Hydriding Induced Corrosion Failures in BWR Fuel

Size: px
Start display at page:

Download "Hydriding Induced Corrosion Failures in BWR Fuel"

Transcription

1 ASTM 17th International Symposium on Zirconium in the Nuclear Industry, Hyderabad, India Hydriding Induced Corrosion Failures in BWR Fuel Dan Lutz 1, Yang-Pi Lin 2, Randy Dunavant 2, Rob Schneider 2, Hartney Yeager 2, Aylin Kucuk 3, Bo Cheng 3, Jim Lemons 4 1 Global Nuclear Fuel Americas, Sunol, CA. 2 Global Nuclear Fuel Americas, Wilmington, NC 3 Electric Power Research Institute, Palo Alto, CA 4 Tennessee Valley Authority, Chattanooga, TN

2 Outline Key features of corrosion failures Summary of failure mechanism Unresolved issues Environmental factor contribution Hydride localization under BWR conditions Results Implications of environmental factor to the BWR fleet 2

3 Corrosion Failures Browns Ferry 2, Cycle 12 ( ) 63 assemblies (9x9), starting at ~30 GWd/MTU, 7 month into 24 month cycle Only second cycle fuel failed Cladding type: Zircaloy-2 with barrier, over 3 million fuel rods experience base Failed rods from multiple ingots; initial failures involved low alloy content, common lot material operated successfully in 12 other reactors to end-of-life Two rounds of hot cell investigations for root cause and failure mechanism Browns Ferry 3, Cycle 11 ( ) 3 assemblies, near end of 3 cycle life 1 assembly discharged after 1 cycle, not corrosion related 3

4 Corrosion Features Elevated, coalesced nodular corrosion and oxide spallation Oxide + Crud thickness Mostly oxide Excessive corrosion, spallation, hydriding and the primary failure perforations near 2.4 m (95 in) and above Low corrosion at assembly spacer (Zircaloy-2 ferrule with Inconel springs) locations Large azimuthal variation in corrosion Spacer Location 4

5 Failure Mechanism Schardt et al, Fuel Corrosion Failures in the Browns Ferry Nuclear Plant, 2004 International Meeting on LWRFP, Orlando, FL, September Lutz et al, Investigation of BWR Fuel Failures, TopFuel 2012, Manchester, UK, September Elevated corrosion on cladding with inadequate corrosion resistance for the exposed environment resulting in high hydrogen pickup Spallation of oxide locally Hydrogen localization leading to hydride lens near cladding OD Cladding stressed from cladding creep down and pellet expansion Failure from cracks initiated in brittle hydride Sound Rod Failed Rod Shared similarity in corrosion characteristics with previous CILC from s: failure involved crud intrusion into oxide, thermal insulation and autocatalytic corrosion 5

6 Unresolved Issues Environmental factor for elevated corrosion Material factor alone cannot explain elevated corrosion; expect coolant water chemistry effect Coolant s role in the failures not readily identifiable from review of water chemistry data Hydrogen localization under BWR conditions Are local thermal gradients at discontinuities in oxide thickness sufficient to cause localization? 6

7 Hydride Localization Oxide plateau Oxide spalling Oxide Crud Cladding 7

8 Objectives and Scope Environmental SIMS analyses of oxide (Failed vs. sound, plants, elevation and common material) Assess hideout return investigation data (water chemistry during reactor shutdown) Hydrogen localization under BWR conditions Finite element simulation 8

9 Samples Analyzed Using SIMS Failed Rod Earlier Reload Another BWR Early discharge Light corrosion Sound Rod, Heavy corrosion Sample Assembly Rod Condition Rod Average Exposure (GWd/MTU) Reactor Insertion Date Residence Time (Days) Sample Elevation (mm/inch) Average Oxide Thickness (microns) A YJS734 H2 Failed 47.3 BF2 May ~2286 / 90 n/a* B YJK363 B3 Sound 35.1 BF2 Oct ~3023 / C YJ1380 D1 Sound 68.9 L Jul ~2413 / D YJS614 G9 Sound 34.5 BF2 May ~3023 / E 724 / YJS616 B8 Sound 41.1 BF2 May F 2311 / 91 23* 1 Cycle Rod Common Lot (to E and F) G 762 / YJN587 E9 Sound 17.0 BF3 Oct H 2362 / I 699 / YJP354 E8 Sound 44.0 H Nov J 2286 / * Severely Spalled 9

10 Li in Oxide: SIMS Line Scans Failed BF2 H2 D, Plant BF2 METAL METAL C, Plant L High Li level in failed BF2 rod Lower Li level in sound, low corrosion BF2 rod Low Li in sound rod from plant L 6 Li implies natural source METAL 10

11 Line Scans High [Li] with 6 Li in sound BF2 rod at upper elevation confirmed. Lower elevation E: BF2 B8 ~2 ppm Upper elevation F: BF2 B8 ~300 ppm 1 cycle BF3 rod shows lower [Li], higher and with 6 Li at upper elevation G: BF3 E9 <1 ppm H: BF3 E9 ~3 ppm 3 cycle Plant H rod shows low [Li]; trend with elevation reversed compared with BF2 and BF3 rods I: H E8 ~3 ppm J: H E8 <1 ppm 11

12 BF2 B8 Upper elevation SIMS Imaging OXIDE CRUD BF3 E9 Upper elevation METAL OXIDE Association of Li and B implies similar trapping site and entry path from coolant 12

13 After: Shutdown Hideout Return Chemistry at Browns Ferry-2 and Hatch-2. EPRI, Palo Alto, CA: Hideout Return, BF2 and H, Feb 2003 Plant H had cladding common to failed BF2 rods No corrosion failures at Plant H Reactor Cold Zero power Reactor Hot Zero power Reactor Hot Full power Li below detection limit at Plant H in hideout return study (detected in SIMS) Sulfate, chloride, Ca return also higher at BF2 relative to Plant H Steady state sulfate and chloride reflect hideout return results Concentration factor predicted for BWR; possible ppm level of Li in BF2 Together with SIMS results indicative of environmental factor effect 13

14 Hydrogen Localization Model Fuel Cladding Thickness, TH 0.7 mm Fuel Cladding Section Axial Height Modeled, HT 25 mm ID O/M Interface Thick Oxide Thickness, OX THICK Spall Depth or Oxide Plateau Height, OX THIN /15 15 microns microns Axial Height of Spall or Thin Oxide Plateau, OX HT 2.5 mm, Spall 0.25 mm, Plateau mm LHGR 250 (7.6) W/cm (kw/ft) (0) DD αα cccc ss Ref. [3] cccc2 (0) 3 DD δδ ss QQ αα Ref. [3] RR 4170KK Ref. [3] QQ δδ RR 5730KK Ref. [3] QQ QQ αα RR 3015KK Ref. [3] QQ δδ RR 653KK Ref. [3] TTTTTTTT TTTTTTTT ee RRRR pppppp Ref. [4] ee RRRR pppppp TT 533KK ee RRRR pppppp TT > 533KK Ref. [4] 14

15 Thermal Gradient Oxide Spalling: T increases with oxide discontinuity (OX THICK OX THIN ) Oxide thickness (OX THICK ) affects temp at ID and O/M interface T not strongly sensitive to oxide thickness (OX THICK ) Axial Height, HT/2 (mm) C 317 C 349 C 317 C 250 W/cm 60 µm oxide layer 15 µm discontinuity OD T = 5 ID T = Clad Thickness, TH (mm) ID Axial Height, HT/2 (mm) W/cm 60 µm oxide layer 45 µm discontinuity OD T = 16 ID T = Clad Thickness, TH (mm) O/M Interface Oxide Plateau: Similar trend as Oxide Spalling cases T less than corresponding Oxide Spalling cases Axial Height, HT/2 (mm) C 317 C 349 C 317 C 250 W/cm 60 µm oxide layer 15 µm discontinuity OD T = 3 ID T = 1 Axial Height, HT/2 (mm) W/cm 60 µm oxide layer 45 µm discontinuity OD T = 10 ID T = Clad Thickness, TH (mm) Clad Thickness, TH (mm) 15

16 Hydrogen Diffusion Agreement with data from Sawatzky and Vogt (Mathematics of Thermal Diffusion of Hydrogen in Zircaloy-2, AECL-1411, 1971.) 16

17 Oxide Discontinuity and Hydride Localization Oxide Spalling: Localization increases with H content and T 5 o C T can cause localization with 200 ppm H No significant localization at 100 ppm H, even at 16 o C T Localization not sensitive to oxide thickness (OX THICK ), similar T Oxide Plateau: Similar trend as Oxide Spalling cases Less localization due to lower T 3 o C T insufficient to cause localization with 200 ppm H 60/15 micron oxide/spall 60/15 micron oxide/plateau 1000 hrs 1000 hrs 60/45 micron oxide/spall 60/45 micron oxide/plateau 17

18 Hydride Localization Hydride Localization Hydride localization possible if H is high (above 100 ppm) and with sufficient T from elevated corrosion and spalling (and high heat flux) all needed for failure mechanism to operate 18

19 Implications / Discussion Corrosion failures result from inadequate cladding corrosion resistance for the environment Li in oxide associated with corrosion failures Control blade absorber tube leakage (B 4 C) can be a source of 7 Li, consistent with Li and B association but B 4 C leakage is common; presence of 6 Li implies natural source Evidence of underperforming water chemistry cleanup system, e.g. sulfate spikes indicative of degraded cation resin Since the corrosion failures, cladding corrosion resistance and water chemistry control improved (Fe & Zn control, filter efficiency, monitoring) Confirmation of environmental factor provides added assurance for preventive measures No recurrence in ~10 years 19

20 Summary Elevated corrosion associated with Li in oxide, indicating environmental factor; Li as initiator not proven Hideout return study supported higher concentration of multiple species, including Li, than in reference plant Finite element modeling confirm significant thermal gradient associated with oxide spall or plateau can cause hydride localization if sufficient hydrogen is present Failure mechanism: cracking of brittle hydride under stress; hydride localization from thermal gradient at discontinuities of thick oxide; elevated corrosion from inadequate cladding corrosion resistance under demanding environment 20

Fuel Reliability (QA)

Fuel 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 information

3D Printing of Components and Coating Applications at Westinghouse

3D 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 information

2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea

2017 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 information

Behavior 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 - 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 information

WM2014 Conference, March 2 6, 2014, Phoenix, Arizona, USA

WM2014 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 information

MICROSTRUCTURE CHARACTERIZATION OF ZIRLO STRUCTURAL COMPONENTS IRRADIATED TO HIGH BURNUP

MICROSTRUCTURE CHARACTERIZATION OF ZIRLO STRUCTURAL COMPONENTS IRRADIATED TO HIGH BURNUP 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

More information

2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea

2017 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 information

Evaluations of Mo-alloy for light water reactor fuel cladding to enhance accident tolerance

Evaluations 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 information

Database 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/ 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 information

Regulatory Challenges. and Fuel Performance

Regulatory Challenges. and Fuel Performance IAEA Technical Meeting on Flexible (Non-Baseload) Operation Approaches for Nuclear Power Plants Regulatory Challenges and Fuel Performance Paul Clifford United States of America Agenda 1. Regulatory Challenges

More information

In-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 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 information

Effect 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 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 information

Chapter 22. Waterside Corrosion and Hydriding of Zr Alloy Cladding

Chapter 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 information

2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea

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 information

Regulatory Perspective on CANDU Feeder Pipe Degradation due to FAC and IGSCC

Regulatory Perspective on CANDU Feeder Pipe Degradation due to FAC and IGSCC Regulatory Perspective on CANDU Feeder Pipe Degradation due to FAC and IGSCC John C. Jin, Specialist Raoul Awad, Director, Canadian Nuclear Safety Commission Commission canadienne de surete nucleaire IAEA

More information

ASTM Conference, Feb , Hyderabad, India

ASTM Conference, Feb , Hyderabad, India ASTM Conference, Feb 6 2013, Hyderabad, India Effect of Hydrogen on Dimensional Changes of Zirconium and the Influence of Alloying Elements: First-principles and Classical Simulations of Point Defects,

More information

Simulating Pellet and Clad Mechanical Interactions of Nuclear Fuel Rod for Pressure Water Reactors

Simulating 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 information

Low Cross-linked Resin Reduces Iron, Maintains Low Reactor Water Sulfate

Low Cross-linked Resin Reduces Iron, Maintains Low Reactor Water Sulfate Case History Low Cross-linked Resin Reduces Iron, Maintains Low Reactor Water Sulfate Site Information Location New York, USA Purpose Reduce feed water iron from 2-3 ppb without increased sulfate concentration

More information

2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea

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 information

EVOLUTION OF HYDROGEN PICKUP FRACTION WITH OXIDATION RATE ON ZIRCONIUM ALLOYS ABSTRACT

EVOLUTION 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 information

High Temperature Corrosion in Gasifiers

High Temperature Corrosion in Gasifiers Vol. Materials 7, No. Research, 1, 2004Vol. 7, No. 1, 53-59, 2004. High Temperature Corrosion in Gasifiers 2004 53 High Temperature Corrosion in Gasifiers Wate Bakker* EPRI 3412 Hillview Avenue Palo Alto,

More information

Irradiation Testing of Structural Materials in Fast Breeder Test Reactor

Irradiation Testing of Structural Materials in Fast Breeder Test Reactor Irradiation Testing of Structural Materials in Fast Breeder Test Reactor IAEA Technical Meet (TM 34779) Nov 17-21, 2008 IAEA, Vienna S.Murugan, V. Karthik, K.A.Gopal, N.G. Muralidharan, S. Venugopal, K.V.

More information

Hideout of Sodium Phosphates in Steam Generator Crevices

Hideout of Sodium Phosphates in Steam Generator Crevices Hideout of Sodium Phosphates in Steam Generator Crevices By Gwendy Harrington Department of Chemical Engineering, University of New Brunswick, P.O. Box 4400, Fredericton, New Brunswick, E3B 5A3 Abstract

More information

Dry storage systems and aging management

Dry 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 information

IRRADIATION 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 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 information

ON-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 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 information

A Comparative Analysis of CABRI CIP0-1 and NSRR VA-2 Reactivity Initiated Accident tests

A 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 information

Fracture of Zircaloy-4 fuel cladding tubes with hydride blisters

Fracture 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 information

INFLUENCE OF STEAM PRESSURE ON THE HIGH POST-COOLING MECHANICAL PROPERTIES OF ZIRCALOY-4 AND M5 CLADDING (LOCA CONDITIONS)

INFLUENCE OF STEAM PRESSURE ON THE HIGH POST-COOLING MECHANICAL PROPERTIES OF ZIRCALOY-4 AND M5 CLADDING (LOCA CONDITIONS) INFLUENCE OF STEAM PRESSURE ON THE HIGH TEMPERATURE OXIDATION AND POST-COOLING MECHANICAL PROPERTIES OF ZIRCALOY-4 AND M5 CLADDING (LOCA CONDITIONS) M. Le Saux 1*, V. Vandenberghe 1, P. Crébier 2, J.C.

More information

Acceptance Criteria in DBA

Acceptance 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 information

Dissimilar Metal Welds (DMW) in German LWR s. Design Types, Disbonding, NDT. W. Mayinger, K.J. Metzner E.ON Kernkraft, Hannover, Germany

Dissimilar Metal Welds (DMW) in German LWR s. Design Types, Disbonding, NDT. W. Mayinger, K.J. Metzner E.ON Kernkraft, Hannover, Germany Dissimilar Metal Welds (DMW) in German LWR s Design Types, Disbonding, NDT W. Mayinger, K.J. Metzner E.ON Kernkraft, Hannover, Germany 2 Scope Scope Flaws/Cracks in Dissimilar Metal Welds (DMW) 1. Interfacial

More information

Irradiation Assisted Stress Corrosion Cracking. By Topan Setiadipura [09M51695] (Obara Lab., Nuclear Engineering Dept., Tokyo Tech.

Irradiation 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 information

Journal of American Science 2014;10(2) Burn-up credit in criticality safety of PWR spent fuel.

Journal 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 information

Irradiation capabilities at the Halden reactor and testing possibilities under supercritical water conditions

Irradiation 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 information

E-BRITE E-BRITE. Technical Data Sheet. Stainless Steel: Superferritic GENERAL PROPERTIES PLANAR SOLID OXIDE FUEL CELLS CHEMICAL COMPOSITION

E-BRITE E-BRITE. Technical Data Sheet. Stainless Steel: Superferritic GENERAL PROPERTIES PLANAR SOLID OXIDE FUEL CELLS CHEMICAL COMPOSITION E-BRITE Stainless Steel: Superferritic (UNS 44627, ASTM Type XM-27) GENERAL PROPERTIES E-BRITE alloy is a high purity ferritic stainless steel which combines excellent resistance to corrosion and oxidation

More information

Materials Issues Related to Reactor Design, Operation & Safety

Materials Issues Related to Reactor Design, Operation & Safety Materials Issues Related to Reactor Design, Operation & Safety Professor R. G. Ballinger Department of Nuclear Engineering Department of Materials Science & Engineering 22.39 Lecture 1-1 Objective To Develop

More information

CASL: Consortium for the Advanced Simulation of Light Water Reactors A DOE Energy Innovation Hub

CASL: Consortium for the Advanced Simulation of Light Water Reactors A DOE Energy Innovation Hub CASL: Consortium for the Advanced Simulation of Light Water Reactors A DOE Energy Innovation Hub Achievements in Addressing Challenges Facing the Light Water Reactor Industry Dave Pointer, PhD. Deputy

More information

THE 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 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 information

Reactor Technology --- Materials, Fuel and Safety

Reactor 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 information

ZRO 2 AND UO 2 DISSOLUTION BY MOLTEN ZIRCALLOY

ZRO 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 information

Recovery Boiler Corrosion Chemistry

Recovery Boiler Corrosion Chemistry Recovery Boiler Corrosion Chemistry Sandy Sharp and Honghi Tran SharpConsultant University of Toronto Pulp and Paper Centre Symposium on Corrosion in Pulp and Paper Mills and Biorefineries, RBI at Georgia

More information

FUEL ROD PERFORMANCE MEASUREMENTS AND RE-INSTRUMENTATION CAPABILITIES AT THE HALDEN PROJECT

FUEL 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 information

Evaluation of corrosion on the fuel performance of stainless steel cladding

Evaluation of corrosion on the fuel performance of stainless steel cladding EPJ Nuclear Sci. Technol. 2, 4 (216) D. de Souza Gomes et al., published by EDP Sciences, 216 DOI: 1.151/epjn/21633 Nuclear Sciences & Technologies Available online at: http://www.epj-n.org REGULAR ARTICLE

More information

MOISTURE IN PAPER ASSESSMENT FROM CONTINUOUS MONITORING OF MOISTURE IN OIL. GE Energy

MOISTURE IN PAPER ASSESSMENT FROM CONTINUOUS MONITORING OF MOISTURE IN OIL. GE Energy MOISTURE IN PAPER ASSESSMENT FROM CONTINUOUS MONITORING OF MOISTURE IN OIL Jacques Aubin GE Energy Brian D Sparling Abstract Moisture content of solid insulation is a persistent concern for power transformer

More information

High Temperature Effects on Vessel Integrity. Marc Levin, Ayman Cheta Mary Kay O Connor Process Safety Center 2009 International Symposium

High Temperature Effects on Vessel Integrity. Marc Levin, Ayman Cheta Mary Kay O Connor Process Safety Center 2009 International Symposium High Temperature Effects on Vessel Integrity Marc Levin, Ayman Cheta Mary Kay O Connor Process Safety Center 2009 International Symposium Outline Motivation Basics / Basis for Pressure Vessel Design Conditions

More information

severe accident progression in the BWR lower plenum and the modes of vessel failure

severe 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 information

Ensuring Spent Fuel Pool Safety

Ensuring 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 information

Assessment 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 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 information

Reactivity requirements can be broken down into several areas:

Reactivity requirements can be broken down into several areas: Reactivity Control (1) Reactivity Requirements Reactivity requirements can be broken down into several areas: (A) Sufficient initial reactivity should be installed to offset the depletion of U 235 and

More information

Task 1 Progress: Analysis of TREAT Minimum Critical and M8CAL Cores with SERPENT and SERPENT/PARCS

Task 1 Progress: Analysis of TREAT Minimum Critical and M8CAL Cores with SERPENT and SERPENT/PARCS Task 1 Progress: Analysis of TREAT Minimum Critical and M8CAL Cores with SERPENT and SERPENT/PARCS Volkan Seker, Matt Neuman, Nicholas Kucinski, Hunter Smith, Tom Downar University of Michigan May 24,

More information

AREVA 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 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 information

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

Fuel Management of VVER-1000 Reactors of Kudankulam Nuclear Power Plant, India. Fuel Management of VVER-1000 Reactors of Kudankulam Nuclear Power Plant, India. Y. K. Pandey & Ashok Chauhan Safety Group, Directorate of Projects, Nuclear Power of Corporation of India Limited, Nabhikiya

More information

EXPERIMENTS ON AIR INGRESS DURING SEVERE ACCIDENTS

EXPERIMENTS ON AIR INGRESS DURING SEVERE ACCIDENTS 13 th International Conference on Nuclear Engineering Beijing, China, May 16-20, 2005 ICONE13-50080 EXPERIMENTS ON AIR INGRESS DURING SEVERE ACCIDENTS Martin Steinbrück *, Alexei Miassoedov **, Gerhard

More information

The 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 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 information

Application 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 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 information

The Effects of Microstructure and Operating Conditions on Irradiation

The 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 information

Analytical support to experiment QUENCH-17 and first post-test calculations with ATHLET-CD

Analytical support to experiment QUENCH-17 and first post-test calculations with ATHLET-CD Analytical support to experiment QUENCH-17 and first post-test calculations with ATHLET-CD C. Bals, T. Hollands, H. Austregesilo Gesellschaft für Anlagen- und Reaktorsicherheit (GRS), Germany Content Short

More information

Solutions. for Severe Corrosion

Solutions. for Severe Corrosion Solutions for Severe Corrosion Linas Mazeika, President, 3L&T Inc., USA, reveals how to prevent equipment corrosion caused by hot combustion gases in a cement plant. Summary The serious economic consequences

More information

2017 Water Reactor Fuel Performance Meeting September 10 (Sun) 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea

2017 Water Reactor Fuel Performance Meeting September 10 (Sun) 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea Extended Storage Collaboration Program for Addressing Long-Term Dry Storage Issues Hatice Akkurt 1 1 Electric Power Research Institute (EPRI), 1300 W WT Harris Blvd, Charlotte, NC 28262, hakkurt@epri.com

More information

Module 06 Boiling Water Reactors (BWR)

Module 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 information

Fundamental Research Program for Removal of Fuel Debris

Fundamental Research Program for Removal of Fuel Debris International Symposium on the Decommissioning of TEPCO s Fukushima Daiichi Nuclear Power Plant Unit 1-4 1 Fundamental Research Program for Removal of Fuel Debris March 14, 2012 Tadahiro Washiya Japan

More information

THERMAK 17. High Temperature Strength. Superior Oxidation Resistance. Excellent Thermal Fatigue Resistance. Equiaxed Microstructure

THERMAK 17. High Temperature Strength. Superior Oxidation Resistance. Excellent Thermal Fatigue Resistance. Equiaxed Microstructure THERMAK 17 Stainless STEEL Product Data Bulletin Applications Potential High Temperature Strength Superior Oxidation Resistance Excellent Thermal Fatigue Resistance Equiaxed Microstructure THERMAK 17 Stainless

More information

STAINLESS STEEL SHEETS

STAINLESS STEEL SHEETS STAINLESS & NICKEL ALLOY CHATHAM STEEL CORPORATION 59 STAINLESS STEEL SHEETS TYPE 04, 04L, 6, 6L No. 2B Finish Cold Rolled, Annealed *No. 2D Finish No. Finish Polished One Side No. 4 Finish Polished One

More information

ATOM-PROBE ANALYSIS OF ZIRCALOY

ATOM-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 information

CHAPTER 5: DIFFUSION IN SOLIDS

CHAPTER 5: DIFFUSION IN SOLIDS CHAPTER 5: DIFFUSION IN SOLIDS ISSUES TO ADDRESS... How does diffusion occur? Why is it an important part of processing? How can the rate of diffusion be predicted for some simple cases? How does diffusion

More information

Design Considerations for Advanced Combined Cycle Plant Using Fast Start Drum Type HRSGs

Design Considerations for Advanced Combined Cycle Plant Using Fast Start Drum Type HRSGs Presented To: Hot Topic Hour A Babcock Power Inc. Company By: Deron Johnston Date: Mar.7/13 Design Considerations for Advanced Combined Cycle Plant Using Fast Start Drum Type HRSGs Authors: Akber Pasha,

More information

Stress Corrosion Cracking in a Dissimilar Metal Butt Weld in a 2 inch Nozzle. Master of Engineering in Mechanical Engineering

Stress Corrosion Cracking in a Dissimilar Metal Butt Weld in a 2 inch Nozzle. Master of Engineering in Mechanical Engineering Stress Corrosion Cracking in a Dissimilar Metal Butt Weld in a 2 inch Nozzle by Thomas E. Demers An Engineering Project Submitted to the Graduate Faculty of Rensselaer Polytechnic Institute in Fulfillment

More information

Metal Transitions HIGH TEMPERATURE

Metal 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 information

High Temperature Oxidation of Zr-2.5%wt Nb Alloys Doped with Yttrium

High 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 information

Introduction to Joining Processes

Introduction to Joining Processes 4. TEST METHODS Joints are generally designed to support a load, and must be tested to evaluate their load-supporting capabilities. However, it is also important to evaluate, not the joint, but rather

More information

Nuclear Power Reactors. Kaleem Ahmad

Nuclear Power Reactors. Kaleem Ahmad Nuclear Power Reactors Kaleem Ahmad Outline Significance of Nuclear Energy Nuclear Fission Nuclear Fuel Cycle Nuclear Power Reactors Conclusions Kaleem Ahmad, Sustainable Energy Technologies Center Key

More information

CLAD STAINLESS STEELS AND HIGH-NI-ALLOYS FOR WELDED TUBE APPLICATION

CLAD STAINLESS STEELS AND HIGH-NI-ALLOYS FOR WELDED TUBE APPLICATION CLAD STAINLESS STEELS AND HIGHNIALLOYS FOR WELDED TUBE APPLICATION Wolfgang Bretz Wickeder Westfalenstahl GmbH Hauptstrasse 6 D58739 Wickede, Germany Keywords: Cladding, Laser/TIG Welding, Combined SolderingWelding

More information

Catalyzed-assisted Manufacture of Olefins (CAMOL): Year-4 Update on Furnace Installations

Catalyzed-assisted Manufacture of Olefins (CAMOL): Year-4 Update on Furnace Installations Catalyzed-assisted Manufacture of Olefins (CAMOL): Year-4 Update on Furnace Installations Steve Petrone, Robert L. Deuis, Fuwing Kong and Peter Unwin Quantiam Technologies Inc. Edmonton, Alberta, Canada

More information

Segregation in Vacuum Arc Remelted Zirconium Alloy Ingots

Segregation in Vacuum Arc Remelted Zirconium Alloy Ingots 16 th International Symposium on Zirconium in the Nuclear Industry Chengdu, China Segregation in Vacuum Arc Remelted Zirconium Alloy Ingots A. Jardy 1, F. Leclerc 2, M. Revil-Baudard 1-2, P. Guerin 2,

More information

The international program Phebus FP (fission

The 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 information

Titanium 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 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 information

Nickel Based Superalloy Incoloy 800 (UNS N08800)

Nickel Based Superalloy Incoloy 800 (UNS N08800) Nickel Based Superalloy Incoloy 800 (UNS N08800) Nickel-Iron-Chromium alloy Incoloy 800 has fine strength and suitable resistance to oxidation and carburization at high temperatures. It offers elevated

More information

PROCESS CHANGES AND CORROSION OF MATERIALS IN RECOVERY BOILERS

PROCESS CHANGES AND CORROSION OF MATERIALS IN RECOVERY BOILERS PROCESS CHANGES AND CORROSION OF MATERIALS IN RECOVERY BOILERS Black Liquor Combustion Coloquium Park City, Utah May 13, 23 Doug Singbeil and Joey Kish Paprican Jim Keiser ORNL 2 1 3 Recovery boiler evolution:

More information

EVALUATION OF A CHEMICAL CLEANING FORMULATION FOR THE STEAM GENERATOR OF NUCLEAR POWER PLANTS

EVALUATION OF A CHEMICAL CLEANING FORMULATION FOR THE STEAM GENERATOR OF NUCLEAR POWER PLANTS EVALUATION OF A CHEMICAL CLEANING FORMULATION FOR THE STEAM GENERATOR OF NUCLEAR POWER PLANTS A.L.Rufus, H.Subramanian, V.S.Sathyaseelan, Padma S.Kumar, S.Veena B.Anupkumar, M.P.Srinivasan, S.Velmurugan

More information

IN-PILE PERFORMANCE OF HANA CLADDING TESTED IN HALDEN REACTOR

IN-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 information

QUICK REFERENCE GUIDE THE ALLOY SPECIALISTS

QUICK REFERENCE GUIDE THE ALLOY SPECIALISTS QUICK REFERENCE GUIDE THE ALLOY SPECIALISTS CONTENTS SOLUTIONS TO MATERIALS PROBLEMS Corrosion-Resistant Alloys...4 Heat-Resistant Alloys...7 High-Performance & Special Purpose Alloys... 10 Special Metals

More information

System Analysis of Pb-Bi Cooled Fast Reactor PEACER

System Analysis of Pb-Bi Cooled Fast Reactor PEACER OE-INES-1 International Symposium on Innovative Nuclear Energy Systems for Sustainable Development of the World Tokyo, Japan, October 31 - November 4, 2004 System Analysis of Pb-Bi ooled Fast Reactor PEAER

More information

Radioactive Materials from U.S. Nuclear Plants

Radioactive Materials from U.S. Nuclear Plants Routine Releases of Radioactive Materials from U.S. Nuclear Plants Dave Lochbaum Union of Concerned Scientists August 2014 Revision i 1 1 The idea for this material came during a November 2013 workshop

More information

UPDATED CASE STUDY OF FIRESIDE CORROSION MANAGEMENT IN AN RDF FIRED ENERGY-FROM-WASTE BOILER

UPDATED CASE STUDY OF FIRESIDE CORROSION MANAGEMENT IN AN RDF FIRED ENERGY-FROM-WASTE BOILER UPDATED CASE STUDY OF FIRESIDE CORROSION MANAGEMENT IN AN RDF FIRED ENERGY-FROM-WASTE BOILER Steve Vrchota Great River Energy Elk River, MN, United States Tim Peterson Great River Energy Elk River MN,

More information

Long-Term Nucle ar O p erations

Long-Term Nucle ar O p erations Long-Term Nucle ar O p erations E x e c u t i v e S u m m a r y Long-Term Nuclear Operations What is a feasible life span for a nuclear plant? I would say 80 years, without question. And I think if we

More information

Combinations of alloys and environments subject to dealloying and elements preferentially removed

Combinations of alloys and environments subject to dealloying and elements preferentially removed ;~ page 8-1 SELECTIVE LEACHING ("Dealloying", "Parting") Corrosion in which one constituent of an alloy is preferentially removed, leaving behind an altered (weakened) residual structure. Can occur in

More information

Boiler Life and Availability Improvement Program - Program 63

Boiler Life and Availability Improvement Program - Program 63 Boiler Life and Availability Improvement Program - Program 63 Program Description Program Overview Safety and availability loss due to pressure part failures are two key issues driving R&D on major fossil

More information

Safety design approach for JSFR toward the realization of GEN-IV SFR

Safety 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 information

Microstructural Characterization of Aluminum Powder Liquid Coating on IN 738 Superalloy

Microstructural Characterization of Aluminum Powder Liquid Coating on IN 738 Superalloy Journal of Metals, Materials and Minerals. Vol.17 No.2 pp. 75-79, 2007 Microstructural Characterization of Aluminum Powder Liquid Coating on IN 738 Superalloy Patama VISUTTIPITUKUL 1*, Nuntiya LIMVANUTPONG

More information

AGEING MANAGEMENT PROGRAM TO REACTOR PRESSURE VESSEL INTERNALS COMPONENTS IN A BWR NUCLEAR POWER PLANT

AGEING MANAGEMENT PROGRAM TO REACTOR PRESSURE VESSEL INTERNALS COMPONENTS IN A BWR NUCLEAR POWER PLANT AGEING MANAGEMENT PROGRAM TO REACTOR PRESSURE VESSEL INTERNALS COMPONENTS IN A BWR NUCLEAR POWER PLANT C. R. Arganis J. a, J. A. Aguilar T. a, M. A. Sanchez M. b a Instituto Nacional de Investigaciones

More information

Understanding the effects of reflooding in a reactor core beyond LOCA conditions

Understanding 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 information

Caustic Gouging. PPChem. PPChem 101 BOILER AND HRSG TUBE FAILURES LESSON 5: INTRODUCTION TYPICAL CHARACTERISTICS OF DAMAGE

Caustic Gouging. PPChem. PPChem 101 BOILER AND HRSG TUBE FAILURES LESSON 5: INTRODUCTION TYPICAL CHARACTERISTICS OF DAMAGE 101 BOILER AND HRSG TUBE FAILURES LESSON 5: Caustic Gouging R. Barry Dooley and Albert Bursik INTRODUCTION In Lesson 3 of this course Underdeposit Corrosion A General Introduction presented in the December

More information

Protection Effectiveness of Vapor Corrosion Inhibitor VpCI 619 for Corrosion Under Insulation at Elevated Temperatures

Protection Effectiveness of Vapor Corrosion Inhibitor VpCI 619 for Corrosion Under Insulation at Elevated Temperatures VpCI 619 Protection Effectiveness of Vapor Corrosion Inhibitor VpCI 619 for Corrosion Under Insulation at Elevated Temperatures For: CORTEC Corporation by Behzad Bavarian, California State University,

More information

Multiphase Flow Dynamics 4

Multiphase Flow Dynamics 4 Multiphase Flow Dynamics 4 Nuclear Thermal Hydraulics von Nikolay I Kolev 1. Auflage Multiphase Flow Dynamics 4 Kolev schnell und portofrei erhältlich bei beck-shop.de DIE FACHBUCHHANDLUNG Thematische

More information

Structural Integrity Research for Reactor Pressure Vessel under In-Vessel Melt Retention

Structural Integrity Research for Reactor Pressure Vessel under In-Vessel Melt Retention IAEA Technical Meeting on In-Vessel Melt Retention and Ex-Vessel Corium Cooling Oct. 17-21, 2016, Shanghai, CHINA Structural Integrity Research for Reactor Pressure Vessel under In-Vessel Melt Retention

More information

EXPERIMENTAL RESIDUAL STRESS EVALUATION OF HYDRAULIC EXPANSION TRANSITIONS IN ALLOY 690 STEAM GENERATOR TUBING

EXPERIMENTAL RESIDUAL STRESS EVALUATION OF HYDRAULIC EXPANSION TRANSITIONS IN ALLOY 690 STEAM GENERATOR TUBING EXPERIMENTAL RESIDUAL STRESS EVALUATION OF HYDRAULIC EXPANSION TRANSITIONS IN ALLOY 9 STEAM GENERATOR TUBING Rod McGregor, Babcock & Wilcox International, Doug Hornbach, Lambda Research Usama Abdelsalam,

More information

Material Selection and Parameter Optimization for Reliable TMV Pop Assembly

Material Selection and Parameter Optimization for Reliable TMV Pop Assembly Selection and Parameter Optimization for Reliable TMV Pop Assembly Brian Roggeman, David Vicari Universal Instruments Corp. Binghamton, NY, USA Roggeman@uic.com Martin Anselm, Ph.D. - S09_02.doc Lee Smith,

More information

Texas A&M University, Department of Nuclear engineering, Ph.D. Qualifying Examination, Fall 2016

Texas A&M University, Department of Nuclear engineering, Ph.D. Qualifying Examination, Fall 2016 Part 2 of 2 100 points of the total exam worth of 200 points Research Area Specific Problems Select and answer any 4 problems from the provided 15 problems focusing on the topics of research tracks in

More information

INSULATED ELEMENT TWIST WELD DUPLEX INSULATED TIG WELD 1 1/2 NOMINAL ORDER CODES. INSULATOR DIMENSIONS (inches) SINGLE DUPLEX

INSULATED ELEMENT TWIST WELD DUPLEX INSULATED TIG WELD 1 1/2 NOMINAL ORDER CODES. INSULATOR DIMENSIONS (inches) SINGLE DUPLEX Straight Base Metal Thermocouple Elements The straight base metal thermocouple elements illustrated on this catalog page are replacement elements for use in Pyromation's complete industrial thermocouple

More information

SIMULATION 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 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 information