Astm stp 1529 2012

Journal of ASTM International

Selected Technical Papers STP1529

Zirconium in the Nuclear Industry:

16th International Symposium

JAI Guest Editors:

Magnus Limbäck

Pierre Barbéris

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Second Printing, April 2012

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Foreword

This publication, Zirconium in the Nuclear Industry: 16th International

Symposium, contains papers presented at the symposium with the same

name held in Chengdu, Sichuan Province, China, May 9-13, 2010. The

sponsor of the symposium was ASTM International Committee B10 on

Reactive and Refractory Metals and Alloys.

The Symposium Chairman was Magnus Limbäck, Westinghouse Electric

Sweden and Co-Chairman Zhao Wenjin, Nuclear Power Institute of China

(NPIC), Chengdu, Sichuan Province, China. Serving as Guest Editors of

this publication are Magnus Limbäck and Pierre Barbéris, Areva/Cezus

Research Centre, Ugine, France. Arthur Motta, Pennsylvania State

University, acted as Associate Editor for the publication of these papers

in Journal of ASTM International (JAI).

Contents

Overview ............................................................ ix

Kroll Award Papers

Explosion Cladding: An Enabling Technology for Zirconium in the Chemical Process

Industry

J. G. Banker ........................................................ 3

Performance of Zirconium Alloys in Light Water Reactors with a Review

of Nodular Corrosion

D. G. Franklin ....................................................... 17

The Evolution of Microstructure and Deformation Stability in Zr–Nb–(Sn,Fe) Alloys

Under Neutron Irradiation

V. N. Shishov ....................................................... 37

The Development of Zr-2.5Nb Pressure Tubes for CANDU Reactors

B. A. Cheadle ....................................................... 67

Schemel Award Paper

Photoelectrochemical Investigation of Radiation-Enhanced Shadow Corrosion

Phenomenon

Y.-J. Kim, R. Rebak, Y.-P. Lin, D. Lutz, D. Crawford, A. Kucuk, and B. Cheng.......... 91

Basic Metallurgy

Dynamic Recrystallization in Zirconium Alloys

J. K. Chakravartty, R. Kapoor, A. Sarkar, and S. Banerjee ....................... 121

Measurement and Modeling of Second Phase Precipitation Kinetics in Zirconium

Niobium Alloys

M. Ivermark, J. Robson, and M. Preuss .................................... 150

Texture Evolution of Zircaloy-2 During Beta Quenching: Effect of Process Variables

J. Romero, M. Preuss, J. Quinta da Fonseca, R. J. Comstock, M. Dahlbäck,

and L. Hallstadius.................................................... 176

In Situ Studies of Variant Selection During the α-β-α Phase Transformation

in Zr-2.5Nb

P. Mosbrucker, M. R. Daymond, and R. A. Holt ............................... 195

Fabrication and Mechanical Properties

Segregation in Vacuum Arc Remelted Zirconium Alloy Ingots

A. Jardy, F. Leclerc, M. Revil-Baudard, P. Guerin, H. Combeau, and V. Rebeyrolle ...... 219

Damage Build-Up in Zirconium Alloys During Mechanical Processing: Application

to Cold Pilgering

A. Gaillac, C. Lemaignan, and P. Barberis ................................... 244

Multiscale Analysis of Viscoplastic Behavior of Recrystallized Zircaloy-4 at 400°C

M. Priser, M. Rautenberg, J.-M. Cloué, P. Pilvin, X. Feaugas, and D. Poquillon ........ 269

Polycrystalline Modeling of the Effect of Texture and Dislocation Microstructure

on Anisotropic Thermal Creep of Pressurized Zr-2.5Nb Tubes

W. Li, R. A. Holt, and S. Tracy ........................................... 298

Improved Zr-2.5Nb Pressure Tubes for Reduced Diametral Strain in Advanced

CANDU Reactors

G. A. Bickel, M. Griffi ths, A. Douchant, S. Douglas, O. T. Woo, and A. Buyers ......... 327

Microstructural Studies of Heat Treated Zr-2.5Nb Alloy for Pressure Tube Applications

N. Saibaba, S. K. Jha, S. Tonpe, K. Vaibhaw, V. Deshmukh, S. V. Ramana Rao,

K. V. Mani Krishna, S. Neogy, D. Srivastava, G. K. Dey, R. V. Kulkarni, B. B. Rath,

E. Ramadasan, and S. A. Anantharaman ................................... 349

Hydriding – Hydrogen Effect

High Temperature Aqueous Corrosion and Deuterium Uptake of Coupons Prepared

from the Front and Back Ends of Zr-2.5Nb Pressure Tubes

H. M. Nordin, A. J. Elliot, and S. G. Bergin .................................. 373

Hydrogen Absorption Mechanism of Zirconium Alloys Based on Characterization

of Oxide Layer

K. Une, K. Sakamoto, M. Aomi, J. Matsunaga, Y. Etoh, I. Takagi, S. Miyamura,

T. Kobayashi, and K. Ito ............................................... 401

In Situ Scanning Electron Microscope Observation and Finite Element Method

Analysis of Delayed Hydride Cracking Propagation in Zircaloy-2 Fuel Cladding Tubes

T. Kubo, H. Muta, S. Yamanaka, M. Uno, and K. Ogata ......................... 433

Study on the Role of Second Phase Particles in Hydrogen Uptake Behavior

of Zirconium Alloys

M. Y. Yao, J. H. Wang, J. C. Peng, B. X. Zhou, and Q. Li ........................ 466

Hydride Platelet Reorientation in Zircaloy Studied with Synchrotron Radiation

Diffraction

K. B. Colas, A. T. Motta, M. R. Daymond, M. Kerr, and J. D. Almer ................. 496

Statistical Analysis of Hydride Reorientation Properties in Irradiated Zircaloy-2

S. Valance, J. Bertsch, and A. M. Alam ..................................... 523

The Effect of Microstructure on Delayed Hydride Cracking Behavior of Zircaloy-4

Fuel Cladding––An International Atomic Energy Agency Coordinated Research

Programme

C. Coleman, V. Grigoriev, V. Inozemtsev, V. Markelov, M. Roth, V. Makarevicius,

Y. S. Kim, K. L. Ali, J. K. Chakravarrty, R. Mizrahi, and R. Lalgudi ................. 544

Neutron Radiography: A Powerful Tool for Fast, Quantitative and Non-Destructive

Determination of the Hydrogen Concentration and Distribution in Zirconium Alloys

M. Grosse ......................................................... 575

Corrosion – Oxide Layer Characterization

Detailed Analysis of the Microstructure of the Metal/Oxide Interface Region in Zir￾caloy-2 after Autoclave Corrosion Testing

P. Tejland, M. Thuvander, H.-O. Andrén, S. Ciurea, T. Andersson, M. Dahlbäck,

and L. Hallstadius.................................................... 595

Study of the Initial Stage and Anisotropic Growth of Oxide Layers Formed on

Zircaloy-4

B. X. Zhou, J. C. Peng, M. Y. Yao, Q. Li, S. Xia, C. X. Du, and G. Xu ................ 620

Studies Regarding Corrosion Mechanisms in Zirconium Alloys

M. Preuss, P. Frankel, S. Lozano-Perez, D. Hudson, E. Polatidis, N. Ni , J. Wei,

C. English, S. Storer, K. B. Chong, M. Fitzpatrick, P. Wang, J. Smith, C. Grovenor,

G. Smith, J. Sykes, B. Cottis, S. Lyon, L. Hallstadius, B. Comstock, A. Ambard,

and M. Blat-Yrieix.................................................... 649

Understanding Crack Formation at the Metal/Oxide Interface During Corrosion of

Zircaloy-4 Using a Simple Mechanical Model

A. Ly, A. Ambard, M. Blat-Yrieix, L. Legras, P. Frankel, M. Preuss, C. Curfs, G. Parry,

and Y. Bréchet ...................................................... 682

In Pile Behaviour

Optimization of Zry-2 for High Burnups

F. Garzarolli, B. Cox, and P. Rudling ....................................... 711

Effects of Secondary Phase Particle Dissolution on the In-Reactor Performance

of BWR Cladding

S. Valizadeh, G. Ledergerber, S. Abolhassan, D. Jädernäs, M. Dahlbäck, E. V. Mader,

G. Zhou, J. Wright, and L. Hallstadius ..................................... 729

Hydrogen Solubility and Microstructural Changes in Zircaloy-4 Due to Neutron

Irradiation

P. Vizcaíno, A. V. Flores, P. B. Bozzano, A. D. Banchik, R. A. Versaci, and R. O. Ríos .... 754

Advanced Zirconium Alloy for PWR Application

A. M. Garde, R. J. Comstock, G. Pan, R. Baranwal, L. Hallstadius, T. Cook,

and F. Carrera ....................................................... 784

Ultra Low Tin Quaternary Alloys PWR Performance—Impact of Tin Content on

Corrosion Resistance, Irradiation Growth, and Mechanical Properties

V. Chabretou, P. B. Hoffmann, S. Trapp-Pritsching, G. Garner, P. Barberis,

V. Rebeyrolle, and J. J. Vermoyal ......................................... 801

Radiation Damage of E635 Alloy Under High Dose Irradiation in the VVER-1000 and

BOR-60 Reactors

G. P. Kobylyansky, A. E. Novoselov, A. V. Obukhov, Z. E. Ostrovsky, V. N. Shishov,

M. M. Peregud, and V. A. Markelov ....................................... 827

Creep and Deformation

ZIRLO Irradiation Creep Stress Dependence in Compression and Tension

J. P. Foster and R. Baranwal ............................................ 853

Experimental Investigation of Irradiation Creep and Growth of Recrystallized

Zircaloy-4 Guide Tubes Pre-Irradiated in PWR

M. A. McGrath and S. Yagnik ............................................ 875

REFLET Experiment in OSIRIS: Relaxation under Flux as a Method for Determining

Creep Behavior of Zircaloy Assembly Components

S. Carassou, C. Duguay, P. Yvon, F. Rozenblum, J. M. Cloué, V. Chabretou,

C. Bernaudat, B. Levasseur, A. Maurice, P. Bouffi oux, and K. Audic ................ 899

Impact of the Irradiation Damage Recovery During Transportation on the Subsequent

Room Temperature Tensile Behavior of Irradiated Zirconium Alloys

B. Bourdiliau, F. Onimus, C. Cappelaere, V. Pivetaud, P. Bouffi oux, V. Chabretou,

and A. Miquet ...................................................... 929

Shadow Corrosion-Induced Bow of Zircaloy-2 Channels

S. T. Mahmood, P. E. Cantonwine, Y.-P. Lin, D. C. Crawford, E. V. Mader,

and K. Edsinger ..................................................... 954

Failure Mechanisms and Transients

Characterization of Oxygen Distribution in LOCA Situations

C. Duriez, S. Guilbert, A. Stern, C. Grandjean, L. Beˇlovský, and J. Desquines ........ 993

Effect of Hydrides on Mechanical Properties and Failure Morphology of BWR Fuel

Cladding at Very High Strain Rate

M. Nakatsuka and S. Yagnik ............................................ 1021

Simulation of Outside-in Cracking in Boiling Water Reactor Fuel Cladding Tubes

under Power Ramp

K. Sakamoto, M. Nakatsuka, and T. Higuchi ................................. 1054

RIA Failure of High Burnup Fuel Rod Irradiated the Leibstadt Reactor: Out-of-Pile

Mechanical Simulation and Comparison with Pulse Reactor Tests

V. Grigoriev, R. Jakobsson, D. Schrire, G. Ledergerber, T. Sugiyama, F. Nagase,

T. Fuketa, L. Hallstadius, and S. Valizadeh .................................. 1073

Author Index ......................................................... 1093

Subject Index .......................................................... 1097

ix

Overview

This STP contains the papers presented at the 16th International Sympo￾sium on Zirconium in the Nuclear Industry held in in Chengdu, Sichuan

Province, China, May 9-13, 2010. The fi rst symposium was held in Phila￾delphia in 1968, and symposia have been held ever since in two to three

year intervals. The proceedings of each symposium in the series have been

documented with an STP.

This symposium series remains, after forty years, one of the top pres￾entation and information source for the research in the area of zirconium

alloy performance in a nuclear reactor environment. 42 papers and 32

posters were selected for presentation at the 16th Symposium from 130

abstracts submitted. The forty-two papers published in these proceedings

were peer reviewed and edited, and are also published in the ASTM online

journal, JAI. In addition, the most signifi cant parts of the discussions that

followed the oral presentation of each paper at the symposium are included

in these proceedings.

Four experts in zirconium area received their Kroll Awards at the 16th

Symposium: J. Banker, D. Franklin, V. Shishov and B. Cheadle. Notewor￾thy is the fact that the fi rst one deals with zirconium outside the nuclear

industry. These papers as well as the 26 previous Kroll award papers are

now gathered in “The Kroll Medal Papers 1975-2010” published by ASTM,

and covering all aspects of zirconium technology.

137 attendants from 19 countries attended the 16th Symposium. North

and South America, Europe, and Asia were represented.

The papers were presented during seven sessions, covering the whole

spectrum of zirconium metallurgy, from basic metallurgy to accidental con￾ditions and transport, through fabrication, creep and growth, and corrosion

and hydriding. Looking back from the beginning of this symposium series, it

appears that these topics remained quite constant over the time.

Besides the historical alloys (Zircaloy-2, Zircaloy-4, Zr2.5Nb and Zr-lNb),

several studies were devoted to advanced alloys and alloys under develop￾ment: ZIRLO™, M5™, X5A, VB, N18, N36, NZ2, E635, ZrNbSnFe with low

tin content, Ziron. Some optimisation of Zircaloy-2 was proposed.

Modelling appears more and more intricate with the experiments, from

precipitation to VAR melting, from the effect of texture and dislocations on

creep to the oxygen distribution during LOCA situations or the outside-in

cracking in BWR fuel cladding.

As noted in the past few symposia, advanced techniques are more system￾atically utilised: high temperatures studies of phase transformations using

x

synchrotron radiation or neutrons diffraction were presented; EBSD allows

measuring local texture, studies on corrosion and hydriding benefi t from the

precise positioning of thin foils with FIB.

Last, it is worth noting that several studies were presented by enthusias￾tic young searchers or PhD students, which demonstrates the dynamism of

the research in the zirconium metallurgy area.

In the fi eld of basic metallurgy, progresses were reported in the following

fi elds:

– plotting the dynamic recrystallisation domain thanks to processing

maps, which were shown to depend on the alloy composition.

– experimental quantifi cation of the precipitation of beta Nb second phase

particles (SPP) in Zr-Nb alloys, which was investigated by synchrotron

radiation and then modelled.

– texture change and variant selection through the α-β-α phase transfor￾mation, wich generated some controversial results on the infl uence of

an externally applied stress.

The processing of zirconium was illustrated fi rst by the VAR melting, with

modelling of the alloying element segregation allowing effi cient control on

the ingot chemistry, then by the investigation of the damage mechanism

during pilgering, which is alloy dependent through the number and size of

the second phase particles. Transmission electron microscopy was used with

the aim of investigating the dislocation microstructure of recrystallised Zir￾caloy-4 after 400°C creep test to develop a model of the anisotropic visco￾plastic behaviour.

Three papers were devoted to Zr-2.5Nb tubes. The fi rst one presented a

self-consistent modelling of their anisotropic behaviour, which takes into ac￾count not only the texture but also the dislocation microstructure result￾ing from the last cold drawing pass. A second paper showed that the tubes

for the future ADC will procure reduced creep variability, owing to a better

understanding of its dependence on the microstructure and to an improved

extrusion process. A detailed study of the microstructure evolution during

annealing and of its relation with texture and mechanical properties of the

tubes was also conductive to improvement in the tube fabrication sequence.

The investigation of hydrogen effects and hydriding constituted an im￾portant topic of this symposium with eight papers dealing with this topic.

Besides a paper showing the capability of neutron radiography to investi￾gate this area, an investigation of Zr-2.5Nb hydriding during autoclave tests

showed that the hydrogen pick-up was linked to the sample position in the

tube (back end/front end) and correlated to various microstructural param￾eters. An investigation of autoclave corroded samples from different alloys

xi

by TEM, RAMAN and SIMS led to the hypothesis according to which the

rate controlling process for hydrogen absorption was the diffusion of hydro￾gen ions in the oxide barrier layer, the best alloys having a higher protec￾tive layer due to compressive stresses and Fe dissolution from the second

phase particles. Another paper confi rms that the hydrogen pick up during

corrosion is closely related to the size, area fraction, and compositions of the

second phase particles.

The hydride dissolution, re-orientation and stress could be followed in

situ by synchrotron radiation diffraction during temperature and stress cy￾cles. The hydride re-orientation was the subject of a second paper coupling

mechanical tests on irradiated cladding and fi nite element modelling, show￾ing the importance of the hoop stress, and evidencing some other param￾eters. Two papers dealt with delayed hydride cracking (DHC). The fi rst one

which was derived from an IAEA coordinated research program investigated

the effect of the microstructure on DHC, showing its main role is to control

the material strength, while the second one presented in situ observation in

an SEM of the crack propagation during a DHC experiment and a FEM of

the accepted mechanism.

Two papers on the corrosion mechanism and oxide layer characteriza￾tion showed on the one hand that the delayed oxidation of the SPPs in the

oxide layer lead to the formation of small cracks, and on the other hand the

infl uence of the grain substrate orientation on the oxide layer thickness and

epitaxy during steam corrosion tests.

The in-pile behaviour was the subject of numerous studies. In the aim of

mitigating the accelerated HPUF tendency at high burn up, it was proposed

to decrease the nickel content and increase the iron and chromium contents

in Zircaloy-2 within the ASTM specifi cation. Zircaloy-2 was also the sub￾ject of a detailed study of the SPPs’ evolution, showing that Zr(Fe,Cr)2 SPP

amorphize while Zr2(Fe,Ni) SPP remain crystalline during irradiation. The

shadow corrosion due to galvanic coupling between a zirconium alloy and

a more noble metal was investigated by photo-electrochemistry; the alloy

infl uence was evidenced, and it was postulated that a coating on the fuel as￾sembly spacer may mitigate the shadow corrosion.

Annealing and DSC experiments on irradiated Zircaloy-4 enabled the

study of the infl uence of the neutron damage on the hydrogen solubility,

while the morphology of hydrides and SPP was investigated by TEM.

The in pile behaviour of advanced alloys were presented: X5A (for PWR)

with two different fi nal heat treatments showed improved properties and

that the irradiation creep of ZIRLO™ is linear with the (deviatoric hoop)

stress. The same phenomenon is observed in tension and compression. The

increase in tin content up to 0.3 % in Zr1NbSnFe does not signifi cantly mod￾ify the corrosion resistance nor the hydrogen pick-up compared to Zr1Nb

xii

alloy, while ensuring a higher creep resistance and an improved dimensional

stability.

Radiation damage of E635 in VVER (SPP evolution, 〈c〉-dislocation loop

formation) was investigated and found in agreement with the model samples

in the BOR-60 reactor.

The creep and deformation during or after irradiation, already evoked in

the previous section, was illustrated by four more papers. No effect of com￾mercial reactor irradiation temperature or hydrogen content was found on

the Zircaloy-4 guide tube creep. Zircaloy-4 irradiation creep law was deduced

from relaxation experiments on bent beam specimens in OSIRIS. Creep de￾formation of Zr-1%Nb and Zircaloy-4 during transportation was associated

to a signifi cant recovery of irradiation damage, preventing the dislocation

channelling. Shadow corrosion-induced bow of Zircaloy-2 channels was as￾sociated to differential hydrogen concentration on channels sides adjacent to

and away from the control blades.

Four papers dealt with accidental conditions. In LOCA conditions, a diffu￾sion model was developed to compute the oxygen distribution, closely related

to the mechanical properties, and compared with experiments. An investiga￾tion on the effect of hydrides on mechanical properties during RIA concluded

that high strain rates did not seem to impact the stress strain behaviour

when the hydrogen content is higher than 400ppm, threshold over which the

failure elongation at room temperature decreases drastically. When radially

oriented hydrides are present in BWR cladding, an outside-in cracking can

occur, which was investigated, in isothermal conditions or with a radial ther￾mal gradient: the outside-in cracking during the power ramp seems strongly

dependent on the distribution of dissolved hydrogen as a result of thermal

diffusion. Finally, a high burn-up BWR fuel rod, subjected to RIA tests in a

research reactor, resulted in cladding failure at room temperature, but not

at elevated temperature. A mechanical test was developed to reproduce the

ramp test and was shown to be predictive.

The John Schemel Award is awarded following each symposium for the

best paper presented at the symposium. The selection is based upon the

technical content of the paper, the usefulness of the work reported to the

worldwide reactor components community, and the technical diffi culty in do￾ing the work. This year, a committee of technical experts in several aspects

of the zirconium industry selected the paper entitled “Photoelectrochemical

Investigation of Radiation Enhanced Shadow Corrosion Phenomenon” by

Y.-J. Kim, R. Rebak, Y-P. Lin, D. Lutz, D. Crawford, A. Kucuk and B. Cheng to

receive the John Schemel Award.

Pierre Barbéris

Areva/Cezus Research Centre

Guest Editor

KROLL AWARD PAPERS