Involvement of EDF in the Halden Reactor Project: a long-term cooperation in R&D
|
|
- Rafe Elliott
- 6 years ago
- Views:
Transcription
1 Involvement of EDF in the Halden Reactor Project: a long-term cooperation in R&D Atoms For the Future 2016 Session: Nuclear Fuel 28/06/2016 Alexandre Lavoil alexandre.lavoil@edf.fr EDF SEPTEN/CN
2 OUTLINE 1. PRESENTATION OF THE HRP 2. COLLABORATION EDF HRP 3. FUEL ROD CONCEPTION : SOME WORDS 4. EDF INDUSTRIAL NEEDS AND CONTRIBUTION OF HRP 5. CONCLUSIONS 2
3 PRESENTATION OF THE HRP HRP: Halden Reactor Project, started in 1958 International collaborative research for the safe and reliable operation of Nuclear Power Plants Affiliated to the OECD NEA in Paris Jointly funded by the membership 20 countries including Norway >100 organizations (utilities, vendors, R&D, licensing/regulatory authorities) Budget for is 431 MNOK (~46 M ) Norway contributes 35% of HRP budget 3
4 PRESENTATION OF THE HRP HBWR : Halden Boiling Water Reactor 20 MW heavy water reactor (34 bar, 235 C) 35 simultaneous tests possible (IFA : Instrumented Fuel Assembly) Instrumentation provides data on performance of nuclear fuels and materials under normal, transient and accident conditions (dry-out & LOCA) 4
5 PRESENTATION OF THE HRP Man-Technology-Organisation Objective: provide knowledge about how and why accidents occur, with the aim of preventing them from happening in the operation of complex processing facilities Nuclear Fuels and Materials (F&M) Objective: provide knowledge about the performance of nuclear fuels and materials under normal, transient and accident conditions, with the aim of increasing safety and reliability plus improving the nuclear fuel cycle 5
6 COLLABORATION EDF - HRP HRP : Decisive and strategic topics related to fuel behaviour Major actors of the nuclear industry involved (regulators, utilities, manufacturers) Necessity for EDF to be involved in this project! Collaboration between EDF and HRP started in June st EDF representative sent in October
7 FUEL ROD CONCEPTION : SOME WORDS Fuel assembly rods - ~4.5 meters high Fuel rods mm large - ~4 meters high 7
8 Fuel thermomechanical criteria which needs to be verified Codes & Models used to demonstrate the respect of these criteria How are these codes validated? Experiences! Ø 1 mm thermocouple 8
9 Internal pressure: The pressure in the rod has to remain lower than the limit pressure which could lead to a gap re-opening One phenomena which can lead to pressure increase : Fission Gas Release (FGR) Important to understand the FGR mechanism in order to implement models in the fuel thermomechanical codes to be able to simulate it IFA-716 : Fission gas release experiment : Objectives : Onset of fission gas release Densification and swelling behaviour Inclusion of novel fuel types with improved thermal conductivity (BeO doped) (Lower fuel T => reduced fission gas release) 9
10 IFA-716 : Fission gas release experiment 6 instrumented rods with types of fuel Different grain size Different additive Different levels of Cr content UO2 large grain 10
11 Reactivity control: Gadolinium is a neutronic poison, used in some of the French PWR to control the reactivity in the beginning of life Some UO 2 Gd 2 O 3 can be inserted in the fuel assemblies, but gadolinium can modify the thermomechanical behaviour of the fuel Important to understand this behaviour in order to be able to add some specific models in the thermomechanical codes IFA : UO 2 Gd 2 O 3 fuel behaviour : objectives : Comparison between UO 2, UO 2 + 2% Gd 2 O 3, UO 2 + 8% Gd 2 O 3 Evaluation of gadolinium on: Fuel densification & swelling FGR 11
12 IFA-681.1: UO 2 Gd 2 O 3 fuel behaviour Delay of the temperature increase of the Gd fuel Similar T for the UO 2 & UO 2 +2% Lower T for the UO 2 +8% 12
13 IFA-681.1: UO 2 Gd 2 O 3 fuel behaviour No densification for UO 2 -Gd 2 O 3 Fuel swelling seems to decrease with Gadolinium content 13
14 LOCA (Loss Of Coolant Accident): Fuel rods not in contact with the coolant Important consequences on the fuel rods behaviour Heat-up phase Ballooning of the rod Fuel Fragmentation Relocation & Dispersal (FFRD) 14
15 Linear Power Cladding temperature INDUSTRIAL NEEDS / CONTRIBUTION OF HRP IFA-650 : LOCA experiment : Objectives Behaviour of a fuel rod in LOCA conditions Find the Burn-Up threshold for FFRD during a LOCA Experimental procedure: Single rod in a rig Blowdown (~100 s) SCRAM LOCA In-pile measurements: Cladding temperature Forced circulation (imposed by HBWR conditions) Natural circulation: Isolation of the loop Heat-up phase Cooldown phase Fuel rod Pressure Fuel rod elongation Activity in the loop 50 P APRP temps 15
16 IFA-650 : LOCA experiment : Post-Irradiation Examinations (PIE): 16
17 Fukushima s accident Post-Fukushima studies: Development of ATF (Accident Tolerant Fuel) 17
18 ATF (Accident Tolerant Fuel): Objectives: Pellets: High melting fuel temperature Increase thermal conductivity (to decrease centerline temperature) Enhance retention of fission products Cladding: High melting cladding temperature Maintain coolable geometry (limit ballooning & burst) Retain high post-quench ductility Reduce production of H 2 Coating of the cladding is one of the solution imaginated by the supplier 18
19 IFA-796: ATF experiment Objective: demonstrate that the in-reactor behaviour of ATF claddings is at least as good as Zr-based claddings in use today under prototypic PWR operation conditions Measurements: Oxide thickness increase Dimensional behaviour 6 rods provided by 6 manufacturers (industrials, R&D, labs ), varying: Cladding compositions Coating types and thicknesses Uncoated Zy4 Non-Optimized coating Optimized coating Experiment to be started in 2016/ IDARRAGA, LE FLEM, BRACHET et al. American Nuclear Society, Top Fuel 2013
20 CONCLUSIONS EDF HRP: a long-term cooperation in R&D HRP brings experimental answers to the industrial issues which can touch EDF Major topics of the nuclear fuel industry treated in the experiments of the HRP (LOCA, ATF ) Mains actors of the nuclear field (manufacturers, providers, regulators, research centers) of all the major countries of the nuclear world Decisive collaboration for EDF! 20
21 THANK YOU! 21
2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea
HALDEN S IN-PILE TEST TECHNOLOGY FOR DEMONSTRATING THE ENHANCED SAFETY OF WATER REACTOR FUELS Margaret A. McGrath 1 1 OECD Halden Reactor Project, IFE: Os Alle 5/P.O. Box 173, 1751 Halden, Norway, Margaret.mcgrath@ife.no
More informationEnhanced Accident Tolerant Fuel at AREVA NP. Dr. Elmar Schweitzer, Dr. Jeremy Bischoff COP23, Bonn, 11/08/2017
Enhanced Accident Tolerant Fuel at Dr. Elmar Schweitzer, Dr. Jeremy Bischoff COP23, Bonn, 11/08/2017 Why Develop eatf Solutions? Zr alloy eatf solution p.2 eatf Program u Evolutionary Concept (Near-term
More informationCONTRIBUTION OF RESEARCH REACTORS TO THE PROGRAMMES FOR RESEARCH AND TECHNOLOGICAL DEVELOPMENT ON SAFETY
CONTRIBUTION OF RESEARCH REACTORS TO THE PROGRAMMES FOR RESEARCH AND TECHNOLOGICAL DEVELOPMENT ON SAFETY J. Couturier, F. Pichereau, C. Getrey, J. Papin, B. Clément INSTITUT DE RADIOPROTECTION ET DE SURETE
More informationA New Method Taking into Account Physical Phenomena Related to Fuel Behaviour During LOCA
S. BOUTIN S. GRAFF A. BUIRON A New Method Taking into Account Physical Phenomena Related to Fuel Behaviour During LOCA Seminar 1a - Nuclear Installation Safety - Assessment AGENDA 1. Context 2. Development
More informationDesign bases and general design criteria for nuclear fuel. 1 General 3. 2 General design criteria 3
GUIDE 1 Nov. 1999 YVL 6.2 Design bases and general design criteria for nuclear fuel 1 General 3 2 General design criteria 3 3 Design criteria for normal operational conditions 4 4 Design criteria for operational
More information2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea
DIMENSIONAL BEHAVIOUR TESTING OF ACCIDENT TOLERANT FUEL (ATF) IN THE HALDEN REACTOR R. Szőke, M. A. McGrath, P. Bennett Institute for Energy Technology OECD Halden Reactor Project ABSTRACT In order to
More informationThe Norwegian Thorium Initiative
Thor Energy The Norwegian Thorium Initiative Saleem Drera, VP R&D ThEC15 Mumbai India, October 2015 Thanks to: The International Thorium Consortium Established in 2012 by Thor Energy Objective: To jointly
More informationSIMULATION OF FUEL BEHAVIOURS UNDER LOCA AND RIA USING FRAPTRAN AND UNCERTAINTY ANALYSIS WITH DAKOTA
SIMULATION OF FUEL BEHAVIOURS UNDER LOCA AND RIA USING FRAPTRAN AND UNCERTAINTY ANALYSIS WITH DAKOTA IAEA Technical Meeting on Modelling of Water-Cooled Fuel Including Design Basis and Severe Accidents,
More informationWestinghouse ACCIDENT TOLERANT FUEL PROGRAM
Westinghouse ACCIDENT TOLERANT FUEL PROGRAM Fausto Franceschini Consulting Engineer Global Technology Development Westinghouse Electric Co. International Workshop on Advanced Reactor Systems and Future
More information(printed) (electronic)
Performing Organisation lnstitutt for Energiteknikk Halden Document no.: Date IFE/HR/E -2011 /005 2011/09/23 ProjecUContract no. and name ClienUSponsor Organisation and reference: Title and subtitle Upgrading
More informationSandrine BOUTIN Stéphanie GRAFF Aude TAISNE Olivier DUBOIS REVIEW OF FUEL SAFETY CRITERIA IN FRANCE
Sandrine BOUTIN Stéphanie GRAFF Aude TAISNE Olivier DUBOIS REVIEW OF FUEL SAFETY CRITERIA IN FRANCE AGENDA 1. About French rulemaking 2. Review of all acceptance criteria in France 3. Summary 2 ABOUT FRENCH
More informationMixed-oxide (MOX) fuel performance benchmarks
Mixed-oxide (MOX) fuel performance benchmarks L. J. Ott a,*, Terje Tverberg b, Enrico Sartori c a Oak Ridge National Laboratory, Oak Ridge, Tennessee, U.S.A. b OECD Halden Reactor Project, Halden, Norway
More informationPost-test analysis of the Halden LOCA experiment IFA using the Falcon code. Abstract
F2.2 Post-test analysis of the Halden LOCA experiment IFA-65.7 using the Falcon code. G. Khvostov, a * W. Wiesenack, b B.C.Oberländer, c E. Kolstad, b G. Ledergerber, d M.A. Zimmermann a a Paul Scherrer
More informationWestinghouse Advanced Doped Pellet Characteristics and Irradiation behaviour
Westinghouse Advanced Doped Pellet Characteristics and Irradiation behaviour Karin Backman 1, Lars Hallstadius 1 and Gunnar Rönnberg 2 1. Westinghouse Electric Sweden, 2. OKG AB Sweden IAEA - Technical
More informationTopic 1: Fuel Fabrication. Daniel Mathers and Richard Stainsby
Topic 1: Fuel Fabrication Daniel Mathers and Richard Stainsby CEIDEN NNL meeting, Sellapark, 1 st February 2016 Level of Benefit / Ambition UK Fuel Ambition: Development of Fuels with Enhanced Safety,
More informationIn-core measurements of fuel-clad interactions in the Halden reactor
In-core measurements of fuel-clad interactions in the Halden reactor Peter Bennett Halden Project IAEA Technical Meeting on Fuel Rod Instrumentation and In-Pile Measurement Techniques Halden, Norway 3
More informationON-GOING STUDIES AT CEA ON CHROMIUM COATED ZIRCONIUM BASED NUCLEAR FUEL CLADDINGS FOR ENHANCED ACCIDENT TOLERANT LWRS FUEL
ON-GOING STUDIES AT CEA ON CHROMIUM COATED ZIRCONIUM BASED NUCLEAR FUEL CLADDINGS FOR ENHANCED ACCIDENT TOLERANT LWRS FUEL J.C. Brachet *, M. Le Saux, M. Le Flem, S. Urvoy, E. Rouesne, T. Guilbert, C.
More informationDeveloping Fuels with Enhanced Accident Tolerance. Fiona Rayment and Dave Goddard
Developing Fuels with Enhanced Accident Tolerance Fiona Rayment and Dave Goddard NNL Technical Conference 30 th April 2015 UK Fuel Ambition: Development of Fuels with Enhanced Safety, Economic & sustainability
More informationStatus of NEA Nuclear Science activities related to accident tolerant fuels
Status of NEA Nuclear Science activities related to accident tolerant fuels Jim Gulliford, Head of Nuclear Science OECD-NEA 1 Outline OECD-NEA Nuclear Science & Data Bank Activities related to innovative
More informationThorium-Plutonium LWR Fuel
Thorium-Plutonium LWR Fuel Irradiation Testing Imminent October 2012 Julian F. Kelly, Chief Technology Officer What Why How Overview Testing ceramic (Th,Pu)O2 fuel with prototypical LWR composition & microstructure
More informationIn-pile testing of CrN, TiAlN and AlCrN coatings on Zircaloy cladding in the Halden Reactor
In-pile testing of CrN, TiAlN and AlCrN coatings on Zircaloy cladding in the Halden Reactor R. Van Nieuwenhove, V. Andersson, J. Balak, B. Oberländer Sector Nuclear Technology, Physics and Safety Institutt
More informationUS Transient Testing Program
www.inl.gov US Transient Testing Program Dan Wachs National Technical lead for Transient Testing Idaho National Laboratory 18 th IGORR Meeting, Sydney, Australia December 7, 2017 Fuel Safety Research Objective:
More informationSustaining Material Testing Capacity in France: From OSIRIS to JHR
Sustaining Material Testing Capacity in France: From OSIRIS to JHR to support industry and public organizations in R&D irradiation programs on nuclear fuel and materials Stéphanie MARTIN, French Alternative
More informationNuclear Fuel Diagnostics (MåBIL-project)
Nuclear Fuel Diagnostics (MåBIL-project) SKC symposium October 11-12, 2016 Prof. Ane Håkansson, UU Doc. Staffan Jacobsson Svärd, UU Dr. Peter Andersson, UU Outline Background of MÅBiL Nuclear Fuel Diagnostics
More informationRECENT ACTIVITIES AND PLAN WITH FRAPCON/FRAPTRAN
RECENT ACTIVITIES AND PLAN WITH FRAPCON/FRAPTRAN FRAPCON/FRAPTRAN User Group Meeting 2014, Sendai, Japan, September 18, 2014 Presented by Jinzhao Zhang (jinzhao.zhang@gdfsuez.com) Co-authors: Adrien Dethioux,
More informationACTIVITIES in NUCLEAR FUEL BEHAVIOUR
ACTIVITIES in NUCLEAR FUEL BEHAVIOUR Nuclear Science Committee Status: October 2002 Presented by Wolfgang Wiesenack R&D Needs for Current and Future Nuclear Systems, Nov. 2002 1 Outline Introduction -
More informationCEA ACTIVITIES SUPPORTING THE OPERATING FLEET OF NPPS
CEA ACTIVITIES SUPPORTING THE OPERATING FLEET OF NPPS Colloque SFEN Atoms for the future Christophe Béhar 24 OCTOBRE 2012 Christophe Béhar - October 24th, 2012 PAGE 1 DEN ASSIGNMENTS Nuclear Energy Support
More informationEnergy for the future IFE / The Halden Reactor Project Instruments and Measurements for Nuclear Research and Development
Energy for the future IFE / The Halden Reactor Project Instruments and Measurements for Nuclear Research and Deelopment EURAMET Symposium, Oslo, May 26 th 2016 Contents of the presentation Nuclear power
More informationUnderstanding the effects of reflooding in a reactor core beyond LOCA conditions
Understanding the effects of reflooding in a reactor core beyond LOCA conditions F. Fichot 1, O. Coindreau 1, G. Repetto 1, M. Steinbrück 2, W. Hering 2, M. Buck 3, M. Bürger 3 1 - IRSN, Cadarache (FR)
More informationJoint Opening Session Margaret McGrath, Halden Project: Welcome and Introduction
EHPG Sandefjord 2016 Technical Program Fuel and Materials Monday May 9 0830-1200 Joint Opening Session Paper No.: 01 Joint Opening Session Margaret McGrath, Halden Project: Welcome and Introduction Session
More informationFission gas release from high burnup fuel during normal and power ramp conditions
1 Fission gas release from high burnup fuel during normal and power ramp conditions M. Amaya, J. Nakamura, F Nagase Japan Atomic Energy Agency (JAEA) amaya.masaki@jaea.go.jp This study was conducted as
More information2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea
DEVELOPMENT STATUS OF MICRO-CELL UO2 PELLET FOR ACCIDENT TOLERANT FUEL Dong-Joo Kim, Keon Sik Kim, Dong-Seok Kim, Jang Soo Oh, Jong Hun Kim, Jae Ho Yang, Yang-Hyun Koo Korea Atomic Energy Research Institute,
More informationAREVA NP S ENHANCED ACCIDENT TOLERANT FUEL DEVELOPMENTS: FOCUS ON CR- COATED M5 CLADDING
AREVA NP S ENHANCED ACCIDENT TOLERANT FUEL DEVELOPMENTS: FOCUS ON CR- COATED M5 CLADDING Jeremy Bischoff 1, Christine Delafoy 2, Christine Vauglin 3, Pierre Barberis 4, Cédric Roubeyrie 5, Delphine Perche
More informationPROPOSAL OF A GUIDE TO PERFORMANCE ASSESSMENT OF FUEL RODS FOR NUCLEAR POWER PLANTS
2013 International Nuclear Atlantic Conference - INAC 2013 Recife, PE, Brazil, November 24-29, 2013 ASSOCIAÇÃO BRASILEIRA DE ENERGIA NUCLEAR - ABEN ISBN: 978-85-99141-05-2 PROPOSAL OF A GUIDE TO PERFORMANCE
More information2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea
ANALYSIS OF THE COMBINED EFFECTS ON THE FUEL PERFORMANCE OF UO 2 -BeO AS FUEL AND IRON-BASED ALLOY AS CLADDING Claudia Giovedi 1, Alfredo Abe 2, Rafael O. R. Muniz 2, Daniel S. Gomes 2, Antonio Teixeira
More information2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea
PRACTICAL APPLICATION OF DETAILED THERMOMECHANICAL FEM MODEL OF FUEL ROD Martin Dostál 1, Jan Klouzal 1, Vítězslav Matocha 1 1 ÚJV Řež, a. s., Severe Accidents and Thermomechanics Department, Hlavní 130,
More informationSafety design approach for JSFR toward the realization of GEN-IV SFR
Safety design approach for JSFR toward the realization of GEN-IV SFR Advanced Fast Reactor Cycle System R&D Center Japan Atomic Energy Agency (JAEA) Shigenobu KUBO Contents 1. Introduction 2. Safety design
More informationA RIA Failure Criterion based on Cladding Strain
A RIA Failure Criterion based on Cladding Strain by C. Vitanza OECD Halden Reactor Project (1) Paper to be presented at the IAEA Technical Committee Meeting on Fuel Behaviour under Transient and LOCA Conditions
More information2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea
Recent Progress of the CNNC Accident Tolerant Fuel(ATF) Program Yongjun Jiao 1, Wenjie Li 2, Ruiqian Zhang 2, Ping Chen 1, Shixin Gao 1 1 Science and Technology on Reactor System Design Technology Laboratory
More informationCalculation of Pellet Radial Power Distributions with Monte Carlo and Deterministic Codes
Progress in NUCLEAR SCIENCE and TECHNOLOGY, Vol. 2, pp.301-305 (2011) TECHNICAL MATERIAL Calculation of Pellet Radial Power Distributions with Monte Carlo and Deterministic Codes Motomu SUZUKI *, Toru
More informationR&D activities related to nuclear fuel performance and technology at the DG JRC. Paul VAN UFFELEN
R&D activities related to nuclear fuel performance and technology at the DG JRC Paul VAN UFFELEN 1 Introduction 2 JRC Core Staff (2004) Institute for Reference Materials and Measurements Institute for
More information15-Nov English text only. Safety Significance of the Halden IFA-650 LOCA Test Results. English text only JT
Unclassified English text only Unclassified NEA/CSNI/R(2010)5 Organisation de Coopération et de Développement Économiques Organisation for Economic Co-operation and Development 15-Nov-2010 English text
More informationFUEL ROD PERFORMANCE MEASUREMENTS AND RE-INSTRUMENTATION CAPABILITIES AT THE HALDEN PROJECT
FUEL ROD PERFORMANCE MEASUREMENTS AND RE-INSTRUMENTATION CAPABILITIES AT THE HALDEN PROJECT Olav Aarrestad and Helge Thoresen OECD Halden Reactor Project Norway Abstract In the area of instrumentation
More informationEnsuring Spent Fuel Pool Safety
Ensuring Spent Fuel Pool Safety Michael Weber Deputy Executive Director for Operations U.S. Nuclear Regulatory Commission American Nuclear Society Meeting June 28, 2011 1 Insights from Fukushima Nuclear
More informationFuel Reliability (QA)
Program Description Fuel Reliability (QA) Program Overview Fuel failures and other fuel-related issues can have significant operational impacts on nuclear power plants. Failures, for example, can cost
More informationNUCLEAR FUEL AND REACTOR
NUCLEAR FUEL AND REACTOR 1 Introduction 3 2 Scope of application 3 3 Requirements for the reactor and reactivity control systems 4 3.1 Structural compatibility of reactor and nuclear fuel 4 3.2 Reactivity
More informationIrradiation capabilities at the Halden reactor and testing possibilities under supercritical water conditions
The 7th International Symposium on Supercritical Water-Cooled Reactors ISSCWR-7 15-18 March 2015, Helsinki, Finland ISSCWR7-2036 Irradiation capabilities at the Halden reactor and testing possibilities
More informationJournal of American Science 2014;10(2) Burn-up credit in criticality safety of PWR spent fuel.
Burn-up credit in criticality safety of PWR spent fuel Rowayda F. Mahmoud 1, Mohamed K.Shaat 2, M. E. Nagy 3, S. A. Agamy 3 and Adel A. Abdelrahman 1 1 Metallurgy Department, Nuclear Research Center, Atomic
More informationMaterial characterization Capabilities at IFE Kjeller (NMAT)
Material characterization Capabilities at IFE Kjeller (NMAT) NOMAGE4, Halden 31.10&1.11.2011 Institute for Energy Technology Sector: Nuclear Safety & Reliability NUSP, Head: Dr. M.McGrath Department: Nuclear
More informationOUT-OF-PILE R&D ON COATED NUCLEAR FUEL ZIRCONIUM BASED CLADDINGS FOR ENHANCED ACCIDENT TOLERANCE IN LWRS
OUT-OF-PILE R&D ON COATED NUCLEAR FUEL ZIRCONIUM BASED CLADDINGS FOR ENHANCED ACCIDENT TOLERANCE IN LWRS Université Paris-Saclay J.C. Brachet(*), I. Idarraga(**), M. Le Flem, M. Le Saux, F. Schuster, F.
More informationNURETH Progress on Severe Accident Code Benchmarking in the Current OECD TMI-2 Exercise
NURETH-15 544 Progress on Severe Accident Code Benchmarking in the Current OECD TMI-2 Exercise G. Bandini (ENEA), S. Weber, H. Austregesilo (GRS), P. Drai (IRSN), M. Buck (IKE), M. Barnak, P. Matejovic
More informationPARAMETRIC STUDY OF THERMO-MECHANICAL BEHAVIOUR OF 19- ELEMENT PHWR FUEL BUNDLE HAVING AHWR FUEL MATERIAL
PARAMETRIC STUDY OF THERMO-MECHANICAL BEHAVIOUR OF 19- ELEMENT PHWR FUEL BUNDLE HAVING AHWR FUEL MATERIAL R. M. Tripathi *, P. N. Prasad, Ashok Chauhan Fuel Cycle Management & Safeguards, Directorate of
More informationNUMERICAL STUDY OF IN-VESSEL CORIUM RETENTION IN BWR REACTOR
NUMERICAL STUDY OF IN-VESSEL CORIUM RETENTION IN BWR REACTOR M. VALINČIUS Lithuanian Energy Institute Kaunas, Lithuania Email: mindaugas.valincius@lei.lt A. KALIATKA Lithuanian Energy Institute Kaunas,
More informationA Brief Summary of Analysis of FK-1 and FK-2 by RANNS
A Brief Summary of Analysis of FK- and by RANNS Motoe Suzuki, JAEA. Introduction For the purpose of benchmarking the RANS code, FK- and experiments conducted at NSRR were analyzed. Emphasis was placed
More informationCANDU Safety #12: Large Loss of Coolant Accident F. J. Doria Atomic Energy of Canada Limited
CANDU Safety #12: Large Loss of Coolant Accident F. J. Doria Atomic Energy of Canada Limited 24-May-01 CANDU Safety - #12 - Large LOCA.ppt Rev. 0 1 Overview Event sequence for a large break loss-of of-coolant
More information3D Printing of Components and Coating Applications at Westinghouse
3D Printing of Components and Coating Applications at Westinghouse Zeses Karoutas Chief Engineer, Fuel Engineering and Safety Analysis MIT Workshop on New Cross-cutting Technologies for Nuclear Power Plants
More informationFAST: The Merger of NRC s Fuel Performance Codes FRAPCON and FRAPTRAN for Scoping and Regulatory Decision Making
FAST: The Merger of NRC s Fuel Performance Codes FRAPCON and FRAPTRAN for Scoping and Regulatory Decision Making Ian E. Porter, Ph.D. United States Nuclear Regulatory Commission (U.S.NRC) Washington, DC,
More informationSafety Analysis Results of Representative DEC Accidental Transients for the ALFRED Reactor
FR13 - TECHNICAL SESSION 3.5: Fast reactor safety: post-fukushima lessons and goals for next-generation reactors Paper n. IAEA-CN-199/260 Safety Analysis Results of Representative DEC Accidental Transients
More informationFuel and material irradiation hosting systems in the Jules Horowitz reactor
Fuel and material irradiation hosting systems in the Jules Horowitz reactor CEA/Cadarache, DEN/DER/SRJH, F-13108 St Paul Lez Durance 14 FÉVRIER 2014 PAGE 1 CONTENTS Fuel and material irradiation hosting
More informationCore Management and Fuel Handling for Research Reactors
Core Management and Fuel Handling for Research Reactors A. M. Shokr Research Reactor Safety Section Division of Nuclear Installation Safety International Atomic Energy Agency Outline Introduction Safety
More informationBehavior of high burnup fuel during LOCA - Key observations and test plan at JAEA -
Behavior of high burnup fuel during LOCA - Key observations and test plan at JAEA - Fumihisa Nagase Japan Atomic Energy Agency IAEA Technical Meeting on Fuel Behaviour and Modelling under Severe Transient
More informationReview Article Fuel R&D Needs and Strategy towards a Revision of Acceptance Criteria
Science and Technology of Nuclear Installations Volume 2010, Article ID 646971, 7 pages doi:10.1155/2010/646971 Review Article Fuel R&D Needs and Strategy towards a Revision of Acceptance Criteria François
More information2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea
NUCLEAR DESIGN AND SAFETY ANALYSIS OF ACCIDENT TOLERANT FUEL CANDIDATES IN OPR1000 Wang-Kee In 1, Ser-Gi Hong 2, Tae-Wan Kim 3, Tae-Hyun Chun 1, Chang-Hwan Shin 1 1 Korea Atomic Energy Research Institute:
More informationEvaluations of Mo-alloy for light water reactor fuel cladding to enhance accident tolerance
EPJ Nuclear Sci. Technol. 2, 5 (2016) B. Cheng et al., published by EDP Sciences, 2016 DOI: 10.1051/epjn/e2015-50060-7 Nuclear Sciences & Technologies Available online at: http://www.epj-n.org REGULAR
More informationThe DENOPI project: a research program on SFP under loss-of-cooling and loss-of-coolant accident conditions
The DENOPI project: a research program on SFP under loss-of-cooling and loss-of-coolant accident conditions NAS meeting March 2015 N. Trégourès, H. Mutelle, C. Duriez, S. Tillard IRSN / Nuclear Safety
More informationAppendix 1: Development of LWR Fuels with Enhanced Accident Tolerance; Task 1 Technical Concept Description
Appendix 1: Development of LWR Fuels with Enhanced Accident Tolerance; Task 1 Technical Concept Description Westinghouse Non-Proprietary Class 3 Award Number DE-NE0000566 Development of LWR Fuels with
More informationModule 06 Boiling Water Reactors (BWR)
Module 06 Boiling Water Reactors (BWR) 1.3.2017 Prof.Dr. Böck Technical University Vienna Atominstitut Stadionallee 2, 1020 Vienna, Austria ph: ++43-1-58801 141368 boeck@ati.ac.at Contents BWR Basics Technical
More informationModule 06 Boiling Water Reactors (BWR)
Module 06 Boiling Water Reactors (BWR) 1.10.2015 Prof.Dr. Böck Vienna University oftechnology Atominstitute Stadionallee 2 A-1020 Vienna, Austria ph: ++43-1-58801 141368 boeck@ati.ac.at Contents BWR Basics
More informationFuel data needs for Posiva s postclosure. B. Pastina (Posiva) IGD-TP 5th Exchange Forum Kalmar
Fuel data needs for Posiva s postclosure safety case B. Pastina (Posiva) IGD-TP 5th Exchange Forum Kalmar 28-29.10.2014 Disposal system at Olkiluoto, Finland TURVA-2012 Safety case report portfolio now
More information2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea
A Parametric Sensitivity Analysis of Nuclear Fuel under RIA with Commercial LWR Conditions Chando Jung 1, Okjoo Kim 2, Jaemyeong Choi 2, Kyuseok Lee 2, Sangwon Park 2 1 KEPCO NF, 242, Daedeok-daero 989beon-gil,
More informationCore Management and Fuel handling for Research Reactors
Core Management and Fuel handling for Research Reactors W. Kennedy, Research Reactor Safety Section Division of Nuclear Installation Safety Yogyakarta, Indonesia 23/09/2013 Outline Introduction Safety
More informationModeling of IFA-409 by Means of TRANSURANUS Code
Modeling of IFA-49 by Means of TRANSURANUS Code Davide ROZZIA 1, Alessandro DEL NEVO 2, Alessandro ARDIZZONE 3, Pietro AGOSTINI 2 1-Dipartimento Ingegneria Meccanica Nucleare e della Produzione, UNIPI
More information2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea
Behavior of High-burnup Advanced LWR Fuels under Design-basis Accident Conditions Masaki Amaya 1, Yutaka Udagawa 1, Takafumi Narukawa 1, Takeshi Mihara 1, Yoshinori Taniguchi 1 1 Nuclear Safety Research
More informationAP1000 European 19. Probabilistic Risk Assessment Design Control Document
19.39 In-Vessel Retention of Molten Core Debris 19.39.1 Introduction In-vessel retention of molten core debris through water cooling of the external surface of the reactor vessel is a severe accident management
More information2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea
NEUTRONIC ANALYSIS OF THE CANDIDATE MULTI-LAYER CLADDING MATERIALS WITH ENHANCED ACCIDENT TOLERANCE FOR VVER REACTORS Ondřej Novák 1, Martin Ševeček 1,2 1 Department of Nuclear Reactors, Faculty of Nuclear
More informationDRAFT PROJECT PLAN TO PREPARE THE U.S. NUCLEAR REGULATORY COMMISSION TO LICENSE AND REGULATE ACCIDENT TOLERANT FUEL
DRAFT PROJECT PLAN TO PREPARE THE U.S. NUCLEAR REGULATORY COMMISSION TO LICENSE AND REGULATE ACCIDENT TOLERANT FUEL The Offices of Nuclear Reactor Regulation (NRR), New Reactors, Nuclear Material Safety
More informationEffects of Pre-Irradiation on Irradiation Growth & Creep of Re-Crystallized Zircaloy-4
Effects of Pre-Irradiation on Irradiation Growth & Creep of Re-Crystallized Zircaloy-4 Margaret A. McGrath 1, Suresh Yagnik 2, Håkon Jenssen 1 1 OECD Halden Reactor Project 2 Electric Power Research Institute
More informationDry storage systems and aging management
Dry storage systems and aging management H.Issard, AREVA TN, France IAEA TM 47934 LESSONS LEARNED IN SPENT FUEL MANAGEMENT Vienna, 8-10 July 2014 AREVA TN Summary Dry storage systems and AREVA Experience
More informationPredictability of CNEA PHWR MOX Experiments by Mean of TRANSURANUS Code, From the IFPE Database. Rozzia D, M. Adorni, A. Del Nevo, F.
Predictability of CNEA PHWR MO Experiments by Mean of TRANSURANUS Code, From the IFPE Database Rozzia D, M. Adorni, A. Del Nevo, F. D Auria University of Pisa Gruppo di Ricerca Nucleare di San Piero a
More informationAEN WPRS Sodium Fast Reactor Core Definitions (version 1.2 September 19 th )
AEN WPRS Sodium Fast Reactor Core Definitions (version 1.2 September 19 th ) David BLANCHET, Laurent BUIRON, Nicolas STAUFF CEA Cadarache Email: laurent.buiron@cea.fr 1. Introduction and main objectives
More informationStudsvik Report. SCIP IV Technical Description. Public. Compiled by Hans-Urs Zwicky DRAFT AS A BASIS FOR DISCUSSION
STUDSVIK/N-18/027 SCIP IV Technical Description Compiled by Hans-Urs Zwicky DRAFT AS A BASIS FOR DISCUSSION Studsvik Report STUDSVIK NUCLEAR AB DRAFT AS A BASIS FOR DISCUSSION STUDSVIK/N-18/027 2018-01-31
More informationOverview of ATF research and ongoing experiments at the Halden reactor project
Overview of ATF research and ongoing experiments at the Halden reactor project Dr. Rudi Van Nieuwenhove Chief Scientist Department Research and Development Sector Nuclear Technology, Physics and Safety
More informationAssessment of the V&V Challenges of Accident Tolerant Fuels
Assessment of the V&V Challenges of Accident Tolerant Fuels Koroush Shirvan Principal Research Scientist Director of MIT ATF IRP Department of Nuclear Science and Engineering Multiphysics Model Validation
More information설계연구실 한전원자력연료. KEPCO NF Proprietary Information 0
2014. 7. 17 설계연구실 한전원자력연료 KEPCO NF Proprietary Information 0 노심설계및안전해석코드개요 노심설계 / 안전해석코드현황및계획 제 1 세대기술도입기 (80 s~90 s) 해외사코드도입 - KWU -CE -WH 제 2 세대기술개량화 ( 99~ 04) 대체코드개발 - 미국정부제한코드 - WH 와공동개발 제 3 세대원천기술확보
More informationCORE COOLABILITY IN LOSS OF COOLANT ACCIDENT: THE COAL EXPERIMENTS
CORE COOLABILITY IN LOSS OF COOLANT ACCIDENT: THE COAL EXPERIMENTS Repetto G., Marquié Ch., Bruyère B. and Glantz T. Institut de Radioprotection et de Sureté Nucléaire, Cadarache, 13115, BP3, Saint Paul
More informationAcceptance Criteria in DBA
IAEA Safety Assessment Education and Training (SAET) Programme Joint ICTP-IAEA Essential Knowledge Workshop on Deterministic Safety Assessment and Engineering Aspects Important to Safety Acceptance Criteria
More informationFuels and Materials Programme Achievements 2012
Institutt for energiteknikk OECD HALDEN REACTOR PROJECT HP-1378 vol. 1 For use within the Halden Project Member Organisations only OECD HALDEN REACTOR PROJECT Fuels and Materials Programme Achievements
More informationOverview of the BISON Multidimensional Fuel Performance Code
www.inl.gov Overview of the BISON Multidimensional Fuel Performance Code Rich Williamson BISON Team Jason Hales, Steve Novascone, Ben Spencer, Danielle Perez, Giovanni Pastore IAEA Technical Meeting: Modeling
More informationTask 3: Licensing Plan for Accident Tolerant Fuel
Westinghouse Non Proprietary Class 3 Award Number DE-NE0000566 Development of LWR Fuels with Enhanced Accident Tolerance Task 3: Licensing Plan for Accident Tolerant Fuel RT TR 13 19 September 30, 2013
More informationCOMPARATIVE STUDY OF TRANSIENT ANALYSIS OF PAKISTAN RESEARCH REACTOR-1 (PARR-1) WITH HIGH DENSITY FUEL
COMPARATIVE STUDY OF TRANSIENT ANALYSIS OF PAKISTAN RESEARCH REACTOR-1 (PARR-1) WITH HIGH DENSITY FUEL M. Iqbal, A. Muhammad, T. Mahmood Nuclear Engineering Division, Pakistan Institute of Nuclear Science
More informationModelling an Unprotected Loss-of-Flow Accident in Research Reactors using the Eureka-2/Rr Code
Journal of Physical Science, Vol. 26(2), 73 87, 2015 Modelling an Unprotected Loss-of-Flow Accident in Research Reactors using the Eureka-2/Rr Code Badrun Nahar Hamid, 1* Md. Altaf Hossen, 1 Sheikh Md.
More informationJ. Stuckert, M. Große, M. Steinbrück
Bundle reflood tests QUENCH-14 and QUENCH-15 with advanced cladding materials: comparable overview J. Stuckert, M. Große, M. Steinbrück Institute for Materials Research KIT University of of the State of
More informationVerification calculations for the WWER version of the TRANSURANUS code. D.Elenkov, St.Boneva, M.Georgieva, St.Georgiev. A. Schubert, P.
Verification calculations for the WWER version of the TRANSURANUS code D.Elenkov, St.Boneva, M.Georgieva, St.Georgiev Institute of Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences, Sofia,
More informationFUMEX 2 IAEA Coordinated Research Programme Nuclear Fuel Cycle and Material Section
FUMEX 2 IAEA Coordinated Research Programme 2002-2006 Nuclear Fuel Cycle and Material Section Purpose Describe the IAEA fuel modelling project Show some of the participants Code Predictions Discuss PCI
More informationLOCA analysis of high temperature reactor cooled and moderated by supercritical light water
GENES4/ANP23, Sep. 15-19, Kyoto, JAPAN Paper 116 LOCA analysis of high temperature reactor cooled and moderated by supercritical light water Yuki Ishiwatari 1*, Yoshiaki Oka 1 and Seiichi Koshizuka 1 1
More informationASTEC Model Development for the Severe Accident Progression in a Generic AP1000-Like
ASTEC Model Development for the Severe Accident Progression in a Generic AP1000-Like Lucas Albright a,b, Dr. Polina Wilhelm b, Dr. Tatjana Jevremovic a,c a Nuclear Engineering Program b Helmholtz-ZentrumDresden-Rossendorf
More informationMIT- CASL EDUCATION ACTIVITIES
MIT- CASL EDUCATION ACTIVITIES CASL Education Team Mujid Kazimi & Koroush Shirvan April 30 2015 Massachusetts Institute of Technology NSE Nuclear Science & Engineering at MIT science : systems : society
More informationCSNI Technical Opinion Papers
Nuclear Safety 2011 CSNI Technical Opinion Papers No. 13 LOCA Criteria Basis and Test Methodology 200 μm Zry-4 ZIRLO N U C L E A R E N E R G Y A G E N C Y Nuclear Safety ISBN 978-92-64-99154-5 NEA/CSNI/R(2011)7
More informationFuels and Materials Programme Achievements 2015
Institutt for energiteknikk OECD HALDEN REACTOR PROJECT HP-1487 vol. 1 For use within the Halden Project Member Organisations only OECD HALDEN REACTOR PROJECT Fuels and Materials Programme Achievements
More informationApplying RISMC Methods, Tools, and Data to Enhance Safety and Economics through Industry Application Demonstrations
Session 2-2: Economics of Plant Life Management Paper # IAEA-CN-246-149 Applying RISMC Methods, Tools, and Data to Enhance Safety and Economics through Industry Application Demonstrations Ronaldo Szilard,
More information