INFLUENCE OF THE HYDRIDE PRECIPITATION ON THE CORROSION KINETICS OF ZIRCALOY-4:
|
|
- Abigayle Crawford
- 5 years ago
- Views:
Transcription
1 INFLUENCE OF THE HYDRIDE PRECIPITATION ON THE CORROSION KINETICS OF ZIRCALOY-4: EFFECT OF THE NANOSTRUCTURE AND GRAIN BOUNDARY PROPERTIES OF ZIRCONIUM OXIDE LAYER ON THE OXYGEN DIFFUSION FLUX M. Jublot, G. Zumpicchiat, M. Tupin, S. Pascal, C. Berdin, C. Bisor, M. Blat ASTM : 18 th International Symposium on Zirconium in the Nuclear Industry 16 TH MAY 2016
2 BACKGROUND Pressurized Water Reactor (PWR) Fuel cladding material : ZIRCALOY-4 (Zy4) Fuel Assembly Image : Areva Primary coolant loop: liquid water - ~ 320 C; 155 bars ppm B - 2 ppm Li - [H 2 ] = 25 cc/kg Alloying elements Sn, wt% Fe, wt% Cr, wt% O, wt% Corrosion of the Zy-4 fuel cladding CEA 16 th May 2016 PAGE 2 H, wt.ppm Zircaloy
3 560 µm BACKGROUND Pressurized Water Reactor (PWR) Reaction of oxidation : Zr + 2 H 2 0 ZrO H 2 ZrO 2 Zr + ZrH x Zy-4 [Bernaudat et al., Topfuel 2009] Fuel rod Burnup Cross-section of a Zy-4 cladding oxidized in Reactor [Bossis, ASTM 2005] CEA 16 th May 2016 PAGE 3
4 BACKGROUND Pressurized Water Reactor (PWR) Potential factors of the «High Burn-Up» acceleration of Zy4 Dissolution of the Zr(Fe,Cr) 2 precipitates Tin content Li effect Irradiation impact on the microstructure Hydride accumulation at the oxide/metal interface Reaction of oxidation : Zr + 2 H 2 0 ZrO H 2 ZrO 2 Zr + ZrH x How the hydride accumulation affects the microstructure of the zirconium oxide? What is the impact on the corrosion kinetics? Zy-4 TEM investigation of the oxide with an Automated Crystal Orientation Mapping tool (ACOM-TEM) Cross-section of a Zy-4 cladding oxidized in Reactor [Bossis, ASTM 2005] Grain size distribution Grain orientations the Grain boundary misorientation CEA 16 th May 2016 PAGE 4
5 OUTLINE How the hydride accumulation affects the microstructure of the zirconium oxide? Materials & techniques Results: The oxide nanostructure The grain boundary misorientation The oxygen diffusion simulation as a function of the nanostructure Resume CEA 16 th May 2016 PAGE 5
6 MATERIALS & TECHNIQUES 2 samples Reference Zy4 (Zy4) Recrystallized sheets of Zy4 [Bisor C. Phd (2010)] Hydrided Zy4 (Zy4-h) [Blat et al. ASTM 1996 p.319] Cathodic charging technique ~ 8 µm thick 8 µm d-zrh 1,66 Oxydation in PWR conditions (Autoclave : 360 C; 190bars; Li; B) Corrosion kinetics In pre-transition phase Hydrided Zy4 Zy4 [Tupin et al., Corrosion Science 98 (2015)] dx dt X=1µm = 1,8 dx dt X=1µm Zy4 Zy4-h The oxidation rate is higher on hydrided Zy4 CEA 16 th May 2016 PAGE 6
7 MATERIALS & TECHNIQUES Cross-section analysis Fractography ZrO 2 / Hydride Fractography ZrO 2 / Hydride ZrO nm d-zrh 1,66 ZrO 2 / Zy4 ZrO nm a-zr [Bisor C. Phd (2010)] CEA 16 th May 2016 PAGE 7
8 MATERIALS & TECHNIQUES Cross-section analysis TEM Bright field ZrO 2 Fractography ZrO 2 / Hydride ZrO 2 : Columnar grains Width : nm Length : nm [De Gabory et al. JNM 456 (2015) p.272] 200 nm ZrO 2 d-zrh 1,66 TEM lamella thickness: From FIB preparation : ~100 nm ZrO 2 ZrO 2 / Zy4 ZrO nm a-zr CEA 16 th May 2016 PAGE 8 [Bisor C. Phd (2010)]
9 MATERIALS & TECHNIQUES Plan-view TEM sample preparation FIB tool ~300 nm ZrO 2 / Zy4 ZrO 2 / Hydride ZrO 2 Thickness ~ 60 nm Thickness ~ 55 nm Advantages of the plan-view analysis : - To analyse single grains through the FIB foil thickness - To investigate the properties of the grain boundaries which control the corrosion kinetics of Zy4 alloy. - To scan a wide zone of interest for a better statistic (~30 µm 2 ) CEA 16 th May 2016 PAGE 9
10 MATERIALS & TECHNIQUES Plan-view TEM analysis ACOM-TEM technique [E. Rauch et al., Microsc Anal, 22, 2008] ASTAR TM tool TEM FEI tecnai 30 G2 Automated Crystal Orientation Mapping (ACOM-TEM) To index the crystal phase To index the crystal orientation Principle e - 1 µm Acquired pattern Pre-calculated templates Orientation map Index map: highlighting the grain boundaries Reliability map: CEA 16 th May 2016 PAGE 10
11 4.2 µm 5.2 µm MATERIALS & TECHNIQUES Scanning conditions e - ZrO 2 / Zy4 ZrO 2 / Hydride 6.0 µm Beam size : 9 nm Scan step : 5 nm 1 µm 6.4 µm Scanned area: 27 µm 2 31 µm 2 Orientation maps Monoclinic phase of ZrO 2 ~300 nm Monoclinic phase of ZrO 2 (a-zro 2 ) Tetragonal phase 200 nm a-zr ZrO 2 a = 5.15 Å b = 5.21 Å c = 5.32 Å b = b y z x a a = 5.08 Å b = 5.08 Å c = 5.17 Å b = 90 b z y x a c c CEA 16 th May 2016 PAGE 11
12 RESULTS THE OXIDE NANOSTRUCTURE Orientation maps ZrO 2 / Zy4 Monoclinic phase of ZrO 2 ZrO 2 / Hydride 500 nm ~ 9000 indexed grains Euler angles ZrO 2 monoclinic F1 0 F 0 F nm ~ indexed grains CEA 16 th May 2016 PAGE
13 RESULTS THE OXIDE NANOSTRUCTURE Index map The grain size distribution ZrO 2 / Zy4 Columnar oxide grains Base shape - not a regular polygon - Spread size distribution 500 nm 100 nm CEA 16 th May 2016 PAGE 13
14 RESULTS THE OXIDE NANOSTRUCTURE The grain size distribution ZrO 2 / Zy4 ~ 9000 indexed grains Columnar oxide grains Base shape - not a regular polygon - Spread size distribution 50 % of grains Average diameter nm 34.6 nm Conditions - Base shape converted as a circular shape - Grain diameters > 15 nm - Misorientation angle between adjacent grains > 10 > 10 CEA 16 th May 2016 PAGE 14
15 RESULTS THE OXIDE NANOSTRUCTURE The grain size distribution Index map ZrO 2 / Hydride Columnar oxide grains Base shape 500 nm CEA 16 th May 2016 PAGE 15
16 RESULTS THE OXIDE NANOSTRUCTURE The grain size distribution ZrO 2 / Hydride Columnar oxide grains Base shape - ~ regular shape - Smaller size ZrO 2 / Zy4 500 nm CEA 16 th May 2016 PAGE 16
17 RESULTS THE OXIDE NANOSTRUCTURE The grain size distribution ZrO 2 / Hydride ~ indexed grains - ZrO 2 / Hydride - ZrO 2 / Zy4 Columnar oxide grains Base shape - ~ regular shape - Smaller size 50 % of grains Average diameter Ø nm 34.6 nm Ø nm 27.8 nm CEA 16 th May 2016 PAGE 17
18 RESULTS THE OXIDE NANOSTRUCTURE The grain size distribution ZrO 2 / Hydride Columnar oxide grains Base shape - ~ regular shape - Smaller size 50 % of grains Average diameter Ø nm 34.6 nm Ø nm 27.8 nm Consequences on the corrosion kinetic of Zy4 Oxygen diffuses through the grain boundaries Base size grain boundary density Surface fraction of the oxide grain boundaries f ZrO2/Hydride = 2.9 % - Base shape converted as a hexagonal shape - Intergranular space of 0.5 nm 0.5 nm f ZrO2/Zy4 = 1.8 % = + 60% CEA 16 th May 2016 PAGE 18 Partially explain the higher corrosion kinetic of the massive hydride
19 RESULTS GRAIN BOUNDARY MISORIENTATION Misorientation angles between adjacent grains [Sainfort, 1984] Consequences on the corrosion kinetic on Zy4 Base size grain boundary density 0.5 nm Oxygen diffuse through the grain boundaries Surface fraction of the oxide grain boundaries f ZrO2/Hydride = 2.9 % - Base shape converted as a hexagonal shape - Intergranular space of 0.5 nm f ZrO2/Zy4 = 1.8 % = + 60% CEA 16 th May 2016 PAGE 19 Partially explain the higher corrosion kinetic of the massive hydride
20 RESULTS GRAIN BOUNDARY MISORIENTATION Misorientation angles distribution between adjacent grains Not randomly distributed Angular range ( ) Distribution (%) on Zy4 on hydride % 29 % 20 % 9 % 15 % 19 % 28 % 9 % CEA 16 th May 2016 PAGE 20
21 b c y a z x RESULTS GRAIN BOUNDARY MISORIENTATION Misorientation angles distribution between adjacent grains Not randomly distributed 90 [001] b y z x c a a y x b z x b c a c 180 [101] z y Low coherent misorientation angles Twins tetragonal to monoclinic phase transformation Low interfacial energy Lower activation energy for the diffusion of oxygen Diffusion limited through these grain boundaries CEA 16 th May 2016 PAGE 21
22 b c y a z x RESULTS GRAIN BOUNDARY MISORIENTATION Misorientation angles distribution between adjacent grains Not randomly distributed 90 [001] b y z x c a a y x b z c a c 180 [101] x b z y Lower activation energy for the diffusion of oxygen Low coherent misorientation angles + 32 % in ZrO 2 / Hydride Twins tetragonal to monoclinic phase transformation Diffusion limited through these grain boundaries CEA 16 th May 2016 PAGE 22 Participate to the higher corrosion kinetic of the massive hydride
23 RESULTS OXYGEN DIFFUSION SIMULATION Oxygen diffusion experiments [Bisor C. Phd (2010)] ZrO 2 Isotopic exposure in H 2 18 O 6h; 360 C; 190 bars SIMS profile of 18 O After 6 h ZrO 2 / Zy4 ZrO 2 / Hydride Diffusion profile of 18 O characteristic of a diffusion through short-circuits (grain boundaries) Influence of the columnar grain width? Second Fick's law: x 18O = x s x s x 0. erf x 18O = x s for x = 0 x 18O = x 0 for x = x 2 D a t 18 O apparent diffusion coefficient Da ratio : D ZrO2/Hydride = D ZrO2/Zy4 = cm²/s cm²/s = +80% CEA 16 th May 2016 PAGE 23
24 RESULTS OXYGEN DIFFUSION SIMULATION ZrO 2 Modelisation with Voronoï cell aggregate ZrO 2 / Zy4 ZrO 2 / Hydride Conditions - Sample ZrO2 / Zy4 ZrO2 / Hydride Average grain size 34.6 nm 27.8 nm Voronoï cells aggregates - Thickness of the grain boundaries : 0.5 nm - Diffusion coefficient of oxygen in: Volume : cm²/s Grain boundaries : 4.3x10-14 cm²/s CEA 16 th May 2016 PAGE 24
25 RESULTS OXYGEN DIFFUSION SIMULATION ZrO 2 Modelisation with Voronoï cells aggregate Fickian diffusion solved with the finite element Cast3M, during 6h www-cast3m.cea.fr 18 O apparent diffusion coefficient Da ratio : D ZrO2/Hydride = D ZrO2/Zy4 = Simulated: cm²/s cm²/s = +30% Lower than experience: +80% After 6 h Num ZrO 2 / Zy4 Num ZrO 2 / hydride ZrO 2 / Zy4 ZrO 2 / hydride Experience Simulation Confirms the diffusion process occurs mainly through the grain boundaries Confirms an effect of the grain size lower ratio of diffusion coefficient CEA 16 th May 2016 PAGE 25
26 RESUME Precipitation of a massive hydride on the surface (+ 80%) Higher corrosion kinetic PWR conditions Zircaloy-4 Modification of the grains boundary components of the monoclinic oxide layer Higher (+ 60%) concentration lower grain size distribution Less coherence of the misorientation angles distribution between adjacent grains Increase the diffusion kinetics of oxygen through the oxide layer The simulation with Cast3M confirms the role of the grain boundaries associated to a lower grain size distribution To be improved ACOM-TEM Informations on the oxide microstructure - grain size - grain boundary component - grain orientation (texture) CEA 16 th May 2016 PAGE 26
27 THANK YOU PAGE 27 CEA 10 AVRIL MAI 2016 Commissariat à l énergie atomique et aux énergies alternatives Centre de Saclay Gif-sur-Yvette Cedex T. +33 (0) Etablissement public à caractère industriel et commercial RCS Paris B DEN DMN SEMI
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 informationMICROSTRUCTURAL EVOLUTION OF Q12 ALLOY IRRADIATED IN PWR AND COMPARISON WITH OTHER Zr BASE ALLOYS
MICROSTRUCTURAL EVOLUTION OF Q12 ALLOY IRRADIATED IN PWR AND COMPARISON WITH OTHER Zr BASE ALLOYS Authors: S. Doriot, B. Verhaeghe, A. Soniak, P. Bossis, D. Gilbon, V. Chabretou, J. P. Mardon, M. Ton-That,
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 informationImpact of hydrogen pick up and applied stress on c component loops: radiation induced growth of recrystallized zirconium alloys
Impact of hydrogen pick up and applied stress on c component loops: Toward a better understanding of the radiation induced growth of recrystallized zirconium alloys L. Tournadre 1, F. Onimus 1, J.L. Béchade
More informationAdvanced Zirconium Alloy for PWR Application
Advanced Zirconium Alloy for PWR Application Anand Garde Westinghouse Nuclear Fuel Columbia, South Carolina, 29209, USA 16 th Zr International Symposium Chengdu, China, May 9-13, 2010 1 Outline Advanced
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 informationMECHANICAL BEHAVIOR AT HIGH TEMPERATURE OF HIGHLY OXYGEN- OR HYDROGEN- ENRICHED α AND (PRIOR-) β PHASES OF ZIRCONIUM ALLOYS
MECHANICAL BEHAVIOR AT HIGH TEMPERATURE OF HIGHLY OXYGEN- OR HYDROGEN- ENRICHED α AND (PRIOR-) β PHASES OF ZIRCONIUM ALLOYS I. Turque 1,2, R. Chosson 1,2,3, M. Le Saux 1*, J.C. Brachet 1, V. Vandenberghe
More informationOxidation Mechanisms in Zircaloy-2 - The Effect of SPP Size Distribution
Oxidation Mechanisms in Zircaloy-2 - The Effect of SPP Size Distribution Pia Tejland 1,2, Hans-Olof Andrén 2, Gustav Sundell 2, Mattias Thuvander 2, Bertil Josefsson 3, Lars Hallstadius 4, Maria Ivermark
More informationStudy of Structure-Phase State of Oxide Films on E110 and E635 Alloys at Pre- and Post-Irradiation Stages
A.A. BOCHVAR HIGH-TECHNOLOGY RESEARCH INSTITUTE OF INORGANIC MATERIALS (SC «VNIINM») 18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY «ROSATOM» STATE ATOMIC ENERGY CORPORATION MAY 15-19,
More informationCHARACTERIZATION OF OXYGEN DISTRIBUTION IN LOCA SITUATIONS
CHARACTERIZATION OF OXYGEN DISTRIBUTION IN LOCA SITUATIONS Duriez C. 1, Guilbert S. 1, Stern A. 2, Grandjean C. 1, Bělovský L. 3, Desquines J. 1 1 IRSN ² IRSN post-doctorate, now at CEA 3 ALIAS Cz Scope
More informationThe Pennsylvania State University. The Graduate School. Department of Mechanical and Nuclear Engineering
The Pennsylvania State University The Graduate School Department of Mechanical and Nuclear Engineering TEM EXAMINATION OF OXIDES FORMED ON ZIRCONIUM ALLOYS A Thesis in Nuclear Engineering by Benoît Le
More informationImpact of the irradiation damage recovery during transportation on the subsequent room temperature tensile behavior of irradiated zirconium alloys
Impact of the irradiation damage recovery during transportation on the subsequent room temperature tensile behavior of irradiated zirconium alloys B. Bourdiliau 1, F. Onimus 2, C. Cappelaere 1, V. Pivetaud
More informationsteam oxidation and post-quench mechanical
Effect of pre-oxide on Zircaloy-4 4high htemperature t steam oxidation and post-quench mechanical properties Guilbert S., Lacote P., Montigny G., Duriez C., Desquines J., Grandjean C. Institut de Radioprotection
More informationPost Quench Ductility of Zirconium Alloy Cladding Materials
Post Quench Ductility of Zirconium Alloy Cladding Materials A. Mueller D. Mitchell J. Romero* A. Garde J. Partezana A. Atwood G. Pan 1 18 th International Symposium on Zirconium in the Nuclear Industry
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 informationASTM 18 TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY, MAY TH 2016, HILTON HEAD, SC, USA
MICROSTRUCTURE EVOLUTION IN ION- IRRADIATED OXIDIZED ZIRCALOY-4 STUDIED WITH SYNCHROTRON RADIATION MICRO-DIFFRACTION AND TRANSMISSION ELECTRON MICROSCOPY K.COLAS, R. VERLET, M. TUPIN, M. JUBLOT Département
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 informationPost-Irradiation analysis of fission gases in nuclear fuels
Post-Irradiation analysis of fission gases in nuclear fuels Ch. VALOT, J. NOIROT, Y. PONTILLON MINOS Workshop, Materials Innovation for Nuclear Optimized Systems December 5-7, 212, CEA INSTN Saclay, France
More informationDamage Build-up in Zirconium Alloys Mechanical Processing and Impacts on Quality of the Cold Pilgering Product
Damage Build-up in Zirconium Alloys Mechanical Processing and Impacts on Quality of the Cold Pilgering Product ASTM 16th International Symposium on Zirconium in the Nuclear Industry Chengdu, China, 10-05-2010
More informationStudy of the Initial Stage and an Anisotropic Growth of Oxide Layers Formed on Zircaloy-4
16 th International Symposium on Zirconium in the Nuclear Industry, Chengdu, P. R. China, May 10-13, 2010 Study of the Initial Stage and an Anisotropic Growth of Oxide Layers Formed on Zircaloy-4 B. X.
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 informationNeutron Irradiation Effects on the Corrosion of Zircaloy-4 in a PWR Environment
Neutron Irradiation Effects on the Corrosion of Zircaloy-4 in a PWR Environment B.F. Kammenzind, J.A. Gruber, R. Bajaj, J.D. Smee Bechtel Marine Propulsion Corporation Bettis Laboratory Knolls Laboratory
More informationTransmission electron microscopy examination of oxide layers formed on Zr alloys
Journal of Nuclear Materials 349 (2006) 265 281 www.elsevier.com/locate/jnucmat Transmission electron microscopy examination of oxide layers formed on Zr alloys Aylin Yilmazbayhan a, Else Breval a, Arthur
More informationTrapping of Hydrogen at Irradiation Induced Defects
Trapping of Hydrogen at Irradiation Induced Defects B.F. Kammenzind, W.J. Duffin (Retired) Bechtel Marine Propulsion Corporation Bettis Laboratory Knolls Laboratory 18 th International Symposium on Zirconium
More informationFOR LIFETIME EXTENSION»
PAGE «ROLE OF MATERIALS FOR LIFETIME EXTENSION» Pascal YVON, Bernard MARINI and Benoit TANGUY Department of Materials for Nuclear Applications, CEA SACLAY OUTLINE Context Effect of neutrons on materials
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 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 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 informationTHE NUMODIS PROJECT DECEMBER 10, A. Etcheverry, P. Blanchard, O. Coulaud, M. Bletry, M. Fivel, E. Ferrié, L. Dupuy and many others
THE NUMODIS PROJECT A. Etcheverry, P. Blanchard, O. Coulaud, M. Bletry, M. Fivel, E. Ferrié, L. Dupuy and many others International Workshop on Dislocation Dynamics Simulations Laurent Dupuy DECEMBER 10,
More informationDeviations from the parabolic kinetics during oxidation
Deviations from the parabolic kinetics during oxidation of zirconium alloys Martin Steinbrück, Mirco Große Karlsruhe Institute of Technology,, Germany 17th International ti lsymposium on Zirconium i in
More informationCORRELATIONS BETWEEN OXIDE STRUCTURE, IRON DISTRIBUTIONS, AND ZIRCONIUM OXIDE GROWTH. Yan Dong
CORRELATIONS BETWEEN OXIDE STRUCTURE, IRON DISTRIBUTIONS, AND ZIRCONIUM OXIDE GROWTH by Yan Dong A dissertation submitted in partial fulfillment of the requirements for the degree of Doctor of Philosophy
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 informationMicrobeam X-ray Absorption Near-Edge Spectroscopic Studies of High-Burnup Zircaloy-2 Oxide Layers
Microbeam X-ray Absorption Near-Edge Spectroscopic Studies of High-Burnup Zircaloy-2 Oxide Layers A.P. Shivprasad 1, A.T. Motta 1, A. Kucuk 2, S. Yagnik 2, and Z. Cai 3 1 The Pennsylvania State University
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 informationInfluence of irradiation on stainless steel corrosion in PWR primary conditions
EPJ Web of Conferences 115, 04006 (2016) DOI: 10.1051/epjconf/201611504006 Owned by the authors, published by EDP Sciences, 2016 2 nd Int. Workshop Irradiation of Nuclear Materials: Flux and Dose Effect
More informationMICROSTRUCTURE AND PROPERTIES OF A 3-LAYERS NUCLEAR FUEL CLADDING PROTOTYPE CONTAINING ERBIUM AS A NEUTRONIC BURNABLE POISON.
MICROSTRUCTURE AND PROPERTIES OF A 3-LAYERS NUCLEAR FUEL CLADDING PROTOTYPE CONTAINING ERBIUM AS A 1 CEA- DEN, Department for Nuclear Materials, Section for Applied Metallurgy Research, CEA-Saclay, F-91191
More informationPhysical Properties. Can increase the strength by cold working but the recrystallization temperature is 400 to 500 C
Zirconium Cladding Why? Physical Properties Corrosion Resistance Radiation Effects ----------------------------------------------- In the early 1950Õs the Navy was looking for a material with low σ a high
More informationLecture 20: Eutectoid Transformation in Steels: kinetics of phase growth
Lecture 0: Eutectoid Transformation in Steels: kinetics of phase growth Today s topics The growth of cellular precipitates requires the portioning of solute to the tips of the precipitates in contact with
More informationNew characterizations at the MARS* beamline (SOLEIL synchrotron radiation)
New characterizations at the MARS* beamline (SOLEIL synchrotron radiation) J.-L. Béchade 1, D. Menut 1, B. Sitaud 2, S. Schlutig 2, I. Llorens 2, M.-L. Lescoat 1, J. Ribis 1, N. Jonquères 3, D. Leterme
More informationFeasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production
Idaho National Engineering and Environmental Laboratory Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production Nuclear Energy Research Initiative
More informationCOMPARISON OF THE MECHANICAL PROPERTIES AND CORROSION RESISTANCE OF ZIRLO AND OTHER ZIRCONIUM ALLOYS
2007 International Nuclear Atlantic Conference - INAC 2007 Santos, SP, Brazil, September 30 to October 5, 2007 ASSOCIAÇÃO BRASILEIRA DE ENERGIA NUCLEAR - ABEN ISBN: 978-85-99141-02-1 COMPARISON OF THE
More informationEffects of Electric Field Treatment on Corrosion Behavior of a Ni-Cr-W-Mo Superalloy
Materials Transactions, Vol. 50, No. 7 (2009) pp. 1644 to 1648 Special Issue on New Functions and Properties of Engineering Materials Created by Designing and Processing #2009 The Japan Institute of Metals
More informationInfluence of Alloy Microstructure on Oxide Growth in HCM12A in Supercritical Water
Mater. Res. Soc. Symp. Proc. Vol. 1125 2009 Materials Research Society 1125-R06-05 Influence of Alloy Microstructure on Oxide Growth in HCM12A in Supercritical Water Jeremy Bischoff 1, Arthur T. Motta
More informationMechanisms of Hydride Reorientation in Zircaloy-4 Studied In-Situ
Mechanisms of Hydride Reorientation in Zircaloy-4 Studied In-Situ 500 µm K. B. Colas 1 *, A. T. Motta 1, M. R. Daymond 2, J. D. Almer 3, 1. Department of Mechanical and Nuclear Engineering, Penn State
More informationIrradiation embrittlement of austenitic stainless steels in PWR vessel s internals. Experiments and modelling from micro to mesoscale
Irradiation embrittlement of austenitic stainless steels in PWR vessel s internals Experiments and modelling from micro to mesoscale Benoit Tanguy, J. Hure, X. Han, C. Ling, P-O Barrioz, in collaboration
More informationCHARACTERIZATION OF SILICON CARBIDE AND PYROCARBON COATINGS FOR FUEL PARTICLES FOR HIGH TEMPERATURE REACTORS (HTR)
CHARACTERIZATION OF SILICON CARBIDE AND PYROCARBON COATINGS FOR FUEL PARTICLES FOR HIGH TEMPERATURE REACTORS (HTR) D. Hélary 1,2, X. Bourrat 1, O. Dugne 2, G. Maveyraud 1,2, F. Charollais 3, M. Pérez 4,
More informationModelling of the contamination transfer in nuclear reactors: The OSCAR code Applications to SFR and ITER
Modelling of the contamination transfer in nuclear reactors: The OSCAR code Applications to SFR and ITER F. Dacquait, J.B. Génin, L. Brissonneau CEA/DEN/Cadarache www.cea.fr 1 st IAEA Workshop on Challenges
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 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 informationTheoretical Modeling of Protective Oxide Layer Growth in Non-isothermal Lead-Alloys Coolant Systems: Quarterly Progress Report (01/01/05-03/31/05)
Transmutation Sciences Materials (TRP) Transmutation Research Program Projects 3-31-005 Theoretical Modeling of Protective Oxide Layer Growth in Non-isothermal Lead-Alloys oolant Systems: Quarterly Progress
More informationAdditive Element Effects on Electronic Conductivity of Zirconium Oxide Film
Journal of NUCLEAR SCIENCE and TECHNOLOGY, 31[6], pp. 546~551 (June 1994). Additive Element Effects on Electronic Conductivity of Zirconium Oxide Film Yusuke ISOBE, Motomasa FUSE and Kinya KOBAYASHI Energy
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 informationUranium corrosion. Dr N.Harker NRC-UK PONI Nuclear futures conference
Uranium corrosion Dr N.Harker NRC-UK PONI Nuclear futures conference Corrosion in air Oxide Uranium Uranium U + ( 2+x ) O 2 UO 2+x 2 Corrosion in air Oxygen diffuses to reach the metal UO 2 O 2 O 2- New
More informationOxford University Materials 2005
Oxford University Materials 2005 Fusion materials Fission materials Materials for Fusion and Fission Power 2013 Fusion materials Fission materials CCFE Materials for Fission & Fusion Power Steve Roberts
More informationCharacterization of Nanoscale Electrolytes for Solid Oxide Fuel Cell Membranes
Characterization of Nanoscale Electrolytes for Solid Oxide Fuel Cell Membranes Cynthia N. Ginestra 1 Michael Shandalov 1 Ann F. Marshall 1 Changhyun Ko 2 Shriram Ramanathan 2 Paul C. McIntyre 1 1 Department
More informationDESTRUCTIVE EXAMINATION OF EXPERIMENTAL CANDU FUEL ELEMENTS IRRADIATED IN TRIGA-SSR REACTOR
DESTRUCTIVE EXAMINATION OF EXPERIMENTAL CANDU FUEL ELEMENTS IRRADIATED IN TRIGA-SSR REACTOR S. IONESCU, O. UTA, C. GENTEA, M. MINCU, M. PARVAN, L. DINU Institute for Nuclear Research Pitesti - Romania
More informationFrench R&D program on SFR and the
PRESENT STATUS OF FRENCH PROGRAM ASTRID JAPAN FRANCE COOPERATION EXPECTATIONS ON MONJU MONJU ASTRID ALAIN PORRACCHIA DIRECTOR FOR INNOVATION AND INDUSTRIAL SUPPORT CEA/DEN French R&D program on SFR and
More informationChapter 2. Literature Review. 2.1 Overview
Chapter 2 Literature Review 2.1 Overview This chapter gives a comprehensive review on alloying elements, impurity elements, uniform oxidation, localized corrosion and hydrogen intake behaviour of important
More informationJULES HOROWITZ REACTOR (JHR)
JULES HOROWITZ REACTOR (JHR) RCC-MRx [1] APPLICABILITY FOR THE DESIGN PHASE OF EXPERIMENTAL DEVICES CEA, DEN, DTN - Nuclear Technology Department, CEA Cadarache FRANCE Sébastien GAY Mechanical Design Leader
More informationM. Négyesi, J. Krejčí, S. Linhart, L. Novotný, A. Přibyl, J. Burda, V. Klouček, J. Lorinčík, J. Sopoušek, J. Adámek, J. Siegl, V.
CHEMCOMEX, NRI, UNIPETROL, AS, JEPU, MU Contribution to the Study of the Pseudobinary Zr1Nb O Phase Diagram and Its Application to Numerical Modeling of the High Temperature Steam Oxidation of Zr1Nb Fuel
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 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 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 informationZirconium in the Nuclear Industry: Thirteenth International Symposium
STP 1423 Zirconium in the Nuclear Industry: Thirteenth International Symposium Gerry D. Moan and Peter Rudling, editors ASTM Stock #: STP1423 ASTM International 100 Barr Harbor Drive West Conshohocken,
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 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 informationHydriding Induced Corrosion Failures in BWR Fuel
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
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 informationDRAFT: SEVERE FUEL DAMAGE EXPERIMENTS WITH ADVANCED CLADDING MATERIALS TO BE PERFORMED IN THE QUENCH FACILITY (QUENCH-ACM)
Proceedings of the 16th International Conference on Nuclear Engineering ICONE16 May 11-15, 2008, Orlando, Florida, USA ICONE16-48074 DRAFT: SEVERE FUEL DAMAGE EXPERIMENTS WITH ADVANCED CLADDING MATERIALS
More informationCHAPTER 7. Conclusions, Summary and Scope for Future Work
CHAPTER 7 Conclusions, Summary and Scope for Future Work Conclusions, Summary and Scope for Future Study This Chapter gives the salient conclusions drawn from the results of the investigations carried
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 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 informationEffects of chemistry and microstructure on corrosion performance of Zircaloy-2 based BWR cladding
Effects of chemistry and microstructure on corrosion performance of Zircaloy-2 based BWR cladding Yang-Pi Lin, David White, Dan Lutz, Global Nuclear Fuel Americas ASTM 18th International Symposium on Zirconium
More informationOxide Surface Peeling of Advanced
Oxide Surface Peeling of Advanced Zirconium Alloy Cladding Oxides after High Burnup Irradiation A.M.Garde, G.Pan, A.J.Mueller & L.Hallstadius Westinghouse Electric Company Platform Presentation at the
More informationTRIBOCORROSION MECHANISM STUDY OF STELLITE-6 AND ZIRCALOY-4 A COMPARISON IN LiOH-H 3 BO 3 SOLUTIONS
THE ANNALS OF UNIVERSITY DUNĂREA DE JOS OF GALAŢI 35 TRIBOCORROSION MECHANISM STUDY OF STELLITE-6 AND ZIRCALOY-4 A COMPARISON IN LiOH-H 3 BO 3 SOLUTIONS Lidia BENEA 1, Viorel-Eugen IORDACHE 2, François
More informationNEW TEXTILE STRUCTURES AND FILM-BOILING DENSIFICATION FOR SIC/SIC COMPONENTS (IV Gen.)
NEW TEXTILE STRUCTURES AND FILM-BOILING DENSIFICATION FOR SIC/SIC COMPONENTS (IV Gen.) P. David 1, J. Blein 1, D. Rochais 1, Y. Pierre 1, M. Zabiego 2 Commissariat à l énergie atomique et aux énergies
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 informationIrradiation-Assisted Stress Corrosion Cracking
Irradiation-Assisted Stress Corrosion Cracking Elin Toijer 1;2 Pär Olsson 1 Mats Jonsson 2 1 Reactor Physics KTH Stockholm 2 Applied Physical Chemistry KTH Stockholm SKC Symposium, 2015-10-09 Elin Toijer
More informationInterfaces: Corrosion in Pb-alloy cooled nuclear reactors and advanced mitigation measures
Interfaces: Corrosion in Pb-alloy cooled nuclear reactors and advanced mitigation measures and G. Müller KIT KIT Universität des Landes Baden-Württemberg und nationales Forschungszentrum in der Helmholtz-Gemeinschaft
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 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 informationIGSCC OF INCONEL 718 HOLD DOWN SPRING IN 400 O C STEAM ENVIRONMENT
IGSCC OF INCONEL 718 HOLD DOWN SPRING IN 400 O C STEAM ENVIRONMENT H. Jang 1, J.D. Hong 1, S.S. Kim 2, K.B. Eom 2, O.H. Kwon 2, C. Jang 1 1 Department of Nuclear and Quantum Engineering, KAIST, Daejeon,
More informationCladding embrittlement, swelling and creep
Cladding embrittlement, swelling and creep Workshop on radiation effects in nuclear waste forms and their consequences for storage and disposal, 12-16 September 2016, Trieste, Italy Scope Spent fuel, the
More informationPhase field modeling of Microstructure Evolution in Zirconium base alloys
Phase field modeling of Microstructure Evolution in Zirconium base alloys Gargi Choudhuri, S.Chakraborty, B.K.Shah, D. Si Srivastava, GKD G.K.Dey Bhabha Atomic Research Centre Mumbai, India- 400085 17
More informationJoint Research Centre
Joint Research Centre the European Commission's in-house science service Serving society Stimulating innovation Supporting legislation Dissolution rate of MOX and Cr-doped UO 2 fuel D. H. Wegen JRC-ITU
More informationElectronic band structure of photoanode heterojunctions dedicated to water splitting
Electronic band structure of photoanode heterojunctions dedicated to water splitting M. Rioult 1, H. Magnan 1, D. Stanescu 1, P. Le Fèvre 2, A. Barbier 1, 1 : CEA Saclay / DSM / IRAMIS / SPEC / LISO, F
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 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 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 informationavailable online at MODELLING OF NUCLEAR FUEL CLADDING TUBES CORROSION Martin Sevecek a,b
Acta Polytechnica CU Proceedings 4:13 18, 2016 Czech echnical University in Prague, 2016 doi:10.14311/ap.2016.4.0013 available online at http://ojs.cvut.cz/ojs/index.php/app MODELLING OF NUCLEAR FUEL CLADDING
More informationGENERATION-IV SODIUM-COOLED FAST REACTORS AND THE ASTRID PROJECT
GENERATION-IV SODIUM-COOLED FAST REACTORS AND THE ASTRID PROJECT French-Swedish Seminar on Future Nuclear Systems Pascal Anzieu pascal.anzieu@cea.fr KTH, STOCKHOLM, DECEMBER 3, 2013 ESNII SNETP ESNII European
More informationIs Spent Nuclear Fuel Immune from Delayed Hydride Cracking (DHC) during Dry Storage? An IAEA Coordinated Research Project
Is Spent Nuclear Fuel Immune from Delayed Hydride Cracking (DHC) during Dry Storage? An Coordinated Research Project C. Coleman, V. Markelov, M. Roth, V. Makarevicius, Z. He, J.K. Chakravartty, A.M. Alvarez-Holston,
More informationPREVENTION OF SCC OCCURRING IN A EXPANSION TRANSITION REGION OF STEAM GENERATOR TUBING BY Ni-PLATING IN PWRS
PREVENTION OF SCC OCCURRING IN A EXPANSION TRANSITION REGION OF STEAM GENERATOR TUBING BY Ni-PLATING IN PWRS J. S. Kim, M. J. Kim, D. J. Kim, H. P. Kim Korea Atomic Energy Research Institute(KAERI), Korea
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 informationAtomic Scale Degradation of Zirconium Alloys for Nuclear Applications
THESIS FOR THE DEGREE OF DOCTOR OF ENGINEERING Atomic Scale Degradation of Zirconium Alloys for Nuclear Applications GUSTAV SUNDELL Department of Applied Physics CHALMERS UNIVERSITY OF TECHNOLOGY Göteborg,
More informationOxidation of Chromium
Oxidation of Chromium Oxidation of chromium is very simple as it usually forms a single oxide Cr 2 O 3, It is a p-type of oxide with Cr 3+ ions diffusing outward. Since the defect concentration is so low
More informationCLADDING PROPERTIES AFTER HIGH TEMPERATURE OXIDATION (Session 3) Chairperson. G. HACHE France
CLADDING PROPERTIES AFTER HIGH TEMPERATURE OXIDATION (Session 3) Chairperson G. HACHE France MECHANICAL BEHAVIOUR AT ROOM TEMPERATURE AND METALLURGICAL STUDY OF LOW-TIN ZY-4 AND M5 TM (ZR-NbO) ALLOYS
More informationBurn up Analysis for Fuel Assembly Unit in a Pressurized Heavy Water CANDU Reactor
Burn up Analysis for Fuel Assembly Unit in a Pressurized Heavy Water CANDU Reactor A. A. EL-Khawlani a, Moustafa Aziz b, M. Ismail c and A. Y. Ellithi c a Physics Department, Faculty of Science, High Education,
More informationBUBBLE FORMATION IN ZR ALLOYS UNDER HEAVY ION IMPLANTATION
BUBBLE FORMATION IN ZR ALLOYS UNDER HEAVY ION IMPLANTATION Luciano Pagano, Jr. 1, Arthur T.Motta 1 and Robert C. Birtcher 2 1. Dept. of Nuclear Engineering, Pennsylvania State University, University Park,
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 information