modelling of RBMK-1500 specific transients taking place at Ignalina NPP. These transients include: measurements of void and fast power reactivity coefficients, change of graphite cooling conditions and reactor power reduction transients. The simulation of these transients was performed using RELAP5-3D code model of RBMK-1500 reactor. At the Ignalina NPP void and fast power reactivity coefficients are measured on a regular basis and, based on the total reactor power, reactivity, control and protection system control rods positions and the main circulation circuit parameter changes during the experiments, the actual values of these reactivity coefficients are determined. Graphite temperature reactivity coefficient at the plant is determined by changing graphite cooling conditions in the reactor cavity. This type of transient is very unique and important from the gap between fuel channel and the graphite bricks model validation point of view. The measurement results, obtained during this transient, allowed to determine the thermal conductivity coefficient for this gap and to validate the graphite temperature reactivity feedback model. Reactor power reduction is a regular operation procedure during the entire lifetime of the reactor. In all cases it starts by either a scram or a power reduction signal activation by the reactor control and protection system or by an operator. The obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviours of the separate MCC thermal-hydraulic parameters as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. Reasonable agreement of the measured and the calculated total reactor power change in time demonstrates the correct modelling of the neutronic processes taking place in RBMK- 1500 reactor core. And finally, the performed validation of RELAP5-3D model of Ignalina NPP RBMK-1500

International cooperation in accident analysis of RBMK reactors

International Nuclear Information System (INIS)

Kaliatka, A.; Isag

2005-01-01

Chouha Michel (Institute for Radiological Protection and Nuclear Safety), D'Auria Francesco (Institute of Pisa), Kaliatka Algirdas (Lithuanian Energy Institute), Uspuras Eugenijus (Lithuanian Energy Institute). The safety of nuclear power plants is a primary concern of the European Union (EU) and its Member States. In the early 1990s, the European Union decided to take a prominent role in international efforts to help the New Independent States (NIS) and countries of central Europe to ensure the safety of their nuclear reactors. The Commission's approach to nuclear safety in central and Eastern Europe and the NIS is based on two main objectives, which are fully in line with the policy of the international community as decided by the G7 in 1992: 1) In the short term, to improve operational safety; to make near term technical improvements to plants based on safety assessments and to enhance regulatory regimes; 2) In the longer term, to examine the scope for replacing less safe plants by the development of alternative energy sources and more efficient use of energy and to examine the potential for upgrading plants of more recent design. In this paper the safety concerns, related to RBMK type reactors (Russian acronym for 'Channelized Large Power Reactor) are discussed. These reactors were not exported and were built exclusively in the territory of the former Soviet Union. There are presently plants at Saint Petersburg (Sosnovy Bor), Kursk, Chernobyl and Smolensk. A total of 17 such reactors have been built and 12 are currently in operation. Two international projects: TACIS project 'Development of a code system for severe accident analysis in RBMK reactors' and PHARE projects 'Support to VATESI for Important Tasks Relevant to the Licensing Activities of Ignalina Nuclear Power Plant' are presented. The aim of the TACIS project is to help the Russian Authorities to build such capabilities, for their RBMK nuclear power plants (NPPs). The drawing of the Tacis nuclear

The Soviet RBMK-1000 containment system

International Nuclear Information System (INIS)

Joosten, J.K.

1988-01-01

Following the accident in April, 1986, considerable attention was focused on the failure of the containment at the Chernobyl RBMK-1000 nuclear power plant. Conflicting statements arose regarding the nature of the plant's containment system primarily because of terminology differences, translation difficulties and lack of reliable information. This article, based on reports and briefings by the Soviet delegation, during the post-accident review meetings in Vienna and prior publications is intended to clarify perceptions of the Soviet RMBK-1000 nuclear power plant containment system design, and its relevance to containment management concepts. (author)

Application of 3D coupled code ATHLET-QUABOX/CUBBOX for RBMK-1000 transients after graphite block modernization

Energy Technology Data Exchange (ETDEWEB)

Samokhin, Aleksei [Scientific and Engineering Centre for Nuclear and Radiation Safety (SEC NRS), Moscow (Russian Federation); Zilly, Matias [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany)

2016-11-15

This work describes the application and the results of transient calculations for the RBMK-1000 with the coupled code system ATHLET 2.2A-QUABOX/CUBBOX which was developed in GRS. Within these studies the planned modernization of the graphite blocks of the RBMK-1000 reactor is taken into account. During the long-term operation of the uranium-graphite reactors RBMK-1000, a change of physical and mechanical properties of the reactor graphite blocks is observed due to the impact of radiation and temperature effects. These have led to a deformation of the reactor graphite columns and, as a result, a deformation of the control and protection system (CPS) and of fuel channels. Potentially, this deformation can lead to problems affecting the smooth movement of the control rods in the CPS channels and problems during the loading and unloading of fuel assemblies. The present paper analyzes two reactivity insertion transients, each taking into account three graphite removal scenarios. The presented work is directly connected with the modernization program of the RBMK- 1000 reactors and has an important contribution to the assessment of the safety-relevant parameters after the modification of the core graphite blocks.

Temperature control of the graphite stack of the reactor RBMK-1500

International Nuclear Information System (INIS)

Lesnoj, S.

1998-01-01

The paper includes general information about RBMK-1500 reactor, construction features and main technical data; graphite moderator stack, temperature channel, thermocouple TXA-1379, its basic technical and metrologic parameters as well as its advantages and disadvantages

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

Generic repository concept for RBMK-1500 spent nuclear fuel disposal in crystalline rocks in Lithuania

International Nuclear Information System (INIS)

Poskas, P.; Brazauskaite, A.; Narkunas, E.; Smaizys, A.; Sirvydas, A.

2006-01-01

During 2002-2005 investigations on possibilities to dispose of spent nuclear fuel (SNF) in Lithuania were performed with support of Swedish experts. Disposal concept for RBMK-1500 SNF in crystalline rocks in Lithuania is based on Swedish KBS-3 concept with SNF emplacement into the copper canister with cast iron insert. The bentonite and its mixture with crushed rock are also foreseen as buffer and backfill material. In this paper modelling results on thermal, criticality and other important disposal characteristics for RBMK-1500 SNF fuel emplaced in copper canisters are presented. Based on thermal calculations, the distances between the canisters and between the tunnels were justified. Criticality calculations for the canister with fresh fuel with 2.8 % 235 U enrichment demonstrated that effective neutron multiplication factor k eff values are less than allowable value of 0.95. Dose calculations have shown that total equivalent dose rate from the canister with 50 years stored RBMK-1500 SNF is rather high and is defined mainly by the γ radiation. (author)

Verification of RBMK-1500 reactor main circulation circuit model with Cathare V1.3L

International Nuclear Information System (INIS)

Jasiulevicius, A.

2001-01-01

Among other computer codes, French code CATHARE is also applied for RBMK reactor calculations. In this paper results of such application for Ignalina NPP reactor (RBMK-1500 type) main circulation circuit are presented. Three transients calculations were performed: all main circulation pumps (MCP) trip, trip of one main circulation pump and trip of one main circulation pump without a closure of check valve on the pump line. Calculation results were compared to data from the Ignalina NPP, where all these transients were recorded in the years 1986, 1996 and 1998. The presented studies prove the capability of the CATHARE code to treat thermal-hydraulic transients with a reactor scram in the RBMK, in case of single or multiple pump trips. However, the presented model needs further improvements in order to simulate loss of coolant accidents. For this reason, emergency core cooling system should be included in the model. Additional model improvement is also needed in order to gain more independent pressure behavior in both loops. Also, flow rates through the reactor channels should be modeled by dividing channels into several groups, referring to channel power (in RBMK power produced in a channel, located in different parts of the core is not the same). The point-neutron kinetic model of the CATHARE code is not suitable to predict transients when the reactor is operating at a nominal power level. Such transients would require the use of 3D-neutron kinetics model to describe properly the strong space-time effect on the power distribution in the reactor core

Study of possibility using LANL PSA-methodology for accident probability RBMK researches

International Nuclear Information System (INIS)

Petrin, S.V.; Yuferev, V.Y.; Zlobin, A.M.

1995-01-01

The reactor facility probabilistic safety analysis methodologies are considered which are used at U.S. LANL and RF NIKIET. The methodologies are compared in order to reveal their similarity and differences, determine possibilities of using the LANL technique for RBMK type reactor safety analysis. It is found that at the PSA-1 level the methodologies practically do not differ. At LANL the PHA, HAZOP hazards analysis methods are used for more complete specification of the accounted initial event list which can be also useful at performance of PSA for RBMK. Exchange of information regarding the methodology of detection of dependent faults and consideration of human factor impact on reactor safety is reasonable. It is accepted as useful to make a comparative study result analysis for test problems or PSA fragments using various computer programs employed at NIKIET and LANL

RELAP5-3D code validation of RBMK-1500 reactor reactivity measurement transients

International Nuclear Information System (INIS)

Kaliatka, Algirdas; Bubelis, Evaldas; Uspuras, Eugenijus

2003-01-01

This paper deals with the modeling of transients taking place during the measurements of the void and fast power reactivity coefficients performed at Ignalina NPP. The simulation of these transients was performed using RELAP5-3D code model of RBMK-1500 reactor. At the Ignalina NPP void and fast power reactivity coefficients are measured on a regular basis and, based on the total reactor power, reactivity, control and protection system control rods positions and the main circulation circuit parameter changes during the experiments, the actual values of these reactivity coefficients are determined. Following the simulation of the two above mentioned transients with RELAP5-3D code, a conclusion was made that the obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviors of the separate MCC thermal-hydraulic parameters as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. The calculated reactivity and the total reactor core power behavior in time are also in reasonable agreement with the measured plant data. Despite of the small differences, RELAP5-3D code predicts reactivity and the total reactor core power behavior during the transients in a reasonable manner. Reasonable agreement of the measured and the calculated total reactor power change in time demonstrates the correct modeling of the neutronic processes taking place in RBMK-1500 reactor core

Safety assessment of proposed improvements to RBMK nuclear power plants

International Nuclear Information System (INIS)

1993-03-01

The purpose of this report is to summarize the findings and recommendations of a Consultants Meeting convened by the IAEA in Vienna (27 October - 5 November 1992) to review new design features and modifications proposed or already implemented for RBMK reactors. This information was provided in four technical areas, namely: Core Monitoring and Control, Pressure Boundary Integrity, Accident Mitigation and Electric Power Supply. The report also presents the status of the modifications at the plants as given by the RBMK specialists. The limited information available and the time constraints did not allow the review to be conducted at the level of a peer review, and the findings and recommendations made reflect the limited scope of the review. More detailed reviews and analysis focusing on selected safety issues are required and should be conducted on a generic and plant specific basis as appropriate. In Chapters 2-5 of the report the main findings and recommendations for the four topical areas reviewed are summarized. Appendices I-IV reflect the results of the discussions held at the meeting and provide more detailed information on the review. 17 refs, 27 figs, 17 tabs

Tests of the RBMK-1500 reactor fuel assemblies in the Leningrad reactor

International Nuclear Information System (INIS)

Aden, V.C.; Varovin, I.A.; Vorontsov, B.A.

1981-01-01

Test of fuel assemblies of the RBMK-1500 reactor is conducted in the reactor of the Leningrad NPP unit 2 for proving the calculational values of critical power of the RBMK-1500 reactor fuel assemblies adopted in design. The experiment presupposes the maximal approximation of the fuel assembly operation parameters to the calculational critical parameters without bringing into the mode of heat transfer crisis. The experiments are carried out at 500, 850 and 900 MW(el) of the reactor. The maximal channel power made up 472 kW at 20.5 t/h coolant flow rate and 49% mass steam content at the outlet of the channel. It was concluded that there was supply up to the heat transfer crisis in all the investigated modes. Data of temperature measurings of the fuel element cans, readings of the devices of the failure control system of the fuel element cans and external inspection of the assemblies after the tests testify to it [ru

Status and development of RBMK fuel rods and reactor materials

International Nuclear Information System (INIS)

Bibilashvili, Yu.K.; Reshetnikov, F.G.; Ioltukhovsky, A.G.

1998-01-01

The paper presents current status and development of RBMK fuel rods and reactor materials. With regard to fuel rod cladding the following issues have been discussed: corrosion, tensile properties, welding technology and testing of an alternative cladding alloy with a composition of Zr-Nb-Sn-Fe. Erbium doped fuel has been suggested for safety improvement. Also analysis of fuel reliability is presented in the paper. (author)

Calculation of the Thermal State of the Graphite Moderator of the RBMK Reactor

Directory of Open Access Journals (Sweden)

Vorobiev Alexander V.

2017-01-01

Full Text Available This work is devoted to study the temperature field of the graphite stack of the RBMK reactor. In work was analyzed the influence of contact pressure between the components of the masonry on the temperature of the graphite moderator.

Actinide-only and full burn-up credit in criticality assessment of RBMK-1500 spent nuclear fuel storage cask using axial burn-up profile

Energy Technology Data Exchange (ETDEWEB)

Barkauskas, V., E-mail: vytenis.barkauskas@ftmc.lt; Plukiene, R., E-mail: rita.plukiene@ftmc.lt; Plukis, A., E-mail: arturas.plukis@ftmc.lt

2016-10-15

Highlights: • RBMK-1500 fuel burn-up impact on k{sub eff} in the SNF cask was calculated using SCALE 6.1. • Positive end effect was noticed at certain burn-up for the RBMK-1500 spent nuclear fuel. • The non-uniform uranium depletion is responsible for the end effect in RBMK-1500 SNF. • k{sub eff} in the SNF cask does not exceed a value of 0.95 which is set in the safety requirements. - Abstract: Safe long-term storage of spent nuclear fuel (SNF) is one of the main issues in the field of nuclear safety. Burn-up credit application in criticality analysis of SNF reduces conservatism of usually used fresh fuel assumption and implies a positive economic impact for the SNF storage. Criticality calculations of spent nuclear fuel in the CONSTOR® RBMK-1500/M2 cask were performed using pre-generated ORIGEN-ARP spent nuclear fuel composition libraries, and the results of the RBMK-1500 burn-up credit impact on the effective neutron multiplication factor (k{sub eff}) have been obtained and are presented in the paper. SCALE 6.1 code package with the STARBUCKS burn-up credit evaluation tool was used for modeling. Pre-generated ARP (Automatic Rapid Processing) crosssection libraries based on ENDF/B-VII cross section library were used for fast burn-up inventory modeling. Different conditions in the SNF cask were modeled: 2.0% and 2.8% initial enrichment fuel of various burn-up and water density inside cavities of the SNF cask. The fuel composition for the criticality analysis was chosen taking into account main actinides and most important fission products used in burn-up calculations. A significant positive end effect is noticed from 15 GWd/tU burn-up for 2.8% enrichment fuel and from 9 GWd/tU for 2.0% enrichment fuel applying the actinide-only approach. The obtained results may be applied in further evaluations of the RBMK type reactor SNF storage as well as help to optimize the SNF storage volume inside the CONSTOR® RBMK-1500/M2 cask without compromising criticality

Actinide-only and full burn-up credit in criticality assessment of RBMK-1500 spent nuclear fuel storage cask using axial burn-up profile

International Nuclear Information System (INIS)

Barkauskas, V.; Plukiene, R.; Plukis, A.

2016-01-01

Highlights: • RBMK-1500 fuel burn-up impact on k_e_f_f in the SNF cask was calculated using SCALE 6.1. • Positive end effect was noticed at certain burn-up for the RBMK-1500 spent nuclear fuel. • The non-uniform uranium depletion is responsible for the end effect in RBMK-1500 SNF. • k_e_f_f in the SNF cask does not exceed a value of 0.95 which is set in the safety requirements. - Abstract: Safe long-term storage of spent nuclear fuel (SNF) is one of the main issues in the field of nuclear safety. Burn-up credit application in criticality analysis of SNF reduces conservatism of usually used fresh fuel assumption and implies a positive economic impact for the SNF storage. Criticality calculations of spent nuclear fuel in the CONSTOR® RBMK-1500/M2 cask were performed using pre-generated ORIGEN-ARP spent nuclear fuel composition libraries, and the results of the RBMK-1500 burn-up credit impact on the effective neutron multiplication factor (k_e_f_f) have been obtained and are presented in the paper. SCALE 6.1 code package with the STARBUCKS burn-up credit evaluation tool was used for modeling. Pre-generated ARP (Automatic Rapid Processing) crosssection libraries based on ENDF/B-VII cross section library were used for fast burn-up inventory modeling. Different conditions in the SNF cask were modeled: 2.0% and 2.8% initial enrichment fuel of various burn-up and water density inside cavities of the SNF cask. The fuel composition for the criticality analysis was chosen taking into account main actinides and most important fission products used in burn-up calculations. A significant positive end effect is noticed from 15 GWd/tU burn-up for 2.8% enrichment fuel and from 9 GWd/tU for 2.0% enrichment fuel applying the actinide-only approach. The obtained results may be applied in further evaluations of the RBMK type reactor SNF storage as well as help to optimize the SNF storage volume inside the CONSTOR® RBMK-1500/M2 cask without compromising criticality safety.

Probabilistic safety analysis applied to RBMK reactors

International Nuclear Information System (INIS)

Gerez Martin, L.; Fernandez Ramos, P.

1995-01-01

The project financed by the European Union ''Revision of RBMK Reactor Safety was divided into nine Topic Groups dealing with different aspects of safety. The area covered by Topic Group 9 was Probabilistic Safety Analysis. TG9 will have touched on some of the problems discussed by other groups, although in terms of the systematic quantification of the impact of design characteristics and RBMK reactor operating practices on the risk of core damage. On account of the reduced time scale and the resources available for the project, the analysis was made using a simplified method based on the results of PSAs conducted in Western countries and on the judgement of the group members. The simplifies method is based on the concepts of Qualification, Redundancy and Automatic Actuation of the systems considered. PSA experience shows that systems complying with the above-mentioned concepts have a failure probability of 1.0E-3 when redundancy is simple, ie two similar equipment items capable of carrying out the same function. In general terms, this value can be considered to be dominated by potential common cause failures. The value considered above changes according to factors that have a positive effect upon it, such as an additional redundancy with a different equipment item (eg a turbo pumps and a motor pump), individual trains with good separations, etc, or a negative effect, such as the absence of suitable periodical tests, the need for operators to perform manual operations, etc. Similarly, possible actions required by the operator during accident sequences are assigned failure probability values between 1 and 1.0E-4, according to the complexity of the action (including local actions to be performed outside the control room) and the time available

A limit load analysis of RBMK-1500 reactor structures

International Nuclear Information System (INIS)

Petkevichius, K.; Dundulis, G.; Marchertas, A.

1996-09-01

Presented is a mathematical model of Ignalina NPP facilities where the transported hermetic containers CASTOR RBMK will be located. Analysis of the mathematical model provides resultant stresses caused by free falling container with spent fuel. The result yield wall deflections and maximum stresses in the reinforcing bars of the structure, which maintains the integrity of these facilities of the Ignalina NPP. They indicate the excessive deflections of the walls and stresses in reinforcement in certain areas of the facilities. The ALGOR computer code is used for the calculation. (author). 3 figs., 6 refs

Analysis of realization of the water chemistry modes in the NPP with the RBMK-1000 and main directions of their improvement

International Nuclear Information System (INIS)

Kritskij, V.G.; Tyapkov, V.F.; Belous, V.N.; Egorova, T.M.; Gost'kov, V.V.; Tishkov, V.M.; Yatsko, O.V.

2005-01-01

Paper deals with the analysis of normalization of the RBMK reactor NPP water chemistry conditions. One analyzed the imposed restrictions at deviation of the normalized parameters from the ones recommended for the normal operating conditions. Paper contains data on water chemistry management and describes measures to improve radiation situation near NPP reactor equipment. One studied the reasons of corrosion damage of the RBMK-1000 reactor NPP pipelines and the ways to prevent them via optimization and improvement of water chemistry conditions [ru

Impact of Zr + 2.5% Nb alloy corrosion upon operability of RBMK-1000 fuel channels

International Nuclear Information System (INIS)

Kovyrshin, V.; Zaritsky, N.

1999-01-01

The basic components of RBMK-1000 core (fuel channels, bimetal adapters, claddings of fuel elements, etc.) are of zirconium alloys. Their corrosion is one of factors influencing upon fuel channels operability. Dynamics of channel tubes nodular corrosion development is presented by the results of in-reactor investigation at ChNPP. Radiation-induced mechanism of corrosion damage of tubes surface in contact with coolant was formulated and substantiated by data of post-reactor studies. Within the certain time period of operation corrosion of zirconium alloy of lower bimetal adapter along with removal from there of corrosion products are predominant within the whole process of reactor elements corrosion. The experimental and calculating method was proposed and substantiated to predict time duration up to loss of fuel channels leak tightness. The approaches were generalized to control state of fuel channels material to assess their operability under operation of RBMK-1000 reactors. (author)

Power distribution monitoring and control in the RBMK type reactors

International Nuclear Information System (INIS)

Emel'yanov, I.Ya.; Postnikov, V.V.; Volod'ko, Yu.I.

1980-01-01

Considered are the structures of monitoring and control systems for the RBMK-1000 reactor including three main systems with high independence: the control and safety system (CSS); the system for physical control of energy distribution (SPCED) as well as the Scala system for centralized control (SCC). Main functions and peculiarities of each system are discussed. Main attention is paid to new structural solutions and new equipment components used in these systems. Described are the RBMK operation software and routine of energy distribution control in it. It is noted that the set of reactor control and monitoring systems has a hierarchical structure, the first level of which includes analog systems (CSS and SPCED) normalizing and transmitting detector signals to the systems of the second level based on computers and realizing computer data processing, data representation to the operator, automatic (through CSS) control for energy distribution, diagnostics of equipment condition and local safety with provision for existing reserves with respect to crisis and thermal loading of fuel assemblies. The third level includes a power computer carrying out complex physical and optimization calculations and providing interconnections with the external computer of power system. A typical feature of the complex is the provision of local automatic safety of the reactor from erroneous withdrawal of any control rod. The complex is designed for complete automatization of energy distribution control in reactor in steady and transient operation conditions

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

International Nuclear Information System (INIS)

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

1983-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1983-12-01

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

Seismic response analyses of turbine hall and electrical building of RBMK-1000 MW type NPP

International Nuclear Information System (INIS)

Jordanov, M.J.; Karparov, K.T.

2003-01-01

This paper addresses results obtained during the study of turbine hall and electrical building of RBMK-1000 MW pair units at Leningradskaya NPP (LNPP) for seismic event. The study was performed in the frame of the Coordinated Research Program of the International Atomic Agency (IAEA) on Safety of RBMK type Nuclear Power Plants (NPP) in Relation of External Events. A 3-D finite element model of Main Building Complex was developed and seismic response analyses were performed taking into account the soil-structure interaction (SSI). The standard mode superposition method was used for evaluation of dynamic response of structure in time domain. The structure was assumed surface founded at the basemat level. Seismic response analyses were carried out considering shear wave propagation pattern for the input motion. The in-structure time histories and response spectra were generated in referenced locations. Conclusions are drawn for the reliability of the structural response evaluation considering the soil-structure interaction effects. (author)

About water chemistry influence on equipment reliability of NPP with RBMK-1000

International Nuclear Information System (INIS)

Berezina, I.G.; Styazhkin, P.S.; Kritskij, V.G.

2001-01-01

In the paper the experience of a quantitative valuation of coolant quality influence on a reliability of some equipment elements of NPP with RBMK-1000 is offered. The choice is made of coolant quality integral parameter. The connection between indices values of coolant quality and reliability of major elements of circulation circuit equipment (including fuel claddings) is established. The reliability improvement of equipment elements operation is supported by high water chemistry quality. (orig.)

Results of comparative RBMK neutron computation using VNIIEF codes (cell computation, 3D statics, 3D kinetics). Final report

Energy Technology Data Exchange (ETDEWEB)

Grebennikov, A.N.; Zhitnik, A.K.; Zvenigorodskaya, O.A. [and others

1995-12-31

In conformity with the protocol of the Workshop under Contract {open_quotes}Assessment of RBMK reactor safety using modern Western Codes{close_quotes} VNIIEF performed a neutronics computation series to compare western and VNIIEF codes and assess whether VNIIEF codes are suitable for RBMK type reactor safety assessment computation. The work was carried out in close collaboration with M.I. Rozhdestvensky and L.M. Podlazov, NIKIET employees. The effort involved: (1) cell computations with the WIMS, EKRAN codes (improved modification of the LOMA code) and the S-90 code (VNIIEF Monte Carlo). Cell, polycell, burnup computation; (2) 3D computation of static states with the KORAT-3D and NEU codes and comparison with results of computation with the NESTLE code (USA). The computations were performed in the geometry and using the neutron constants presented by the American party; (3) 3D computation of neutron kinetics with the KORAT-3D and NEU codes. These computations were performed in two formulations, both being developed in collaboration with NIKIET. Formulation of the first problem maximally possibly agrees with one of NESTLE problems and imitates gas bubble travel through a core. The second problem is a model of the RBMK as a whole with imitation of control and protection system controls (CPS) movement in a core.

Improvement of Algorithms for Pressure Maintenance Systems in Drum-Separators of RBMK-1000 Reactors

Energy Technology Data Exchange (ETDEWEB)

Aleksakov, A. N., E-mail: yankovskiy.k@nikiet.ru; Yankovskiy, K. I. [JSC “N. A. Dollezhal Research and Development Institute of Power Engineering (NIKIET),” (Russian Federation); Dunaev, V. I.; Kushbasov, A. N. [JSC “Diakont,” (Russian Federation)

2015-05-15

The main tasks and challenges for pressure regulation in the drum-separators of RBMK-1000 reactors are described. New approaches to constructing algorithms for pressure control in drum-separators by electro-hydraulic turbine control systems are discussed. Results are provided from tests of the operation of modernized pressure regulators during fast transients with reductions in reactor power.

Thermal-Hydraulic Analysis of Coolant Flow Decrease in Fuel Channels of Smolensk-3 RBMK during GDH Blockage Event

International Nuclear Information System (INIS)

Costa, A. L.; Cherubini, M.; D'Auria, F.; Giannotti, W.; Moskalev, A.

2007-01-01

One of the transients that have received considerable attention in the safety evaluation of RBMK reactors is the partial break of a group distribution header (GDH). The coolant flow rate blockage in one GDH might lead to excessive heat-up of the pressure tubes and can result in multiple fuel channels (FC) ruptures. In this work, the GDH flow blockage transient has been studied considering the Smolensk-3 RBMK NPP (nuclear power plant). In the RBMK, each GDH distributes coolant to 40-43 FC. To investigate the behavior of each FC belonging to the damaged GDH and to have a more realistic trend, one (affected) GDH has been schematised with its forty-two FC, one by one. The calculations were performed using the 0-D NK (neutron kinetic) model of the RELAP5-3.3 stand-alone code. The results show that, during the event, the mass flow rate is disturbed differently according to the power distribution established for each FC in the schematization. The start time of the oscillations in mass flow rate depends strongly on the attributed power to each FC. It was also observed that, during the event, the fuel channels at higher thermal power values tend to undergo first cladding rupture leaving the reactor to scram and safeguarding all the other FCs connected to the affected GDH.

Soil-structural interaction analysis of RBMK type NPP for seismic event. Progress report. From 1 July 1998 - 30 June 1999

International Nuclear Information System (INIS)

1999-01-01

The objective of the project is to assess the structural behavior and safety capacity of a RBMK-1000 MW Main Building Complex under critical combination of loads including seismic events. This project is part of the Coordinated Research Program carried out by International Atomic Energy Agency on safety of RBMK Type Nuclear Power Plants (NPP) in Relation to External Events. The nuclear power plant considered for this study is the Sosnovy Bor NPP, located near St.Petersburg, Russia. The Soviet standard design RBMK-1000 MW type units installed in Sosnovy Bor NPP were originally designed for a Safe Shutdown Earthquake (SSE) with a peak ground acceleration (PGA) of 0.1 g. The relevant response spectra are not available for reference and assessment. The new international requirements for nuclear power plants in operation require site specific seismic hazard studies as a basis for the definition of a Review Level Earthquake (RLE) for reassessment of the structures and safety related equipment Ell - As the RLE site specific seismic data is still not available, the RLE earthquake spectra for Kozloduy NPP scaled to PGA=0.1 g were used in this study. This value is intentionally chosen for comparison purposes. The Russian design requirements (if design floor response spectra are available) will be compared with the international regulations. The scope of the study is to perform a Soil-Structure Interaction (SSI) seismic response analysis of the referenced RBMK-11000 MW. Main Building Complex to evaluate the effect on the structural response of a greater than design earthquake. The analysis is focused on a realistic assessment of the structural response to a potentially higher earthquake level instead of a conservative design type analysis. Special attention is paid on the seismic response of the sub-structures in the safe shutdown path, as well as on the locations of the heavy equipment

Leak-before-break assessment of RBMK-1500 fuel channel in case of delayed hydride cracking

International Nuclear Information System (INIS)

Klimasauskas, A.; Grybenas, A.; Makarevicius, V.; Nedzinskas, L.; Levinskas, R.; Kiselev, V.

2003-01-01

One of the factors determining remaining lifetime of Zr-2.5% Nb fuel channel (FC) is the amount of hydrogen dissolved during corrosion process. When the concentration of hydrogen exceeds the terminal solid solubility limit zirconium hydrides are precipitated. As a result form necessary conditions for delayed hydride cracking (DHC). Data from the RBMK-1500 fuel channel tubes (removed from service) shows that hydrogen in some cases distributes unevenly and hydrogen concentration can differ several times between individual FC tubes or separate zones of the same tube and possibly, can reach dangerous levels in the future. Consequently, lacking statistical research data, it is difficult to forecast increase of hydrogen concentration and formation of DHC. So it is important to verify if under the most unfavorable situation leak before break condition will be satisfied in the case of DHC. To estimate possible DHC rates in RBMK 1500 FC pressure tubes experiments were done in the following order: hydriding of the Zr-2.5Nb pressure tube material to the required hydrogen concentration; hydrogen analysis; machining of specimens, fatigue crack formation in the axial direction, DHC testing; average crack length measurement and DHC velocity calculation. During the tests in average DHC values were determined at 283, 250 and 144 degC (with hydrogen concentrations correspondingly 76, 54 and 27 ppm). The fracture resistance dependence from hydrogen concentration was measured at 20 degC. To calculate leak through the postulated flaw, statistical distribution of DHC surface irregularity was determined. Leak before break analysis was carried out according to requirements of RBMK 1500 regulatory documents. J integral and crack opening were calculated using finite element method. Loading of the FC was determined using RELAP5 code. Critical crack length was calculated using R6 and J-integral methods. Coolant flow rate through the postulated crack was estimated using SQUIRT software

Thermal-hydraulic codes validation for safety analysis of NPPs with RBMK

International Nuclear Information System (INIS)

Brus, N.A.; Ioussoupov, O.E.

2000-01-01

This work is devoted to validation of western thermal-hydraulic codes (RELAP5/MOD3 .2 and ATHLET 1.1 Cycle C) in application to Russian designed light water reactors. Such validation is needed due to features of RBMK reactor design and thermal-hydraulics in comparison with PWR and BWR reactors, for which these codes were developed and validated. These validation studies are concluded with a comparison of calculation results of modeling with the thermal-hydraulics codes with the experiments performed earlier using the thermal-hydraulics test facilities with the experimental data. (authors)

Assessment of different mechanisms of C-14 production in irradiated graphite of RBMK-1500 reactors

International Nuclear Information System (INIS)

Narkunas, Ernestas; Smaizys, Arturas; Poskas, Povilas; Kilda, Raimondas

2010-01-01

Two RBMK-1500 water-cooled graphite-moderated channel-type power reactors at the Ignalina Nuclear Power Plant (INPP) are under decommissioning now. The total mass of irradiated graphite in the cores of both units is more than 3600 tons. The main source of uncertainty in the numerical assessment of graphite activity is the uncertainty of the initial impurities content in graphite. Nitrogen is one of the most important impurities, having a large neutron capture cross-section. This impurity may become the dominant source of C-14 production. RBMK reactors graphite stacks operate in the cooling mixture of helium-nitrogen gases and this may additionally increase the quantity of the nitrogen impurity. In this paper the results of the numerical modelling of graphite activation for the INPP Unit I reactor are presented. In order to evaluate the C-14 activity dependence on the nitrogen impurity content, several cases with different nitrogen content were modelled taking into account initial nitrogen impurity quantities in the graphite matrix and possible nitrogen quantities entrapped in the graphite pores from cooling gases. (orig.)

Numeric modeling of HfO2 neutron flux sensor parameters during sensor burnup in the RBMK-1500 reactor

International Nuclear Information System (INIS)

Jurkevicius, A.; Remeikis, V.

2001-01-01

The isotopic composition of hafnium in the radial neutron flux sensor of the RBMK-1500 reactor, the rates of the neutron absorption on Hf isotopes and the neutron spectrum in the sensor were numerically modeled. The sequence SAS2 (Shielding Analysis Sequence) from the package SCALE 4.3 was used for calculations. It has been obtained that the main neutron absorber 167 Er isotope practically burns up completely at the 18 MW d/kgU burnup depth, and at that time the capture rate of thermal neutrons in erbium decreases ten-fold. The average neutron flux density was calculated 7.6*10 13 neutrons. Cm -2 S -1 in the RBMK-1500 reactor grating, when the nuclear fuel enriched with 235 U by 2.4% and with Er by 0.4% is used in a fuel assembly. When the sensor burnup reaches 28 MW d/kgU, the neutron absorption rate of 178 Hf exceeds the rate of 177 Hf. The overall neutron absorption rate in hafnium decreases 2.53 times due to the sensor burnup to 56 MW d/kgU. The corrective factors ξ d (I) at different integral flux I of the sensors were calculated. The obtained dependence ξ d (I) calculated numerically was compared to the experimental one determined by processing repeated calibration results of Hf sensors in RBMK-1500 reactors, as well as compared to the theoretical one currently used in the Ignalina NPP special mathematical algorithms. (author)

Methods for estimating the reliability of the RBMK fuel assemblies and elements

International Nuclear Information System (INIS)

Klemin, A.I.; Sitkarev, A.G.

1985-01-01

Applied non-parametric methods for calculation of point and interval estimations for the basic nomenclature of reliability factors for the RBMK fuel assemblies and elements are described. As the fuel assembly and element reliability factors, the average lifetime is considered at a preset operating time up to unloading due to fuel burnout as well as the average lifetime at the reactor transient operation and at the steady-state fuel reloading mode of reactor operation. The formulae obtained are included into the special standardized engineering documentation

Russian RBMK reactor design information

International Nuclear Information System (INIS)

1993-11-01

This document concerns the systems, design, and operations of the graphite-moderated, boiling, water-cooled, channel-type (RBMK) reactors located in the former Soviet Union (FSU). The Russian Academy of Sciences Nuclear Safety Institute (NSI) in Moscow, Russia, researched specific technical questions that were formulated by the Pacific Northwest Laboratory (PNL) and provided detailed technical answers to those questions. The Russian response was prepared in English by NSI in a question-and-answer format. This report presents the results of that technical exchange in the context they were received from the NSI organization. Pacific Northwest Laboratory is generating this document to support the US Department of Energy (DOE) community in responding to requests from FSU states, which are seeking Western technological and financial assistance to improve the safety systems of the Russian-designed reactors. This report expands upon information that was previously available to the United States through bilateral information exchanges, international nuclear society meetings, International Atomic Energy Agency (IAEA) reactor safety programs, and Research and Development Institute of Power Engineering (RDIPE) reports. The response to the PNL questions have not been edited or reviewed for technical consistency or accuracy by PNL staff or other US organizations, but are provided for use by the DOE community in the form they were received

The safety of WWER and RBMK nuclear power plants. Progress report on the IAEA extrabudgetary programme on the safety of WWER and RBMK nuclear power plants, 1992-1994

International Nuclear Information System (INIS)

1994-11-01

This report, prepared by the IAEA secretariat, provides an overview of the IAEA Extrabudgetary Programme activities and results from 1992 until June 1994. The report describes the scope, the current status of the implementation and major findings and recommendations of the Programme. Though this report concentrates on the results of the Extrabudgetary Programme, it also refers to the significant related activities carried out under IAEA Technical Co-operation projects, as well as other international activities relevant to the safety of WWER and RBMK reactors. 13 figs, 9 tabs

Strategy for Handling and Treatment of INPP RBMK-1500 Irradiated Graphite

International Nuclear Information System (INIS)

Oryšaka, A.

2016-01-01

There are two RBMK-1500 water-cooled graphite-moderated channel-type power reactors at Ignalina NPP. After the final shutdown of the INPP, radioactive i-graphite dismantling, handling, conditioning, storage and disposal is an important part of the decommissioning activities. The core of the INPP unit 1 and 2 contains about 3600 tons of i-graphite. Formation of activation products strongly depends on the contents of impurities, operational mode and concentration of impurities in the graphite. The case study for INPP envisages the analysis of possibilities of graphite handling and treatment in the context of immediate decommissioning. (author)

Qualitative and quantitative characteristics of fission products in spent nuclear fuel from RBMK-type reactor

International Nuclear Information System (INIS)

Adlys, G.; Adliene, D.

2002-01-01

Well-known empirical models or experimental instruments and methods for the estimation of fission product yields do not allow prediction of the behavior and evaluation of the time-dependent qualitative and quantitative characteristics of all fission products in spent nuclear fuel during long-term storage. Several computer codes were developed in different countries to solve this problem. French codes APOLLO1 and PEPIN were used in this work for modeling the characteristics of spent nuclear fuel in the RBMK reactor. The modeling results of qualitative and quantitative characteristics of long-lived fission products for different cooling periods of spent nuclear fuel, including 50-year cooling period, are presented in this paper. The 50-year cooling period conforms to the foreseen time of storage of spent nuclear fuel in CONSTOR and CASTOR casks at the Ignalina NPP. These results correlate well with evaluated quantities for the well-known yields of the nuclides and could be used for the compilation of the database for long-lived fission products in spent nuclear fuel from the RBMK-type reactor. They allow one to predict and to solve effectively safety problems concerning with long-term spent nuclear fuel storage in casks. (author)

IAEA/USDOE senior management workshop on promotion of safety culture for the NPPS with RBMK reactors. Working material

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-09-01

The current workshop, co-sponsored by the IAEA and USDOE, was a continuation of the previous effort for further promotion of safety culture at RBMK NPPs. The objective of the workshop was to provide a forum for senior managers from governmental organizations and operating organizations to further exchange experience in understanding the factors influencing safety culture, in assessing safety culture at their own organizations and developing safety culture at RBMK NPPs. The workshop consisted of a broad scope of presentations to review the basic concepts and major elements of safety culture (ownership, accountability, pride, job satisfaction, trust, openness, etc.), to identify and discuss the various approaches used in different countries in attaining a strong safety culture, and to explain, through the use of practical examples, what the benefits of a strong safety culture are; how to improve the behavior of people, how to gain trust and openness, how to overcome difficulties in changing staff`s attitudes, and how to manage safety culture. 2 figs.

IAEA/USDOE senior management workshop on promotion of safety culture for the NPPS with RBMK reactors. Working material

International Nuclear Information System (INIS)

1997-01-01

The current workshop, co-sponsored by the IAEA and USDOE, was a continuation of the previous effort for further promotion of safety culture at RBMK NPPs. The objective of the workshop was to provide a forum for senior managers from governmental organizations and operating organizations to further exchange experience in understanding the factors influencing safety culture, in assessing safety culture at their own organizations and developing safety culture at RBMK NPPs. The workshop consisted of a broad scope of presentations to review the basic concepts and major elements of safety culture (ownership, accountability, pride, job satisfaction, trust, openness, etc.), to identify and discuss the various approaches used in different countries in attaining a strong safety culture, and to explain, through the use of practical examples, what the benefits of a strong safety culture are; how to improve the behavior of people, how to gain trust and openness, how to overcome difficulties in changing staff's attitudes, and how to manage safety culture. 2 figs

Screening of external hazards for NPP with bank type reactor. Modeling of safety related systems and equipment for RBMK. Probabilistic assessment of NPP safety on aircraft impact. Progress report

International Nuclear Information System (INIS)

Kostarev, V.

1999-01-01

This progress report was produced within the frame of IAEA research project on screening the hazards for NPP with bank type reactor. It covers the following tasks; development of the model for the primary loop system of RBMK; developing the models for safety related equipment of RBMK; developing of models for safety related models of EGP-6 type reactor (Bilibinskaya Nuclear Co-generated heat and Power Plant); and probabilistic assessment of NPP safety on aircraft impact

RBMK nuclear reactors: Proposals for instrumentation and control improvements to enhanced safety and availability. IEC technical report of type 3. Working material

International Nuclear Information System (INIS)

1995-01-01

The present material presents a CD+V draft report ''RBMK nuclear reactors: Proposals for instrumentation and control improvements to enhance safety and availability'' prepared by the Joint IEC/IAEA team during 1993-1995. Experience has demonstrated the need to improve the safety instrumentation of the RBMK type reactors using well proven modern technology. The working group identified the upgrades and changes of the highest priority based on the evaluation of the RBMK systems and the events where the instrumentation was found to be inadequate for safe operation. The subjects discussed in this document were not selected on a systematic basis but were selected by the IEC and IAEA experts as considered to be appropriate to the activities of the IEC and for which technical experience was available. The items identified therefore do not reflect any ranking of the safety issues or any priority or impact on safety of any of the measures were they to be implemented. Many important safety issued and areas where physical measures are required to improve safety have been omitted and indeed not even acknowledged in this document. The recommendations presented in the document differ from those normally produced by the IEC in the form of standards as they are of a transitory nature and some have already been overtaken by the continuing process of improvements to plant safety. Figs and tabs

Effect of eccentric location of the RBMK CPS displacer graphite block in the shielding sheath

International Nuclear Information System (INIS)

Dostov, A.I.

2001-01-01

Temperature conditions and accumulation of Wigner energy in the graphite block of the RBMK reactor CPS (control power system) displacer is examined. It is shown, that at eccentric location of the block in the shielding sheath average temperature of the block drops sharply. Due to the design demerit quantity of the stored energy in the block may be so great, that its release will result in melting of the displacer tube. (author)

ABB engagement in efforts to improve the safety of RBMK reactors

International Nuclear Information System (INIS)

Tiren, L.I.; Bioere, S.; Molin, J.

1993-01-01

ABB Atom is engaged in safety analysis for the Ignalinsk (RBMK) nuclear power plant. The analysis is done within the framework of two different initiatives of the Swedish Nuclear Power Inspectorate, namely: probabilistic safety assessment, i.e. the BARSELINA project, and analysis of containment safety issues. The aim is to enable decisions to be made for specific hardware modifications. The following items were considered by the Swedish Nuclear Power Inspectorate to be the most significant with regard to safety and were thus selected for further study or action: nondestructive testing of primary system components, fire and flooding protection, pressure relief from the reactor cavity in certain accident sequences, Accident Localization System improvements, and a separate auxiliary feedwater system. (Z.S.) 1 fig

United States Department of Energy's reactor core protection evaluation methodology for fires at RBMK and VVER nuclear power plants. Revision 1

International Nuclear Information System (INIS)

1997-06-01

This document provides operators of Soviet-designed RBMK (graphite moderated light water boiling water reactor) and VVER (pressurized light water reactor) nuclear power plants with a systematic Methodology to qualitatively evaluate plant response to fires and to identify remedies to protect the reactor core from fire-initiated damage

Analysis of fuel pin mechanics in case of flow blockage of a single RBMK channel

International Nuclear Information System (INIS)

Pierro, F.; Moretti, F.; Mazzini, D.; D'Auria, F.

2005-01-01

The evaluation of the consequences of the pressure tube rupture due to accidental overheating is one of the key elements for addressing an RBMK safety analysis, since it causes the lost of design boundaries against the fission products release. Several events are expected to take place: thermal hydraulic crisis (energy unbalance), fuel overheating, fuel rod damage, pressure tube overheating, pressure tube failure and graphite stack damage, Hydrogen and fission products release. The present work deals with the research activity carried out at ''Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione'' (DIMNP) of the University of Pisa aimed at assessing numerical models for safety analysis of the RBMK-1000. The attention is focused on the modelling of (1) a single fuel channel and its surrounding graphite column for evaluating the transient conditions enabling the different damaging phenomena, (2) a single fuel rod for investigating fuel pin behaviour, (3) the ruptured fuel channel for figuring the magnitude of the hydrodynamic loads acting on fuel rods. Different codes were employed to cover the competences for the investigation of each field; in particular, RELAP5 code for thermal-hydraulics, FRAPCON-3 and FRAPTRAN1-2 codes for fuel pin mechanics, FLUENT-6 for fluid dynamics. The paper discusses the numerical models, the analysis capabilities of numerical models in comparison with available data about the Leningrad NPP 1992 accident. Furthermore, the possibility to draw a failure map identifying the range of the cladding safety respect to the transient condition is outlined. (author)

Damage mechanism of piping welded joints made from austenitic Steel for the type RBMK reactor

International Nuclear Information System (INIS)

Karzov, G.; Timofeev, B.; Gorbakony, A.; Petrov, V.; Chernaenko, T.

1999-01-01

In the process of operation of RBMK reactors the damages were taking place on welded piping, produced from austenitic stainless steel of the type 08X18H10T. The inspection of damaged sections in piping has shown that in most cases crack-like defects are of corrosion and mechanical character. The paper considers in details the reasons of damages appearance and their development for this type of welded joints of downcomers 325xl6 mm, which were fabricated from austenitic stainless steel using TlG and MAW welding methods. (author)

State of the Art of the Ignalina RBMK-1500 Safety

International Nuclear Information System (INIS)

Uspuras, E.

2010-01-01

Ignalina NPP is the only nuclear power plant in Lithuania consisting of two units, commissioned in 1983 and 1987. Unit 1 of Ignalina NPP was shut down for decommissioning at the end of 2004 and Unit 2 is to be operated until the end of 2009. Both units are equipped with channel-type graphite-moderated boiling water reactors RBMK-1500. The paper summarizing the results of deterministic and probabilistic analyses is developed within 1991 to 2007 by specialists from Lithuanian Energy Institute. The main operational safety aspects, including analyses performed according the Ignalina Safety Improvement Programs, development and installation of the Second Shutdown System and Guidelines on Severe Accidents Management are discussed. Also the phenomena related to the closure of the gap between fuel channel and graphite bricks, multiple fuel channel tube rupture, and containment issues as well as implication of the external events to the Ignalina NPP safety are discussed separately.

Extending the features of RBMK refuelling machine simulator with a training tool based on virtual reality

International Nuclear Information System (INIS)

Khoudiakov, M.; Slonimsky, V.; Mitrofanov, S.

2004-01-01

The paper describes a continuation of efforts of an international Russian - Norwegian joint team to improve operational safety during the refuelling process of an RBMK-type reactor by implementing a training simulator based on an innovative Virtual Reality (VR) approach. During the preceding 1st stage of the project a display-based simulator was extended with VR models of the real Refuelling Machine (RM) and its environment in order to improve both the learning process and operation's effectiveness. The simulator's challenge is to support the performance (operational activity) of RM operational staff firstly by helping them to develop basic knowledge and skills as well as to keep skilled staff in close touch with the complex machinery of the Refuelling Machine. During the 2nd stage of the joint project the functional scope of the VR-simulator was greatly enhanced - firstly, by connecting to the RBMK-unit full-scope simulator, and, secondly, by including a training program and simulator model upgrade. The present 3rd stage of the Project is primarily oriented towards the improvement of the training process for maintenance and operational personnel by means of a development of the Training Support Methodology and Courses (TSMC) to be based on Virtual Reality and enlarged functionality of 3D and process modelling. The TMSC development is based on Russian and International Regulatory Bodies requirements and recommendations. Design, development and creation of a specialised VR-based Training System for RM Maintenance Personnel are very important for the Russian RBMK plants. The main goal is to create a powerful, autonomous VR-based simulator for training technical maintenance personnel on the Refuelling Machine. VR based training is expected to improve the effect of training compared to the current training based on traditional methods using printed documentation. The LNPP management and the Regulatory Bodies supported this goal. The VR-based Training System should

RBMK fuel channel blockage analysis by MCNP5, DRAGON and RELAP5-3D codes

International Nuclear Information System (INIS)

Parisi, C.; D'Auria, F.

2007-01-01

The aim of this work was to perform precise criticality analyses by Monte-Carlo code MCNP5 for a Fuel Channel (FC) flow blockage accident, considering as calculation domain a single FC and a 3x3 lattice of RBMK cells. Boundary conditions for MCNP5 input were derived by a previous transient calculation by state-of-the-art codes HELIOS/RELAP5-3D. In a preliminary phase, suitable MCNP5 models of a single cell and of a small lattice of RBMK cells were set-up; criticality analyses were performed at reference conditions for 2.0% and 2.4% enriched fuel. These analyses were compared with results obtained by University of Pisa (UNIPI) using deterministic transport code DRAGON and with results obtained by NIKIET Institute using MCNP4C. Then, the changes of the main physical parameters (e.g. fuel and water/steam temperature, water density, graphite temperature) at different time intervals of the FC blockage transient were evaluated by a RELAP5-3D calculation. This information was used to set up further MCNP5 inputs. Criticality analyses were performed for different systems (single channel and lattice) at those transient' states, obtaining global criticality versus transient time. Finally the weight of each parameter's change (fuel overheating and channel voiding) on global criticality was assessed. The results showed that reactivity of a blocked FC is always negative; nevertheless, when considering the effect of neighboring channels, the global reactivity trend reverts, becoming slightly positive or not changing at all, depending in inverse relation to the fuel enrichment. (author)

Analysis of water hammer phenomena in RBMK-1500 reactor main circulation circuit

International Nuclear Information System (INIS)

Kaliatka, A.; Uspuras, E.; Vaisnoras, M.

2006-01-01

Water hammer can occur in any thermal-hydraulic systems. Water hammer can reach pressure levels far exceeding the pressure range of a pipe given by the manufacturer, and it can lead to the failure of the pipeline integrity. In the past three decades, since a large number of water hammer events occurred in the light-water- reactor power plants, a number of comprehensive studies on the phenomena associated with water hammer events have been performed. There are three basic types of severe water hammer occurring at power plants that can result in significant plant damage: rapid valve operation events; void-induced water hammer; condensation-induced water hammer. Correct prediction of water hammer transients, is therefore of paramount importance for the safe operation of the plant. Therefore verifying of computer codes capability to simulate water hammer type transients is very important issue at performing of safety analyses for nuclear power plants. Verification of RELAP5/MOD3.3 code capability to simulate water hammer type transients employing the experimental investigations is presented. Experience gained from benchmarking analyses has been used at development of the detail RELAP5 code RBMK-1500 model for simulation of water hammer effects in reactor main circulation circuit. Analysis of reactor cooling system shows, that water hammers can occur in main circulation circuit of RBMK-1500 reactor in cases of: (1) Guillotine break of the inlet piping upstream of the Group Distribution Header and (2) Guillotine break of the pressure piping upstream the Main Circulation Pump check valve. Analysis of above mentioned accident scenarios is presented in this paper. First scenario of the accident potentially is more dangerous, because the pressure pulses influence not only the reactor cooling circuit, but also the piping of safety related system (Emergency Core Cooling System pipeline) connected to affected Group Distribution Header. The performed analysis using RELAP5 code

Aircraft impact qualification of RBMK systems and components. Technical report. Rev. 00, May 1999

International Nuclear Information System (INIS)

1999-01-01

In the present report, the problem of qualification procedures of electrical equipment with respect to the dynamic excitation subsequent to an aircraft impact (ACC) on a Nuclear Power Plant (NPP) is approached, within the context of IAEA Benchmark on vulnerability of equipment and structures of RBMK-type NPP against the aircraft impact. After a short description of the main objectives of the work and the relevant area of concern (Chapter 1), the safety related equipment more commonly installed in a NPP are grouped in few classes, according to widely accepted classification criteria and the relevant failure modes are described (Chapter 2). Taking as reference a deeply studied RBMK reactor (Ignalina NPP), an overview of its main characteristics and of the equipment ensemble housed in is given in Chapter 3. An overview of the worldwide most used qualification standards for safety related equipment for NPPs is reported in Chapter 4, and a comparison of the practices used in Europe for the qualification of safety related electrical and I and C equipment is described with special attention to seismic and impact qualification (Chapter 5). In the hypothesis that the equipment to qualify against impact excitation has been already qualified against seismic excitation, the problems relevant to the different nature of earthquake and shock phenomena are listed, together with the main criteria to implement a procedure which, based on standardized shock pulses, could be applied for ACC qualification purposes (Chapter 6). Consequently, a possible ACC qualification procedure is outlined (Chapter 7) and the interface data (data coming from numerical analysis and seismic qualification, to be used for ACC qualification purposes) are listed (Chapter 8). Finally, the main conclusions of the work are described (Chapter 9). The main references are listed in Chapter 10. (author)

Final report of the programme on the safety of WWER and RBMK nuclear power plants. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

International Nuclear Information System (INIS)

1999-05-01

The review of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants focuses on the wide scope of the activities aimed at identifying safety deficiencies, ranking their importance on the results of safety improvement programmes and on areas where future work is necessary. The information in the report reflects to a large extent, the situation as it stood when individual IAEA tasks actually took place. It deals with the IAEA activities and it discusses selected safety issues and safety review results as they apply to each reactor type. The results, recommendations and conclusions resulting from the IAEA Programme are intended to assist national decision makers who have the sole responsibilities for the regulation and safety operation of their nuclear power plants

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

International Nuclear Information System (INIS)

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

2012-01-01

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

Plant level of automated control system at a NPP with RBMK reactor

International Nuclear Information System (INIS)

Vorob'ev, V.P.; Gorbunov, V.P.; Dmitriev, V.M.; Litvin, A.S.

1987-01-01

The functional structure of plant level automated control system (ACS) at NPP with RBMK-1000 reactors, its binding with the on-line control system of higher and lower levels, as well as engineering requirements to software and recommendations on composition of hardware components, are considered. NPP ACS is an organizational-engineering system consisting of computer facilities and binding aimed at solving management, economical, organizational and physical-engineering problems to control NPP more effectively. The system carries out data acquisition, preliminary processing, analysis, transmission and representation for users to accept solutions for NPP operation by operative and management personnel. The main aim of integrated NPP ACS is the control development and increase of NPP economical efficiency, the increase of electric and heat energy production, the optimization of the production distribution between units, the development of production and economic NPP control

Burnup credit calculations for criticality safety justification for RBMK-1000 spent fuel of transport and storage systems

Directory of Open Access Journals (Sweden)

V. V. Galchenko

2010-12-01

Full Text Available In present paper the burnup credit calculations for TK-8 transport container and SVJP-1 spent fuel storage fa-cility of pool type with RBMK-1000 spent fuel during 100-years of cooling time were performed for criticality safety analysis purpose using MCNP and SCALE codes. Only actinides were taken into account for these critical systems. Two approaches were analyzed with isotopes distribution calculations along fuel assembly height and without it. The results show that subcriticality margin is increased considerably using burnup credit and isotopes distribution along fuel assembly height made this value more reasonable.

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

International Nuclear Information System (INIS)

Petri, M.C.; Binder, J.L.; Pasedag, W.F.

2001-01-01

Throughout the years 1990 the U.S. Department of Energy has worked build capability in countries of the former Soviet Union to assess the safety of their VVER and RBMK reactors. Through this Plant Safety Evaluation Program, deterministic and probabilistic analyses have been used to provide a documented plant risk profile to support safe plant operation and to set priorities for safety upgrades. Work has been sponsored at thirteen nuclear power plant sites in eight countries. The Plant Safety Evaluation Program has resulted in immediate and long-term safety benefits for the Soviet-designed nuclear plants. (author)

Determining the residual lifetime for the main RBMK components in frame of life extension of power units

International Nuclear Information System (INIS)

Baldin, V.D.; Petrov, A.A.; Potapov, A.A.

2005-01-01

At present time 11 power units of NPPs with RBMK reactors produce annually 140 billion kWh of electricity, which is about half of all electricity generated at Russian NPPs. Commissioned during the period of 1973 to 1990, they have the specified service life of 30 years. The power units have been regularly upgraded during operation to raise their safety and reliability. As the result now they meet modem requirements in terms of safety and their lifetime can considerably exceed their specified service life. The task of the plant life extension becomes more urgent. The main nonreplaceable RBMK reactor components that determine the reactor life, are graphite stack and metal support structures. The aging of these structures is monitored and the structures residual lifetime is predicted in accordance with procedures approved by regulatory authorities. The lead unit among all RBMK power units is Leningrad NPP Unit 1 commissioned in 1973. A complex of activities was performed for this power unit in 2002-2003 to prove the feasibility of its life extension. In terms of defining the residual lifetime of the reactor components, the activities were performed in three major areas: - examination of the state of the reactor structures; - materials studies of irradiated metal and graphite specimens in hot chambers; - performing the calculation and prediction evaluations. The results of examining the metal structures and studying specimens have shown that: 1)The metal and the welds do not have visible defects. The state of protective coatings is largely satisfactory. 2)The ultimate strength and hardness of steel have grown. 3)The brittle fracture transition temperature shift is at the allowable level. 4)The leak-tightness of the metal structures has not been violated. The results of the graphite stack examination have shown that the graphite is at the hardening stage. No tendency for deterioration of mechanical properties has been recorded. According to the results of measuring the

Heat transfer in the core graphite structures of RBMK nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Knoglinger, E., E-mail: ernst.knoglinger@a1.net [Am Winklerwald 15, A 4020 Linz (Austria); Wölfl, H., E-mail: herbert.woelfl@tele2.at [Berg, Im Weideland 19, A 4060 Linz (Austria); Kaliatka, A., E-mail: algirdas.kaliatka@lei.lt [Laboratory of Nuclear Installation Safety, Lithuanian Energy Institute, Breslaujos 3, LT-44403 Kaunas (Lithuania)

2015-11-15

Highlights: • Proposed solution of heat transfer model from a hollow cylinder to a fluid through narrow duct. • Thermal conductance of annular gaps, filled by two component gas was discussed. • Xenon transient preceding the Chernobyl Accident was analyzed. • Reactivity balance during power manoeuvres and potenrial causes of the accident were discussed. - Abstract: Conductive and combined radiative/conductive gap conductance models are presented and discussed in great detail. The heat resistance concept and an exact solution to the one dimensional heat conduction equation for a 3-region composite hollow cylinder are used to calculate gap conductance in function of gap gas composition and fuel burn up. The study includes the back calculation of a reactor experiment performed at the Ignalina NPP Unit-1 which provides some insight in the function of the RBMK nitrogen supply and regulating device and an investigation of the role the graphite temperature played during the power manoeuvres preceding the Chernobyl Accident.

Behaviour of the RBMK-1000 plant during reactivity disturbances under part load reduction - completing investigations

International Nuclear Information System (INIS)

Clemente, M.; Langenbuch, S.

1989-01-01

This report describes investigations of the behavior of a RBMK-1000 reactor core during reactivity initiated accidents and completes earlier studies of the Chernobyl accident. Special questions related to this accident are studied, e.g. the effect of Xenon dynamics during the delayed load reduction and the coarse of the experiment as planned with the coast-down of four main recirculaton pumps at nominal part load conditions. The main interest is the detailed analysis of reactivity initiated accidents in the low power range till 25% during start-up. In the calculations no reactor trip is taken into account. The results confirm the unfavourable effects of the positive void coefficient, which are amplified in the low power range. Finally the results are discussed in comparison to other positive reactivity effects. (orig.) [de

Automatic optimization of constants and special mathematic ensuring algorithms SKALA-micro system of RBMK-1000 reactor self-certification in operation

International Nuclear Information System (INIS)

Aleksandrov, S.I.; Dmitrenko, V.V.; Postnikov, V.V.; Sviridenkov, A.N.; Yurkin, G.V.; Yakunin, I.S.

2007-01-01

Paper dwells upon problems dealing with accuracy improvement of the energy release distribution and the safety margin of the RBMK-1000 operation. The accuracy is improved through the automatic optimization of some constants used in the SKALA-micro system special mathematic ensuring program and the regular self-validation of the algorithm to determine the energy release distribution calculation error. The validation based on the regular scanning of the reactor core by a calibrating detector and through the sequence disabling of the internal detectors is shown to give the close results [ru

Calculation and experimental study of the RBMK-1500 reactor emergency cooling at maximum designed accident

International Nuclear Information System (INIS)

Cherkashov, Yu.M.; Vasilevskij, V.P.; Labazov, V.H.; Loninov, A.Ya.; Molochnikov, Yu.S.; Novosel'skij, O.Yu.; Podlazov, L.N.; Pavlov, V.B.; Pushkarev, V.I.

1981-01-01

The analysis of thermohydraulic and neutron-physical processes occurring in the RBMK-1500 reactor during the reactor emergency cooling system triggering (RECS) after the maximum designed accident (MDA) is conducted. The MDA means hypothetical instant hilliotine break of the main circulating pump head collector. During the whole cooling down period the RECS should provide the temperature level of the fuel elements not exceeding 1200 deg C and the channel pipe temperature - 600 deg C. The principal flowsheet of the balloon type RECS is described. Calculations of the valve fast response effect on the RECS productivity are carried out. It is concluded that the chosen balloon RECS provides reliable temperature modes of fuel elements naand channel pipes under the MDA conditions. At the same time a momentary splash of neutron power by the value not more than 10% can take place [ru

Ways of decreasing the labour content and construction duration for the RBMK-1000 NPP

International Nuclear Information System (INIS)

Chernyshenko, V.M.

1984-01-01

Problems associated with reducing the labour content and duration of construction for the RBMK-1000 NPPs are considered. General and specific labour contents for construction of the first units at the Kursk and Chernobylsk, NPPs as well as progress chart for construction-installation work at the 1-st unit of the Smolensk and 3-d unit of the Chernobylsk NPPs are presented. The analysis has shown that reduction in the general labour costs and therefore, duration of construction can be attained by reducing the number of auxiliary objects, increasing the level and mechanization of construction (with optimum utilization of gantry and turret cranes) as well as by mechanization of placing concrete mortars and use of large-block structures. According to preliminary calculations, the introduction of new solutions would ensure reduction of construction periods to 24 months for the second units and reduction of labour content by 8 to 10% at the Kursk and Chernobylsk NPPs

Prestart-up hydrogen peroxide solution washing of NPP unit with the RBMK-type reactor

International Nuclear Information System (INIS)

Gruzdev, N.I.; Man'kina, N.N.; Al'tshuller, M.A.

1979-01-01

Presented are the results of industrial hydrogen peroxide solution washing of condensating-feed system conducted on the second unit of the Kursk NPP. Duration of the washing constituted 8 hours. The hydrogen peroxide concentration during first 4 hours was 10-20 mg/kg at a flow rate of 260 m 2 /h, during the following 4 hours it constituted 2-5 mg/kg at a flow rate of 1000 m 3 /h. It is found out that prestart-up hydrogen peroxide washing of NPP power units with the RBMK-type reactor permits: to simplify essentially the technology and scheme of washing process; to reduce a flow rate of desalt washing water; to except environmental contamination with washing solutions and reagents being neutralized; to reduce the time of washing process; to reduce the time necessary for the achievement of reference water condition factors, and to increase the unit reliability and to improve a radiation situation

Using the coolant temperature noise for measuring the flow rate in the RBMK technological channels

International Nuclear Information System (INIS)

Selivanov, V.M.; Karlov, N.P.; Martynov, A.D.; Prostyakov, V.V.; Lysikov, B.V.; Kuznetsov, B.A.; Pallagi, D.; Khorani, Sh.; Khargitai, T.; Tezher, Sh.

1983-01-01

The problems are considered connected with the possibility of using thermometric correlation method to measure the coolant flow rate in the RBMK reactor technological channels. The main attention is paid to the study of the physical nature of the coolant temperature pulsations and to estimation of the effect of parameters of the primary thermaelectrical converter (TEC) on the results of measurements. In the process of reactor inspections made using the thermometric correlation flowmeter of a special design, the temperature noise distribution in the points of flow rate measurement is studied, the noise intensity and physical nature are determined, as well as the effect of different TEC parameters (TEC inertia and base distance between them) on the measurement accuracy. On the basis of the analysis of the effect on the results of the TEC thermal inertia measured value divergence, tausub(α) and transport time, tau sub(T), a conclusion is made on the necessity of choosing the base distance between TEC with tausub(T)>tausub(d)

Dynamic model for tritium build-up at NPP with RBMK type reactors and its enviromental beraviour

International Nuclear Information System (INIS)

Badyaev, V.V.; Egorov, Yu.A.; Ivanov, E.A.; Stegachev, G.F.; Tolstykh, V.D.

1982-01-01

A model of tritium production dynamics for a high power NPP with RBMK type reactors is proposed and investigated. The main ''skeleton'' model structure for forecasting tritium buildup at a NPP and its exchange with the environment has been singled out at a heuristic level. Decomposition and layout of the units have been performed by global functional relations of the investigated objects (NPP and environment). the model accounts for only oxidized tritium forms. Water exchange between the NPP subsystems and environment is the main mechanism for tritium migration. The model does not account for scheduled periodic maintenance work effects, presence of stagnant zones in the station circuits, fuel burn-up, etc. The parametric identification method applied in the model makes the model adaptable to particular situations and considered systems of the NPP and environment. Completing the model with necessary and sufficient experimental data one can pass to certain forecasting problems and to NPP control as a tritium source in the environment

Some problems of software development for the plant-level automated control system of NPPs with the RBMK reactors

International Nuclear Information System (INIS)

Gorbunov, V.P.; Egorov, A.K.; Isaev, N.V.; Saprykin, E.M.

1987-01-01

Problems on development and operation of automated control system (ACS) software of NPPs with the RBMK reactors are discussed. The ES computer with large on-line storage (not less than 1 Mbite) and fast response (not less than 300.000 of operations per a second) should enter the ACS composition. Several program complexes are used in the NPP ACS. The programs collected into the EhNERGIYa library are used to provide central control system operation. The information-retrival system called the Fuel file is used to automate NPP fuel motion account, as well as to estimate efficiency of fuel application, to carry out calculations of a fuel component of electric and heat energy production cost. The automated information system for unit operation efficiency analysis, which solves both plant and unit-level problems, including engineering and economical factors and complexing of operation parameter bank, is under trial operation

Detection and localization of leak of pipelines of RBMK reactor. Methods of processing of acoustic noise

International Nuclear Information System (INIS)

Tcherkaschov, Y.M.; Strelkov, B.P.; Chimanski, S.B.; Lebedev, V.I.; Belyanin, L.A.

1997-01-01

For realization of leak detection of input pipelines and output pipelines of RBMK reactor the method, based on detection and control of acoustic leak signals, was designed. In this report the review of methods of processing and analysis of acoustic noise is submitted. These methods were included in the software of the leak detection system and are used for the decision of the following problems: leak detection by method of sound pressure level in conditions of powerful background noise and strong attenuation of a signal; detection of a small leak in early stage by high-sensitivity correlation method; determination of a point of a sound source in conditions of strong reflection of a signal by a correlation method and sound pressure method; evaluation of leak size by the analysis of a sound level and point of a sound source. The work of considered techniques is illustrated on an example of test results of a fragment of the leak detection system. This test was executed on a Leningrad NPP, operated at power levels of 460, 700, 890 and 1000 MWe. 16 figs

Calculation of the real states of Ignalina NPP Unit 1 and Unit 2 RBMK-1500 reactors in the verification process of QUABOX/CUBBOX code

International Nuclear Information System (INIS)

Bubelis, E.; Pabarcius, R.; Demcenko, M.

2001-01-01

Calculations of the main neutron-physical characteristics of RBMK-1500 reactors of Ignalina NPP Unit 1 and Unit 2 were performed, taking real reactor core states as the basis for these calculations. Comparison of the calculation results, obtained using QUABOX/CUBBOX code, with experimental data and the calculation results, obtained using STEPAN code, showed that all the main neutron-physical characteristics of the reactors of Unit 1 and Unit 2 of Ignalina NPP are in the safe deviation range of die analyzed parameters, and that reactors of Ignalina NPP, during the process of the reactor core composition change, are operated in a safe and stable manner. (author)

Joint US/Russian study on the development of a decommissioning strategy plan for RBMK-1000 unit No. 1 at the Leningrad Nuclear Power Plant

International Nuclear Information System (INIS)

1997-12-01

The objective of this joint U.S./Russian study was to develop a safe, technically feasible, economically acceptable strategy for decommissioning Leningrad Nuclear Power Plant (LNPP) Unit No. 1 as a representative first-generation RBMK-1000 reactor. The ultimate goal in developing the decommissioning strategy was to select the most suitable decommissioning alternative and end state, taking into account the socioeconomic conditions, the regulatory environment, and decommissioning experience in Russia. This study was performed by a group of Russian and American experts led by Kurchatov Institute for the Russian efforts and by the Pacific Northwest National Laboratory for the U.S. efforts and for the overall project

Dynamic reliability and risk assessment of the accident localization system of the Ignalina NPP RBMK-1500 reactor

International Nuclear Information System (INIS)

Kopustinskas, V.; Augutis, J.; Rimkevicius, S.

2005-01-01

The paper presents reliability and risk analysis of the RBMK-1500 reactor accident localization system (ALS) (confinement), which prevents radioactive releases to the environment. Reliability of the system was estimated and compared by two methods: the conventional fault tree method and an innovative dynamic reliability model, based on stochastic differential equations. Frequency of radioactive release through ALS was also estimated. The results of the study indicate that conventional fault tree modeling techniques in this case apply high degree of conservatism in the system reliability estimates. One of the purposes of the ALS reliability study was to demonstrate advantages of the dynamic reliability analysis against the conventional fault/event tree methods. The Markovian framework to deal with dynamic aspects of system behavior is presented. Although not analyzed in detail, the framework is also capable of accounting for non-constant component failure rates. Computational methods are proposed to solve stochastic differential equations, including analytical solution, which is possible only for relatively small and simple systems. Other numerical methods, like Monte Carlo and numerical schemes of differential equations are analyzed and compared. The study is finalized with concluding remarks regarding both the studied system reliability and computational methods used

On using continuoas Markov processes for unit service life evaluation taking as an example the RBMK-1000 gate-regulating valve

International Nuclear Information System (INIS)

Klemin, A.I.; Emel'yanov, V.S.; Rabchun, A.V.

1984-01-01

A technique is sugfested for estimating service life indices of equipment based on describing the process of the equipment ageing by continuous Markov diffusion process. It is noted that a number of problems on estimating durability indices of products is reduced to problems of estimating characteristics of the time of the first attainment of the preset boundary (boundaries) by a random process describing the ageing of a product. The methods of statistic estimation of the drift and diffusion coefficient in the continuous Markov diffusion process are considered formulae for their point and interval estimates are presented. A special description is given for a case of a stationary process and determining in this case mathematical expectation and dispersion of the time of the first attainment of a boundary (boundaries). The method of numerical simulation of the diffusion process with constant drift and diffusion coefficients is also described; results obtained on the basis of such a simulation are discussed. An example of using the suggested technique for quantitative estimate of the service life for the RBMK-1000 gate-regulating value is given

Rearrangement of fuel assemblies in the RBMK type reactors to flatten power distribution and improve the fuel cycle

International Nuclear Information System (INIS)

Mityaev, Yu.I.; Vikulov, V.K.

1982-01-01

A possibility of increasing the burnup of uranium fuel unloaded from the RBMK type reactors is investigated. Three variants of a two-zone reactor-refueling are considered: 1. the simplest variant of continuous refueling used at present, when the central and peripherical reactor zones are additionally fueled independently by similar fuel assemblies (FA); 2. the variant under which new FA are loaded to the peripherical zone and are used there up to the same burnup as in the first case, then all the peripherical FA (PFA) are rearranged to the centre and they are used there up to maximum burnup; 3. the same as in the second variant, but not all the PFA are rearranged to the centre but only FA with small fuel burnup. It is shown by calculation that average fuel burnup for the third refueling variant is several per cent higher at the optimal burnup of rearranged FA. Besides, flattening of fuel channel power is improved in this case, that permits to increase uranium enrichment and burnup at the same maximum power. It essentially improves economic parameters of the reactor. It is concluded that realization of the considered variant of fuel refueling will produce the most essential effect for reactors refueled without shutdown

RBMK power unit operational performance investigation while false coming into action of the emergency reactor cooling system

International Nuclear Information System (INIS)

Emel'yanov, I.Ya.; Aleksakov, A.N.; Vasilevskij, V.P.; Labazov, V.N.; Nikolaev, E.V.; Podlazov, L.N.; Rogov, V.D.; Shevchenko, V.V.

1984-01-01

Regimes of RBMK reactor operation during false coming into action of the emergency reactor cooling system (ERCS), which might occur in the case of faults in the automation systems or erroneous actions of operator have been investigated. At that, thepe exists a probability of water supply from ERCS to one half of the reactor, which results in a sharp change of boiling regime, and due to void reactivity effect it causes the neutron field disturbance. Change in flow rate and enthalpy of coolant, as well as changes of neutron flux in the left and right halves of the reactor at ERCS response and during operation of the whole system of automatic control of power and system of local automatic control of power - local emergency protection - have been studied. The investigations have been carried out for different values of vapour effect of void reactivity effect and for time ranges from 0 to 40 s. The calculations are made using a model, describing spatial dynamics of reactor in two-dimensional approximation with 54 nodes. The model describes neutron-physics and thermohydraulic processes and it is realized using the BEhSM-6 computer. It is pointed out that one system of automatic control of power or local emergency protection (i.e. without the shut-off system), is insufficient for the compensation of disturbances appearing as a result of false ERCS coming into operation

Progress in the U.S. department of energy sponsored in-depth safety assessments of VVER and RBMK reactors

International Nuclear Information System (INIS)

Binder, J.L.; Petri, M.C.; Pasedag, W.F.

2001-01-01

Since the disastrous accident at Chernobyl Nuclear Power Plant Unit 4 in 1986, there has been international recognition of the safety concerns posed by the operation of 67 Soviet-designed commercial nuclear reactors. These reactors are operated in eight countries from the former Soviet Union and its former satellite states in Central and Eastern Europe. The majority of these plants are in the Russian Federation (30 units) and Ukraine (14 units). New plants are in various stages of construction. U.S. support to improve the safety of Soviet-designed reactors over the past decade has been intended to enhance operational safety, provide for risk-reduction measures, and enhance regulatory capability. The U.S. approach to improving the safety of Soviet-designed reactors has matured into a large multi-year program known as the Soviet-Designed Reactor Safety Program that is managed by the U.S. Department of Energy (US DOE). The mission of the program is to implement a self-sustaining nuclear safety improvement program that would lead to internationally accepted safety practices at the plants. Those practices would create a safety culture that would be reflected in the operation, regulation, and professional attitudes of the designers, operators, and regulators of the nuclear facilities. A key component of this larger program has been the Plant Safety Evaluation Program, which supports in-depth safety assessments of VVER and RBMK plants. (author)

Relap5 Analysis of Processes in Reactor Cooling Circuit and Reactor Cavity in Case of Station Blackout in RBMK-1500

International Nuclear Information System (INIS)

Kaliatka, A.

2007-01-01

Ignalina NPP is equipped with channel-type boiling-water graphite-moderated reactor RBMK-1500. Results of the level-1 probabilistic safety assessment of the Ignalina NPP have shown that in topography of the risk, the transients with failure of long-term core cooling other than LOCA are the main contributors to the core damage frequency. The total loss of off-site power with a failure to start any diesel generator, that is station blackout, is the event which could lead to the loss of long-term core cooling. Such accident could lead to multiple ruptures of fuel channels with severe consequences and should be analyzed in order to estimate the timing of the key events and the possibilities for accident management. This paper presents the results of the analysis of station blackout at Ignalina NPP. Analysis was performed using thermal-hydraulic state-of-the-art RELAP5/MOD3.2 code. The response of reactor cooling system and the processes in the reactor cavity and its venting system in case of a few fuel-channel ruptures due to overheating were demonstrated. The possible measures for prevention of the development of this beyond design basis accident (BDBA) to a severe accident are discussed

IAEA/SiP senior managers workshop on international promotion of safety culture for the NPPs with RBMK reactors. Working material

International Nuclear Information System (INIS)

1996-01-01

The IAEA/SiP Senior Managers Workshop on International Promotion of Safety Culture for the NPPs with RMBK reactors was organized in the frame of the IAEA Technical Cooperation Regional Project RER/9/035 and the IAEA Extrabudgetary Project on WWER and RBMK Safety in co-operation with Swedish International Project Nuclear Safety (SiP). It took place at the Forsmark NPP, Sweden, from 1 to 4 October 1996. The objectives of the workshop were to provide a forum for senior managers to exchange national and international experience on factors influencing safety culture, to better understand these factors and to further enhance promotion of safety culture. Twenty-three specialists participated in the workshop from six countries (Canada, Lithuania, Russian Federation, Sweden, Ukraine and USA) and from two international organizations (WANO, EC-G24 coordination). Participants were from regulatory bodies, ministries and operational organizations of respective countries. The INSAG-4 definition of safety culture was taken as a starting point for the discussions, but at the start of the workshop participants did not seem to have the same understanding of what is contained in the safety culture context. Specifically the difference between measures taken to improve safety and establishing a proper safety culture level was discussed with useful results. Some participants proposed quantitative safety culture indicators, but there was no agreement at this stage about how to define them. Refs

IAEA/SiP senior managers workshop on international promotion of safety culture for the NPPs with RBMK reactors. Working material

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-12-31

The IAEA/SiP Senior Managers Workshop on International Promotion of Safety Culture for the NPPs with RMBK reactors was organized in the frame of the IAEA Technical Cooperation Regional Project RER/9/035 and the IAEA Extrabudgetary Project on WWER and RBMK Safety in co-operation with Swedish International Project Nuclear Safety (SiP). It took place at the Forsmark NPP, Sweden, from 1 to 4 October 1996. The objectives of the workshop were to provide a forum for senior managers to exchange national and international experience on factors influencing safety culture, to better understand these factors and to further enhance promotion of safety culture. Twenty-three specialists participated in the workshop from six countries (Canada, Lithuania, Russian Federation, Sweden, Ukraine and USA) and from two international organizations (WANO, EC-G24 coordination). Participants were from regulatory bodies, ministries and operational organizations of respective countries. The INSAG-4 definition of safety culture was taken as a starting point for the discussions, but at the start of the workshop participants did not seem to have the same understanding of what is contained in the safety culture context. Specifically the difference between measures taken to improve safety and establishing a proper safety culture level was discussed with useful results. Some participants proposed quantitative safety culture indicators, but there was no agreement at this stage about how to define them. Refs.

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

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

A Review of Root Causes of SCC Phenomena in BWR/RBMK: An Overview of Radiation-Induced Long Cell Action Relevant to SCC

International Nuclear Information System (INIS)

Genn Saji

2004-01-01

The author suggests a new hypothetical mechanism: radiation-induced 'long cell action' may cause electrolytic corrosion. In this mechanism, SCC (stress corrosion cracking) results from auto-catalytic growth of cracks in crevice water chemistry that is kept acidic by a combination of hydration of cations released from crack tips. The acidic chemistry is maintained by radiation-induced 'long cell action' in pits which are maintained by a trans-passive corrosion process under a stress field. The pivotal point of the thesis is 'long cell action' which appears not to have been investigated in the nuclear community. It is because the reactor water used in BWR/RBMK systems has a very low electrical conductivity. For 'long cell action' to take place, there must be an unknown ion transport mechanism. One potential mechanism can be the high flow rate of the reactor water, carrying ionic species from the anode to the cathode. The other is the effective removal of ferrous ions by deposition as crud, which enhanced by the decomposition of H 2 O 2 . There are also some surprising similarities between SCC in the reactor systems and the basic mechanism of underground corrosion by long cell action. In this mechanism, the 'long cell action' is induced by a difference in availability of oxygen inside the soil. Conduction of electrons through an electric conductor over a long distance plays a significant role as they are released by dissolution of metallic ions and sucked up from the metal surface. (author)

Safety assessment of proposed modifications of Ignalina nuclear power plant. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

International Nuclear Information System (INIS)

1995-09-01

The objective of the meeting was to further discuss previous findings and recommendations and their application to the particular situation of the Ignalina NPP. Since design information and a series of proposed modifications for INPP had been prepared by the main RBMK designer, Research and Development Institute for Power Engineering (RDIPE), it was considered appropriate to conduct the meeting in two parts, the first from 17 to 22 October 1994, at RDIPE headquarters in Moscow and the second from 24 to 28 October 1994, at the plant site in Lithuania. Twelve international experts and IAEA staff participated in the meetings, together with a large group of RDIPE specialists and plant staff. The review covered five topical areas: core monitoring and control; pressure boundary integrity; accident migration; safety and support systems and instrumentation and control. A summary of the reviews in each technical areas is given in this report. Appendices 1 and 5 present the records of the reviews and detailed findings and recommendations in each topical area. The experts strongly supported the effort to develop a new extended safety analysis report. They also stressed the need for close monitoring of the fuel channel conditions and the need for an integrated approach for the upgrading of the control and safety systems. Refs, figs, tabs

Effect of nonlinear void reactivity on bifurcation characteristics of a lumped-parameter model of a BWR: A study relevant to RBMK

Energy Technology Data Exchange (ETDEWEB)

Verma, Dinkar, E-mail: dinkar@iitk.ac.in [Nuclear Engineering and Technology Program, Indian Institute of Technology Kanpur, Kanpur 208 016 (India); Kalra, Manjeet Singh, E-mail: drmanjeet.singh@dituniversity.edu.in [DIT University, Dehradun 248 009 (India); Wahi, Pankaj, E-mail: wahi@iitk.ac.in [Department of Mechanical Engineering, Indian Institute of Technology Kanpur, Kanpur 208 016 (India)

2017-04-15

Highlights: • A simplified model with nonlinear void reactivity feedback is studied. • Method of multiple scales for nonlinear analysis and oscillation characteristics. • Second order void reactivity dominates in determining system dynamics. • Opposing signs of linear and quadratic void reactivity enhances global safety. - Abstract: In the present work, the effect of nonlinear void reactivity on the dynamics of a simplified lumped-parameter model for a boiling water reactor (BWR) is investigated. A mathematical model of five differential equations comprising of neutronics and thermal-hydraulics encompassing the nonlinearities associated with both the reactivity feedbacks and the heat transfer process has been used. To this end, we have considered parameters relevant to RBMK for which the void reactivity is known to be nonlinear. A nonlinear analysis of the model exploiting the method of multiple time scales (MMTS) predicts the occurrence of the two types of Hopf bifurcation, namely subcritical and supercritical, leading to the evolution of limit cycles for a range of parameters. Numerical simulations have been performed to verify the analytical results obtained by MMTS. The study shows that the nonlinear reactivity has a significant influence on the system dynamics. A parametric study with varying nominal reactor power and operating conditions in coolant channel has also been performed which shows the effect of change in concerned parameter on the boundary between regions of sub- and super-critical Hopf bifurcations in the space constituted by the two coefficients of reactivities viz. the void and the Doppler coefficient of reactivities. In particular, we find that introduction of a negative quadratic term in the void reactivity feedback significantly increases the supercritical region and dominates in determining the system dynamics.

Chernobyl reactor transient simulation study

International Nuclear Information System (INIS)

Gaber, F.A.; El Messiry, A.M.

1988-01-01

This paper deals with the Chernobyl nuclear power station transient simulation study. The Chernobyl (RBMK) reactor is a graphite moderated pressure tube type reactor. It is cooled by circulating light water that boils in the upper parts of vertical pressure tubes to produce steam. At equilibrium fuel irradiation, the RBMK reactor has a positive void reactivity coefficient. However, the fuel temperature coefficient is negative and the net effect of a power change depends upon the power level. Under normal operating conditions the net effect (power coefficient) is negative at full power and becomes positive under certain transient conditions. A series of dynamic performance transient analysis for RBMK reactor, pressurized water reactor (PWR) and fast breeder reactor (FBR) have been performed using digital simulator codes, the purpose of this transient study is to show that an accident of Chernobyl's severity does not occur in PWR or FBR nuclear power reactors. This appears from the study of the inherent, stability of RBMK, PWR and FBR under certain transient conditions. This inherent stability is related to the effect of the feed back reactivity. The power distribution stability in the graphite RBMK reactor is difficult to maintain throughout its entire life, so the reactor has an inherent instability. PWR has larger negative temperature coefficient of reactivity, therefore, the PWR by itself has a large amount of natural stability, so PWR is inherently safe. FBR has positive sodium expansion coefficient, therefore it has insufficient stability it has been concluded that PWR has safe operation than FBR and RBMK reactors

Exploitation questions regarding channel type reactors: water graphite channel reactors (operation, reconstruction, advantages and disadvantages)

International Nuclear Information System (INIS)

Chichindaev, D.A.

2001-01-01

An overview of up-grade of the RBMK-type reactors is given. I this paper the core design and core monitoring, pressure boundary integrity, RBMK basic design and safety improvements emergency core cooling system (ECCS) as well as reactor cavity overpressure protection system (RCOPS) are discussed

Operator support systems in nuclear power plants. RDIPE's activities in the field of CSS

International Nuclear Information System (INIS)

Gorelov, A.

1996-01-01

Research and Development Institute of Power Engineering (RDIPE) works in the field of computerized support system for the RBMK NPP during last 15 years. The first system which had some special features to be classified as CSS was designed for the Ignalina NPP in the beginning of the 1980s. THese efforts were mainly aimed to create standards for CSS design and V and V; to establish general requirements for human-machine interface tools; to upgrade support of operator as well as maintenance and administrative staff by implementing new hardware and software into existing RBMK computer-based monitoring systems and to develop the SPDS for all RBMK units. (author)

Reactor safety in Eastern Europe. Proceedings

International Nuclear Information System (INIS)

1995-02-01

The papers presented to the GRS colloquium refer to the cooperative activities for reactor accident analysis and modification of the GRS computer codes for their application to reactors of the Russian design types of WWER or RBMK. Another topic is the safety of RBMK reactors in particular, and the current status of investigations and studies addressing the containment of unit 4 of the Chernobyl reactor station. (HP) [de

The Barselina Project Phase 4 Summary report. Ignalina Unit 2 Probabilistic Safety Analysis

Energy Technology Data Exchange (ETDEWEB)

Johansson, Gunnar [ES-Konsult AB, Stockholm (Sweden); Hellstroem, P. [RELCON AB, Solna (Sweden); Zheltobriuch, G.; Bagdonas, A. [Ignalina Power Plant, Visaginas (Lithuania)

1996-12-01

The Barselina Project was initiated in the summer of 1991. The project is a multilateral co-operation between Lithuania, Russia and Sweden. The long range objective is to establish common perspectives and unified bases for assessment of severe accident risks and needs for remedial measures for the RBMK reactors. The Swedish BWR Barsebaeck is used as reference plant and the Lithuanian RBMK Ignalina as application plant. During phase 3, from March, 1993 to June, 1994, a full scope Probabilistic Safety Analysis (PSA) model of the Ignalina Nuclear Power Plant unit 2 (INPP-2) was developed to identify possible safety improvement of risk importance. The probabilistic methodology was applied on a plant specific basis for a channel type reactor of RBMK design. To increase the realism of the risk model a set of deterministic analyses were performed and plant/RBMK-specific data bases were developed and used. A general concept for analysing this type of reactor was developed. During phase 4, July 1994 to September 1996, the PSA was further developed, taking into account plant changes, improved modeling methods and extended plant information concerning dependencies (area events, dynamic effects, electrical and signal dependencies). The updated model is quantified and new results and conclusions are evaluated.

Chernobyl and the safety of nuclear reactors in OECD countries

International Nuclear Information System (INIS)

1987-01-01

This report assesses the possible bearing of the Chernobyl accident on the safety of nuclear reactors in OECD countries. It discusses analyses of the accident performed in several countries as well as improvements to the safety of RBMK reactors announced by the USSR. Several remaining questions are identified. The report compares RBMK safety features with those of commercial reactors in OECD countries and evaluates a number of issues raised by the Chernobyl accident

Reactor safety in Eastern Europe

International Nuclear Information System (INIS)

1995-02-01

The papers presented to the GRS colloquium refer to the cooperative activities for reactor accident analysis and modification of the GRS computer codes for their application to reactors of the Russian design types of WWER or RBMK. Another topic is the safety of RBMK reactors in particular, and the current status of investigations and studies addressing the containment of unit 4 of the Chernobyl reactor station. All papers are indexed separately in report GRS--117. (HP)

Potential consequences in Norway after a hypothetical accident at Leningrad nuclear power plant. Potential release, fallout and predicted impacts on the environment

Energy Technology Data Exchange (ETDEWEB)

Nalbandyan, A.; Ytre-Eide, M.A.; Thoerring, H.; Liland, A.; Bartnicki, J.; Balonov, M.

2012-06-15

The report describes different hypothetical accident scenarios at the Leningrad nuclear power plant for both RBMK and VVER-1200 reactors. The estimated release is combined with different meteorological scenarios to predict possible fallout of radioactive substances in Norway. For a hypothetical catastrophic accident at an RBMK reactor combined with a meteorological worst case scenario, the consequences in Norway could be considerable. Foodstuffs in many regions would be contaminated above the food intervention levels for radioactive cesium in Norway. (Author)

Potential consequences in Norway after a hypothetical accident at Leningrad nuclear power plant. Potential release, fallout and predicted impacts on the environment

International Nuclear Information System (INIS)

Nalbandyan, A.; Ytre-Eide, M.A.; Thoerring, H.; Liland, A.; Bartnicki, J.; Balonov, M.

2012-06-01

The report describes different hypothetical accident scenarios at the Leningrad nuclear power plant for both RBMK and VVER-1200 reactors. The estimated release is combined with different meteorological scenarios to predict possible fallout of radioactive substances in Norway. For a hypothetical catastrophic accident at an RBMK reactor combined with a meteorological worst case scenario, the consequences in Norway could be considerable. Foodstuffs in many regions would be contaminated above the food intervention levels for radioactive cesium in Norway. (Author)

Evaluation of special safety features of the SNR-300 in view of the Chernobyl accident

International Nuclear Information System (INIS)

Vossebrecker, H.

1987-03-01

A comparison of those characteristics, which decisively influenced the accident in the RMBK-1000 reactor, with the safety features of SNR-300 has been performed. The conclusions of this comparison are presented in the present report. The SNR-300 is characterized by a stable reactivity behaviour and good controllability, whereas RBMK-1000 has an instable behaviour and complex spatial dependencies in the core. Among other points, design deficiencies in the protection and emergency shutdown systems were responsible for the Chernobyl accident. The protection and scram systems of the SNR-300 are unquestionably superior to those of the RBMK-1000 with regard to redundancy, diversity, degree of automation, separation of operational and safety-relevant tasks, protection against inadmissible interventions, effectiveness and safety reserves. Therefore, excursion accidents can be classified as hypothetical for SNR-300. Due to elementary physical properties, possible energy releases during hypothetical excursions are substantially lower for SNR-300 and would be controlled by the design of the primary system and containment systems. No damage limiting measures are provided in the RBMK-100 for excursion accidents. Finally, exothermal processes augmented the consequences of the accident in the RBMK-1000 and the long-lasting graphite fire intensified the release of radioactivity. In the SNR-300, however, inertisation of the containment, the steel plate lining and the floor troughs ensure that activity enclosure inside the containment after leakage or hypothetical excursion accident is not endangered by exothermal reactions. Further safety aspects are presented in the report, which can be linked with the accident in Chernobyl. In summary, it is obvious that the disadvantageous physical and technical features of the RBMK-1000 do either not exist in the SNR-300 or are covered by the safety design

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

Improvements in the methods for optimising dose rate CTOR NPPs in the Russian Federation

International Nuclear Information System (INIS)

Noskov, A.A.

1995-01-01

The paper describes the results of a model investigation and an occupational radiation exposure analysis for reducing the collective doses on LWGMR nuclear power plants. The first RBMK NPP had been commissioned more than twenty years ago. It was the new generation unit of the Light Water Graphite Moderated Reactor (LWGMR). The RBMK-1000 occupational exposures were the same as or smaller than BWR doses. For example in 1980-1983 US BWR collective doses were 12.30-10.17 man-Sv per unit-year and the equivalent RBMK-1000 were 5.00-9.05 man-Sv per unit-year. A new generation of LWGMR, the MKER-800, is currently at the design stage and the design target for the plant is 1.0 man-Sv per GW(e)-year. There were three principal trends in the investigation of the occupational radiation exposure mechanism on LWMGR units. They were: (1) ionizing radiation sources; (2) radiation field distribution in the NPP compartments; (3) occupational radiation exposure profile. (author)

Human factors in RBNK plants

International Nuclear Information System (INIS)

Demitrack, T.

1995-01-01

The Safety of RBMK nuclear power plants in the Russian Federation, The Ukraine and Lithuanian is a topic of concern to the European Union and other Western European countries. The European Commission, Sweden, Finland and Canada financed the project Safety Design Solutions and Operation of NPP with RBMK Reactors. The project examined nine issues and recommended safety improvements which will form the basis of future European Commission spending on these power plants. During its year of work, the project examined these issues: 1. Systems Engineering and progression of accidents 2. Protection System 3. Core Physics 4. External Events 5. Engineering Quality 6. Operating Experience 7. Human Factors 8. Regulatory Interface 9. Probabilistic Safety analysis Empresarios Agrupados, in collaboration with other western European firms, the Russian Federation and Lithuanian took part in two of these groups, Human Factors and Probabilistic Safety Analysis. This presentation gives a brief description of the most important aspects of human factors in RBMK plants, focusing on operations organization, training and education

Graphite-water steam-generating reactor in the USSR

Energy Technology Data Exchange (ETDEWEB)

Dollezhal, N A [AN SSSR, Moscow

1981-10-01

One of the types of power reactor used in the USSR is the graphite-water steam-generating reactor RBMK. This produces saturated steam at a pressure of 7MPa. Reactors giving 1GWe each have been installed at the Leningrad, Kursk, Chernobyl and other power stations. Further stations using reactors of this type are being built. A description is given of the fuel element design, and of the layout of the plant. The main characteristics of RBMK reactors using fuel of rated and higher enrichment are listed.

Papers submitted to the international forum ''one decade after Chernobyl: nuclear safety aspects''. Working material

International Nuclear Information System (INIS)

1996-01-01

The objective of the forum is to review the remedial measures taken since the Chernobyl accident for improving the safety of RBMK reactors and the Chernobyl containment structure (sarcophagus). The forum will also serve to exchange information on national, bilateral and multilateral efforts for the enhancement of RBMK safety. The conclusions and recommendations will serve as a basis for a background paper to be prepared for presentation, by the forum chairman, at the International Conference ''One decade after Chernobyl'' to held in Vienna from 8-12 April 1996. Refs, figs, tabs

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

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-09-01

This report presents the safety issues in 'small series' WWER-1000 nuclear power plants (NPPs). Safety issues are deviations from current recognized safety practices in design and operation judged to be safety significant by their impact on the plants' defence in depth. This report is intended to serve as reference for the development of plant specific safety improvement programmes and for the evaluation of measures proposed and/or implemented. The identification of safety issues is based on safety studies conducted by the operators of 'small series' WWER-1000 units and by organizations dealing with these reactors, on findings of IAEA safety missions to 'small series' WWER-1000 plants in South Ukraine, at Novovoronezh and Kalinin, and on information obtained from specialists from various countries during an IAEA consultants meeting, 8-12 September 1997 in Vienna, within the framework of the Extra budgetary Programme on the Safety of WWER and RBMK NPPs. Safety issues are first presented according to their impact on the main safety functions and are then described individually. The safety issues are characterized by issue title and specified by issue clarification. Safety issues connected with plant design are followed by the ranking of the issue and ranking justification. Altogether 85 safety issues have been identified, 12 of which are in Category III (defence in depth is insufficient, immediate corrective action is necessary), 38 in Category 11 (defence in depth is degraded, action is needed to resolve the issue) and 22 in Category I (departure from international practices, to be addressed as part of actions to resolve higher priority issues). In the case of operational safety issues (13 safety issues) no ranking is provided as the available material was considered insufficient. For each safety issue, comments and recommendations are made by the IAEA; the status of corresponding measures to improve safety implemented or planned at each site are presented in the

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

International Nuclear Information System (INIS)

2000-09-01

This report presents the safety issues in 'small series' WWER-1000 nuclear power plants (NPPs). Safety issues are deviations from current recognized safety practices in design and operation judged to be safety significant by their impact on the plants' defence in depth. This report is intended to serve as reference for the development of plant specific safety improvement programmes and for the evaluation of measures proposed and/or implemented. The identification of safety issues is based on safety studies conducted by the operators of 'small series' WWER-1000 units and by organizations dealing with these reactors, on findings of IAEA safety missions to 'small series' WWER-1000 plants in South Ukraine, at Novovoronezh and Kalinin, and on information obtained from specialists from various countries during an IAEA consultants meeting, 8-12 September 1997 in Vienna, within the framework of the Extra budgetary Programme on the Safety of WWER and RBMK NPPs. Safety issues are first presented according to their impact on the main safety functions and are then described individually. The safety issues are characterized by issue title and specified by issue clarification. Safety issues connected with plant design are followed by the ranking of the issue and ranking justification. Altogether 85 safety issues have been identified, 12 of which are in Category III (defence in depth is insufficient, immediate corrective action is necessary), 38 in Category 11 (defence in depth is degraded, action is needed to resolve the issue) and 22 in Category I (departure from international practices, to be addressed as part of actions to resolve higher priority issues). In the case of operational safety issues (13 safety issues) no ranking is provided as the available material was considered insufficient. For each safety issue, comments and recommendations are made by the IAEA; the status of corresponding measures to improve safety implemented or planned at each site are presented in the

Safety research needs for Russian-designed reactors. Requirements situation

International Nuclear Information System (INIS)

Brown, R. Allan; Holmstrom, Heikki; Reocreux, Michel; Schulz, Helmut; Liesch, Klaus; Santarossa, Giampiero; Hayamizu, Yoshitaka; Asmolov, Vladimir; Bolshov, Leonid; Strizhov, Valerii; Bougaenko, Sergei; Nikitin, Yuri N.; Proklov, Vladimir; Potapov, Alexandre; Kinnersly, Stephen R.; Voronin, Leonid M.; Honekamp, John R.; Frescura, Gianni M.; Maki, Nobuo; Reig, Javier; ); Bekjord, Eric S.; Rosinger, Herbert E.

1998-01-01

In June 1995, an OECD Support Group was set up to perform a broad study of the safety research needs of Russian-designed reactors. The emphasis of the study is on the VVER-type reactors in part because of the larger base of knowledge within the NEA Member countries related to LWRs. For the RBMKs, the study does not make the judgement that such reactors can be brought to acceptable levels of safety but focuses on near term efforts that can contribute to reducing the risk to the public. The need for the safety research must be evaluated in context of the lifetime of the reactors. The principal outcome of the work of the Support Group is the identification of a number of research topics which the members believe should receive priority attention over the next several years if risk levels are to be reduced and public safety enhanced. These appear in the Conclusions and Recommendations section of the report, and are the following: - The most important near-term need for VVER and RBMK safety research is to establish a sound technical basis for the emergency operating procedures used by the plant staff to prevent or halt the progression of accidents (i.e., Accident Management) and for plant safety improvements. - Co-operation of Western and Eastern experts should help to avoid East-West know-how gaps in the future, as safety technology continues to improve. - Safety research in Eastern countries will make an important contribution to public safety as it has in OECD countries. - RBMK safety research, including verification of codes, starts from a smaller base of experience than VVER, and is at an earlier stage of development. Technical Conclusions: - Research to improve human performance and operational safety of VVER and RBMK plants is extremely important. - VVER thermal-hydraulic and reactor physics research should focus on full validation of codes to VVER-specific features, and on extension of experimental data base. - Methods of assessing VVER pressure boundary

Coupled neutronic and thermal-hydraulic code benchmark activities at the International Nuclear Safety Center

International Nuclear Information System (INIS)

Podlazov, L. N.

1998-01-01

Two realistic benchmark problems are defined and used to assess the performance of coupled thermal-hydraulic and neutronic codes used in simulating dynamic processes in VVER-1000 and RBMK reactor systems. One of the problems simulates a design basis accident involving the ejection of three control and protection system rods from a VVER-1000 reactor. The other is based on a postulated rod withdrawal from an operating RBMK reactor. Preliminary results calculated by various codes are compared. While these results show significant differences, the intercomparisons performed so far provide a basis for further evaluation of code limitations and modeling assumptions

Thermal-hydraulic instabilities in pressure tube graphite - moderated boiling water reactors

Energy Technology Data Exchange (ETDEWEB)

Tsiklauri, G.; Schmitt, B.

1995-09-01

Thermally induced two-phase instabilities in non-uniformly heated boiling channels in RBMK-1000 reactor have been analyzed using RELAP5/MOD3 code. The RELAP5 model of a RBMK-1000 reactor was developed to investigate low flow in a distribution group header (DGH) supplying 44 fuel pressure tubes. The model was evaluated against experimental data. The results of the calculations indicate that the period of oscillation for the high power tube varied from 3.1s to 2.6s, over the power range of 2.0 MW to 3.0 MW, respectively. The amplitude of the flow oscillation for the high powered tube varied from +100% to -150% of the tube average flow. Reverse flow did not occur in the lower power tubes. The amplitude of oscillation in the subcooled region at the inlet to the fuel region is higher than in the saturated region at the outlet. In the upper fuel region and outlet connectors the flow oscillations are dissipated. The threshold of flow instability for the high powered tubes of a RBMK reactor is compared to Japanese data and appears to be in good agreement.

Pressure-tube reactors as a part of Russian nuclear fleet

International Nuclear Information System (INIS)

Gmyrko, V.E.; Grozdov, I.I.; Nikitin, Yu.M.; Petrov, A.A.; Potapov, A.A.; Finyakin, A.F.

2007-01-01

The place and role of channel reactors in nuclear power in our country and the main measures for upgrading and improving the power generating units of nuclear power plants with RBMK reactors are described. It is shown that the risk indicators for serious damage to the core of power generating units with RBMK reactors are lower after upgrading and the corresponding IAEA criterion established for operating nuclear power plants. Upgrading and implementation of a service life extension program has made it possible to obtain licenses for continuing operation of power generating units with first-generation RBMK reactors and predicting a service life increase to 45 years. The characteristics of nuclear power plants with channel reactors with more highly developed internal and natural safety properties are shown in evolutionary designs of the power generating units MCER-860,-1000, and-1500, which have protective shells and which meet all requirements for power generating units built today. It is shown that innovative solutions for the channel reactor concept can be implemented on the basis of the designs of power generating units with nuclear superheating of steam or on the basis of designs for developing reactors with supercritical parameters [ru

Thermal-hydraulic instabilities in pressure tube graphite-moderated boiling water reactors

International Nuclear Information System (INIS)

Tsiklauri, G.; Schmitt, B.

1995-09-01

Thermally induced two-phase instabilities in non-uniformly heated boiling charmers in RBMK-1000 reactor have been analyzed using RELAP5/MOD3 code. The RELAP5 model of a RBMK-1000 reactor was developed to investigate low flow in a distribution group header (DGH) supplying 44 fuel pressure tubes. The model was evaluated against experimental data. The results of the calculations indicate that the period of oscillation for the high power tube varied from 3.1s to 2.6s, over the power range of 2.0 MW to 3.0 MW, respectively. The amplitude of the flow oscillation for the high powered tube varied from +100% to -150% of the tube average flow. Reverse flow did not occur in the lower power tubes. The amplitude of oscillation in the subcooled region at the inlet to the fuel region is higher than in the saturated region at the outlet. In the upper fuel region and outlet connectors the flow oscillations are dissipated. The threshold of flow instability for the high powered tubes of a RBMK reactor is compared to Japanese data and appears to be in good agreement

Multimedia system for the visitors' centre at the Ignalina NPP

International Nuclear Information System (INIS)

Alvers, Margareta

1999-01-01

The contents illustrated with video clips, animations, photographs, show the follwing: History of Ignalina NPP (INPP) growing; Visaginas - how the town came into being; Lake Druksiai; Development of nuclear power; Technical data of INPP; Description of INPP; Characteristic features of RBMK reactors; Reactor design; Technical parametres of RBMK-1500 reactor; Nuclear reaction and nuclear fission; Types of nuclear reactors; Circuits and systems; Radiation safety; Safety systems at the INPP; Upgrading nuclear safety at INPP following the Chernobyl accident; Safety problems at MP; Radioactive waste management in the world; RW Management at MP; Energy in Lithuania (thermal power stations, cogeneration plants, producing biogas from organic waste)

Measurement of Neutron Field Characteristics at Nuclear-Physics Instalations for Personal Radiation Monitoring

CERN Document Server

Alekseev, A G; Britvich, G I; Kosyanenko, E V; Pikalov, V A; Gomonov, I P

2003-01-01

n this work the observed data of neutron spectra on Rostov NEP, Kursk NEP and Smolensk NEP and on the reactor IRT MIPHI are submitted. For measurement of neutron spectra two types of spectrometer were used: SHANS (IHEP design ) and SDN-MS01 (FEI design). The comparison of the data measurements per-formed by those spectrometers above one-type cells on the reactor RBMK is submitted. On the basis of the 1-st horizontal experimental channel HEC-1 of the IRT reactor 4 reference fields of neutrons are investigated. It is shown, that spectra of neutrons of reference fields can be used for imitation of neutron spectra for conditions of NEP with VVER and RBMK type reactors.

Chernobyl

International Nuclear Information System (INIS)

1986-01-01

This leaflet has been prepared by the Central Electricity Generating Board. Following the accident at Chernobyl nuclear power station in the Soviet Union people are concerned about the safety of the UK's nuclear power stations. This leaflet explains that Chernobyl is unlike any nuclear station operating or planned in the UK and under the CEGB's stringent safety rules it could not have been built in the UK. The leaflet explains what happened at Chernobyl and compares the RBMK design and British reactors. The bodies concerned with reactor safety are noted. The containment of radioactivity and emergency procedures are explained. The PWR design for Sizewell-B is stated to be much safer than the RBMK Chernobyl design. (UK)

From USA operation experience of industrial uranium-graphite reactors

International Nuclear Information System (INIS)

Burdakov, N.S.

1996-01-01

The review on materials, presented by a group of the USA specialists at the seminar in Moscow on October 9-11, 1995 is considered. The above specialists shared their experience in operation of the Hanford industrial reactors, aimed at plutonium production for atomic bombs. The purpose of the above visit consisted in providing assistance to the Russian specialists by evaluation and modernization of operational conditions safety improvement of the RBMK type reactors. Special attention is paid to the behaviour of the graphite lining and channel tubes with an account of possible channel power interaction with the reactor structural units. The information on the experience of the Hanford reactor operation may be useful for specialists, operating the RBMK type reactors

Analysis of the RBMK-1500 type reactor emergency core cooling system behavior, taking into account the specified hydraulic characteristics of fast acting motor valves

International Nuclear Information System (INIS)

Kaliatka, A.; Ognerubov, V.; Adomavicius, A.; Ziedelis, S.

2005-01-01

During the accident analysis of nuclear power plants, reliability and uncertainty of results depends on adequateness of mathematical models of main elements and phenomena in systems important to safety. The best way for qualification of these models is collation with relevant experimental data. However, at the case of lack of such data modern computational fluid dynamics codes can be used for this purpose. This paper presents the results of an attempt to specify the hydraulic characteristics of the fast acting motor valves as well as to demonstrate the impact of these characteristics to transient processes in emergency core cooling system of the RBMK-1500 type reactor. For these purposes the finite element model of fast acting motor valve was developed and analyzed, using two separate computational fluid dynamics codes in parallel: CFX5 and COSMOS/FLOWORKS. Both all main design particularities and changes of flow structure during valve opening (closure) process were taken into account. It was demonstrated, that the obtained dependencies of changes of hydraulic loss coefficient in respect of relative valve opening (closure) rate substantially differ from those commonly used in thermal-hydraulic calculations of nuclear reactors. This difference is extremely big at the square one of the valve opening process, when the value of the valve hydraulic resistance is most important to flow of coolant channelized to the group distribution header. The series of thermal-hydraulic calculations of the maximum design-basis accident initiated by full break of main circulation pump pressure header were performed. The obtained dependencies of changes of hydraulic loss coefficient in respect of relative valve opening (closure) rate as well as those commonly used in thermal-hydraulic code RELAP5 were used. The results of calculations show, that in the initial stage of accident flow of coolant going from emergency core cooling system via fast acting motor valves to group distribution

New generation main control room of enhanced safety NPP with MKER reactor

International Nuclear Information System (INIS)

Golovanev, V.E.; Gorelov, A.I.; Proshin, V.A.

1994-01-01

Russia is planning to begin the gradual substitution RMBK NPP units, whose resources were worked out itself, to NPP units with a 800 MW multiloop boiling water power reactor (MKER-800) enhanced safety at next ten-year period. Main drawbacks of RBMK Reactor were completely removed in design of MKER-800 reactor. Moreover some special decisions were made to give MKER-800 self-safety properties. The proposed design of the MKER-800 enhanced safety reactor is not only fully free from the drawbacks of the RBMK reactors, but also show a number of advantages of channel-type reactors. This Paper presents some preliminary proposals of MCR Design, that developed Research and Development Institute of Power Energy (RDIPE). 6 refs, 2 figs

Optimization of the power distribution in a large power reactor core

International Nuclear Information System (INIS)

Emel'yanov, I.Ya.; Nazaryan, V.G.; Postnikov, V.V.

1978-01-01

The reactor power distribution optimization problem is solved for the case of the RBMK-1000 reactor. The algorithm is written in terms of the linear programming method. The algorithm rests on two assumptions: 1) the relative power change of each fuel assembly is a linear function of reactivity increment caused by displacement of a regulating rod; 2) the change is an additive value. The algorithm is written in ALGOL for the BESM-6 computer. The optimum reactivity gain for the RBMK reactor has proved to equal the reactivity of 35-40 control rods. The results obtained confirm the validity of the assumptions. It is noted that the total computation time on the BESM-6 can be reduced to 20 min

Main problems of increasing labour productivity in the power plant construction

International Nuclear Information System (INIS)

Falaleev, P.P.

1984-01-01

The reserve for labour productivity growth in power-, industrial- and civil engineering in the USSR Minenergo system are discussed. Such reserve comprises: introduction of effective designs, increase of technological readiness of structures; a higher mechanization level in construction, improvement of industrial organization, economical and social aspects. Decrease of labour inputs in NPP construction will be attained by using unified designs of serial WWER-1000, RBMK-1000- and RBMK-1500 reactors as well as by developing nuclear power construction complexes-industrial-construction enterprises for manufacturing and transport of special structures as well as for performing civil engineering and installation work on the ground part of the reactor building and special structure. Other potentialities for increasing labour productivity in NPP construction are considered

Reserves of labour content reduction in NPP construction

International Nuclear Information System (INIS)

Bekerman, R.E.

1986-01-01

Specific labour contents when constructing NPP with RBMK-1000 and WWER-1000 type reactors are presented. Factors affecting labour content of NPP construction are shown. Measures aimed at labour content decrease are suggested

Specific energy released in power reactors

International Nuclear Information System (INIS)

Zaritskaya, T.S.; Kiselev, G.V.; Rudik, A.P.; Tsenter, Eh.M.

1986-01-01

Technique of determination are described and analysis of specific energy for different methods of critically maintance of RBMK and WWER-440 reactors are conducted. Characteristics of the optimal mode of critically maintanance are determined

Lithuanian requirements for ageing management of systems and components important to safety of nuclear power plant

International Nuclear Information System (INIS)

Ramanauskiene, A.

2000-01-01

In this paper the Lithuanian requirements for ageing management of systems and components important to safety of Ignalina nuclear power plant (two RBMK-1500 water-cooled graphite moderated channel-type power reactors) are presented

Accounting for the residual stress effects on the creep deformation of channel tubes

International Nuclear Information System (INIS)

Knizhnikov, Yu.N.; Platonov, P.A.; Ul'yanov, A.I.

1985-01-01

The effect of the first kind residual stresses arising in the walls of the zirconium base alloy fules in the process of fabrication on the RBMK type reactor channel tube creep is investigated. Models for calculation of the reactor component creep with account for the relaxation of residual stresses distributed by the wall thickness as well as the radiation and temperature fields are developed. On the basis of the analysis of the data obtained it is concluded that the effect of the residual stresses on the RBMK channel tube deformation for a long-term operation is negligible. But for the short-term fests the results can be noticeably distorted by this factor. The role of internal stresses can also manifest when determining the deformation of radiation elongation of the zirconium base alloy samples

Exploding Chernobyl myths

International Nuclear Information System (INIS)

Arnott, D.

1991-01-01

Misconceptions about the way thermal reactors really work, and the use of misleading terminology, have allowed the western nuclear industry to claim that the accident at the RBMK (water cooled, graphite moderated) type reactor at Chernobyl would not be possible in western type pressurized water reactors. The author contends that control of thermal reactors is only possible because a small but consistent fraction of the secondary neutrons are delayed. If the delayed neutron reaction is overridden by the prompt neutron reaction, control is irretrievably lost and a nuclear explosion, such as at Chernobyl, results. Parallels between the PWR and RBMK are drawn. The consequences of the Chernobyl explosion are discussed and the question is asked: can any combination of circumstances, however improbable, produce a prompt neutron explosion in any western reactors? (UK)

Nuclear power

International Nuclear Information System (INIS)

1987-01-01

''Nuclear Power'' describes how a reactor works and examines the different designs including Magnox, AGR, RBMK and PWR. It charts the growth of nuclear generation in the world and its contributions to world energy resources. (author)

Research contract: Seismic stability of nuclear power plants in Eastern Europe. Final report. From 15 April 1998 to 15 April 1999

International Nuclear Information System (INIS)

Ambriashvili, Y.

1999-01-01

General scientific scope of the presented program is assessment of stability and functionality of the nuclear power plants with RBMK type reactors in relation to External Evens including following: seismic capacity of structures, equipment and distribution systems; capacity of structures for impact type loading; capacity of structures for blast type loading. For the analyses only structures, equipment and distribution systems which are responsible for safety shutdown path are used. Since 1980 Atomenergoproject has been participated in development and carrying out the Research Program related to investigation of seismic stability of RBMK NPPs. This investigation was done for Smolensk and Kursk NPPs. It is known that the design basis of seismic analyses is investigation of dynamic characteristic (main frequencies and main modes) of structures, equipment and distribution systems. Therefore the assessment of capacity of structures and systems can be based on the results of seismic stability investigations. In the present final report some main results of dynamic analyses reactor building, electrical buildings, storage building, pipe lines, separators, electrical equipment, etc. are described. On the basis of the data site seismological investigation, calculations and different type of testing structures and equipment it was accepted, that the RBMK NPP is the Seismic Stability NPP up to 6 intensity by MSK64 scale. The results of calculation and testing investigations can be use for the dynamic analyses on the external events blast and impact loading. Using the results the probability analysis should be done

One decade after Chernobyl: Summing up the consequences of the accident

International Nuclear Information System (INIS)

1996-07-01

This summary is the results of the International Conference ''One decade after Chernobyl''. It includes topics on initial responses, radioactive releases, absorbed radiation doses and health effects, socio-economic impacts as well as safety of RBMK type reactors

Implementation of post-Chernobyl first-priority safety improvement measures at the Leningrad NPP

International Nuclear Information System (INIS)

Eperin, A.P.

1996-01-01

After the severe accident at Chernobyl Unit 4 and in order to ensure RBMK-1000 safe operation, a set of measures was worked out aimed at preclusion of such an accident recurrence. Implementation of these measures is described

20 years after Chernobyl Accident. Future outlook. National Report of Ukraine

Energy Technology Data Exchange (ETDEWEB)

Baloga, V I [ed.

2006-07-01

The scale of the Chernobyl catastrophe - the most severe man made nuclear accident in the history of mankind - is well known to both scientists and politicians worldwide. The basic causes of the catastrophe were as follows: Conduction an incompletely and incorrectly prepared electrical experiment; The low professional level of operators, and of the NPP management and the officials of the Ministry of Electrification as a whole in the area of NPP safety; Insufficient safety level of the graphite-uranium reactor RBMK-1000; Constructive faults RBMK-1000; Personnel mistakes. The report describes and reviews the actions of the governments of the USSR, Ukraine, and the Verkhovna Rada of Ukraine; the activities of scientists in elimination of the accident consequences; and elimination of the additional experience gained over the past years. Mistakes made during these activities are highlighted.

Water channel reactor fuels and fuel channels: Design, performance, research and development. Proceedings of a technical committee meeting

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-01-01

The International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended holding a Technical Committee Meeting on Water Channel Reactor Fuel including into this category fuels and pressure tubes/fuel channels for Atucha-I and II, BWR, CANDU, FUGEN and RBMK reactors. The IWGFPT considered that even if the characteristics of Atucha, CANDUs, BWRs, FUGEN and RBMKs differ considerably, there are also common features. These features include materials aspects, as well as core, fuel assembly and fuel rod design, and some safety issues. There is also some similarity in fuel power history and operating conditions (Atucha-I and II, FUGEN and RBMK). Experts from 11 countries participated at the meeting and presented papers on technology, performance, safety and design, and materials aspects of fuels and pressure tubes/fuel channels for the above types of water channel reactors. Refs, figs, tabs.

Water channel reactor fuels and fuel channels: Design, performance, research and development. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1998-01-01

The International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended holding a Technical Committee Meeting on Water Channel Reactor Fuel including into this category fuels and pressure tubes/fuel channels for Atucha-I and II, BWR, CANDU, FUGEN and RBMK reactors. The IWGFPT considered that even if the characteristics of Atucha, CANDUs, BWRs, FUGEN and RBMKs differ considerably, there are also common features. These features include materials aspects, as well as core, fuel assembly and fuel rod design, and some safety issues. There is also some similarity in fuel power history and operating conditions (Atucha-I and II, FUGEN and RBMK). Experts from 11 countries participated at the meeting and presented papers on technology, performance, safety and design, and materials aspects of fuels and pressure tubes/fuel channels for the above types of water channel reactors

20 years after Chernobyl Accident. Future outlook. National Report of Ukraine

International Nuclear Information System (INIS)

Baloga, V.I.

2006-01-01

The scale of the Chernobyl catastrophe - the most severe man made nuclear accident in the history of mankind - is well known to both scientists and politicians worldwide. The basic causes of the catastrophe were as follows: Conduction an incompletely and incorrectly prepared electrical experiment; The low professional level of operators, and of the NPP management and the officials of the Ministry of Electrification as a whole in the area of NPP safety; Insufficient safety level of the graphite-uranium reactor RBMK-1000; Constructive faults RBMK-1000; Personnel mistakes. The report describes and reviews the actions of the governments of the USSR, Ukraine, and the Verkhovna Rada of Ukraine; the activities of scientists in elimination of the accident consequences; and elimination of the additional experience gained over the past years. Mistakes made during these activities are highlighted

The Chernobyl-4 Reactor and the possible causes of the accident

International Nuclear Information System (INIS)

Motte, F.

1986-01-01

A description and information about the Chernobyl nuclear reactor is given. Some comparison elements between the RBMK reactor type and GCR, CANDU, SGHWR and Hanford N reactor types are presented. A scenario of the possible causes of the accident is discussed. (A.F.)

Benchmark analysis of three main circulation pump sequential trip event at Ignalina NPP

International Nuclear Information System (INIS)

Uspuras, E.; Kaliatka, A.; Urbonas, R.

2001-01-01

The Ignalina Nuclear Power Plant is a twin-unit with two RBMK-1500 reactors. The primary circuit consists of two symmetrical loops. Eight Main Circulation Pumps (MCPs) at the Ignalina NPP are employed for the coolant water forced circulation through the reactor core. The MCPs are joined in groups of four pumps for each loop (three for normal operation and one on standby). This paper presents the benchmark analysis of three main circulation pump sequential trip event at RBMK-1500 using RELAP5 code. During this event all three MCPs in one circulation loop at Unit 2 Ignalina NPP were tripped one after another, because of inadvertent activation of the fire protection system. The comparison of calculated and measured parameters led us to establish realistic thermal hydraulic characteristics of different main circulation circuit components and to verify the model of drum separators pressure and water level controllers.(author)

Nuclear and thermal power plant power ramping capability

International Nuclear Information System (INIS)

Golovach, E.A.

1983-01-01

The possibilities of step power increase by NPP and TPP units under emergency conditions of power grids operation are considered. The data analysis has shown that power units ramping capability with WWER-440, WWER-1000 and RBMK-1000 reactors is higher than that of 300 MW power units on fossil fuel, at the initial time interval (0-30 s). These NPP power units satisfy as to ramping capability the energy system requirements. Higher NPP power units ramping capability is explained by the fact that relative pressure before turbine valves is decreased less than in straight-through boilers while the steam volumes time constant of steam separator-superheaters is less than that of intermediate superheatings. Higher power unit ramping capability with WWER-440 and RBMK-1000 reactors as compared with the WWER-1000 reactor is pointed out as well as the increase of WWER-1000 power unit capability using high-speed turbines

Investigation on Conversion of I-Graphite from Decommissioning of Chernobyl NPP into a Stable Waste Form Acceptable for Long Term Storage and Disposal

International Nuclear Information System (INIS)

Zlobenko, Borys; Fedorenko, Yriy; Yatzenko, Victor; Shabalin, Borys; Skripkin, Vadim

2016-01-01

For Ukraine, the main radiocarbon ( 14 C) source is irradiated graphite from Chernobyl Nuclear Power Plant. The ChNPP is a decommissioned nuclear power station about 14 km northwest of the city of Chernobyl, and 110 km north of Kyiv. The ChNPP had four RBMK reactor units. The commissioning of the first reactor in 1977 was followed by reactor No. 2 (1978), No. 3 (1981), and No.4 (1983). Reactors No.3 and 4 were second generation units, whereas Nos.1 and 2 were first-generation units. RBMK is an acronym for ''High Power Channel-type Reactor'' of a class of graphite-moderated nuclear power reactor with individual fuel channels that uses ordinary water as its coolant and graphite as its moderator. The combination of graphite moderator and water coolant is found in no other type of nuclear reactor

Development of passive condensers for accident localization systems at nuclear power plants in the former USSR

International Nuclear Information System (INIS)

Kuznecov, M.V.

1992-01-01

The development is summarized of passive condensers for accident localization systems at nuclear power plants (with RBMK and WWER reactors) in the former USSR. Basic properties and criteria defining their availability are described, as are experimental tests and technical solution optimization results. (author) 2 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)

Transferring manual ultrasonic inspection procedures - results of a pilot study

International Nuclear Information System (INIS)

Anderson, M.; Taylor, T.; Kadenko, I.

2002-01-01

Results of a manual ultrasonic pilot study for NDE specialists at RBMK nuclear reactor sites are presented. Probabilities of detection and false calls, using two different grading criteria, are estimated. Analyses of performance parameters lead to conclusions regarding attributes for improved test discrimination capabilities. (orig.)

Best estimate approach for the evaluation of critical heat flux phenomenon in the boiling water reactors

Energy Technology Data Exchange (ETDEWEB)

Kaliatka, Tadas; Kaliatka, Algirdas; Uspuras, Eudenijus; Vaisnoras, Mindaugas [Lithuanian Energy Institute, Kaunas (Lithuania); Mochizuki, Hiroyasu; Rooijen, W.F.G. van [Fukui Univ. (Japan). Research Inst. of Nuclear Engineering

2017-05-15

Because of the uncertainties associated with the definition of Critical Heat Flux (CHF), the best estimate approach should be used. In this paper the application of best-estimate approach for the analysis of CHF phenomenon in the boiling water reactors is presented. At first, the nodalization of RBMK-1500, BWR-5 and ABWR fuel assemblies were developed using RELAP5 code. Using developed models the CHF and Critical Heat Flux Ratio (CHFR) for different types of reactors were evaluated. The calculation results of CHF were compared with the well-known experimental data for light water reactors. The uncertainty and sensitivity analysis of ABWR 8 x 8 fuel assembly CHFR calculation result was performed using the GRS (Germany) methodology with the SUSA tool. Finally, the values of Minimum Critical Power Ratio (MCPR) were calculated for RBMK-1500, BWR-5 and ABWR fuel assemblies. The paper demonstrate how, using the results of sensitivity analysis, to receive the MCPR values, which covers all uncertainties and remains best estimated.

Simulation of Thermal, Neutronic and Radiation Characteristics in Spent Nuclear Fuel and Radwaste Facilities

International Nuclear Information System (INIS)

Poskas, P.; Bartkus, G.

1999-01-01

The overview of the activities in the Division of Thermo hydro-mechanics related with the assessment of thermal, neutronic and radiation characteristics in spent nuclear fuel and radwaste facilities are performed. Also some new data about radiation characteristics of the RBMK-1500 spent nuclear fuel are presented. (author)

Questions about the reactor accident with Chernobyl-4

International Nuclear Information System (INIS)

Heijboer, R.J.

1986-01-01

The author presents an inventory of existing information about the Chernobyl-4 accident. Several possible scenarios are described and a comparison is drawn with the Three Mile Island-2 accident. The author concludes that the event is connected to an inherent instability of the RBMK-1000 reactor type. (G.J.P.)

Smolensk gets GOMIS

International Nuclear Information System (INIS)

Dynan, John; Francis, Arthur

1993-01-01

Improving safety at the Soviet designed RBMK reactors is not simply a matter of backfits. Effective management information systems and maintenance schedules will be needed, and not just for increased safety. A computer-based maintenance management information system is described, called GOMIS as are the introduction of various Nondestructive Testing programs. (Author)

Calculational-theoretical studies of the system of local automated regulators and lateral ionization chambers

International Nuclear Information System (INIS)

Aleksakov, A.N.; Emel'yanov, I.Ya.; Nikolaev, E.V.; Panin, V.M.; Podlazov, L.N.; Rogova, V.D.

1987-01-01

Methods of engineering synthesis of the systems for nuclear reactor local automated power regulation and radial-azimuthal energy distribution stabilization operating according to lateral ionization chamber signals are described. Results of calculational-theoretical investigations into the system efficiency and peculiarities of its reaction to some perturbations typical of the RBMK type reactors are considered

IAEA databases on safety issues and plant status for Central and Eastern European NPPs

International Nuclear Information System (INIS)

Czibolya, L.

1995-01-01

The main principles of the database development for safety issues of WWER and RBMK reactors are described. The IAEA contribution to the G-24 Project Data Bank and an analysis of gaps and overlaps in assistance projects is given. As an example the database on WWER-440/VV-230 type reactors is shown. (HP)

Identification of fast power reactivity effect in nuclear power reactor

International Nuclear Information System (INIS)

Efanov, A.I.; Kaminskas, V.A.; Lavrukhin, V.S.; Rimidis, A.P.; Yanitskene, D.Yu.

1987-01-01

A nuclear power reactor is an object of control with distributed parameters, characteristics of which vary during operation time. At the same time the reactor as the object of control has internal feedback circuits, which are formed as a result of the effects of fuel parameters and a coolant (pressure, temperature, steam content) on the reactor breeding properties. The problem of internal feedback circuit identification in a nuclear power reactor is considered. Conditions for a point reactor identification are obtained and algorithms of parametric identification are constructed. Examples of identification of fast power reactivity effect for the RBMK-1000 reactor are given. Results of experimental testing have shown that the developed method of fast power reactivity effect identification permits according to the data of normal operation to construct adaptive models for the point nuclear reactor, designed for its behaviour prediction in stationary and transition operational conditions. Therefore, the models considered can be used for creating control systems of nuclear power reactor thermal capacity (of RBMK type reactor, in particular) which can be adapted to the change in the internal feedback circuit characteristics

Characterization of Ignalina NPP RBMK Reactors Graphite

International Nuclear Information System (INIS)

Hacker, P.J.; Neighbour, G.B.; Levinskas, R.; Milcius, D.

2001-01-01

The paper concentrates on the investigations of the initial physical properties of graphite used in production of graphite bricks of Ignalina NPP. These graphite bricks are used as nuclear moderator and major core structural components. Graphite bulk density is calculated by mensuration, pore volumes are measured by investigation of helium gas penetration in graphite pore network, the Young's modulus is determined using an ultrasonic time of flight method, the coefficient of thermal expansion is determined using a Netzsch dilatometer 402C, the fractured and machined graphite surfaces are studied using SEM, impurities are investigated qualitatively by EDAX, the degree of graphitization of the material is tested using X-ray diffraction. (author)

Modernization for safety purposes of Russian nuclear power plants with channel-type reactors

International Nuclear Information System (INIS)

Riakhin, V.M.

1999-01-01

The nineties have crucially changed the Russian policy towards channel-type reactors known as RBMK. After the period of intensive commissioning the new Units (Kursk NPP: 1976, 1979, 1983,1985; Smolensk NPP 1982, 1985, 1990), the main financial flow was directed into reconstruction of these units. Safety upgrade of the units of Kursk NPP is presented in more details

Comparison of accident risks in different energy systems: Comments from Russian specialists

International Nuclear Information System (INIS)

2000-01-01

Many articles on accident risk analysis of different energy systems in comparison with nuclear power share certain stereotypical features. For example: When assessing the risks associated with the operation of such facilities, they ignore the effects of the upgrading of RBMK reactors which was carried out after the Chernobyl accident. In their integrated assessment of the radiological consequences of the Chernobyl accident they use numerous studies which frequently contain unreliable source data and unfounded predictions, and they ignore many socio-political factors which considerably increased the damage caused by the accident. Unfortunately, the study in question, despite its topicality and originality of approach, is also not without such shortcomings. After the Chernobyl accident, reconstruction and safety enhancement measures were implemented at nuclear power plants with RBMK reactors which were without precedent in world practice and have continued to this day. According to probabilistic safety assessments (PSA) carried out with the assistance of international experts, the probability of serious accidents at RBMKs has decreased by a factor of two or more thanks to the above mentioned measures. The mean weighted safety index for all operational RBMK reactors is 10 -4 l/year and is decreasing thanks to the ongoing and planned reconstruction of all units. All operational nuclear power plants with RBMK reactors are thus on a par with the successfully operating Soviet WWERs and western boiling water reactors (BWRs) and pressurized water reactors (PWRs), and satisfy the IAEA recommendations regarding the risk level of older generation nuclear power plants. The authors of the IAEA Bulletin article give estimates of the remote radiological consequences of the Chernobyl accident which range from an estimated 10,000 to 30,000 fatal cases of radiation-induced cancer, and the literature on the subject contains even more extreme estimates. However, our 14 years

Current state and perspectives of spent fuel storage in Russia

International Nuclear Information System (INIS)

Kurnosov, V.A.; Tichonov, N.S.; Makarchuk, T.F.

1999-01-01

Twenty-nine power units at nine nuclear power plants, having a total installed capacity of 22 GW(e), are now in operation in the Russian Federation. They produce approximately 12% of the generated electricity in the country. The annual spent fuel arising is approximately 790 tU. The concept of the closed fuel cycle was adopted as the basis for nuclear power development in the Russian Federation, but until now this concept is only implemented for the fuel cycles of WWER-440 and BN-600 reactors. The WWER-1000 spent fuel is planned to be reprocessed at the reprocessing plant RT-2 which is under construction near Krasnoyarsk. The RBMK-1000 spent fuel is not reprocessed. It is meant to be stored in intermediate storage facilities at the NPP sites. The status of the spent fuel (SF) stored in the storage facilities is given in the paper. The principal characteristics of the fuel cycles of the Russian NPPs in the period up to 2015 is also given in the report. The key variant of the current spent fuel management at RBMK-1000 NPPs is storage in at-reactor and in away-from-reactor wet storage facilities at the power plant site with a capacity of 2,000 W. The storage capacity at the operating RBMKs (including the increase due to denser fuel assembly arrangement) will provide SF reception from the NPPs only up to 2005. For RBMK spent fuel, intermediate dry storage is foreseen at power plant sites in metallic concrete casks and thereafter transportation to the central storage facility at the RT-2 plant for long-term storage. The SF will be reprocessing after completion of the reprocessing plant at RT-2. In the Programme of Nuclear Power Development in the Russian Federation for the period 1998 to 2005 and for the period until 2010 year, provisions are made for the construction of a central dry storage facility before 2010. The facility will have a design capacity of 30,000 tU for WWER-1000 and RBMK-1000 spent fuel and is part of the reprocessing plant RT-2. The paper considers

To selecting the characteristics of saturated steam direct cycle NPPs for operation under variable loads

International Nuclear Information System (INIS)

Khrustalev, V.A.; Demidov, O.I.

1986-01-01

Problems for operating process optimization of NPPs with RBMK type reactors under load swings in the power system is considered. Determination technique for optimal values of such parameters as initial steam pressure and fuel enrichment for NPP different load factors is developed. Optimization of these parameters gives a 150000 rouble saving of annual expenditures per each 3200 MW of reactor heat output

Effect of high-temperature filtration on impurity composition in the primary circuit coolant of power units with WWER-1000 reactors

International Nuclear Information System (INIS)

Efimov, A.A.; Moskvin, L.N.; Gusev, B.A.; Leont'ev, G.G.; Nekrest'yanov, S.N.

1992-01-01

The effects of high-temperature filtration on changes in dispersive, chemical, radioisotope and phase compositions of impurities in primary circuit coolant of NPP with the WWER-1000 reactor are studied. Special filters are used for the studies. The data obtained confirm the applicability of high-temperature filtration for purification of WWER reactor water and steam separators at NPPs with RBMK reactors

Safety analysis of reactor's cooling system

International Nuclear Information System (INIS)

1999-01-01

Results of the analysis of reactor's RBMK-1500 coolant system during normal operation mode, hydrodynamic testing and in the case of earthquake are presented. Analysis was performed using RELAP5 code. Calculations showed the most vulnerable place in the reactor's coolant system. It was found that in the case of earthquake the horizontal support system of drum separator could be damaged

Nuclear safety in eastern countries. Background of IPSN's actions

International Nuclear Information System (INIS)

1999-01-01

In this document, IPSN presents its opinion about the safety level that might be reached by the nuclear power plants situated in the former-USSR countries. In these countries 2 types of fission reactors are operating: VVER and RBMK with respectively 46 units and 14 units. 3 generations of VVER-type reactors are coexisting: 440 MWe-230, 440 MWe-213 and 1000 MWe-320. The first generation (440 MWe-230) which involves 11 operating units are the least safe and by no means is it possible to make them reach the western standard of safety. The second generation (440 MWe-213) require technical modifications to near western safety standards. The last generation (1000 MWe-320) has safety levels very similar to PWR's if operating procedures are modified and adapted. RBMK-type reactors have been designed in the years 60-70, they suffer from generic defects due to their design, the poor quality of materials and their low reliability. IPSN fears that any incident in such reactors might turn into a major accident. In order to improve nuclear safety in eastern countries, the European Union has launched an international cooperation, the programmes PHARE and TACIS are presented. (A.C.)

Review of the conclusions of the 1996 workshop on safety culture, in Forsmark, Sweden

Energy Technology Data Exchange (ETDEWEB)

Eckered, T [PROMENT LTD (Sweden)

1997-09-01

The IAEA/SiP Senior Managers Workshop on International Promotion of Safety Culture for the NPPs with RBMK reactors was organized by IAEA and the Swedish International Project Nuclear Safety (SiP). It took place at the Forsmark NPP, Sweden, from 1 to 4 October 1996. The objective of the workshop were to provide a forum managers to exchange national and international experience on factors influencing safety culture, to better understand these factors and to further enhance promotion of safety culture. The Workshop participants started work by agreeing to seek the answers to the following three questions: 1. What constitutes a good Safety Culture? 2. What is good and bad in our own countries and plants from a Safety Culture point of view? 3. Where can we find advice and help from our colleagues to improve our own Safety Culture? This was the first workshop specifically addressing Safety Culture in RBMK countries. The aim was therefore not to produce good practices, but to lay a foundation for further work and development. A follow-up workshop should deepen the understanding of the SC concept and address specific SC matters identified at this Workshop.

Extent estimation of different factors influence on the corrosion cracking of steels type X18H10T in NPP

International Nuclear Information System (INIS)

Stjazhkin, P.S.; Kritski, V.G.; Simanovski, V.M.; Kovalev, S.M.; Malov, M.Ju.; Butorin, S.L.

2002-01-01

1. Results of the performed complex study of chemical and physical factors effect on pipelines DU-300 integrity proves that in HT and going to power conditions under reactor RBMK-1000 operation there are most favourable conditions for origin and further undergrowth of crack in near-joint zones of pipelines and equipment made of steel X18H10T. These conditions are defined by high concentration of oxygen and electric conductivity of the coolant, higher in comparison with NOC (operation at power) stressed-and-strained state of units being loaded with additional dynamic loads. 2. Results received are of an evaluating character, they show only that under start-up operations after thermal shutdown an accelerated growth of cracks is most probable. 3. In view of a general trend of prolongation of operation terms of NPP with RBMK-1000 (LNPP including), realization of procedures to decrease probability of occurrence and evolution of CCS in equipment and pipelines and introduction of a system of corrosion monitoring is one of the actual tasks. 4. An additional measure can be a validated transition to hydro-pressing of the circuit at lower temperature (80-90 C). (authors)

Probabilistic safety assessment activities at Ignalina NPP

International Nuclear Information System (INIS)

Bagdonas, A.

1999-01-01

The Barselina Project was initiated in the summer 1991. The project was a multilateral co-operation between Lithuania, Russia and Sweden up until phase 3, and phase 4 has been performed as a bilateral between Lithuania and Sweden. The long-range objective is to establish common perspectives and unified bases for assessment of severe accident risks and needs for remedial measures for the RBMK reactors. During phase 3, from 1993 to 1994, a full scope Probabilistic Safety Analysis (PSA) model of the Ignalina Nuclear Power Plant unit 2 was developed to identify possible safety improvement of risk importance. The probabilistic methodology was applied on a plant specific basis for a channel type reactor of RBMK design. During phase 4, from 1994 to 1996, the PSA was further developed, taking into account plant changes, improved modelling methods and extended plant information concerning dependencies (area events, dynamic effects, electrical and signal dependencies). The model reflected the plant status before the outage 1996. During phase 4+, 1998 to 1999 the PSA model was upgraded taking into account the newest plant modifications. The new PSA model of CPS/AZRT was developed. Modelling was based on the Single Failure Analysis

The safety of nuclear power plants in Eastern Europe

International Nuclear Information System (INIS)

Hoehn, J.; Niehaus, F.

1997-01-01

Nuclear power plant operators and nuclear organizations from the West and from the East cooperate at many levels. The G7 and G24 nations have taken it upon themselves to improve the safety of Eastern nuclear power plants. The European Union has launched support programs, i.e. Technical Assistance to the Commonwealth of Independent States (Tacis) and Pologne-Hangrie: Aide a la Reconstruction Economique (Phare), and founded the European Bank for Reconstruction and Development. The countries of Central and Eastern Europe operate nuclear power plants equipped with VVER-type pressurized water reactors and those equipped with RBMK-type reactors. The safety of these two types of plants is judged very differently. Among the VVER plants, a distinction is made between the older and the more recent 440 MWe lines and the 1000 MWe line. Especially the RBMK plants (Chernobyl-type plants) differ greatly as a function of location and year of construction. Even though they do not meet Western safety standards and at best can be backfitted up to a certain level, it must yet be assumed that they will remain in operation to the end of their projected service lives for economic reasons. (orig.) [de

Analysis of Elektrogorsk 108 test facility experimental data

International Nuclear Information System (INIS)

Urbonas, R.

2001-01-01

In the paper an evaluation of experimental data obtained at Russian Elektrogorsk 108 (E-108) test facility is presented. E-108 facility is a scaled model of Russian RBMK design reactor. An attempt to validate state-of-the-art thermal hydraulic codes on the basis of E-108 test facility was made. Originally these codes were developed and validated for BWRs and PWRs. Since state-of-art thermal hydraulic codes are widely used for simulation of RBMK reactors further codes' implementation and validation is required. The facility was modelled by employing RELAP5 (INEEL, USA) thermal hydraulic system analysis best estimate code. The results show dependence from number of nodes used in the heated channels, frictional and form losses employed. The obtained oscillatory behaviour is resulted by density wave and critical heat flux. It is shown that codes are able to predict thermal hydraulic instability and sudden heat structure temperature excursion, when critical heat flux is approached, well. In addition, an uncertainty analysis of one of the experiments was performed by employing GRS developed System for Uncertainty and Sensitivity Analysis (SUSA). It was one of the first attempts to use this statistic-based methodology in Lithuania.(author)

Review of the conclusions of the 1996 workshop on safety culture, in Forsmark, Sweden

International Nuclear Information System (INIS)

Eckered, T.

1997-01-01

The IAEA/SiP Senior Managers Workshop on International Promotion of Safety Culture for the NPPs with RBMK reactors was organized by IAEA and the Swedish International Project Nuclear Safety (SiP). It took place at the Forsmark NPP, Sweden, from 1 to 4 October 1996. The objective of the workshop were to provide a forum managers to exchange national and international experience on factors influencing safety culture, to better understand these factors and to further enhance promotion of safety culture. The Workshop participants started work by agreeing to seek the answers to the following three questions: 1. What constitutes a good Safety Culture? 2. What is good and bad in our own countries and plants from a Safety Culture point of view? 3. Where can we find advice and help from our colleagues to improve our own Safety Culture? This was the first workshop specifically addressing Safety Culture in RBMK countries. The aim was therefore not to produce good practices, but to lay a foundation for further work and development. A follow-up workshop should deepen the understanding of the SC concept and address specific SC matters identified at this Workshop

Dealing with local and international media - experience of Lithuania

International Nuclear Information System (INIS)

Bieliauskas, V.

2000-01-01

Public acceptance is one of most serious conditions for nuclear power introduction or successful operation of nuclear power plants. After restoring it's independence, Lithuania inherited large nuclear power plant RBMK type reactors widely known as Chernobyl type reactors. Several small case studies of dealing with the media on nuclear matters are presented and efforts to adopt an active attitude on different levels are described. (author)

IAEA Newsbriefs. V. 11, no. 2(71). Apr-May 1996

International Nuclear Information System (INIS)

1996-01-01

This issue gives brief information on the following topics: Chernobyl Conference sums up Scientific Understanding, International Forum on the Safety of RBMK Reactors, IAEA Board of Governors, Radiological Study of the Mururoa and Fangataufa Atolls, Safeguards Support, Workshop on Modelling Methods for Water Systems, Uranium Market Trends, Security Council Resolution on Iraq, South East Asian NWFZ, Radioactive Waste Management, Nuclear Seminar in Poland, and other short information

Top-Level Software for VVER-1000 In-core Monitoring System under Implementation of Expanded Nuclear Fuel Diversification Program in Ukraine

International Nuclear Information System (INIS)

Khalimonchuk, V.A.

2015-01-01

The paper considers the possibility and expediency of developing mathematical software for VVER-1000 ICMS in Ukraine. This mathematical software is among the most important conditions for implementation of the expanded nuclear fuel diversification program. The top-level software is to be developed based on SSTC own studies in the development of codes for power distribution recovery, which were successfully used previously for RBMK-1000 safety analysis

Accident at the Chernobyl AES and its consequences. Data prepared for the International Atomic Energy Agency Expert Conference (25-29 August 1986, Vienna)

International Nuclear Information System (INIS)

1986-01-01

This report on the accident at the Chernobyl nuclear power plant describes the plant and associated RBMK-1000 reactors and gives a chronology of the development of the accident. The causes of the accident are discussed as well as an analysis of the process of development of the accident. Also discussed are measures adopted to increase power plant safety and prevent development of similar accidents

German safety engineering in comparison to the Chernobyl reactor; Deutsche Sicherheitstechnik im Vergleich zum Tschernobyl-Reaktor

Energy Technology Data Exchange (ETDEWEB)

NONE

2014-04-15

In Germany the Russian RBMK would not have been licensed, since the safety standards in Germany were superior to those in the former USSR. The most significant differences - the containment, the control rod system, the void coefficient and the emergency cooling system are shortly summarized. The consequences for the population are cumulative environmental radiation exposure are reported using the official data of IAEA, WHO and UNDP.

The safety of RMBK reactors 10 years after Chernobyl

International Nuclear Information System (INIS)

Lederman, L.

1996-01-01

In April 1986 the Unit 4 of Chernobyl NPP was destroyed in the worst accident in history of commercial nuclear power. Unit 4 started operation in 1983 and was a RBMK nuclear power plant (NPP). Over the years, three generations of reactors have emerged which have significant differences, particularly with respect to the safety provisions built into their design. The electric power of the RBMK reactors is 1000 MW(e) except for Ignalina whose power is 1500 MW(e). development of the Kursk Unit 5, currently under construction, has led to many design changes hence it can be thought of as a fourth generation. The first generation units (Leningrad-l and -2, Kursk-1 and Chernobyl-l and -2) designed and built before 1982 when new standards on the design and construction of Nuclear Power Plants (NPPs) OPB-82 were introduced in the Soviet Union. Since then other units have designed and constructed in accordance to these requirements. The safety standards in the U were revised again in 1988 (OPB-88). Since the Chernobyl accident a considerable amount of work has been carried out by Ru designers and PTSMK operators to improve RBMK reactor safety and to eliminate the causes o accident. As a result, major design modifications and operational changes have been implemented. However, safety concerns remain, particularly related to first generation units. In the framework of a Programme on PTSMK safety initiated by the IAEA in 1992, a total of 58 safety issues related to seven topical areas were identified. The issues related to the six design areas were further ranked according to their perceived impact on plant safety. Safety issues connected to operational areas, particularly those related to ensuring that a high safety culture is an underlying basis for operation, were considered very important. It was stressed that all efforts should be made to implement the related recommendations along with d modifications (author)

The safety of RMBK reactors 10 years after Chernobyl

Energy Technology Data Exchange (ETDEWEB)

Lederman, L [International Atomic Energy Agency, Vienna (Austria)

1996-12-01

In April 1986 the Unit 4 of Chernobyl NPP was destroyed in the worst accident in history of commercial nuclear power. Unit 4 started operation in 1983 and was a RBMK nuclear power plant (NPP). Over the years, three generations of reactors have emerged which have significant differences, particularly with respect to the safety provisions built into their design. The electric power of the RBMK reactors is 1000 MW(e) except for Ignalina whose power is 1500 MW(e). development of the Kursk Unit 5, currently under construction, has led to many design changes hence it can be thought of as a fourth generation. The first generation units (Leningrad-l and -2, Kursk-1 and Chernobyl-l and -2) designed and built before 1982 when new standards on the design and construction of Nuclear Power Plants (NPPs) OPB-82 were introduced in the Soviet Union. Since then other units have designed and constructed in accordance to these requirements. The safety standards in the U were revised again in 1988 (OPB-88). Since the Chernobyl accident a considerable amount of work has been carried out by Ru designers and PTSMK operators to improve RBMK reactor safety and to eliminate the causes o accident. As a result, major design modifications and operational changes have been implemented. However, safety concerns remain, particularly related to first generation units. In the framework of a Programme on PTSMK safety initiated by the IAEA in 1992, a total of 58 safety issues related to seven topical areas were identified. The issues related to the six design areas were further ranked according to their perceived impact on plant safety. Safety issues connected to operational areas, particularly those related to ensuring that a high safety culture is an underlying basis for operation, were considered very important. It was stressed that all efforts should be made to implement the related recommendations along with d modifications (author).

International conference on the strengthening of nuclear safety in Eastern Europe. Keynote papers. Regulatory aspects of NPP safety, status of safety improvements, status of safety analysis report

International Nuclear Information System (INIS)

1999-06-01

The Objective of the Conference was to assess the past decade of nuclear safety efforts in countries operating WWER and RBMK nuclear reactors and to address remaining safety issues which require further work. A particular focus of the Conference was on international co-operation and assistance and where such efforts should be focused in the future. All Eastern European countries that operate RBMK or WWER reactors participated in the Conference, and presented papers on three key areas of nuclear safety: Regulatory Aspects of Nuclear Power Plant Safety; Status of Safety Improvements; and Status of Safety Analysis Reports. In addition, representatives from 18 additional countries that provide financial and/or technical assistance and co-operation in the area of WWER and RBMK safety offered the most extensive commentary. Key international (IAEA, World Association of Nuclear Operators, the Nuclear Energy Agency, the G-24 NUSAC, the European Commission, and the EBRD) organizations that provide nuclear safety assistance for WWER and RBMK reactors also made presentations. There is no question that considerable progress on nuclear safety has been made in Eastern Europe. Special mention should be made of successful efforts to strengthen the independence and technical competence of the nuclear regulatory authorities. Efforts should now concentrate on improving the depth and scope of the technical abilities of the regulatory authorities. More attention by governments is needed to ensure that the regulatory authorities have the financial resources and enforcement authority to fully execute their missions. In respect to the operators of the nuclear power plants, they have demonstrated clear progress in operational safety improvements. Significant additional efforts are required to maintain and enhance an effective safety culture. Design safety improvement programmes are in place in all countries. Implementation of these programmes has varied and is particularly affected by

Chernobyl: before and after. Information sheet No. 1

Energy Technology Data Exchange (ETDEWEB)

Devine, J [comp.

1986-01-01

Complied in June 1986 the information sheet lists title, authors, and journal details and gives brief details of all relevant published articles. Part 1 concerns material relevant before the accident in April 1986 - the construction, design and safety of the Chernobyl RBMK nuclear power station (20 references). Part 2 lists articles published after the accident concerning the impact of the disaster on safety in the nuclear power industry (23 references).

Chernobyl, 17 after

International Nuclear Information System (INIS)

2003-04-01

This information document takes stock on the Chernobyl accident effects, 17 years after the reactor accident. The domains concerned are: the Chernobyl power plant, the sanitary consequences of the accident in the most exposed countries, the Chernobyl environment and the polluted regions management, the Chernobyl accident consequences in France; Some data and technical sheets on the RBMK reactors and the international cooperation are also provided. (A.L.B.)

Analysis of the oscillation causes in automatic controller of reactor power

International Nuclear Information System (INIS)

Aleksakov, A.N.; Nikolaev, E.V.; Podlazov, L.N.

1991-01-01

Conditions for occurence of oscillations in automatic controller of reactor power are determined. Graphic-analytical method for calculating the stability of non-linear system, which enables one to reveal the most important factors determining the stability, is used. The practical results of the analysis are obtained for the system of local automatic comtrollers, used in the RBMK reactors. A simple method providing for the required stability margin, is suggested

The role of the shot effect in the accumulation of nuclide masses and activities in the thermal reactor core

International Nuclear Information System (INIS)

Nepoleao, P.; Rudak, E.; Wiley, J.

1998-01-01

A method is proposed for estimating masses and activities of nuclides in the thermal reactor core with an arbitrary dependence of specific masses on the burnout depth. The method takes into account the statistical character of micro processes accompanying the fuel burning out and accumulation of fission and activation products. For the RBMK reactor of Chernobyl NPP the method gives practically the same results as exact numerical calculations. (author)

Ignalina NPP Safety Analysis: Models and Results

International Nuclear Information System (INIS)

Uspuras, E.

1999-01-01

Research directions, linked to safety assessment of the Ignalina NPP, of the scientific safety analysis group are presented: Thermal-hydraulic analysis of accidents and operational transients; Thermal-hydraulic assessment of Ignalina NPP Accident Localization System and other compartments; Structural analysis of plant components, piping and other parts of Main Circulation Circuit; Assessment of RBMK-1500 reactor core and other. Models and main works carried out last year are described. (author)

Bootstrap and Order Statistics for Quantifying Thermal-Hydraulic Code Uncertainties in the Estimation of Safety Margins

Directory of Open Access Journals (Sweden)

Enrico Zio

2008-01-01

Full Text Available In the present work, the uncertainties affecting the safety margins estimated from thermal-hydraulic code calculations are captured quantitatively by resorting to the order statistics and the bootstrap technique. The proposed framework of analysis is applied to the estimation of the safety margin, with its confidence interval, of the maximum fuel cladding temperature reached during a complete group distribution blockage scenario in a RBMK-1500 nuclear reactor.

International conference on the strengthening of nuclear safety in Eastern Europe. Keynote papers. Regulatory aspects of NPP safety, status of safety improvements, status of safety analysis report

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-06-01

The Objective of the Conference was to assess the past decade of nuclear safety efforts in countries operating WWER and RBMK nuclear reactors and to address remaining safety issues which require further work. A particular focus of the Conference was on international co-operation and assistance and where such efforts should be focused in the future. All Eastern European countries that operate RBMK or WWER reactors participated in the Conference, and presented papers on three key areas of nuclear safety: Regulatory Aspects of Nuclear Power Plant Safety; Status of Safety Improvements; and Status of Safety Analysis Reports. In addition, representatives from 18 additional countries that provide financial and/or technical assistance and co-operation in the area of WWER and RBMK safety offered the most extensive commentary. Key international (IAEA, World Association of Nuclear Operators, the Nuclear Energy Agency, the G-24 NUSAC, the European Commission, and the EBRD) organizations that provide nuclear safety assistance for WWER and RBMK reactors also made presentations. There is no question that considerable progress on nuclear safety has been made in Eastern Europe. Special mention should be made of successful efforts to strengthen the independence and technical competence of the nuclear regulatory authorities. Efforts should now concentrate on improving the depth and scope of the technical abilities of the regulatory authorities. More attention by governments is needed to ensure that the regulatory authorities have the financial resources and enforcement authority to fully execute their missions. In respect to the operators of the nuclear power plants, they have demonstrated clear progress in operational safety improvements. Significant additional efforts are required to maintain and enhance an effective safety culture. Design safety improvement programmes are in place in all countries. Implementation of these programmes has varied and is particularly affected by

Analysis of NPP pipes and equipment damage in life time prolongation

International Nuclear Information System (INIS)

Tkachev, V.V.; Zheltukhin, K.K.

2008-01-01

Paper describes a procedure to calculate the probability of pipes and equipment failure taking account of both the service records of the structures under various conditions and their aging. The parameters characterizing applied loads, failures, as well as metal strength, mechanical and thermal properties serve as the arbitrary values used in the described procedure. Paper presents an example of the probability calculation of failure of the RBMK emergency feed pump recirculation pipes when their service life is prolonged [ru

Ignalina RBMK-1500 building capability in retaining radioactive releases

International Nuclear Information System (INIS)

Nilsson, Lars; Johansson, K.

1993-01-01

The Ignalina reactor building structures are capable of retaining substantial fractions of radioactive emissions from the fuel core, in those accident sequences where pressurization failure of structures can be averted by pressure relief arrangements. In stage 1 of the IBBA project it was demonstrated that enhanced retention of radioactive fission products within the plant can be achieved if natural convection is facilitated in the upper building compartments. In this report of stage 2 is discussed for which accident sequences the introduction of natural convection in combination with the existing forced convection ventilation and the accident localization system can improve the total safety of Ignalina 1-2. The purpose of this stage is to provide a basis for further review and more detailed studies of the natural convection concept, its benefits and disadvantages, and of the feasability to introduce the concept in existing plants

Problems and experience of ensuring nuclear safety in NPP spent fuel storage facilities in Russia

International Nuclear Information System (INIS)

Vnukov, Victor S.; Ryazanov, Boris G.

2003-01-01

The amount of Nuclear Power Plant (NPP) spent fuel in special storage facilities of Russia runs to more than 15000 tons and the annual growth is equal to about 850 tons. The storage facilities for spent nuclear fuel from the main nuclear reactors of Russia (RBMK-1000, VVER-1000, BN-600, EGP-6) were designed in the 60s - 70s. In the last years when the concept of closed fuel cycle and safety requirements had changed, the need was generated to have the nuclear storage facilities more crowded. First of all it is due to the necessity to increase the storage capacity because the RBMK-1000, VVER-1000, EGP-6 fuel is not reprocessed. So there comes the need for the facilities of a bigger capacity which meet the current safety requirements. The paper presents the results of studies of the most important nuclear safety issues, in particular: development of regulatory requirements; analysis of design-basis and beyond-the design-basis accidents (DBA and BDBA); computation code development and verification; justification of nuclear safety when water density goes down; the use of burn-up fraction values; the necessity and possibility to experimentally study the storage facility subcriticality; development of storage norms and rules for new types of fuel assemblies with mixed fuel and burnable poison. (author)

Improvements of MMI and operator support systems at the Leningrad NPP

International Nuclear Information System (INIS)

Rakitin, I.D.; Malkin, S.D.; Shalia, V.V.; Fedorov, E.M.; Koudiakov, M.M.; Stebenev, A.S.

1998-01-01

A practical need of MMI up-grade and inclusion of new Operator Support Systems is of utmost importance for the existing NPPs under the new safety related Russian and International demands, requirements and regulations. The given paper describes RandD work for RBMK-type reactors with using full scope simulator features. But its main results could be well implemented for other reactor types as well. Significant efforts to up-grade safety of RBMK and implement a set of additional Safety Support Systems are provided by the Russian Project Design Institutes and by the International Organizations and Communities. But these projects have been mostly developed without a proper verification and validation against the real plant operation modes and real Control Room circumstances, and with no justification of an operating crew demands and expectations. That unfavorable situation should be successfully changed by now with using the Training Support Center (TSC) created at the Leningrad NPP. It incorporates the full-scope and analytical simulators working in parallel with the prototypes of the expert and interactive systems to provide a new scope of RandD work. The development and adjustment of two state-of-the-art Operators' Support Systems with using the Simulators are described in the Paper. These systems have been developed by the joint RRC KI and LNPP team. (author)

Chernobyl lesson

Energy Technology Data Exchange (ETDEWEB)

Vajda, G

1986-01-01

Structure and major technological parameters of the RBMK-1000 type Chernobylsk reactor, description of different phases of the reactor accident, the causes and consequences of the catastrophe and the measures taken to cease the fire, to stop the chain reaction, to prevent the inhabitants and the environment from radiation exposure and contamination are discussed. Major development projects at the Paks Nuclear Power Plant to support human control activities and to increase the operational safety are listed. (V.N.). 2 refs.

Interim report on fallout situation in Finland from April 26 to May 4 1986

International Nuclear Information System (INIS)

1986-05-01

As known, a reactor unit of type RBMK 1000 ignited on April 26, 1986, approximately at 5 am Finnish time in the reactor site in Chernobyl, some 130 kilometers from the city of Kiev in the USSR. The Soviet authorities gave an official announcement saying that the reactor fire had ended on May 5 and the reaction had stopped. This report presents results from external radiation measurements and analysis of environmental samples in Finland from April 26 to May 4, 1986. (L.K)

State of stress and strain and vibration resistance of manifolds and tube assemblies of the RMBK-1000 plant exemplified by Leningrad, Kursk and Chernobyl nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Shul' tsev, D N; Egorov, M F; Kw' min, Yu S; Sidorov, A A

1982-05-01

The design calculations of RBMK-1000 circulation loop manifolds and tube assemblies are summed up. It was found that the most stressed places of the multiple forced circulation loop were the pump intake branch connection and the contact section between the elbow and the horizontal part of the pressure head part before the gate valve. Vibration resistance appears to be generally adequate. Calculations were made in accordance with an M-222 computer program realizing an algorithm based on the Castiglians principle for three-dimensional rod systems.

Application of the low disturbances theory in operation calculations of the BMK reactor

International Nuclear Information System (INIS)

Isaev, N.V.; Shmonin, Yu.V.; Pogosbekyan, L.R.; Druzhinin, V.E.

1985-01-01

Calculation algorithm of direct and contingent tasks in a two-group diffusion approximation for RBMK-1000 of Smolensk-1 nuclear power plant is presented. Examples of numeric calculation of the reactivity change caused by neutron field disturbance reactivity effect in case of refueling, refueling, rate and reactivity reserve on control rods are given. Calculations are made according to PEPO-program. The program is written in FORTRAN-4 for ES computer. The modificated low disturbances theory used in this program allows to reduce sufficiently the calculation error

The accident of Chernobyl

International Nuclear Information System (INIS)

1986-10-01

RBMK reactors (reactor control, protection systems, containment) and the nuclear power plant of Chernobyl are first presented. The scenario of the accident is given with a detailed chronology. The actions and consequences on the site are reviewed. This report then give the results of the source term estimation (fision product release, core inventory, trajectories, meteorological data...), the radioactivity measurements obtained in France. Health consequences for the French population are evoked. The medical consequences for the population who have received a high level of doses are reviewed [fr

Development of the international Chornobyl center in Slavutych its role in nuclear facilities decommissioning

International Nuclear Information System (INIS)

Nosovskij, A.

2002-01-01

ChNPP site should be actively used for development and realizing the efforts on RBMK reactors decommissioning technologies, as well as Unit 'Shelter' transformation into ecologically safe system. Those technologies and methods will be required in future both in Ukraine and abroad. So, it is wise to employ another direction of ChNPP site utilization as a polygon for development and implementing new technologies associated with decommissioning. Besides, it could be used for exercising the methods and technologies of large-scale man-caused accidents elimination

Influence of metallurgical variables on the velocity of crack propagation by delayed hydride cracking (DHC) in Zr-Nb

International Nuclear Information System (INIS)

Cirimelo, Pablo G.

2002-01-01

In the present thesis work the propagation of cracks due to the delayed hydride cracking (DHC) mechanism in Zr-2,5 % Nb pressure tubes is analyzed. For this purpose two different type of tubes of different origin were used: CANDU type (Canada) and RBMK type (Russia). The analyzed figurative parameters were: critical temperature Tc (highest temperature at which DHC phenomenon could occur) and crack propagation velocity by DHC, Vp, in the axial direction. The influence of the memory effect (phenomenon proper of hydride precipitation) was studied, as well as the type of cracks (fatigue or DHC) on Tc. However, no influence of these effects was found. Instead, it was found that Tc varies with the hydrogen content of the specimen, in agreement with previous works. Samples obtained from tubes with different microstructures and similar amounts of hydrogen presented similar Tc values. It was also shown that DHC propagation could occur without precipitated hydrides in the volume. Besides, Vp determinations were performed in temperature ranges and hydrogen amounts of technological importance. Two techniques were set up in order to determine Vp at different temperatures in a single specimen, thus saving time and material. An Arrhenius type variation was found for Vp vs. temperature, for temperatures lower than that corresponding to precipitation. For higher temperatures, but lower than the critical one, velocity decreases with temperature. Determination of Vp vs. temperature was performed for the two above-mentioned materials, whose microstructure and hardness were previously characterized. For RBMK material, which presents a spheroidal β phase, the velocity was lower than the corresponding to CANDU material, in which β phase is formed by continuous plates. In addition, yield stress σ Y is lower in RBMK material, which presents lower Vp. However, it is considered that the effect of microstructure is more important on Vp since it highly affects diffusion of hydrogen from the

AGR core models and their application to HTRs and RBMKs

International Nuclear Information System (INIS)

Baylis, Samuel

2014-01-01

EDF Energy operates 14 AGRs, commissioned between 1976 and 1989. The graphite moderators of these gas cooled reactors are subjected to a number of ageing processes under fast neutron irradiation in a high temperature CO2 environment. As the graphite ages, continued safe operation requires an advanced whole-core modeling capability to enable accurate assessments of the core’s ability to fulfil fundamental nuclear safety requirements. This is also essential in evaluating the reactor's remaining economic lifetime, and similar assessments are useful for HTRs in the design stage. A number of computational and physical models of AGR graphite cores have been developed or are in development, allowing simulation of the reactors in normal, fault and seismic conditions. Many of the techniques developed are applicable to other graphite moderated reactors. Modeling of the RBMK allows validation against a core in a more advanced state of ageing than the AGRs, while there is also an opportunity to adapt the models for high temperature reactors. As an example, a finite element model of the HTR-PM side reflector based on rigid bodies and nonlinear springs is developed, allowing rapid assessments of distortion in the structure to be made. A model of the RBMK moderator has also been produced using an established AGR code based on similar methods. In addition, this paper discusses the limitations of these techniques and the development of more complex core models that address these limitations, along with the lessons that can be applied to HTRs. (author)

Solid radioactive waste processing facility of the NPP Leningrad

International Nuclear Information System (INIS)

Weichard, Swetlana

2008-01-01

On behalf of the Russian Company Rosenergoatom NUKEM Technologies GmbH is planning and constructing a complete facility for the processing of solid low- and medium-active radioactive wastes. The NPP Leningrad comprises 4 units of RBMK-1000 reactors, the plant life has been extended by 15 years, the first unit is to be decommissioned in 2018. The construction of four new units is planned. NUKEM is in charge of planning, manufacture, construction and startup of the following facilities: sorting, internal transport, combustion and waste gas cleaning, emission surveillance, compacting, packaging and radiological measurement.

Skoda JS in 2016

International Nuclear Information System (INIS)

Perlik, Josef

2016-01-01

The presentation describes the main areas of activities of the company, which are: Engineering: Construction of WWER rectors; Modernization and reconstruction of units with WWER (including replacement of control systems); spent fuel storage facilities; Construction of research and training reactors; Production: Equipment for nuclear power plants with WWER and RBMK reactors; Equipment for nuclear power plants with reactor type PWR and BWR; The equipment for storage of spent nuclear fuel; Equipment for the petrochemical industry projects Service: Planned maintenance management; Maintenance and repair of the reactor equipment; Upgrading reactor components; life management for equipment

The concepts of leak before break and absolute reliability of NPP equipment and piping

International Nuclear Information System (INIS)

Getman, A.F.; Komarov, O.V.; Sokov, L.M.

1997-01-01

This paper describes the absolute reliability (AR) concept for ensuring safe operation of nuclear plant equipment and piping. The AR of a pipeline or component is defined as the level of reliability when the probability of an instantaneous double-ended break is near zero. AR analysis has been applied to Russian RBMK and VVER type reactors. It is proposed that analyses required for application of the leak before break concept should be included in AR implementation. The basic principles, methods, and approaches that provide the basis for implementing the AR concept are described

The Ninth International scientific and technical conference Safety, efficiency and economy of atomic energy. Book of abstracts

International Nuclear Information System (INIS)

2014-01-01

The abstracts of the Ninth International scientific and technical conference Safety, efficiency and economy of atomic energy are present. The conference took place in Moscow, 21-23 May, 2014. The problems of WWER, RBMK, BN and EhGP-6 NPPs operation, maintenance and repair; materials testing and metallic structures control; radioactive wastes and spent fuel management; NPP decommissioning; radiation safety, NPP ecology, emergency preparedness were discussed on the conference. The great attention was paid to the problems of atomic energy economy and its developing, international cooperation for NPP safety and young NPP specialists training [ru

Effect of irradiation on corrosion of low-activation austenite Cr-Mn steel in technological liquid mediums of nuclear power plant

International Nuclear Information System (INIS)

Demina, E.V.; Prusakova, M.D.; Vinogradova, N.A.; Orlova, G.D.; Nechaev, A.F.; Doil'nitsyn, V.A.

2008-01-01

Effect of γ-radiation on corrosion rate in cold-worked and annealed low-activation austenite 12Cr-20Mn steel has been studied. Corrosion tests were carried out in water solutions which simulate the coolant medium in the primary coolant circuit of WWER power reactor and in the circuit of multiple forced circulation of RBMK-1000 reactor as well as an aquatic environment in cooling pond for spent fuel. The worst radiation effect was observed in the cooling pond environment where the value of corrosion rate is increased by tens or hundreds times

The concepts of leak before break and absolute reliability of NPP equipment and piping

Energy Technology Data Exchange (ETDEWEB)

Getman, A.F.; Komarov, O.V.; Sokov, L.M. [and others

1997-04-01

This paper describes the absolute reliability (AR) concept for ensuring safe operation of nuclear plant equipment and piping. The AR of a pipeline or component is defined as the level of reliability when the probability of an instantaneous double-ended break is near zero. AR analysis has been applied to Russian RBMK and VVER type reactors. It is proposed that analyses required for application of the leak before break concept should be included in AR implementation. The basic principles, methods, and approaches that provide the basis for implementing the AR concept are described.

Reasons for the RBMK reactor accident at the Chernobyl NPP

International Nuclear Information System (INIS)

Novikov, I.I.; Kruzhilin, G.N.; Anan'ev, E.P.

1995-01-01

This analysis of the reasons for the Chernobyl Reactor accident in 1986 places the blame firmly with the reactor operators, who, it is argued, made a number of dramatic mistakes while controlling the reactor. The report also included an additional analysis of the causes of the accident. (UK)

The 1986 Chernobyl accident; Der Unfall von Tschernobyl 1986

Energy Technology Data Exchange (ETDEWEB)

Kerner, Alexander; Stueck, Reinhard; Weiss, Frank-Peter [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Garching bei Muenchen, Koeln (Germany). Bereich Reaktorsicherheitsanalysen; Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Koeln (Germany)

2011-02-15

April 26, 2011 marks the 25th anniversary of the Chernobyl reactor accident, the worst incident in the history of the peaceful utilization of nuclear power. While investigations of the course of events and the causes of the accident largely present a uniform picture, descriptions still vary widely when it comes to the impact on the population and the environment. This treatment of the Chernobyl accident constitutes a summary of facts about the initiation of the accident and the sequence of events that followed. In addition, measures are described which were taken to exclude any repetition of a disaster of this kind. The health consequences and the socio-economic impact of the accident are not discussed in any detail. The first section contains an introduction and an overview of the Soviet RBMK (Chernobyl) reactor line. In section 2, fundamental characteristics of this special type of reactor, which was exclusively built in the former Soviet Union, are discussed. This information is necessary to understand the sequence of accident events and provides an answer to the frequent question whether that accident could be transferred to reactors in this country. The third section outlines the history of the accident caused ultimately by a commissioning test never performed before. The section is completed by a brief description of radiological releases and the state of the plant after the accident when entombed in the ''sarcophagus.'' The different causes are then summarized and the modifications afterwards made to RBMK reactors are outlined. (orig.)

The modeling experience of fuel element units operation under MSC.MARC and MENTAT 2008R1

International Nuclear Information System (INIS)

Kulakov, G.; Kashirin, B.; Kosaurov, A.; Konovalov, Y.; Kuznetsov, A.; Medvedev, A.; Novikov, V.; Vatulin, A.

2009-01-01

MSC Software is leading developer of CAE-software in the world, so behaviour of fuel elements modeling with MSC.MARC use is of great practical importance. Behaviour of fuel elements usually is modeled in the elastic-viscous-plastic statement with account on fuel swelling during irradiation. For container type fuel elements contact interaction between fuel pellets and cladding or other parts of fuel element in top and bottom plugs must be in account. Results of simulated behaviour of various type fuel elements - container type fuel elements for PWR and RBMK reactors, dispersion type fuel elements for research reactors are presented. (authors)

Corrosion strength monitoring of NPP component residual lifetime

International Nuclear Information System (INIS)

Denisov, V.G.; Belous, V.N.; Arzhaev, A.I.; Shuvalov, V.A.

1994-01-01

Importance of corrosion and fatigue monitoring; types of corrosion determine the NPP equipment life; why automated on-line corrosion and fatigue monitoring is preferable; major stages of lifetime monitoring system development; major groups of sensors for corrosion and strength monitoring system; high temperature on-line monitoring of water chemistry and corrosion; the RBMK-1000 NPP unit automatic water chemistry and corrosion monitoring scheme; examples of pitting, crevice and general corrosion forecast calculations on the basis of corrosion monitoring data; scheme of an experimental facility for water chemistry and corrosion monitoring sensor testing. 2 figs., 4 tabs

Safety parameter display system (SPDS) for Russian-designed NPPs

International Nuclear Information System (INIS)

Anikanov, S.S.; Catullo, W.J.; Pelusi, J.L.

1997-01-01

As part of the programs aimed at improving the safety of Russian-designed reactors, the US DoE has sponsored a project of providing a safety parameter display system (SPDS) for nuclear power plants with such reactors. The present paper is focused mostly on the system architecture design features of SPDS systems for WWER-1000 and RBMK-1000 reactors. The function and the operating modes of the SPDS are outlined, and a description of the display system is given. The system architecture and system design of both an integrated and a stand-alone IandC system is explained. (A.K.)

Ventilation of nuclear power plants

International Nuclear Information System (INIS)

Madoyan, A.A.; Vlasik, V.F.

1984-01-01

Foundations and calculation methods of ventilation of rooms with different degree of heat and gas release with the change of operation mode of NPP main equipment, as well as problems of NPP site and adjoining area aerodynamics, have been presented. Systems of air ventilation and conditioning, cooling equipment, are considered. The main points of designing are described and determination of economic efficiency of the ventilation systems are made. Technical characteristics of the ventilators, conditioners, filters and air heaters used, are presented. Organization of adjustment, tests, operation and maintenance of the ventilation systems of NPP with RBMK and WWER-type reactors, is described

RBMK-1500 accident management for loss of long-term core cooling

International Nuclear Information System (INIS)

Uspuras, E.; Kaliatka, A.

2001-01-01

Results of the Level 1 probabilistic safety assessment of the Ignalina NPP has shown that in topography of the risk, transients dominate above the accidents with LOCAs and failure of the core long-term cooling are the main factors to frequency of the core damage. Previous analyses have shown, that after initial event, as a rule, the reactivity control, as well as short-term and intermediate cooling are provided. However, the acceptance criteria of the long-term cooling are not always carried out. It means that from this point of view the most dangerous accident scenarios are the scenarios related to loss of the core long-term cooling. On the other hand, the transition to the core condition due to loss of the long-term cooling specifies potential opportunities for the management of the accident consequences. Hence, accident management for the mitigation of the accident consequences should be considered and developed. The most likely initiating event, which probably leads to the loss of long term cooling accident, is station blackout. The station blackout is the loss of normal electrical power supply for local needs with an additional failure on start-up of all diesel generators. In the case of loss of electrical power supply MCPs, the circulating pumps of the service water system and MFWPs are switched-off. At the same time, TCV of both turbines are closed. Failure of diesel generators leads to the non-operability of the ECCS long-term cooling subsystem. It means the impossibility to feed MCC by water. The analysis of the station blackout for Ignalina NPP was performed using RELAP5 code. (author)

First international workshop on severe accidents and their consequences. [Chernobyl Accident

Energy Technology Data Exchange (ETDEWEB)

1989-07-01

An international workshop on past severe nuclear accidents and their consequences was held in Dagomys region of Sochi, USSR on October 30--November 3, 1989. The plan of this meeting was approved by the USSR Academy of Sciences and by the USSR State Committee of the Utilization of Atomic Energy. The meeting was held under the umbrella of the ANS-SNS agreement of cooperation. Topics covered include analysis of the Chernobyl accident, safety measures for RBMK type reactors and consequences of the Chernobyl accident including analysis of the ecological, genetic and psycho-social factors. Separate reports are processed separately for the data bases. (CBS)

Corrosion Characteristics of the SMART Materials

Energy Technology Data Exchange (ETDEWEB)

Baek, Jong Hyuk; Jeong, Y. H.; Choi, B. K.; Soh, J. R.; Lee, D. J.; Choi, B. S

2000-05-01

This report summarized the corrosion characteristics of the candidate steam generator tubes (PT-7M, ASTM Gr.2, Inconel-600), which are considering as the core materials in SMART. Also, this evaluated the waterchemstry conditions of commercial power plant including the PWR, BWR, WWER, PHWR, RBMK plants in comparison with that of SMART. And this report described that the microstructures of as-received PT-7M, ASTM Gr.2, and Inconel-600 as the candidate materials of fuel cladding and steam generator tubes and characterized the corrosion properties of the materials, which were tested systematically in the conditions of standard, ammonia solution and ammonia nodular to evaluate the corrosion resistance.

Application of the leak-before-break concept to the primary circuit piping of the Leningrad NPP

Energy Technology Data Exchange (ETDEWEB)

Eperin, A.P.; Zakharzhevsky, Yu.O.; Arzhaev, A.I. [and others

1997-04-01

A two-year Finnish-Russian cooperation program has been initiated in 1995 to demonstrate the applicability of the leak-before-break concept (LBB) to the primary circuit piping of the Leningrad NPP. The program includes J-R curve testing of authentic pipe materials at full operating temperature, screening and computational LBB analyses complying with the USNRC Standard Review Plan 3.6.3, and exchange of LBB-related information with emphasis on NDE. Domestic computer codes are mainly used, and all tests and analyses are independently carried out by each party. The results are believed to apply generally to RBMK type plants of the first generation.

Effect of ionite decomposition products on the reactor coolant pH in a boiling-water reactor

International Nuclear Information System (INIS)

Bredikhin, V.Ya.; Moskvin, L.N.

1982-01-01

The effect of products resulting from thermal radiolysis of ionites on water-chemical regime of NPP with RBMK is considered basing on investigations conducted in a boiling type experimental reactor. Data are presented on dynamics of changes in the specific electric conductivity and pH of the coolant following destruction of ion exchange groups and ionite matrix under the effect of reactor radiation. The authors draw a conclusion that radiation destruction of ionito fine disperse suspension or high-molecular soluble compounds in the reactor are, probably, one of the main reasons for variations in pH values of the coolant at NPP in non-correction water chemical regime

Corrosion Characteristics of the SMART Materials

International Nuclear Information System (INIS)

Baek, Jong Hyuk; Jeong, Y. H.; Choi, B. K.; Soh, J. R.; Lee, D. J.; Choi, B. S.

2000-05-01

This report summarized the corrosion characteristics of the candidate steam generator tubes (PT-7M, ASTM Gr.2, Inconel-600), which are considering as the core materials in SMART. Also, this evaluated the waterchemstry conditions of commercial power plant including the PWR, BWR, WWER, PHWR, RBMK plants in comparison with that of SMART. And this report described that the microstructures of as-received PT-7M, ASTM Gr.2, and Inconel-600 as the candidate materials of fuel cladding and steam generator tubes and characterized the corrosion properties of the materials, which were tested systematically in the conditions of standard, ammonia solution and ammonia nodular to evaluate the corrosion resistance

Detection of non-stationary leak signals at NPP primary circuit by cross-correlation analysis

International Nuclear Information System (INIS)

Shimanskij, S.B.

2007-01-01

A leak-detection system employing high-temperature microphones has been developed for the RBMK and ATR (Japan) reactors. Further improvement of the system focused on using cross-correlation analysis of the spectral components of the signal to detect a small leak at an early stage of development. Since envelope processes are less affected by distortions than are wave processes, they give a higher-degree of correlation and can be used to detect leaks with lower signal-noise ratios. Many simulation tests performed at nuclear power plants have shown that the proposed methods can be used to detect and find the location of a small leak [ru

Joint U.S./Russian Study on the Development of a Preliminary Cost Estimate of the SAFSTOR Decommissioning Alternative for the Leningrad Nuclear Power Plant Unit #1

Energy Technology Data Exchange (ETDEWEB)

SM Garrett

1998-09-28

The objectives of the two joint Russian/U.S. Leningrad Nuclear Power Plant (NPP) Unit #1 studies were the development of a safe, technically feasible, economically acceptable decom missioning strategy, and the preliminary cost evaluation of the developed strategy. The first study, resulting in the decommissioning strategy, was performed in 1996 and 1997. The preliminary cost estimation study, described in this report, was performed in 1997 and 1998. The decommissioning strategy study included the analyses of three basic RBM.K decommission- ing alternatives, refined for the Leningrad NPP Unit #1. The analyses included analysis of the requirements for the planning and preparation as well as the decommissioning phases.

The US nuclear safety approach to upgrading the Russian and Ukrainian reactors

International Nuclear Information System (INIS)

Baron, S.

1993-01-01

Brookhaven National Laboratory reporting to the Department of Energy has the technical and administrative management responsibilities for improving the operational and design safety systems of RBMK and the VVER reactors in Russia and the Ukraine. U.S. experts and industry interact with the Russian/Ukraine designers and operators to jointly develop the detailed requirements for system upgrades. When available, indigenous equipment and materials will be utilized. The construction and installation of most upgrades will be accomplished by Russia and the Ukraine with U.S. support and participation. This will maximize technology transfer, provide funds to U.S. and recipient country industry, and limit the nuclear liability of U.S. industry. (author)

Method for calculating the critical heat flux in mixed rod assemblies based on the tables of crisis in bundles

International Nuclear Information System (INIS)

Bobkov, V.P.

2000-01-01

The method for calculating the critical heat flux in the mixed rod assemblies, for example RBMK, containing three-four angle and peripheral macrocells, is presented. The method is based on generalization of experimental data in form of tables for the rods beams. It is recommended for the areas of parameters both provided for by experimental data and for others, where the data are absent. The advantages of the table method as follows: it is acceptable within a wide range of parameters and provides for smooth description of dependence of critical heat fluxes on these parameters; it is characterized by clearness, high reliability and accuracy and is easy in application [ru

The accident at the Chernobyl' nuclear power plant and its consequences. Pt. 1. General material

International Nuclear Information System (INIS)

1986-01-01

The report contains a presentation of the Chernobyl' nuclear power station and of the RBMK-1000 reactor, including its principal physical characteristics, the safety systems and a description of the site and of the surrounding region. After a chronological account of the events which led to the accident and an analysis of the accident using a mathematical model it is concluded that the prime cause of the accident was an extremely improbable combination of violations of instructions and operating rules committed by the staff of the unit. Technical and organizational measures for improving the safety of nuclear power plants with RBMK reactors have been taken. A detailed description of the actions taken to contain the accident and to alleviate its consequences is given and includes the fire fighting at the nuclear power station, the evaluation of the state of the fuel after the accident, the actions taken to limit the consequences of the accident in the core, the measures taken at units 1, 2 and 3 of the nuclear power station, the monitoring and diagnosis of the state of the damaged unit, the decontamination of the site and of the 30 km zone and the long-term entombment of the damaged unit. The measures taken for environmental radioactive contamination monitoring, starting by the assessment of the quantity, composition and dynamics of fission products release from the damaged reactor are described, including the main characteristics of the radioactive contamination of the atmosphere and of the ground, the possible ecological consequences and data on the exposure of plant and emergency service personnel and of the population in the 30 km zone around the plant. The last part of the report presents some recommendations for improving nuclear power safety, including scientific, technical and organizational aspects and international measures. Finally, an overview of the development of nuclear power in the USSR is given

The causes of the Chernobyl event

International Nuclear Information System (INIS)

Frot, J.

2000-11-01

The Chernobylsk event has two components, the explosion of the RBMK type nuclear reactor number 4 and the sanitary damages that resulted. The causes of the explosion are of three kinds: conception error, management fault, exploitation personnel mistakes and political causes. For the sanitary damages there are the immediate causes and the deep causes. No emergency planning to answer to a such disaster and no iodinated tablets delivery to protect the thyroid for the direct causes. The secret culture made that the knowledge developed by the Soviet researchers was not diffused to the medical and nuclear communities of USSR. The civil authorities were not aware of it or they neglected it. (N.C.)

An environmental perspective on Lithuania's energy options

International Nuclear Information System (INIS)

Banks, A.; Todd, J.

1995-01-01

The views of experts on Lithuania's energy options are reviewed. On the one hand, nuclear energy is seen as an island of stability in the power industry in the conditions of economic crisis, and some decision-makers believe that Lithuania cannot survive without nuclear. On the other hand, the Ignalina NPP is the largest Chernobyl-type RBMK plant within the former Soviet Union, posing a dangerous environmental hazard to the Baltic Sea region, and no upgrading seems to be capable of bringing the reactors up to the safety standards of today's Western reactors. Many experts believe that the only solution is to shut the reactors down as soon as possible. (P.A.) 33 refs

Application of ecological interface design in nuclear power plant (NPP operator support system

Directory of Open Access Journals (Sweden)

Alexey Anokhin

2018-05-01

Full Text Available Most publications confirm that an ecological interface is a very efficient tool to supporting operators in recognition of complex and unusual situations and in decision-making. The present article describes the experience of implementation of an ecological interface concept for visualization of material balance in a drum separator of RBMK-type NPPs. Functional analysis of the domain area was carried out and revealed main factors and contributors to the balance. The proposed ecological display was designed to facilitate execution of the most complicated cognitive operations, such as comparison, summarizing, prediction, etc. The experimental series carried out at NPPs demonstrated considerable reduction of operators' mental load, time of reaction, and error rate. Keywords: Ecological Interface Design, Experimental Evaluation, Model, Work Domain Analysis

Management of spent fuel from power and research reactors using CASTOR and CONSTOR casks and licensing experience worldwide

International Nuclear Information System (INIS)

Becher, D.

2003-01-01

An overview of the spent fuel storage in CASTOR and CONSTOR casks during the last 30 years is made. Design characteristics of the both types of casks are presented. CASTOR casks fulfill both the requirements for type B packages according to the IAEA requirements covering different accident situations in storage sites. Analyses of nuclear and thermal behavior and strength are carried out for CONSTOR concept. Special experimental program for verification of mechanical and thermomechanical properties is implemented. Licensing experience of the casks in German storage facilities is presented. Special modifications of CASTOR casks for WWER-440 and RBMK fuel assemblies have been designed for implementation in Eastern Europe. Contracts for GNB spent fuel casks delivery are concluded with Czech Republic, Slovakia, Hungary and Lithuania

Perspectives of nuclear energy in Lithuania

International Nuclear Information System (INIS)

Bieliauskas, V.; Marchenas, V.

1998-01-01

Description of present status of nuclear power in Lithuania and prospects for future are presented. Lithuania operate two reactors of RBMK-1500 type. Since regaining of independence in 1990 Lithuania made a great efforts in developing legal framework for nuclear power regulation and improving safety of both reactors at Ignalina NPP. The main ideas of the draft of a new energy strategy are summarized. As regards nuclear power development in Lithuania there are two scenarios in the draft strategy: operation of the plant till the end of its design lifetime and operation of the plant till the gap closure between fuel channel and graphite and non re channeling of the reactors. Comparison of the cost and implications to the country's economy of both scenarios is discussed

Challenges of Ignalina NPP Decommissioning - View of Lithuanian Operator

International Nuclear Information System (INIS)

Aksionov, P.

2017-01-01

The state enterprise Ignalina Nuclear Power Plant (INPP) operates 2 similar design units of RBMK-1500 water-cooled graphite-moderated channel-type power reactors (1500 MW electrical power). INPP is carrying out the decommissioning project of the 2 reactors which includes: -) the retrieval of the spent nuclear fuel from the power units and its transportation into the Interim Spent Fuel Storage Facility; -) equipment and building decontamination and dismantling; -) radioactive waste treatment and storage; and -) the operation of key systems to ensure nuclear, radiation and fire protection. Ignalina NPP decommissioning project is planned to be completed by 2038. The presentation will be focused on the ongoing decommissioning activities at Ignalina NPP. The overview of main aspects and challenges of INPP decommissioning will be provided

Calculation support for industrial production of cobalt-60 at Leningrad NPP

International Nuclear Information System (INIS)

Artemov, Vladimir; Elshin, Alexander; Ivanov, Alexander; Gorbunov, Evgeny; Ikonnikov, Roman; Pimenov, Alexander

2008-01-01

Cobalt-60 is industrially produced at the Leningrad NPP by irradiation of cobalt-59 in special-purpose facilities loaded into the RBMK reactor core (all 4 units). The paper describes calculation methods used to determine the current activity of cobalt in irradiation assemblies for their timely unloading. The described peculiarities of core calculation model account for continuous refueling, overloading of irradiation assemblies and individual thermohydraulics in each channel under variation of reactor power. Fuel burnup in the core is calculated with a time step of about 24 hours. The resulting values for cobalt activity and uncertainties are presented in the paper as well. Deviation of calculated cobalt activity from measured activity is within the experimental accuracy of 10% (at confidence probability of 0.95). (authors)

15 years after Chernobyl, nuclear power plant safety improved world-wide, but regional strains on health, economy and environment remain

International Nuclear Information System (INIS)

2001-01-01

Fifteen years after the Chernobyl accident, exhaustive studies by the IAEA and others provide a solid understanding of the causes and consequences of the accident, which stemmed from design deficiencies in the reactor compounded by violation of operating procedures. These deficiencies and the lack of an international notification mechanism led to the speedy adoption of early Notification and Assistance Conventions as well as later establishment of the landmark Convention on Nuclear Safety. Lessons learned from the accident were also a significant driving force behind a decade of IAEA assistance to the countries of Central and eastern Europe and the Former Soviet Union. Much of this work was focused on identifying the weaknesses in and improving the design safety of WWER and RBMK reactors

Design, construction and commissioning of an interim spent fuel store for the decommissioning of Ignalina NPP, Lithuania

International Nuclear Information System (INIS)

Rainer Goehring; Martin Beverungen; Phil Smith

2006-01-01

The contract for the design, construction and commissioning (turn-key) of an interim spent fuel store facility (ISFSF) has been awarded to a Consortium of GNS Gesellschaft fuer Nuklear Service and RWE NUKEM GmbH under the lead of RWE NUKEM. The contract was signed on the 12.01.2005. The Interim Spent Fuel Storage Facility (ISFSF) is financed by the Ignalina Decommissioning Support Fund which is managed by EBRD. All spent fuel assemblies, currently stored in the spent fuel pits at the reactors plus future arising (about 18000 in total) will be loaded in the CONSTOR R RBMK1500/M2 containers, which are stored in the new facility. The initial contract has been awarded for 3500 spent fuel assemblies. (authors)

Presentations provided

International Nuclear Information System (INIS)

Hashemian, H.; Beverly, D.

1998-01-01

The following topics covered in detail at the workshop included: temperature instrumentation; pressure instrumentation; in-situ calibration and response time testing of RTDs and pressure transmitters; on-line performance monitoring and preventive maintenance of critical equipment; automated measurement of critical parameters; nuclear power plant infrastructure, management and Quality Assurance issues and recent developments for WWER and RBMK reactors. Conclusions drawn were: aging can adversely affect the performance of nuclear plant pressure transmitters; current testing interval of once in every fuel cycle is adequate for aging management; in-situ response time measurements and on-line calibration testing methods have been developed and validated for nuclear plant pressure transmitters; NUREG/CR-5851 should be taken into account for details of aging research on pressure transmitters

Presentations provided

Energy Technology Data Exchange (ETDEWEB)

Hashemian, H; Beverly, D [Analysis and Measurement Services Corp., Knoxville, TN (United States)

1999-12-31

The following topics covered in detail at the workshop included: temperature instrumentation; pressure instrumentation; in-situ calibration and response time testing of RTDs and pressure transmitters; on-line performance monitoring and preventive maintenance of critical equipment; automated measurement of critical parameters; nuclear power plant infrastructure, management and Quality Assurance issues and recent developments for WWER and RBMK reactors. Conclusions drawn were: aging can adversely affect the performance of nuclear plant pressure transmitters; current testing interval of once in every fuel cycle is adequate for aging management; in-situ response time measurements and on-line calibration testing methods have been developed and validated for nuclear plant pressure transmitters; NUREG/CR-5851 should be taken into account for details of aging research on pressure transmitters

Chemical technologies and life management of Ukrainian NPPs

International Nuclear Information System (INIS)

Arkhipenko, A.V.; Barbashev, S.V.; Litvinskij, L.L.; Masko, A.N.

2000-01-01

Now 11 units with WWER-1000 reactors, 2 units with WWER-440 and 1 unit with RBMK-1000 are operated in Ukraine. State of chemical technologies of NPPs essentially influences on unit operating resource by the next ways: decreasing of corrosion intensity of equipment metal; decreasing of contamination on thermal exchanged surfaces of equipment; decreasing of amounts of radioactive waste. Improvement of these parameters can be achieved by the next measures: improvement of purification schemes for feed water of main systems; introduction of more effective water-chemical regimes (WCR); implementation of new methods and instruments of chemical monitoring for WCR; providing of without-scale regime of thermal-exchanged equipment operating by reactor division users through optimisation of the WCR of the NPP spray pool. (author)

United Kingdom and USSR reactor types

International Nuclear Information System (INIS)

Lewins, Jeffery

1988-01-01

The features of the RBMK reactor operated at Chernobyl are compared with reactor types pertinent to the UK. The UK reactors covered are in three classes: the commercial reactors now built and operated or in commission (Magnox and Advanced Gas-cooled Reactor (AGR)); the prototype Steam Generating Heavy Water Reactor (SGHWR) and Prototype Fast Reactor (PFR) that have comparable performance to commercial reactors; and the proposed Pressurised Water Reactor (PWR) or Sizewell 'B' design which, it will be recollected, is different in detail from PWRs built elsewhere. We do not include research and test reactors nor the Royal Navy PWRs. The appendices explain resonances, Doppler and Xenon effects, the reactor physics of Chernobyl and positive void coefficients all of which are relevant to the comparisons. (author)

Thermal strain measurements in graphite using electronic speckle pattern interferometry

International Nuclear Information System (INIS)

Tamulevicius, S.; Augulis, L.; Augulis, R.; Zabarskas, V.; Levinskas, R.; Poskas, P.

2001-01-01

Two 1500 MW(e) RBMK Units are operated at Ignalina NPP in Lithuania. Due to recent decision of the Parliament on the earlier closure of Unit 1, preparatory work for decommissioning have been initiated. Preferred decommissioning strategy is based on delayed dismantling after rather long safe enclosure period. Since graphite is one of the basic and probably the most voluminous components of the reactor internals, a sufficient information on status and behaviour of graphite moderator and reflector during long time safe enclosure period is of special significance. In this context, thermal strain in graphite is one of the parameters requiring particular interest. Electronic speckle pattern interferometry has been proposed and successfully tested to control this parameter using the real samples of graphite from Ignalina NPP Units. (author)

New challenges in energy future of Lithuania

International Nuclear Information System (INIS)

Gylys, J.; Ziedelis, S.; Adomavicius, A.

2004-01-01

Lithuania is a relatively small country with the population of 3,5 mln, disproportionately powerful energy industry and low energy consumption. Installed electricity generating capacities are more than 6 GW, but total power demand is less than 2 GW. Lithuania with average electricity consumption about 2900 kWh per person occupies one of the last places in Europe. Nuclear is the main source of electric energy in Lithuania: it covers 60 - 86% of total electricity production. Comparing consumption of all primary energy sources in all branches of economy nuclear covers about one third (32 - 37%) of the whole alongside with oil (31 - 33%) and natural gas (30 - 31%). At Ignalina NPP Lithuania are operating two the RBMK-1500 type reactors - the most advanced version of the former Soviet Union channel type reactor design series. The designed electrical power of RBMK-1500 reactor (1500 MW) is the biggest in the world for the single unit. The first unit of INPP was put into operation by the end of 1983 and the second unit in 1987. After Chernobyl accident the maximal allowed electrical power of each reactor at INPP was reduced to 1350 MW. The initial RBMK-1500 design at Ignalina NPP at the present time is substantially improved. More than 200 million US dollars of western countries support were spent, and numerous safety features were implemented. Nowadays both Lithuanian and foreign experts agree, that the safety level of Ignalina NPP is very similar to the western type NPP's of the same age. During accession process, one of the main EU requirements to energy sector of Lithuania was to close both reactors of Ignalina NPP, which were decided to be unsafe in principle. Despite all efforts of Lithuanian specialists and negotiators shutdown of the 1st reactor of INPP is foreseen at the end of 2004 and shutdown of the 2nd reactor is foreseen at the end of 2009. Closure of Ignalina NPP will decrease maximal power generating capacity to 2641 MW in 2010 and will cause a complex of

Specific features of RBMK severe accidents progression and approach to the accident management

International Nuclear Information System (INIS)

Vasilevskij, V.P.; Nikitin, Yu.M.; Petrov, A.A.; Potapov, A.A.; Cherkashov, Yu.M.

2001-01-01

Fundamental construction features of the LWGR facilities (absence of common external containment shell, disintegrated circulation circuit and multichannel reactor core, positive vapor reactivity coefficient, high mass of thermally capacious graphite moderator) predetermining development of assumed heavy non-projected accidents and handling them are treated. Rating the categories of the reactor core damages for non-projected accidents and accident types producing specific grope of damages is given. Passing standard non-projected accidents, possible methods of attack accident consequences, as well as methods of calculated analysis of non-projected accidents are demonstrated [ru

Determination of water chemistry parameters which influence on failure intensity of RBMK equipment

Energy Technology Data Exchange (ETDEWEB)

Kovalev, S.M.; Kritski, V.G.; Berezina, I.G.; Stjazhkin, P.S.; Olejnik, P.V. [All-Russian Design and Scientific Research Inst. of Complex Power Technology (VNIPIET), St. Petersburg (Russian Federation)

2002-07-01

The coolant quality has an effect on intensity of austenitic stainless steel intergranular corrosion. The correlation between rate of intergranular stress corrosion cracking and water electro-conductivity (on generalized data) for non-stabilized stainless steel 304SS type and stabilized stainless steel X18H10T type is shown in Fig. 1. The cracking rate is increased with electro-conductivity rise and this is equivalently to reduction of pipelines lifetime, increase of amount and depth of cracks in welds. The increased values of water electro-conductivity correspond to high concentrations of iron corrosion product, corrosion-active anions (chlorides, sulphates et. al.), arrived to circulation circuit due to failure of equipment elements, for example, of condenser tubes or during start-up. (author)

Analysis and first evaluation of the course of the Chernobyl accident up to the excursion. Interim report. Analyse und erste Bewertung des Unfallablaufs in Tschernobyl bis zur Leistungsexkursion. Zwischenbericht

Energy Technology Data Exchange (ETDEWEB)

Clemente, M; Frisch, W; Langenbuch, S; Weber, J P

1986-01-01

This report contains a description and an evaluation of the course of the Tschernobyl accident up to the excursion. It is based on information obtained during the IAEA conference in Vienna in August 1986 and includes a first qualitative evaluation of the course of the accident as well as results of analyses carried out at GRS. This work was done with the aim to better understand the particular phases of the accident and to demonstrate the typical dynamic behaviour of the RBMK-1000 type reactor with a positive void coefficient in contrast to the behaviour of german BWRs with negativ void coefficients. The calculations also contribute to the evaluation of the consequences of the violations and errors executed by the operating team and the consequences of design weaknesses of the plant.

The Hatch-Smolensk exchange

International Nuclear Information System (INIS)

Sproles, A.

1993-01-01

During summer 1992, the World Association of Nuclear Operators (WANO) sponsored an exchange visit between Georgia Power Company's Edwin I. Hatch nuclear plant, a two-unit boiling water reactor site, and the Smolensk atomic energy station, a three-unit RBMK (graphite-moderated and light-water-cooled) plant located 350 km west of Moscow, in Desnogorsk, Russia. The Plant Hatch team included Glenn Goode, manager of engineering support; Curtis Coggin, manager of training and emergency preparedness; Wayne Kirkley, manager of health physics and chemistry; John Lewis, manager of operations; Ray Baker, coordinator of nuclear fuels and contracts; and Bruce McLeod, manager of nuclear maintenance support. Also traveling with the team was Jerald Towgood, of WANO's Atlanta Centre. The Hatch team visited the Smolensk plant during the week of July 27, 1992

Investigations of possibilities to dispose of spent nuclear fuel in Lithuania: a model case. Volume 2, Concept of Repository in Crystalline Rocks

International Nuclear Information System (INIS)

Motiejunas, S.; Poskas, P.

2005-01-01

The aim is to present the generic repository concept in crystalline rocks in Lithuania and cost assessment of the disposal of spent nuclear fuel and long-lived intermediate level waste in this repository. Due to limited budget of this project the repository concept development for Lithuania was based mostly on the experience of foreign countries. In this Volume a review of the existing information on disposal concept in crystalline rocks from various countries is presented. Described repository concept for crystalline rocks in Lithuania covers repository layout, backfill, canister, construction materials and auxiliary buildings. Costs calculations for disposal of spent nuclear fuel and long-lived intermediate-level wastes from Ignalina NPP are presented too. Thermal, criticality and other important disposal evaluations for RBMK-1500 spent nuclear fuel emplaced in copper canister were performed and described

Adaptation of GRS calculation codes for Soviet reactors

International Nuclear Information System (INIS)

Langenbuch, S.; Petri, A.; Steinborn, J.; Stenbok, I.A.; Suslow, A.I.

1994-01-01

The use of ATHLET for incident calculation of WWER has been tested and verified in numerous calculations. Further adaptation may be needed for the WWER 1000 plants. Coupling ATHLET with the 3D nuclear model BIPR-8 for WWER cores clearly improves studies of the influence of neutron kinetics. In the case of FBMK reactors ATHLET calculations show that typical incidents in the complex RMBK reactors can be calculated even though verification still has to be worked on. Results of the 3D-core model QUABOX/CUBBOX-HYCA show good correlation of calculated and measured values in reactor plants. Calculations carried out to date were used to check essential parameters influencing RBMK core behaviour especially dependence of effective voidre activity on the number of control rods. (orig./HP) [de

Review of the safety analysis of the Ignalina Nuclear Power Plant

International Nuclear Information System (INIS)

Weber, J.P.

1999-01-01

Description of the review of safety analysis report (SAR) of Ignalina NPP is presented. this review, called RSR, was conducted by independent group of international experts. SAR and RSR represented a unique international effort in a very short time. It was first attempt to provide Western-type SAR for any Soviet designed NPP. SAR was completed in December 1996, RSR was completed in March 1997. SAR has produced 85 reports, available on CR-ROM and RSR produced 86 reports also available on CD-ROM. SAR presented a large list of recommendations. RSR agreed with most of SAR recommendations and added further recommendations. SAR and RSR contributed significantly to better understanding of RBMK plant behaviour. Ignalina NPP has responded to the SAR/RSR recommendations by an ambitious Safety Improvement Programme SIP-2 which currently is under implementation

The use of energy analysis and indexes of energy efficiency in nuclear power

International Nuclear Information System (INIS)

D'yakonov, E.I.; Ignatenko, E.I.

1991-01-01

The results of calculating the indexes of energy efficiency for NPPs with the WWER-1000 and RBMK-1000 reactors, heat and power NPPs with the WWER-1000 and dictrict heating NPPs with the AST-500 reactor in three fuel cycles, namely, the open one and with uranium and plutonium recycles, are considered. Complex account for the quantity and quality of produced and consumed energy provides for objective evaluation of the indexes of energy efficiency during comparative analysis of nuclear power plants with different types of reactors. It is shown that complex use of the energy produced at a NPP provides for increase of indexes of energy efficiency. The highest indexes are obtained for heat and power NPP with the WWER-1000 reactor in the open fuel cycle, with uranium and plutonium recycle and for NPP with the WWER-1000 reactor with plutonium recycle

Preliminary report about nuclear accident of Chernobylsk reactor

International Nuclear Information System (INIS)

Oliveira, A.R. de.

1986-07-01

The preliminary report of nuclear accident at Chernobyl, in URSS is presented. The Chernobyl site is located geographically and the RBMK type reactors - initials of russian words which mean high power pressure tube reactors are described. The conditions of reactor operation in beginning of accident, the events which lead to reactor destruction, the means to finish the fire, the measurements adopted by Russian in the accident location, the estimative of radioactive wastes, the meteorological conditions during the accident, the victims and medical assistence, the sanitary aspects and consequences for population, the evaluation of radiation doses received at small and medium distance and the estimative of reffered doses by population attained are presented. The official communication of Russian Minister Council and the declaration of IAEA general manager during a collective interview in Moscou are annexed. (M.C.K.) [pt

Chernobyl: getting to the heart of the matter

International Nuclear Information System (INIS)

North, Richard.

1996-01-01

In the second of two linked articles on the aftermath of the Chernobyl nuclear reactor accident of 1986, the author explores the effects on local agriculture and the health of populations affected by the contamination from the fall-out, especially children. Agriculture around Chernobyl has resumed, with workers moving back from the cities to areas where radiation doses are similar to parts of Cornwall. Concern continues about the safety of milk from cows grazing contaminated grass and eating local mushrooms. The largest risk to children's health is not birth deformaties, but leukaemia, possibly in part due to iodine deficiency in their diet prior to contamination. Concern also continues about keeping power supplies going in areas heavily dependent on nuclear power. Reactor safety issues remaining operational RBMK reactors and the sarcophagus around Chernobyl-4 itself have yet to be resolved. (UK)

Comprehensive survey of the Russian nuclear industry

International Nuclear Information System (INIS)

2004-03-01

This document presents the organization of nuclear activities in the Russian federation: Minatom and its replacement by the federal agency of atomic energy, personnel, nuclear power plants (VVER, RBMK, fast neutron and mixed reactors), availability and power production, export of activities (construction of nuclear power plants in Slovakia, Iran, China, India, project in Viet Nam), expansion of the nuclear power plants park (improvement of plants safety, increase of service life), completion of uncompleted plants, the construction of which was stopped after the Chernobyl accident and the reorganization of the former-USSR, construction of new generation power plants (VVER-640, -1000 and -1500), fuel cycle facilities (geographical distribution, production of natural uranium, conversion and enrichment), fuel fabrication, reprocessing processes and spent fuel storage, management of radioactive wastes (leasing), R and D activities (organizations and institutes), research programs of the international scientific and technical center, nuclear safety authority (Gosatomnadzor - GAN). (J.S.)

Validation of the reactor dynamics code TRAB

International Nuclear Information System (INIS)

Raety, H.; Kyrki-Rajamaeki, R.; Rajamaeki, M.

1991-05-01

The one-dimensional reactor dynamics code TRAB (Transient Analysis code for BWRs) developed at VTT was originally designed for BWR analyses, but it can in its present version be used for various modelling purposes. The core model of TRAB can be used separately for LWR calculations. For PWR modelling the core model of TRAB has been coupled to circuit model SMABRE to form the SMATRA code. The versatile modelling capabilities of TRAB have been utilized also in analyses of e.g. the heating reactor SECURE and the RBMK-type reactor (Chernobyl). The report summarizes the extensive validation of TRAB. TRAB has been validated with benchmark problems, comparative calculations against independent analyses, analyses of start-up experiments of nuclear power plants and real plant transients. Comparative RBMES type reactor calculations have been made against Soviet simulations and the initial power excursion of the Chernobyl reactor accident has also been calculated with TRAB

ILK statement on the consequences of the Chernobyl accident. Taking stock after twenty years

International Nuclear Information System (INIS)

2006-01-01

The Chernobyl reactor accident was the consequence of a reactor design which was not inherently safe, and of a lack of 'safety culture'. The RBMK-type reactor (a Russian graphite-moderated light water reactor design: reaktor bolshoi moshnosty kanalny=high-power channel reactor) had not been designed to a satisfactory safety level, and the operating staff were not informed on the weak spots in plant design. The combination of these factors caused the worst nuclear accident, completely destroying the reactor. The consequences may be seen as the product of two severe accidents superimposed upon each other: the explosion of the reactor, and core melt-down associated with an intense, persistent fire of the graphite moderator. The Statement contains analyses of these points: Release, Propagation and Deposition of Radioactive Materials; Protective Measures; Impact on the Environment and Agriculture; Assessment of Radiation Exposure; Health Impact; Psychological and Societal Impacts; Potential Residual Risks. (orig.)

Status of Ignalina's safety analysis reports

International Nuclear Information System (INIS)

Uspuras, E.

1999-01-01

Ignalina NPP is unique among RBMK type reactors in the scope and comprehensiveness of international studies which have been performed to verify its design parameters and analyze risk levels. International assistance took several forms, a very valuable mod of assistance utilized the knowledge of international experts in extensive international studies whose purpose was: collection, systematization and verification of plant design data; analysis of risk levels; recommendations leading to improvements in the safety lave; transfer of state of the art analytical methodology to Lithuanian specialists. The major large scale international studies include: probabilistic risk analysis; extensive international study meant to provide comprehensive overview of plant status with special emphasis on safety aspects; an extensive review of the Safety Analysis Report by an independent group of international experts. In spite of the safety improvements and analyses which have been performed at the Ignalina NPP, much remains to be done in the nearest future

Comprehensive survey of the Russian nuclear industry; Le panorama nucleaire russe

Energy Technology Data Exchange (ETDEWEB)

NONE

2004-03-01

This document presents the organization of nuclear activities in the Russian federation: Minatom and its replacement by the federal agency of atomic energy, personnel, nuclear power plants (VVER, RBMK, fast neutron and mixed reactors), availability and power production, export of activities (construction of nuclear power plants in Slovakia, Iran, China, India, project in Viet Nam), expansion of the nuclear power plants park (improvement of plants safety, increase of service life), completion of uncompleted plants, the construction of which was stopped after the Chernobyl accident and the reorganization of the former-USSR, construction of new generation power plants (VVER-640, -1000 and -1500), fuel cycle facilities (geographical distribution, production of natural uranium, conversion and enrichment), fuel fabrication, reprocessing processes and spent fuel storage, management of radioactive wastes (leasing), R and D activities (organizations and institutes), research programs of the international scientific and technical center, nuclear safety authority (Gosatomnadzor - GAN). (J.S.)

Treatment of NPP wastes using vitrification

International Nuclear Information System (INIS)

Sobolev, I.A.; Lifanov, F.A.; Stefanovsky, S.V.; Kobelev, A.P.; Savkin, A.E.; Kornev, V.I.

1998-01-01

Glass-based materials to immobilize various liquid and solid radioactive wastes generated at nuclear power plants (NPP) were designed. Glassy waste forms can be produced using electric melting including a cold crucible melting. Leach rate of cesium was found to be 10 -5 -10 -6 g/(cm 2 day) (IAEA technique). Volume reduction factor after vitrification reached 4-5. Various technologies for NPP waste vitrification were developed. Direct vitrification means feeding of source waste into the melter with formation of glassy waste form to be disposed. Joule heated ceramic melter, and cold crucible were tested. Process variables at treatment of Kursk, Chernobyl (RBMK), Kalinin, Novovoronezh (VVER) NPP wastes were determined. The most promising melter was found to be the cold crucible. Pilot plant based on the cold crucibles has been designed and constructed. Solid burnable NPP wastes are incinerated and slags are incorporated in glass. (author)

Improvements to the transient solution in the PANTHER space-time code

International Nuclear Information System (INIS)

Kutt, P.K.; Knight, M.P.

1993-01-01

The three dimensional, two-group, nodal diffusion code PANTHER has been developed for the analysis of almost all thermal reactor types [pressurized water reactor (PWR), boiling water reactor, VVER, RBMK, advanced gas-cooled reactor, MAGNOX]. It can perform a comprehensive range of calculations for fuel management, operational support including on-line application, and transient analysis. Transient results for a number of light water reactor (LWR) benchmark problems have been reported previously. This paper outlines some recent developments of the transient solution in PANTHER, showing results for two LWR benchmark problems. Recently, PANTHER results have been accepted as the reference solutions for a Nuclear Energy Agency Committee on Reactor Physics (NEACRP) rod ejection benchmark Unlike previous simplified rod ejection benchmarks, it represents a real PWR with a detailed thermal model and cross sections dependent on boron, fuel temperature, and water density and temperature. This reference solution was computed with fine time steps

Safety analysis of Ignalina NPP during shutdown conditions

International Nuclear Information System (INIS)

Kaliatka, A.; Uspuras, E.

2000-01-01

The accident analysis for the Ignalina NPP with RBMK-1500 reactors at normal operating conditions and at minimum controlled power level (during startup of the reactor) has been performed in the frame of the project I n-Depth Safety Assessment of the Ignalina NPP , which was completed in 1996. However, the plant conditions during the reactor shutdown differ from conditions during reactor operation at full power (equipment status in protection systems, set points for actuation of safety and protection systems, etc.). Results of RELAP5 simulation of two worst initiating events during reactor shutdown - Pressure Header rupture in case of steam reactor cooldown as well as Pressure Header rupture in case of water reactor cooldown are discussed in the paper. Results of analysis shown that reactor are reliably cooled in both cases. Further analysis for all range of initial events during reactor shutdown and at shutdown conditions is recommended. (author)

Assessment of core damage frequency owing to possible fires at NPP with RBMK type reactors

Energy Technology Data Exchange (ETDEWEB)

Vinnikov, B. [National Research Centre Kurchatov Inst., 1, Kurchatov Square, Moscow, 123 182 (Russian Federation); NRC Kurchatov Inst. (Russian Federation)

2012-07-01

According to Scientific and Technical Cooperation between the USA and Russia in the field of nuclear engineering the Idaho National Laboratory has transferred to the possession of the National Research Center ' Kurchatov Inst. ' the SAPHIRE software without any fee. With the help of the software Kurchatov Inst. developed a Pilot Living PSA- Model of Leningrad NPP Unit 1. Computations of core damage frequencies were carried out for additional Initiating Events. In the submitted paper such additional Initiating Events are fires in various compartments of the NPP. During the computations of each fire, structure of the PSA - Model was not changed, but Fault Trees for the appropriate systems, which are removed from service during the fire, were changed. It follows from the computations, that for ten fires Core Damaged Frequencies (CDF) are not changed. Other six fires will cause additional core damage. On the basis of the calculated results it is possible to determine a degree of importance of these fires and to establish sequence of performance of fire-prevention measures in various places of the NPP. (authors)

Assessment of core damage frequency owing to possible fires at NPP with RBMK type reactors

International Nuclear Information System (INIS)

Vinnikov, B.

2012-01-01

According to Scientific and Technical Cooperation between the USA and Russia in the field of nuclear engineering the Idaho National Laboratory has transferred to the possession of the National Research Center ' Kurchatov Inst. ' the SAPHIRE software without any fee. With the help of the software Kurchatov Inst. developed a Pilot Living PSA- Model of Leningrad NPP Unit 1. Computations of core damage frequencies were carried out for additional Initiating Events. In the submitted paper such additional Initiating Events are fires in various compartments of the NPP. During the computations of each fire, structure of the PSA - Model was not changed, but Fault Trees for the appropriate systems, which are removed from service during the fire, were changed. It follows from the computations, that for ten fires Core Damaged Frequencies (CDF) are not changed. Other six fires will cause additional core damage. On the basis of the calculated results it is possible to determine a degree of importance of these fires and to establish sequence of performance of fire-prevention measures in various places of the NPP. (authors)

Investigation of cascade regions of damage in alpha iron by a computer simulation method (crystal model). Issledovaniye kaskadnykh oblastey povrezhdeniya v. cap alpha. -zheleze metodom mashinnogo modelirovaniya (kristallicheskaya model)

Energy Technology Data Exchange (ETDEWEB)

Kevorkyan, Yu R

1974-01-01

A SPIKE program is used to study regions of structural damage that arise as a result of cascades of atomic collisions in single-crystal alpha iron. The model of the cascade process realized in the program uses a pair collision approximation and accounts for the influence of the crystal structure of the material. The following characteristics of regions of damage are found as a function of the energy of the primary knock-on atom: volume of the region, displacement effectiveness, size distribution of complexes of vacancies and injections. The results are compared with data in the literature. An appendix gives the text of the SPIKE program in FORTRAN.

Energy supply, nuclear power, and the international energy situation

International Nuclear Information System (INIS)

Pierer, H. von

1991-01-01

The Chernobyl accident has greatly intensified the readiness for international cooperation on problems of reactor safety and for exchanges of operating experience. That accident was more than a regional event. If all psychological and political consequences are taken into account, its international significance is apparent. In principle, it demonstrated not the lack of safety of nuclear power plants generally, but rather that of the Soviet RBMK reactor line, which would not have been licensed in any Western country because of its inherent unsafety. In the long run, the worldwide acceptance of nuclear power can be regained and stabilized only by an open dialog and by international exchanges of experience. The pronounced growth of the world's population requires energy policy to think beyond national frontiers. The rising energy requirement permits of no other decision than to exploit all technically feasible and economically viable as well as ecologically tolerable sources of energy. This includes nuclear power as well as solar energy. (orig.) [de

Derivation of the mass factors for decommissioning cost estimation of low contaminated auxiliary systems

International Nuclear Information System (INIS)

Poskas, G.

2015-01-01

Ignalina NPP was operating two RBMK-1500 reactors. Unit 1 was closed at the end of 2004, and Unit 2 - at the end of 2009. Now they are under decommissioning. Decommissioning has been started from the reactor's periphery, with dismantling of non-contaminated and low contaminated equipment and installations. This paper discusses a methodology for derivation of mass factors for preliminary decommissioning costing at NPP when the number of inventory items is significant, and separate consideration of each inventory item is impossible or impractical for preliminary decommissioning plan, especially when the level of radioactive contamination is very low. The methodology is based on detailed data analysis of building V1 taking into account period and inventory based activities, investment and consumables and other decommissioning approach- related properties for building average mass factors. The methodology can be used for cost estimation of preliminary decommissioning planning of NPP auxiliary buildings with mostly very low level contamination. (authors)

The Chernobyl and Fukushima Daiichi nuclear accidents and their tragic consequences

CERN Multimedia

CERN. Geneva

2016-01-01

On April 26, 1986, the Unit 4 of the RBMK nuclear power plant of Chernobyl, in Ukraine, went out of control during a test at low-power, leading to an explosion and fire. The reactor building was totally demolished and very large amounts of radiation were released into the atmosphere for several hundred kilometres around the site including the nearby town of Pripyat. The explosion leaving tons of nuclear waste and spent fuel residues without any protection and control totally contaminating the entire area. Several hundred thousand people were affected by the radiation fall out. The radioactive cloud spread across Europe affecting most of the Northern, Central and Eastern European countries. Some areas of southern Switzerland, of northern Italy as well as western France were subject to radioactive contamination. The initiative of the G7 countries to launch and important programme for the closure of some Soviet built nuclear plants was accepted by several donor countries. A team of engineers was established wi...

Experience of upgrading existing Russian designed nuclear plants

International Nuclear Information System (INIS)

Yanev, P.I.; Facer, R.I.

1993-01-01

From the reviewed experiences of upgrading existing Russian designed nuclear plants both of WWER and RBMK type the conclusions drawn are as follows. For the countries operating Russian designed plants it is necessary to adopt a pragmatic approach where all changes must be demonstrated to improve the safety of the plant and safety must be demonstrably improving. Care must be taken to avoid the pitfalls of excessive regulatory demands which are not satisfied and the development of an attitude of disregarding requirements on the basis that they are not enforced. For the lending countries and organizations, it is necessary to ensure that assistance is given to the operating organizations so that the most effective use of funds can be achieved. The experience in the West is that over-regulation and excessive expenditure do not necessarily lead to improved safety. They can lead to significant waste of resources. The use of western technology is recommended but where it is necessary and where it provides the greatest benefit

Power generation by nuclear power plants

International Nuclear Information System (INIS)

Bacher, P.

2004-01-01

Nuclear power plays an important role in the world, European (33%) and French (75%) power generation. This article aims at presenting in a synthetic way the main reactor types with their respective advantages with respect to the objectives foreseen (power generation, resources valorization, waste management). It makes a fast review of 50 years of nuclear development, thanks to which the nuclear industry has become one of the safest and less environmentally harmful industry which allows to produce low cost electricity: 1 - simplified description of a nuclear power generation plant: nuclear reactor, heat transfer system, power generation system, interface with the power distribution grid; 2 - first historical developments of nuclear power; 3 - industrial development and experience feedback (1965-1995): water reactors (PWR, BWR, Candu), RBMK, fast neutron reactors, high temperature demonstration reactors, costs of industrial reactors; 4 - service life of nuclear power plants and replacement: technical, regulatory and economical lifetime, problems linked with the replacement; 5 - conclusion. (J.S.)

Performance of CASTOR {sup registered} HAW cask cold trials for loading, Transport and storage of HAW canisters

Energy Technology Data Exchange (ETDEWEB)

Wilmsmeier, Marco; Horn, Thomas; Graf, Wilhelm [GNS Gesellschaft fuer Nuklear-Service mbH (Germany)

2009-07-01

With over 30 years of experience in the design, manufacturing, assembly and loading of CASTOR {sup registered} casks, GNS is one of the worldwide leading suppliers of casks for the transport and storage of spent fuel assemblies as well as for canisters with vitrified high active wastes (meanwhile over 1.000 casks loaded and stored and more than 1.500 ordered). GNS's products are used at around 30 sites worldwide for a wide range of inventories from pressurised and boiling water reactor fuels (PWR, VVER and BWR, RBMK), thorium high-temperature reactor fuels (THTR) and research reactor fuels (MTR) to vitrified high active wastes (HAW) from reprocessing plants. GNS is responsible for all nuclear wastes resulting from German Nuclear Power Plants and assists and/or performs in the loading and dispatch of CASTOR {sup registered} casks as well as their transport to and storage at central interim storage facilities and local interim storage areas. (orig.)

Critical analysis of major incidents risks in civil nuclear energy

International Nuclear Information System (INIS)

2000-09-01

The differences existing between the PWR type reactors and the RBMK type reactors are explained as well as the risk associated to each type when it exists. The Ines scale, tool to give the level of an accident gravity comprises seven levels, the number seven is the most serious and corresponds to the Chernobyl accident; The number zero is of no consequence but must be mentioned as a matter of form. The incidents from 1 to 3 concern increasing incidents, affecting the nuclear power plant but not the external public. The accidents from 4 to 7 have a nature to affect the nuclear power plant and the environment. An efficient tool exists between nuclear operators it is made of the reports on incidents encountered by close reactors. Two others type reactors are coming, the high temperature type reactors and the fast neutrons reactors. different risks are evoked, terrorism, proliferation, transport and radioactive wastes. (N.C.)

Investigation of CTF void fraction prediction by ENTEK BM experiment data

International Nuclear Information System (INIS)

Hoang Minh Giang; Hoang Tan Hung; Nguyen Phu Khanh

2015-01-01

Recently, CTF, a version of COBRA-TF code is reviewed to validate its simulation models by several experiments such as Castellana 4x4 rod bundle, EPRI 5x5 bundle tests, PSBT bundle tests and TPTF experiment. These above experiments provide enthalpy, mass flux (Castellana), temperature (EPRI) and void fraction (PSBT, TPTF) at exit channel only. In order to simulate PWR rod bundle flow behavior, it is necessary to review CTF with more experiment in high pressure condition and it is found that the ENTEK BM facility is suitable for this purpose. The ENTEK BM facility is used to simulate Russia RBMK and VVER rod bundle two phase flow with pressure at 3 and 7 MPa and it gives measured void fraction distribution along the channel. This study focus on two points: (a) accuracy assessment between CTF void fraction distribution predictions versus experiment void fraction distributions and (b) investigation of void fraction prediction uncertainty from propagation of input deviations caused by measured accuracy. (author)

Inspection of fuel elements in the cooling pond of a research reactor

International Nuclear Information System (INIS)

Pavlov, S.V.; Mestnikov, A.V.

1992-01-01

Nondestructive testing methods for fuel bundles and fuel elements in the cooling ponds of atomic power plants, using special inspection stands, have come into widespread use during the past decade. This paper describes a methodological stand that was built for the laboratory development of methods and individual units of inspection stands for fuel bundles of RBMK and VVER-1000 reactors. A complex of equipment was developed for the study of irradiated fuel elements, thus creating a methodological base for developing techniques for nondestructive testing of irradiated fuel elements and equipment to obtain information about the state of the fuel elements in a reactor expeditiously. The time required to inspect a fuel element can be shortened using some techniques simultaneously. The length of a fuel element can be measured simultaneously with visual inspection, eddy-current flaw detection can be preformed at the same time as the tranverse size of the fuel element is being determined. 6 refs., 5 figs

Safety assurance in radioactive waste management at nuclear power plants of the Northwest region of Russia

Energy Technology Data Exchange (ETDEWEB)

Safonov, Igor

1999-07-01

This presentation describes the two large operating nuclear power plants (NPP) in Northwest Russia, the Kola NPP and the Leningrad NPP. The four units at Kola are tank-type pressurised water reactors of 440 MW (electric) while the four Leningrad reactors are 1000 MW (electric) of RBMK type. Gosatomnadzor of Russia regularly conducts so-called target inspections on safety assurance for radioactive waste management at NNP. Among the many items checked during such inspections are the existence and realisation of an action plan for waste reduction, the technical state of equipment and the compliance with previous directions. The management of liquid, solid and gaseous radioactive wastes is described in some detail, and so are the violations revealed at both sites. There is also some discussion of modernisation plans for waste management. It is stated that the ecological impact of the plants is negligible and there is no hazard to people or environment. The presentation concludes with some suggestions for improving the licensing requirements for waste management.

Safety assurance in radioactive waste management at nuclear power plants of the Northwest region of Russia

International Nuclear Information System (INIS)

Safonov, Igor

1999-01-01

This presentation describes the two large operating nuclear power plants (NPP) in Northwest Russia, the Kola NPP and the Leningrad NPP. The four units at Kola are tank-type pressurised water reactors of 440 MW (electric) while the four Leningrad reactors are 1000 MW (electric) of RBMK type. Gosatomnadzor of Russia regularly conducts so-called target inspections on safety assurance for radioactive waste management at NNP. Among the many items checked during such inspections are the existence and realisation of an action plan for waste reduction, the technical state of equipment and the compliance with previous directions. The management of liquid, solid and gaseous radioactive wastes is described in some detail, and so are the violations revealed at both sites. There is also some discussion of modernisation plans for waste management. It is stated that the ecological impact of the plants is negligible and there is no hazard to people or environment. The presentation concludes with some suggestions for improving the licensing requirements for waste management

Current state of spent fuel management in the Russian Federation

International Nuclear Information System (INIS)

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

1998-01-01

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

A defence in depth approach to safety assessment of existing nuclear power plant

International Nuclear Information System (INIS)

Butcher, P.; Holloway, N.J.

1998-01-01

The safety assessment of plant built to earlier standards requires an approach to prioritisation of upgrades that is based on sound engineering and safety principles. The principles of defence in depth are universally accepted and can form the basis of a prioritisation scheme for safety issues, and hence for the upgrading required to address them. The described scheme includes criteria for acceptability and issue prioritisation that are based on the number of lines of defence and the consequences of their failure. They are thus equivalent in concept to risk criteria, but are based on deterministic principles. This scheme has been applied successfully to the RBMK plant at Ignalina in Lithuania, for which a Western-style Safety Analysis Report has recently been produced and reviewed by joint Western and Eastern teams. An extended Safety Improvement Programme (SIP2) has been developed and agreed, based on prioritisations from the defence in depth assessment. (author)

Failed fuel monitoring at nuclear power plants with RBMK reactors: operating parameters, requirements and decision making criteria

International Nuclear Information System (INIS)

Zhukov, I.V.

1993-01-01

The procedure for estimating the number of failed fuel rods in the core and the prediction of their discharge efficiency during operation is presented. The procedure is based on the FFM data base and the I-131 and Xe-133 coolant activity. (author)

Effect of local automatic control rods on three-dimensional calculations of the power distribution in an RBMK

International Nuclear Information System (INIS)

Pogosbekyan, L.R.; Lysov, D.A.; Bronitskii, L.L.

1993-01-01

Numerical simulators and information systems that support nuclear reactor operators must have fast models to estimate how fuel reloads and control rod displacement affect neutron and power distributions in the core. The consequences of reloads and control rod displacement cannot be evaluated correctly without considering local automatic control-rod operations in maintaining the radial power distribution. Fast three-dimensional models to estimate the effects of reloads and displacement of the control and safety rods have already been examined. I.V. Zonov et al. used the following assumptions in their calculational model: (1) the full-scale problem could be reduced a three-dimensional fragment of a locally perturbed core, and (2) the boundary conditions of the fragment and its total power were constant. The last assumption considers approximately how local automatic control rods stabilize the radial power distribution, but three dimensional calculations with these rods are not considered. These assumptions were introduced to obtain high computational speed. I.L. Bronitskii et al. considered in more detail how moving the local automatic control rods affect the power dimensional in the three-dimensional fragment, because, with on-line monitoring of the reload process, information on control rod positions is periodically renewed, and the calculations are done in real time. This model to predict the three-dimensional power distribution to (1) do a preliminary reload analysis, and (2) prepare the core for reloading did not consider the effect of perturbations from the local automatic control rods. Here we examine a model of a stationary neutron distribution. On one hand it gives results in an acceptable computation time; on the other it is a full-scale three-dimensional model and considers how local automatic control rods affect both the radial and axial power distribution

Upgrading safety of NPPs with RBMK-1000 reactors by implementation of the first priority measures and activities

International Nuclear Information System (INIS)

1996-01-01

After the accident at the Chernobyl Unit 4 reactor, extensive debates were in place about the future of nuclear power industry, its safety and the role of nuclear power in human life. The major conclusion drawn from those discussions is that the energy demands and ecological problems could not be resolved without further development of nuclear industry. However, the continued development of nuclear power industry, first and foremost, should rest on a wide range of actions aimed at assuring the quality of design and construction of new NPPs, the quality of operation of the existing plants and by means of their backfitting. 1 ref., 3 figs, 1 tab

Specific features and techneiueq improving the reinforced s forenclosing structures for NPP with the RBMK-type reactors

International Nuclear Information System (INIS)

Kirillov, A.P.; Konviz, V.S.; Semenov, V.P.

1982-01-01

A study is made on a new technique of construction of NPP enclosing structures (walls and ceilings) by the assembled- monolith method. Its advantages over the monolith method are presented. The design philosophy of the assembled-monolith constructions is described. It is shown on the basis of the experience of Kursk and Chernobylsk NPP construction that during the introduction of assembled-monolith constructions of massive walls and ceilings the speed of reactor blocks construction increased practically twice

Several perspectives on water-chemical cycles for nuclear power stations equipped with type VVER and RBMK reactors

International Nuclear Information System (INIS)

Mamet, A.P.; Mamet, V.A.; Pashevich, V.I.; Nazarenko, P.N.

1982-01-01

Water-chemical cycles for loops I and II of VVER reactors are discussed. These cycles are mixed ammonia-sodium with a variable concentration of boric acid and ammonia hydrazine with a pH factor of 9.1 +/- 0.1. New water-chemical cycles are considered for use in both existing and new nuclear power plants. Application of these new water-chemical cycles showed produce a significant improvement in operating conditions of nuclear power plants. Upon accumulation of sufficient operating experience with these cycles, it should be possible to raise the issue of revising applicable standard documentation

Chernobyl - a Canadian technical perspective

International Nuclear Information System (INIS)

Snell, V.G.; Howieson, J.Q.

1987-01-01

On April 26, 1986, the Number 4 reactor at the Chernobyl Nuclear Power Station in the Soviet Union suffered a severe accident which destroyed the reactor core and led to a loss of life. The four reactors at this station are of the RBMK-1000 type - boiling-light-water cooled, graphite moderated, vertical pressure-tube reactors, each generating 1000 MW of electricity through two turbines. AECL has carefully studied the accident, and the design of Chernobyl, to see if anything has been overlooked in the CANDU design. This report reviews the results of that study, in particular the relevant features of the Chernobyl design which exacerbated the accident, and compares them to the CANDU 600 design. A number of issues (the sign of the void coefficent and the pressure-tube design) have also been given some international prominence in the post-Chernobyl analysis; these are discussed in this report and shown to be irrelevant to the CANDU design. Finally this report describes the subjects identified for further design follow-up in Canada

International measures for supporting the Ukraine in decommissioning Chernobyl nuclear power plant

International Nuclear Information System (INIS)

Wolf, J.

2006-01-01

The destruction of Block 4 of the Ukranian nuclear power plant in Chernobyl on 26 April 1986 was the largest and most momentous accident in the civil use of nuclear energy. Its far-reaching and lasting ecological, heath-related and economic effects confronted the then Soviet and later the Ukraine with grave problems. Particularly after the dissolution of the Eastern Bloc and the emergence of information about the safety shortcomings of RBMK-type (Chernobyl-type) reactors the Western states pressed for the decommissioning of these reactors. At the G7 summit in Naples in 1994 the Ukraine was offered an action plan of support if it were willing to close down Chernobyl nuclear power plant. This initiative led to the signing on 20 December 1995 of a Memorandum of Understanding on the Closure of Chernobyl Nuclear Power Plant between the G7 states, the European Commission and the Ukraine. It contained an assurance by President Kuchma that Chernobyl nuclear power plant would be closed by the year 2000

Swiss operating experience: availability and post-Chernobyl upgrading

International Nuclear Information System (INIS)

Wenger, H.

1988-01-01

Switzerland started its era of nuclear power with the foundation stone for the country's first nuclear power unit (Beznau-1) onSeptember 6, 1965. Up to that date, Switzerland was the classic country for hydropower, negligible amounts of electricity being produced by fossil-fuelled plants. Today, nuclear accounts for close to 40 % of Swiss total electricity generation. Whwn credits for lifetime capacity factors of each individual plant are combined, Switzerland tops the world list for light water reactor performance over many years. The Chernobyl reactor type RBMK-1000 has very little in common with the light water reactors operating in Switzerland, so one would certainly not expect any direct influence on Swiss plant design, operation or maintenance as an immediate consequence of the accident. Some important safety measures against severe accidents are currently being implemented. These measures were not a direct outcome of the Chernobyl accident and were already in discussion quite some time before. With this action, the proper position of nuclear power to meet the ever increasing demand for electricity in Switzerland will hopefully again find greater public acceptance. 1 tab

Effects of B4C control rod degradation under severe accident

Energy Technology Data Exchange (ETDEWEB)

Seo, Si-Won; Park, Sang-Gil; Han, Sang-Ku [Atomic Creative Technology Co., Daejeon (Korea, Republic of)

2016-10-15

Boron carbide (B4C) is widely used as absorber material in western boiling water reactor (BWR), some PWR, EPR and Russian RBMK and VVERs. B4C oxidation is one of the important phenomena of in-vessel. In the present paper, the main results and knowledge gained regarding the B4C control rod degradation from above mentioned experiments are reviewed and arranged to inform its significance on the severe accident consequences. In this paper, the role of B4C control rod oxidation and the subsequent degradation on the severe accident consequences is reviewed with available literature and report of previous experimental program regarding the B4C oxidation. From this review, it seems that the contribution of this B4C oxidation on the accident progression to the further severe accident situation is not negligible. For the future work, the extensive experimental data interpretation will be performed to assess quantitatively the effect of B4C oxidation and degradation on the various postulated severe accident conditions.

Thermophysical instruments for non-destructive examination of tightness and internal gas pressure or irradiated power reactor fuel rods

International Nuclear Information System (INIS)

Pastoushin, V.V.; Novikov, A.Yu.; Bibilashvili, Yu.K.

1998-01-01

The developed thermophysical method and technical instruments for non-destructive leak-tightness and gas pressure inspection inside irradiated power reactor fuel rods and FAs under poolside and hot cell conditions are described. The method of gas pressure measuring based on the examination of parameters of thermal convection that aroused in gas volume of rod plenum by special technical instruments. The developed method and technique allows accurate value determination of not only one of the main critical rod parameters, namely total internal gas pressure, that forms rod mean life in the reactor core, but also the partial pressure of every main constituent of gaseous mixture inside irradiated fuel rod, that provides the feasibility of authentic and reliable leak-tightness detection. The described techniques were experimentally checked during the examination of all types power reactor fuel rods existing in Russia (WWER, BN, RBMK) and could form the basis for new technique development for non-destructive examination of PWR (and other) type rods and FAs having gas plenum filled with spring or another elements of design. (author)

Development of heat treated Zr-2.5% Nb alloy tubes for pressure tubes

International Nuclear Information System (INIS)

Saibaba, N.; Jha, S.K.; Tonpe, S.

2011-01-01

Zr-2.5% Nb alloy is the candidate material for pressure tubes of Pressurized Heavy Water Reactors (PHWR), and are manufactured in cold working condition while heat treated pressure tubes are used in RBMK and FUGEN type of reactors. The diametral creep of these tubes is the life limiting factor. This paper presents the extensive work carried out for the optimization of process parameters to manufacture heat treated Zr-2.5% Nb pressure tubes. Extensive dilactometry study was carried out to establish the transus temperature for the alloy and the effect of soaking temperature and cooling rate on the microstructure was characterized. On the basis of the study, water quenching (at 883 deg C) in the a b region with 20-25% primary a phase was selected, further cold worked, aged and finally autoclaved. Mechanical properties of the finished tubes were found to be comparable to the cold worked route. Large number of full sized tubes of about 700 - 800 mm long was produced to establish the repeatability. (author)

Nuclear reactor types

International Nuclear Information System (INIS)

Jones, P.M.S.

1987-01-01

The characteristics of different reactor types designed to exploit controlled fission reactions are explained. Reactors vary from low power research devices to high power devices especially designed to produce heat, either for direct use or to produce steam to drive turbines to generate electricity or propel ships. A general outline of basic reactors (thermal and fast) is given and then the different designs considered. The first are gas cooled, including the Magnox reactors (a list of UK Magnox stations and reactor performance is given), advanced gas cooled reactors (a list of UK AGRs is given) and the high temperature reactor. Light water cooled reactors (pressurized water [PWR] and boiling water [BWR] reactors) are considered next. Heavy water reactors are explained and listed. The pressurized heavy water reactors (including CANDU type reactors), boiling light water, steam generating heavy water reactors and gas cooled heavy water reactors all come into this category. Fast reactors (liquid metal fast breeder reactors and gas cooled fast reactors) and then water-cooled graphite-moderated reactors (RBMK) (the type at Chernobyl-4) are discussed. (U.K.)

Cluster-cell calculation using the method of generalized homogenization

International Nuclear Information System (INIS)

Laletin, N.I.; Boyarinov, V.F.

1988-01-01

The generalized-homogenization method (GHM), used for solving the neutron transfer equation, was applied to calculating the neutron distribution in the cluster cell with a series of cylindrical cells with cylindrically coaxial zones. Single-group calculations of the technological channel of the cell of an RBMK reactor were performed using GHM. The technological channel was understood to be the reactor channel, comprised of the zirconium rod, the water or steam-water mixture, the uranium dioxide fuel element, and the zirconium tube, together with the adjacent graphite layer. Calculations were performed for channels with no internal sources and with unit incoming current at the external boundary as well as for channels with internal sources and zero current at the external boundary. The PRAKTINETs program was used to calculate the symmetric neutron distributions in the microcell and in channels with homogenized annular zones. The ORAR-TsM program was used to calculate the antisymmetric distribution in the microcell. The accuracy of the calculations were compared for the two channel versions

Kursk ASSET brings praise for plant operators

International Nuclear Information System (INIS)

Anon.

1992-01-01

An Assessment of Safety Significant Events Team (ASSET) from the International Atomic Energy Agency (IAEA) visited Kursk on 19-31 July 1992 - the first time such a group had been to a RBMK reactor site. Kursk is a four unit station and the IAEA Team was able to consider safety significant events over its 45 reactor-year history. all four units have good operating records, with lifetime load factors of 70-80%. The ASSET's aim was to consider the plants current safety provisions for prevention of accidents and incidents. ASSET recommendations may cover the design and operability of the plant, personnel or operating procedures. At Kursk the Team found that ''highly qualified plant management and a very dedicated and knowledgeable operating staff'' was a major asset. They found a sound maintenance programme although some aspects of the facility appeared neglected. This was attributed to Russian industrial priorities that focused on functionality rather than appearance. it was in the test and maintenance personnel that the ASSET mission found the safety culture weakest. Some of their recommendations are reported in this article. (Author)

East European nuclear power plant review

International Nuclear Information System (INIS)

Thomas, Steve

1993-01-01

Western public opinion regards East European nuclear power plants as inefficient and dangerous. However the plants achieve consistently good operating performances. The load factors achieved by each type of plant by country in 1991 are tabulated. These are shown to be good, especially the Hungarian plant. Load factors seem to be dependent on the type of plant rather than where they were installed. WWER 213s worked better than the WWER 320s. This was because of long shutdowns to try and bring the safety standards up to acceptable levels. RBMK performances were depressed because of a 30% derating by safety authorities on 8 out of the 15 units operating. Overall the picture in Eastern Europe is encouraging with improvements in safety related indicators such as break-down frequency whilst the plants still achieve respectable load factors. The performance of the WWER 320s is particularly encouraging. Good load factors from this type of plant in Russia, the Ukraine and Bulgaria may allow older unsafe plant to be phased out. (UK)

Nuclear power in the Ukraine: Problems and prospects

International Nuclear Information System (INIS)

Nigmatullin, N.R.

1995-01-01

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

The Chernobyl reactor accident and its effects on the Bremen area

International Nuclear Information System (INIS)

Fischer, H.; Moser, D.; Urbach, M.

1986-01-01

Chapter 2 of the report gives an outline of the design of the RBMK-1000 reactor and its inventory of radionuclides at the time the accident happened, together with a brief scenario of possible events leading to the accident, and an assessment of total radionuclide release. Chapter 3 explains the measurement campaigns made in the Bremen area in the given period and the consequences to be drawn from measured data up to present time. The measuring campaigns are described by a full-test report, graphical illustration, and a table of measured data. The information covers all data collected from onset of radioactivity release up to the 9th of Sept. 1986. Chapter 4 describes the assessment of dose commitment by the Bremen population, on the basis of measured radionuclide concentrations in the environment. Chapter 5 discusses the possible health hazard to the population in accordance with current knowledge of radiation exposure and its effects. Chapter 6 summarizes and interprets the results, and chapter 8 presents definitions of concepts and terminology. (orig./HP) [de

French-Finnish colloquium on safety of French and Russian type nuclear power plants

International Nuclear Information System (INIS)

Lukka, M.; Jaervinen, M.; Minkkinen, P.; Ukkola, A.; Levomaeki, L.

1994-01-01

The French-Finnish Colloquium on Safety of French and Russian Type Nuclear Power Plants was held in June, 14th - 16th, 1994, in Lappeenranta, Finland. The main topics of the colloquium were: VVER and RBMK reactors; Industrial safety studies for VVER's in FRAMATOME; Structural safety analysis of Ignalina NPP; Thermalhydraulic system (BETHSY) and analytical experiments for French NPP; Test facilities simulating VVER plants during accidents; PACTEL - facility for VVER thermal hydraulics; High burn-up fuel and reactivity accidents; Overview of severe accident research at Nuclear Protection and Safety Institute of CEA; Research of severe accidents in Finland; Review of main activities concerning computer codes used for VVER thermal-hydraulic safety analysis in OKB Gidropress; CATHARE code; APROS computer code, new developments; TRIO and TOLBIAC computer codes; ESTET and N3S softwares; HEXTRAN - 3D reactor dynamics code for VVER accident analysis; An overview the boron dilution issue in PWRs; Boron mixing transients in a 900 MW PWR vessel for a reactor start-up operation; and Problem of boric acid dilution in IVO

Atomic science and engineering in the economy of the Soviet Union

International Nuclear Information System (INIS)

Kruglov, A.K.

1976-01-01

The main achievements of Soviet atomic science and engineering are presented. Even now, due to the development of the atomic industry, it is possible to produce at atomic stations cheaper energy in kWh cost than at thermal electrical stations. The successful operation of the VVER reactor at the Novo Voronezh Atomic Station and the RBMK reactor at the Leningrad Atomic Station, makes it possible to proceed to the development of more economic thermal reactors with a unit power over 1,500,000 kW. Methods are analysed allowing the atomic industry to be supplied with cheap nuclear fuel on the basis of poor uranium ores. The introduction of radioactive isotopes into the national economy has allowed a number of industries to automate control, to improve technologies and safety measures, etc. Isotopes are being more and more widely used in medicine. Some aspects are considered of using nuclear explosions in the gas and oil industry, in constructing hydraulic engineering works and creating places for the disposal of harmful or radioacmive wastes

The nuclear safety account and the Chernobyl nuclear power plant

International Nuclear Information System (INIS)

Maltini, F.

1996-01-01

In 1993, the G-7 officially proposed that the European Bank for Reconstruction and Development set up the Nuclear Safety Account (NSA) and act as the Account's secretariat. The Bank's Board of Directors approved this proposal and the Rules of the NSA on 22 March 1993 and the NSA became effective on 14 April 1993. The NSA finances, through grants, operational and near-term technical safety improvements for Soviet-designed nuclear reactors in the countries of the former Soviet Union, central and eastern Europe. Priority is given to those reactors which present the highest level of risk that can be significantly reduced by short-term and cost-effective safety improvements, and which are necessary to ensure the continuing electricity supply in the region. Efforts are therefore focused on WWER 440/230 and RBMK types of reactors and on the purchase of equipment as opposed to studies, which a number of donors already fund. Finance from the NSA is not used to extend the operating lifetime of unsafe reactors

Quantifying uncertainties in the estimation of safety parameters by using bootstrapped artificial neural networks

Energy Technology Data Exchange (ETDEWEB)

Secchi, Piercesare [MOX, Department of Mathematics, Polytechnic of Milan (Italy); Zio, Enrico [Department of Energy, Polytechnic of Milan, Via Ponzio 34/3, 20133 Milano (Italy)], E-mail: enrico.zio@polimi.it; Di Maio, Francesco [Department of Energy, Polytechnic of Milan, Via Ponzio 34/3, 20133 Milano (Italy)

2008-12-15

For licensing purposes, safety cases of Nuclear Power Plants (NPPs) must be presented at the Regulatory Authority with the necessary confidence on the models used to describe the plant safety behavior. In principle, this requires the repetition of a large number of model runs to account for the uncertainties inherent in the model description of the true plant behavior. The present paper propounds the use of bootstrapped Artificial Neural Networks (ANNs) for performing the numerous model output calculations needed for estimating safety margins with appropriate confidence intervals. Account is given both to the uncertainties inherent in the plant model and to those introduced by the ANN regression models used for performing the repeated safety parameter evaluations. The proposed framework of analysis is first illustrated with reference to a simple analytical model and then to the estimation of the safety margin on the maximum fuel cladding temperature reached during a complete group distribution header blockage scenario in a RBMK-1500 nuclear reactor. The results are compared with those obtained by a traditional parametric approach.

Safety culture in an RBMK perspective; Sostoyanie i perspektivy razvitiya sistemy povysheniya kul`tury bezopasnosti AEhS Rossii

Energy Technology Data Exchange (ETDEWEB)

Porokhin, V G [Rosenergoatom, Moscow (Russian Federation)

1997-12-31

The presentation discusses the following issues: state and perspectives of the development of the system on Russia NPP safety culture enhancement; steps of Rosehnergoatom on development of system on Russia NPP Russia NPP safety culture enhancement, qualitative and quantitative evaluation of the safety culture, the nearest perspectives on safety culture enhancement in Russia.

Radioactive and other environmental threats to the United States and the Arctic resulting from past Soviet activities

International Nuclear Information System (INIS)

1993-01-01

Earlier this year the Senate Intelligence Committee began to receive reports from environmental and nuclear scientists in Russia detailing the reckless nuclear waste disposal practices, nuclear accidents and the use of nuclear detonations. We found that information disturbing to say the least. Also troubling is the fact that 15 Chernobyl style RBMK nuclear power reactors continue to operate in the former Soviet Union today. These reactors lack a containment structure and they're designed in such a way that nuclear reaction can actually increase when the reactor overheats. As scientists here at the University of Alaska have documented, polar air masses and prevailing weather patterns provide a pathway for radioactive contaminants from Eastern Europe and Western Russia, where many of these reactors are located. The threats presented by those potential radioactive risks are just a part of a larger Arctic pollution problem. Every day, industrial activities of the former Soviet Union continue to create pollutants. I think we should face up to the reality that in a country struggling for economic survival, environment protection isn't necessarily the high priority. And that could be very troubling news for the Arctic in the future

Shadow corrosion testing in the INCA facility in the Studsvik R2 reactor

International Nuclear Information System (INIS)

Nystrand, A.C.; Lassing, A.

1999-01-01

Shadow corrosion is a phenomenon which occurs when zirconium alloys are in contact with or in proximity to other metallic objects in a boiling water reactor environment (BWR, RBMK, SGHWR etc.). An enhanced corrosion occurs on the zirconium alloy with the appearance of a 'shadow' of the metallic object. The magnitude of the shadow corrosion can be significant, and is potentially limiting for the lifetime of certain zirconium alloy components in BWRs and other reactors with a similar water chemistry. In order to evaluate the suitability of the In-Core Autoclave (INCA) in the Studsvik R2 materials testing reactor as an experimental facility for studying shadow corrosion, a demonstration test has been performed. A number of test specimens consisting of Zircaloy-2 tubing in contact with Inconel were exposed in an oxidising water chemistry. Some of the specimens were placed within the reactor core and some above the core. The conclusion of this experiment after post irradiation examination is that it is possible to use the INCA facility in the Studsvik R2 reactor to develop a significant level of shadow corrosion after only 800 hours of irradiation. (author)

From Core to Capture: Graphite Management by Gasification and Carbon Capture & Storage (CCS)

International Nuclear Information System (INIS)

Goodwin, J.; Bradbury, D.; Black, S.; Tomlinson, T.; Livesey, B.; Robinson, J.; Lindberg, M.; Newton, C.; Jones, A.; Wickham, A.

2016-01-01

Radioactive graphite waste arises principally from the moderators of graphite/gas-cooled reactors at the end of life of the reactors. Commercial power producing reactors (for example, Magnox, AGR and RBMK) have graphite moderators, each containing several thousand tonnes of graphite, with the UK having the largest inventory of over 90,000 tonnes. Additionally, there are smaller quantities of graphite arising from other sources such as fuel element components. The current long term strategy for management of reactor graphite in the UK is for these wastes to be conditioned for disposal followed by transfer to a geological disposal facility (GDF). With this baseline position, these wastes will account for about 30% of the ILW inventory in a GDF. As the volume of the graphite waste is so large, it is not currently economic to retrieve and process the graphite in advance of the availability of a geological disposal facility. Recent work by the NDA has ascribed a much smaller “incremental” volume of 2% due to graphite, calculated on the basis that the GDF has to be a certain size anyway in order to dissipate the decay heat from high level waste

Scaling and uncertainty in BWR instability problems

International Nuclear Information System (INIS)

Di Auria, F.; Pellicoro, V.

1995-01-01

This paper deals with a critical review of activities, performed at the DCMN of Pisa University, in relation to the thermo-hydraulic oscillations in two-phase systems. Stability analyses, including model development and achievement of experimental data, are generally performed for BWRs in order to achieve the following objectives: to reach a common understanding in relation to the predictive capabilities of system codes and to the influence of various parameters on the instability; to establish a data base for the qualification of the analytical tools already or becoming available; to set-up qualified tools (code/models + nodalization + user assumption) suitable for predicting the unstable behaviour of the nuclear plants of interest (current BWR, SBWR, ABWR and RBMK). These considerations have been the basis for the following researches: 1) proposal of the Boiling Instability Program (BIP) (1) 2) evaluation of stability tests in PIPER-ONE apparatus (2) 3) coupled thermal-hydraulic and neutronic instabilities in the LaSalle-2 BWR plant (3) 4) participation to the NEA-OECD BWR Benchmark (4) The RELAP/MOD2 and RELAP5/MOD3 codes have been used. (author)

The accident at the Chernobyl' nuclear power plant and its consequences

International Nuclear Information System (INIS)

1986-08-01

The material is taken from the conclusions of the Government Commission on the causes of the accident at the fourth unit of the Chernobyl' nuclear power plant and was prepared by a team of experts appointed by the USSR State Committee on the Utilization of Atomic Energy. It contains general material describing the accident, its causes, the action taken to contain the accident and to alleviate its consequences, the radioactive contamination and health of the population and some recommendations for improving nuclear power safety. 7 annexes are devoted to the following topics: water-graphite channel reactors and operating experience with RBMK reactors, design of the reactor plant, elimination of the consequences of the accident and decontamination, estimate of the amount, composition and dynamics of the discharge of radioactive substances from the damaged reactor, atmospheric transport and radioactive contamination of the atmosphere and of the ground, expert evaluation and prediction of the radioecological state of the environment in the area of the radiation plume from the Chernobyl' nuclear power station, medical-biological problems. A separate abstract was prepared for each of these annexes. The slides presented at the post-accident review meeting are grouped in two separate volumes

Ecological radiation protection criteria for nuclear power

International Nuclear Information System (INIS)

Kryshev, I.I.

1993-01-01

By now a large quantity of radioactive hazards of all sizes and shapes has accumulated in Russia. They include RBMK, VVER, and BN (fast-neutron) nuclear power plants, nuclear fuel processing plants, radioactive waste dumps, ships with nuclear power units, etc. In order to evaluate the radioecological situation correctly, the characteristics of the radioactive contamination must be compiled in these areas with some system of criteria which will provide an acceptable level of ecological safety. Currently health criteria for radiation protection are, which are oriented to man's radiation protection, predominate. Here the concept of a thresholdless linear dose-response dependence, which has been confirmed experimentally only at rather high doses (above 1 Gy), is taken as the theoretical basis for evaluating and normalizing radiation effects. According to one opinion, protecting people against radiation is sufficient to protect other types of organisms, although they are not necessarily of the same species. However, from the viewpoint of ecology, this approach is incorrect, because it does not consider radiation dose differences between man and other living organisms. The article discusses dose-response dependences for various organisms, biological effects of ionizing radiation, and appropriate radiation protection criteria

New partnerships improve eastern European training

International Nuclear Information System (INIS)

Smith, G.; Zinger, V.; Kumar, A.; Jenkins, T.

1992-01-01

In 1991, General Physics International Engineering and Simulation (GPI) won a contract for the first Western-supplied full-scope simulator and training system for a Soviet-designed RBMK. GPI was chosen by the former Soviet Ministry of Nuclear Power and Industry in open competition with other Western simulation companies. The Leningrad power plant near St Petersburg will be the host site for the simulator and training system. The total training system consists of: A full-scope, high fidelity simulator featuring: replica panels; advanced RISC technology computers; a UNIX-based, graphical X-Window, Russian text, instructor station; engineering workstations; GPI simulation software technology; and UNIX simulation development tools. Hardware and software for one analytical simulator, including instructor station, engineering workstations and several remote PC stations. Hardware and software for an interactive graphical training system. Methodology for expert system training. A psychophysiological system. Training facility equipment, including TV monitors, video players and printers. Training in hardware, software and instructor skills. This project will be carried out in co-operation with several Russian partners. Similar co-operative agreements have also been set up with Czechoslovakia and Taiwan. (author)

Expert system for diagnostics and status monitoring of NPP water chemistry condition

International Nuclear Information System (INIS)

Shvedova, M.N.; Kritski, V.G.; Zakharova, S.V.; Benediktov, V.B.; Nikolaev, F.V.

2002-01-01

Water chemistry condition (WCC) has been the subject of constant study and improvement up to the present day. It is connected with the presence of a direct relationship between the violation of water chemistry regulation on the one hand and components reliability of the circuit's equipment and cost-effectiveness of their operation on the other. It dictates the necessity to apply different optimization methods in the field of monitoring and use of information analytical and diagnostic systems to assess WCC quality, control and support. LI ''VNIPIET'' employees have, for several years, been developing an expert diagnostic system for supporting WCC and status monitoring of RBMK - reactor NPPs [2]. This system has not only conveniently organized the traditional functions of information acquisition and storage, a complete presentation of information in the form of tables, graphs of a dynamical changes of parameters and formation regular reports, diagnostic functions and issuing recommendations on WCC correction, but it also allows the assessment of confidence in the diagnosis made, relying on a wide range of numerical estimates, which were calculated by the use of expert data, and to make a credible prediction of an existing situation development. (authors)

Simulating thermal behavior of AECL's spent fuel dry storage system with CATHENA

International Nuclear Information System (INIS)

Sabourin, G.

1998-01-01

This paper documents the comparisons between CATHENA predictions and temperature measurements taken at the Gentilly-2 NPP spent fuel dry storage facility and in a mock--up of a storage basket placed inside a storage cylinder. It also presents CATHENA temperature predictions related to the storage of spent fuel in MACSTOR modules as planned for Ignalina NPP, Lithuania. CATHENA has been chosen because it can simulate many noncondensable gases including air and helium, and because of its great flexibility in the representation of the MACSTOR module geometry. The results of the simulations show good agreement with the experimental measurements. The two comparisons indicate that CATHENA can be used to simulate heat transfer from the fuel to the external air circuit of the spent fuel dry storage system. For the Ignalina MACSTOR module, containing RBMK fuel having higher heat release than typical CANDU fuel, CATHENA predicts that the maximum fuel temperature is expected to be around 240 deg C, giving an acceptable margin below the maximum allowed temperature of 300 deg C. In conclusion, this paper shows that the thermalhydraulic code CATHENA can accurately predict the thermal behavior AECL's air cooled spent fuel dry storage system. (author)

Simulating thermal behavior of AECL's spent fuel dry storage system with CATHENA

Energy Technology Data Exchange (ETDEWEB)

Sabourin, G. [Atomic Energy of Canada Limited, Montreal, PQ (Canada)

1998-07-01

This paper documents the comparisons between CATHENA predictions and temperature measurements taken at the Gentilly-2 NPP spent fuel dry storage facility and in a mock--up of a storage basket placed inside a storage cylinder. It also presents CATHENA temperature predictions related to the storage of spent fuel in MACSTOR modules as planned for Ignalina NPP, Lithuania. CATHENA has been chosen because it can simulate many noncondensable gases including air and helium, and because of its great flexibility in the representation of the MACSTOR module geometry. The results of the simulations show good agreement with the experimental measurements. The two comparisons indicate that CATHENA can be used to simulate heat transfer from the fuel to the external air circuit of the spent fuel dry storage system. For the Ignalina MACSTOR module, containing RBMK fuel having higher heat release than typical CANDU fuel, CATHENA predicts that the maximum fuel temperature is expected to be around 240 deg C, giving an acceptable margin below the maximum allowed temperature of 300 deg C. In conclusion, this paper shows that the thermalhydraulic code CATHENA can accurately predict the thermal behavior AECL's air cooled spent fuel dry storage system. (author)

Strategy on radioactive waste management in Lithuania

International Nuclear Information System (INIS)

Poskas, P.; Adomaitis, J.E.

2003-01-01

In Lithuania about 70-80% of all electricity is generated at a single power station, Ignalian NPP which has two non-upgradable RBMK-1500 type reactors. The unit 1 will be closed by 2005. The decision on unit 2 should be made in Lithuanian Parliament very soon taking into consideration substantial long-term financial assistance from the EU, G7 and other states as well as international institutions. The Government approved the Strategy on Radioactive Waste Management in 2002. Objectives of this strategy are to develop the radioactive waste management infrastructure based on modern technologies and provide for the set of practical actions that shall bring management of radioactive waste in Lithuania in compliance with radioactive waste management principles of IAEA and with good practices in force in EU Member States. Ignalina NPP is undertaking a program of decommissioning support projects, financed by grants from the International Ignalina Decommissioning Support Fund, administered by the European Bank for Reconstruction and Development. This program comprises also the implementation of investment projects in a number of pre-decommissioning facilities including the management of radioactive waste and spent nuclear fuel. (orig.)

Radioactive and other environmental threats to the United States and the Arctic resulting from past Soviet activities

Energy Technology Data Exchange (ETDEWEB)

NONE

1993-12-31

Earlier this year the Senate Intelligence Committee began to receive reports from environmental and nuclear scientists in Russia detailing the reckless nuclear waste disposal practices, nuclear accidents and the use of nuclear detonations. We found that information disturbing to say the least. Also troubling is the fact that 15 Chernobyl style RBMK nuclear power reactors continue to operate in the former Soviet Union today. These reactors lack a containment structure and they`re designed in such a way that nuclear reaction can actually increase when the reactor overheats. As scientists here at the University of Alaska have documented, polar air masses and prevailing weather patterns provide a pathway for radioactive contaminants from Eastern Europe and Western Russia, where many of these reactors are located. The threats presented by those potential radioactive risks are just a part of a larger Arctic pollution problem. Every day, industrial activities of the former Soviet Union continue to create pollutants. I think we should face up to the reality that in a country struggling for economic survival, environment protection isn`t necessarily the high priority. And that could be very troubling news for the Arctic in the future.

The Performance of Major Plant Items at Calder Hall; Fonctionnement des elements principaux de la centrale de Calder Hall; Kharakteristika osnovnykh uzlov ustanovki v Kolder-Kholle; Rendimiento de los principales elementos de la central de Calder Hall

Energy Technology Data Exchange (ETDEWEB)

Desbruslais, E. L. [United Kingdom Atomic Energy Authority, Seascale, Cumberland (United Kingdom)

1963-10-15

. Opravdannoj yavlyaetsya ustanovka sdvoennogo vspomogatel'nogo oborudovaniya vmeste s ob{sup e}dineniem parovoj i vodnoj sistem. Oborudovanie B.C.D.G. okazalos' v vysshej stepeni udovletvoritel'nym. Vvidu togo, chto chuvstvitel'nost' gruppy kanalov vyshe chem neobkhodimo, v nastoyashchee vremya planirugtsya izmeneniya s tsel' ukorocheniya tsiklov za schet nekotoroj chuvstvitel'nosti. Obsuzhdayutsya meropriyatiya po pereoborudovaniyu sistemy podachi ehnergii dlya priborov reaktora i konturov bezopasnosti. Rekomenduyutsya klassifikatsiya stepeni opasnosti i vazhnosti i ustrojstvo po pervomu vyklyucheniyu reaktora. Vyyasnyaetsya neobkhodimost' imet' sistemu s dvumya zashchitnymi liniyami i ustanavlivat' tri ili bolee takikh linij. Statistika- pokazala polnuyu nadezhnost' oborudovaniya registratsii dannykh. Opravdyvaetsya neobkhodimost' bol'shoj standartizatsii oborudovaniya po peregruzke topliva. Ispol'zovanie spetsial'nykh ballonov i spetsial'nogo oborudovaniya po razgruzke trebuet mnogo vremeni, i ot ehtogo sleduet otkazat'sya. Opisyvayutsya izmeneniya/ vnesennye v mekhanizmy po zagruzke/razgruzke i vo vspomogatel'nuyu ustanovku. Podcherkivaetsya neobkhodimost' tshchatel'noj obrabotki topliva vo vremya zagruzki. Primenenie televizionnykh kamer v toplivnykh kanalakh i spetsial'nykh zakhvatov pozvolilo sokratit' vremya razgruzki. Nebol'shie neispravnosti otmecheny lish' pri ehkspluatatsii slozhnykh mekhanizmov upravlyayushchikh sterzhnej i sootvetstvuyushchego oborudovaniya po kontrolyu. Naskol'ko ehto vozmozhno, osnovnye sistemy upravleniya ustanovki dolzhny byt' tsentralizovany, i osnovnye tsentry upravleniya i uzly upravleniya ustanovki dolzhny byt' zashchishcheny ot sluchajnogo povrezhdeniya pri obluchenii vneshnimi istochnikami. Priznayutsya neobkhodimymi avtomaticheskie puskovoe i parallel'noe ustrojstva dlya avarijnoj dizel'noj ustanovki, odnako ustanovka mogla by byt' tsentralizovana s pol'zoj dlya dela. (author)

The Electronic Library of the Thermal Physical Databases

International Nuclear Information System (INIS)

Zhuravleva, Y.; Mingaleeva, G.; Mokrousov, K.; Yashnikov, D.

2008-01-01

Up-to-date quality assurance procedure requires the permanent verification of the best-estimate thermal-hydraulic system codes and the uncertainty analysis of results. Therefore, the researches need the growing up amount of the experimental data. Over the last years RDIPE has been carried out the verification of RELAP5/mod3.2 code and safety analysis for NPP with RBMK reactor. Moreover, these activities include both Russian (Puchok, Korsar, RATEG) and foreign codes (RELAP, MELCOR, ATHLET). Such activities require of the accumulation and the assessment of the large amount of experimental data. Electronic data base library was created in order to unify and keep the large amount of the primary experimental data. The special attention was given to completeness and sufficiency of information for modelling of the experiments. Generally this activity was carried out in the collaboration with the authors of experiment. First of all the experimental data for the additional verification of Russian and foreign codes relating to RBMK reactor safety analysis were included in the library. The following phenomena are specific and important: outflow from the main circulation circuit including critical flow of water, two phases mixture and vapour through the break, flow limiters, long channels with/ without local resistance and other circuit elements; thermal hydraulic process in reactor channels: pressure-drop, relative movement of phases, countercurrent flow, reflooding; heat transfer in fuel bundles including radiation heat transfer; heat transfer before and after critical heat flux transition in the rod bundle; variation of steam-water level in drum separator. These phenomena were studied at the test sites of KPI (Ukraine), Lithuanian Energy Institute, RDIPE (Russia), Russian Research Center 'Kurchatov Institute', EREC (Russia) and others. Transient modes data from operating power plants became the important part of the library. The authors of the electronic thermal physical

The Chernobyl NPP decommissioning: Current status and alternatives

International Nuclear Information System (INIS)

Mikolaitchouk, H.; Steinberg, N.

1996-01-01

After the Chernobyl accident of April 26, 1986, many contradictory decisions were taken concerning the Chernobyl nuclear power plant (NPP) future. The principal source of contradictions was a deadline for a final shutdown of the Chernobyl NPP units. Alterations in a political and socioeconomic environment resulted in the latest decision of the Ukrainian Authorities about 2000 as a deadline for a beginning of the Chernobyl NPP decommissioning. The date seems a sound compromise among the parties concerned. However, in order to meet the data a lot of work should be done. First of all, a decommissioning strategy has to be established. The problem is complicated due to both site-specific aspects and an absence of proven solutions for the RBMK-type reactor decommissioning. In the paper the problem of decommissioning option selection is considered taking into account an influence of the following factors: relevant legislative and regulatory requirements; resources required to carry out decommissioning (man-power, equipment, technologies, waste management infrastructure, etc.); radiological and physical status of the plant, including structural integrity and predictable age and weather effects; impact of planned activities at the destroyed unit 4 and within the 30-km exclusion zone of the Chernobyl NPP; planed use of the site; socio-economic considerations

Non-Destructive Techniques in the Tacis and Phare Nuclear Safety Programmes

International Nuclear Information System (INIS)

Bieth, Michel

2002-01-01

Decisions regarding the verification of design plant lifetime and potential license renewal periods involve a determination of the component and circuit condition. In Service Inspection of key reactor components becomes a crucial consideration for continued safe plant operation. The determination of the equipment properties by Non Destructive Techniques during periodic intervals is an important aspect of the assessment of fitness-for-service and safe operation of nuclear power plants The Tacis and Phare were established since 1991 by the European Union as support mechanisms through which projects could be identified and addressed satisfactorily. In Nuclear Safety, the countries mainly concerned are Russia, Ukraine, Armenia, and Kazakhstan for the Tacis programme, and Bulgaria, Czech Republic, Hungary, Slovak Republic, Lithuania, Romania and Slovenia for the Phare programme. The Tacis and Phare programs concerning the Nuclear Power Plants consist of: - On Site Assistance and Operational Safety, - Design Safety, - Regulatory Authorities, - Waste management, and are focused on reactor safety issues, contributing to the improvement in the safety of East European reactors and providing technology and safety culture transfer. The main parts of these programmes are related to the On-Site Assistance and to the Design Safety of VVER and RBMK Nuclear power plants where Non Destructive Techniques for In Service Inspection of the primary circuit components are addressed. (authors)

The nuclear technology development program in the U.S.S.R

International Nuclear Information System (INIS)

Lukonin, N.F.

1987-01-01

The trend of strategy on the nuclear power generation in USSR is not changed in spite of the accident in Chernobyl Nuclear Power Station. In 1986, the electric power generated by nuclear power generation was 162 billion kWh, and the heat supply by nuclear energy was 29 million Gcal. The development of nuclear power generation in USSR for 30 years proved that the atomic energy is technically omnipotent, and the economical substitution of the demand of fossil fuel with nuclear fuel is possible. As of January 1, 1987, 17 nuclear power stations were in operation in USSR, and the total power output was 31,000 MW. The share of nuclear power generation in the total electric power generation was 1/9. 11 nuclear power stations are under construction. The accelerating development of nuclear power generation is the base of meeting the electric power demand in the European region of USSR together with the power transmission from the eastern region. The nuclear power generation in USSR is based on two types of nuclear reactors, that is, water-water type VVER and water-graphite type RBMK. The accident in Chernobyl Nuclear Power Station and the situation thereafter are reported. The development of nuclear power generation in future is discussed. (Kako, I.)

Disposal and long-term storage in geological formations of solidified radioactive wastes

International Nuclear Information System (INIS)

Shischits, I.

1996-01-01

The special depository near Krasnoyarsk contains temporarily about 1,100 tons of spent nuclear fuel (SNF) from WWR- should be solidified and for the most part buried in geological formations. Solid wastes and SNF from RBMK reactors are assumed to be buried as well. For this purpose special technologies and underground constructions are required. They are to be created in the geological plots within the territory of Russian Federation and adjacent areas of CIS, meeting the developed list of requirements. The burial structures will vary greatly depending on the geological formation, the amount of wastes and their isotope composition. The well-known constructions such as deep wells, shafts, mines and cavities can be mentioned. There is a need to design constructions, which have no analog in the world practice. In the course of the Project fulfillment the following work will be conducted: -theoretical work followed by code creation for mathematical simulation of processes; - modelling on the base of prototypes made from equivalent materials with the help of simulators; - bench study; - experiments in real conditions; - examination of massif properties in particular plots using achievements of geophysics, including gamma-gamma density detectors and geo locators. Finally, ecological-economical model will be given for designing burial sites

International review of Kursk unit 1 in-depth safety analysis report

International Nuclear Information System (INIS)

Chouha, M.; Bolshov, L.; Butcher, P.; Janke, R.; Parsons, T.; Weber, J.P.

2004-01-01

The paper presents the objectives, organisation, main findings and conclusions of the international review of the Kursk unit 1 safety analysis report (K1IRSR). The K1IRSR was administered by RISKAUDIT IRSN/GRS international and carried out by international experts from 7 western countries plus the Russian Federation, under the supervision of the safety review group (SRG) of the European bank for reconstruction and development (EBRD). The project was financed by the nuclear safety account (NSA) administered by the EBRD. The Russian experts worked under a contract with IBRAE financed by Rosenergoatom. The main conclusions were that the SAR followed a correct approach, broadly in line with Russian and international guidance documents, but needed improvement in structure and content. It established that the safety level of the unit has been increased significantly by the modernisation programme. The important deviations of the unit from current Russian regulations and the IAEA safety issues for RBMK are either fully resolved or are being addressed to the extent possible by compensatory measures to further reduce the risk. The K1IRSR experts have made a number of recommendations for improvement of the K1SAR. The authors agreed to take the recommendations into account in future revision of the K1SAR. (orig.)

Degradation of aged plants by corrosion: 'Long cell action' in unresolved corrosion issues

International Nuclear Information System (INIS)

Saji, Genn

2009-01-01

In a series of previously published papers the author has identified that 'long cell action' corrosion plays a pivotal role in practically all unresolved corrosion issues for all types of nuclear power plants (e.g. PWR/VVER, BWR/RBMK and CANDU). Some of these unresolved issues are IGSCC, PWSCC, AOA and FAC (erosion-corrosion). In conventional corrosion science it is well established that 'long cell action' can seriously accelerate or suppress the local cell corrosion activities. Although long cell action is another fundamental mechanism of corrosion, especially in a 'soil corrosion' arena, potential involvement of this corrosion process has never been studied in nuclear and fossil power plants as far as the author has been able to establish. The author believes that the omission of this basic corrosion mechanism is the root cause of practically all un-resolved corrosion issues. In this paper, the author further elaborated on his assessment to other key corrosion issues, e.g. steam generator and turbine corrosion issues, while briefly summarizing previous discussions for completeness purposes, as well as introducing additional experimental and theoretical evidence of this basic corrosion mechanism. Due to the importance of this potential mechanism the author is calling for institutional review activities and further verification experiments in the form of a joint international project.

Multifunctional optimised scope simulators in Central and Eastern Europe

International Nuclear Information System (INIS)

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

1999-01-01

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

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.

Sensitivity and uncertainty analysis for Ignalina NPP confinement in case of loss of coolant accident

International Nuclear Information System (INIS)

Urbonavicius, E.; Babilas, E.; Rimkevicius, S.

2003-01-01

At present the best-estimate approach in the safety analysis of nuclear power plants is widely used around the world. The application of such approach requires to estimate the uncertainty of the calculated results. Various methodologies are applied in order to determine the uncertainty with the required accuracy. One of them is the statistical methodology developed at GRS mbH in Germany and integrated into the SUSA tool, which was applied for the sensitivity and uncertainty analysis of the thermal-hydraulic parameters inside the confinement (Accident Localisation System) of Ignalina NPP with RBMK-1500 reactor in case of Maximum Design Basis Accident (break of 900 mm diameter pipe). Several parameters that could potentially influence the calculated results were selected for the analysis. A set of input data with different initial values of the selected parameters was generated. In order to receive the results with 95 % probability and 95 % accuracy, 100 runs were performed with COCOSYS code developed at GRS mbH. The calculated results were processed with SUSA tool. The performed analysis showed a rather low dispersion of the results and only in the initial period of the accident. Besides, the analysis showed that there is no threat to the building structures of Ignalina NPP confinement in case of the considered accident scenario. (author)

Ignalina Safety Analysis Group

International Nuclear Information System (INIS)

Ushpuras, E.

1995-01-01

The article describes the fields of activities of Ignalina NPP Safety Analysis Group (ISAG) in the Lithuanian Energy Institute and overview the main achievements gained since the group establishment in 1992. The group is working under the following guidelines: in-depth analysis of the fundamental physical processes of RBMK-1500 reactors; collection, systematization and verification of the design and operational data; simulation and analysis of potential accident consequences; analysis of thermohydraulic and neutronic characteristics of the plant; provision of technical and scientific consultations to VATESI, Governmental authorities, and also international institutions, participating in various projects aiming at Ignalina NPP safety enhancement. The ISAG is performing broad scientific co-operation programs with both Eastern and Western scientific groups, supplying engineering assistance for Ignalina NPP. ISAG is also participating in the joint Lithuanian - Swedish - Russian project - Barselina, the first Probabilistic Safety Assessment (PSA) study of Ignalina NPP. The work is underway together with Maryland University (USA) for assessment of the accident confinement system for a range of breaks in the primary circuit. At present the ISAG personnel is also involved in the project under the grant from the Nuclear Safety Account, administered by the European Bank for reconstruction and development for the preparation and review of an in-depth safety assessment of the Ignalina plant

Comparison of technical and economical factors of 1000-MW steam turbines at 3000 and 1500 r.p.m. for nuclear power plants

International Nuclear Information System (INIS)

Markov, N.M.; Safonov, L.P.

1980-01-01

The problem of unification of the low-pressure cilinders (LPC) for turbo-generator units of nuclear power plants with power of 1000 MW on base of the WWER and RBMK type reactors is discussed. The results of the comparison of the K-1000-60/1500 and K-1000-60/3000 turbines in the thermal efficiency of flow passages and arrangements masses and dimensions, static and dynamic strength manoeurrability and reliability are given. To cerry out the correct comparison methods adoped as branch standards thermal calculations, calculation of low-potential part and thermal arrangements, calculations of temperature fields and of low cycle fatigue calculation of the erosion failure accumulation of blades calculation of the blades for the last steps have been used. A conclusion is made that in the nearest future it is necessary to produce the K-1000-60/1500 and K-1000-60/3000 turbines simultaneously. The low-speed lurbines with three LPC are preferable for the nuclear power plants with average annual temperatures of water up to 20 deg C and the high-speed turbines and the K-1000-60/1500 units with two LPC are expedient for nuclear power plants with temperatures higher than 20 deg C. Introduction of the turboplants with reduced number of LPC in the nuclear power engineering provides the increase of reliability, maintenance fitness and the decrease of building costs and transport expenses

Chernobyl operators mesmerized by mind-set

International Nuclear Information System (INIS)

Rippon, Simon.

1986-01-01

This post mortem report is based mainly on the information presented by Soviet specialists to a post accident review conference organized by the International Atomic Energy Agency (IAEA) in Vienna in August 1986, and an analysis of that information reported to a special session of the IAEA General Conference in early September 1986. The Chernobyl accident is blamed on operator mind-set - a situation where control room operators fix their minds on one interpretation of instrument readings and doggedly follow a set of procedures without fully appreciating what is actually happening in the plant. The background to, and the step-by-step sequence of, operator actions, the reasons for taking them, the operating procedures these violated and the consequences of the actions are described. The RBMK type reactor design is also described to show how the operations relate to this particular reactor design. The main reactor parameters are given. A diary of events, starting at 01.00h on 25 April 1986 and leading to explosions at about 01.24h on 26 April is set out. The damage to the reactor building and the subsequent actions to put out the fire caused by the explosions is described. By May 6 the actions had resulted in temperature stabilization at the reactor and a reduction to a low level of the radioactivity release. (UK)

Report of the review of WWER-1000 safety issues resolution at Temelin nuclear power plant, Temelin, Czech Republic 11 to 15 March 1996

International Nuclear Information System (INIS)

Almeida, C.; Hoehn, J.; Seiberling, R.; Chambon, J.L.; Fil, N.S.; Munoz, A.; Roennberg, G.; Wenk, W.

1996-01-01

At the request of the Government of the Czech Republic, the IAEA conducted, in the period of 11-15 March 1996, a mission to review the resolution of WWER-1000 safety issues at Temelin NPP. These safety issues have been identified for WWER-1000 model 320 reactors and ranked according to their importance to safety in the frameworks of the IAEA Extrabudgetary Programme on Safety of WWER and RBMK Nuclear Power Plants. The Temelin NPP is a WWER-1000 and was originally designed according to standards of the former Soviet Union. After a series of reviews in the 1980s, a decision was taken by the Temelin NPP management to upgrade the design of Temelin, including the supply of fuel and instrumentation and control by a western company. The objective of the mission was to review the response of Temelin to the safety issues identified by the IAEA. The mission assessed the current Temelin design, including proposed modifications and plans for operation at Temelin, in the light of the IAEA recommendations for each relevant issue. The present report contains the mission's general conclusions and recommendations and an overview of the review performed in each topical area. The attachment contains a brief summary of the discussions on each individual safety issue and associated conclusions and recommendations. 3 refs

Leak detection in the primary reactor coolant piping of nuclear power plant by applying beam-microphone technology

International Nuclear Information System (INIS)

Kasai, Yoshimitsu; Shimanskiy, Sergey; Naoi, Yosuke; Kanazawa, Junichi

2004-01-01

A microphone leak detection method was applied to the inlet piping of the ATR-prototype reactor, Fugen. Statistical analysis results showed that the cross-correlation method provided the effective results for detection of a small leakage. However, such a technique has limited application due to significant distortion of the signals on the reactor site. As one of the alternative methods, the beam-microphone provides necessary spatial selectivity and its performance is less affected by signal distortion. A prototype of the beam-microphone was developed and then tested at the O-arai Engineering Center of the Japan Nuclear Cycle Development Institute (JNC). On-site testing of the beam-microphone was carried out in the inlet piping room of an RBMK reactor of the Leningrad Nuclear Power Plant (LNPP) in Russia. A leak sound imitator was used to simulate the leakage sound under the leakage flow condition of 1-3 gpm (0.23-0.7 m 3 /h). Analysis showed that signal distortion does not seriously affect the performance of this method, and that sound reflection may result in the appearance of ghost sound sources. The test results showed that the influences of sound reflection and background noise were smaller at the high frequencies where the leakage location could be estimated with an angular accuracy of 5deg which is the range of localization accuracy required for the leak detection system. (author)

Ignalina plant licensing process, international co-operation and assistance

International Nuclear Information System (INIS)

Bystedt, P.

1999-01-01

The challenge for Lithuania as a country with regained independence was to perform a licensing review in a way never done before in the country and in a time schedule that was extremely short. The work included establishing of the licensing base, strengthening the regulatory authority and organising the technical support, establish and implement a safety improvement program, production of the safety case and review of the safety case, and to derive a conclusion regarding whether to issue a licence or not. This was to be done together with other tasks, such as implementation of modifications included in the safety improvement programme at Ignalina, implementation of a new storage for spent fuel and, most important of all, to manage the operational safety at the plant. The achievements are impressive seen in view of the point of start and in view of the time and resources that have been available. Lithuania has put forward a unique safety documentation of an RBMK reactor and presented an in-depth safety evaluation in full openness to Western experts, giving the unique possibility to compare the safety of the Ignalina reactors to Western standards. The co-operation that has been established between Lithuania and Western experts through different assistance programmes is of outmost value, for all involved parties. Co-operation should continue as one element of the challenges for the future

Analysis of the source term in the Chernobyl-4 accident

International Nuclear Information System (INIS)

Alonso, A.; Lopez Montero, J.V.; Pinedo Garrido, P.

1990-01-01

The report presents the analysis of the Chernobyl accident and of the phenomena with major influence on the source term, including the chemical effects of materials dumped over the reactor, carried out by the Chair of Nuclear Technology at Madrid University under a contract with the CEC. It also includes the comparison of the ratio (Cs-137/Cs-134) between measurements performed by Soviet authorities and countries belonging to the Community and OECD area. Chapter II contains a summary of both isotope measurements (Cs-134 and Cs-137), and their ratios, in samples of air, water, soil and agricultural and animal products collected by the Soviets in their report presented in Vienna (1986). Chapter III reports on the inventories of cesium isotopes in the core, while Chapter IV analyses the transient, especially the fuel temperature reached, as a way to deduce the mechanisms which took place in the cesium escape. The cesium source term is analyzed in Chapter V. Normal conditions have been considered, as well as the transient and the post-accidental period, including the effects of deposited materials. The conclusion of this study is that Chernobyl accidental sequence is specific of the RBMK type of reactors, and that in the Western world, basic research on fuel behaviour for reactivity transients has already been carried out

Accident analysis for nuclear power plants

International Nuclear Information System (INIS)

2002-01-01

Deterministic safety analysis (frequently referred to as accident analysis) is an important tool for confirming the adequacy and efficiency of provisions within the defence in depth concept for the safety of nuclear power plants (NPPs). Owing to the close interrelation between accident analysis and safety, an analysis that lacks consistency, is incomplete or is of poor quality is considered a safety issue for a given NPP. Developing IAEA guidance documents for accident analysis is thus an important step towards resolving this issue. Requirements and guidelines pertaining to the scope and content of accident analysis have, in the past, been partially described in various IAEA documents. Several guidelines relevant to WWER and RBMK type reactors have been developed within the IAEA Extrabudgetary Programme on the Safety of WWER and RBMK NPPs. To a certain extent, accident analysis is also covered in several documents of the revised NUSS series, for example, in the Safety Requirements on Safety of Nuclear Power Plants: Design (NS-R-1) and in the Safety Guide on Safety Assessment and Verification for Nuclear Power Plants (NS-G-1.2). Consistent with these documents, the IAEA has developed the present Safety Report on Accident Analysis for Nuclear Power Plants. Many experts have contributed to the development of this Safety Report. Besides several consultants meetings, comments were collected from more than fifty selected organizations. The report was also reviewed at the IAEA Technical Committee Meeting on Accident Analysis held in Vienna from 30 August to 3 September 1999. The present IAEA Safety Report is aimed at providing practical guidance for performing accident analyses. The guidance is based on present good practice worldwide. The report covers all the steps required to perform accident analyses, i.e. selection of initiating events and acceptance criteria, selection of computer codes and modelling assumptions, preparation of input data and presentation of the

On a possibility to ground a reliable and safe disposing of a spent nuclear fuel from nuclear reactors RBMK in deep boreholes

International Nuclear Information System (INIS)

Kedrovskij, O.L.

1998-01-01

In order to isolate a spent nuclear fuel (SNF), it is proposed to dispose it, after 30 years keeping on the day surface, in boreholes of up to 4 km depth and 350-1020 mm diameter drilled in low permeable platform basement crystalline rocks, that allows one to localize SNF radionuclides till their full decay. It is shown that the method requires relatively low investments and enables the volume of a burial being increased during wastes income. Along with consideration and assessment of hydrodynamic, geological and hydrogeological parameters of a rock massive and rocks preferable for the new method, engineering solutions for the borehole design and hermetizing structure and assessments of the technogeneous influence on the environment are given. The questions are also considered of a possible shortening of the terms of keeping SNF in a surface storage before their burying, that enables one to decrease expenses for surface storage constructions and to promote the cleaning of the sphere of human habitation from the most active and dangerous wastes

Nuclear power in eastern and central Europe. Background paper

International Nuclear Information System (INIS)

Myers, L.C.

1993-11-01

The breakup of the former Soviet Union and other political changes in eastern and central Europe have opened up the area to closer scrutiny than was previously possible. Because of the accident at Chernobyl, nuclear power is one of the subjects that western nations have had a great deal of interest in exploring. The former Soviet Union designed and/or helped build more than 60 civilian reactors in the region. Most of these reactors follow one of two distinctly different designs: the VVER, or pressurized water reactor series; and the RBMK, which is a graphite-moderated, multi-channel reactor (the so-called Chernobyl type). In addition, there are two fast-breeder reactors and four graphite-moderated boiling water reactors for combined heat and power in operation in Russia. These last two designs are not widely distributed and so are not discussed in detail in this report. As noted above, the safety of Soviet-designed reactors has been of great concern around the world since the catastrophic events at Chernobyl in 1986. This paper will briefly describe the technology involved. It will also examine the main safety concerns, both technical and organizational, associated with each reactor type. In addition, the paper will review the nuclear power programs in the new countries emerging from the former Soviet Union and its satellites and discuss the international efforts underway to address the most pressing problems. (author). 1 tab

Irradiation effects on Zr-2.5Nb in power reactors

Energy Technology Data Exchange (ETDEWEB)

Song, C., E-mail: Carol.Song@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

2016-06-15

Zirconium alloys are widely used as structural materials in nuclear applications because of their attractive properties such as a low absorption cross-section for thermal neutrons, excellent corrosion resistance in water, and good mechanical properties at reactor operating temperatures. Zr-2.5Nb is one of the most commonly used zirconium alloys and has been used for pressure tube materials in CANDU (Canada Deuterium Uranium) and RBMK (Reaktor Bolshoy Moshchnosti Kanalnyy, 'High Power Channel-type Reactor') reactors for over 40 years. In a recent report from the Electric Power Research Institute, Zr-2.5Nb was identified as one of the candidate materials for use in normal structural applications in light-water reactors owing to its increased resistance to irradiation-induced degradation as compared with currently used materials. Historically, the largest program of in-reactor tests on zirconium alloys was performed by Atomic Energy of Canada Limited. Over many years of in-reactor testing and CANDU operating experience with Zr- 2.5Nb, extensive research has been conducted on the irradiation effects on its microstructures, mechanical properties, deformation behaviours, fracture toughness, delayed hydride cracking, and corrosion. Most of the results on Zr-2.5Nb obtained from CANDU experience could be used to predict the material performance under light water reactors. This paper reviews the irradiation effects on Zr-2.5Nb in power reactors (including heavy-water and light-water reactors) and summarizes the current state of knowledge. (author)

Nuclear safety in EU candidate countries

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-10-01

Nuclear safety in the candidate countries to the European Union is a major issue that needs to be addressed in the framework of the enlargement process. Therefore WENRA members considered it was their duty to offer their technical assistance to their Governments and the European Union Institutions. They decided to express their collective opinion on nuclear safety in those candidate countries having at least one nuclear power plant: Bulgaria, the Czech Republic, Hungary, Lithuania, Romania, Slovakia and Slovenia. The report is structured as follows: A foreword including background information, structure of the report and the methodology used, General conclusions of WENRA members reflecting their collective opinion, For each candidate country, an executive summary, a chapter on the status of the regulatory regime and regulatory body, and a chapter on the nuclear power plant safety status. Two annexes are added to address the generic safety characteristics and safety issues for RBMK and VVER plants. The report does not cover radiation protection and decommissioning issues, while safety aspects of spent fuel and radioactive waste management are only covered as regards on-site provisions. In order to produce this report, WENRA used different means: For the chapters on the regulatory regimes and regulatory bodies, experts from WENRA did the work. For the chapters on nuclear power plant safety status, experts from WENRA and from French and German technical support organisations did the work. Taking into account the contents of these chapters, WENRA has formulated its general conclusions in this report.

A hypothetical severe reactor accident in Sosnovyj Bor, Russia

International Nuclear Information System (INIS)

Lahtinen, J.; Toivonen, H.; Poellaenen, R.; Nordlund, G.

1993-12-01

Individual doses and short-term radiological consequences from a hypothetical severe accident at the Russian nuclear power plant in Sosnovyj Bor were estimated for two sites in Finland. The sites are Kotka, located 140 km from the plant, and Helsinki, 220 km from the plant. The release was assumed to start immediately after the shutdown of the reactor (a 1000 MW RBMK unit) which had been operating at nominal power level for a long time. An effective release height of 500 m was assumed. The prevailing meteorological conditions during the release were taken to present the situation typical of the area (effective wind speed 9 m/s, neutral dispersion conditions). The release fractions applied in the study were of the same order as in the Chernobyl accident, i.e. 100% for noble gases, 60% for iodines, 40% for cesium and 1-10% for other radiologically important nuclides. The release was assumed to last 24 hours. However, half of the nuclides were released during the first hour. No attention was paid to the actual sequence of events that could lead to such release characteristics and time behaviour. The concentration and dose calculations were performed with a modified version of the computer code OIVA developed in Finnish Centre for Radiation and Nuclear Safety. Inhalation dose and external doses from the release plume and from the deposited activity were calculated for adults only, and no sheltering was considered. (11 refs., 4 figs., 6 tabs.)

FLICA-4 (version 1) a computer code for three dimensional thermal analysis of nuclear reactor cores

International Nuclear Information System (INIS)

Raymond, P.; Allaire, G.; Boudsocq, G.

1995-01-01

FLICA-4 is a thermal-hydraulic computer code developed at the French Energy Atomic Commission (CEA) for three dimensional steady state or transient two phase flow for design and safety thermal analysis of nuclear reactor cores. The two phase flow model of FLICA-4 is based on four balance equations for the fluid which includes: three balance equations for the mixture and a mass balance equation for the less concentrated phase which permits the calculation of non-equilibrium flows as sub cooled boiling and superheated steam. A drift velocity model takes into account the velocity disequilibrium between phases. The thermal behaviour of fuel elements can be computed by a one dimensional heat conduction equation in plane, cylindrical or spherical geometries and coupled to the fluid flow calculation. Convection and diffusion of solution products which are transported either by the liquid or by the gas, can be evaluated by solving specific mass conservation equations. A one dimensional two phase flow model can also be used to compute 1-D flow in pipes, guide tubes, BWR assemblies or RBMK channels. The FLICA-4 computer code uses fast running time steam-water functions. Phasic and saturation physical properties are computed by using bi-cubic spline functions. Polynomial coefficients are tabulated from 0.1 to 22 MPa and 0 to 800 degrees C. Specific modules can be utilised in order to generate the spline coefficients for any other fluid properties

Technical and economical conditions of nuclear energy usage continuation in Lithuania

International Nuclear Information System (INIS)

Gylys, J.; Ziedelis, S.; Klevas, V.

2005-01-01

The main producer of electric energy in Lithuania is Ignalina NPP with its two RBMK-1500 type reactors. It covers up to 86% of total annual electricity production. The compulsory premature closure of Ignalina NPP due to the decision of EU authorities will decrease maximum power generating capacity to 3273 MW in the year 2010 (slump of 42% in respect to 5698 MW of the year 2000) and it will cause a complex of serious technical, economical, ecological, and social consequences. The most important ones for energy sector are the negative power balance and the shortage of power generating capacity which can emerge straight after closure of the second unit of Ignalina NPP. An attempt has been taken to prove, that the most realistic way for replacement of lost power generating capacities is the construction of new nuclear or combined cycle gas turbine power plants. The results of the comparative analysis of their effectiveness and competitiveness are presented in the paper. Estimating the basic prevailing technical and economical factors and three possible scenarios of economy growth, the changes of power balance and levelised cost of produced electricity are compared. It is demonstrated that a new modern nuclear power plant would be competitive and it would be even a more favourable option in respect to a combined cycle gas turbine power plant due to the relatively lower energy production cost, especially when estimating the possible future growth of price for fossil fuel. (authors)

Nuclear power in eastern and central Europe. Background paper

Energy Technology Data Exchange (ETDEWEB)

Myers, L C [Library of Parliament, Ottawa, ON (Canada). Science and Technology Div.

1993-11-01

The breakup of the former Soviet Union and other political changes in eastern and central Europe have opened up the area to closer scrutiny than was previously possible. Because of the accident at Chernobyl, nuclear power is one of the subjects that western nations have had a great deal of interest in exploring. The former Soviet Union designed and/or helped build more than 60 civilian reactors in the region. Most of these reactors follow one of two distinctly different designs: the VVER, or pressurized water reactor series; and the RBMK, which is a graphite-moderated, multi-channel reactor (the so-called Chernobyl type). In addition, there are two fast-breeder reactors and four graphite-moderated boiling water reactors for combined heat and power in operation in Russia. These last two designs are not widely distributed and so are not discussed in detail in this report. As noted above, the safety of Soviet-designed reactors has been of great concern around the world since the catastrophic events at Chernobyl in 1986. This paper will briefly describe the technology involved. It will also examine the main safety concerns, both technical and organizational, associated with each reactor type. In addition, the paper will review the nuclear power programs in the new countries emerging from the former Soviet Union and its satellites and discuss the international efforts underway to address the most pressing problems. (author). 1 tab.

Modelling of zirconium alloys corrosion in LWRs

International Nuclear Information System (INIS)

Kritskij, V.G.; Berezina, I.G.; Kritskij, A.V.; Stjagkin, P.S.

1999-01-01

Chemical parameters, that exerted effect on Zr+1%Nb alloy corrosion and deserved consideration during reactor operation, were defined and a model was developed to describe the influence of physical and chemical parameters on zirconium alloys corrosion in nuclear power plants. The model is based on the correlation between the zirconium oxide solubility in high-temperature water under the influence of the chemical parameters and the measured values of fuel cladding corrosion under LWR conditions. The intensity of fuel cladding corrosion in the primary circuits depends on the coolant water quality, growth of iron oxide deposits and vaporization portion. Mathematically, the oxidation rate can be expressed as a sum of heat and radiation components. The temperature dependence on the oxidation rate can be described by the Arrenius equation. The radiation component of Zr uniform corrosion equation is a function of several factors such as neutron fluency, the temperature the metallurgical composition and et. We assume that the main factor is the changing of water chemistry and the H 2 O 2 concentration play the determinative role. Probably, the influence of H 2 O 2 is based on the formation of unstable compound ZrO 3 ·nH 2 O and Zr(OH) 4 with high solubility. The validity of the used formulae was confirmed by corrosion measurements on WWER and RBMK fuel cladding. The model can be applied for calculating the reliability of nuclear fuel operation. (author)

Upgrade of Control and Protection System of the Ignalina Nuclear Power Plant Units 1 and 2

International Nuclear Information System (INIS)

Wright, Ronald E.; Fletcher, Norman; Sidnev, Victor E.; Bickel, John H.; Vianello, Aldo; Pearsall, Raymond D.

2003-01-01

The Ignalina nuclear power plant (NPP) Units 1 and 2 are Soviet-designed, RBMK (Reaktor Bolshoi Moschnosti Kipyashchiy), channelized, large power-type reactors. The original-design electrical capacity for each unit was 1500 MW. Unit 1 began operating in 1983, and Unit 2 was started up in 1987. In 1994, the government of Lithuania agreed to accept grant support for the Ignalina NPP Safety Improvement Program with funding supplied by the Nuclear Safety Account of the European Bank for Reconstruction and Development (EBRD). As conditions for receiving this funding, the Ignalina NPP agreed to prepare a comprehensive safety analysis report that would undergo independent peer review after it was issued. The EBRD Safety Panel oversaw preparation and review of the report. In 1996, the safety analysis report for Unit 1 was completed and delivered to the EBRD. Part of the analyses covered anticipated transients without scram (ATWS). The analysis showed that some ATWS scenarios could lead to unacceptable consequences in <1 min. The EBRD Safety Panel recommended to the government of Lithuania that the Ignalina NPP develop and implement a program of compensatory measures for the control and protection system before the unit would be allowed to return to operation following its 1998 maintenance outage. A compensatory control and protection system that would mitigate the unacceptable consequences was designed, procured, manufactured, tested, and installed. The project was funded by U.S. Department of Energy

Rhodium self-powered detector for monitoring neutron fluence, energy production, and isotopic composition of fuel

International Nuclear Information System (INIS)

Sokolov, A.P.; Pochivalin, G.P.; Shipovskikh, Yu.M.; Garusov, Yu.V.; Chernikov, O.G.; Shevchenko, V.G.

1993-01-01

The use of self-powered detectors (SPDs) with a rhodium emitter customarily involves monitoring of neutron fields in the core of a nuclear reactor. Since current in an SPD is generated primarily because of the neutron flux, which is responsible for the dynamics of particular nuclear transformations, including fission reactions of heavy isotopes, the detector signal can be attributed unambiguously to energy release at the location of the detector. Computation modeling performed with the KOMDPS package of programs of the current formation in a rhodium SPD along with the neutron-physical processes that occur in the reactor core makes it possible to take account of the effect of the principal factors characterizing the operating conditions and the design features of the fuel channel and the detector, reveal quantitative relations between the generated signal and individual physical parameters, and determine the metrological parameters of the detector. The formation and transport of changed particles in the sensitive part of the SPC is calculated by the Monte Carlo method. The emitter activation, neutron transport, and dynamics of the isotopic composition in the fuel channel containing the SPD are determined by solving the kinetic equation in the multigroup representation of the neutron spectrum, using the discrete ordinate method. In this work the authors consider the operation of a rhodium SPD in a bundle of 49 fuel channels of the RBMK-1000 reactor with a fuel enrichment of 2.4% from the time it is inserted into a fresh channel

Analysis of the first stage in the reactor accident development at the Chernobyl NPP fourth unit

International Nuclear Information System (INIS)

Adamov, E.O.; Vasilevskij, V.P.; Ionov, A.I.

1988-01-01

Results of analyzing possible development of the first stage of the accident at the Chernobyl NPP fourth unit from the moment of pressing the Az-5 push button are presented. Calculations were conducted using the TRIADA three-dimensional dynamic program both for conditions without pump switching off and with their switching off. Distribution of neutron field over the core volume was determined according to actual readings of in-core detectors immediately before turbogenerator switching off. It is shown that sufficient reconstruction of neutron field begins immediately after pressing the Az-5 push button. Prohibitive decrease of operative reactivity margin which was admitted by personnel in the accident resulted in the growth of neutron power in reactor lower part within 1.5 s, predominating over power decrease in the upper part. Thus, the average integral power grows achieving the maximum during 7.5 s, after which its sharp decrease begins. Conditions with switching off of 4 circulating pumps lead to intesive growth of power and reactor runaway, initiated in the lower part of the core, which safety rods have not managed to reach. Fuel element temperature at that exceeds fuel melting point in the most power-intensive regions. This causes extremely intensive process of steam generation and overheating, pressure growth in the circuit, short-time decrease of the rate of operating pumps, destruction of fuel channels and the whole reactor. Primary measures assuring RBMK ractor safety were formulated on the basis of conducted investigation

Electrolysis byproduct D2O provides a third way to mitigate CO2

International Nuclear Information System (INIS)

Schenewerk, William Ernest

2009-01-01

Rapid atomic power deployment may be possible without using fast breeder reactors or making undue demands on uranium resource. Using by-product D2O and thorium-U233 in CANDU and RBMK piles may circumvent need for either fast breeder reactors or seawater uranium. Atmospheric CO2 is presently increasing 2.25%/year in proportion to 2.25%/year exponential fossil fuel consumption increase. Roughly 1/3 anthropologic CO2 is removed by various CO2 sinks. CO2 removal is modelled as being proportional to 45-year-earlier CO2 amount above 280 ppm-C Water electrolysis produces roughly 0.1 kg-D20/kWe-y. Material balance assumes each electrolysis stage increases D2O bottoms concentration times 3. Except for first two electrolysis stages, all water from hydrogen consumption is returned to electrolysis. The unique characteristic of this process is the ability to economically burn all deuterium-enriched H2 in vehicles. Condensate from vehicles returns to appropriate electrolysis stage. Fuel cell condensate originally from reformed natural gas may augment second-sage feed. Atomic power expansion is 5%/year, giving 55000 GWe by 2100. World primary energy increases 2.25%/y, exceeding 4000 EJ/y by 2100. CO2 maximum is roughly 600 ppm-C around year 2085. CO2 declines back below 300 ppm-C by 2145 if the 45-year-delay seawater sink remains effective

Nuclear safety in EU candidate countries

International Nuclear Information System (INIS)

2000-10-01

Nuclear safety in the candidate countries to the European Union is a major issue that needs to be addressed in the framework of the enlargement process. Therefore WENRA members considered it was their duty to offer their technical assistance to their Governments and the European Union Institutions. They decided to express their collective opinion on nuclear safety in those candidate countries having at least one nuclear power plant: Bulgaria, the Czech Republic, Hungary, Lithuania, Romania, Slovakia and Slovenia. The report is structured as follows: A foreword including background information, structure of the report and the methodology used, General conclusions of WENRA members reflecting their collective opinion, For each candidate country, an executive summary, a chapter on the status of the regulatory regime and regulatory body, and a chapter on the nuclear power plant safety status. Two annexes are added to address the generic safety characteristics and safety issues for RBMK and VVER plants. The report does not cover radiation protection and decommissioning issues, while safety aspects of spent fuel and radioactive waste management are only covered as regards on-site provisions. In order to produce this report, WENRA used different means: For the chapters on the regulatory regimes and regulatory bodies, experts from WENRA did the work. For the chapters on nuclear power plant safety status, experts from WENRA and from French and German technical support organisations did the work. Taking into account the contents of these chapters, WENRA has formulated its general conclusions in this report

Damages of electrical insulation of cable products used at NPP`s and technique of their detection and operative control; Povrezhdeniya v ehlektricheskoj izolyatsii kabel`nykh izdulij, ehkspluatirue mykh na atomnykh ehlektrostantsiyakh i metody ikh obnaruzheniya i operativnogo kontro lya

Energy Technology Data Exchange (ETDEWEB)

Valeev, R S; Filatov, N I

1994-12-31

Analysis of possible damages in electrical insulation of cable products under their application at NPP`s is conducted. Basic methods for detecting such damages and rapid control of technical condition of cable products during the operation are considered.

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 show that the ATWS could contribute essentially to the probability of core damage or to off-site radioactive releases (Russia) or in cases where there is only one fast acting shutdown system for which the Regulatory Authority stipulates that failure must be assumed. The development of a national policy on the ATWS issue in line with international practices is recommended for national regulatory authorities. This policy should consider the systems necessary to mitigate consequences and/or requirements for acceptable plant performance in an ATWS event. This publication provides guidance on the performance of ATWS analyses of transients for licensing purposes, on the initiating events identified for those WWER reactors, and on the best estimate approach to the transient analysis of ATWS events. Further guidance is directed at reliability assessment of instrumentation and control (I and C) related to ATWS, including its common mode failure potential, and the qualification of systems and components necessary for operation under accident conditions to mitigate ATWS. While focusing on WWER-1000 reactors, this publication also provides guidance for ATWS events in WWER-440 reactors, taking into consideration the differences of this reactor type and its instrumentation and control

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

International Nuclear Information System (INIS)

1996-08-01

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

Chernobyl: 30 years after - Proceedings of the technical meeting of the French Society of Radiation Protection

International Nuclear Information System (INIS)

Champion, Didier; Chouha, Michel; Damette, Guy; Durand, Vanessa; Besnus, Francois; Renaud, Philippe; Adam-Guillermin, Christelle; Laurier, Dominique; Chabrier, Patrick; Chauveau, Thomas; Thiry, Yves; Menetrier, Florence; Chevillard, Sylvie; Lesueur, Fabienne; Schneider, Thierry

2016-03-01

The French Society of Radiation Protection (SFRP) organized a technical meeting on the present day situation of the Chernobyl site, 30 years after the accident of the nuclear power plant. The review deals with the situation of the facility and of its safety works, the environment, the management of wastes, the workers and populations exposure, and the health monitoring of the exposed populations. This document brings together the abstracts and the presentations (slides) of the different talks given at the meeting: 1 - The main highlights 30 years after the Chernobyl accident (Didier CHAMPION, SFRP); 2 - Circumstances, progress and consequences of the Chernobyl accident - Lessons and experience feedback for the other RBMK reactors (Michel CHOUHA, IRSN); 3 - Chernobyl, a confinement arch for 100 years (Patrick CHABRIER, Thomas CHAUVEAU - BOUYGUES); 4 - The reactor wastes management and the dismantling operations (Guy DAMETTE - IRSN); 5 - Environment contamination in the vicinity of the site (Yves THIRY - ANDRA); 6 - Impact of the accident on agriculture (Vanessa DURAND - IRSN); 7 - The fate of remediation wastes (Francois BESNUS - IRSN); 8 - Chernobyl fallouts in France (Philippe RENAUD - IRSN); 9 - The ecological consequences of the Chernobyl accident (Christelle ADAM-GUILLERMIN - IRSN); 10 - Results of liquidators and populations exposure (Florence MENETRIER - CEA); 11 - Thyroid cancers monitoring in the Chernobyl area and the role of modifying genetic factors (Fabienne LESUEUR - Institut Curie); 12 - Results of the Chernobyl accident health impact studies (Dominique LAURIER - IRSN); 13 - Impact on populations living condition (Thierry SCHNEIDER - CEPN); 14 - Molecular signature of radiation induced thyroid tumors (Sylvie CHEVILLARD - CEA)

Static analytical and experimental research of shock absorber to safeguard the nuclear fuel assemblies

Energy Technology Data Exchange (ETDEWEB)

Dundulis, Gintautas [Laboratory of Nuclear Installation Safety, Lithuanian Energy Institute, Breslaujos Street 3, LT-44403 Kaunas (Lithuania)], E-mail: gintas@mail.lei.lt; Grybenas, Albertas [Laboratory of Materials Research and Testing, Lithuanian Energy Institute, Breslaujos Street 3, LT-44403 Kaunas (Lithuania); Karalevicius, Renatas [Laboratory of Nuclear Installation Safety, Lithuanian Energy Institute, Breslaujos Street 3, LT-44403 Kaunas (Lithuania); Makarevicius, Vidas [Laboratory of Materials Research and Testing, Lithuanian Energy Institute, Breslaujos Street 3, LT-44403 Kaunas (Lithuania); Rimkevicius, Sigitas; Uspuras, Eugenijus [Laboratory of Nuclear Installation Safety, Lithuanian Energy Institute, Breslaujos Street 3, LT-44403 Kaunas (Lithuania)

2009-01-15

The Ignalina Nuclear Power Plant (NPP) has two RBMK-1500 graphite-moderated boiling water multi-channel reactors. The Ignalina NPP Unit 1 was shutdown at the end of 2004, while Unit 2 is foreseen to be shutdown at the end of 2009. At the Ignalina NPP Unit 1 remains approximately 1000 spent fuel assemblies with low burn-up depth. A special set of equipment was developed to reuse these assemblies in the reactor of Unit 2. One of most important items of this set is a container, which is used for the transportation of spent fuel assemblies between the reactors of Unit 1 and Unit 2. A special shock absorber was designed to avoid failure of fuel assemblies in case of hypothetical spent fuel assemblies drop accident during uploading/unloading of spent fuel assemblies to/from container. This shock absorber was examined by using scaled experiments. The objective of this article is the estimation whether the proposed design of shock absorber fulfils the function of the absorber and the optimization of its geometrical parameters using the results of the performed investigations. Static analytical and experimental investigations are presented in the article. The finite element code BRIGADE/Plus was used for the analytical analysis. The calculation model was verified by comparing the experimental investigation and simulation results for further employment of this finite element model in the development of an optimum design of shock absorber. Static simulation was used to perform primary optimization of design and dimension of the shock absorber.

Progress in Low and Intermediate Level Operational Waste Characterization and Preparation for Disposal at Ignalina NPP

International Nuclear Information System (INIS)

Poskas, P.; Adomaitis, J. E.; Ragaisis, V.

2003-01-01

In Lithuania about 70-80% of all electricity is generated at a single power station, Ignalina NPP, which has two RBMK-1500 type reactors. Units 1 and 2 will be closed by 2005 and 2010, respectively, taking into account the conditions of the long-term substantial financial assistance rendered by the European Union, G-7 countries and other states as well as international institutions. The Government approved the Strategy on Radioactive Waste Management. Objectives of this strategy are to develop the radioactive waste management infrastructure based on modern technologies and provide for the set of practical actions that shall bring management of radioactive waste in Lithuania in compliance with radioactive waste management principles of IAEA and with good practices in force in European Union Member States. SKB-SWECO International-Westinghouse Atom Joint Venture with participation of Lithuanian Energy Institute has prepared a reference design of a near surface repository for short-lived low and intermediate level waste. This reference design is applicable to the needs in Lithuania, considering its hydro-geological, climatic and other environmental conditions and is able to cover the expected needs in Lithuania for at least thirty years ahead. Development of waste acceptance criteria is in practice an iterative process concerning characterization of existing waste, repository development, safety and environmental impact assessment etc. This paper describes the position in Lithuania with regard to the long-term management of low and intermediate level waste in the absence of finalized waste acceptance criteria and a near surface repository

CASTOR(r) and CONSTOR(r) type transport and storage casks for spent fuel and high active waste

International Nuclear Information System (INIS)

Kuehne, B.; Sowa, W.

2002-01-01

The German company GNB has developed, tested, licensed, fabricated, loaded, transported and stored a large number of casks for spent fuel and high-level waste. CASTOR(r) casks are used at 18 sites on three continents. Spent fuel assemblies of the types PWR, BWR, VVER, RBMK, MTR and THTR as well as vitrified high active waste (HAW) containers are stored in these kinds of casks. More than 600 CASTOR(r) casks have been loaded for long-term storage. The two decades of storage have shown that the basic requirements, which are safe confinement, criticality safety, sufficient shielding and appropriate heat transfer have been fulfilled in each case. There is no indication that problems will arise in the future. Of course, the experience of 20 years has resulted in improvements of the cask design. One basic improvement is GNB's development since the mid 1990s of a sandwich cask design using heavy concrete and steel as basic materials, for economical and technical reasons. This CONSTOR(r) cask concept also fulfils all design criteria for transport and storage given by the IAEA recommendations and national authorities. By May 2002 40 CONSTOR(r) casks had been delivered and 15 had been successfully loaded and stored. In this paper the different types of casks are presented. Experiences gained during the large number of cask loadings and more than 4000 cask-years of storage will be summarised. The presentation of recent and future development shows the optimisation potential of the CASTOR(r) and CONSTOR(r) cask families for safe and economical management of spent fuel. (author)

Implementation of U.S. Department of Energy physical protection upgrades in Lithuania and Uzbekistan

International Nuclear Information System (INIS)

Haase, M.; Romesberg, L.; Showalter, R.; Soo Hoo, M.S.; Corey, J.; Engling, E.

1996-01-01

Since 1994, the U.S. Department of Energy (DOE) has provided cooperative assistance to the non-nuclear weapons states of the Former Soviet Union. This effort, within DOE's program of Material Protection, Control, and Accounting (MPC ampersand A), identified the Institute of Nuclear Physics (INP) in Uzbekistan and the Ignalina Nuclear Power Plant (INPP) in Lithuania as sites for cooperative MPC ampersand A projects. The INP, located just outside of Tashkent, is the site of a 10-megawatt WWR-SM research reactor. This reactor is expected to remain operational as a major nuclear research and isotope production reactor for Central Asia. The INPP, located 100 kilometers northeast of the capital city of Vilnius, consists of two Russian-made RBMK reactors with a combined power output of 3,000 megawatts (electric). This power plant has been the subject of international safety and security concerns, which prompted DOE's cooperative assistance effort. This paper describes U.S. progress in a multi-national effort directed at implementing physical protection upgrades in Lithuania and Uzbekistan. The upgrades agreed upon between DOE and the INP and between DOE and the INPP have been designed to interface with upgrades being implemented by other donor countries. DOE/INPP upgrade projects include providing training on U.S. approaches to physical protection, access control through the main vehicle portal, a hardened central alarm station, and improved guard force communications. DOE/INP upgrade projects in Uzbekistan include an access control system, a hardened fresh fuel storage vault, an interior intrusion detection and assessment system, and an integrated alarm display and assessment system

The full stories on Armenia and Beloyarsk

International Nuclear Information System (INIS)

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

1995-01-01

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

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

Lessons learnt from Ignalina NPP decommissioning project

International Nuclear Information System (INIS)

NAISSE, Jean-Claude

2007-01-01

The Ignalina Nuclear Power Plant (INPP) is located in Lithuania, 130 km north of Vilnius, and consists of two 1500 MWe RBMK type units, commissioned respectively in December 1983 and August 1987. On the 1. of May 2004, the Republic of Lithuania became a member of the European Union. With the protocol on the Ignalina Nuclear Power in Lithuania which is annexed to the Accession Treaty, the Contracting Parties have agreed: - On Lithuanian side, to commit closure of unit 1 of INPP before 2005 and of Unit 2 by 31 December 2009; - On European Union side, to provide adequate additional Community assistance to the efforts of Lithuania to decommission INPP. The paper is divided in two parts. The first part describes how, starting from this agreement, the project was launched and organized, what is its present status and which activities are planned to reach the final ambitious objective of a green field. To give a global picture, the content of the different projects that were defined and the licensing process will also be presented. In the second part, the paper will focus on the lessons learnt. It will explain the difficulties encountered to define the decommissioning strategy, considering both immediate or differed dismantling options and why the first option was finally selected. The paper will mention other challenges and problems that the different actors of the project faced and how they were managed and solved. The paper will be written by representatives of the Ignalina NPP and of the Project Management Unit. (author)

Proceedings of TopSafe 2008 Transactions

International Nuclear Information System (INIS)

2008-01-01

The aim of the conference is to provide a forum for addressing the current status and future perspectives with regards to safety at nuclear installations worldwide. Previous TopSafe editions took place in Budapest (1995) and Valencia (1998). The conference is directed at a broad range of experts in the area of nuclear safety, including professionals from the different disciplines involved in the safety of nuclear power plants, installations in other parts of the fuel cycle, and research reactors. It is aimed at professionals coming from the research organisations, universities, vendors, operators, regulatory bodies as well as policy makers. Top level representatives of the Countries that are constructing new nuclear power plants are invited. Regulators of all individual Countries with nuclear programme are expected to contribute the Conference. The topics of the conference are: Safety Issues of Operating Power Plants PWR and BWR, CANDU, WWER, RBMK; Application of European Utilities Requirements; Probabilistic and Deterministic Analysis; Shutdown Safety; Advances in Safety: Analysis Codes and Techniques; Severe Accidents Management; International Safety Studies; Emergency Planning; Risk Informed Application and Licensing; Regulatory Safety Requirements; Ageing and Life Extension; Power Upgrades and Relevant Topics; Management of Safety and Quality; Safety Culture and Self Assessment; Political and Public Perception of Nuclear Energy; Nuclear Power Plant Security; Safety Issues of Future Power Plants-Near term deployment reactors (EPR, SWR1000, AP1000, ESBWR, SBWR, ACR-1000) and Generation IV reactors; Safety Issues of Research Reactors (pool type and others); Fuel Cycle Facilities Safety-Uranium mining and conversion, enrichment and fuel production, reprocessing and transmutation, waste disposal. (authors)

Radiation safety practice at nuclear power stations and estimation of dose burdens to the USSR general public in the context of the country's nuclear power development plans

International Nuclear Information System (INIS)

Vorob'ev, E.I.; Il'in, L.A.; Turovskij, V.D.; Buldakov, L.A.; Lusev, N.G.; Pavlovskij, O.A.; Parkhomenko, G.M.

1983-01-01

The paper sets forth the main features of the State system of health protection for staff and the general public, and likewise the essentials of environmental protection. The principles of standardizing radiation factors are given for power station personnel and for the general public, together with the main provisions of the health Standards and Rules for radiation protection at present valid in the USSR. Data are quoted on the radiation situation at nuclear power stations and on the size of releases of radioactive aerosols and liquid effluents to the environment. The paper pays particular attention to analyses of the radiation situation in districts where nuclear power stations are situated and also to the type and scope of monitoring of radioactive environmental contamination. An analysis of the coefficients achieved with Soviet pressurized water (WWER), high-power channel-type (RBMK) and fast (BN) reactors currently in large-scale use shows that in terms both of release levels of radioactive substances and of the dose burdens to staff and general public these reactors are comparable with the best foreign nuclear power installations. Values actually measured and values calculated for the basic parameters of the radiation situation in areas of the USSR where nuclear power stations are situated confirm the safety of these facilities as regards the health of the general public and the extremely low levels of their effects on the environment. In conclusion, the paper quotes estimates of the collective effective dose equivalent to the USSR population expected to result from implementation of the country's nuclear power programme up to the year 2000. Radiation safety problems associated with nuclear power production which still require solution are enumerated. (author)

Information on the Chernobyl NPP accident and its consequencies prepared for IAEA

Energy Technology Data Exchange (ETDEWEB)

1986-11-01

The information on the accident at the 4th power unit of the Chernobyl NPP and its consequences prepared for IAEA on the basis of the conclusions made by the Government commission constituted for investigating the accident causes and implementing the necessary emergency and reconstruction measures is given. The accident with reactor core disruption and partial destruction of the building Lappened on 26.04.86 at 1 hour and 23 minutes. The accident occurred before reactor shut-down for planned repairs during the testing of one of turbogenerators. The design features of the RBMK-1000 reactor plant, its main physical characteristics and parameters of the NPP safety system are considered. The chronology of the accident development and the results of analysis carried out using a mathematical model are given. The causes of the accident are analyzed. The measures for preventing the accident development and lessening its consequences as well as those for the environment radioactive contamination control and sanitary provisions are described in detail. The conclusion is made that the original cause of the accident is highly improbable combination of disorder and errors in operational conditions made by the personnel of the power unit. It is emphasized that development of the world nuclear engineering, besides advantages in the field of power supply and natural resources conservation, incurs also damages of international character. Among these are transboundary radioactivity transport, in particular, during serious radiation accidents and the danger of international terrorism and specific radiation hazard of nuclear objects under war conditions. All this defines the key necessity of deep international cooperation in the field of nuclear power engineering and its safeguarding.

Nuclear power in Kazakhstan and current status of the BN-350 fast reactor

International Nuclear Information System (INIS)

Krechetov, S.

1998-01-01

Atomic scientific-industrial complex of Republic of Kazakhstan consist of: Number of natural uranium mines, two plants of U 3 O 8 production at Aktau and Stepnogorsk towns, metallurgical plant producing fuel pellets for RBMK and WWER fuel assemblies. Fast breeder reactor with sodium coolant BN - 350 at Aktau. The average share of BN-350 in total electricity production is 0.7%. Taking into account common condition industrial in Kazakhstan have no significant improvement the total electricity production on goal and oil station stayed on the same level as in 1996. According to government decision in 1998 the following structure of atomic complex have been established. Several rather serious events should be mentioned. In January 1998 the Provision of licensing in nuclear field was signed by Prime Ministry and now Kazakhstan have all necessary acts for starting this process. In April 1998 the General Program of development atomic scientific and industrial complex of Kazakhstan had been reported to Government and got approval in whole. In particular this program are including the design and construction NPP for electricity production on the lake Balhash, and two NPP for heating Almaty and new capital Akmola. In April 1998 the law on Radiation protection had got approval of Parliament and now President should sign it. In January the Nuclear Technologies Safety Center (NTSC) had been established by group of organizations such as KAEA, NNC, University, Nuclear Society of Kazakhstan, Center of standardization and Almaty local administration. NTSC have established as a society independent experts in the field nuclear safety. With cooperation with ANL an expertise on nuclear safety of BN-350 will be done related to long-term spent fuel storage

Evaluation of release amount of radioactivity from Chernobyl accident and of resulting radiological consequence in China

International Nuclear Information System (INIS)

Zhang Yongxing

1987-01-01

Three kinds of methods are used to evaluate the release amount from Chernoby1 RBMK-1000 reactor accident, i.e. (1) estimation by comparison with Windscale accident; (2) estimation in terms of the stock in the core; and (3) estimation according to the available monitoring data form adjacent countries such as Poland and Finland. The results obtained are as follows: the release of I-131 was 0.1-1.5 EBq, which is approximately 4-50% of the stock in the core; the release amount of Ru-103 was comparable to that of Cs-137, both approximately 5-10% of that of I-131; the volatile nuclides such as Mo-99, Ru-103, Te-132, Cs-137 etc., were in the order of 0.4 EBq; involatile nuclides were 0.2 EBq; noble gases and other fission products 10 EBq; and the total amount released was about 20 EBq, which taken together 8% of the stock in the core. The radioactive cloud cluster passed through over the area of China in the beginning of May. It was estimated that the total amount of I-131 in air over China area was about 1.6 PB q , Cs-137 about 0.3 PB q , Ru-103 about 0.2 PB q ; the total fallout in the area of China was about 3 PB q for I-131, about 0.1 PB q for Cs-137, about 0.3 PB q for Ru-103. The resulting effective dose equavalent commitment to critical group individual was about 60 μSv, collective effective dose equavalent commitment received by the population of China was about 1 x 10 4 man.Sv

HELIOS2: Benchmarking against experiments for hexagonal and square lattices

International Nuclear Information System (INIS)

Simeonov, T.

2009-01-01

HELIOS2, is a 2D transport theory program for fuel burnup and gamma-flux calculation. It solves the neutron and gamma transport equations in a general, two-dimensional geometry bounded by a polygon of straight lines. The applied transport solver may be chosen between: The Method of Collision Probabilities (CP) and The Method of Characteristics(MoC). The former is well known for its successful application for preparation of cross section data banks for 3D simulators for all types lattices for WWERs, PWRs, BWRs, AGRs, RBMK and CANDU reactors. The later, MoC, helps in the areas where the requirements of CP for computational power become too large of practical application. The application of HELIOS2 and The Method of Characteristics for some large from calculation point of view benchmarks is presented in this paper. The analysis combines comparisons to measured data from the Hungarian ZR-6 reactor and JAERI facility of Tank type Critical Assembly (TCA) to verify and validate HELIOS2 and MOC for WWER assembly imitators; configurations with different absorber types- ZrB 2 , B 4 C, Eu 2 O 3 and Gd 2 O 3 ; and critical configurations with stainless steel in the reflector. Core eigenvalues and reaction rates are compared. With the account for the uncertainties the results are generally excellent. Special place in this paper is given to the effect of Iron-made radial reflector. Comparisons to measurements from TIC and TCA for stainless steel and Iron reflected cores are presented. The calculated by HELIOS-2 reactivity effect is in very good agreement with the measurements. (author)

HELIOS2: Benchmarking Against Experiments for Hexagonal and Square Lattices

International Nuclear Information System (INIS)

Simeonov, T.

2009-01-01

HELIOS2, is a 2D transport theory program for fuel burnup and gamma-flux calculation. It solves the neutron and gamma transport equations in a general, two-dimensional geometry bounded by a polygon of straight lines. The applied transport solver may be chosen between: The Method of Collision Probabilities and The Method of Characteristics. The former is well known for its successful application for preparation of cross section data banks for 3D simulators for all types lattices for WWER's, PWR's, BWR's, AGR's, RBMK and CANDU reactors. The later, method of characteristics, helps in the areas where the requirements of collision probability for computational power become too large of practical application. The application of HELIOS2 and The method of characteristics for some large from calculation point of view benchmarks is presented in this paper. The analysis combines comparisons to measured data from the Hungarian ZR-6 reactor and JAERI's facility of tanktype critical assembly to verify and validate HELIOS2 and method of characteristics for WWER assembly imitators; configurations with different absorber types-ZrB2, B4C, Eu2O3 and Gd2O3; and critical configurations with stainless steel in the reflector. Core eigenvalues and reaction rates are compared. With the account for the uncertainties the results are generally excellent. Special place in this paper is given to the effect of Iron-made radial reflector. Comparisons to measurements from The Temporary International Collective and tanktype critical assembly for stainless steel and Iron reflected cores are presented. The calculated by HELIOS-2 reactivity effect is in very good agreement with the measurements. (Authors)

The nuclear accidents: Causes and consequences

International Nuclear Information System (INIS)

Rochd, M.

1988-01-01

The author discussed and compared the real causes of T.M.I. and Chernobyl accidents and cited their consequences. To better understand how these accidents occurred, a brief description of PWR type (reactor type of T.M.I.) and of RBMK type (reactor type of Chernobyl) has been presented. The author has also set out briefly the safety analysis objectives and the three barriers established to protect the public against the radiological consequences. To distinguish failures that cause severe accidents and to analyze them in details, it is necessary to classify the accidents. There are many ways to do it according to their initiator event, or to their frequency, or to their degree of gravity. The safety criteria adopted by nuclear industry have been explained. These criteria specify the limits of certain physical parameters that should not be exceeded in case of incidents or accidents. To compare the real causes of T.M.I. and Chernobyl accidents, the events that led to both have been presented. As observed the main common contributing factors in both cases are that the operators did not pay attention to warnings and signals that were available to them and that they were not trained to handle these accident sequences. The essential conclusions derived from these severe accidents are: -The improvement of operators competence contribute to reduce the accident risks; -The rapid and correct diagnosis of real conditions at each point of the accidents permits an appropriate behavior that would bring the plant to a stable state; -Competent technical teams have to intervene and to assist the operators in case of emergency; -Emergency plans and an international collaboration are necessary to limit the accident risks. 11 figs. (author)

Nuclear safety research collaborations between the U.S. and Russian Federation International Nuclear Safety Centers

International Nuclear Information System (INIS)

Hill, D. J.; Braun, J. C.; Klickman, A. E.; Bougaenko, S. E.; Kabonov, L. P.; Kraev, A. G.

2000-01-01

The Russian Federation Ministry for Atomic Energy (MINATOM) and the US Department of Energy (USDOE) have formed International Nuclear Safety Centers to collaborate on nuclear safety research. USDOE established the US Center (ISINSC) at Argonne National Laboratory (ANL) in October 1995. MINATOM established the Russian Center (RINSC) at the Research and Development Institute of Power Engineering (RDIPE) in Moscow in July 1996. In April 1998 the Russian center became a semi-independent, autonomous organization under MINATOM. The goals of the center are to: Cooperate in the development of technologies associated with nuclear safety in nuclear power engineering; Be international centers for the collection of information important for safety and technical improvements in nuclear power engineering; and Maintain a base for fundamental knowledge needed to design nuclear reactors. The strategic approach is being used to accomplish these goals is for the two centers to work together to use the resources and the talents of the scientists associated with the US Center and the Russian Center to do collaborative research to improve the safety of Russian-designed nuclear reactors. The two centers started conducting joint research and development projects in January 1997. Since that time the following ten joint projects have been initiated: INSC databases--web server and computing center; Coupled codes--Neutronic and thermal-hydraulic; Severe accident management for Soviet-designed reactors; Transient management and advanced control; Survey of relevant nuclear safety research facilities in the Russian Federation; Computer code validation for transient analysis of VVER and RBMK reactors; Advanced structural analysis; Development of a nuclear safety research and development plan for MINATOM; Properties and applications of heavy liquid metal coolants; and Material properties measurement and assessment. Currently, there is activity in eight of these projects. Details on each of these

Dynamic analytical and experimental research of shock absorber to safeguard the nuclear fuel assemblies

Energy Technology Data Exchange (ETDEWEB)

Dundulis, Gintautas, E-mail: gintas@mail.lei.lt [Laboratory of Nuclear Installation Safety, Lithuanian Energy Institute, Breslaujos str. 3, LT-44403 Kaunas (Lithuania); Grybenas, Albertas [Laboratory of Materials Research and Testing, Lithuanian Energy Institute, Breslaujos str. 3, LT-44403 Kaunas (Lithuania); Karalevicius, Renatas [Laboratory of Nuclear Installation Safety, Lithuanian Energy Institute, Breslaujos str. 3, LT-44403 Kaunas (Lithuania); Makarevicius, Vidas [Laboratory of Materials Research and Testing, Lithuanian Energy Institute, Breslaujos str. 3, LT-44403 Kaunas (Lithuania); Rimkevicius, Sigitas; Uspuras, Eugenijus [Laboratory of Nuclear Installation Safety, Lithuanian Energy Institute, Breslaujos str. 3, LT-44403 Kaunas (Lithuania)

2013-07-15

Highlights: • Plastical deformation of the shock absorber. • Dynamic testing of the scaled shock absorber. • Dynamic simulation of the shock absorber using finite element method. • Strain-rate evaluation in dynamic analysis. • Variation of displacement, acceleration and velocity during dynamic impact. -- Abstract: The Ignalina Nuclear Power Plant (NPP) has two RBMK-1500 graphite moderated boiling water multi-channel reactors. The Ignalina NPP Unit 1 was shut down at the end of 2004 while Unit 2 has been in operation for over 5 years. After shutdown at the Unit 1 remained spent fuel assemblies with low burn-up depth. In order to reuse these assemblies in the reactor of Unit 2 a special set of equipment was developed. One of the most important items of this set is a container, which is used for the transportation of spent fuel assemblies between the reactors of Unit 1 and Unit 2. A special shock absorber was designed to avoid failure of fuel assemblies in case of hypothetical spent fuel assemblies drop accident during uploading/unloading of spent fuel assemblies to/from container. This shock absorber was examined using scaled experiments and finite element analysis. Static and dynamic investigations of the shock absorber were performed for the estimation and optimization of its geometrical parameters. The objective of this work is the estimation whether the proposed design of shock absorber can fulfil the stopping function of the spent fuel assemblies and is capable to withstand the dynamics load. Experimental testing of scaled shock absorber models and dynamic analytical investigations using the finite element code ABAQUS/Explicit were performed. The simulation model was verified by comparing the experimental and simulation results and it was concluded that the shock absorber is capable to withstand the dynamic load, i.e. successful force suppression function in case of accident.

Study and optimization of operating regimes of NPP district heating system

International Nuclear Information System (INIS)

Bunin, V.S.; Vasil'ev, M.K.; Kudryavtsev, A.A.; Gorbashev, Yu.B.; Gadzhij, V.M.

1980-01-01

Thermal tests of the system with two reactors and four turbines have been carried out for the purpose of verification of operating regimes of the NPP district heating system with boiling single-curcuit RBMK-1000 reactors and K-500-65/3000 turbines. The system is designed for heat supply of habitable settlement and industrial site. The data processing have been carried out by the BESM-6 computer representing distributions of heat flow, steam, water and their parameters and determining the main energy indices of the system. Calculations of the system operating regime variables during the year have been carried out with the help of this program. It has been expected that the system provided heat consumption of 232 MW at calculated regime of thermal loading of the district, temperature regime of the system water of 130/170 deg C, relative load of hot water supply of 0.2 and duration of heating period of 4800 h. Calculations demonstrated that distric heat supply by NPP allowed one to supplant about 85 thous. of reference fuel/year of organic fuel. About 63 thous. of reference fuel/year are required for compensation of decrease of electric energy production in a condensation cycle. It has been also shown, that replacing the four-stroke system heaters by one-stroke heaters permits to drop system water underheating 1.5 times and, respectively, electric energy underproduction to 72 mln Mj (20 mln, kWxh). It produces additional economy of 6.6 thous. reference fuel/year. Calculations of its heat system have been conducted in order to determine the influence of water consumption in an intermediate circuit on the system efficiency. It has been shown that with the increase of water consumption energy power losses decrease. Thus, the above studied have demonstrated that the use of the single-circuit NPP district heating systems leads to considerable economy of fuel

Constor steel concrete sandwich cask concept for transport and storage of spent nuclear fuel

International Nuclear Information System (INIS)

Diersch, R.; Dreier, G.; Gluschke, K.; Zubkov, A.; Danilin, B.; Fromzel, V.

1998-01-01

A spent nuclear fuel transport and storage sandwich cask concept has been developed together with the Russian company CKTI. Special consideration was given to an economical and effective way of manufacturing by using conventional mechanical engineering technologies and common materials. The main objective of this development was to fabricate these casks in countries not having highly specialized industries. Nevertheless, this sandwich cask concept fulfills both the internationally valid IAEA criteria for transportation and the German criteria for long-term intermediate storage. The basic cask concept has been designed for adaptation to different spent fuel specifications as well as handling conditions in the NPP. Recently, adaptations have been made for spent fuel from the RBMK and VVER reactors, and also for BWR spent fuel. The analyses of nuclear and thermal behaviour as well as of strength according to IAEA examination requirements (9-m-drop, 1-m-pin drop, 800 deg. C-fire test) and of the behaviour during accident scenarios at the storage site (drop, fire, gas cloud explosion, side impact) were carried out by means of recognized calculation methods and programmes. In a special experimental programme, the mechanical and thermodynamic properties of heavy concrete were examined and the reference values required for safety analyses were determined. The results of the safety analysis after drop tests according to IAEA-regulations as well as after 1 m-drops at the storage site were confirmed by means of a test programme using a scale model. The fabrication technology has been tested with help of a half scale cask model. The model has been prefabricated in Russia and completed in Germany. It has been shown that the CONSTOR cask can be fabricated in an effective and economic way. (authors)

WANO. Development, programs, challenges

International Nuclear Information System (INIS)

Haferburg, Manfred

2011-01-01

In the wake of the accident at the Soviet RBMK reactor unit 4 in Chernobyl the nuclear industry founded the World Association of Nuclear Operators (WANO). To this day, the purpose of the organization has been to enhance worldwide cooperation of nuclear industry and, in this way, strengthen the safety and availability of nuclear power plants. Following some first steps after 1986, the charter of the organization was signed at the WANO constituent assembly in Moscow on May 15 and 16, 1989. The member companies thus committed themselves to support WANO's mission. WANO was established for these purposes: ''The mission of WANO is to maximize the safety and reliability of nuclear power plants worldwide by working together to assess, benchmark and improve performance through mutual support, exchange of information, and emulation of best practices.'' The WANO programs developed speedily thereafter. The focus was on peer reviews. In 2000, the first interim objective had been reached: Fifty percent of all member nuclear power plants had undergone peer reviews. In addition, plant-related peer reviews were extended throughout all operator organizations, and corporate peer reviews were developed. The other WANO programs as well, i.e. exchanges of experience, technical support, and performance indicators, exerted more and more influence on industry. Peer reviews covered entire operator organizations, and corporate peer reviews were developed. The worldwide paradigm shift in evaluating the use of nuclear power, and the associated construction programs for new nuclear power plants already in their implementation phase, assigned a new quality to the work of WANO. The organization is preparing a long-term strategy in the face of the challenges to be expected. The ultimate objective of these efforts is to support member organizations from the first preparations of a nuclear power plant project to the end of commercial operation. (orig.)

From the experience of spokesman's interactions with journalists

International Nuclear Information System (INIS)

Latek, Stanislaw

2000-01-01

In Poland, the relations between the nuclear community (and its spokesmen) and the media are rather special, as the nuclear energy industry does not exist. Since the withdrawal from nuclear power plant construction in Zarnowiec (in 1990), which was forced by firm public opposition against the application of Russian technology in Poland, nuclear community relations with media have been focused mainly on convincing journalists that it is necessary to familiarize the public with ionizing radiation. All sociological studies indicate that Poles associate the radiation (and nuclear energy) with hazards for human health and environment. The situation is further complicated by the fact that Poland borders Russia, Lithuania and Ukraine, i.e. the countries where RBMK reactors are in operation, which - especially after the Chernobyl disaster - has been the reason of grave public concerns. The fears of radiation hazards probably account for the aversion of the majority of Poles to nuclear power in our country and for their susceptibility to various unfounded (and often improbable) rumours of nuclear accidents in the neighbouring countries. A way to overcome the fear of nuclear is to demonstrate the benefits gained from nuclear technologies and to show by personal testimony that nuclear reactors are no more dangerous than other industrial facilities are. Journalists have a special role to play here, as their readers, listeners and viewers trust them more than representatives of nuclear sector. Thus, before journalists decided to persuade the public, it was necessary to persuade the journalists themselves. The following factors are considered important in the contacts with the media: prompt, accurate and specific answers, the respondent's authority, honesty, truthfulness and knowledge, and - as far as possible - 'face to face' communication. The ability to supply additional background information is also valuable, as well as that to communicate in a language easily

Sound velocity in the coolant of boiling nuclear reactors

International Nuclear Information System (INIS)

Proskuryakov, K.N.; Parshin, D.A.; Novikov, K.S.; Galivec, E.Yu.

2009-01-01

To prevent resonant interaction between acoustic resonance and natural frequencies of FE, FA and RI oscillations, it is necessary to determine the value of EACPO. Based on results of calculations of EACPO and natural frequencies of FR, FA and RI oscillations values, it would be possible to reveal the dynamical loadings on metal that are dangerous for the initiation of cracking process in the early stage of negative condition appearance. To calculate EACPO it is necessary to know the Speed Velocity in Coolant. Now we do not have any data about real values of such important parameter as pressure pulsations propagation velocity in two phase environments, especially in conditions with variations of steam content along the length of FR, with taking into account the type of local resistances, flow geometry etc. While areas of resonant interaction of the single-phase liquid coolant with equipment and internals vibrations are estimated well enough, similar estimations in the conditions of presence of a gas and steam phase in the liquid coolant are inconvenient till now. Paper presents results of calculation of velocity of pressure pulsations distribution in two-phase flow formed in core of RBMK-1000 reactors. Feature of the developed techniques is that not only thermodynamic factors and effect of a speed difference between water and steam in a two phase flow but also geometrical features of core, local resistance, non heterogeneity in the two phase environment and power level of a reactor are considered. Obtained results evidence noticeable decreasing of velocity propagation of pressure pulsations in the presence of steam actions in the liquids. Such estimations for real RC of boiling nuclear reactors with steam-liquid coolant are obtained for the first time. (author)

Safety assessment and improvement of Ignalina NPP against downcomer ruptures outside Accident Localisation System

International Nuclear Information System (INIS)

Rimkevicius, S.; Urbonavicius, E.

2002-01-01

Accident Localisation System (ALS) of Ignalina NPP is a pressure suppression type confinement, designed to prevent the release of contaminated steam-water mixture to the environment in case of Loss-of-Coolant Accident (LOCA). One of the peculiarities of Ignalina NPP with RBMK-1500 reactors is that not all of the reactor coolant circuit is enclosed within ALS. Some part of downcomers, that connect Drum Separator (DS) and suction header of main circulation pump is located outside ALS. In case of downcomer rupture in DS compartment the discharge is not confined, but flows to the environment through the safety panels installed in the ceiling of DS compartments. Numerous safety analyses were performed to assess the safety of Ignalina NPP against downcomer break outside ALS, and results were used for different applications in order to improve the safety of the plant. This paper presents the overview of the performed analyses, recommendations raised and safety improvements made to enhance the safety level of NPP. One of the applications is to present the recommendations for safety improvement if maximal allowable pressure limits are exceeded. The calculations results demonstrate that in the case of two downcomers rupture in drum separators compartment the maximum permissible pressure in the reactor hall could be exceeded. The knock-out panels from the reactor hall to the environment were recommended and installed for reactor hall overpressure protection. The evaluation of the drainage system efficiency from DS compartments was performed. In this case the especial attention was paid to analyse the water collection and drainage system behaviour in long term after postulated breaks. The analysis results showed that the modernization of the drainage system prevents the accumulation of the released water in the compartments even in the case of two downcomer pipes ruptures, and decreases the release of radioactive fission products (FP) to the environment.(author)

Review of the IAEA Nuclear Fuel Cycle Materials Section activities related to WWER fuel

International Nuclear Information System (INIS)

Killeen, J.

2003-01-01

The IAEA Nuclear Fuel Cycle Programme, designated as Programme B, has the main objective of supporting Member States in policy making, strategic planning, developing technology and addressing issues with respect to safe, reliable, economically efficient, proliferation resistant and environmentally sound nuclear fuel cycle. This paper is concentrated on describing the work within Sub-programme B.2 'Fuel Performance and Technology'. Two Technical Working Groups assist in the preparation of the IAEA programme in the nuclear fuel cycle area - Technical Working Group on Water Reactor Fuel Performance and Technology and Technical Working Group on Nuclear Fuel Cycle Options. The activities of the Unit within the Nuclear Fuel Cycle and Materials Section working on Fuel Performance and Technology are given, based on the sub-programme structure of the Agency programme and budget for 2002-2003. Within the framework of Co-ordinated Research Projects a study of the delayed hydride cracking (DHC) of the zirconium alloys used in pressurised heavy water reactors (PHWR) involving 10 countries has been completed. It achieved very effective transfer of know-how at the laboratory level in three technologically important areas: 1) Controlled hydriding of samples to predetermined levels; 2) Accurate measurement of hydrogen concentrations at the relatively low levels found in pressure tubes and RBMK channel tubes; and 3) In the determination of DHC rates under various conditions of temperature and stress. A new project has been started on the 'Improvement of Models used for Fuel Behaviour Simulation' (FUMEX II) to assist Member States in improving the predictive capabilities of computer codes used in modelling fuel behaviour for extended burnup. The IAEA also collaborates with organisations in the Member States to support activities and meetings on nuclear fuel cycle related topics

News of the world

International Nuclear Information System (INIS)

Anon.

2012-01-01

This document gathers pieces of news from the nuclear industry around the world. The most relevant are the following. EDF has inaugurated a logistic hub for the supply of spare parts for its 58 operating reactors. Russia has opened a new site to store spent fuels from RBMK reactors. This site is located at Zheleznogorsk near Krasnoiarsk in Siberia. The capacity of the La Hague fuel reprocessing plant is 1700 tonnes a year but the plant processes only between 800 and 1000 tones because most of its foreign contracts have come to an end and have not been renewed. In 2012 the plant is expected to process 1003 tonnes for EDF and 12 tonnes for The Netherlands. AREVA has delivered to the CNNC Chinese company 700 fuel assemblies and 800 control rod clusters. The French Institute for Radioprotection and Nuclear Safety (IRSN) said that there was neither health nor environmental hazards on French soil due to the Fukushima accident. The French Academy of Sciences has highlighted the least sanitary impact of nuclear power compared to other energies. The American Nuclear Safety Association has stated that the American nuclear power plants are safe and that the probability of a severe accident is very low. A new study shows an excess of cases of leukemia near nuclear power stations in France. This study rests on very few statistical cases. An opinion survey in the United Kingdom shows that the construction of nuclear power stations is considered as the best investment in infrastructures. EDF has planned to recruit in 2012 about 6000 people essentially in the nuclear sector. The Netherlands government has given its consent for the construction of the high flux reactor Pallas on the Petten site, this reactor will replace the HFR whose lifetime is over 50 years. (A.C.)

Report of a consultants meeting on control rod insertion reliability for WWER-1000 nuclear power plants. Extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

International Nuclear Information System (INIS)

1995-09-01

Starting from 1992, an increased drop time of control rods exceeding the design limit of four seconds has been observed in most of the operating WWER-1000 reactors in Russia and in the Ukraine. In some cases a dropped control rod became stuck in an intermediate position near the bottom of the core. In October 1994, a similar control rod problem was also observed at Unit 6 of the Kozloduy NPP. The issue of control rod insertion reliability was considered at a consultants' meeting on ''Core Control and Protection Strategy of WWER-1000 Reactors'' in April 1994. A consultants' meeting specifically focused on ''Control Rod Insertion Reliability'' was convened in Vienna in February 1995 attended by 15 international experts. The objectives of this meeting were: The exchange of international experience on problems and solutions related to anomalous control rod insertion; judgement of the safety concern of this issue for WWER-1000 reactors based on safety analyses; consideration of regulatory requirements and interim measures to continue operation in short term including modifications implemented or planned; and, status of root cause analyses and pending problems. The technical discussions were held in plenary sessions and in three working groups devoted to specific aspects of the issue. Refs, figs, tabs

Safety issues and their ranking for WWER-1000 model 320 nuclear power plants. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

International Nuclear Information System (INIS)

1996-03-01

The objective of this report is to present a consolidated list of safety deficiencies, called safety issues, ranked according to their safety significance and the corrective measures to improve overall safety. It is intended for use as a reference to facilitate the development of plant specific safety improvement programmes and to serve as a basis for reviewing their implementation. To the extent that information was made available to the IAEA, the country/plant specific status with respect to each safety issue is described. Section 2 provides an overview of the impact of the relevant issues on the main safety functions in different operational conditions and other aspects important to overall plant safety. A summary of the safety issues and their respective ranking is given in Tables 1 and 2 at the end of Section 2. Section 3 deals with individual safety issues identified in the design which are presented according to the structure below. Section 4 presents the safety issues related to operational safety according to a similar structure but without the ranking. 73 refs, 3 tabs

Safety issues and their ranking for WWER-1000 model 320 nuclear power plants. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

International Nuclear Information System (INIS)

1997-04-01

The objective of this report is to present a consolidated list of safety deficiencies, called safety issues, ranked according to their safety significance and the corrective measures to improve overall safety. It is intended for use as a reference to facilitate the development of plant specific safety improvement programmes and to serve as a basis for reviewing their implementation. To the extent that information was made available to the IAEA, the country/plant specific status with respect to each safety issue is described. Section 2 provides an overview of the impact of the relevant issues on the main safety functions in different operational conditions and other aspects important to overall plant safety. A summary of the safety issues and their respective ranking is given in Tables 1 and 2 at the end of Section 2. Section 3 deals with individual safety issues identified in the design which are presented according to the structure below. Section 4 presents the safety issues related to operational safety according to a similar structure but without the ranking

Safety issues and their ranking for WWER-440 model 213 nuclear power plants. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

International Nuclear Information System (INIS)

1996-04-01

The objective of this report is to present a consolidated list of generic safety concerns, called safety issues, ranked according to their safety significance and the corrective measures to improve safety. It is intended for use as a reference to facilitate the development of plant specific safety improvement programmes and to serve as a basis for reviewing their implementation. Section 2 provides and overview of the impact of all relevant issues on the main safety functions and other aspects important to overall plant safety. Section 3 presents safety issues identified in design according to the structure described below. Section 4 presents the safety issues in the area of operation, according to the same structure except that no ranking is given. At the end of Section 2, Tables 1 and 2 present a summary of all safety issues in a tabular form. 138 refs, tabs

Implications of the Chernobyl accident for Protective Action Guidance

International Nuclear Information System (INIS)

Miller, Charles W.; Pepper, Andrea J.

1989-01-01

The accident that occurred at Unit 4 of the nuclear power station at Chernobyl in the Union of Soviet Socialist Republics on April 26, 1986, was the worst accident in the history of nuclear power. Thirty-one workers and emergency personnel died and more than 200 site personnel were hospitalized as a result of this event Approximately 135,000 persons within 30 km around the reactor were evacuated, and radioactive debris was spread throughout the Northern Hemisphere. There was much public concern generated around the world, and an increased risk of fatal cancel in the world's population is possible as a result of exposure to Chernobyl fallout (USNRC, 1987a). Since the time the Chernobyl accident occurred, many authoritative studies have been published, e.g. USNRC, 1987a. In these studies, differences in design between commercial U.S. reactors and the RBMK pressure-tube reactor at Chernobyl have been emphasized, e.g. USNRC, 1987b. While significant differences in design do exist between these reactors, we believe there are still significant lessons to be learned from the Chernobyl accident for U.S. reactors. The purpose of this paper is to summarize some of the major lessons to be learned related to protective action guidance. The Illinois Department of Nuclear Safety (IDNS) has identified three areas related to protective action guidance for food and water where implications can be drawn from Chernobyl for the U.S.: (1) uniformity of Protective Action Guides (PAGs), (2) incompleteness of U.S. PAGs, and (3) international communications. Following the Chernobyl accident, a variety of protective actions were undertaken by various nations. Furthermore, these actions were initiated, modified, and terminated at different times in different places and, in some instances, were applied on a local or regional basis rather than a national basis (Goldman et al., 1987). One result of this differing application of PAGs was the generation of considerable confusion among decision

DDG-NS statement at the opening of the international conference 'Chernobyl: Looking back to go forwards'

International Nuclear Information System (INIS)

Taniguchi, T.

2005-01-01

The IAEA, as a specialised nuclear-related technical UN agency, has been involved in the mitigation of the Chernobyl accident consequences since early May 1986 when former Director General Hans Blix had visited Chernobyl in order to observe the physical damage and to discuss further actions. The IAEA took on many projects related to technical assistance, technical co-operation and research - with several immediate and longer term goals: first, to mitigate the accident's radiological, environmental and health consequences; second, to improve the overall safety of other RBMK reactors; and third, to understand and disseminate globally those lessons that could be learned from the Chernobyl experience. The projects executed between 1986 and 2005 covered the full range of topics: radiation, waste and nuclear safety; monitoring human exposure; environmental restoration of contaminated land; treatment of people living in the affected areas; and development of special measures to reduce exposure levels. The largest project took place in 1990. Over a two year period, the Agency coordinated the efforts of some 200 international experts over a two year period to complete an independent assessment of the consequences of the Chernobyl accident. Many missions to the three most affected countries were conducted and many meetings were held. The Agency has also organized or supported numerous international meetings to foster information exchange and to promote further assessment of the accident's radiological consequences. The Agency continues its ongoing activities regarding the mitigation of the accident's radiological consequences as part of the UN strategy 'Human Consequences of the Chernobyl Accident - A Strategy for Recovery' launched in 2002. Further IAEA commitment in continued Chernobyl-related activities, mainly in nuclear and radiation safety fields, may involve the following areas: Safety of Shelter decommissioning, Safety of radioactive waste management in the Chernobyl

Modeling of irradiated graphite {sup 14}C transfer through engineered barriers of a generic geological repository in crystalline rocks

Energy Technology Data Exchange (ETDEWEB)

Poskas, Povilas; Grigaliuniene, Dalia, E-mail: Dalia.Grigaliuniene@lei.lt; Narkuniene, Asta; Kilda, Raimondas; Justinavicius, Darius

2016-11-01

There are two RBMK-1500 type graphite moderated reactors at the Ignalina nuclear power plant in Lithuania, and they are under decommissioning now. The graphite cannot be disposed of in a near surface repository, because of large amounts of {sup 14}C. Therefore, disposal of the graphite in a geological repository is a reasonable solution. This study presents evaluation of the {sup 14}C transfer by the groundwater pathway into the geosphere from the irradiated graphite in a generic geological repository in crystalline rocks and demonstration of the role of the different components of the engineered barrier system by performing local sensitivity analysis. The speciation of the released {sup 14}C into organic and inorganic compounds as well as the most recent information on {sup 14}C source term was taken into account. Two alternatives were considered in the analysis: disposal of graphite in containers with encapsulant and without it. It was evaluated that the maximal fractional flux of inorganic {sup 14}C into the geosphere can vary from 10{sup −} {sup 11} y{sup −} {sup 1} (for non-encapsulated graphite) to 10{sup −} {sup 12} y{sup −} {sup 1} (for encapsulated graphite) while of organic {sup 14}C it was about 10{sup −} {sup 3} y{sup −} {sup 1} of its inventory. Such difference demonstrates that investigations on the {sup 14}C inventory and chemical form in which it is released are especially important. The parameter with the highest influence on the maximal flux into the geosphere for inorganic {sup 14}C transfer was the sorption coefficient in the backfill and for organic {sup 14}C transfer – the backfill hydraulic conductivity. - Highlights: • Graphite moderated nuclear reactors are being decommissioned. • We studied interaction of disposed material with surrounding environment. • Specifically {sup 14}C transfer through engineered barriers of a geological repository. • Organic {sup 14}C flux to geosphere is considerably higher than inorganic

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

Science.gov (United States)

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

2016-12-01

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

Main ways for solution of nuclear power plant safety in the USSR

International Nuclear Information System (INIS)

Sidorenko, V.A.; Kovalevich, O.M.; Kramerov, A.Ya.; Bagdasarov, Yu.E.

1977-01-01

In this paper the principles of provisions of nuclear power plant (NPP) safety in the USSR based on the accumulated scientific and engineering experience of design, construction and operating of NPP are discussed. Methods for the decision of the problem were different in various stages of development of nuclear energy and they will be changed in the light of future experience and when the scope of NPP utilization will be broadened. It was formulated the complex of means and ways for providing the real safety of NPP. It includes both the technical and organizing measures and all they are important to solve the NPP safety problem. The first two of these conditions of providing the safety - the high quality of NPP equipment and its constant control during the whole operation of NPP - have to minimize possible failures of power plant elements capable of resulting in an accident conditions. Design and utilization of effective safety systems at NPP and localization systems - these are the next two conditions to provide the safety called upon either to exclude or to attenuate up to the reasonable level the radioactive product yield outside the determined boundaries. The normalization of safety problems and the effective state inspection after keeping the required actions are the next necessary links of this complex called upon to ensure the fulfillment of all safety requirements. The report concerns the realization of formulated procedure to the provision of NPP safety with reactors of WWER and RBM-K types are being put into operation in the Soviet Union. The versatility of approach to the safety provision allows to attain the required safety by various means depending on the type of reactor, its siting and other features concentrating attention to one or another factors. Alongside with the above-mentioned, the paper considers the special requirements and safety measures regarding reactors of next generation such as liquid metal breeders of BN-type and other systems [ru

Experience of international projects implementation at Leningrad Nuclear Power Plant

International Nuclear Information System (INIS)

Zavialov, L.A.

2008-01-01

During the period of 1992-2007 more than 60 different projects of different specificity and budget have been successfully implemented in frames of Technical Assistance for the Commonwealth of Independent States (TACIS) Program, Project financed by European Bank for Reconstruction and Development (EBRD), as well as in frames of Agreements on Cooperation between Leningrad NPP and Radiation and Nuclear safety Authority of Finland (STUK) and Swedish Nuclear Power Inspectorate, International Co-operation Program SKI-ICP(SIP). All these projects were directed to the safety increasing of the Leningrad NPP reactor, type RBMK-1000. Implementation of the technical aid projects has been performed by different foreign companies such as Aarsleff Oy, (Finland), SGN (France), Nukem (Germany), Jergo AB (Sweden), SABAROS (Switzerland), Westinghouse (USA), Nordion (Canada), Bruel and Kjer (Denmark), Data System and Solutions (UK), SVT Braundshuz (Germany) WICOTEC (Sweden), Studsvik (Sweden) and etc. which has enough technical and organizational experience in implementation of such projects, as well as all necessary certificates and licenses for works performance. Selection of a Contractor/Supplier for a joined work performance has been carried out in accordance with the tender procedure, technical specification and a planned budget. Project financing was covered by foreign Consolidated Funds and Authorities interested in increasing of Leningrad NPP safety, which have valid intergovernmental agreements with Russian Federation on the technical assistance to be provided to the NPPs. At present time all joined international projects implemented at Leningrad NPP are financed jointly with LNPP. All projects can be divided into technical aid projects connected with development and turnkey implementation of systems and complexes and projects for supply of equipment which has no analogues in Russia but successfully used all over the world. Positive experience of the joined projects

Experience of international projects implementation at Leningrad Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

Zavialov, L.A. [Leningrad Nuclear Power Plant ' Rosenergoatom' , Leningrad Region, 188540, Sosnovy Bor (Russian Federation)

2008-07-01

During the period of 1992-2007 more than 60 different projects of different specificity and budget have been successfully implemented in frames of Technical Assistance for the Commonwealth of Independent States (TACIS) Program, Project financed by European Bank for Reconstruction and Development (EBRD), as well as in frames of Agreements on Cooperation between Leningrad NPP and Radiation and Nuclear safety Authority of Finland (STUK) and Swedish Nuclear Power Inspectorate, International Co-operation Program SKI-ICP(SIP). All these projects were directed to the safety increasing of the Leningrad NPP reactor, type RBMK-1000. Implementation of the technical aid projects has been performed by different foreign companies such as Aarsleff Oy, (Finland), SGN (France), Nukem (Germany), Jergo AB (Sweden), SABAROS (Switzerland), Westinghouse (USA), Nordion (Canada), Bruel and Kjer (Denmark), Data System and Solutions (UK), SVT Braundshuz (Germany) WICOTEC (Sweden), Studsvik (Sweden) and etc. which has enough technical and organizational experience in implementation of such projects, as well as all necessary certificates and licenses for works performance. Selection of a Contractor/Supplier for a joined work performance has been carried out in accordance with the tender procedure, technical specification and a planned budget. Project financing was covered by foreign Consolidated Funds and Authorities interested in increasing of Leningrad NPP safety, which have valid intergovernmental agreements with Russian Federation on the technical assistance to be provided to the NPPs. At present time all joined international projects implemented at Leningrad NPP are financed jointly with LNPP. All projects can be divided into technical aid projects connected with development and turnkey implementation of systems and complexes and projects for supply of equipment which has no analogues in Russia but successfully used all over the world. Positive experience of the joined projects

Microstructure and textural characterization of hot extruded Zr-2.5Nb alloy PHWR pressure tube fabricated by various ingot processing route

International Nuclear Information System (INIS)

Vaibhaw, Kumar; Jha, S.K.; Saibaba, N.; Neogy, S.; Mani Krishna, K.V.; Srivastava, D.; Dey, G.K.

2011-01-01

Zr-2.5 Nb alloys finds its applications as a pressure tube component in pressure tube type thermal reactors such as PHWRs and RBMK due to properties attributed such as low neutron absorption cross section, high temperature strength and corrosion resistance etc. Manufacturing of this life time components involves series of thermo-mechanical processes of hot working and cold working with intermediate annealing. The life time of Pressure tube are limited due to their diametral creep properties which is governed by metallurgical characteristics such as texture, microstructure dislocation density etc. The primary breakdown of cast structure in Vacuum Arc Melted ingot can be effected by either hot extrusion or forging in single or multiple stages before final hot extrusion step into the blank for manufacturing of seamless pressure tube. Elevated temperature deformation carried out in hot working above the recrystallization temperature would enable impositions of large strains in single step. This deformation causes a significant change in the microstructure of the material and depends on process parameters such as extrusion ratio, temperature and strain rate. Basic microstructure developed at this deformation stage has significant bearing on the final properties of the material fabricated with subsequent cold working steps. The major texture in α+β Zr-2.5 Nb alloy is established during final extrusion to blank which does not change significantly during subsequent cold pilgering. However, microstructure is modified significantly in subsequent cold working which can be effected by cold pilgering or cold drawing in single or multiple steps. Present paper brings out the various ingot processing routes using forging and or extrusion followed for fabrication of pressure tubes. The development of texture and microstructures has been discussed at the blank stage from these processing routes and also with respect to varying extrusion variable such as extrusion ratio

Pyrolysis and its potential use in nuclear graphite disposal

International Nuclear Information System (INIS)

Mason, J.B.; Bradbury, D.

2001-01-01

Graphite is used as a moderator material in a number of nuclear reactor designs, such as MAGNOX and AGR gas cooled reactors in the United Kingdom and the RBMK design in Russia. During construction the moderator of the reactor is usually installed as an interlocking structure of graphite bricks. At the end of reactor life the graphite moderator, weighing typically 2,000 tonnes, is a radioactive waste which requires eventual management. Radioactive graphite disposal options conventionally include: In-situ SAFESTORE for extended periods to permit manual disassembly of the graphite moderator through decay of short-lived radionuclides. Robotic or manual disassembly of the reactor core followed by disposal of the graphite blocks. Robotic or manual disassembly of the reactor core followed by incineration of the graphite and release of the resulting carbon dioxide Studsvik, Inc. is a nuclear waste management and waste processing company organised to serve the US nuclear utility and government facilities. Studsvik's management and technical staff have a wealth of experience in processing liquid, slurry and solid low level radioactive waste using (amongst others) pyrolysis and steam reforming techniques. Bradtec is a UK company specialising in decontamination and waste management. This paper describes the use of pyrolysis and steam reforming techniques to gasify graphite leading to a low volume off-gas product. This allows the following options/advantages. Safe release of any stored Wigner energy in the graphite. The process can accept small pieces or a water-slurry of graphite, which enables the graphite to be removed from the reactor core by mechanical machining or water cutting techniques, applied remotely in the reactor fuel channels. In certain situations the process could be used to gasify the reactor moderator in-situ. The low volume of the off-gas product enables non-carbon radioactive impurities to be efficiently separated from the off-gas. The off-gas product can

Report of a consultants' meeting on insights from PSA results on the programmes for safety upgrading of WWER NPPs. Extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

International Nuclear Information System (INIS)

1996-11-01

The objective of the meeting was to compare the insight from plant specific PSA studies with the safety upgrading programme of WWER NPPs. The PSAs were reviewed considering the scope, level and detail of PSA models and results of IAEA peer reviews. Safety improvements which are not normally included in PSAs were also considered. The review specifically considers for each plant specific PSA: the dominant initiating events and accident sequences contributing to core damage; and, the importance of systems, components and human actions to be used for prioritizing actions. 4 refs, tabs

Technical basis for instrumentation and control design improvements in WWER-440/230 nuclear power plants. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

International Nuclear Information System (INIS)

1996-01-01

Instrumentation and control (I and C) has been recognized as an area which requires substantial improvements in WWER NNPs, particularly for model 230 plants. Under contract with the IAEA the Spanish company Empresarios Agrupados (EA) developed a basic document proposing a technical basis for improvements related to the following most significant aspects of I and C: criteria for safety classification; remote shutdown panel; I and C support to operation and control room design; instrumentation set point margins; accident monitoring instrumentation. This publication is derived from the original report of EA which was circulated by the IAEA for review by staff members and experts from various Member States. It was finally agreed upon at a Consultants' meeting convened by the IAEA in Vienna in May 1994 with the participation of experts from France, Germany and Spain. The guidance expressed in this report is based on the IAEA/NUSS standards, safety guides and practices, and on regulations in use in various Member States. It is proposed as a way of carrying out the necessary studies to improve safety by upgrading the vital part of instrumentation an control in WWER-440 model 230 nuclear power plants. 28 refs, 3 figs

Chernobyl, 13 years after

International Nuclear Information System (INIS)

Regniault-Lacharme, Mireille; Metivier, Henri

1999-04-01

This is an annual report, regularly issued by IPSN, that presents the ecological and health consequences of the Chernobyl Nuclear Accident. The present status of the Chernobyl Nuclear Plant, which Ukraine engaged to stop definitively in year 2000, is summarized. The only reactor unit now in operation is Chernobylsk-3 Reactor which poses two safety questions: evolution of cracks in part of the tubing and behaviour of the pressure tubes. Although, some improvements in the RBMK reactor types were introduced, problems remain that make IPSN to stress the requirement of stopping this NPP completely. In the contaminated territories surrounding Chernobyl incidence rate of infant thyroid cancers continues to grow, reaching values 10 to 100 times higher than the natural rate. In France the IPSN analyzed 60,000 records carried out in 17 sites during May 1986 and April 1989. It was estimated that the individual dose received during 60 years (1986-2046) by the inhabitants of the most affected zone (eastern France) is lower than 1.5 mSv, a value lower than 1% of the natural cosmic and telluric radioactivity exposure for the same period. For the persons assumed to live in the most attacked forests (from eastern France) and nourishing daily with venison and mushrooms the highest estimate is 1 mSv a year. Concerning the 'hot spots', identified in mountains by IPSN and CRIIRAD, the doses received by excursionists are around 0.015 mSv. For an average inhabitant of the country the dose piled up in the thyroid due to iodine-131 fallout is estimated to 0.5-2 mSv for an adult and 6.5-16 mSv for an infant. These doses are 100 to 1000 times lower than the ones to which the infants living in the neighbourhood of Chernobyl are exposed to. The contents of the report is displayed in the following six chapters: 1. Chernobyl in some figures; 2. The 'sarcophagus' and the reactors of the Chernobyl NPP; 3. Health consequences of the Chernobyl accident;. 4. The impact of Chernobyl fallout in France

Influence of temperature on δ-hydride habit plane in α-Zirconium

International Nuclear Information System (INIS)

Singh, R. N.; Stahle, P.; Banerjee, S.; Ristmanaa, Matti; Sauramd, K.

2008-01-01

Dilute Zr-alloy with hcp α-Zr as major phase is used as pressure boundary for hot coolant in CANDU, PHWR and RBMK reactors. Hydrogen / deuterium ingress during service makes the pressure boundary components like pressure tubes of the aforementioned reactors susceptible to hydride embrittlement. Hydride acquires plate shaped morphology and the broad face of the hydride plate coincides with certain crystallographic plane of α-Zr crystal, which is called habit plane. Hydride plate oriented normal to tensile stress significantly increases the degree of embrittlement. Thus key to mitigating the damage due to hydride embrittlement is to avoid the formation of hydride plates normal to tensile stress. Two different theoretical approaches are used to determine the habit plane of precipitates viz., geometrical and solid mechanics. For the geometrical approach invariant plane and invariant-line criteria have been applied successfully and for the solid mechanics approach strain energy minimization criteria have been used successfully. Solid mechanics approach using strain energy computed by FEM technique has been applied to hydride precipitation in Zr-alloys, but the emphasis has been to understand the solvus hysteresis. The objective of the present investigation is to predict the habit plane of δ-hydride precipitating in α-Zr at 25, 300, 400 and 450 .deg. C. using strain energy minimization technique. The δ-hydride phase is modeled to undergo isotropic elastic and plastic deformation. The α-Zr phase was modeled to undergo transverse isotropic elastic deformation. Both isotropic plastic and transverse isotropic plastic deformations of α-Zr were considered. Further, both perfect and linear work-hardening plastic behaviors were considered. Accommodation strain energy of δ-hydrides forming in α-Zr crystal was computed using initial strain method as a function of hydride nuclei orientation. Hydride was modeled as disk with circular edge. The simulation was carried out

A fire hazard analysis at the Ignalina nuclear power plant

International Nuclear Information System (INIS)

Joerud, F.; Magnusson, T.

1998-01-01

The fire hazard analysis (FHA) of the Ignalina Nuclear Power Plant (INPP) Unit no.1 was initiated during 1997 and is estimated to finalise in summer 1998. The reason for starting a FHA was a recommendation in the Safety Analysis Report and its review to prioritise a systematic FHA. Fire protection improvements had earlier been based on engineering assessments, but further improvements required a systematic FHA. It is also required by the regulator for licensing of unit no.1. In preparation of the analysis it was decided to perform a deterministic FHA to fulfil the requirements in the IAEA draft of a Safety Practice ''Preparation of Fire Hazard Analyses for Nuclear Power Plants''. As a supporting document the United States Department of Energy Reactor Core Protection Evaluation Methodology for Fires at RBMK and WWER Nuclear Power Plants (RCPEM) was agreed to be used. The assistance of the project is performed as a bilateral activity between Sweden and UK. The project management is the responsibility of the INPP. In order to transfer knowledge to the INPP project group, training activities are arranged by the western team. The project will be documented as a safety case. The project consists of parties from INPP, Sweden, UK and Russia which makes the project very dependent of good communication procedures. The most difficult problems is except from the problems with translation, the problems with different standards and lack of testing protocols of the fire protection installations and problems to set the right level of screening criteria. There is also the new dimension of making it possible to take credit for the fire brigade in the safety case, which can bring the project into difficulties. The most interesting challenges for the project are to set the most sensible safety levels in the screening phase, to handle the huge volume of rooms for survey and screening, to maintain the good exchange of fire- and nuclear safety information between all the parties involved

Outline of the presentation on experiences with safety culture from an EDF viewpoint

International Nuclear Information System (INIS)

Tanguy, P.Y.

1994-01-01

Safe operation must draw on an industrial culture based professionalism and strict enforcement. Experience, from abroad and in France, shows that this is not enough to reach the degree of quality we are aiming at. In addition we need an attitude from both men an organizations that questions, when necessary, ways of thinking and working, to give safety the attention it deserves according to its importance. Before TMI, the nuclear community was convinced that the safety approach that has been progressively built up, on the basis of the research performed in various countries, and with the assistance of multiple committees directing their attention to safety matters, has reached the point where the prevention was so effective that a major accident has become 'incredible'. The enquiry commissions that were convened after the accident identified a common mindset: the 'beyond design accident' will not occur. Today we all know that the TMI accident was 'due to happen' some day or another, in some western nuclear power plant. Before Chernobyl, the nuclear community did not think possible a single accident in a nuclear power plant could have socioeconomical consequences 1500 kilometres away from the site, where people would throw away food products because of their radioactivity content. The first independent Soviet enquiry commission on Chernobyl, under the chairmanship of Dr. Steinberg, published its report in Moscow in June 1991, and concluded that the Chernobyl accident was due to happen some time in some RBMK plant. Today we have no excuse if we pretend that, since all measures have been taken to prevent severe accidents in LWR's, we don't have to worry any more about them. We have to be prepared to face them ,and to protect efficiently the plant personnel, the site environment and the public if such a severe accident does occur in an unpredictable way. But we should not consider that such accident is as sure as fate. We know the way to prevent it. The answer is

Earthquake experience and seismic qualification by indirect methods in nuclear installations

International Nuclear Information System (INIS)

2003-01-01

In recent years, many operational nuclear power plants around the world have conducted seismic re-evaluation programmes either as part of a review of seismic hazards or to comply with best international nuclear safety practices. In this connection, Member States have called on the IAEA to carry out several seismic review missions at their plants, primarily those of WWER and RBMK design. One of the critical safety issues that arose during these missions was that of seismic qualification (determination of fitness for service) of already installed plant distribution systems, equipment and components. The qualification of new components, equipment and distribution systems cannot be replicated for equipment that is already installed and operational in plants, as this process is neither feasible nor appropriate. For this reason, seismic safety experts have developed new procedures for the qualification of installed equipment: these procedures seek to demonstrate that installed equipment, through a process of comparison with new equipment, is apt for service. However, these procedures require large sets of criteria and qualification databases and call for the use of engineering judgement and experience, all of which open the door to wide margins of interpretation. In order to guarantee a sound technical basis for the qualification of in-plant equipment, currently applied to 70% to 80% of all plant equipment, the regulatory review of this type of qualification process calls for a detailed assessment of the technical procedures applied. Such an assessment is the first step towards eliminating the risk of large differences in qualification results between different plants, operators and countries, and guaranteeing the reliability of seismic re-evaluation programmes. Bearing this in mind, in 1999, the IAEA convened a seminar and technical meeting on seismic qualification under the auspices of the IAEA Technical Co-operation programme. Altogether 66 senior experts attended the

Scheme of higher-density storage of spent nuclear fuel in Chernobyl NPP interim storage facility no. 1

International Nuclear Information System (INIS)

Britan, P.M.

2008-01-01

On 29. March 2000 the Cabinet of Ministers of Ukraine issued a decree prescribing that the last operating unit of Chernobyl NPP be shut down before its design lifetime expiry. In accordance with the Contract concluded on 14 June 1999 between the National Energy-generating Company 'Energoatom' and the Consortium of Framatome, Campenon Bernard-SGE and Bouygues, in order to store the spent ChNPP fuel a new interim dry storage facility (ISF-2) for spent ChNPP fuel would be built. Currently the spent nuclear fuel (spent fuel assemblies - SFAs) is stored in reactor cooling pools and in the reactors on Units 1, 2, 3, as well as in the wet Interim Storage Facility (ISF-1). Taking into account the expected delay with the commissioning of ISF-2, and in connection with the scheduled activities to build the New Safe Confinement (including the taking-down of the existing ventilation stack of ChNPP Units 3 and 4) and the expiry of the design operation life of Units 1 and 2, it is expedient to remove the nuclear fuel from Units 1, 2 and 3. This is essential to improve nuclear safety and ensure that the schedule of construction of the New Safe Confinement is met. The design capacity of ISF-1 (17 800 SFAs) is insufficient to store all SFAs (21 284) currently on ChNPP. A technically feasible option that has been applied on other RBMK plants is denser storage of spent nuclear fuel in the cooling ponds of the existing ISF-1. The purpose of the proposed modifications is to introduce changes to the ISF-1 design supported by necessary justifications required by the Ukrainian codes with the objective of enabling the storage of additional SFAs in the existing storage space (cooling pools). The need for the modification is caused by the requirement to remove nuclear fuel from the ChNPP units as soon as possible, before the work begins to decommission these units, as well as to create safe conditions for the construction of the New Safe Confinement over the existing Shelter Unit. (author)

Blowdown and rewetting characteristics for AHWR under postulated LOCA - an analytical study

International Nuclear Information System (INIS)

Mukhopadhyay, D.; Chatterjee, B.

2015-01-01

Advanced Heavy Water Reactor (AHWR) is a thorium fuelled, natural circulation driven and heavy water moderated reactor. The cooling of the nuclear fuel is achieved through natural circulation mode for the tube type reactor where hot and cold leg of the reactor has been designed to be long and high enough to avail the gravity head desired to overcome the hydraulic resistances in the flow path. The natural circulation cooling mode makes AHWR very different as compared to other tube type reactors with forced circulation e.g RBMK. This cooling feature which calls for longer pipes length and elevation head is having an influence on the blowdown characteristic and the initial fuel heatup characteristic of the reactor. Analyses of Loss of Coolant Accident carried out for different break sizes in the inlet header of the reactor identifies two competing transient forces namely 'blowdown force' and 'natural circulation' which act against each other due to virtue of the break location. The flow in the reactor channel is being decided by these two forces and eventually the flow condition decides the fuel heatup. It has been observed through analyses that variation of break sizes from moving smaller break sizes to bigger one (30% to 200%), causes an enhancement in blowdown forces and weakening of driving force for natural circulation as quality appears in cold leg section. A balance of these two forces is observed for 200% break case, causing a sustained flow stagnation condition leading to maximum fuel heat up among all the break cases. The blowdown characterization study is being carried out with RELAP5/mod3.4 code and the influences of transient forces on the fuel heatup are presented. It is concluded that the fuel heat up during blowdown phase is significantly dependent on the two competing forces namely blowdown and natural circulation which eventually depend on break sizes. The mist flow regime remains for a longer period during rewetting phase and the

Chernobyl, 13 years after; Tchernobyl, 13 ans apres

Energy Technology Data Exchange (ETDEWEB)

Regniault-Lacharme, Mireille; Metivier, Henri [Inst. de Protection et de Surete Nucleaire, CEA Centre d' Etudes de Fontenay-aux-Roses, 92 (France)

1999-04-01

This is an annual report, regularly issued by IPSN, that presents the ecological and health consequences of the Chernobyl Nuclear Accident. The present status of the Chernobyl Nuclear Plant, which Ukraine engaged to stop definitively in year 2000, is summarized. The only reactor unit now in operation is Chernobylsk-3 Reactor which poses two safety questions: evolution of cracks in part of the tubing and behaviour of the pressure tubes. Although, some improvements in the RBMK reactor types were introduced, problems remain that make IPSN to stress the requirement of stopping this NPP completely. In the contaminated territories surrounding Chernobyl incidence rate of infant thyroid cancers continues to grow, reaching values 10 to 100 times higher than the natural rate. In France the IPSN analyzed 60,000 records carried out in 17 sites during May 1986 and April 1989. It was estimated that the individual dose received during 60 years (1986-2046) by the inhabitants of the most affected zone (eastern France) is lower than 1.5 mSv, a value lower than 1% of the natural cosmic and telluric radioactivity exposure for the same period. For the persons assumed to live in the most attacked forests (from eastern France) and nourishing daily with venison and mushrooms the highest estimate is 1 mSv a year. Concerning the 'hot spots', identified in mountains by IPSN and CRIIRAD, the doses received by excursionists are around 0.015 mSv. For an average inhabitant of the country the dose piled up in the thyroid due to iodine-131 fallout is estimated to 0.5-2 mSv for an adult and 6.5-16 mSv for an infant. These doses are 100 to 1000 times lower than the ones to which the infants living in the neighbourhood of Chernobyl are exposed to. The contents of the report is displayed in the following six chapters: 1. Chernobyl in some figures; 2. The 'sarcophagus' and the reactors of the Chernobyl NPP; 3. Health consequences of the Chernobyl accident;. 4. The impact of

Upgrading the SPP-500-1 moisture separators-steam reheaters used in the Leningrad NPP turbine units

Science.gov (United States)

Legkostupova, V. V.; Sudakov, A. V.

2015-03-01

The specific features of existing designs of moisture separators-steam reheaters (MSRs) and experience gained with using them at nuclear power plants are considered. Main factors causing damage to and failures of MSRs are described: nonuniform distribution of wet steam flow among the separation modules, breakthrough of moisture through the separator (and sometimes also through the steam reheater), which may lead to the occurrence of additional thermal stresses and, hence, to thermal-fatigue damage to or stress corrosion cracking of metal. MSR failure results in a less efficient operation of the turbine unit as a whole and have an adverse effect on the reliability of the low-pressure cylinder's last-stage blades. By the time the design service life of the SPP-500-1 MSRs had been exhausted in power units equipped with RBMK-1000 reactors, the number of damages inflicted to both the separation part and to the pipework and heating surface tubes was so large, that a considerable drop of MSR effectiveness and turbine unit efficiency as a whole occurred. The design of the upgraded separation part used in the SPP-500-1 MSR at the Leningrad NPP is described and its effectiveness is shown, which was confirmed by tests. First, efforts taken to achieve more uniform distribution of moisture content over the perimeter and height of steam space downstream of the separation modules and to bring it to values close to the design ones were met with success. Second, no noticeable effect of the individual specific features of separation modules on the moisture content was revealed. Recommendations on elaborating advanced designs of moisture separators-steam reheaters are given: an MSR arrangement in which the separator is placed under or on the side from the steam reheater; axial admission of wet steam for ensuring its uniform distribution among the separation modules; inlet chambers with an extended preliminary separation system and devices for uniformly distributing steam flows in the

Experience in arranging shipments of spent fuel assemblies of commercial and research reactors

International Nuclear Information System (INIS)

Komarov, S.; Barinkov, O.; Eshcherkin, A.; Lozhnikov, V.; Smirnov, A.

2008-01-01

At present the key activities of Sosny Company are to inspect physical conditions, handle and arrange shipment of SFA including failed SFA. In 2003 after obtaining the license of Gosatomnadzor (Rostechnadzor now) entitled to handle nuclear materials in the process of their shipment, Sosny Company started preparing certification and arranging SFA shipment on its own. About 40 shipments of SFA were performed with participation of Sosny Company. Experience in handling failed SFA - an example of development of a new technology could be the transport and technological scheme of RBMK-1000 SFA shipment from Leningradskaya NPP that was designed by Sosny Company. TUK-11 cask was selected for this shipment. The example of change of transport and technological scheme is modification of the technology for handling and shipment of WWER-440 SFA from Kola NPP. Experience in arranging transportation - based on the results of development of logistics schemes for shipping SFA of reactor facilities Sosny Company justified and implemented composition of mixed trains containing rail cars of many types that enabled to perform shipment more efficiently in time and cost. Experience in arranging handling and shipment of research reactor SFA - over the past years the activity of Sosny Company was aimed at implementing international Russian Research Reactor Fuel Return (RRRFR) program. Since equipment of the majority of research centers doesn't allow for the large casks to be accepted and loaded, special casks of less mass and dimensions are used to ship SFA from research reactors. In RRRFR program it is assumed to use different casks for RR SFA such as Russian TUK- 19, TUK-128 and foreign SKODA VPVR/M and NAC-LWT. At present Sosny Company is involved in coordination of the efforts of the affected organizations in creating the type 'C' package for RR SFA in the RF. Conclusion: Under conditions of constant increase of the requirements to shipment safety and complication of regulations of all

Risk-Informed Decisions Optimization in Inspection and Maintenance

International Nuclear Information System (INIS)

Robertas Alzbutas

2002-01-01

partial case is used for the construction and research of the models related to inspections and maintenance planning of Ignalina Nuclear Power Plant (RBMK-1500) piping components. The discussed example is related to risk analysis and inspection program improvements for selected pipe systems. The new risk-informed inspection and maintenance program for selected pipe systems are compared with various alternatives. The usage of risk evaluations to optimize the selection of inspection locations, the inspection interval, and the changes in risk and cost due suggested modifications are demonstrated. The proposed integrated modeling methodic and general model of inspection process can be used as a base for other risk-informed models of inspection process control and risk monitors of complex dynamic systems. (authors)

Expert System for Diagnostics and Status Monitoring of NPP Water Chemistry Condition

International Nuclear Information System (INIS)

Shvedova, M.N.; Kritski, V.G.; Zakharova, S.V.; Nikolaev, F.V.; Benediktov, V.B.

2002-01-01

Water chemistry condition (WCC) has been the subject of constant study and improvement up to the present day. It is connected with the presence of a direct relationship between the violation of water chemistry regulation on the one hand and components reliability of the circuit's equipment and cost-effectiveness of their operation on the other. It dictates the necessity to apply different optimization methods in the field of monitoring and use of information - analytical and diagnostic systems to assess WCC quality, control and support. By now NPP experts have broad experience in revealing and removing the causes of WCC disturbances. However this knowledge is often of an intuitive, non-classified nature, scattered among various working documents, which makes their transfer difficult. Based on what has been mentioned above, special attention is currently being paid to the problem of creating expert diagnostic systems for supporting the optimum WCC. The existing developments in this field (DIWA, Smart chem Works, the water quality control system at the Onagava NPP etc. [1,3,4,5] are based on wide use of experts' knowledge. Such expert diagnostic systems for supporting WCC refer to the new generation of intellectual control methods, which allow the incorporation of the latest achievements both in the field of water chemistry simulation and in the field of artificial intelligence and computer technologies. LI 'VNIPIET' employees have, for several years, been developing an expert diagnostic system for supporting WCC and status monitoring of RBMK - reactor NPPs [2]. This system has not only conveniently organized the traditional functions of information acquisition and storage, a complete presentation of information in the form of tables, graphs of a dynamical changes of parameters and formation regular reports, diagnostic functions and issuing recommendations on WCC correction, but it also allows the assessment of confidence in the diagnosis made, relying on a wide

Fourth Generation Reactor Concepts

International Nuclear Information System (INIS)

Furtek, A.

2008-01-01

Concerns over energy resources availability, climate changes and energy supply security suggest an important role for nuclear energy in future energy supplies. So far nuclear energy evolved through three generations and is still evolving into new generation that is now being extensively studied. Nuclear Power Plants are producing 16% of the world's electricity. Today the world is moving towards hydrogen economy. Nuclear technologies can provide energy to dissociate water into oxygen and hydrogen and to production of synthetic fuel from coal gasification. The introduction of breeder reactors would turn nuclear energy from depletable energy supply into an unlimited supply. From the early beginnings of nuclear energy in the 1940s to the present, three generations of nuclear power reactors have been developed: First generation reactors: introduced during the period 1950-1970. Second generation: includes commercial power reactors built during 1970-1990 (PWR, BWR, Candu, Russian RBMK and VVER). Third generation: started being deployed in the 1990s and is composed of Advanced LWR (ALWR), Advanced BWR (ABWR) and Passive AP600 to be deployed in 2010-2030. Future advances of the nuclear technology designs can broaden opportunities for use of nuclear energy. The fourth generation reactors are expected to be deployed by 2030 in time to replace ageing reactors built in the 1970s and 1980s. The new reactors are to be designed with a view of the following objectives: economic competitiveness, enhanced safety, minimal radioactive waste production, proliferation resistance. The Generation IV International Forum (GIF) was established in January 2000 to investigate innovative nuclear energy system concepts. GIF members include Argentina, Brazil, Canada, Euratom, France Japan, South Africa, South Korea, Switzerland, United Kingdom and United States with the IAEA and OECD's NEA as permanent observers. China and Russia are expected to join the GIF initiative. The following six systems

Nuclear Criticality Safety Assessment Using the SCALE Computer Code Package. A demonstration based on an independent review of a real application

International Nuclear Information System (INIS)

Mennerdahl, Dennis

1998-06-01

The purpose of this project was to instruct a young scientist from the Lithuanian Energy Institute (LEI) on how to carry out an independent review of a safety report. In particular, emphasis, was to be put on how to use the personal computer version of the calculation system SCALE 4.3 in this process. Nuclear criticality safety together with radiation shielding from gamma and neutron sources were areas of interest. This report concentrates on nuclear criticality safety aspects while a separate report covers radiation shielding. The application was a proposed storage cask for irradiated fuel assemblies from the Ignalina RBMK reactors in Lithuania. The safety report contained various documents involving many design and safety considerations. A few other documents describing the Ignalina reactors and their operation were available. The time for the project was limited to approximately one month, starting 'clean' with a SCALE 4.3 CD-ROM, a thick safety report and a fast personal computer. The results should be of general interest to Swedish authorities, in particular related to shielding where experience in using advanced computer codes like those available in SCALE is limited. It has been known for many years that criticality safety is very complicated, and that independent reviews are absolutely necessary to reduce the risk from quite common errors in the safety assessments. Several important results were obtained during the project. Concerning use of SCALE 4.3, it was confirmed that a young scientist, without extensive previous experience in the code system, can learn to use essentially all options. During the project, it was obvious that familiarity with personal computers, operating systems (including network system) and office software (word processing, spreadsheet and Internet browser software) saved a lot of time. Some of the Monte Carlo calculations took several hours. Experience is valuable in quickly picking out input or source document errors. Understanding

Reliability analysis of pipe whip impacts

International Nuclear Information System (INIS)

Alzbutas, R.; Dundulis, G.; Kulak, R.F.; Marchertas, P.V.

2003-01-01

A probabilistic analysis of a group distribution header (GDH) guillotine break and the damage resulting from the failed GDH impacting against a neighbouring wall was carried out for the Ignalita RBMK-1500 reactor. The NEPTUNE software system was used for the deterministic transient analysis of a GDH guillotine break. Many deterministic analyses were performed using different values of the random variables that were specified by ProFES software. All the deterministic results were transferred to the ProFES system, which then performed probabilistic analyses of piping failure and wall damage. The Monte Carlo Simulation (MCS) method was used to study the sensitivity of the response variables and the effect of uncertainties of material properties and geometry parameters to the probability of limit states. The First Order Reliability Method (FORM) was used to study the probability of failure of the impacted-wall and the support-wall. The Response Surface (RS/MCS) method was used in order to express failure probability as function and to investigate the dependence between impact load and failure probability. The results of the probability analyses for a whipping GDH impacting onto an adjacent wall show that: (i) there is a 0.982 probability that after a GDH guillotine break contact between GDH and wall will occur; (ii) there is a probability of 0.013 that the ultimate tensile strength of concrete at the impact location will be reached, and a through-crack may open; (iii) there is a probability of 0.0126 that the ultimate compressive strength of concrete at the GDH support location will be reached, and the concrete may fail; (iv) at the impact location in the adjacent wall, there is a probability of 0.327 that the ultimate tensile strength of the rebars in the first layer will be reached and the rebars will fail; (v) at the GDH support location, there is a probability of 0.11 that the ultimate stress of the rebars in the first layer will be reached and the rebars will fail

Ideas in support to the definition of the Phase 6

International Nuclear Information System (INIS)

Rimpault, G.

2004-01-01

Hybrid UOX/MOX fuelled core of the BN-600 reactor was endorsed as an international benchmark. Phases 1 and 2 consist of RZ and HEX-Z homogeneous models of the hybrid version of the BN-600 reactor. Phase 3 consists of RZ and HEX-Z heterogeneous models of the hybrid version of the BN-600 reactor. Phase 4 consists of RZ and HEX-Z heterogeneous models of the full MOX version of the BN-600 reactor. Phase 5 consists of the Analysis of BFS-62 hybrid configuration in support to Phase 3 studies. The background strategy was defined to make the world safer by using weapon grade Plutonium for civil application. Make that use safe by checking the behaviour of the BN-600 core with limited (hybrid core: Phases 1, 2 and 3) and then full use of MOX (Phase 4); Verify uncertainties on reactivity coefficients and especially on SVRE with some BFS-62 experiments (Phase 5) and use of Minor Actinides in the fuel (Phase 6 and possibly Phase 7). The French Strategy was make the link between existing reactors PWR and GEN-IV ones. From 2030 - 2040, Introduction of 4th generation systems was planned. The P4 and N4 PWR reactors will reach 40 years lifetime at 2025-2035. Lifetime extension to 50 years is considered. The replacement of PWR reactors by Gen IV systems will be effective. Proposal of Phase 6 considers to develop a strategy in connection with GEN IV criteria, use BN-600 as a demonstrator of GEN IV cores, use spent fuels from WWERs, RBMKs as a fuel for use in LMFBR (BN-600 being the first in the row). In Russia, there are roughly 9 GWe WWER and 10.2 GWe RBMK reactors. UOX is being used (no MOX being used), burn up rate is 45 GWd/ton. At the moment, no reprocessing is performed but a reasonable scenario is to develop a simplified dry reprocessing or a dry reprocessing to extract both MA and Pu resulting in no separation and limited Proliferation. Pu vector will no longer be weapon grade. There will be no blanket as far as possible. Study the BN-600 behaviour with this type of fuel

Acoustic noise diagnostics of leaks. Report of the E O Paton Electric Welding Institute, Kiev, 1997

International Nuclear Information System (INIS)

Volkov, L.P.; Zverev, A.F.; Kozeletskii, V.F.; Leonov, A.N.

1999-01-01

Special attention has been paid in recent years to safe service of nuclear energy plant (NEP): various safety programmes are being developed and applied, measures are taken in reactors of the WWER-RBMK type, the Gospromatomnadzor is introducing a package of new standard requirements and documents, etc. Formation of leaks in equipment and pipelines, actual leakage and the discharge of the heat carrier outside the limits of contour I of the NEP leads to undesirable and sometimes serious consequences. In most cases, elimination of these consequences is associated with the need to use a large amount of material, personnel, shutdown of nuclear equipment, disruption of the ecological situation, and partial or complete loss of electric or thermal power. The 'leak prior to fracture' concept, accepted several years ago in the EEC countries and the USA has been used as a strong impetus for the rapid development of diagnostic means of inspection of pipelines and equipment, including acoustic equipment. The development of means of acoustic leak detection as a variety of noise diagnostics is still important because the utilisation of the possibilities of the methods of noise diagnostics makes it possible to solve a complex problem: recording the actual leakage event, determination of its position and the degree of risk. The majority of modern leak inspection methods, having a low sensitivity threshold, require development of special testing conditions, the use of indicator or test substances and, most importantly, they should not be remote controlled. For example, the mass spectrometric, halide, luminescent, bubble and capillary methods have a sensitivity threshold of 6.7·10 -11 -6.7·10 -4 m 3 Pa/s, but require the presence of an operator in the inspection area. The use of these methods for determining leakage in pipelines and equipment of nuclear power plant is out of the question and it is not possible to develop a system for remote control of NEP equipment using these

Steam water cycle chemistry of liquid metal cooled innovative nuclear power reactors

International Nuclear Information System (INIS)

Yurmanov, Victor; Lemekhov, Vadim; Smykov, Vladimir

2012-09-01

selection of chemistry controls is vital for NPPs with liquid metal cooled reactors. This paper highlights principles and approaches to chemistry controls in steam/water cycles of future NPPs with innovative liquid metal cooled reactors. The recommendations on how to arrange chemistry controls in steam/water cycles of future NPPs with innovative liquid metal cooled reactors are based taking into account: - the experience with operation of fossil power industry; - secondary side water chemistry of lead-bismuth eutectics cooled nuclear reactors at submarines; - steam/water cycles of NPPs with sodium cooled fast breeders BN-350 and BN-600; - secondary water chemistry at conventional NPPs with WER, RBMK and some other reactors. (authors)

A study of radionuclide dispersion by river systems, using GIS and remote sensing techniques

International Nuclear Information System (INIS)

Borghuis, Sander; Brown, Justin; Steenhuisen, Frits; Skorve, Johnny

2000-01-01

The Krasnoyarsk Mining and Chemical Combine in Zheleznogorsk, Russia, is situated on the banks of the Yenisey river. The combine consists of three RBMK-type graphite moderate reactors, a reprocessing plant for the production of weapons-grade plutonium and storage facilities for nuclear waste. Discharges of radionuclides into the Yenisey river were either part of normal operation procedures or caused by accidental releases (Strand et al., 1997). So far, little is known about the transport and fate of the radioactive contaminants in the areas downstream of the Krasnoyarsk CC that are influenced by the Yenisey river system. Aim is to comprehend the dispersion of radionuclides through the river system. Remotely sensed and field study information are combined in a geographical information system (GIS) to study the processes leading to the dispersion of sediment-bound radionuclides carried by the river system. Since the extent of the study area is several thousands or kilometres of river and adjacent flood plains, use is made of a record of remotely sensed (satellite) images that are handled by the GIS. Panchromatic, high resolution satellite images as well as multispectral Landsat MSS and TM images were compiled for the area of interest. The panchromatic images were taken in a period during which the facility was in operation (1960-1972) and obtained for intervals of circa 6 months. A time series of satellite images enables the identification of erosion and sedimentation zones. The behaviour and fate of particle-reactive radionuclides, e.g. 239,240 Pu and to large extent 137 Cs, will be closely related to the movement of sediment. With respect to the behaviour and fate of more conservative radionuclides as 90 Sr, information is required accounting for fractionation between the particulate and aqueous phases. Stereo images are used to comprehend the geomorphology of the Yenisey river systems, focused on classification of sedimentary deposits. Landsat MSS and TM with five

Technical and economical problems of decommissioning nuclear power plants (NPP) in Russia

International Nuclear Information System (INIS)

Vaneev, M.

2001-01-01

under the direction of the senior lecturer Mikhail A. Skachek was estimated the basic economic parameters of decommissioning process NPP with the various types of reactors (WWR-440, WWR-1000, RBMK-1000) Now we are researching the subscription of disposal the nuclear waste in a total cost of decommissioning NPP. The estimation of expenses for decommissioning NPP was carried out with the helps of the program DECOST - adapted on faculty NPP MPEI to a Russian economy conditions of the transition period. Program 'DECOST' is developed for an estimation of expenses and payments for removal from operation of nuclear power installations in different conditions. (author)

RELAP5 / MOD3.2 analysis of INSC standard problem INSCSP - V4 : investigation of heat transfer for partly uncovered VVER-1000 core at the test facility KS (RRC K1)

International Nuclear Information System (INIS)

Tentner, A.; Ahrens, J. W.

2002-01-01

The RELAP5/MOD3.2 computer program has been used to analyze a series of tests investigating heat-transfer from a partly uncovered VVER-1000 core in the KS test facility at the Russian Research Center ''Kurchatov Institute'' (RRC-KI). The analysis documented represents VVER Standard Problem 4 in Joint Project 6, which is the investigation of Computer Code Validation for Transient Analysis of RBMK and VVER Reactors, between the United States and Russian International Nuclear Safety Centers. The experiment facility and data, RELAP5 nodalization, and results are shown for one of the six tests defined in Standard Problem 4. Only part of the data was analyzed due to our conclusion that the available experimental data is not sufficient to allow the modeling of the actual experiment sequence. The experiment initial conditions were reached through a series of transient processes, about which no quantitative information was available. This has required the modeling of an arbitrary computational transient, with the goal of reaching initial conditions similar to those observed during the experiment. The use of an arbitrary transient introduces many degrees of freedom in the analysis, i.e. initial computational values that influence the entire sequence of events, including the loop behavior during the experiment time window. Reasonable agreement between RELAP5 and the experiment data can be obtained by manipulating a number of initial computational values, including the liquid level in the fuel assembly model, the liquid level in the annular region, the quality of the saturated vapor in the voided loop regions, etc. Our study has focused on exploring the sensitivity of results to changes in these initial values which are not based on experimental information, but are selected with the goal of matching the experimentally observed behavior during the experiment time window. We have shown that changes in several initial arbitrary values can lead to similar changes in the

Licensing of spent nuclear fuel dry storage in Russia

International Nuclear Information System (INIS)

Kislov, A.I.; Kolesnikov, A.S.

1999-01-01

The Federal nuclear and radiation safety authority of Russia (Gosatomnadzor) being the state regulation body, organizes and carries out the state regulation and supervision for safety at handling, transport and storage of spent nuclear fuel. In Russia, the use of dry storage in casks will be the primary spent nuclear fuel storage option for the next twenty years. The cask for spent nuclear fuel must be applied for licensing by Gosatomnadzor for both storage and transportation. There are a number of regulations for transportation and storage of spent nuclear fuel in Russia. Up to now, there are no special regulations for dry storage of spent nuclear fuel. Such regulations will be prepared up to the end of 1998. Principally, it will be required that only type B(U)F, packages can be used for interim storage of spent nuclear fuel. Recently, there are two dual-purpose cask designs under consideration in Russia. One of them is the CONSTOR steel concrete cask, developed in Russia (NPO CKTI) under the leadership of GNB, Germany. The other cask design is the TUK-104 cask of KBSM, Russia. Both cask types were designed for spent nuclear RBMK fuel. The CONSTOR steel concrete cask was designed to be in full compliance with both Russian and IAEA regulations for transport of packages for radioactive material. The evaluation of the design criteria by Russian experts for the CONSTOR steel concrete cask project was performed at a first stage of licensing (1995 - 1997). The CONSTOR cask design has been assessed (strength analysis, thermal physics, nuclear physics and others) by different Russian experts. To show finally the compliance of the CONSTOR steel concrete cask with Russian and IAEA regulations, six drop tests have been performed with a 1:2 scale model manufactured in Russia. A test report was prepared. The test results have shown that the CONSTOR cask integrity is guaranteed under both transport and storage accident conditions. The final stage of the certification procedure

Radioactive preparation of defects in solids; Creation de defauts dans les solides au moyen de radioisotopes; Ispol'zovanie radioaktivnosti dlya obrazovaniya defektov v tverdykh telakh; Creacion de deiectos en los solidos mediante radioisotopos

Energy Technology Data Exchange (ETDEWEB)

Lambe, J [Physics Department, Ford Motor Company, Scientific Laboratory, Dearborn, MI (United States)

1962-01-15

moleculaire solide. Les experiences effectuees montrent que l'emploi de ce procede est tout a fait possible et devrait convenir plus particulierement a la production de radicaux libres dans les matieres organiques dont la radioresistance est assez forte. (author) [Spanish] El estudio de los defectos que aparecen en los cristales casi perfectos constituye uno de los capitulos mas interesantes de la fisica del estado solido. Por lo tanto, la creacion controlada de los defectos deseados representa un aspecto primordial de esos estudios. Dos de los metodos que mas se emplean para introducir defectos de esta clase consisten en agregar deliberadamente impurezas quimicas y en radioinducir danos. El autor ha estudiado la posililidad de valerse de la desintegracion radiactiva para crear defectos en los solidos, La tecnica se basa simplemente en obtener un material lo mas perfecto posible, pero en el que algunos de los atomos del cristal consistan en isotopos radiactivos del mismo elemento. Al desintegrarse estos atomos, el elemento descendiente origina un defecto. A fin de comprobar la posibilidad de aplicar esta tecnica, el autor preparo atomos de tritio incorporados en cristales de tritio molecular solido. Los experimentos indican que el metodo es viable, y que deberia poder aplicarse ante todo a la separacion de aquellos radicales libres contenidos en sustancias organicas que sean bastante resistentes a los danos ocasionados por las radiaciones. (author) [Russian] Odnoj iz naibolee interesnykh oblastej fiziki tverdogo sostoyaniya yavlyaetsya izuchenie defektov v pochti sovershennykh kristallakh. Takim obrazom, umyshlennoe obrazovanie takikh defektov yavlyaetsya vazhnym aspektom ehtikh issledovanij. Dlya proizvodstva takikh defektov byli shiroko rasprostraneny metody khimicheskikh dobavlenij i povrezhdeniya radiatsiej. V nastoyashchem trude rassmatrivayutsya nekotorye vozmozhnosti ispol'zovaniya radioaktivnogo raspada dlya obrazovaniya defektov v tverdykh telakh. Metod ehtot

Validity aspects in Chernobyl at twenty years of the accident

International Nuclear Information System (INIS)

Arredondo, C.

2006-01-01

For April 25, 1986 the annual stop of the unit 4 of the nuclear power plant of Chernobyl was programmed, in order to carry out maintenance tasks. This unit was equipped with a reactor of 1000 MW, type RBMK, developed in the former Soviet Union, this type of reactors uses graphite like moderator, the core is refrigerated with common water in boil, and the fuel is uranium enriched to 2%. Also it had been programmed to carry out, before stopping the operation of the power station, a test with one of the two turbogenerators, which would not affect to the reactor. However, the intrinsic characteristics of the design of the reactor and the fact that the operators disconnected intentionally several systems of security that had stopped the reactor automatically, caused a decontrolled increase of the power (a factor 1000 in 4 seconds), with the consequent fusion of the fuel and the generation of a shock wave, produced by the fast evaporation of the refrigeration water and caused by the interaction of the fuel fused with the same one. It broke the core in pieces and destroy the structure of the reactor building that was not resistant to the pressure. When being exposed to the air, the graphite of the moderator entered in combustion, while the radioactive material was dispersed in the environment. The radionuclides liberation was prolong during 10 days, and only it was stopped by means of the one poured from helicopters, of some 5000 tons of absorbent materials on the destroyed reactor, as long as tunnels were dug to carry out the cooling of the core with liquid nitrogen. Later on, the whole building of the damaged reactor was contained inside a concrete building. The immediate consequence of the accident was the death of 31 people, between operators of the nuclear power station and firemen. One of people died as consequence of the explosion and 30 died by cause of the irradiation, with dose of the order of 16 Gy. The liberated radioactive material was the entirety of the

Validity aspects in Chernobyl at twenty years of the accident; Aspectos vigentes en Chernobyl a veinte anos del accidente

Energy Technology Data Exchange (ETDEWEB)

Arredondo, C [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

2006-07-01

For April 25, 1986 the annual stop of the unit 4 of the nuclear power plant of Chernobyl was programmed, in order to carry out maintenance tasks. This unit was equipped with a reactor of 1000 MW, type RBMK, developed in the former Soviet Union, this type of reactors uses graphite like moderator, the core is refrigerated with common water in boil, and the fuel is uranium enriched to 2%. Also it had been programmed to carry out, before stopping the operation of the power station, a test with one of the two turbogenerators, which would not affect to the reactor. However, the intrinsic characteristics of the design of the reactor and the fact that the operators disconnected intentionally several systems of security that had stopped the reactor automatically, caused a decontrolled increase of the power (a factor 1000 in 4 seconds), with the consequent fusion of the fuel and the generation of a shock wave, produced by the fast evaporation of the refrigeration water and caused by the interaction of the fuel fused with the same one. It broke the core in pieces and destroy the structure of the reactor building that was not resistant to the pressure. When being exposed to the air, the graphite of the moderator entered in combustion, while the radioactive material was dispersed in the environment. The radionuclides liberation was prolong during 10 days, and only it was stopped by means of the one poured from helicopters, of some 5000 tons of absorbent materials on the destroyed reactor, as long as tunnels were dug to carry out the cooling of the core with liquid nitrogen. Later on, the whole building of the damaged reactor was contained inside a concrete building. The immediate consequence of the accident was the death of 31 people, between operators of the nuclear power station and firemen. One of people died as consequence of the explosion and 30 died by cause of the irradiation, with dose of the order of 16 Gy. The liberated radioactive material was the entirety of the

Fuel development for reactors of new generation in Ukraine

International Nuclear Information System (INIS)

Odeychuk, N.P.

2006-01-01

Full text: On the background of critical situation in traditional power engineering due to deficiency of organic fuel, physical and moral ageing of the of thermal power stations equipment and their harmful influence on the ecology of environment, the nuclear engineering works stably enough and, by keeping all safety measures, is the most non-polluting energy source. In Ukraine the atomic engineering became one of main sources of energy production and is the important factor of guarantee the power engineering independence of the state. The main center on development of the components of nuclear reactors active zones is the National scientific center K harkov institute of Physics and Technology . The significant place in institutes' investigations was occupied with works on creation the constructional materials and nuclear fuel for heavy water reactors E-circumflexS-150, OR-1000, OR-2000, light water reactors WWER-1000 and RBMK-1500, high-temperature gas cooled reactors ABTU and HTGR, gas reactors on fast neutrons BGR and BRGD, and also the reactor - converter ROMASHKA and other special reactors of special assignment. Radiation tests and post-irradiation research confirm intended material-study, technological and design decisions and fuel elements capacity work on the whole. Nevertheless, by the present conditions, it is necessary to pay special attention to development of the new, safe guaranteed nuclear energy sources. In Ukraine proceed works on research and development of new safe nuclear reactors: basing the underground nuclear thermal power stations; development the reactors with managed chain reaction of nucleus division in an active zone with the help of an external source of neutrons; power thermonuclear installations; high-temperature helium reactors which are especially actual now from the point of view of the hydrogen production; the advanced pressure water reactors, heavy water reactors. In the paper also discussed the state of works in Ukraine on fuel

Chernobyl and status of nuclear power development in the USSR

International Nuclear Information System (INIS)

Gagarinskii, A.Yu.

1989-01-01

The Chernobyl accident has seriously affected development of the USSR nuclear power program. But it has not eliminated the basic prerequisites for nuclear power development in the USSR which are: - resources and consumption territorial disproportions; - large share of oil and gas in electricity generation; - negative ecological aspects of coal plants; - high power industry development rate. At the same time it has aggravated the old problems and has given rise to some new-ones of which the most important are: - increased safety requirements; rise in costs; longer construction schedules; public opinion. On the whole for further safe development of nuclear power a detailed analysis of the Chernobyl accident is required, including studies of long-term accident consequences and measures of their mitigation and elimination. A necessary condition for NPP operation to be continued would also be development and rapid implementation of technical approaches which would permit to eliminate the design shortcomings in the RBMK NPPs both operating and those under construction. At the same time we have to ensure their competitiveness with other energy sources and possibility of expansion of their applications. The problem of public opinion should be emphasised. After the Chernobyl accident we have faced a social phenomenon which is quite new in this country. There is almost no site where the population was not opposed to NPP construction. For us these problems are especially difficult as we have had no experience of this kind of interactions with the public. We are planning and begin to realize a program basing on the current world experience. This program includes primarily a wide series of publications on the problems of nuclear energy its ecologic and economic advantages as compared with conventional and alternative energy sources,, using all cur-rent media. Centers of public information discussion clubs, exhibitions etc are being organized. In particular, our Institute has

Experience of Bohunice V-1 NPP

International Nuclear Information System (INIS)

Dobik, Dobroslav

2000-01-01

. Since 'European nuclear safety criteria' don't exist, all international recommendations are used in Bohunice, and our results were highly quoted at the following events: IAEA April - Meeting of contracting parties to Convention on Nuclear Safety; IAEA Final Report of the Programme on the safety of WWER and RBMK NPPs in May; IAEA Conference on Nuclear Safety in the Middle and East European Countries in June. There are two activities, both on a voluntary basis - WENRA's assessment of some applicants countries NPPs/Bohunice, Kozloduy, Ignalina and another activity trying to define safety criteria for NPPs large producers and operating utilities are involved in. As far as there is not a unique legislation in this area existing in EU, it cannot be an official item of the accession negotiation

Chernobyl and status of nuclear power development in the USSR

Energy Technology Data Exchange (ETDEWEB)

Gagarinskii, A Yu [I.V. Kurchatov Institute of Atomic Energy, Kurchatov Square, 123182 Moscow (Russian Federation)

1989-07-01

The Chernobyl accident has seriously affected development of the USSR nuclear power program. But it has not eliminated the basic prerequisites for nuclear power development in the USSR which are: - resources and consumption territorial disproportions; - large share of oil and gas in electricity generation; - negative ecological aspects of coal plants; - high power industry development rate. At the same time it has aggravated the old problems and has given rise to some new-ones of which the most important are: - increased safety requirements; rise in costs; longer construction schedules; public opinion. On the whole for further safe development of nuclear power a detailed analysis of the Chernobyl accident is required, including studies of long-term accident consequences and measures of their mitigation and elimination. A necessary condition for NPP operation to be continued would also be development and rapid implementation of technical approaches which would permit to eliminate the design shortcomings in the RBMK NPPs both operating and those under construction. At the same time we have to ensure their competitiveness with other energy sources and possibility of expansion of their applications. The problem of public opinion should be emphasised. After the Chernobyl accident we have faced a social phenomenon which is quite new in this country. There is almost no site where the population was not opposed to NPP construction. For us these problems are especially difficult as we have had no experience of this kind of interactions with the public. We are planning and begin to realize a program basing on the current world experience. This program includes primarily a wide series of publications on the problems of nuclear energy its ecologic and economic advantages as compared with conventional and alternative energy sources,, using all cur-rent media. Centers of public information discussion clubs, exhibitions etc are being organized. In particular, our Institute has

Neutron Tests at the Start-Up of EDF1; Les essais neutroniques au demarrage du reacteur EDF1; Nejtronnye izmereniya pri puske reaktora EDF1; Ensayos neutronicos efectuados durante la puesta en marcha del reactor EDF1

Energy Technology Data Exchange (ETDEWEB)

Teste du Bailler, A. [Centre d' Etudes Nucleaires de Saclay (France); Janin, R. [Electricite de France, Paris (France)

1963-10-15

ehksperimental'nykh metodov, razrabotannykh na reaktorakh v Markule, byla provedena vo vremya puska reaktora EDF.1. Izmereniya kasalis' glavnym obrazom ehffektivnosti upravlyayushchikh sterzhnej pri razlichnykh p ogruzh eniyakh. Opredelyaetsya skhema pod''ema s terzhnej, kotoraya pozvolyaet poluchat' polnuyu moshchnost' prk soblyudenii opredelennykh ogranichenij v temperature obolochek i gaza. Parallel'no byli provedeny izmereniya potoka pri razlichnykh polozheniyakh kompensiruyushchikh sterzhnej i pri razlichnykh zagruzkakh poglotitelej v opredelennykh kanalakh, v zavisimosti ot predvaritel'nykh raschetov v dvukhraehmernom izmerenii. Ehti izmereniya byli polucheny putem aktivatsii tochechnykh detektorov s primeneniem obychnogo metoda otravleniya vozdukha. Pri nekotorykh urovnyakh temperatury (do 140{sup o}C) byli provedeny izmereniya koehffitsientov reaktivnosti i ehffektivnosti reguliruyushchikh sterzhnej. V to zhe vremya putem aktivatsii sootvetstvuyushchikh detektorov (uran, plutonij, lyutetsij, marganets, indij, zoloto) byli provedeny izmereniya koehffitsientov spektra. Dlya izmereniya ehffektivnosti nekotorykh kompensiruyushchikh sterzhnej byl ispol'zovan ostsillyatsionnyj metod. Nakonets, s tselyakh izucheniya zashchity i povrezhdeniya grafita byli provedeny izmereniya potoka bystrykh nejtronov. (author)

Development and assessment of ASTEC code for severe accident simulation

International Nuclear Information System (INIS)

Van Dorsselaere, J.P.; Pignet, S.; Seropian, C.; Montanelli, T.; Giordano, P.; Jacq, F.; Schwinges, B.

2005-01-01

the initiating event up to the aerosol and iodine behaviour in the containment, which shows the code capabilities for plant applications. It describes also in details the results of the CESAR-DIVA application to the phases 1 and 2 of the TMI-2 accident. Main efforts in 2004-05 will focus on consolidation of the version through multiplication of calculations of different reactor sequences. In parallel, progress will be made on one hand on modeling of reflooding of intact and degraded cores and on the other hand on models closely linked to preparation or interpretation of experimental programs (CHIP, EPICUR, ARTIST, ARTEMIS...). Beyond, long-term improvements will concern the remaining safety key-issues such as corium coolability, iodine species behaviour in the primary circuit, long-term MCCI. Specifications to extend the application scope to other reactors such as BWR, CANDU and RBMK will be prepared in 2004-05 in SARNET frame for implementation of new models in the future ASTEC V2 version. (authors)

Major changes in the world's nuclear power at the beginning of the new century

International Nuclear Information System (INIS)

Dumitrache, Ion

2002-01-01

In the last decade of the 20th century the world nuclear power recorded some characteristic trends among which one can mention the following: - Almost total absence of investments in new NPPs in the industrialized countries except Japan and South Korea; - Policy of some governments to decrease the nuclear power sector in their countries up to a complete stop of electricity production in a foreseeable future (as in case of Sweden, Germany, Nederland and Belgium); - Projections indicating a steady decline of nuclear share in the national power production as for instance in USA, Germany, Great Britain, and other industrialized countries; - pressures upon countries late owners of soviet type NPPs in order to shut down completely the RBMK and WWER reactors; - a drastic reduction of the funds afforded for research dedicated to fission reactors of new concept, except Japan and South Korea; - almost negligible effects of the Kyoto protocol upon nuclear power, hopes being directed towards renewable energy sources. After second half of the year 1998 modest signals of future changes in the energy policy occurred. The US government admitted on basis of performance assessments and projections that the important role of nuclear power in US will be extended still for long after the years 2020-2030. Consequently, research concerning the future demand for fission based power began be financed. Gradually the countries of EU and Canada modified also their official position towards the role of nuclear fission in ensuring the electric energy needs of the future. The beginning of the new century was marked by a significant acceleration of changes of opinions in favor of nuclear power. Japan and South Korea stated that at least in the first half of the 21th century the fission NPP's will play a major role. Russia promoted new WWER reactor types of safety standards equivalent or higher than the western ones. Also China and India launched ambitious plans for building new NPPs. These new

The Enrico Fermi Atomic Power Plant; La centrais nucleaire Enrico Fermi; Atomnaya ehlektrostantsiya im Ehnriko Fermi.; La central nucleoelectrica Enrico Fermi

Energy Technology Data Exchange (ETDEWEB)

Hartwell, R. W. [Power Reactor Development Company, Detroit, MI (United States)

1963-10-15

nejtronakh moshchnost'yu 100 mgvt (ehl.) bylo v osnovnom zakoncheno v dekabre 1961 goda. V techenie poslednikh 16 mesyatsev provodilis' shirokie ispytaniya sistem i komponentov. Ehta predpuskovaya programma ispytaniya okazalas' ochen' poleznoj dlya proverki konstruktsii i dlya opredeleniya neobkhodimykh izmenenij. Vse voznikshie problemy okazalis' razreshimymi. V doklade kratko osveshchayutsya naibolee vatnye izmeneniya. Grafitovaya zashchita. V dekabre 1960 goda pervyj kontur byl zapolnen natriem i byli nachaty shirokie ispytaniya. Kogda byl snova otkryt zashchitnyj bak pervogo kontura posle ispytaniya pervogo kontura pri temperature 1000{sup o}F, bylo obnaruzheno, chto bol'shaya chast' grafitovogo bloka zashchity, ustanovlennogo vokrug reaktora, povrezhdena. Vysokotemperaturnye bloki, nasyshchennye borom, uvelichilis' v ob''eme i poteryali prochnost'. Provedennyj tshchatel'nyj analiz pokazal, chto grafitovaya svyaz' povrezhdena. Bylo resheno zamenit' ves' grafit, ispol'zovat' karbid bora v kachestve soedineniya, soderzhashchego bor, ustanovit' blok s pomoshch'yu mekhanicheskogo krepleniya i dovesti vlazhnost' do minimuma. Izmeneniya v korpuse reaktora. Byl proveden remont i vneseny izmeneniya v konstruktsiyu dlya ustraneniya prichiny zaedaniya sborok, dlya likvidatsii povrezhdeniya, kotoroe yavilos' rezul'tatom ehtogo, i dlya ustraneniya v dal'nejshem neispravnosti v peremeshchayushchem mekhanizme upravleniya. Pered remontom byl udalen peremeshchayushchij mekhanizm upravleniya, korpus reaktora byl osushen ot natriya. Posle ehtogo podgotovlennyj peroonvl v spetsial'nykh zashchitnykh kostyumakh byl dopushchen vnutr' korpusa reaktora. Vkhod v rabochuyu zonu obespechivalsya s pomoshch'yu spetsial'nogo- vozdushnogo shl'za, poskol'ku v korpuse reaktora imelsya argon. Izmeneniya parogeneratora. Vo vremya gidrostaticheskikh ispytanij parogeneratora No. 2 byli obnaruzheny techi v neskol'kikh trubkakh. Prichinoj neispravnosti trubok bylo ikh rastreskivanie v rezul'tate korrozii ot

11. international topical meeting on nuclear reactor thermal-hydraulics (NURETH-11)

International Nuclear Information System (INIS)

Lemonnier, H.

2005-01-01

; aerosol transport, deposition and re-entrainment; steam generators thermal-hydraulics; system codes development and assessment; uncertainties analysis; diffuse interface methods and interface tracking methods; C - severe accidents and fires: molten core natural convection and physico-chemical phenomena, modeling and experiments; fuel coolant interaction, modeling and experiments; debris bed cooling; combustion and fires, modeling and experiments; molten corium concrete interaction; D - advanced code developments: fast transient modelling and experiments; multidimensional single-phase or two-phase flow and heat transfer modeling; neutronics and thermal-hydraulics coupling; fluid and structures mechanical interactions; coupled thermal-hydraulics of fluids and structures; thermal-hydraulic dependent corrosion and ablation; E - operation and safety of existing reactors: instabilities and nonlinear dynamics; NPP transients and accidents analysis; RBMK and VVER safety analysis, including the OECD benchmark; F - experimental thermal-hydraulics: boiling heat transfer; CHF and post-CHF heat transfer; condensation heat transfer; integral testing; vibrations, wear and thermal fatigue phenomena; fuel design and performance; G - advanced reactors thermal-hydraulics (gen IV, INPRO, fusion, hydrogen production): accelerator driven reactors; advanced pressurized water reactors thermal-hydraulics; gas cooled fast reactors; gas cooled high temperature reactors; lead and lead-bismuth cooled reactors; future and existing sodium cooled reactors; molten salt reactors; H - waste management thermal-hydraulics: thermal-hydraulics problems related to waste processing and storage; I - thermal-hydraulics of non electricity generating nuclear equipment: sono-fusion (cavitation induced bubble fusion; hydrogen producing nuclear reactors

Fuel Cycle of Reactor SVBR-100

Energy Technology Data Exchange (ETDEWEB)

Zrodnikov, A.V.; Toshinsky, G.I.; Komlev, O.G. [FSUE State Scientific Center Institute for Physics and Power Engineering, 1, Bondarenko sq., Obninsk, Kaluga rg., 249033 (Russian Federation)

2009-06-15

the cheap resources of natural uranium will be expired till the middle of the century that will cause increase in the uranium cost, the period of FRs operating in the open NFC have to be shortened to the maximal possible extent. Changeover to the closed NFC will be cheaper if the plutonium extracted from the own SNF of uranium loads is used to form the first MOX fuel loads. When uranium oxide fuel is used for operation, comparatively high breeding ratio (BR) ({approx}0,84) of reactor SVBR 100 provides by the end of lifetime the large content of plutonium in the SNF, which can be used in next fuel lifetimes in case of organizing the closed NFC. Moreover, in the own SNF of starting loads made of oxide uranium fuel there are much of unburned uranium 235, which is also expedient to use for formation of the load for the next lifetime. That approach to organization of fuel cycles with complete reprocessing of the own SNF will considerably reduce integral consumption of natural uranium and thus provide competitiveness of NPPs based on RIs of the SVBR 100 type with NPPs based on RIs with TRs. The report demonstrates that in the closed NFC instead of waste pile uranium the TR SNF (of both WWER and RBMK) can be used (utilized) without partitioning uranium, plutonium, minor actinides and fission products (FP) similarly to the DUPIC technology for reactors Candu. (authors)

Thermal Shock Tests on UO{sub 2} Small Spheres; Essais de choc thermique sur des elements spheriques de UO{sub 2}; Ispytaniya nebol'shikh sharikov iz UO{sub 2} teplovykh udarom; Ensayo de pequenas esferas de UO{sub 2} por choque.termico

Energy Technology Data Exchange (ETDEWEB)

Perona, G.; Brutto, E.; Galbusera, U.; Palladino, G.; Sesini, R. [Centro Informazioni Studi Esperienze, Milan (Italy)

1963-11-15

exponen los resultados obtenidos. La aplicacion de este metodo presenta, al parecer, considerable interes, sobre todo en lo que concierne a las investigaciones encaminadas a mejorar las caracteristicas de las esferas de UO{sub 2} por medio de aditivos. En e fecto, permite verificar el efecto global con una sola medicion. (author) [Russian] Ispol'zuya malye pariki iz UO{sub 2} v kachestve yadernogo topliva v reaktore, gde oni nakhodyatsya v soprikosnovenii s teplonositelem, neobkhodimo znat' maksimal'nuyu pri rabochem rezhime reaktora velichinu termicheskikh napryazhenij, kotorye mogut vyderzhivat' bez povrezhdeniya ehti shariki. Esli izvestny fizicheskie svojstva materiala, to mozhno rasschitat' ehti napryazheniya pri rabochem rezhime. Odnako vvidu mnogochislennosti podlezhashchikh uchetu faktorov i neizbezhnoj doli neopredelennosti kazhdogo iz nikh predstavlyaetsya predpochtitel'nym provesti neposredstvennye ispytaniya ehtikh sharikov, podvergnuv ikh tem zhe napryazheniyam, kakie oni ispytyvayut v reaktore. V nastoyashchej rabote byl izuchen metod teplovogo udara v primenenii k malym sharikam i ukazyvayutsya usloviya, pri kotorykh ehtot metod pozvolyaet proizvesti napryazheniya, neposredstvenno sravnimye s temi, kotorye sushchestvuyut v reaktore. V sluchae malykh sharikov zatrudnenie zaklyuchaetsya v osushchestvlenii okhlazhdeniya, pozvolyayushchego dostigat' ochen' bol'shikh znachenij koehffitsienta poverkhnostnoj teploperedachi. Opisyvayutsya ehksperimental'nye metody i soobshchayutsya poluchennye rezul'taty. Primenenie ehtogo' metoda, po-vidimomu., predstavlyaet bol'shoj interes, v osobennosti v oblasti tekhnologicheskikh izyskanij s tsel'yu uluchsheniya svojstv malykh sharikov iz UO{sub 2} putem vklyucheniya dobavochnykh komponentov. Fakticheski,ehtot metod daet vozmozhnost' pri pomoshchi odnogo tol'ko izmereniya kontrolirovat' izuchaemoe vozdejstvie. (author)

Cooperation in Nuclear Waste Management, Radiation Protection, Emergency Preparedness, Reactor Safety and Nuclear Non-Proliferation in Eastern Europe

International Nuclear Information System (INIS)

Dassen, Lars van; Delalic, Zlatan; Ekblad, Christer; Keyser, Peter; Turner, Roland; Rosengaard, Ulf; German, Olga; Grapengiesser, Sten; Andersson, Sarmite; Sandberg, Viviana; Olsson, Kjell; Stenberg, Tor

2009-10-01

The Swedish Radiation Safety Authority (SSM) is trusted with the task of implementing Sweden's bilateral assistance to Russia, Ukraine, Georgia, Belarus and Armenia in the fields of reactor safety, nuclear waste management, nuclear non-proliferation as well as radiation protection and emergency preparedness. In these fields, SSM also participates in various projects financed by the European Union. The purpose of this project-oriented report is to provide the Swedish Government and other funding agencies as well as other interested audiences in Sweden and abroad with an encompassing understanding of our work and in particular the work performed during 2008. the activities are divided into four subfields: Nuclear waste management; Reactor safety; Radiation safety and emergency preparedness; and, Nuclear non-proliferation. SSM implements projects in the field of spent nuclear fuel and radioactive waste management in Russia. The problems in this field also exist in other countries, yet the concentration of nuclear and radioactive materials are nowhere higher than in north-west Russia. And given the fact that most of these materials stem from the Cold War era and remain stored under conditions that vary from 'possibly acceptable' to 'wildly appalling' it is obvious that Sweden's first priority in the field of managing nuclear spent fuel and radioactive waste lies in this part of Russia. The prioritisation and selection of projects in reactor safety are established following thorough discussions with the partners in Russia and Ukraine. For specific guidance on safety and recommended safety improvements at RBMK and VVER reactors, SSM relies on analyses and handbooks established by the IAEA in the 1990s. In 2008, there were 16 projects in reactor safety. SSM implements a large number of projects in the field of radiation protection and emergency preparedness. The activities are at a first glance at some distance from the activities covered and foreseen by for instance the

Cooperation in Nuclear Waste Management, Radiation Protection, Emergency Preparedness, Reactor Safety and Nuclear Non-Proliferation in Eastern Europe

Energy Technology Data Exchange (ETDEWEB)

Dassen, Lars van; Delalic, Zlatan; Ekblad, Christer; Keyser, Peter; Turner, Roland; Rosengaard, Ulf; German, Olga; Grapengiesser, Sten; Andersson, Sarmite; Sandberg, Viviana; Olsson, Kjell; Stenberg, Tor

2009-10-15

The Swedish Radiation Safety Authority (SSM) is trusted with the task of implementing Sweden's bilateral assistance to Russia, Ukraine, Georgia, Belarus and Armenia in the fields of reactor safety, nuclear waste management, nuclear non-proliferation as well as radiation protection and emergency preparedness. In these fields, SSM also participates in various projects financed by the European Union. The purpose of this project-oriented report is to provide the Swedish Government and other funding agencies as well as other interested audiences in Sweden and abroad with an encompassing understanding of our work and in particular the work performed during 2008. the activities are divided into four subfields: Nuclear waste management; Reactor safety; Radiation safety and emergency preparedness; and, Nuclear non-proliferation. SSM implements projects in the field of spent nuclear fuel and radioactive waste management in Russia. The problems in this field also exist in other countries, yet the concentration of nuclear and radioactive materials are nowhere higher than in north-west Russia. And given the fact that most of these materials stem from the Cold War era and remain stored under conditions that vary from 'possibly acceptable' to 'wildly appalling' it is obvious that Sweden's first priority in the field of managing nuclear spent fuel and radioactive waste lies in this part of Russia. The prioritisation and selection of projects in reactor safety are established following thorough discussions with the partners in Russia and Ukraine. For specific guidance on safety and recommended safety improvements at RBMK and VVER reactors, SSM relies on analyses and handbooks established by the IAEA in the 1990s. In 2008, there were 16 projects in reactor safety. SSM implements a large number of projects in the field of radiation protection and emergency preparedness. The activities are at a first glance at some distance from the activities covered and

VSOP, Neutron Spectra, 2-D Flux Synthesis, Fuel Management, Thermohydraulics Calculation

International Nuclear Information System (INIS)

Teuchert, E.; Haas, K.A.

1995-01-01

1 - Description of problem or function: VSOP (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories. It comprises neutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D and 3-D diffusion calculation, depletion and shut-down features, in- core and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (steady state and transient). Various techniques have been employed to accelerate the iterative processes and to optimize the internal data transfer. The code system has been used extensively for comparison studies of thermal reactors, their fuel cycles, thermal transients, and safety assessment. Besides its use in research and development work for the Gas Cooled High Temperature Reactor, the system has been applied successfully to Light Water and Heavy Water Reactors, MAGNOX, and RBMK. 2 - Method of solution: The nuclear data for 184 isotopes are contained in two libraries. Fast and epithermal data in a 68 group GAM-I structure have been prepared mainly from ENDF/B-V and JEF-1. Resonance cross section data are given as input. Thermal data in a 30 group THERMOS structure have been collapsed from a 96 group THERMALIZATION (GATHER) library by a relevant neutron energy spectrum generated by the THERMALIZATION code. Graphite scattering matrices are based on the Young phonon spectrum in graphite. The neutron spectrum is calculated by a combination of the GAM and THERMOS codes. They can simultaneously be employed for many core regions differing in temperature, burnup, and fuel element lay-out. The thermal cell code THERMOS has been extended to treat the grain structure of the coated particles inside the fuel elements, and the epithermal GAM code uses modified cross sections for the resonance absorbers prepared from double heterogeneous ZUT-DGL calculations. The diffusion module of the code is CITATION with 2 - 8 energy groups. It provides the neutron

Peculiarities of physical protection assurance of the nuclear materials at nuclear installation decommissioning stage

International Nuclear Information System (INIS)

Pinchuk, M.G.

2001-01-01

On December 15, 2000 Unit 3 of Chernobyl NPP, which is the last one in Ukraine having RBMK-type reactor, was permanently shutdown before the end of its lifetime. A number of projects related to establishing infrastructure for the plant decommissioning are being implemented in compliance with the Ukraine's commitments. Decommissioning stage includes activities on fuel unloading from the cores of Unit I and Unit 3, fuel cooling in the ponds followed by the fuel transportation to the spent fuel dry storage facility (currently under construction) for its safe long-term storage. Special facilities are being created for liquid and solid radioactive waste treatment. Besides, it is planned to implement a number of projects to convert Shelter Object in environmentally safe structure. Physical protection work being an essential part of the nuclear material management is organized in line with the recommendations of the IAEA, and the Laws of Ukraine 'On Nuclear Energy Utilization and Radiation Safety', 'On Physical Protection of Nuclear Installations and Materials', 'Regulations on Physical Protection of Nuclear Materials and Installations', other codes and standards. While organizing physical protection on ChNPP decommissioning stage we have to deal with some specific features, namely: Significant amount of fuel assemblies, which are continuously transferred between various storage and operation facilities; Big amount of odd nuclear material at Shelter Object; 'Theft of new fuel fragments from the central hall of the Shelter Object in 1995 with the intention of their further sale. The thieves were detained and sentenced. The stolen material was withdrawn, that prevented its possible proliferation and illicit trafficking. At present physical protection of ChNPP does not fully satisfy the needs of the decommissioning stage and Ukraine's commitments on non-admission of illicit trafficking. Work is carried out, aimed to improve nuclear material physical protection, whose main

Diagnosis of Intracranial Lesions by Gamma-Encephalography using Human Serum Albumin Labelled with Iodine-131; Diagnostic des lesions intracraniennes par la gamma-encephalographie a l'aide de la serumalbumine humaine marquee a l'iode 131; Diagnoz vnutricherepnykh povrezhdenij putem gamma-ehntsefalografii pri pomoshchi mechenoj iodom-131 albuminovoj syvorotki cheloveka; Diagnostico de las lesiones intracraneanas por gammaencefalografia mediante sero- albumina humana marcada con yodo-131

Energy Technology Data Exchange (ETDEWEB)

Planiol, Therese [Institut National d' Hygiene, Paris (France)

1959-07-01

sto pyat'desyat sopovozhdalis' ochagom dlitel'noj povyshennoj gamma-aktivnosti. Opukholi, kotorye ne poddalis' obsledovaniyu, yavlyayutsya napolovinu opukholyami polushariya mozga, a napolovinu - opukholyami zadnej polosti i shishkovidnoj zhelezy. Byli obnaruzheny gnojniki, krovyanye opukholi, mestnye vetvistye anevrizmy. Povrezhdenie mozgovoj tkani i krovenosnykh sosudov vyzvalo v odnoj treti sluchaev anomalii: polovina ehtikh anormal'nykh grafikov imeet formu, kharakternuyu dlya razmyagcheniya godovnogo mozga (ili tromboz), nalichie kotorogo mozhet byt' takim obrazom ustanovleno; drugie zhe napominayut grafiki opukholej s toj edinstvennoj raznitsej, chto anomalii ischezayut cherez neskol'ko nedel' v sluchayakh povrezhdeniya krovenosnykh sosudov; dve treti ehtikh povrezhdenij dayut otritsatel'nuyu gamma- ehntsefalogrammu. Rasshirenie arterij, bezopukhol'ny e ochagi, vyzyvayushchie pripadki ehpillepsii i razlichnye zabolevaniya chisto nevrologicheskog o kharaktera dali v 96% sluchaev normal'nye rezul'taty. Obsledovanie mozga pri pomoshchi radioaktivnog o albumina mozhet dat' tsennye svedeniya ne tol'ko o nalichii i tochnom mestonakhozhdeni i nevrokhirurgicheskikh povreyasdenij, no i v otnoshenii ikh kharaktera. V chastnosti, ono daet bol'shie vozmozhnosti dlya opredeleniya opukholej mozgovykh obolochek, gliom i metastazov. Ehti vozmozhnosti, sovmestno so svedeniyami, kasayushchimisya polozhitel'nogo i topograficheskogo diagnoza, pridayut dannomu vidu obsledovaniya osobyj interes po sravneniyu s drugimi metodami nevrologicheskog o diagnoza. Krome togo gamma-ehntsefalografi ya obeshchaet stat' odnim iz naibolee tochnykh sredstv dlya obnaruzheniya retsidivov i dlya nablyudeniya za khodom lecheniya obychnymi metodami ili radioterapii. (author)

The practical experience with assistance programs: view from a non-nuclear weapons-state with a significant nuclear infrastructure

International Nuclear Information System (INIS)

Chetvergov, S.

2002-01-01

Full text: In the area of nuclear energy utilization, the Republic of Kazakhstan follows international legislation standards. Since December 13, 1993 Kazakhstan has been a participant of the Nuclear Weapon Non-proliferation Treaty and does not have nuclear weapons. In the framework of this treaty, Kazakhstan provides measures to ensure the non-proliferation regime. The republic signed the agreement with IAEA on the guarantees, that were ratified by a presidential decree in 1995. Nuclear objects in Kazakhstan have the following characteristics: Ulba Metallurgical Plant, located in the eastern area of Kazakhstan, manufactures fuel pellets of uranium dioxide for heat release assemblies of RBMK and LWR reactor types. These pellets have an enrichment of U235 1.6-4.4 %. Ulba also has a radioactive waste disposal storage site; a power plant for heat and power supply, and water desalination is based at the BN-350 fast breeder reactor. This reactor is located in Aktau city on the Caspian Sea. Since April 1999, the reactor has been in the process of being decommissioned. There is a lot of spent fuel with highly radioactive and toxic weapon plutonium there. There are also research reactors of National Nuclear Centre, located in the north-eastern area of Kazakhstan, near Semipalatinsk city. These research reactors have nuclear materials of the first category, which are attractive to criminal groups: IVG.1 M - light-water heterogeneous reactor of vessel type on thermal neutrons, with light water moderator and coolant, maximum power is 35 MW; IGR - impulse homogeneous graphite reactor on thermal neutrons, with graphite reflector; RA - high temperature gas cooled reactor on thermal neutrons, 0.5 MW power. There is also a research reactor site near Almaty city, with LWR-K - light-water reactor, with 10 MW power, uses highly enriched uranium (up to 36 % of U-235). The following activity was accomplished in the framework of physical security modernization for nuclear objects and

Dispersions of Oxides in Oxide Matrices as High-Temperature Reactor Fuels; Dispersions d'oxyde dans des matrices d'oxyde, utilisees comme combustibles dans des reacteurs a haute temperature; Dispersiya okisej v okislovykh matritsakh v kachestve topliva dlya vysokotemperaturnogo reaktora; Empleo de dispersiones de oxidos en matrices de oxidos, como combustibles para reactores de elevada temperatura

Energy Technology Data Exchange (ETDEWEB)

Williams, J. [Atomic Energy Research Establishment, Harwell (United Kingdom)

1963-11-15

incorporarlas en matrices de elevada densidad. Los trabajos sobre metodos de elaboracion en escala experimental se hallan bastante avanzados. (author) [Russian] Daetsya obzor vozmozhnosti primeneniya dispersij PuO{sub 2},UO{sub 2}, ThO{sub 2} v matritsakh iz BeO, Al{sub 2}O{sub 3}, MgO i SiO{sub 2} s tochki zreniya sokhraneniya tselostnosti takogo topliva i sposobov ego izgotovleniya. Neizmennost' razmerov i sposobnost' uderzhaniya produktov deleniya yavlyayutsya naibolee vazhnymi svojstvami s tochki zreniya sokhraneniya tselostnosti topliva. Sovmestimost' sostavnykh ehlementov topliva drug s drugom i s teplonositelem okazyvayut vliyanie na neizmennost' razmerov, no v ehtom otnoshenii okislovye vidy topliva obladayut znachitel'nymi preimushchestvami. Na izmenenie razmerov pod dejstviem oblucheniya okazyvayut vliyanie: povrezhdeniya matritsy pod dejstviem nejtronov i oskolkov deleniya; radiatsionnoe povrezhdenie fazy delyashchikhsya veshchestv vosproizvodyashchikh materialov i nakoplenie produktov deleniya v gazoobraznom sostoyanii. Termicheskie napryazheniya takzhe mogut vyzyvat' izmeneniya formy. Odnako svedeniya o mekhanizme relaksatsii napryazhenij slishkom ogranicheny, chtoby mozhno bylo dat' kakuyu-libo priemlimuyu teoreticheskuyu otsenku povedeniyu topliva. Issledovaniya vykhoda produktov deleniya kak v sluchae legkogo oblucheniya, tak i pri sil'nom vygoranii okisej delyashchikhsya veshchestv/vosproizvodyashchikh materialov ogranichivalis' glavnym obrazom gazoobraznymi produktami deleniya, preimushchestvenno ksenonom. Dannye o vykhode drugikh produktov deleniya, a takzhe svedeniya o prokhozhdenii produktov deleniya voobshche cherez vozmozhnye materialy dlya matrits ochen' ogranicheny. Issledovaniya pronitsaemosti chistykh spekshikhsya okisej pokazyvayut, chto dlya ustraneniya otkrytoj poristosti takikh matrits potrebovalos' by dostizhenie plotnostej, dokhodyashchikh po men'shej mere do 95, a to i do 98% ot teoreticheski osushchestvimoj. Dlya izgotovleniya chastits

Preliminary Note on the Use of Radioisotopes to Study Some Cotton-Plant Pests in Africa; Note preliminaire sur l 'utilisation des radioisotopes dans l 'etude des parasites du cotonnier en afrique; Predvaritel'nye zamechaniya o primenenii radioizotopov dlya izucheniya nekotorykh vreditelej khlopchatnika v afrike.; Nota preliminar sobre el empleo de radioisotopos en el estudio de parasitos del algodonero en africa

Energy Technology Data Exchange (ETDEWEB)

Delattre, R. [Institut de Recherches du Coton et Textiles, Paris (France)

1963-09-15

el follaje de algodoneros jovenes. Hay orugas filofagas (Silepta derogata, Prodenia litura) que no retienen la radiactividad, pero las que se alimentan de organos fructiferos (Heliothis armigera, Earias insulana, Diparopsis watersi, etc.) se pueden detectar facilmente tres meses despues de la pulverizacion. En el segundo ensayo se aplicaron {sup 32}P y {sup 35}S a algodoneros viejos, inmediatamente antes de comenzar el periodo de diapausa natural de Diparopsis. Los resultados obtenidos hacen pensar que no sera muy dificil distinguir, entre las poblaciones de crisalidas recogidas del suelo, las que se han nutrido en el algodonero antes de ser marcado, es decir, las que habran sufrido una diapausa. (author) [Russian] Gusenitsa Diparopsis watersi (Roth) nanosit vred khlopchatniku, unichtozhaya tsvet i korobochku. Ehta,prakticheski,monofagovaya sovka provodit mezhsezonnyj period libo v zemle v vide kukolki v sostoyanii diapauzy, nachinaya s 10 noyabrya, libo v vide novykh (polivol'tinnykh) pokolenij na ne vykopannykh iz zemli rasteniyakh. Laboratornye issledovaniya pozvolili vyyasnit' osnovnye mekhanizmy vozniknoveniya i prekrashcheniya diapauzy. V prirode intensivnoe razmnozhenie pri razvedenii kul'tur proiskhodit u odnikh osobej s odnim pokoleniem v god, vyplansivayushchikhsya k aprelyu-mayu, u vtorykh - poyavlenie pokolenij bez diapauzy; otnositel'nuyu rol' ehtikh dvukh razlichnykh vetvej osobej sledovalo by tochno opredelit' dlya vybora metoda bor'by: vykorchevka khlopchatnika, unichtozhenie kukolok v sostoyanii diapauzy v zemle, sokrashchenie srokov sel'skokhozyajstvennoj kampanii i t.d. Mechenie radioizotopami gusenits, kotorye v kriticheskij period dostigayut zrelosti, dolzhno suzit' ehtu problemu v prakticheskom plane. V Tikeme (Respublike Chad) byli provedeny predvaritel'nye opyty dlya opredeleniya prostykh metodov mecheniya. V pervom opyte byl ispol'zovan P{sup 32} pri pryamom vodnom obryzgivanii listvy molodogo khlopchatnika. Nesmotrya na dozhd', pogloshchenie

Establishing design basis threats for the physical protection of nuclear materials and facilities

International Nuclear Information System (INIS)

Chetvergov, S.

2001-01-01

In the area of nuclear energy utilization, the Republic of Kazakhstan follows the standards of international legislation and is a participant of the Nuclear Weapons Non-proliferation Treaty as a country that does not have nuclear weapons. In the framework of this treaty, Kazakhstan provides for the measures to ensure the regime of nonproliferation. The Republic signed the Agreement with the IAEA on the guarantee that was ratified by the Presidential Decree in 1995. Now the Government of the RK is considering the Convention on Physical Protection of Nuclear Materials. Kazakhstan legislation in the area of nuclear energy utilization is represented by a set of laws: the main of them is the Law of the Republic of Kazakhstan 'On the utilization of atomic energy', dated April 14, 1997. According to the Law, the issues of physical protection are regulated by interdepartmental guideline documents. Nuclear science and industry of RK include: Enterprises on uranium mining and processing; Ulba metallurgical plant, manufacturing fuel pellets of uranium dioxide for heat release assemblies of RBMK and WWR reactor types, with the enrichment on U235 1.6-4.4%; Power plant in Aktau for heat and power supply and water desalination, based on fast breeder reactor BN-350; Research reactors of National Nuclear Center: WWR-K - water-water reactor, with 10 MW power, uses highly enriched uranium (up to 36% of U-235); IVG.1M - water-water heterogeneous reactor of vessel type on thermal neutrons, maximum power is 35 MW; IGR - impulse homogeneous graphite reactor on thermal neutrons, with graphite reflector; RA - high temperature gas cooled reactor on thermal neutrons, 0.5 MW power. The establishment of design basis threats for nuclear objects in the Republic of Kazakhstan is an urgent problem because of the developing military-political situation in the region. It is necessary to specify important elements affecting the specific features of the design basis threat: military operations of

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

Energy Technology Data Exchange (ETDEWEB)

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

1990-03-01

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

Some aspects of RF radiation safety guidelines on urgent protective measures in case of radiation emergency at NPP

International Nuclear Information System (INIS)

Bulgakov, V.G.; Klepikova, N.V.; Shershakov, V.M.; Ivanov, E.A.

2003-01-01

VR or RBMK reactors using the forces of NPP before radiation safety experts become involved. This method was developed with the purpose to 1.) Improve radically fast response in terms of preparation of recommendations an intervention measures in the first hours into an accident. 2.) Reduce stress for the NPP administration and liberate them for efficient management of an accident The task of preparing recommendations an emergency measures in the first hours of an accident by NPP forces can be efficiently executed based on preliminary study of a set of initial data, which am key for radiation consequences of an accident and developing a set of recommendations for each set. The presentation describes a method for selection of a set of initial data and identifying corresponding areas in which emergency measures would be reasonable to implement. The considered key factors influencing the radiation situation in the early phase of an accident at NPP include: the activity of 131 I release to the atmosphere (contribution of other iodine radioisotopes can be allowed for by using a correction factor), physical and chemical forms of iodine in the release, a release duration, time of the day during which the release is dispersed in the atmosphere, the population age group, the atmospheric boundary layer stability category, wind speed and the underlying surface roughness length. In selecting a set of initial data preference was given to standard information about meteorological conditions of pollutant dispersion in the atmosphere available to meteorological stations and dosimetry services of NPPs. Analysis was performed for joint realization of atmospheric stability category and classification of wind speed at vane's height. Results of analysis are presented. tab. 1 (author)

Implementing and measuring safety goals and safety culture. 1. Lessons to Learn from Three Mile Island, Chernobyl, and Tokaimura and the New Era of the European Nuclear Industry

International Nuclear Information System (INIS)

Reisch, Frigyes

2001-01-01

profile. The operators were not aware of this because of lack of instrumentation. There had been several precursors, i.e., incidents at other RBMK plants, that had not been properly evaluated, and the lessons had not been applied. After the accident, all the blame was placed at the grass-roots level. Nevertheless, major changes were made in the design. However, there is much left to do. Tokaimura, 1999: The workers during the night shift at Tokaimura were cutting corners to save time. Because a nonexistent weighing instrument did not exist, they were not aware that they had overcharged a vessel, nor were they aware that they were handling material with higher enrichment than usual because there was no information anywhere about the enrichment. Their training was so insufficient that they were hardly aware of the dangers that erroneous handling of fissionable material posed. There had been many previous criticality accidents that could serve as precursors. After the accident, most of the blame was placed on the workers. Accountability and responsibility of the management and the authorities were not discussed, nor were necessary improvements of the instrumentation. The dawn of a new era with low electricity prices will hopefully mean a more relaxed attitude and improved safety culture. The entire European electrical grid is integrated now. Reactor manufacturers have become fewer and bigger. For example, the original suppliers of the Swedish reactors, ABB and Westinghouse, are now owned by British Nuclear Fuels, Ltd. Germany's Siemens is controlled now by Framatome. Both BNFL and Framatome have plants all over the world. (author)

Effects of Ionizing Radiation on Insects and Other Arthropods; Effet des rayonnements ionisants sur les insectes et autres arthropodes; Vozdejstvie ioniziruntsej radiatsii na nasekomykh i drugikh chlenistonogikh; Efectos de las radiaciones ionizantes sobre los insectos y otros artropodos

Energy Technology Data Exchange (ETDEWEB)

Stone, William E. [United States Department of Agriculture Laboratories, Mexico City, D.F (Mexico)

1963-09-15

radiaciones gamma en la capacidad de reproduccion, los instintos sexuales, el vigor y la longevidad de la mosca oriental de la fruta, Dacus dorsalis Hendel, la mosca del melon, Dacus cucurbitae Coq., la mosca mediterranea, Ceratitis capitata Wied., la mosca de la fruta mejicana, Anastrepha ludens Loew, y el Anopheles quadrimaculatus Say, e indica los resultados de la campada de exterminio obtenidos en la practica liberando machos esteriles. Revisa tambien los progresos realizados en los Estados Unidos en la campana para exterminar la Cochliomya hominivorax Cqrl., y en los estudios para desarrollar cepas vigorosas, marcadas geneticamente, que permitan identificar con facilidad las moscas esteriles liberadas. Se discuten tambien los resultados de las investigaciones sobre la irradiacion de otras seis especies que atacan a frutas, verduras y otros cultivos agricolas y forestales, otras tres que atacan al ganado, y tres mas que atacan principalmente al hombre. Se trata asimismo de la irradiacion del escorpion, Centruroides limp idus Karsch y del aracnido Amblyomma americanum L., y de la posibilidad de emplear radiaciones ionizantes como tratamiento de cuarentena para las frutas y verduras infestadas con la mosca de la fruta y para los mangos infestados con el gorgojo Stemochetus mangiferae Fabricius. (author) [Russian] V nastoyashchee vremya provodyatsya issledovaniya vozmozhnosti primeneniya metoda sterilizatsii posredstvom oblucheniya dlya unichtozheniya populyatsij tselogo ryada nasekomykh, porazhaptsikh cheloveka, zhivotnykh i razlichnye kul'tury. Ehti predvaritel'nye issledovaniya pokazali, chto ioniziruyushchee obluchenie privodit k sterilizatsii, no chto dlya ehtoj tseli trebuyutsya chrezvychajno raznoobraznye dozy. Okazalos', chto v nekotorykh sluchayakh radiatsionnye povrezhdeniya mogut isklyuchit' vozmozhnost' primeneniya ehtogo metoda u nekotorykh nasekomykh. Prepyatstvie, kotoroe zachastuyu prikhoditsya preodolevat', zaklyuchaetsya v otsutstvii prakticheskikh metodov

Radiotracer Approaches to Carbamate Insecticide Toxicology; Emploi des radio indicateurs pour l'etude de la toxicologie des insecticides a base de carbamates; Primenenie radioaktivnykh indikatorov dlya izucheniya toksikologii karbamatnykh insektitsidov; Estudio con radioindicadores de la toxico logia de los insecticidas a base de carbamatos

Energy Technology Data Exchange (ETDEWEB)

Casida, J. E. [University of Wisconsin, Madison, WI (United States)

1963-09-15

metabolizma karbamatnykh insektitsidov. Sevin (1-naphtyl N-methylcarbamate) byl naibolee tshchatel'no izuchen parallel'no s vozmozhnymi produktami ego gidroliza. Issledovaniya s pomoshch'yu ugleroda-14 ne podtverdili gipotezy o tom, chto pri metabolizme sevina proiskhodit pervonachal'nyj gidroliz, a zatem dal'nejshee razlozhenie produktov gidroliza. Osnovnoj mekhanizm obezvrezhivaniya insektitsida v organizme mlekopitayushchikh i, veroyatno, nasekomykh zaklyuchaetsya v pervonachal'nom okislyayushchem vozdejstvii mikrosom na kehrbamaty v prisutstvii vosstanovlennogo nikotinamid-adenindi-nukleotid fosfata. Sevin bystro podvergaetsya raspadu u mlekopitayushchikh, no sud'ba nekotorykh produktov raspada eshche ne vyyasnena. Nekotorye iz produktov metabolizma mogut byt' obnaruzheny v moloke zhivotnykh. Odnoj iz stupenej metabolizma, po-vidimomu, yavlyaetsya obrazovanie proizvodnogo N-metilola. Predvaritel'nye issledovaniya metabolizma mechennogo radioaktivnymi veshchestvami dime-tilana (2-dimetilkarbamid-3-metilpirazolil-(5)-dimetilkarbamat) i metabolizma podobnogo soedineniya u tarakanov ukazyvaet takzhe na to, chto pri okislenii obrazuyutsya proizvodnye N-metila i N-metilola. Mnogoe ostaetsya sdelat' dlya vyyasneniya svyazi ehtikh reaktsij obezvrezhivaniya s mekhanizmom soprotivlyaemosti, dejstviem sinergistov, selektivnoj toksichnost'yu ehtoj gruppy insektitsidov, a takzhe s prirodoj i znacheniem ostatkov insektitsida. Metabolizm sevina, vvedennogo v rasteniya, nosit, veroyatno, takzhe skoree okislitel'nyj, a ne gidroliticheskij kharakter, ko kharakter produktov i fermentativnye mekhanizmy eshche ne ustanovleny. (author)

Ion irradiation used as surrogate of neutron irradiation in graphite: Consequences on 14C and 36Cl behavior and structural evolution

Science.gov (United States)

Galy, N.; Toulhoat, N.; Moncoffre, N.; Pipon, Y.; Bérerd, N.; Ammar, M. R.; Simon, P.; Deldicque, D.; Sainsot, P.

2018-04-01

Graphite has been widely used as neutron moderator, reflector or fuel matrix in different types of reactors such as gas cooled nuclear reactors (UNGG, Magnox, AGR), RBMK reactors or high temperature gas cooled reactors. Their operation produces a great quantity of irradiated graphite or other carbonaceous waste (around 250,000 tons worldwide) that requires a special management strategy. In the case of disposal, which is a current management strategy, two main radionuclides, 14C and 36Cl might be dose determining at the outlet. Particular attention is paid to 14C due to its long half-life (T∼5730 years) [1] and as major contributor to the radioactive dose. 14C has two main production routes, i) transmutation of nitrogen (14N(n,p)14C) where nitrogen is mainly adsorbed at the surfaces of the irradiated graphite; ii) activation of carbon from the matrix (13C(n,γ)14C). According to leaching tests, it was shown that even if the quantity of 14C released in the solution is low (less than 1% of the initial inventory), around 30% is in the organic form that would be mobile in repository conditions [2,3]. 36Cl is mainly produced through the activation of 35Cl (35Cl(n,γ)36Cl) which is an impurity in nuclear graphite. Its activity is low but it might be highly mobile in clay host rocks. Thus, in order to make informed decisions about the best management process and to anticipate potential radionuclide dissemination during dismantling and in the repository, it is necessary to collect information on 14C and 36Cl location and speciation in graphite, after reactor closure. The goal of the present paper is therefore to use ion irradiation to simulate neutron irradiation and to evaluate the irradiation effects on the behavior of 36Cl and 14C as well as on the induced graphite structure modifications. For that, to understand and model the underlying mechanisms, we used an indirect approach based on 13C or 37Cl implantation to simulate the respective presence of 14C or 36Cl. These

Interesting Developments in UO{sub 2} Technology; Progres interessants dans la technologie du bioxyde d'uranium; Interesnye usovershenstvovaniya tekhnologii UO{sub 2}; Recientes progresos en la tecnologia del UO{sub 2}

Energy Technology Data Exchange (ETDEWEB)

Robertson, J. A.L. [Atomic Energy of Canada Ltd., Chalk River, Ontario (Canada)

1963-11-15

vydelenie iz UO{sub 2} gazov, yavlyayushchikhsya produktami deleniya. V chastnosti, uvelichenie oblucheniya s 10{sup 15} do 10{sup 18} delenij/cm{sup 2} mozhet snizit' ochevidnye skorosti diffuzii dlya ksenona v UO{sub 2} pri posleduyushchikh obzhigakh na koehffitsient 10{sup 3}. Gaz, po-vidimomu, uderzhivaetsya v mel'chajshikh lovushkakh, chast' iz kotorykh sushchestvuet v iskhodnom materiale, a chast' obrazuetsya v rezul'tate radiatsionnogo povrezhdeniya. Tshchatel'nyj analiz pokazal sushchestvovanie medlennoj utechki iz lovushek, chto, veroyatno, ob''yasnyaetsya ogranichennoj rastvorimost'yu ksenona v UO{sub 2}. Vozmozhnost' osushchestvleniya izmerenij v reaktore otkryvaet novuyu fazu eshche bolee vazhnykh ehksperimentov. Oni pokazhut, imeyutsya li kakie-libo potentsial'nye ehkonomicheskie preimushchestva v novykh formakh topliva. V to zhe vremya budut prodolzhat'sya nastojchivye razrabotki spechenoj UO{sub 2} v prostoj geometrii sterzhnya. (author)

Application of Nuclear Radiation to Textile Materials and Processes. Radiation-induced graft copolymerization of vinyl monomers and fibrous polymers; Applications des rayonnements aux textiles. Formation radiochimique de copolymeres ''greffes'' par l'action de monomeres vinyliques sur des polymeres en fibre; Primenenie yadernogo izlucheniya v tekstil'noj promyshlennosti. Obrazovanie privitykh sopolimerov iz vinilovykh monomerov i voloknistykh polimerov pod dejstviem izlucheniya; Aplicaciones de las radiaciones nucleares a los procesos y materiales textiles. Copolimerizacion por injerto radioinducida de monomeros vinilicos y de polimeros fibrosos

Energy Technology Data Exchange (ETDEWEB)

Armstrong, Jr., A. A.; Rutherford, H. A. [University of North Carolina at Raleigh, NC (United States)

1963-11-15

que la modificacion de las propiedades de la fibra mediante el procedimiento descrito podria aprovecharse en el terreno comerical de encontrarse procedimientos que permitan aumentar la velocidad de difusion del monomero en la estructura de la fibra. (author) [Russian] Razrabotana standartnaya metodika radiatsionnoj privivki letuchikh vinilovykh monomerov k voloknistym polimeram. Ehta metodika sostoit v vvedenii organicheskogo soedineniya v parovoj faze; protsess osushchestvlyaetsya ili putem odnovremennogo oblucheniya ili predvaritel'nym oblucheniem ot istochnika Co''6''0. Issledovany sistemy, povyshayushchie skorost' privitoj sopolimeriehatsii. Ustanovleno, chto mozhno vvodit' znachitel'nye kolichestva nekotorykh monomerov, ne vyzyvaya radiatsionnogo povrezhdeniya voloknistogo substrata. Imenno te voloknistye materialy, kotorye sravnitel'no neustojchivo reagiruyut na radiatsiyu naibolee legko prisoedinyayut vinilovye soedineniya. K takim materialam otnosyatsya tsellyulozy, slozhnye ehfiry tsellyulozy, poliamidy i polipropilen. Prisoedinenie monomera k voloknistomu substratu, ochevidno, proiskhodit po svobodno-radikal'nomu mekhanizmu, i, kak uzhe ukazyvalos' vyshe, mozhno snachala obluchit' volokno a zatem provodit' privivku, podvergaya obluchennyj material vozdejstviyu parov mono- mera. Predprinimalis' popytki vyyasnit' dlitel'nost' sushchestvovaniya svobodnykh radikalov i, khotya dovol'no trudno poluchit' tochnye dannye, bylo pokazano, chto svobodnye radikaly prodolzhayut sokhranyat'sya v nekotorykh voloknakh spustya 15 - 20 chasov posle oblucheniya, dazhe pri komnatnoj temperature. Predvaritel'nye opytnye dannye takzhe navodyat na mysl' o tom,chto pri posleradiatsionnom privivanii monomer diffundiruet k svobodno-radikal'nym tsentram s razlichnymi skorostyami v raznykh voloknakh. Volokno khlopka, modifitsirovannoe vvedeniem v ego sostav poliakrilonitrila, ustojchivo k dejstviyu mikroorganizmov. Vvedenie 3,5%-nogo poliakrilonitrila pozvolyaet materialu sokhranyat

Condensed Matter Nuclear Science

Science.gov (United States)

Biberian, Jean-Paul

2006-02-01

. Bloch ions / T. A. Chubb. II. Inhibited diffusion driven surface transmutations / T. A. Chubb. III. Bloch nuclides, Iwamura transmutations, and Oriani showers / T. A. Chubb. Bose-Einstein condensate. Theoretical study of nuclear reactions induced by Bose-Einstein condensation in Pd / K.-I. Tsuchiya and H. Okumura. Proposal for new experimental tests of the Bose-Einstein condensation mechanism for low-energy nuclear reaction and transmutation processes in deuterium loaded micro- and nano-scale cavities / Y. E. Kim ... [et al.]. Mixtures of charged bosons confined in harmonic traps and Bose-Einstein condensation mechanism for low-energy nuclear reactions and transmutation processes in condensed matters / Y. E. Kim and A. L. Zubarev. Alternative interpretation of low-energy nuclear reaction processes with deuterated metals based on the Bose-Einstein condensation mechanism / Y. E. Kim and T. O. Passell. Multi-body fusion. [symbol]He/[symbol]He production ratios by tetrahedral symmetric condensation / A. Takahashi. Phonon coupling. Phonon-exchange models: some new results / P. L. Hagelstein. Neutron clusters. Cold fusion phenomenon and solid state nuclear physics / H. Kozima. Neutrinos, magnetic monopoles. Neutrino-driven nuclear reactions of cold fusion and transmutation / V. Filimonov. Light monopoles theory: an overview of their effects in physics, chemistry, biology, and nuclear science (weak interactions) / G. Lochak. Electrons clusters and magnetic monopoles / M. Rambaut. Others. Effects of atomic electrons on nuclear stability and radioactive decay / D. V. Filippov, L. I. Urutskoev, and A. A. Rukhadze. Search for erzion nuclear catalysis chains from cosmic ray erzions stopping in organic scintillator / Yu. N. Bazhutov and E. V. Pletnikov. Low-energy nuclear reactions resulting as picometer interactions with similarity to K-shell electron capture / H. Hora ... [et al.] -- 5. Other topics. On the possible magnetic mechanism of shortening the runaway of RBMK-1000 reactor

Post-Construction Testing of the Elk River, Hallam and Piqua Power Reactor Plants; Essais apres construction des centrales nucleaires d'Elk River, de Hallam et de Piqua; Predehkspluatatsionnoe ispytanie Ehlk-riverskoj, Khehlpemskoj i Pikuaskoj ehnergeticheskikh reaktornykh ustanovok; Ensayos posteriores a la construccion de las centrales nucleoelectricas de Elk River, Hallam y Piqua

Energy Technology Data Exchange (ETDEWEB)

Pursel, C. A. [United States Atomic Energy Commission, Argonne, IL (United States)

1963-10-15

byli: Ehlk-riverskij reaktor. V nekotorykh chastyakh baka reaktora obnaruzheny treshchiny, chto potrebovalo obshirnykh issledovanij i analizov, a takzhe nekotorogo remonta i modifikatsii baka. Nedostatochnaya proizvoditel'nost' parootdelitelya potrebovala zameny k modifikatsii nekotorykh metallicheskikh detalej baka reaktora. Khehldemskaya yadernaya ehnergeticheskaya ustanovka. Uvlechenie zashchitnogo gaza geliya privelo k neobkhodimosti modifitsirovat' vtorichnye natrievye kontury. Povrezhdenie truboprovoda v promezhutochnom natrievo-natrievom; teploobmennike potrebovalo provedeniya analizov dlya opredeleniya prichiny povrezhdeniya, vsled za chem posledovalo izvlechenie i remont teploobmennika. Pikuaskaya yadernaya ehnergeticheskaya ustanovka. Pri khimicheskoj ochistke sistemy truboprovodov povrezhdeno neskol'ko ventilej, chto potrebovalo ikh remonta ili zameny. Utechki v sisteme organicheskogo teplonositelya i v sisteme indikatsii para priveli k povtornym zaderzhkam. Posle okonchaniya neobkhodimogo remonta i vneseniya izmenenij fakticheskie rabochie kharakteristiki kazhdogo iz trekh reaktorov priblizilis' vplotnuyu k proektnym raschetam. (author)