Computerized monitoring system of residual cyclic life of WWER RP equipment
|
|
- Ruth Baldwin
- 5 years ago
- Views:
Transcription
1 Computerized monitoring system of residual cyclic life of WWER RP equipment
2 The computerized monitoring system of residual cyclic life (SACOR-M) is intended for in-service calculation of cumulative fatigue damage of metal and fatigue extension of defects of WWER RP equipment by actual parameters of heat-andforce loads under real operation conditions. SACOR-M monitors the cumulative fatigue damage at the most stressed points of the structure selected by the results of design strength calculations. The loading parameters are calculated by the indications of standard transducers recording the current state of RP equipment. SACOR-M carries out the stress calculation by using the functional dependences of stresses on loading parameters for each check point. During RP operation the actual course of conditions can differ from that assumed in design strength calculations both by coolant parameters, and by the number of loading cycles. The monitoring of fatigue damage by the real loading of equipment is the urgent problem. SACOR-M main characteristics ÔÔ set of check points on the RP equipment is chosen in the scope sufficient for evaluation of RP residual life by the well-known mechanisms of fatigue damage; ÔÔ conservative models applied for the calculation of loading parameters allow to use the standard transducers that essentially reduces the labour input and dose commitments during SACOR installation and operation; ÔÔ use of Duamel s integral relation in the functional dependences of stresses on loading parameters permits to perform the calculation of stresses in check points directly at NPP; ÔÔ universal methods for determination of coefficients in the functional dependences of stresses on operation parameters permits to use the earlier strength calculations made at the stage of RP design justification; ÔÔ mathematical formula for stress determination permits to take into account all loading factors: weight, pressure, temperature compensation under the conditions of stratification and without it, non-uniformity of temperature field across the component caused by thermal shocks, thermal pulsations and stratification, off-design displacement of components; ÔÔ use of two mechanisms of metal degradation during evaluation of residual life (accumulation of fatigue damage and cyclic growth of initial unsoundness) permits to check the various limiting states; ÔÔ possibility for using the material property database both by certificates, and that obtained during in-service inspection of metal. The results of SACOR-M operation can be used ÔÔ during justification of residual life of RP equipment in case of a single design condition and off-design condition (for example, failure of BRU-A to seat); ÔÔ for optimization of the program of non-destructive testing of RP equipment for the purpose of reducing the period of preventive maintenance (for example, reduction in inspection scope of MCP welded joints); ÔÔ when exceeding the design number of conditions specified in the process specifications (for example, scheduled trips of RCP sets); ÔÔ for optimization of operational conditions and detection of unfavourable loading factors (for example, thermopulsations in the injection pipeline and the connecting pipeline during malfunction of low flow injection controller in PRZ); ÔÔ inspection of damages revealed during operation (for example, area of welded joint No. 111 of SG); ÔÔ when changing over to the daily load-follow mode of operation (automatic registration of cumulative fatigue damage under conditions of power variation); ÔÔ in case of extension of RP service life.
3 Design basis and verification of design formulas Stresses are calculated by using the functional dependences of stresses on the loading parameters, which are verified by the data of design strength calculations of components according to representative sequence of design conditions. The fatigue damage calculated by the method of rain (GOST ) is chosen as a comparison criterion. Calculation Design Comparison of stresses by design calculations and by SACOR for check point on the reactor instrumentation nozzle Calculation of cumulative fatigue damage for current situation No. of standard transducer Check point of equipment Fatigue damage by SACOR as of YA11 1 Hot leg: weld near MCP1 Dnom850 nozzle 0, YA41 13 Joint of surge line into MCP4 hot leg 0, Extract from the calculation log of cumulative fatigue damage YB10 2 Feed water nozzle of SG3 0, YC00 10 Reactor vessel flange 0, YC00 10 Convolved cycle YC00 10 Non-convolved cycle YC00 10 Total Diagram of accumulation of fatigue damage at check point of ECCS nozzle calculated by real loading for Rostov NPP, Unit 1
4 Computerized monitoring system of residual cyclic life of WWER RP equipment Calculation of extension of initial unsoundness and limiting states Repair of minor unsoundness in MCP welded joints is complicated in technology, expensive and its successful realization is out of guarantee. MCP could operate throughout design service life without repair. Method of calculated estimation of extension of initial defects is proposed as a compensating measure that will enable to optimize periodic nondestructive testing of the given welded joints. SACOR-M application experience at NPPs in operation Rostov NPP, Unit 1. SACOR-M has been put into operation at Rostov NPP, Unit 1, since November, It is installed in PC of the automated workstation (PC AWS) of SACOR-M system. SACOR software was developed for operational system Windows and certified at NTC YaRB GAN of Russia (certificate for software No. 161). SACOR, as applied to Rostov NPP, Unit 1, is introduced into project V-320. Power cut of Rostov NPP, Unit 1, from the grid took place on by the protection of generator with reactor scram and failure of BRU-A to close. SACOR-M was used for calculation of cumulative fatigue damage as a result of off-design condition. SG-1 was permitted to operate without instrumentation control of metal state by the results of analysis of accident condition scenario and calculation of fatigue damage. Tianwan NPP, Unit 1 (SACOR-428). SA- COR-428 involved in MCDS has been put into operation since November 2005, it is realized on actual data base obtained from Tianwan NPP standard transducers. SACOR-428 software is developed for operational system Unix (Solaris-8). The system made it possible to execute the on-line monitoring of temperature conditions throughout commissioning work of the reactor plant, to reveal local conditions of temperature pulsations, e.g. under off-design operation of regulator for water injection into pressurizer, to assess thermal stress level of the units and to optimize operation of the regulator. This condition recorded at Tianwan NPP, Unit 1, is shown in figure wherein constant temperature in PRZ (the upper line), pulsating temperature in surge pipeline near MCP ( grey line), pulsating temperature in injection pipeline (green line) and the lower red line temperature in MCP hot leg 4.
5 Computerized monitoring system of residual cyclic life of WWER RP equipment Equipment monitored and check points SACOR-M monitors RP equipment residual life in the following scope - reactor vessel and top head, SG, PRZ, MCP, pressurizing system and ECCS pipelines. Maximum design basis damage, welded joints with initial presence of defects, areas of cold coolant injection, damage areas as per operational feedback are assumed as the criteria for choosing check points. Figure shows a part of check points for assessment of SG fatigue damages. Their total number in V-320 RP equipment amounts to 118. Transfer of transducer indication data to SACOR-M server involved in MCDS In calculating loading factors by indications of standard transducers there are used immerse standard resistance thermometers, pressure pickups in the primary and secondary circuits, displacement transducers on shockabsorbers, indications of flowmeters, position indicators of the valves and surface resistance thermometers. NPP is possible to be additionally equipped with hardware components, placing of transducers and computer aids. TLS-U data collection Standard transducers and SG displacements ICIS transducers TLS-U top level system of the Unit, MCDS monitoring, control and diagnostics system, ICIS - in-core instrumentation system. ICIS server Rejection of error values SACOR-320 server (data sampling and calculation) Consideration of loading factors by indications of standard transducers In calculating stresses the following loading factors are taken into account: from weight, primary and secondary pressure, temperature compensation of pipelines under actual displacement of the equipment, thermal shocks, thermal pulsations and coolant stratification under all operating conditions. New position of MCP nozzle Integration of monitoring of SG actual displacements together with available SACOR-M aids makes it possible to diagnose the stressed state in the area of hot collector nozzle-to-sg vessel welded joint (area of welded joint No.111). Figure shows the example of calculation of displacement of MCP hot nozzle on SG vessel calculated by indications of displacement transducers on shock-absorbers.
6 21, Ordzhonikidze Street, Podolsk, Moscow region, Russian Federation Tel.: (495) , (4967) Fax: (495) , (4967) The following references can be used for more detailed acquaintance with the issue: 1. N.V. Shary, V.P. Semishkin, V.A. Piminov, Yu.G. Dragunov, «Strength of main equipment and pipelines of WWER reactor plants» M.: IzdAT, Chapter Operating experience of system of the automated control of a residual cyclic resource for RP with WWER Bogachev A.V., Bakirov M.B. (VNIIAES), Dranchenko B.N., Semishkin V.P. (OKB Gidropress ). 18th International Conference on Structural Mechanics in Reactor Technology (SMiRT 18). Beijing, China, August 7-12, Development of system SACOR-M. A.V. Bogachev, B.N. Dranchenko, V.P. Semishkin, V.Ya. Berkovitch. Issues of nuclear science and engineering. Series «NPP safety assurance». Scientific and technical collection. Issue 15. WWER reactor plants. Podolsk A.V. Bogachev. Lecture «Implementation of systems of calculated - experimental diagnostics of residual cyclic life of NPP equipment and pipelines». The 2-nd Russian interbranch school - seminar «Operational stability of nuclear power plant components». Collection of summaries of lectures, FSUE «NIIP», Lytkarino V.P. Semishkin, A.V. Bogachev, B.N. Dranchenko. Calculations of RP equipment stressed state performed by using FEM within the framework of creating the computerized monitoring system of residual cyclic life for AES The 5-th International scientific and technical conference Safety assurance of NPP with WWER», Podolsk. May 29 June 1, A.V. Bogachev, V.Ya. Berkovitch, B.N. Dranchenko, V.P. Semishkin. Determination of the loading factors for calculation of stresses by SACOR as applied to AES-2006 RP design. The 5-th International scientific and technical conference «Safety assurance of NPP with WWER», Podolsk. May 29 June 1, 2007.
Reactor plant for NPP WWER-1000
2008 Key assets to support a strategic choice NPPs with WWER-type reactors (WWER-440,, WWER-1200) in operation and under construction Evolutionary, safe and innovative design ÔÔ reactor plant is a reactor
More informationSPECIFIC DEGRADATIONS OF VVER-1000
SPECIFIC DEGRADATIONS OF VVER-1000 (in view of lifetime extension) Dimitar Popov Kozloduy NPP, Bulgaria IAEA Technical Meeting on Degradation of Primary Components of PW cooled NPPs, Vienna, 05-08 Nov,
More informationTechnical potentialities of integration of reactor vessel external cooling in operating WWER-440 plant - Assessment results
Technical potentialities of integration of reactor vessel external cooling in operating WWER-440 plant - Assessment results Speaker Pantyushin S.I. ERMSAR-2013 02-04.10.2013 1. Introduction. Current status
More informationINVESTIGATION OF CRITICAL SAFETY FUNCTION INTEGRITY IN CASE OF STEAM LINE BREAK ACCIDENT FOR VVER 1000/V320
International Conference 12th Symposium of AER, Sunny Beach, pp.99-105, 22-28 September, 2002. INVESTIGATION OF CRITICAL SAFETY FUNCTION INTEGRITY IN CASE OF STEAM LINE BREAK ACCIDENT FOR VVER 1000/V320
More informationDesign, Safety Technology and Operability Features of Advanced VVERs
Technical Cooperation Project INT/4/142 Interregional Workshop on Advanced Nuclear Reactor Technology for Near Term Deployment IAEA Headquarters, Vienna, Austria, 4-8 July 2011 Design, Safety Technology
More informationNEW APPROACHES FOR FLOW-ACCELERATED CORROSION
NEW APPROACHES FOR FLOW-ACCELERATED CORROSION M. Bakirov a, H. Cheng b, V. Levchuk a, L. Selesnev a, A. Eremyn a IAEA-CN-155-052 a Centre of Materials Researches and Lifetime Management (CMSLM), Moscow,
More informationRELAP5/MOD3.2 INVESTIGATION OF A VVER-440 STEAM GENERATOR HEADER COVER LIFTING
Science and Technology Journal of BgNS, Vol. 8, 1, September 2003, ISSN 1310-8727 RELAP5/MOD3.2 INVESTIGATION OF A VVER-440 STEAM GENERATOR HEADER COVER LIFTING Pavlin P. Groudev, Rositsa V. Gencheva,
More informationOperating Nuclear Reactors in Ukraine: Enhancement of Safety and Performance
IAEA-CN-164-6S05 Operating Nuclear Reactors in Ukraine: Enhancement of Safety and Performance S. Bozhkoa, G. Gromovb, S. Sholomitskyb, O.Sevbob, G. Balakanc a State Nuclear Regulatory Authority of Ukraine,
More informationNEUTRON PHYSICS CALCULATION FOR VVER-1000 ABSORBER ELEMENT LIFETIME DETERMINATION ABSTRACT
NEUTRON PHYSICS CALCULATION FOR VVER-1000 ABSORBER ELEMENT LIFETIME DETERMINATION K.Yu. Kurakin, S.A. Kushmanov OKB GIDRORESS ABSTRACT Absorber element (AE) with compound absorber has been operating in
More informationRussian regulatory approach to evaluation of passive systems used for specific BDBA S (SBO, loss of UHS) during safety review of NPP
Federal Environmental, Industrial and Nuclear Supervision Service Scientific and Engineering Centre for Nuclear and Radiation Safety Member of Russian regulatory approach to evaluation of passive systems
More informationHeat exchanger equipment of TPPs & NPPs
Heat exchanger equipment of TPPs & NPPs Lecturer: Professor Alexander Korotkikh Department of Atomic and Thermal Power Plants TPPs Thermal power plants NPPs Nuclear power plants Content Steam Generator
More informationAndrey Afanasiev, Victor Makarov, Leonid Lyakishev, Ivan Matvienko, Michail Puchkov, Dmitry Ivanov
SEISMIC AND HYDRODYNAMIC TESTS OF FA DUMMY AND SHEM-3 RCCA DRIVE FOR AES-2006 PROJECT IN A LARGE-SCALE TESTING FACILITY Andrey Afanasiev, Victor Makarov, Leonid Lyakishev, Ivan Matvienko, Michail Puchkov,
More informationA Research Reactor Simulator for Operators Training and Teaching. Abstract
Organized and hosted by the Canadian Nuclear Society. Vancouver, BC, Canada. 2006 September 10-14 A Research Reactor Simulator for Operators Training and Teaching Ricardo Pinto de Carvalho and José Rubens
More informationManufacturability, safety, reliability and efficiency
Manufacturability, safety, reliability and efficiency VVER reactor plant, related to a group of multi-loop plants, was initially designed as a six-loop plant and then evolved to a fourloop plant. In VVER
More informationGUIDELINES ON PRESSURIZED THERMAL SHOCK ANALYSIS FOR WWER NUCLEAR POWER PLANTS
IAEA-EBP-WWER-08 (Rev. 1) GUIDELINES ON PRESSURIZED THERMAL SHOCK ANALYSIS FOR WWER NUCLEAR POWER PLANTS Revision 1 A PUBLICATION OF THE EXTRABUDGETARY PROGRAMME ON THE SAFETY OF WWER AND RBMK NUCLEAR
More informationStructural Integrity and NDE Reliability II
Structural Integrity and NDE Reliability II NDE Experience and Lessons Learnt from Recent EU Projects Focused on Assistance to the Armenian NPP L. Horacek, Nuclear Research Institute, Czech Republic ABSTRACT
More informationBN-1200 Reactor Power Unit Design Development
International Conference on Fast Reactors and Related Fuel Cycles: Safe Technologies and Sustainable Scenarios (FR13) BN-1200 Reactor Power Unit Design Development B.A. Vasilyev a, S.F. Shepelev a, M.R.
More informationBasic Engineering Solutions in the VBER-500 Power Unit for Regional Power Systems
Basic Engineering Solutions in the VBER-500 Power Unit for Regional Power Systems A.E. Arefyev, V.V. Petrunin, Yu.P. Fadeev (JSC "Afrikantov OKBM") Yu.A. Ivanov, A.V. Yeremin (JSC "NIAEP") Yu.M. Semchenkov,
More informationFatigue Monitoring for Demonstrating
Fatigue Monitoring for Demonstrating Fatigue Design Basis Compliance Timothy J. Griesbach Structural Integrity Associates IAEA 2 nd International Symposium on Nuclear Power Plant Life Mgmt. 15-18 October,
More informationRELAP5 Analysis of Krško Nuclear Power Plant Abnormal Event from 2011
RELAP5 Analysis of Krško Nuclear Power Plant Abnormal Event from 2011 ABSTRACT Andrej Prošek Jožef Stefan Institute Jamova cesta 39 SI-1000, Ljubljana, Slovenia andrej.prosek@ijs.si Marko Matkovič Jožef
More informationSMR/1848-T21b. Course on Natural Circulation Phenomena and Modelling in Water-Cooled Nuclear Reactors June 2007
SMR/1848-T21b Course on Natural Circulation Phenomena and Modelling in Water-Cooled Nuclear Reactors 25-29 June 2007 T21b - Selected Examples of Natural Circulation for Small Break LOCA and Som Severe
More informationLight Water Reactor in Russia
ROSATOM STATE CORPORATION ON NUCLEAR ENERGY ROSATOM Light Water Reactor in Russia Presented M.A.Bykov OKB GIDROPRESS 17 th TWG-LWR Meeting, Vienna, 18-20 June 2012 NPP Electricity Generation in Russia
More informationDESIGN AND SAFETY PRINCIPLES LEONTI CHALOYAN DEPUTY CHIEF ENGINEER ON MODERNIZATION
DESIGN AND SAFETY PRINCIPLES LEONTI CHALOYAN DEPUTY CHIEF ENGINEER ON MODERNIZATION VIENNA OKTOBER 3-6, 2016 1 ANPP * ANPP is located in the western part of Ararat valley 30 km west of Yerevan close to
More informationSteam Generator Ageing Management in Slovakia current practices and related issues
International Atomic Energy Agency Meeting on Update and Upgrade of IAEA-TECDOC on Ageing Management of Steam Generators IAEA, Vienna, 15 18 June 2009 Steam Generator Ageing Management in Slovakia current
More informationAES-2006 PSA LEVEL 1. PRELIMINARY RESULTS AT PSAR STAGE
AES-2006 PSA LEVEL 1. PRELIMINARY RESULTS AT PSAR STAGE A. Kalinkin a*, A. Solodovnikov a, S. Semashko a a JSC "VNIPIET", Saint-Petersburg, Russian Federation Abstract: This report represents PSA level
More informationAdvanced Sodium Fast Reactor Power Unit Concept
International Conference on Fast Reactors and Related Fuel Cycles: Challenges and Opportunities (FR 09) Advanced Sodium Fast Reactor Power Unit Concept V.M. Poplavsky a, A.M. Tsybulya a, Yu.E. Bagdasarov
More informationFederal Environmental, Industrial and Nuclear Supervision Service of Russia FEDERAL STANDARDS AND RULES IN THE FIELD OF USE OF ATOMIC ENERGY
Federal Environmental, Industrial and Nuclear Supervision Service of Russia FEDERAL STANDARDS AND RULES IN THE FIELD OF USE OF ATOMIC ENERGY Approved by Decree of the Federal Environmental, Industrial
More informationCONTENTS CO-GENERATING WATER-DESALINATING FACILITY POWERED BY SVBR-75/100 NUCLEAR REACTOR DESIGN ORGANIZATIONS INVOLVED IN THE PROJECT
CONTENTS CO-generating water-desalinating facility powered by SVBR-75/100 NUCLEAR 1 REACTOR Design organizations involved in the project 1 Layout of co-generating nuclear-powered water desalinating facility
More informationIMPLEMENTATION OF A FATIGUE MONITORING SYSTEM IN MOCHOVCE UNIT 1 AND 2 FIRST RESULTS
International Conference Nuclear Energy in Central Europe 2001 Hoteli Bernardin, Portorož, Slovenia, September 10-13, 2001 www: http://www.drustvo-js.si/port2001/ e-mail: PORT2001@ijs.si tel.:+ 386 1 588
More informationDEVELOPMENT AND APPLICATION OF PROBABILISTIC SAFETY ASSESSMENT PSA IN DAYA BAY NUCLEAR POWER STATION
18th International Conference on Structural Mechanics in Reactor Technology (SMiRT 18) Beijing, China, August 7-12, 2005 SMiRT18-A01-2 DEVELOPMENT AND APPLICATION OF PROBABILISTIC SAFETY ASSESSMENT PSA
More informationPractice on NDE Licensing and Inspection Qualifications for Nuclear Power Plant In-service Inspections
18 th World Conference on Nondestructive Testing, 16-20 April 2012, Duban, South Africa Practice on NDE Licensing and Inspection Qualifications for Nuclear Power Plant In-service Inspections Jinhong LIU,
More informationModule 05 WWER/ VVER (Russian designed Pressurized Water Reactors)
Module 05 WWER/ VVER (Russian designed Pressurized Water Reactors) 1.3.2016 Prof.Dr. Böck Technical University Vienna Atominstitut Stadionallee 2, 1020 Vienna, Austria ph: ++43-1-58801 141368 boeck@ati.ac.at
More informationTERMO-HYDRAULICS AND TERMO-MECHANICAL LOADING OF VVER-440 REACTOR PRESSURE VESSEL
TERMO-HYDRAULICS AND TERMO-MECHANICAL LOADING OF VVER-440 REACTOR PRESSURE VESSEL G. Gálik 1, V. Kutiš 2, J. Paulech 3, V. Goga 4 Abstract: This article describes a pressure thermal shock simulation methodology
More informationIAEA-TECDOC-1361 Assessment and management of ageing of major nuclear power plant components important to safety
IAEA-TECDOC-1361 Assessment and management of ageing of major nuclear power plant components important to safety Primary piping in PWRs July 2003 The originating Section of this publication in the IAEA
More informationSupporting Deterministic T-H Analyses for Level 1 PSA
Supporting Deterministic T-H Analyses for Level 1 PSA ABSTRACT SLAVOMÍR BEBJAK VUJE, a.s. Okružná 5 918 64 Trnava, Slovakia slavomir.bebjak@vuje.sk TOMÁŠ KLIMENT VUJE, a.s. Okružná 5 918 64 Trnava, Slovakia
More informationAnalysis of the State of Steam Generator Tubes
IYNC 28 Interlaken, Switzerland, September 2-26,28 ANNOTATION Analysis of the State of Steam Generator Tubes Bergunker Olga JSC OKB "GIDROPRESS", 14213 Podolsk, Russia grpress@grpress.podolsk.ru The problem
More informationProfile LFR-66 V-200 RUSSIA. V-200 water facility
Profile LFR-66 V-200 RUSSIA GENERAL INFORMATION NAME OF THE FACILITY Three loop water facility V-200 with an integrated model of the fast reactor designed to study thermal hydraulic processes in primary
More informationProfile LFR-62 SGI RUSSIA. Facility SGI for studies of thermohydraulic characteristics of FACILITY. atomic power plants.
Profile LFR-62 SGI RUSSIA General information NAME OF THE Facility SGI for studies of thermohydraulic characteristics of FACILITY atomic power plants. ACRONYM SGI Coolant technology Water LOCATION (address)
More informationRUSSIAN REGULATORY APPROACH TO EXTENSION OF NUCLEAR POWER PLANT SERVICE LIFE
RUSSIAN REGULATORY APPROACH TO EXTENSION OF NUCLEAR POWER PLANT SERVICE LIFE E.V. Vasileva, N.I. Karpunin Federal State Institution Scientific and Engineering Center for Nuclear and Radiation Safety (SEC
More informationCentre of Materials Researches and Lifetime Management (CMSLM), Moscow, Russia b
APPLICATION OF THE LEAK BEFORE BREAK CONCEPT ON NPP UNITS OF THE FIRST GENERATION WITH WWER-440 REACTORS. IMPROVEMENT OF LEAK DETECTION SYSTEMS CONSIDERING NEW ELABORATED APPROACHES M. Bakirov a, V. Levchuk
More informationPTS re-evaluation project for Czech NPPs
PTS re-evaluation project for Czech NPPs Vladislav Pištora, Miroslav Žamboch, Pavel Král, Ladislav Vyskočil Fourth International Conference on Nuclear Power Plant Life Management 23 27 October 2017 Lyon,
More informationRussian Federal Nuclear and Radiation Safety Inspectorate (Gosatomnadzor) FEDERAL RULES AND REGULATIONS ON THE USE OF NUCLEAR ENERGY
Translated from Russian Russian Federal Nuclear and Radiation Safety Inspectorate (Gosatomnadzor) FEDERAL RULES AND REGULATIONS ON THE USE OF NUCLEAR ENERGY APPROVED BY Gosatomnadzor Resolution No. 4 of
More informationStatus report VVER-1200 (V-491) (VVER-1200 (V-491))
Status report 108 - VVER-1200 (V-491) (VVER-1200 (V-491)) Overview Full name Acronym Reactor type Coolant Moderator Neutron spectrum Thermal capacity Gross Electrical capacity Design status Designers VVER-1200
More informationInspection Qualification II
Inspection Qualification II Qualification According to ENIQ of Complete EPR RPV Inspection Using Phased Array F. Mohr, G. Guse, intelligendt System & Services, Germany INTRODUCTION On each new reactor
More informationNuclear I&C Systems Basics. The role of Instrumentation and Control Systems in Nuclear Power Plants, and their Characteristics
Nuclear I&C Systems Basics The role of Instrumentation and Control Systems in Nuclear Power Plants, and their Characteristics Functions of Nuclear I&C Functions and significance of the Instrumentation
More informationStructural Integrity and NDE Reliability I
Structural Integrity and NDE Reliability I Assessment of Failure Occurrence Probability as an Input for RI-ISI at Paks NPP R. Fótos, University of Miskolc, Hungary L. Tóth, P. Trampus, University of Debrecen,
More informationDEVELOPING AND IMPLEMENTATION OF A FATIGUE MONITORING SYSTEM FOR THE NEW EUROPEAN PRESSURIZED WATER REACTOR EPR
DEVELOPING AND IMPLEMENTATION OF A FATIGUE MONITORING SYSTEM FOR THE NEW EUROPEAN PRESSURIZED WATER REACTOR EPR Christian Pöckl, Wilhelm Kleinöder AREVA NP GmbH Freyeslebenstr. 1, 91058 Erlangen, Germany
More information5th Pan American Conference for NDT 2-6 October 2011, Cancun, Mexico. Systems for inspection and repair of WWER type steam generators
5th Pan American Conference for NDT 2-6 October 2011, Cancun, Mexico Systems for inspection and repair of WWER type steam generators Adrian KOVALYK, Pavol JABLONICKY, Peter PILAT Division for diagnostics
More informationFuel Cycle of VVER-1000: technical and economic aspects. E. Kosourov, V. Pavlov, A. Pavlovichev
8th International Conference on WWER Fuel Performance, Modelling and Experimental Support 26 September 04 October 2009, Helena Resort near Burgas, Bulgaria, (in co-operation with the International Atomic
More informationHigh Energy Line Breaks. Design Requirements and Engineering Practice. Alexey Berkovsky 1), Alexander Kultsep 1), Leonid Lyakishev 2)
High Energy Line Breaks. Design Requirements and Engineering Practice. Alexey Berkovsky 1), Alexander Kultsep 1), Leonid Lyakishev 2) 1) CKTI Vibroseism, Saint Petersburg, Russia 2) FSUE OKB "GIDROPRESS",
More informationUp-to-date issues of VVER Technology Development
Up-to-date issues of VVER Technology Development 10 th International Scientific and Technical Conference Safety, Efficiency and Economics of Nuclear Power industry JSC Concern Rosenergoatom May 25-27,
More informationMarch 16, Mr. William M. Dean Director, Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC
ANTHONY R. PIETRANGELO Senior Vice President and Chief Nuclear Officer 1201 F Street, NW, Suite 1100 Washington, DC 20004 P: 202.739.8081 arp@nei.org nei.org March 16, 2015 Mr. William M. Dean Director,
More informationStatus report 85 - VVER-1500 (V-448) (VVER-1500 (V-448))
Status report 85 - VVER-1500 (V-448) (VVER-1500 (V-448)) Overview Full name Acronym Reactor type Coolant Moderator Neutron spectrum Thermal capacity Gross Electrical capacity Design status Designers VVER-1500
More informationYa.I.Shtrombakh. «Russian nuclear power plants life time management» Finnish Nuclear Society s seminar: "Nuclear Russia today and tomorrow"
Ya.I.Shtrombakh «Russian nuclear power plants life time management» Finnish Nuclear Society s seminar: "Nuclear Russia today and tomorrow" Helsinki 2009 Main directions of RRC Kurchatov institute activities
More informationScenarios of Heavy Beyond-Design-Basis Accidents in HTGRs N.G. Kodochigov, Yu.P. Sukharev
Scenarios of Heavy Beyond-Design-Basis Accidents in HTGRs N.G. Kodochigov, Yu.P. Sukharev IAEA Technical Meeting on the Safety of High Temperature Gas Cooled Reactors in the Light of the Fukushima Daiichi
More informationEngineering Flow Solutions. Pumps and Solutions for Nuclear Power Plants
Engineering Flow Solutions Pumps and Solutions for Nuclear Power Plants Pumps and Solutions for Nuclear Power Plants The HMS GROUP is the leading manufacturer of pumps, compressors and modular equipment,
More informationNDE SOLUTIONS INTERCONTRÔLE NONDESTRUCTIVE EXAMINATION
INTERCONTRÔLE NONDESTRUCTIVE EXAMINATION Specialists in nondestructive examination has specialized in automated nondestructive testing of nuclear reactor primary components and more specifically on steam
More informationEngineering Flow Solutions. Pumps and Solutions for Nuclear Power Plants
Engineering Flow Solutions Pumps and Solutions for Nuclear Power Plants Pumps and Solutions for Nuclear Power Plants The HMS GROUP is the leading manufacturer of pumps, compressors and modular equipment,
More informationPSA ANALYSIS FOCUSED ON MOCHOVCE NPP SAFETY MEASURES EVALUATION FROM OPERATIONAL SAFETY POINT OF VIEW
International Conference Nuclear Energy in Central Europe 2001 Hoteli Bernardin, Portorož, Slovenia, September 10-13, 2001 www: http://www.drustvo-js.si/port2001/ e-mail: PORT2001@ijs.si tel.:+ 386 1 588
More informationLifetime-Management and Operational Lifetime Extension at Paks Nuclear Power Plant
Transactions of the 17 th International Conference on Structural Mechanics in Reactor Technology (SMiRT 17) Prague, Czech Republic, August 17 22, 2003 Paper # D02-2 Lifetime-Management and Operational
More informationThermal-hydraulic model of the reactor facility with lead coolant in the ATHLET code
Journal of Physics: Conference Series PAPER OPEN ACCESS Thermal-hydraulic model of the reactor facility with lead coolant in the ATHLET code To cite this article: V A Chudinova and S P Nikonov 2018 J.
More informationIAEA-TECDOC Probabilistic safety assessments of nuclear power plants for low power and shutdown modes
IAEA-TECDOC-1144 Probabilistic safety assessments of nuclear power plants for low power and shutdown modes March 2000 The originating Section of this publication in the IAEA was: Safety Assessment Section
More informationBasic Principles of the Construction of Residual Resource Estimation Systems
17th World Conference on Nondestructive Testing, 25-28 Oct 2008, Shanghai, China Basic Principles of the Construction of Residual Resource Estimation Systems Abstract Vladimir V. KLYUEV, Mikhail V. FILINOV,
More informationGrid Stability and Safety Issues Associated With Nuclear Power Plants
Grid Stability and Safety Issues Associated With Nuclear Power Plants Dr. John H. Bickel, Ph.D. Workshop on International Grid Interconnection in Northeast Asia Beijing, China May 14-16, 2001 1 Items to
More informationECNDT Moscow 2010 ADVANCED APPROACH OF REACTOR PRESSURE VESSEL HEAD INSPECTION AND REPAIR of CRDM J-WELD
ECNDT Moscow 2010 ADVANCED APPROACH OF REACTOR PRESSURE VESSEL HEAD INSPECTION AND REPAIR of CRDM J-WELD Mladen Pajnić, Hrvoje Franjić, Gabrijel Smoljkić, Krunoslav Markulin, Fran Jarnjak INETEC Institute
More informationEXPERIENCE OF SERVICE LIFE PROLONGATION OF NPP UNITS OF THE FIRST GENERATION
IAEA Second International Symposium on Nuclear Power Plant Life Management 15-18 October 2007, Shanghai, China EXPERIENCE OF SERVICE LIFE PROLONGATION OF NPP UNITS OF THE FIRST GENERATION M. Bakirov, V.
More informationSimulation of thermal hydraulics accidental transients: evaluation of MAAP5.02 versus CATHAREv2.5
1/12 Simulation of thermal hydraulics accidental transients: evaluation of MAAP5.02 versus CATHAREv2.5 J. Bittan¹ 1) EDF R&D, Clamart (F) Summary MAAP is a deterministic code developed by EPRI that can
More informationMEASUREMENT AND EVALUATION SYSTEMS FOR NPP COMMISSIONING
MEASUREMENT AND EVALUATION SYSTEMS FOR NPP COMMISSIONING Marek Elko, VUJE, Inc. ABSTRACT Standard core monitoring and information systems are designed with an emphasis on normal operation of nuclear power
More informationValve Group The First Line of Safety
0510 Valve Group The First Line of Safety Farris Engineering Our Company Farris Engineering, a business unit of Curtiss-Wright, has been at the forefront in the design and manufacture of spring-loaded
More informationApplication for Permission to Extend the Operating Period and Application for Approval of Construction Plans of Unit 3 at Mihama Nuclear Power Station
November 26, 2015 The Kansai Electric Power Co., Inc. Application for Permission to Extend the Operating Period and Application for Approval of Construction Plans of Unit 3 at Mihama Nuclear Power Station
More informationThe ultrasonic examination and monitoring of austenite welds of stainless steel pipelines at Russian Nuclear Power Plants.
The ultrasonic examination and monitoring of austenite welds of stainless steel pipelines at Russian Nuclear Power Plants. V.G. Badalyan, A. K. Vopilkinе Scientific and production center Echo+, Russia,
More informationADVANCED I&C SYSTEMS FOR NUCLEAR POWER PLANTS FEEDBACK OF EXPERIENCE
International Conference Nuclear Energy in Central Europe 2001 Hoteli Bernardin, Portorož, Slovenia, September 10-13, 2001 www: http://www.drustvo-js.si/port2001/ e-mail: PORT2001@ijs.si tel.:+ 386 1 588
More informationRosatom Seminar on Russian Nuclear Energy Technologies and Solutions
ROSATOM STATE ATOMIC ENERGY ОАО CORPORATION «ТВЭЛ» - российский ROSATOM производитель ядерного топлива Nuclear fuel for VVER-1200 Rosatom Seminar on Russian Nuclear Energy Technologies and Solutions L.T.
More informationSignificant Events in Rostechnadzor Activity Regarding WWER-type NPPs Operation within the Period from September 2015 up to July 2016
FEDERAL ENVIRONMENTAL, INDUSTRIAL AND NUCLEAR SUPERVISION SERVICE OF RUSSIA Significant Events in Rostechnadzor Activity Regarding WWER-type NPPs Operation within the Period from September 2015 up to July
More informationInternational Atomic Energy Agency. Technical Working Group on Nuclear Power Plant Instrumentation and Control (TWG-NPPIC)
International Atomic Energy Agency Technical Working Group on Nuclear Power Plant Instrumentation and Control (TWG-NPPIC) Main Achievements and Development of I&C in Rosatom Speaker: Alexey Chernyaev,
More informationENVIRONMENTAL SAFETY OF NUCLEAR POWER PLANTS OF RUSSIAN DESIGN
International Journal of Mechanical Engineering and Technology (IJMET) Volume 9, Issue 11, November 2018, pp. 172 176, Article ID: IJMET_09_11_019 Available online at http://www.iaeme.com/ijmet/issues.asp?jtype=ijmet&vtype=9&itype=11
More informationSTRESS TESTS ACTION PLAN LEONTI CHALOYAN DEPUTY CHIEF ENGINEER ON MODERNIZATION
STRESS TESTS ACTION PLAN LEONTI CHALOYAN DEPUTY CHIEF ENGINEER ON MODERNIZATION VIENNA MARCH 27-29, 2017 1 ANPP * The ANPP is located in the western part of Ararat valley, 30 km west of Yerevan, close
More informationExperience feedback: methodology & recent events. Kiev March 22, 2017
Experience feedback: methodology & recent events Kiev March 22, 2017 Nuclear safety stakeholders in France Technical safety organization Regulatory body Others Licensee Kiev March 22, 2016 International
More informationPRESSURE THERMAL SHOCK ANALYSIS FOR NUCLEAR REACTOR PRESSURE VESSEL
PRESSURE THERMAL SHOCK ANALYSIS FOR NUCLEAR REACTOR PRESSURE VESSEL Gabriel Gálik 1, Vladimir Kutiš 1, Jakub Jakubec 1, Juraj Paulech 1, Justín Murín 1 1 Institute of Automobile Mechatronics, Faculty of
More informationQUALIFICATION AND APPLICATION OF IN-SERVICE INSPECTION OF VVER-440 CONTROL ROD DRIVE PROTECTION PIPES
QUALIFICATION AND APPLICATION OF IN-SERVICE INSPECTION OF VVER-440 CONTROL ROD DRIVE PROTECTION PIPES Krunoslav Markulin, Matija Vavrous INETEC-Institute for Nuclear technology, Croatia Jani Pirinen, Petri
More informationStation Blackout Analysis for a 3-Loop Westinghouse PWR Reactor Using TRACE
The Egyptian Arab Journal of Nuclear Sciences and Applications Society of Nuclear Vol 50, 3, (229-239) 2017 Sciences and Applications ISSN 1110-0451 Web site: esnsa-eg.com (ESNSA) Station Blackout Analysis
More informationVerification of VVER-1200 NPP Simulator in Normal Operation and Reactor Coolant Pump Coast-Down Transient
World Journal of Engineering and Technology, 2017, 5, 507-519 http://www.scirp.org/journal/wjet ISSN Online: 2331-4249 ISSN Print: 2331-4222 Verification of VVER-1200 NPP Simulator in Normal Operation
More informationA PRA application to support outage schedule planning at OL1 and OL2 units
A PRA application to support outage schedule planning at OL1 and OL2 units Hannu Tuulensuu a a Teollisuuden Voima Oyj, Eurajoki, Finland Abstract: For Olkiluoto 1 (OL1) and Olkiluoto 2 (OL2) nuclear power
More informationSafety Provisions for the KLT-40S Reactor Plant
6th INPRO Dialogue Forum on Global Nuclear Energy Sustainability: Licensing and Safety Issues for Small and Medium-sized Nuclear Power Reactors (SMRs) 29 July - 2 August 2013 IAEA Headquarters, Vienna,
More informationRELAP 5 ANALYSIS OF PACTEL PRIMARY-TO-SECONDARY LEAKAGE EXPERIMENT PSL-07
Fifth International Seminar on Horizontal Steam Generators 22 March 21, Lappeenranta, Finland. 5 ANALYSIS OF PACTEL PRIMARY-TO-SECONDARY LEAKAGE EXPERIMENT PSL-7 József Bánáti Lappeenranta University of
More informationFATIGUE MONITORING FOR DEMONSTRATING FATIGUE DESIGN BASIS COMPLIANCE
FATIGUE MONITORING FOR DEMONSTRATING FATIGUE DESIGN BASIS COMPLIANCE D. Gerber, G. Stevens, T. Gilman, J. Zhang Structural Integrity Associates,San Jose,USA Structural Integrity Associates,San Jose,USA
More information6.1 Introduction. Control 6-1
Control 6-1 Chapter 6 Control 6.1 Introduction Although the bulk of the process design of the Heat Transport System is done prior to the control design, it is helpful to have the key features of the control
More informationNSSS Design (Ex: PWR) Reactor Coolant System (RCS)
NSSS Design (Ex: PWR) Reactor Coolant System (RCS) Purpose: Remove energy from core Transport energy to S/G to convert to steam of desired pressure (and temperature if superheated) and moisture content
More informationVVER-440/213 - The reactor core
VVER-440/213 - The reactor core The fuel of the reactor is uranium dioxide (UO2), which is compacted to cylindrical pellets of about 9 height and 7.6 mm diameter. In the centreline of the pellets there
More informationIntegrated functional modeling method for NPP plant DiD risk monitor and its application for conventional PWR
Integrated functional modeling method for NPP plant DiD risk monitor and its application for conventional PWR YOSHIKAWA Hidekazu 1, YANG Ming 2, and ZHANG Zhijian 3 1.College of Nuclear Science and Technology,
More informationVerification of the MELCOR Code Against SCDAP/RELAP5 for Severe Accident Analysis
Verification of the Code Against SCDAP/RELAP5 for Severe Accident Analysis Jennifer Johnson COLBERT 1* and Karen VIEROW 2 1 School of Nuclear Engineering, Purdue University, West Lafayette, Indiana 47907-2017,
More informationCorium Retention Strategy on VVER under Severe Accident Conditions
NATIONAL RESEARCH CENTRE «KURCHATOV INSTITUTE» Corium Retention Strategy on VVER under Severe Accident Conditions Yu. Zvonarev, I. Melnikov National Research Center «Kurchatov Institute», Russia, Moscow
More informationOverview of AP1000 Commissioning Inspection in P.R. China. National Nuclear Safety Administration Sept. 2017
Overview of AP1000 Commissioning Inspection in P.R. China National Nuclear Safety Administration Sept. 2017 contents 1 3 2 Status of AP1000 Projects Inspection of AP1000 Commissioning Challenges of AP1000
More informationCONTENTS. CHAPTER I Up-to Day Methods of the Non-Destructive Inspection of Pipelines Systems
CONTENTS CHAPTER I Up-to Day Methods of the Non-Destructive Inspection of Pipelines Systems 1. Standard non-destructive inspection methods used for revealing, locating, identifying and assessing flaws
More informationIMPLEMENTATION OF SAFETY PARAMETER DISPLAY SYSTEM ON RUSSIAN NPPs WITH WER REACTORS
IMPLEMENTATION OF SAFETY PARAMETER DISPLAY SYSTEM ON RUSSIAN NPPs WITH WER REACTORS V.G. DOUNAEV, V.T. NEBOYAN Consyst Co. Ltd, Moscow, Russian Federation Abstract This report gives a short overview of
More informationStress tests specifications Proposal by the WENRA Task Force 21 April 2011
Stress tests specifications Proposal by the WENRA Task Force 21 April 2011 Introduction Considering the accident at the Fukushima nuclear power plant in Japan, the Council of the European Union declared
More informationAP1000 European 21. Construction Verification Process Design Control Document
2.5 Instrumentation and Control Systems 2.5.1 Diverse Actuation System Design Description The diverse actuation system (DAS) initiates reactor trip, actuates selected functions, and provides plant information
More informationNUCLEAR PLANT WITH VK-300 BOILING WATER REACTORS FOR POWER AND DISTRICT HEATING GRIDS
7th International Conference on Nuclear Engineering Tokyo, Japan, April 19-23, 1999 ICONE-7335 NUCLEAR PLANT WITH VK-300 BOILING WATER REACTORS FOR POWER AND DISTRICT HEATING GRIDS Yu.N. Kuznetsov*, F.D.
More informationUse of PSA to Support the Safety Management of Nuclear Power Plants
S ON IMPLEMENTATION OF THE LEGAL REQUIREMENTS Use of PSA to Support the Safety Management of Nuclear Power Plants РР - 6/2010 ÀÃÅÍÖÈß ÇÀ ßÄÐÅÍÎ ÐÅÃÓËÈÐÀÍÅ BULGARIAN NUCLEAR REGULATORY AGENCY TABLE OF CONTENTS
More informationKeywords: Thermalhydraulics, VVER-440, safety, strainer, clogging, downstream effects, fuel element, sump, risk.
SAFETY IMPACT OF THE INSULATION FIBERS PENETRATING SUMP STRAINERS AND ACCUMULATING IN LOVIISA VVER-440 FUEL BUNDLES Seppo Tarkiainen, Olli Hongisto, Timo Hyrsky, Heikki Kantee, Ilkka Paavola Fortum Power,
More information