Zirconium in the N

uclear Industry 17th International Symposium

STP15

43C

omstock | B

arbéris

ASTM INTERNATIONAL Helping our world work better

ISBN: 978-0-8031-7529-7 Stock #: STP1543

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ASTM INTERNATIONAL Selected Technical Papers

Zirconium in the Nuclear Industry 17th International SymposiumSTP 1543Editors Robert J. ComstockPierre Barbéris

SELECTED TECHNICAL PAPERSSTP1543

Editors: Robert Comstock, Pierre Barbéris

Zirconium in the Nuclear Industry: 17th International Symposium

ASTM Stock #STP1543

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ISBN: 978-0-8031-7529-7ISSN: 1050-7558

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Printed in Bay Shore, NYJanuary, 2015

Th is Compilation of Selected Technical Papers, STP1543, Zirconium in the Nuclear Industry: 17th International Symposium, contains peer-reviewed papers that were presented at a symposium held February 3–7, 2013 in Hyderabad, India. Th e sym-posium was sponsored by ASTM International Committee B10 on Reactive and Re-fractory Metals and Alloys and Subcommittee B10.02 on Zirconium and Hafnium.

Th e Symposium Co-Chairmen were Pierre Barbéris, Areva/Cezus Research Centre, Ugine, France and Srikumar Banerjee, Atomic Energy Commission, Anush-akti Bhavan, Mumbai, India.

Th e STP Editors are Robert J. Comstock, Westinghouse Electric Company, Pittsburgh, PA, USA and Pierre Barbéris.

Foreword

Overview xi

Kroll Papers

Refl ections on the Development of the “f ” Texture Factors for Zirconium Components and the Establishment of Properties of the Zirconium–Hydrogen System 3

J. J. Kearns

Displacive and Diff usional Transformations of the Beta Phase in Zirconium Alloys 23

S. Banerjee

Schemel Award Paper

Eff ect of Hydrogen on Dimensional Changes of Zirconium and the Infl uence of Alloying Elements: First-Principles and Classical Simulations of Point Defects, Dislocation Loops, and Hydrides 55

M. Christensen, W. Wolf, C. Freeman, E. Wimmer, R. B. Adamson, L. Hallstadius, P. Cantonwine, and E. V. Mader

Basic Metallurgy and Alloying Eff ects

Phase Field Modeling of Microstructure Evolution in Zirconium Base Alloys 95

G. Choudhuri, S. Chakraborty, B. K. Shah, D. Srivastava, and G. K. Dey

Thermodynamics of Zr Alloys: Application to Heterogeneous Materials 118

P. Barberis, C. Vauglin, P. Fremiot, and P. Guerin

Infl uence of Sn on Deformation Mechanisms During Room Temperature Compression of Binary Zr–Sn Alloys 138

K. V. Mani Krishna, D. G. Leo Prakash, D. Srivastava, N. Saibaba, J. Quinta da Fonseca, G. K. Dey, and M. Preuss

Impact of Iron in M5TM 159

D. Kaczorowski, J. P. Mardon, P. Barberis, P. B. Hoff mann, and J. Stevens

Contents

Microstructure and Properties of a Three-Layer Nuclear Fuel Cladding Prototype Containing Erbium as a Neutronic Burnable Poison 184

J. C. Brachet, P. Olier, V. Vandenberghe, S. Doriot, S. Urvoy, D. Hamon, T. Guilbert, A. Mascaro, J. Jourdan, C. Toff olon-Masclet, M. Tupin, B. Bourdiliau, C. Raepsaet, J. M. Joubert, and J. L. Aubin

Characterizing Quenched Microstructures in Relation to Processing 225

P. Barberis, M. T. Tran, F. Montheillet, D. Piot, and A. Gaillac

Fabrication

Identifi cation of Safe Hot-Working Conditions in Cast Zr–2.5Nb 259

J. K. Chakravartty, R. Kapoor, A. Sarkar, V. Kumar, S. K. Jha, N. Saibaba, and S. Banerjee

A Numerical Study of the Eff ect of Extrusion Parameters on the Temperature Distribution in Zr–2.5Nb 282

N. Saibaba, N. Keskar, K. V. Mani Krishna, V. Raizada, K. Vaibhaw, S. K. Jha, D. Srivastava, and G. K. Dey

Study on Eff ect of Processing on Texture Development in Zirconium-2.5 % Niobium Alloy Tubes 302

N. Saibaba, K. Kapoor, S. V. Ramana Rao, K. Itisri, S. K. Jha, C. Phani Babu, G. N. Ganesha, B. Prahlad, and R. K. Mistry

Numerical Modeling of Fuel Rod Resistance Butt Welding 331

A. Gaillac, D. Duthoo, C. Vauglin, D. Carcey-Collet, F. Bay, and K. Mocellin

Application of Coating Technology on Zirconium-Based Alloy to Decrease High-Temperature Oxidation 346

H.-G. Kim, I.-H. Kim, J.-Y. Park, and Y.-H. Koo

Corrosion and Hydrogen Pickup

Oxidation Mechanism in Zircaloy-2—The Eff ect of SPP Size Distribution 373

P. Tejland, H.-O. Andrén, G. Sundell, M. Thuvander, B. Josefsson, L. Hallstadius, M. Ivermark, and M. Dahlbäck

Eff ect of Sn on Corrosion Mechanisms in Advanced Zr-Cladding for Pressurised Water Reactors 404

P. G. Frankel, J. Wei, E. M. Francis, A. Forsey, N. Ni, S. Lozano-Perez, A. Ambard, M. Blat-Yrieix, R. J. Comstock, L. Hallstadius, R. Moat, C. R. M. Grovenor, S. Lyon, R. A. Cottis, and M. Preuss

Understanding of Corrosion Mechanisms of Zirconium Alloys after Irradiation: Eff ect of Ion Irradiation of the Oxide Layers on the Corrosion Rate 438

M. Tupin, J. Hamann, D. Cuisinier, P. Bossis, M. Blat, A. Ambard, A. Miquet, D. Kaczorowski, and F. Jomard

Eff ect of Alloying Elements on Hydrogen Pickup in Zirconium Alloys 479

A. Couet, A. T. Motta, and R. J. Comstock

Toward a Comprehensive Mechanistic Understanding of Hydrogen Uptake in Zirconium Alloys by Combining Atom Probe Analysis With Electronic Structure Calculations 515

M. Lindgren, G. Sundell, I. Panas, L. Hallstadius, M. Thuvander, and H.-O. Andrén

Corrosion and Hydrogen Uptake in Zirconium Claddings Irradiated in Light Water Reactors 540

S. Abolhassani, G. Bart, J. Bertsch, M. Grosse, L. Hallstadius, A. Hermann, G. Kuri, G. Ledergerber, C. Lemaignan, M. Martin, S. Portier, C. Proff , R. Restani, S. Valance, S. Valizadeh, and H. Wiese

In Reactor Performance

Oxidation and Hydrogen Uptake of ZIRLO Structural Components Irradiated to High Burn-Up 577

J. M. García-Infanta, M. Aulló, D. Schrire, F. Culebras, and A. M. Garde

Performance and Property Evaluation of High-Burnup Optimized ZIRLOTM Cladding 607

G. Pan, A. M. Garde, and A. R. Atwood

Corrosion, Dimensional Stability and Microstructure of VVER-1000 E635 Alloy FA Components at Burnups up to 72 MWday/kgU 628

V. N. Shishov, V. A. Markelov, A. V. Nikulina, V. V. Novikov, M. M. Peregud, A. Y. Shevyakov, I. N. Volkova, G. P. Kobylyansky, A. E. Novoselov, and A. V. Obukhov

Corrosion and Hydriding Model for Zircaloy-2 Pressure Tubes of Indian Pressurised Heavy Water Reactors 651

S. K. Sinha and R. K. Sinha

Oxide Surface Peeling of Advanced Zirconium Alloy Cladding after High Burnup Irradiation in Pressurized Water Reactors 673

A. M. Garde, G. Pan, A. J. Mueller, and L. Hallstadius

The Eff ects of Microstructure and Operating Conditions on Irradiation Creep of Zr-2.5Nb Pressure Tubing 693

L. Walters, G. A. Bickel, and M. Griffi ths

Irradiation and Hydrogen Eff ects

Breakthrough in Understanding Radiation Growth of Zirconium 729

S. I. Golubov, A. V. Barashev, R. E. Stoller, and B. N. Singh

Microstructural Evolution of M5TM7 Alloy Irradiated in PWRs up to High Fluences—Comparison With Other Zr-Based Alloys 759

S. Doriot, B. Verhaeghe, J.-L. Béchade, D. Menut, D. Gilbon, J.-P. Mardon, J.-M. Cloué, A. Miquet, and L. Legras

Modeling Irradiation Damage in Zr-2.5Nb and Its Eff ects on Delayed Hydride Cracking Growth Rate 800

G. A. Bickel, M. Griffi ths, H. Chaput, A. Buyers, and C. E. Coleman

Understanding the Drivers of In-Reactor Growth of β-Quenched Zircaloy-2 BWR Channels 830

J. Romero, M. Dahlbäck, L. Hallstadius, M. Ivermark, and G. Ledergerber

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 853

L. Tournadre, F. Onimus, J.-L. Béchade, D. Gilbon, J.-M. Cloué, J.-P. Mardon, and X. Feaugas

High Temperature Transient Behavior

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 897

M. Négyesi, J. Krejčí, S. Linhart, L. Novotný, A. Přibyl, J. Burda, V. Klouček, J. Lorinčík, J. Sopoušek, J. Adámek, J. Siegl, and V. Vrtílková

Experimental Comparison of the Behavior of E110 and E110G Claddings at High Temperature 932

Z. Hózer, E. Perez-Feró, T. Novotny, I. Nagy, M. Horváth, A. Pintér-Csordás, A. Vimi, M. Kunstár, and T. Kemény

Eff ect of Pre-Oxide on Zircaloy-4 High-Temperature Steam Oxidation and Post- Quench Mechanical Properties 952

S. Guilbert, P. Lacote, G. Montigny, C. Duriez, J. Desquines, and C. Grandjean

Deviations From Parabolic Kinetics During Oxidation of Zirconium Alloys 979

M. Steinbrück and M. Grosse

Infl uence of Steam Pressure on the High Temperature Oxidation and Post-Cooling Mechanical Properties of Zircaloy-4 and M5 Cladding (LOCA Conditions) 1002

M. Le Saux, V. Vandenberghe, P. Crébier, J. C. Brachet, D. Gilbon, J. P. Mardon, P. Jacques, and A. Cabrera

Analysis of the Secondary Cladding Hydrogenation During the Quench–LOCA Bundle Tests With Zircaloy-4 Claddings and its Infl uence on the Cladding Embrittlement 1054

M. Grosse, J. Stuckert, C. Roessger, M. Steinbrueck, M. Walter, and A. Kaestner

Degradation and Failure Mechanisms

Eff ect of Hydride Distribution on the Mechanical Properties of Zirconium-Alloy Fuel Cladding and Guide Tubes 1077

S. K. Yagnik, J.-H. Chen, and R.-C. Kuo

Mechanisms of Hydride Reorientation in Zircaloy-4 Studied in Situ 1107

K. Colas, A. Motta, M. R. Daymond, and J. Almer

Hydriding Induced Corrosion Failures in BWR Fuel 1138

D. Lutz, Y.-P. Lin, R. Dunavant, R. Schneider, H. Yeager, A. Kucuk, B. Cheng, and J. Lemons

Author Index 1173

Subject Index 1177

xi

Th is STP contains the papers presented at the 17th International Symposium on Zirco-nium in the Nuclear Industry held in Hyderabad, Andhra Pradesh, India from Febru-ary 3–7, 2013. Th e fi rst symposium was held in Philadelphia in 1968 with subsequent symposia held every two to three years. Th e proceedings of each symposium in the series have been documented with an STP.

During this symposium, the William J. Kroll Zirconium Medal was presented to John Kearns (2011 winner) and Srikumar Banerjee (2012 winner) for their unique and lasting contributions to the technology of zirconium alloys. Both provided his-torical perspectives of their research during the symposium and contributed papers that are included in the STP.

Th e symposium was truly international; with approximately 130 participants from 17 countries attending and representation from North and South America, Eu-rope, and Asia. Th e 17th Symposium included 42 platform presentations along with a session with 31 posters. Th is STP contains 38 peer reviewed papers along with the papers from the two Kroll winners. In addition, the discussion of each platform pres-entation provided an opportunity for further insight and understanding of the paper. As in past symposia, the questions along with written responses by the authors are included at the end of each paper.

Th e symposium is an opportunity to capture a snapshot of the current research areas that are relevant to the nuclear industry. Th is symposium was no exception. While the papers are grouped into seven categories (Basic Metallurgy and Alloying Eff ects, Fabrication, Corrosion and Hydrogen Pickup, In Reactor Performance, Irra-diation and Hydrogen Eff ects, High Temperature Transient Behavior, and Degradation and Failure Mechanisms), there is oft en overlap between the topics as you will see when you browse through the STP or delve into papers more deeply.

An important component of the symposium is the in-reactor performance of zir-conium alloys with several papers presenting recent results from materials irradiated to high burnups and fl uence.

• Alloy E635 was irradiated to 72 MWd/kgU in VVER-1000 reactors with per-formance data presented from both fuel cladding and structural components. Th e behavior of the material was correlated to both temperature and neutron fl uence.

• Th e evolution of the microstructure of M5™ fuel rod cladding irradiated in a PWR up to 7 cycles with fast fl uence up to 17.1 × 1025 n/m2 was described. In a separate paper, results from M5™ with 1000 ppm Fe (designated M5-Fe)

Overview

xii

included both oxide thickness measurements at burnups of about 65 MWd/tU and free growth at fl uences to 20 × 1025 n/m2 (E > 1Mev).

• ZIRLO® structural components were characterized for both corrosion and di-mensional changes following irradiation in PWRs with a maximum fl uence of 13.6 × 1025 n/m2 (E > 1Mev).

• Optimized ZIRLO™ fuel cladding was irradiated beyond the license limit of 62 GWd/MTU to 70 GWd/MTU. Characterization of the cladding included corrosion, dimensional changes, and mechanical properties.

• A detailed study was presented on the infl uence of temperature and micro-structure on the irradiation creep behavior of Zr-2.5Nb pressure tubes.

A dominant theme in this STP is the role of hydrogen on the performance of zir-conium alloy components. Issues discussed in this volume where performance was dominated by hydrogen included the following:

• Failure of BWR fuel rods was attributed to the localization of hydrides fol-lowing accelerated corrosion and subsequent cracking of the hydride lenses. Despite an extensive investigation, the cause of the accelerated corrosion was not defi nitively identifi ed.

• Th e growth of beta-quenched Zircaloy-2 BWR channels was driven late in life by accelerated hydrogen pickup that coincided with the dissolution of second phase particles.

• As reorientation of hydrides plays an important role during dry storage, in-situ measurements were performed to gain new insights into the reorientation of hydrides in Zircaloy-4.

• Delayed hydride cracking (DHC) growth rate of in-service Zr-2.5Nb CANDU pressure tubes was controlled by thermal and irradiation eff ects on the micro-structure (e.g, decomposition and reconstitution of the beta phase controlling hydrogen diff usion to the crack tip).

In addition to papers that highlight the impact of hydrogen, several papers fo-cused on understanding the mechanisms of hydrogen ingress into the metal or understanding the interaction of hydrogen with point defects and dislocation loops in the matrix. Th e latter has potential implications related to hydrogen assisted ir-radiation growth.

Th e US Nuclear Regulatory Commission proposal of a rule to amend the current requirements governing emergency core cooling system for light nuclear power reac-tors has prompted renewed activity in cladding ductility following a high tempera-ture transient. Papers addressed diff erent aspects of the high temperature oxidation behavior of cladding, including secondary hydriding following rod burst, the detri-mental role of nitrogen on oxidation kinetics, and the impact of a pre-oxide or steam pressure on oxidation and subsequent cladding ductility. One paper demonstrated

xiii

the improved oxidation performance of E110G relative to E110. Unfortunately, the reason for the improvement remains an area for continued research.

In response to the Fukushima Daiichi nuclear accident in March, 2011, signifi cant attention has been given to improving the accident tolerance of zirconium alloy fuel cladding. Researchers presented one approach through the application of coating technology to improve the high temperature oxidation resistance of the cladding. While signifi cant development work remains, this is an area that will likely receive continued attention to identify viable options.

Th e zirconium community continues to push the limits of state-of-the-art ana-lytical techniques to characterize the microstructure of both non-irradiated and ir-radiated zirconium alloys. Techniques such as synchrotron radiation (e.g., μ x-ray absorption near edge spectroscopy, diff raction, and stress measurement), electron back-scattered diff raction, atom probe tomography, secondary ion mass spectros-copy, and electron energy loss spectroscopy have become routine analytical tools in multiple laboratories around the world. Cold neutron prompt gamma activation analysis was successfully used in one study to non-destructively measure hydrogen content in corrosion coupons. Complementing the analytical characterization tech-niques are modelling eff orts designed to facilitate mechanistic understanding of per-formance phenomena.

Following the symposium, a committee of technical experts covering a breadth of experience in the zirconium nuclear industry selected the best paper based upon technical excellence, relevance to the nuclear industry, and ‘groundbreaking’ research. Th e winner of the John H. Schemel Best Paper Award was the paper entitled “Eff ect of Hydrogen on Dimensional Changes of Zirconium and the Infl uence of Alloying Ele-ments: First-Principles and Classical Simulations of Point Defects, Dislocation Loops, and Hydrides” by M. Christensen, W. Wolf, C. Freeman, E. Wimmer, R. B. Adamson, L. Hallstadius, P. Cantonwine, and E. V. Mader. Congratulations to the winners.

Robert J. Comstock Pierre Barbéris

Zirconium in the N

uclear Industry 17th International Symposium

STP15

43C

omstock | B

arbéris

ASTM INTERNATIONAL Helping our world work better

ISBN: 978-0-8031-7529-7 Stock #: STP1543

www.astm.org

ASTM INTERNATIONAL Selected Technical Papers

Zirconium in the Nuclear Industry 17th International SymposiumSTP 1543Editors Robert J. ComstockPierre Barbéris