STP 1505

Bruce Kammenzind Magnus Limbäck

Editors

www.astm.org

ISBN: 978-0-8031-4514-6Stock #: STP1505

Zirconium in the

Nuclear Industry15th International Symposium

STP 1505Z

irconium

in the N

uclear Industry

15th International S

ymp

osium

Bruce Kammenzind Magnus Limbäck

STP 1505

Zirconium in the Nuclear Industry:15th International Symposium

Bruce Kammenzind and Magnus Limback, editors

ASTM Stock Number: STP1505

ASTM International100 Barr Harbor DrivePO Box C700West Conshohocken, PA 19428-2959

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Library of Congress Cataloging-in-Publication DataISBN:978-0-8031-4514-6ISSN:1050-7558

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Printed in Baltimore, MDJune, 2009

Foreword

This publication, Zirconium in the Nuclear Industry: 15th International Symposium,contains papers presented at the symposium with the same name held in Sunriver,Oregon, USA, June 24-28, 2007. The sponsor of the symposium was ASTM InternationalCommittee B10 on Reactive and Refractory Metals and Alloys.

The symposium chairman was Bruce Kammenzind, Bettis Laboratory, the symposiumco-chairman was Jack Tosdale, ATI Wah Chang, and the editorial chairman was MagnusLimbäck, Westinghouse Electric Sweden. Serving as editors of this publication wereBruce Kammenzind and Magnus Limbäck while Arthur Motta, from The PennsylvaniaState University, acted as Associate Editor for the concurrent publication of these papersin Journal of ASTM International.

Contents

Overview ix

KROLL AWARD PAPERS

In-Reactor Deformation of Zirconium Alloy Components—R. A. HOLT 3

Microstructure Evolution in Zr Alloys during Irradiation: Dose, Dose Rate,and Impurity Dependence—M. GRIFFITHS 19

SCHEMEL AWARD PAPER

Characterization of Zirconium Hydrides and Phase Field Approach to aMesoscopic-Scale Modeling of Their Precipitation—Z. ZHAO, M. BLAT-YRIEIX,J.-P. MORNIROLI, A. LEGRIS, L. THUINET, Y. KIHN, A. AMBARD, AND L. LEGRAS 29

LOCA & TRANSIENTS

Influence of Structure Changes in E110 Alloy Claddings on Ductility Loss under LOCAConditions—S. A. NIKULIN, A. B. ROZHNOV, V. A. BELOV, N. V. LYASCHENKO,A. V. NIKULINA, AND A. G. MAL’GIN 53

Investigations of the Microstructure and Mechanical Properties of Prior-� Structure asa Function of the Oxygen Content in Two Zirconium Alloys—A. STERN,J.-C. BRACHET, V. MAILLOT, D. HAMON, F. BARCELO, S. POISSONNET, A. PINEAU,J.-P. MARDON, AND A. LESBROS 71

Hydrogen Content, Preoxidation, and Cooling Scenario Effects on Post-QuenchMicrostructure and Mechanical Properties of Zircaloy-4 and M5® Alloysin LOCA Conditions—J.-C. BRACHET, V. VANDENBERGHE-MAILLOT, L. PORTIER,D. GILBON, A. LESBROS, N. WAECKEL, AND J.-P. MARDON 91

Experimental and Analytical Investigation of the Mechanical Behavior of High-BurnupZircaloy-4 Fuel Cladding—R. S. DAUM, S. MAJUMDAR, AND M. C. BILLONE 119

Effect of Local Hydride Accumulations on Zircaloy Cladding Mechanical Properties—A. HERMANN, S. K. YAGNIK, AND D. GAVILLET 141

Fracture Toughness of Hydrided Zircaloy-4 Sheet Under Through-Thickness CrackGrowth Conditions—P. A. RAYNAUD, D. A. KOSS, A. T. MOTTA,AND K. S. CHAN 163

MECHANICAL PROPERTIES & FAILURE MECHANISMS

Characterization of Local Strain Distribution in Zircaloy-4 and M5® Alloys—K. ELBACHIRI, P. DOUMALIN, J. CRÉPIN, M. BORNERT, P. BARBERIS, V. REBEYROLLE,AND T. BRETHEAU 181

v

In-Pile Criteria for the Initiation of Massive Hydriding of Zr in Steam-HydrogenEnvironment—I. A. EVDOKIMOV AND V. V. LIKHANSKII 193

Round-Robin Testing of Fracture Toughness Characteristics of Thin-Walled Tubing—S. K. YAGNIK, N. RAMASUBRAMANIAN, V. GRIGORIEV, C. SAINTE-CATHERINE,J. BERTSCH, R. ADAMSON, R.-C. KUO, S. T. MAHMOOD, T. FUKUDA, P. EFSING, AND

B. C. OBERLÄNDER 205

Role of Twinning and Slip in Deformation of a Zr-2.5Nb Tube—Y. S. KIM 227

Measurement of Rates of Delayed Hydride Cracking (DHC) in Zr-2.5Nb Alloys—AnIAEA Coordinated Research Project—C. E. COLEMAN AND V. V. INOZEMTSEV 244

CORROSION

A Study of the Structure and Chemistry in Zircaloy-2 and the Resulting Oxide AfterHigh Temperature Corrosion—B. HUTCHINSON, B. LEHTINEN, M. LIMBÄCK, AND

M. DAHLBÄCK 269

Effects of Pt Surface Coverage on Oxidation of Zr and Other Materials—C. ANGHEL,G. HULTQUIST, M. LIMBÄCK, AND P. SZAKALOS 285

Studies of Corrosion of Cladding Materials in Simulated BWR Environment usingImpedance Measurements—S. FORSBERG, E. AHLBERG, AND M. LIMBÄCK 303

In Situ EIS Measurements of Irradiated Zircaloy-4 Post-Transition Corrosion KineticBehavior—D. M. RISHEL, K. L. EKLUND, AND B. F. KAMMENZIND 326

Effect of Water Chemistry and Composition on Microstructural Evolution of Oxide onZr-Alloys—B. X. ZHOU, Q. LI, M. Y. YAO, W. Q. LIU, AND Y. L. CHU 360

The Effect of Hydrogen on the Transition Behavior of the Corrosion Rate of ZirconiumAlloys—M. HARADA AND R. WAKAMATSU 384

PWR CORROSION

A New Model to Predict the Oxidation Kinetics of Zirconium Alloys in PressurizedWater Reactor—V. BOUINEAU, A. AMBARD, G. BÉNIER, D. PÊCHEUR,J. GODLEWSKI, L. FAYETTE, AND T. DUVERNEIX 405

In PWR Comprehensive Study of High Burn-up Corrosion and Growth Behaviorof M5® and Recrystallized Low-Tin Zircaloy-4—P. BOSSIS, B. VERHAEGHE,S. DORIOT, D. GILBON, V. CHABRETOU, A. DALMAIS, J.-P. MARDON, M. BLAT, AND

A. MIQUET 430

ZirloTM Cladding Improvement—J. P. FOSTER, H. KEN YUEH, AND R. J. COMSTOCK 457

Corrosion and Oxide Properties of HANA Alloys—J.-Y. PARK, B.-K. CHOI, S. J. YOO,AND Y. H. JEONG 471

vi CONTENTS

Microstructural Characterization of Oxides Formed on Model Zr Alloys UsingSynchrotron Radiation—A. T. MOTTA, M. J. GOMES DA SILVA,A. YILMAZBAYHAN, R. J. COMSTOCK, Z. CAI, AND B. LAI 486

Chemistry of Waterside Oxide Layers on Pressure Tubes—T. DO, M. SAIDY,AND W. H. HOCKING 507

DEFORMATION MECHANISM

Manufacturing Variability, Microstructure, and Deformation of Zr-2.5Nb PressureTubes—G. A. BICKEL AND M. GRIFFITHS 529

Effect of Irradiation Damage on the Deformation Properties of Zr-2.5Nb PressureTubes—M. GRIFFITHS, N. WANG, A. BUYERS, AND S. A. DONOHUE 541

Determination and Interpretation of Texture Evolution during Deformation of aZirconium Alloy—V. M. ALLEN, J. QUINTA DA FONSECA, M. PREUSS,J. D. ROBSON, M. DAYMOND, AND R. J. COMSTOCK 550

Irradiation-Induced Growth and Microstructure of Recrystallized, Cold Workedand Quenched Zircaloy-2, NSF, and E635 Alloys—G. P. KOBYLYANSKY, A.E. NOVOSELOV, Z. E. OSTROVSKY, A. V. OBUKHOV, V. YU. SHISHIN, V. N. SHISHOV,A. V. NIKULINA, M. M. PEREGUD, S. T. MAHMOOD, D. W. WHITE, Y-P. LIN,AND M. A. DUBECKY 564

Deformation Anisotropy of Annealed Zircaloy-2 as a Function of Fast NeutronFluence—X. WIE, J. R. THEAKER, AND M. GRIFFITHS 583

Toward a Better Understanding of Dimensional Changes in Zircaloy-4: What is theImpact Induced by Hydrides and Oxide Layer?—M. BLAT-YRIEIX,A. AMBARD, F. FOCT, A. MIQUET, S. BEGUIN, AND N. CAYET 594

CASTA DIVA®: Experiments and Modeling of Oxide-Induced Deformation in NuclearComponents—P. BARBERIS, V. REBEYROLLE, J. J. VERMOYAL, V. CHABRETOU,J. P. VASSAULT 612

SPENT FUEL

Cladding Tube Deformation Test for Stress Reorientation of Hydrides—A. M. ALAM

AND C. HELLWIG 635

Evaluation of Hydride Reorientation Behavior and Mechanical Properties forHigh-Burnup Fuel-Cladding Tubes in Interim Dry Storage—M. AOMI, T. BABA,T. MIYASHITA, K. KAMIMURA, T. YASUDA, Y. SHINOHARA, AND T. TAKEDA 651

Experimental and Modeling Approach of Irradiation Defects Recovery in ZirconiumAlloys: Impact of an Applied Stress—J. RIBIS, F. ONIMUS, J.-L. BÉCHADE,S. DORIOT, C. CAPPELAERE, C. LEMAIGNAN, A. BARBU, AND O. RABOUILLE 674

viiCONTENTS

BASIC METALLURGY

Mechanical Properties of Zr-2.5Nb Pressure Tubes Made from Electrolytic Powder—C. COLEMAN, M. GRIFFITHS, V. GRIGORIEV, V. KISELIOV, B. RODCHENKOV,AND V. MARKELOV 699

Structure-Phase State, Corrosion and Irradiation Properties of Zr-Nb-Fe-Sn SystemAlloys—V. N. SHISHOV, M. M. PEREGUD, A. V. NIKULINA, V. F. KON’KOV,V. V. NOVIKOV, V. A. MARKELOV, T. N. KHOKHUNOVA, G. P. KOBYLYANSKY,A. E. NOVOSELOV, Z. E. OSTROVSKY, AND A. V. OBUKHOV 724

Investigation of Structural and Chemical Uniformity of Zr2.5% Nb and E635 Alloyby Radioactive Indicators—V. ARZHAKOVA, A. SHIKOV, A. KABANOV, V. BELOV,M. SHTUTSA, A. ZIGANSHIN, V. KURAPOV, AND L. KURAPOVA 744

Contribution of Thermodynamic Calculations to Metallurgical Studiesof Multi-Component Zirconium Based Alloys—C. TOFFOLON-MASCLET,J. C. BRACHET, C. SERVANT, J. M. JOUBERT, P. BARBERIS, N. DUPIN, AND P. ZELLER 754

Intergranular and Interphase Constraints in Zirconium Alloys—R. A. HOLT,M. R. DAYMOND, F. XU, AND S. CAI 776

Tearing Crack Growth and Fracture Micro-Mechanisms Under Micro Segregation inZr-2.5%Nb Pressure Tube Material—K. KAPOOR, N. SAIBABA, B. P. KASHYAP,AND A. V. RAMANA RAO 796

viii CONTENTS

OverviewThis STP contains the papers presented at the 15th International Symposium on Zirconium in theNuclear Industry held in Sunriver, Oregon, USA, June 24 through 28, 2007. The first symposiumin the series was held in 1968 and the symposiums have been held ever since in two to three yearintervals. The proceedings of each symposium have been documented with an STP.

This symposium series provides one of the most important arenas for presenting results fromresearch performed all over the world on all aspects of zirconium alloy properties and perfor-mance relevant for the nuclear industry. Forty-two papers and twenty-four posters were selectedfor presentation at the 15th Symposium from about 100 abstracts submitted. The 42 paperspublished in these proceedings were peer reviewed and edited, and are also published in theASTM On Line Journal �JAI�. In addition, the most significant parts of the discussions thatfollowed the oral presentation of each paper at the symposium are included in these proceedings.Finally, M. Griffiths and R. Holt, made presentations on historical aspects of research in zirco-nium alloys when they received their Kroll Awards at the 15th Symposium. In all, 168 delegatesfrom 21 countries attended the 15th Symposium with presentations from North America, Europeand Asia, making the conference truly international in scope and content.

Historically, zirconium alloys were chosen as structural materials for water reactors because ofthe combination of low thermal neutron capture cross section, relatively high mechanicalstrength and high corrosion resistance in water and steam at elevated temperatures. Pure zirco-nium can not be used in water reactors due to its low strength and low corrosion resistance inwater environments. Up to the mid-to-late 1980’s basically only four alloys were used in com-mercial reactors, Zircaloy-2, Zircaloy-4 and Zr-1Nb for fuel assembly components, and Zr-2.5Nbfor pressure tubes. At this point in time the increased cladding corrosion observed in hightemperature Pressurized Water Reactors �PWRs� encouraged the fuel vendors to develop morecorrosion resistant materials which were verified in co-operation with utilities. The developmentof Zr alloys has continued and this STP includes several papers on alternative alloys, i.e., otherthan the four alloys mentioned above, that are either commercially available today or are underverification.

The key characteristics of Zr metallurgy come from its strongly anisotropic hexagonal crystalstructure �which during thermo-mechanical processing leads to the development of a texturedmaterial�, from its high reactivity with oxygen, and from the different types of chemical inter-actions with the alloying elements, including both complete solubility and intermetallic com-pound formation. Moreover, the in-reactor performance of zirconium alloys is strongly depen-dent on irradiation induced changes, or irradiation damage. Most of the structural damageinduced during neutron irradiation in nuclear reactors is due to elastic interactions with fastneutrons, which affects the Zr matrix, the intermetallics as well as the oxide �zirconia� layer onthe surface of the component. The progressive changes of the material during in-reactor opera-tion further complicate and prolong the verification of a new alloy. The verification process alsoneeds to cover the dimensional changes during irradiation, e.g. irradiation induced growth andcreep, and their effect on the pellet-cladding mechanical interaction �PCMI� behavior as well asthe lift-off properties of the fuel rod.

Apart from the in-reactor performance during normal operation one also has to consider thebehavior of the cladding tubes during transients and loss-of-coolant accidents �LOCA� as well asthroughout long-term post-irradiation storage.

ix

This symposium series has a tradition of covering all aspects of Zr-based material performancerelevant for nuclear applications and the papers and posters presented at the 15th Symposium,consequently, covered all aspects of the fuel cycle, with the oral presentations divided into sevensessions covering fields from Basic Metallurgy to Spent Fuel via Corrosion, Mechanical Prop-erties, Deformations Mechanisms, Failure Mechanisms, LOCA and Transients.

A number of interesting topics were noticeable during the 15th Symposium. There is a continuedstrive to enhance the usage of existing experimental techniques as well as to search for applica-tions of new techniques that may further increase the detailed understanding of the properties ofZr-based alloys. We also see several examples of studies where experimental techniques arecombined with analytical tools and calculations, which promote enhanced development in bothareas. Throughout the years many interesting studies have been made out-of-reactor, but now wesee a stronger trend towards efforts that enable fundamental in-reactor measurements, which areessential for further development of experimental techniques as well as for the possibility toenhance the development of improved alloys. Last but not least we see an increasing number ofalternative alloys being developed and verified. These trends and possibilities in combinationwith the talented set of researchers, including the older and long established experts to theenthusiastic and energetic younger scientists assure both continued successful development inour field and numerous exciting future symposiums on Zirconium in the Nuclear Industry.

The John Schemel Award is awarded following each symposium for the best paper presented atthe symposium. The selection is based upon the technical content of the paper, the usefulness ofthe work reported to the worldwide reactor components community, and the technical difficultyin doing the work. This year a committee of technical experts in several aspects of the zirconiumindustry selected the paper entitled ‘‘Characterization of Zirconium Hydrides and Phase FieldApproach to a Mesoscopic-Scale Modeling of their Precipitation� by Z. Zhao, M. Blat-Yrieix,J.-P. Morniroli, A. Legris, L. Thuinet, Y. Kihn, A. Ambard, and L. Legras to receive the JohnSchemel Award.

Bruce KammenzindBettis Laboratory

West Mifflin, PA, USASymposium Chairman and STP Editor

Magnus LimbäckWestinghouse Electric Sweden

Västerås, SwedenEditorial Chairman and STP Editor

x

STP 1505

Bruce Kammenzind Magnus Limbäck

Editors

www.astm.org

ISBN: 978-0-8031-4514-6Stock #: STP1505

Zirconium in the

Nuclear Industry15th International Symposium

STP 1505

Zirco

nium in the

Nuclear Ind

ustry15

th International Sym

posium

Bruce Kammenzind Magnus Limbäck