MICROSTRUCTURE CHARACTERIZATION OF ZIRLO STRUCTURAL COMPONENTS IRRADIATED TO HIGH BURNUP
|
|
- Noreen O’Brien’
- 6 years ago
- Views:
Transcription
1 MICROSTRUCTURE CHARACTERIZATION OF ZIRLO STRUCTURAL COMPONENTS IRRADIATED TO HIGH BURNUP J.M. García-Infanta 1, M. Aulló 1, D. Schrire 2, F. Culebras 3 A. M. Garde 4 1 ENUSA Industrias Avanzadas C/ Santiago Rusiñol Madrid (Spain). jmgi@enusa.es 2 Vattenfall Nuclear Fuel AB SE Stockholm (Sweden) 3 Asociación Nuclear Ascó-Vandellós II Edificio Sede L Hospitalet de l Infant 4389 Tarragona (Spain) 4 Westinghouse Electric Company Columbia, SC (USA) 17 th International Symposium on Zirconium in the Nuclear Industry. 3-7 th of February 213, Hyderabad, India
2 OUTLINE OF THE PRESENTATION 1.- Introduction 2.- Experimental scope 3.- Results and discussion - Oxide thickness - Hydriding - Guide tube bulge joints - Grid growth 4.- Conclusions 22
3 INTRODUCTION Objective: Microstructure characterization of ZIRLO samples extracted from the skeleton of assemblies irradiated to usual end-of-life exposures and high burnups. Fuel assembly characteristics and operational information Assembly Id NPP Design Structure material Irradiation Cycles Burnup (GWd/tU) A Ringhals 2 15x15 ZIRLO 4 / annual 54 B Vandellós II 17x17 RFA ZIRLO 3 / eighteen months 52 LTA Vandellós II 17x17 RFA ZIRLO 4 / eighteen months 68 3
4 EXPERIMENTAL SCOPE: ASSEMBLY A Top nozzle Guide tube-top nozzle joint sectioning L1 Top grid T1 T2 Grid 6 Grid 6 IFM 3 Guide tube-top nozzle joint Section T1 3 mm Grid 5 Guide tube IFM 2 Grid 4 IFM 1 Guide tube-grid joint Section T2 Grid 2 Grid 3 Grid 2 Bottom grid P-grid Bottom nozzle Samples from: grid 2, guide tube (between IFM3 and Grid 6), grid 6, guide tube-grid 6 bulge joint and guide tube top nozzle insert bulge joint. Examinations: - Oxide layer thickness (optical microscopy) - Hydrogen content (SEM and hot vacuum extraction) - Metallography for hydrides distribution 4
5 EXPERIMENTAL SCOPE: ASSEMBLY B AND LTA 2.54 mm Top nozzle Top grid Grid 7 Vane extraction Grid 6 Grid 5 Grid 4 Grid 3 Outer strap detail Samples: outer strap grid vanes from grids 5, 6 y 7 (LTA) and grid 7 (assembly B) Grid 2 Examinations: Bottom grid P-grid Bottom nozzle - Oxide layer thickness (optical microscopy) - Hydrogen content (hot vacuum extraction) - Metallography for hydrides distribution - Grid growth (on-site LVDT system) 5
6 EXPARIMENTAL SCOPE: TEMPERATURE AND NEUTRON FLUX Top nozzle Fuel column Neutron flux Temperature Tout Top grid Grid 7 Grid 6 Grid 5 Grid 4 Grid 3 Grid 2 Bottom grid P-grid Bottom nozzle Tin 6
7 One-sided oxide thickness (mm) RESULTS AND DISCUSSION: OXIDE THICKNESS ASSEMBLY A AND B Previous data guide thimble Previous data strap Previous data welds/haz A Guide thimble end piece A Guide thimble at grid 6 A Bulge at grid 6. Guide thimble A Grid 6 outer strap A Grid 2 outer strap A Grid 6 inner strap A Grid 6 dimple A Grid 6 spring A Grid 6 welds/hazs B Grid 7 outer strap vane Elevation (mm) - Consistency with previous experience at equivalent exposures - Maximum values at upper ZIRLO grid elevation: Guide tube 2 mm; Grid strap 3mm; grid HAZ/welds 5 mm - No significant differences between grid strap and cold deformed shapes (dimple/springs) - Guide tube top nozzle bulge joint 1 mm, out of the fuel column. No crevice or galvanic corrosion effects 7
8 One-sided oxide thickness (mm) RESULTS AND DISCUSSION: OXIDE THICKNESS LTA GRID VANES 6 5 Previous data 17x17 ZIRLO grid strap A B LTA grid 6 grid 7 4 grid 5 grid Elevation (mm) - Maximum values at grid 6 elevation 5 mm; - Comparison with Assembly B upper ZIRLO grid (grid 7) oxide thickness 5% higher. Acceleration of oxidation during the 4th LTA cycle. 8
9 One-sided oxide thickness (mm) RESULTS AND DISCUSSION: OXIDE THICKNESS LTA GRID VANES Previous data 17x17 ZIRLO grid strap A B LTA grid 5 grid 6 grid 7 grid 7 5% Elevation (mm) - Maximum values at grid 6 elevation 5 mm; - Comparison with Assembly B upper ZIRLO grid (grid 7) oxide thickness 5% higher. Acceleration of oxidation during the 4th LTA cycle. 9
10 [H] (ppm) RESULTS AND DISCUSSION: HYDROGEN CONCENTRATION IN GUIDE TUBE Previous data 17x17 ZIRLO guide tube Assembly A grid End joint Elevation (mm) - Consistency with previous experience at equivalent exposures at grid 6 elevation. - Maximum value at grid 6 is 6 ppm. - Guide tube top nozzle bulge joint is 2 ppm, out of the fuel column 1
11 [H] (ppm) RESULTS AND DISCUSSION: HYDROGEN CONCENTRATION IN GRID STRAP 1 8 Previous data 17x17 ZIRLO grid strap Assembly A outer strap Assembly B outer strap vanes LTA outer strap vanes LTA grid 7 6 LTA grid 6 B grid 7 4 LTA grid 5 A grid 6 2 A grid Elevation (mm) - Assembly A and B consistent with previous experience at equivalent exposures at upper ZIRLO grid elevation - LTA maximum [H] value is 9 ppm at upper ZIRLO grid elevation. - Comparing LTA with assembly B at upper ZIRLO grid, [H] is 8% higher. Acceleration of H pick-up during the 4th LTA cycle. 11
12 [H] (ppm) RESULTS AND DISCUSSION: HYDROGEN CONCENTRATION IN GRID STRAP Previous data 17x17 ZIRLO grid strap Assembly A outer strap Assembly B outer strap vanes LTA outer strap vanes A grid 2 LTA grid 5 LTA grid 6 B grid 7 LTA grid 7 8% A grid Elevation (mm) - Assembly A and B consistent with previous experience at equivalent exposures at upper ZIRLO grid elevation - LTA maximum [H] value is 9 ppm at upper ZIRLO grid elevation. - Comparing LTA with assembly B at upper ZIRLO grid, [H] is 8% higher. Acceleration of H pick-up during the 4th LTA cycle. 12
13 HPF (%) RESULTS AND DISCUSSION: HYDROGEN PICK-UP FRACTION HPF ZIRLO database High Burnup Assembly (US plant) Assembly A Assembly B LTA LTA grid 6 LTA grid 7 A grid 7 15 LTA grid 5 B grid Temperature (ºF) - Assembly A and B consistent with previous experience at equivalent exposures at upper ZIRLO grid elevation - LTA HPF values are above the previous experience at grid 6 and 7. - LTA show higher HPF than assembly B at upper ZIRLO grid. Acceleration of H pick-up during the 4th LTA cycle. 13
14 RESULTS AND DISCUSSION: HYDRIDE DISTRIBUTION IN ASSEMBLY A Guide tube above grid 6 cross (C) and longitudinal (L) sections [H] ~ 6 ppm Top nozzle Grid 2 and Grid 6 cross sections Grid 6 [H] ~ 4 ppm Grid 2 [H] ~ 15 ppm C-section L-section Top grid Grid 6 IFM 3 Grid 5 IFM 2 Grid 4 IFM 1 Grid 3 Grid 2 Bottom grid P-grid Bottom nozzle Plane of view 14
15 RESULTS AND DISCUSSION: HYDRIDES DISTRIBUTION IN ASSEMBLY B AND LTA GRID STRAP VANES LTA Assembly B Top nozzle 2 mm 918 ppm 2 mm 57 ppm Top grid Grid 7 Grid 6 Grid 5 2 mm 713 ppm Grid 4 Grid 3 Plane of view Grid 2 2 mm 414 ppm Bottom grid P-grid Bottom nozzle 15
16 Avg. 1-sided oxide thickness (mm) Avg. 1-sided oxide thickness (mm) 27º 9º 9º Avg. 1-sided oxide thickness (mm) 27º Avg. 1-sided oxide thickness (mm) RESULTS AND DISCUSSION: GUIDE TUBE BULGE JOINTS. OXIDE THICKNESS a) 315º º T2 45º º 18º 135º Radial position (º) Guide tube to top nozzle insert bulge joint oxide thickness ~1 mm Guide tube-top nozzle joint sectioning T1 T2 L1 Section T1 b) 225º 315º º 18º c) T º 135º Radial position (º) Section T2 3 4 L Axial position (º) Guide tube to grid 6 bulge joint d) Sleeve (outer tube) 5 Guide tube Axial position (º) 16
17 RESULTS AND DISCUSSION: GUIDE TUBE - TOP NOZZLE BULGE JOINT HYDRIDES DISTRIBUTION [H]<2 ppm. Low in comparison with the guide tube at grid 6 elevation. Reason, it is located above the fuel column. - Some radial and 45º oriented hydrides in the plastically deformed areas - Normal hydrides distribution in the non plastically deformed areas 17
18 Grid growth (%) RESULTS AND DISCUSSION: GRID GROWTH CONTRIBUTIONS.8 = irr + [H] + creep.7 LTA grid LTA grid 5 LTA grid 6 irr ~.1% [H] ~.3% creep ~.3%? [H] (ppm) More research is necessary to know exact mechanisms grid growth 18
19 CONCLUSIONS Oxide thickness, hydrogen concentration and hydrogen pick up fraction depend on the axial elevation of the piece as an effect of the coolant temperature rise. Results from LTA show an acceleration of the oxidation reaction and hydrogen pickup, and it is attributed excessive microstructure degradation due high burnup. Good performance of guide tube bulge joints. No galvanic or crevice corrosion observed. Estimation of the different contributors to grid growth reveals that corrosion of the grid straps has a major effect at highest elevations. 19
20 THANK YOU VERY MUCH FOR YOUR ATENTION QUESTIONS? 2
2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea
ASSESSMENT OF THE INTEGRITY OF THE FUEL ROD WITH SPALLED OXIDE UNDER HYPOTHETICAL TRANSPORTATION ACCIDENTS Alfonso Ascarza 1, Alberto Cerracín 1, Jorge Muñoz 1, Leo Carrilho 2, Guirong Pan 2. 1 Technology
More informationA Comparative Analysis of CABRI CIP0-1 and NSRR VA-2 Reactivity Initiated Accident tests
A Comparative Analysis of CABRI CIP-1 and NSRR VA-2 Reactivity Initiated Accident tests M. PETIT*, V. GEORGENTHUM*, T. SUGIYAMA**, M. QUECEDO***, J. DESQUINES* * IRSN, DPAM/SEMCA, BP 3, 13115 Saint-Paul-lez-Durance
More informationEVOLUTION OF HYDROGEN PICKUP FRACTION WITH OXIDATION RATE ON ZIRCONIUM ALLOYS ABSTRACT
Westinghouse Non-Proprietary Class 3 EVOLUTION OF HYDROGEN PICKUP FRACTION WITH OXIDATION RATE ON ZIRCONIUM ALLOYS J. ROMERO 1, J. PARTEZANA 2, R. J. COMSTOCK 2, L. HALLSTADIUS 3, A. MOTTA 4, A. COUET
More informationIRRADIATION TEST RESULTS OF HANA CLADDING IN HALDEN TEST REACTOR AFTER 67 GWD/MTU
IRRADIATION TEST RESULTS OF HANA CLADDING IN HALDEN TEST REACTOR AFTER 67 GWD/MTU HYUN-GIL KIM, JEONG-YONG PARK, YANG-IL JUNG, DONG-JUN PARK, YANG-HYUN KOO LWR Fuel Technology Division, Korea Atomic Energy
More informationWM2014 Conference, March 2 6, 2014, Phoenix, Arizona, USA
Integrity Study of Spent PWR Fuel under Dry Storage Conditions 14236 Jongwon Choi *, Young-Chul Choi *, Dong-Hak Kook * * Korea Atomic Energy Research Institute ABSTRACT Technical issues related to long-term
More informationMICROSTRUCTURAL STUDIES OF HEAT-TREATED Zr-2.5Nb ALLOY FOR PRESSURE TUBE APPLICATIONS. N. Saibaba Nuclear Fuel Complex, Hyderabad, INDIA
MICROSTRUCTURAL STUDIES OF HEAT-TREATED Zr-2.5Nb ALLOY FOR PRESSURE TUBE APPLICATIONS N. Saibaba Nuclear Fuel Complex, Hyderabad, INDIA OUTLINE Introduction Objective Background Optimization of Quenching
More informationTHE EFFECTS OF CREEP AND HYDRIDE ON SPENT FUEL INTEGRITY DURING INTERIM DRY STORAGE
THE EFFECTS OF CREEP AND HYDRIDE ON SPENT FUEL INTEGRITY DURING INTERIM DRY STORAGE HYUN-GIL KIM *, YONG-HWAN JEONG and KYU-TAE KIM 1 Nuclear Convergence Technology Division, Korea Atomic Energy Research
More informationThe Effects of Microstructure and Operating Conditions on Irradiation
The Effects of Microstructure and Operating Conditions on Irradiation Creep of Zr Zr-2.5Nb 2 5Nb Pressure Tubing 17th International Symposium on Zirconium in the Nuclear Industry L.Walters, G.Bickel and
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 informationWestinghouse SMR & Nuclear Fuel Overview
Westinghouse SMR & Nuclear Fuel Overview Carlos Leipner, Westinghouse VP Latin America Presented at LAS-ANS July 2014 Rio de Janeiro, Brasil 2014 Westinghouse Electric Company LLC, All Rights Reserved.
More information2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea
ZIRCALOY-2 CORROSION AND HYDROGEN PICKUP NEAR BWR CORE INLET David Schrire 1, Erik Mader 2, Aylin Kucuk 3, Ron Adamson 4 1 Vattenfall Nuclear Fuel, SE-16287 Stockholm, Sweden, Email: david.schrire@vattenfall.com
More informationPreliminary Irradiation Effect on Corrosion Resistance of Zirconium Alloys
A.A. BOCHVAR HIGH-TECHNOLOGY RESEARCH INSTITUTE OF INORGANIC MATERIALS (SC «VNIINM») 18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY MAY 15-19, 2016 «ROSATOM» STATE ATOMIC ENERGY CORPORATION
More informationChapter 22. Waterside Corrosion and Hydriding of Zr Alloy Cladding
Chapter 22. Waterside Corrosion and Hydriding of Zr Alloy Cladding 22.1 Introduction... 2 22.2 Influence of Alloying Additions on Zirconium Alloy Corrosion... 3 22.3 Uniform Corrosion Mechanism and Oxide
More informationIN-PILE PERFORMANCE OF HANA CLADDING TESTED IN HALDEN REACTOR
http://dx.doi.org/10.5516/net.07.2013.093 INPILE PERFORMANCE OF HANA CLADDING TESTED IN HALDEN REACTOR HYUNGIL KIM 1*, JEONGYONG PARK 1, YONGHWAN JEONG 1, YANGHYUN KOO 1, JONGSUNG YOO 2, YONGKYOON MOK
More informationThe Pennsylvania State University. The Graduate School THE EFFECT OF HYDROGEN ON THE DEFORMATION BEHAVIOR OF ZIRCALOY-4.
The Pennsylvania State University The Graduate School Department of Mechanical and Nuclear Engineering THE EFFECT OF HYDROGEN ON THE DEFORMATION BEHAVIOR OF ZIRCALOY-4 A Thesis in Nuclear Engineering by
More informationImprovement of economic efficiency of the fuel usage in NPP. The new types of nuclear fuel and the fuel cycles
STATE ATOMIC ENERGY CORPORATION ROSATOM Improvement of economic efficiency of the fuel usage in NPP. The new types of nuclear fuel and the fuel cycles Petr Lavrenyuk (JSC «TVEL») Eighth International Forum
More informationEffect of Alloying Elements, Cold Work, and Hydrogen on the Irradiation Growth Behavior of Zirconium Alloy Variants
Effect of Alloying Elements, Cold Work, and Hydrogen on the Irradiation Growth Behavior of Zirconium Alloy Variants Acknowledgments Work performed under auspices of NFIR Program (2005-11) Coauthors: Yagnik,
More informationAssessment of Aging of Zr-2.5Nb Pressure Tubes for Use in Heavy Water Reactor
Assessment of Aging of Zr-2.5Nb Pressure Tubes for Use in Heavy Water Reactor Ahmad Hussain, Dheya Al-Othmany Department of Nuclear Engineering, Faculty of Engineering, King Abdulaziz University, P.O.
More informationSimulating Pellet and Clad Mechanical Interactions of Nuclear Fuel Rod for Pressure Water Reactors
14 th International LS-DYNA Users Conference Session: Simulation Simulating Pellet and Clad Mechanical Interactions of Nuclear Fuel Rod for Pressure Water Reactors W. Zhao, D. Mitchell, R. Oelrich Westinghouse
More informationDevelopment of Radiation Resistant Reactor Core Structural Materials
Development of Radiation Resistant Reactor Core Structural Materials A. Introduction 1. The core of a nuclear reactor is where the fuel is located and where nuclear fission reactions take place. The materials
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 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 informationFracture of Zircaloy-4 fuel cladding tubes with hydride blisters
Fracture of Zircaloy-4 fuel cladding tubes with hydride blisters Vincent Macdonald, David Le Boulch, Arthur Hellouin de Menibus, Jacques Besson, Quentin Auzoux, Jérôme Crépin, Thomas Le Jolu To cite this
More informationCryogenic Pipe Support Systems
Cryogenic Pipe Support Systems Overview Rilco cryogenic pipe supports are developed to support cold piping in various applications ranging from chilled water to liquefied natural gas. The qualified service
More informationHigh Temperature Oxidation of Zr-2.5%wt Nb Alloys Doped with Yttrium
Journal of Materials Science and Engineering A 5 (3-4) (215) 154-158 doi: 1.17265/2161-6213/215.3-4.7 D DAVID PUBLISHING High Temperature Oxidation of Zr-2.5%wt Nb Alloys Doped with Yttrium Djoko Hadi
More informationHigh Energy Piping Inspection Program for Power Generation and Process Industries. For Regulatory Compliance, Safety, and System Longevity
High Energy Piping Inspection Program for Power Generation and Process Industries For Regulatory Compliance, Safety, and System Longevity EAPC Industrial Services offers an inspection plan for assessment
More informationBatch Annealing Model for Cold Rolled Coils and Its Application
China Steel Technical Report, No. 28, pp.13-20, (2015) Chun-Jen Fang and Li-Wen Wu 13 Batch Annealing Model for Cold Rolled Coils and Its Application CHUN-JEN FANG and LI-WEN WU New Materials Research
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 informationBARC/2011/E/016 BARC/2011/E/016
BARC/2011/E/016 BARC/2011/E/016 INFLUENCE OF DEUTERIUM CONTENT ON TENSILE BEHAVIOR OF Zr-2.5Nb PRESSURE TUBE MATERIAL IN THE TEMPERATURE RANGE OF AMBIENT TO 300 C by A.K. Bind, R.N. Singh and J.K. Chakravartty
More informationOXIDATIVE RESISTANCE OF SULFONE POLYMERS TO CHLORINATED POTABLE WATER
OXIDATIVE RESISTANCE OF SULFONE POLYMERS TO CHLORINATED POTABLE WATER S. Chung, J. Couch, J.D. Kim, K. Oliphant and P. Vibien, Jana Laboratories Inc., 280B Industrial Parkway S., Aurora ON, Canada and
More informationKIPT ADS Facility. Yousry Gohar 1, Igor Bolshinsky 2, Ivan Karnaukhov 3. Argonne National Laboratory, USA 2. Idaho National Laboratory, USA 3
KIPT ADS Facility Yousry Gohar 1, Igor Bolshinsky 2, Ivan Karnaukhov 3 1 Argonne National Laboratory, USA 2 Idaho National Laboratory, USA 3 Kharkov Institute of Physics & Technology, Ukraine EuCARD 2
More informationIn-plane testing of precast concrete wall panels with grouted sleeve
In-plane testing of precast concrete wall panels with grouted sleeve P. Seifi, R.S. Henry & J.M. Ingham Department of Civil Engineering, University of Auckland, Auckland. 2017 NZSEE Conference ABSTRACT:
More informationsevere accident progression in the BWR lower plenum and the modes of vessel failure
1 For Presentation at the ERMSAR Conference held in Marseilles, France, March 24-26, 2015 severe accident progression in the BWR lower plenum and the modes of vessel failure B. R. Sehgal S. Bechta Nuclear
More informationEffect of Nb on hydride embrittlement of Zr-xNb alloys
Effect of Nb on hydride embrittlement of Zr-xNb alloys Seungjin Oh 1, Changheui Jang 1*, Jun Hwan Kim 2, Yong Hwan Jeong 2 1 Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science
More informationReactor Technology --- Materials, Fuel and Safety
Reactor Technology --- Materials, Fuel and Safety UCT EEE4101F / EEE4103F April 2015 Emeritus Professor David Aschman Based on lectures by Dr Tony Williams Beznau NPP, Switzerland, 2 x 365 MWe Westinghouse,
More informationLEU Conversion of the University of Wisconsin Nuclear Reactor
LEU Conversion of the University of Wisconsin Nuclear Reactor Paul Wilson U. Wisconsin-Madison Russian-American Symposium on the Conversion of Research Reactors to Low Enriched Uranium Fuel 8-10 June 2011
More informationVALIDATION PROCEDURE FOR THE FABRICATION SEQUENCE OF SELF-POWERED NEUTRON DETECTORS FOR NPP DEVELOPED BY CNEA.
VALIDATION PROCEDURE FOR THE FABRICATION SEQUENCE OF SELF-POWERED NEUTRON DETECTORS FOR NPP DEVELOPED BY CNEA. Marcelo Miller, Abraham Banchik, Daniel Bianchi, Alejandra Flores, Adolfo Marajofsky, Lidia
More informationSpecification for Phase VII Benchmark
Specification for Phase VII Benchmark UO 2 Fuel: Study of spent fuel compositions for long-term disposal John C. Wagner and Georgeta Radulescu (ORNL, USA) November, 2008 1. Introduction The concept of
More informationKIPT ACCELERATOR DRIVEN SYSTEM DESIGN AND PERFORMANCE
KIPT ACCELERATOR DRIVEN SYSTEM DESIGN AND PERFORMANCE Yousry Gohar 1, Igor Bolshinsky 2, Ivan Karnaukhov 3 1 Argonne National Laboratory, USA 2 Idaho National Laboratory, USA 3 Kharkov Institute of Physics
More informationThe pressure tube inspection and integrity evaluation in Fugen
GENES4/ANP2003, Sep. 15-19, 2003, Kyoto, JAPAN Paper 1151 The pressure tube inspection and integrity evaluation in Fugen Nobuo Ishizuka 1), Kouzou Nakai 2), Koichi Takayama 3) Fugen Nuclear Power Station
More informationMetal Transitions HIGH TEMPERATURE
Metal Transitions High Temperature Thermal Solutions of Texas continues to meet the demands of technological advances by developing thermocouples using materials with unusually high performance characteristics
More informationEnglish text only NUCLEAR ENERGY AGENCY COMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATIONS
Unclassified NEA/CSNI/R(2011)10 NEA/CSNI/R(2011)10 Unclassified Organisation de Coopération et de Développement Économiques Organisation for Economic Co-operation and Development 19-Jan-2012 English text
More informationApplication of Coating Technology on the Zirconium-Based Alloy to Decrease High-Temperature Oxidation
Application of Coating Technology on the Zirconium-Based Alloy to Decrease High-Temperature Oxidation Hyun-Gil Kim*, Il-Hyun Kim, Jeong-Yong Park, Yang-Hyun Koo, KAERI, 989-111 Daedeok-daero, Yuseong-gu,
More informationJean Paul MARDON 1, Nesrine GHARBI 2, Thomas JOURDAN 3, Didier GILBON 4, Fabien ONIMUS 2, Xavier FEAUGEAS 5, Rosmarie HENGSTLER-EGER 6
EPJ Web of Conferences 115, 02006 (2016) DOI: 10.1051/epjconf/201611502006 Owned by the authors, published by EDP Sciences, 2016 2 nd Int. Workshop Irradiation of Nuclear Materials: Flux and Dose Effects
More informationSTUDY ON IN-VESSEL RETENTION (IVR) STRATEGY FOR CPR1000
NURETH14-297 STUDY ON IN-VESSEL RETENTION (IVR) STRATEGY FOR CPR1000 X. Chen, H. Zhang, J. Zhang, S. Zhang and J. Lin China Nuclear Power Technology Research Institute (CNPRI), Shenzhen, China Abstract
More informationSTRESS CORROSION CRACKING OF STAINLESS STEELS IN HIGH PRESSURE ALKALINE ELECTROLYSERS
STRESS CORROSION CRACKING OF STAINLESS STEELS IN HIGH PRESSURE ALKALINE ELECTROLYSERS Haraldsen, K. 1 and Leth-Olsen, H. 1 1 Corporate Research Centre, Norsk Hydro ASA, P.O.Box 2560, 3908 Porsgrunn, Norway
More informationA Review of Suitability for PWHT Exemption Requirements in the Aspect of Residual Stresses and Microstructures
Transactions, SMiRT-23 Division IX, Paper ID 612 (inc. assigned division number from I to X) A Review of Suitability for PWHT Exemption Requirements in the Aspect of Residual Stresses and Microstructures
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 informationRESEARCH AND DEVELOPMENT OF COATINGS FOR ZIRCONIUM FUEL CLADDINGS
RESEARCH AND DEVELOPMENT OF COATINGS FOR ZIRCONIUM FUEL CLADDINGS V.А. Belous, V.N. Voyevodin, А.S. Kuprin *, V.D. Ovcharenko, R.L. Vasilenko, V.S. Krasnorutskyy, V.A. Zuyok, Ya.A. Kushtym National Science
More informationVARIATION IN SHAPE DISTORTION DUE TO CORNER THINNING/THICKENING OF PREPREG
VARIATION IN SHAPE DISTORTION DUE TO CORNER THINNING/THICKENING OF PREPREG JM Svanberg a,*, P Hallander b and T Nyman b a Swerea SICOMP AB, Box 271, 941 26 Piteå, Sweden b Saab Aerostructures, SE-581 88
More informationDatabase Enhancements for Improved AREVA NP LWR Deposition Model BG Lockamon 11/17/ p.1
Database Enhancements for Improved AREVA NP LWR Deposition Model BG Lockamon 11/17/2010 - p.1 Database Enhancements for Improved AREVA NP LWR Deposition Model Brian G. Lockamon AREVA NP Inc., Plant Chemistry
More informationRisks of Nuclear Ageing
Risks of Nuclear Ageing Technical characteristics of ageing processes and their possible impacts on nuclear safety in Spain S. Mohr, S. Kurth Greenpeace, Valencia, November 2014 Age of European NPPs (grid
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 informationApplication of CANDLE Burnup to Block-Type High Temperature Gas Cooled Reactor for Incinerating Weapon Grade Plutonium
GENES4/ANP2003, Sep. 15-19, 2003, Kyoto, JAPAN Paper 1079 Application of CANDLE Burnup to Block-Type High Temperature Gas Cooled Reactor for Incinerating Weapon Grade Plutonium Yasunori Ohoka * and Hiroshi
More informationExperimental Investigation of the Galvanization. Effects on the Properties of. Low Carbon Alloy Steel
Adv. Theor. Appl. Mech., Vol. 5, 2012, no. 5, 225-236 Experimental Investigation of the Galvanization Effects on the Properties of Low Carbon Alloy Steel Ghazi S. Al-Marahleh Department of Mechanical Engineering,
More informationThermal ageing of nickel-base Alloy 690 TT
SAFIR2018 - The Finnish Research Programme on Nuclear Power Plant Safety 2015-2018 RG5 Structural Integrity: THELMA (Thermal Ageing of Materials) one topic in the project: Thermal ageing of nickel-base
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 informationSpecification for Phase IID Benchmark. A. BARREAU (CEA, France) J. GULLIFORD (BNFL, UK) J.C. WAGNER (ORNL, USA)
Specification for Phase IID Benchmark PWR-UO 2 Assembly: Study of control rods effects on spent fuel composition A. BARREAU (CEA, France) J. GULLIFORD (BNFL, UK) J.C. WAGNER (ORNL, USA) 1. Introduction
More informationNUCLEAR. Nuclear Reactor Pressure Vessel Seals
NUCLEAR Nuclear Reactor Pressure Vessel Seals SEALING CONCEPT Technetics Group is the world s leading manufacturer of Nuclear Reactor Pressure Vessel (RPV) Closure Head Seals. In addition, Technetics Group
More informationManufacture of Nuclear Fuel Elements in Chile Fabricación de Combustible Nuclear en Chile
Manufacture of Nuclear Fuel Elements in Chile Fabricación de Combustible Nuclear en Chile Luis Olivares, Jorge Marin, Jaime Lisboa, Eric Alcorta Nuclear Fuel Section Chilean Nuclear Energy Commission Workshop
More informationWM2015 Conference, March 15-19, 2015, Phoenix, Arizona, USA
An Integrated Equipment for Massive Segmentation and Packaging of Control Rod Guide Tubes 15161 Joseph Boucau*, Patrick Gobert** Sébastien Bonne *** * Westinghouse Electric Company, 43 rue de l Industrie,
More informationLONG TERM THERMAL EXPOSURE OF HAYNES 282 ALLOY
LONG TERM THERMAL EXPOSURE OF HAYNES 282 ALLOY L. M. Pike Haynes International 12 West Park Ave.; Kokomo, IN, 4694-913 USA Keywords: 282, thermal stability, R-41, Waspaloy, 263 Abstract HAYNES 282 was
More informationJoint ICTP-IAEA School of Nuclear Energy Management August 2011
2257-23 Joint ICTP-IAEA School of Nuclear Energy Management 8-26 August 2011 Nuclear Applications Fundamentals: Materials for fission and fusion technology Danas Ridikas IAEA, Vienna Austria Lecture 3
More information9. VACUUM TANK 9. VACUUM TANK
9. VACUUM TANK This system constitutes the external part of the solenoid cryostat. The vacuum tank, made of stainless steel, is cantilevered from the central ring of the barrel yoke. It houses and supports
More informationWeld Distortion Control Methods and Applications of Weld Modeling
Weld Distortion Control Methods and Applications of Weld Modeling F. W. Brust Paul Scott ABSTRACT A virtual fabrication technology (VFT) modeling procedure is introduced in this paper. It is a state-of-theart
More informationCorrosion Characteristics of PT-7M and PT-3V Titanium Alloys in Ammonia Water Chemistry
Proceedings of the Korean Nuclear Society Spring Meeting Cheju, Korea, May 2001 Corrosion Characteristics of PT-7M and PT-3V Titanium Alloys in Ammonia Water Chemistry Byoung-Kwon Choi, Tae-Kyu Kim, Yong-Hwan
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 international program Phebus FP (fission
1The safety of nuclear reactors 1 6 Results of initial Phebus FP tests FPT-0 and FPT-1 S. BOURDON (IRSN) D. JACQUEMAIN (IRSN) R. ZEYEN (JRC/PETTEN) The international program Phebus FP (fission products)
More informationIrradiation Assisted Stress Corrosion Cracking. By Topan Setiadipura [09M51695] (Obara Lab., Nuclear Engineering Dept., Tokyo Tech.
Introduction Short Review on Irradiation Assisted Stress Corrosion Cracking By Topan Setiadipura [09M51695] (Obara Lab., Nuclear Engineering Dept., Tokyo Tech.) Irradiation-assisted stress-corrosion cracking
More informationFailure Analysis of Cracked Reducer Flange
J Fail. Anal. and Preven. (2010) 10:480 485 DOI 10.1007/s11668-010-9389-9 Failure Analysis of Cracked Reducer Flange Khaled Habib Submitted: 18 May 2010/in revised form: 29 July 2010/Published online:
More informationMechanisms of Hydride Reorientation in Zircaloy-4 Studied in Situ
ZIRCONIUM IN THE NUCLEAR INDUSTRY: 17TH INTERNATIONAL SYMPOSIUM 1107 STP 1543, 2014 / available online at www.astm.org / doi: 10.1520/STP154320120168 Kimberly Colas, 1 Arthur Motta, 2 Mark R. Daymond,
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 informationATOM-PROBE ANALYSIS OF ZIRCALOY
ATOM-PROBE ANALYSIS OF ZIRCALOY H. Andren, L. Mattsson, U. Rolander To cite this version: H. Andren, L. Mattsson, U. Rolander. ATOM-PROBE ANALYSIS OF ZIRCALOY. Journal de Physique Colloques, 1986, 47 (C2),
More informationKAPROS-E: A Modular Program System for Nuclear Reactor Analysis, Status and Results of Selected Applications.
KAPROS-E: A Modular Program System for Nuclear Reactor Analysis, Status and Results of Selected Applications. C.H.M. Broeders, R. Dagan, V. Sanchez, A. Travleev Forschungszentrum Karlsruhe Institut für
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 informationPTFE Pipe Slide Assemblies
PTFE Slide Assemblies Overview Application Anvil PTFE pipe slide assemblies are designed to support the pipe and provide for lateral and axial movement due to thermal expansion and contraction of the piping
More information2017 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 informationFull MOX Core Design in ABWR
GENES4/ANP3, Sep. -9, 3, Kyoto, JAPAN Paper 8 Full MOX Core Design in ABWR Toshiteru Ihara *, Takaaki Mochida, Sadayuki Izutsu 3 and Shingo Fujimaki 3 Nuclear Power Department, Electric Power Development
More informationWorkgroup Thermohydraulics. The thermohydraulic laboratory
Faculty of Mechanical Science and Engineering Institute of Power Engineering Professorship of Nuclear Energy and Hydrogen Technology Workgroup Thermohydraulics The thermohydraulic laboratory Dr.-Ing. Christoph
More informationZRO 2 AND UO 2 DISSOLUTION BY MOLTEN ZIRCALLOY
International Conference Nuclear Energy for New Europe 2002 Kranjska Gora, Slovenia, September 9-12, 2002 www.drustvo-js.si/gora2002 ZRO 2 AND UO 2 DISSOLUTION BY MOLTEN ZIRCALLOY J. Stuckert, A. Miassoedov,
More informationINVESTIGATION OF VOID REACTIVITY BEHAVIOUR IN RBMK REACTORS
INVESTIGATION OF VOID REACTIVITY BEHAVIOUR IN RBMK REACTORS M. Clemente a, S. Langenbuch a, P. Kusnetzov b, I. Stenbock b a) Gesellschaft für Anlagen- und Reaktorsicherheit (GRS)mbH, Garching, E-mail:
More informationTitanium Production Tubing for HPHT Oil & Gas Wells. Jonathan Parry Chevron Jim Grauman TIMET
Titanium Production Tubing for HPHT Oil & Gas Wells Jonathan Parry Chevron Jim Grauman TIMET Overview Titanium alloy seamless pipe manufacturing & application Titanium advantages over high chrome/nickel
More informationTi 6Al-4V is recommended for use at service temperatures up to approximately 350 C (660 F).
Titanium Alloy Ti 6Al-4V Type Analysis Carbon (Maximum) 0.10 % Titanium Balance Aluminum 5.50 to 6.75 % Vanadium 3.50 to 4.50 % Nitrogen 0.05 % Iron (Maximum) 0.40 % Oxygen (Maximum) 0.020 % Hydrogen (Maximum)
More informationMASTERBATCHES. Masterbatch Application and Selection Guide for Irrigation Pipes
MASTERBATCHES Masterbatch Application and Selection Guide for Irrigation Pipes Table of contents Page Nr Introduction 2 Definitions 3 Key Performance Requirements of Irrigation Pipes 3 Drip Irrigation
More informationWWER Fuel Performance and Modelling Activities in Bulgaria
WWER Fuel Performance and Modelling Activities in Bulgaria Maria Manolova INRNEBAS, Sofia, Bulgaria manolova@inrne.bas.bg TWGFTP Meeting Vienna, IAEA, 2729 29 April 2010 Intermediate Meeting of the TWGFTP,
More informationBURNER FLAME TEMPERATURE DURING WARM UP AND HOT STANDBY. Alan D. Mosher KPS Technology & Engineering LLC
BURNER FLAME TEMPERATURE DURING WARM UP AND HOT STANDBY Alan D. Mosher KPS Technology & Engineering LLC Presented at the Brimstone Sulfur Symposium Facilitated by Brimstone STS Limited Vail Colorado September
More informationMetal Powder - the Raw Material of Future Production
Metal Powder - the Raw Material of Future Production BY GÜNTER BUSCH* SYNOPSIS Alongside Mobile Internet, Cloud Computing, Robotics, Energy Storage and Autonomous Vehicles, Additive Manufacturing is one
More informationFENDL NEUTRONICS BENCHMARK: NEUTRON MULTIPLICATION MEASUREMENTS IN BERYLLIUM, BERYLLIUM OXIDE AND LEAD WITH 14-MEV NEUTRONS
fflfornohonoi Atomic cn^oy AQGHCY INDCfNDSl-314 Distrib.: G+F I N DC INTERNATIONAL NUCLEAR DATA COMMITTEE FENDL NEUTRONICS BENCHMARK: NEUTRON MULTIPLICATION MEASUREMENTS IN BERYLLIUM, BERYLLIUM OIDE AND
More informationFRAPCON-4.0: Integral Assessment
PNNL-19418, Vol.2 Rev.2 FRAPCON-4.0: Integral Assessment September 2015 KJ Geelhood WG Luscher Prepared for the U.S. Department of Energy under Contract DE-AC05-76RL01830 PNNL-19418, Vol. 2 Rev.2 FRAPCON-4.0
More informationRESEARCH REACTOR FRJ-1 (MERLIN) THE CORE STRUCTURES OF THE REACTOR BLOCK ARE DISMANTLED
RESEARCH REACTOR FRJ-1 (MERLIN) THE CORE STRUCTURES OF THE REACTOR BLOCK ARE DISMANTLED B. Stahn, R. Printz, K. Matela, C. Zehbe Forschungszentrum Jülich GmbH 52425 Jülich, Germany J. Pöppinghaus Gesellschaft
More informationA Systematic Study to Determine the Remaining Life of a 60 year old Westinghouse-design Steam Chest
A Systematic Study to Determine the Remaining Life of a 60 year old Westinghouse-design Steam Chest Sazzadur Rahman, Ph.D. Waheed Abbasi, Ph.D. Thomas W. Joyce Siemens Energy, Inc., 4400 Alafaya Trail,
More informationNew Developments in Measuring Low Silicon in Process Piping Using Handheld X-ray Fluorescence (XRF)
New Developments in Measuring Low Silicon in Process Piping Using Handheld X-ray Fluorescence (XRF) Presented by: Mark Lessard Business Development Manager Thermo Fisher Scientific mark.lessard@thermofisher.com
More informationAward Number DE-NE
Appendix 2: Development of LWR Fuels with Enhanced Accident Tolerance; Task 2 Description of Research & Development Required to Qualify the Technical Concept Westinghouse Non-Proprietary Class 3 Award
More informationBimetallic Barrel Benchmark Xaloy vs. Reiloy May 28th, 2013
Bimetallic Barrel Benchmark Xaloy vs. Reiloy May 28th, 2013 Scope available parts X-102 barrel Ø 35 mm x 1010 mm X-800 barrel Ø 40 mm x 1010 mm R121 barrel internally RP19.5 R216 barrel evaluated properties
More informationThe risk of reactor pressure vessel material degradation Doel-3/Tihange-2
The risk of reactor pressure vessel material degradation Doel-3/Tihange-2 Ilse Tweer NURIS 2015, Vienna, 16-17 April 2015 Reactor pressure vessel structural integrity The reactor pressure vessel (RPV)
More informationCOLD NEUTRON SOURCE AT CMRR
COLD NEUTRON SOURCE AT CMRR Hu Chunming Shen Wende, Dai Junlong, Liu Xiankun ( 1 ) Vadim Kouzminov, Victor Mityukhlyaev / 2 /, Anatoli Serebrov, Arcady Zakharov ( 3 ) ABSTRACT As an effective means to
More informationI IRRI
THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS Three Park Avenue, New York, N.Y. 10016-6990 99-GT-417 The Society shall not be responsible for statements or opinions advanced In papers or discuszo311 at
More informationPlanning for the Decommissioning of the ASTRA-Reactor
Planning for the Decommissioning of the ASTRA-Reactor Konrad Mück, Jörg Casta Austrian Research Center Seibersdorf Introduction The ASTRA Reactor, a 10 MW multipurpose MTR research reactor at the Austrian
More informationAISI D2 Cold work tool steel
T OOL STEEL FACTS AISI D2 Cold work tool steel Great Tooling Starts Here! This information is based on our present state of knowledge and is intended to provide general notes on our products and their
More information