Committee B10 on Reactive and Refractory Metals and Alloys

    PRESENTATIONS

    1.1 Phase Field Modeling of Microstructure Evolution in Zirconium Base Alloy
    G. Choudhuri, S.Chakraborty, B. K. Shah, D. Srivastava, and G. K. Dey, Bhabha Atomic Research Centre, Mumbai, India

    1.3 Influence of Sn on Deformation Mechanisms and Texture Evolution during Cold Deformation of Binary Zr-Alloys
    K. V. Mani Krishna, D. Srivastava, G. K. Dey, Bhabha Atomic Research Centre, Trombay, Mumbai, India; D G Leo Prakash, J. Quinta da Fonseca, and M. Preuss, The University of Manchester, Manchester, UK; and N. Saibaba, Nuclear Fuel Complex, Hyderabad, India

    1.5 Microstructure and Properties of a 3-Layers Nuclear Fuel Cladding Prototype Containing Erbium as a Neutronic Burnable Poison
    J.C. Brachet, P. Olier, V. Vandenberghe, S. Urvoy, D. Hamon, T. Guilbert, A. Mascaro, C. Toffolon-Masclet, M. Tupin and B. Bourdiliau, CEA- DEN, Gif-sur-Yvette Cedex, France; C. Raepsaet, CEA-DSM, Gif-sur-Yvette Cedex, France; J.M. Joubert; Université Paris-Est, Thiais, France; J.L. Aubin, AREVA-CEZUS, Paimboeuf, France

    2.1 Identification of Safe Hot Working Conditions in Cast Zr-2.5Nb
    J.K. Chakravartty, R. Kapoor, and A. Sarkar, Bhabha Atomic Research Centre, Mumbai, India; S. K. Jha and N. Saibaba, Nuclear Fuel Complex, Hyderabad, India; V. Kumar, RDCIS, Steel Authority of India Limited, Ranchi, Jharkhand, India; and S. Banerjee, Department of Atomic Energy, Mumbai, India

    2.2 Simulation and Experimental Validation of Extrusion of Zr-2.5 Nb Alloy Pressure Tube
    N. Saibaba, V. Raizada and V. Kumar, Nuclear Fuel Complex, Hyderabad, India; and G.K. Dey and N. Keskar, Bhabha Atomic Research Center, Mumbai, India

    2.3 Study on Effect of Processing on Texture Development in Zirconium-2.5% Niobium Alloy Tubes
    K. Kapoor, S. V. Ramana Rao, K. Itisri, B. Prahlad, and N. Saibaba, Nuclear Fuel Complex ECIL(PO), Hyderabad, India

    2.5 Application of Coating Technology on the Zirconium-Based Alloy to Decrease High Temperature Oxidation
    H.-G. Kim, I.-H. Kim, J.-Y. Park and Y.-H. Koo, KAERI, Daejeon, Republic of Korea

    2.6 Development of Composite Zirconium Materials with Increased Level of Properties and Protective Layers for New Generation LWR Active Core Components
    S.V. Ivanova and E.M. Glagovsky, Nuclear Industrial Technology Institute, Moscow, Russia; V.K. Orlov, I.A. Shlepov, K.Y. Nikonorov, V. V. Rozhko, A.A. Bochvar High-Technology Research Institute of Inorganic Materials, Moscow, Russia

    3.1 Oxidation Mechanisms in Zircaloy-2 – The Effect of SPP Size Distribution
    P. Tejland, H.-O. Andrén, G. Sundell and M. Thuvander, Chalmers University of Technology, Gothenburg, Sweden; B. Josefsson, Vattenfall Nuclear Fuel AB, Stockholm, Sweden; and L. Hallstadius, M. Ivermark, and M. Dahlbäck, Westinghouse Electric Sweden, Västerås, Sweden

    3.2 The Effect of Sn on Corrosion Mechanisms in Advanced Zr-Cladding for Pressurized Water Reactors
    P. Frankel, J. Wei, J. Smith, S. Lyon, R. A. Cottis, and M. Preuss, University of Manchester, Manchester, UK; N. Ni, S. Lozano-Perez, K. L. Moore, S. S. Yardley, C. R. M. Grovenor, G. Smith, and J. Sykes, University of Oxford, Oxford, UK; A. Ambard and M. Blat-Yrieix EDF, Moret sur Loing, France; R. J. Comstock, Westinghouse Electric Co., Pittsburgh, PA, USA; and L. Hallstadius, Westinghouse Electric Sweden AB, Västerås, Sweden

    3.3 Understanding of Corrosion Mechanisms under Irradiation: Effect of Ion Irradiation of the Oxide Layers on the Corrosion Rate of Zirconium Alloys
    M. Tupin, J. Hamann, D. Cuisinier, P. Bossis, CEA-DEN, CEA-Saclay, Gif-sur-Yvette, France; M. Blat, A. Ambard, EDF, Moret-sur-Loing, France; A. Miquet, EDF, EDF/SEPTEN Villeurbanne France; D. Kaczorowski, AREVA, AREVA NP, Lyon, France; and F. Jomard, CNRS UMR 8635, Meudon, France

    4.1 Microstructure Characterization of ZIRLO™ Structural Components Irradiated to High Burnup
    J. M. García-Infanta, and M. Aulló, ENUSA Industrias Avanzadas, Madrid, Spain; D. Schrire, Vattenfall, Nuclear Fuel AB, Stockholm, Sweden; F. Culebras, Asociación Nuclear Ascó-Vandellós II Edificio Sede L´Hospitalet del Tarragona, Spain; and A. M. Garde, Westinghouse Electric Company, Hopkins, SC, USA

    4.2 Performance and Property Evaluation of High Burnup Optimized ZIRLO™ Cladding
    G. Pan, A. M. Garde, and A. R. Atwood, Westinghouse Electric Company, Hopkins, SC, USA

    4.3 Corrosion and Structure VVER-1000 FA Components from E635 Alloy at Burnups up to 72 MW·DAY/KGU
    V. N. Shishov, M. M. Peregud, A. Yu. Shevyakov, V. A. Markelov, A. V. Nikulina, and V. V. Novikov, JSC VNIINM, Moscow, Russia; I. N. Volkova, A. E. Novoselov, G. P. Kobylyansky, and A. V. Obukhov, JSC SCC NIIAR, Dimitrovgrad, Russia

    4.4 Corrosion and Hydriding Model for Zircaloy-2 Pressure Tubes of Indian PHWRs
    S. K. Sinha and R. K. Sinha, Bhabha Atomic Research Centre, Mumbai, India

    4.5 Oxide Surface Peeling of Advanced Zirconium Alloy Cladding Oxides after High Burnup Irradiation
    A. M. Garde and G. Pan, Westinghouse Electric Company, Hopkins, SC, USA; A. J. Mueller, Westinghouse Electric Company, Pittsburgh, PA, USA; and L. Hallstadius, Westinghouse Electric Sweden AB, Vasteras, Sweden

    4.6 The Effects of Microstructure and Operating Conditions on Irradiation Creep of Zr-2.5%Nb Pressure Tubing
    L. Walters, G. Bickel and M. Griffiths, Atomic Energy of Canada Limited, ON, Canada

    5.1 Breakthrough in Understanding Radiation Growth of Zirconium
    S. I. Golubov and R. E. Stoller, ORNL, Oak Ridge, TN, USA; A.V. Barashev, University of Tennessee, Knoxville, TN, USA; B. N. Singh, Risø National Laboratory, Roskilde, Denmark

    5.2 Microstructural Evolution of M5™ Alloy Irradiated in PWRs Up to High Fluencies -Comparison with Others Zr Base Alloys
    S. Doriot, J. L. Béchade, D. Menut, B. Verhaeghe and D. Gilbon, CEA-DEN, CEA/Saclay, Gif-sur-Yvette, France; J.-P. Mardon, and J. M. Cloué, AREVA, AREVA NP, Lyon, France;A. Miquet, Electricité de France- DIN Septen, Villeurbanne, France; L. Legras, Electricité de France, Moret sur Loing, France

    5.3 The Influence of Fast Neutron Irradiation on the Plasticity Induced Evolution of Dislocation Densities and Operating Deformation Modes in Zr-2.5Nb
    F. Long, M. R. Daymond and R. A. Holt, Queen’s University, Kingston, ON, Canada;L. Balogh and D.W. Brown, Los Alamos National Laboratory, Los Alamos, NM, USA; and C. D. Judge, Atomic Energy Canada Ltd, Chalk River, ON, Canada

    5.4 Modeling Irradiation Damage in Zr-2.5Nb and its Effects on Delayed Hydride Cracking Growth Rate
    G. A. Bickel, M. Griffiths, H. Chaput, A. Buyers, and C. E. Coleman, AECL, Chalk River, ON, Canada

    5.5 Understanding the Drivers of in-reactor Growth of β-quenched Zircaloy‑2 BWR Channels
    M. Dahlbäck, L. Hallstadius and M. Ivermark, Westinghouse Electric Sweden AB, Västerås, Sweden; J. Romero, Westinghouse Electric Company LLC, Hopkins, SC, USA; and G. Ledergerber, Kernkraftwerk Leibstadt AG, Leibstadt, Switzerland

    5.6 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, F. Onimus, J.-L. Bechade and D. Gilbon, CEA-DEN, Gif-Sur-Yvette, France; J.-M. Cloue and J.-P. Mardon, AREVA NP SAS Fuels, Lyon,  France; and X. Feaugas, Université de La Rochelle, La Rochelle, France

    5.7 Effect of Hydrogen on Dimensional Changes of Zirconium and the Influence of Alloying Elements: First-principles and Classical Simulations of Point Defects, Dislocation Loops, and Hydrides
    W. Wolf, M. Christensen, C. Freeman, and E. Wimmer, Materials Design, Inc., Santa Fe, NM, USA; R. Adamson, Zircology Plus, Fremont, CA, USA; L. Hallstadius, Westinghouse Electric Sweden AB, Vasteras, Sweden; P. Cantonwine, Global Nuclear Fuel, Wilmington, NC, USA; and E. Mader, Electric Power Research Institute, Palo Alto, CA, USA

    6.1 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 Cladding
    M. Négyesi, J. Adámek and J. Siegl, Czech Technical University in Prague, Prague Czech Republic; S. Linhart, L. Novotný, A. Přibyl and V. Vrtílková, UJP PRAHA, Prague, Czech Republic; J. Krejčí, CHEMCOMEX, and Czech Technical University in Prague Praha - Zbraslav, Czech Republic; J. Burda, NRI Rez plc, Řež, Czech Republic; V. Klouček and J. E. Purkinje, University and UNIPETROL RPA, Litvinov, Czech Republic; J. Lorinčík, Institute of Photonics and Electronics, Academy of Sciences of the Czech Republic; Prague, Czech Republic; and J. Sopoušek, Ústav chemie, Brno, Czech Republic

    6.2 Effect of Pre-Oxide on Zircaloy-4 High Temperature Steam Oxidation and Post-Quench Mechanical Properties
    S. Guilbert, P. Lacote, G. Montigny, C. Duriez, J. Desquines and C. Grandjean, Institut de Radioprotection et de Sureté Nucléaire, Saint Paul lez Durance, France

    6.3 Deviations from Parabolic Kinetics during Oxidation of Zirconium Alloys
    M. Steinbrück, M. Grosse, Karlsruhe Institute of Technology, Karlsruhe, Germany

    6.4 Experimental Comparison of the Behavior of E110 and E110G Claddings at High Temperature
    Z. Hózer, E. Perez-Feró, T. Novotny, I. Nagy, M. Horváth, A. Pintér-Csordás, A. Vimi, M. Kunstár, and W. T. Kemény, Hungarian Academy of Sciences, Budapest, Hungary

    6.5 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, V. Vandenberghe, J.C. Brachet, and D. Gilbon, CEA-DEN, Gif-sur-Yvette, France; P. Crébier CEA-DEN, Grenoble, France; J. P. Mardon, AREVA, AREVA NP, Lyon, France; P. Jacques and A. Cabrera, EDF-SEPTEN, Villeurbanne, France

    6.6 Analysis of the Secondary Cladding Hydrogenation during QUENCH-LOCA Tests and its Influence on the Cladding Embrittlement
    M. Grosse, J. Stuckert, C. Roessger, M. Steinbrueck and M. Walter, Karlsruhe Institute of Technology, Karlsruhe, Germany; and A. Kaestner, Paul Scherrer Institute, Villigen, Switzerland

    7.1 Effect of Hydride Distribution on the Mechanical Properties of Zirconium Alloy Fuel Cladding and Guide Tubes
    S. K. Yagnik, EPRI, Palo Alto, CA, USA; J-H Chen and R-C Kuo, INER, Chiaan Village, Lung-tan, Taiwan

    7.2 Mechanisms of Hydride Reorientation in Zircaloy-4 studied In-Situ
    K. Colas and A. Motta, Penn State University, University Park, PA, USA; M. R. Daymond, Queen’s University, Kingston, ON, Canada; and J. Almer, Argonne National Laboratory, Argonne, IL, USA

    7.3 Modeling Zirconium Hydrides Reorientation under Applied Load
    L. Zhang, L. Thuinet, A. Legris, and A. Debacker, Université Lille 1, Villeneuve D'Ascq, France; and M. Blat-Yrieix and A. Ambard, EDF, Moret-sur-Loing, France

    7.4 Hydriding Induced Corrosion Failures in BWR Fuel
    D. Lutz, Global Nuclear Fuel - Americas, Sunol, CA, USA; Y-P Lin, R. Dunavant, R. Schneider and H. Yeager, Global Nuclear Fuel – Americas, Wilmington, NC, USA; A. Kucuk, and B. Cheng, Electric Power Research Institute, Palo Alto, CA, USA; and J. Lemons, Tennessee Valley Authority, Chattanooga, TN, USA

    7.5 Experimental Studies of DHC of Unirradiated and Irradiated Fuel Rod Cladding and Implications to in-pile Operation and Back-End Issues
    V. Grigoriev, A-M. Alvarez and G. Lysell, Studsvik Nuclear AB, Nyköping, Sweden; D. Schrire, Vattenfall Nuclear Fuel AB, Stockholm, Sweden; L. Hallstadius, Westinghouse Electric Sweden AB, Västerås, Sweden; S. T. Mahmood, GNF-A, Vallecitos Nuclear Center, Sunol, CA, USA; and I. Arimescu, AREVA NP Inc., Richland WA, USA


    POSTERS

    01 Macroscopic and Microscopic Behavior of Recrystallized Zircaloy-4 Under Reactivity Initiated Accident Conditions
    E. Bosso and N. Rupin, Electricité De France, Ecuelles, Moret-Sur-Loing,  France; and J. Besson and J. Crépin, Centre des Matériaux Mines Paris, Evry, France

    02 Defect Characterization of Zirconium Alloy Bar Material
    R. K. Chaube, S. A. Acharya, S. K. Reddy, A. Lakshiminarayana, B. Prahlad, N. Saibaba, Nuclear Fuel Complex, Hyderabad, India

    03 Interactions between Gliding Dislocations and Irradiation Induced Loops in Recrystallized Zircaloy-4: in Situ TEM Tensile Tests and Dislocation Dynamic Simulations
    J. Drouet, F. Onimus, L. Dupuy, CEA-Saclay, Gif-Sur-Yvette, France; and F. Mompiou, CEMES-CNRS, Toulouse, France

    04 Micro-Raman Imaging of Oxides Formed at High Temperature on Zircaloy-4 and M5®
    I. Idarraga and C. Duriez, Institut de Radioprotection et de Sûreté Nucléaire (IRSN), St. Paul lez Durance France; M. Mermoux, Laboratoire d’Electrochimie et de Physicochimie des Matériaux et des Interfaces (LEPMI), Saint Martin d’Hères, France; A. Crisci, Science et Ingénierie des Matériaux et Procédés (SIMaP), Saint Martin d’Hères, France; and J.P. Mardon, AREVA, Lyon, France

    05 A Model on Corrosion of Zr-Nb Alloys under in-Pile WWER and PWR Conditions
    T. N. Aliev, I. A. Evdokimov, V. V. Likhanskii, State Research Center of Russian Federation, Troitsk, Moscow Region, Russia; and V. F. Kon’kov, V. A. Markelov, V. V. Novikov, and T. N. Khokhunova, A. A. Bochvar VNIINM, Moscow, Russia

    06 Iron Segregation in Low Sn ZIRLO
    E. Francis, P. Frankel, S. Haigh, and M. Preuss, The University of Manchester, Manchester, UK

    07 Study of Zircaloy Corrosion to Develop Mechanistic Understanding
    V. Allen, M. Bamber, M. Gass, H. Goulding, R. Howells D. Ludlow and P. Platt, Serco Energy, Risley, UK; C. English, J. Hyde, J. Minshull, S. Ortner and H. Thompson, NNL, Didcot, Oxfordshire, UK

    08 Influence of process route and hydrides orientation on fracture toughness for Zirconium-2.5% Niobium for pressure tube material
    K S Reddy, K Kapoor, S V R Rao, B Prahlad, and N Saibaba, Nuclear Fuel Complex, PO ECIL, Hyderabad, India

    09 Strain and Corrosion Behavior of E110 Alloy Claddings under a Long-Term Operation in VVER Reactors
    A. E. Novoselov, S. V. Pavlov, V. S. Polenok, E. A. Zvir, V. A. Zhitelev, G. P. Kobylyansky, I. N. Volkova, Joint Stock Company, State Scientific Center - Research Institute of Atomic Reactors, Dimitrovgrad, Russia

    10 Local Atomic Environment of Ni-Bearing Precipitates in Irradiated Zircaloy-2 Claddings Investigated by Micro-Beam X-Ray Absorption Spectroscopy
    G. Kuri, M. Martin, J. Bertsch, C. N. Borca, Paul Scherrer Institute, Villigen PSI, Switzerland

    11 In-situ electrochemical study of zirconium alloys corrosion in simulated VVER coolant
    J. Macák, P. Sajdl, and V. Renčiuková, Institute of Chemical Technology, Prague, Czech Republic; R. Novotný, Institute for Energy, Petten, The Netherlands; and V. Vrtílková and S. Linhart, UJP PRAHA a.s., Prague, Czech Republic

    12 Multiscale Modelling of Hydrogen Embrittlement in Zirconium Alloys
    J. Majevadia, Imperial College, London, England

    13 Effect of Thermomechanical Processing on the Development of Microstructure and Texture During Fabrication of Zr-1 Wt.% Nb Tube Product
    S. Neogy, K. V. Mani Krishna, D. Srivastava and G. K. Dey, Bhabha Atomic Research Centre, Mumbai, India; B. S. Chandrasekhar, N. Saibaba, Nuclear Fuel Complex, Hyderabad, India; and A. Kumar and I. Samajdar, IIT Bombay, Mumbai, India

    14 Fatigue Crack Initiation Tests on Zr-2.5Nb Material in Simulated CANDU Heat Transport System Conditions
    H.M. Nordin1, C. Helsengreen2, N-W Høgberg2, A.J. Phillion1, M.D. Wright1, and A. Douchant1 1. Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, Canada 2. Institute for Energy Technology, OECD Halden Reactor Project, Halden, Norway

    15 The Effect of Shot Peening on Corrosion and Deuterium Pickup of Zr-2.5Nb Pressure Tube Material
    H. M. Nordin and S. G. Bergin, Atomic Energy of Canada Limited, Chalk River, ON, Canada

    16 Development of Stresses and their Effects in Oxide Layers on Zircaloy-4
    H. Swan, S. Ortner, J. Hyde, and C. English, National Nuclear Laboratory, Didcot, Oxon, UK; and H. Hulme, M. Gass and V. Allen, AMEC, Walton House, Warrington, Cheshire, UK

    17 Micro-mechanical Modelling of ZrO2 Formed during the Oxidation of Zirconium Alloys
    P. Platt, M. Preuss, P. Frankel, University of Manchester, Manchester, UK; and R. Howells, M. Gass, and I. Symington, Serco, Risley, Warrington, UK

    18 Secondary Cyclic Hardening and Dynamic Strain Aging in Annealed Zircaloy-2 at RT
    G. Sudhakar Rao, G. S. Mahobia, K. Chattopadhyay, S. Srinivas, and N. C. V. Singh, Banaras Hindu University, Varanasi, India; J.K. Chakravartty, Bhabha Atomic Research Center, Trombay, Mumbai; N Saibaba, Nuclear Fuel Complex, Hyderabad, India

    19 Direct Observations on Twinning: Split Channel Die Plane Strain Compression of Zircaloy‑4
    J. Singh, I. Samajdar and P. Pant, Indian Institute of Technology Bombay, Powai, Mumbai India; K.V. Mani, D. Srivastava, and G. K. Dey, Bhabha Atomic Research Centre, Mumbai, India; and N. Saibaba, National Fuel Complex, Hyderabad-500 062, India

    20 Characterization of Zircaloy-4 Corrosion Films using Microbeam Synchrotron Radiation
    D. J. Spengler and A. T. Motta, Penn State University, University Park, PA, USA; R. Bajaj, and J. Seidensticker, Bettis Atomic Power Laboratory, West Mifflin, PA, USA; Z. Cai, Argonne National Laboratory, Argonne, IL, USA

    21 Studies on Strain Hardening Behavior of Rolled Zircaloy (Zr4) Sheet Material in Different Directions
    K. Chandrakar, K. S. Lakshmi, A. Lakshminarayana, B. Prahlad, Nuclear Fuel Complex Hyderabad, India