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 ASTM International 100 Barr Harbor Drive PO Box C700 West Conshohocken, PA 19428-2959 Printed in the U.S.A ASTM Stock #: STP1529 Library of Congress Cataloging-in-Publication Data ISBN: 978-0-8031-7515-0 ISSN: 1050-7558 Copyright © 2011 ASTM INTERNATIONAL, West Conshohocken, PA All rights reserved This material may not be reproduced or copied, in whole or in part, in any printed, mechanical, electronic, film, or other distribution and storage media, without the written consent of the publisher Journal of ASTM International (JAI) Scope The JAI is a multi-disciplinary forum to serve the international scientific and engineering community through the timely publication of the results of original research and critical review articles in the physical and life sciences and engineering technologies These peer-reviewed 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Publications The quality of the papers in this publication reflects not only the obvious efforts of the authors and the technical editor(s), but also the work of the peer reviewers In keeping with long-standing publication practices, ASTM International maintains the anonymity of the peer reviewers The ASTM International Committee on Publications acknowledges with appreciation their dedication and contribution of time and effort on behalf of ASTM International Citation of Papers When citing papers from this publication, the appropriate citation includes the paper authors, “paper title”, J ASTM Intl., volume and number, Paper doi, ASTM International, West Conshohocken, PA, Paper, year listed in the footnote of the paper A citation is provided as a footnote on page one of each paper Second Printing, April 2012 Baltimore, MD 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 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 91 121 Basic Metallurgy Dynamic Recrystallization in Zirconium Alloys J K Chakravartty, R Kapoor, A Sarkar, and S Banerjee 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 Damage Build-Up in Zirconium Alloys During Mechanical Processing: Application to Cold Pilgering A Gaillac, C Lemaignan, and P Barberis 219 244 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 Griffiths, 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 Zircaloy-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 Bouffioux, 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 Bouffioux, 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 Overview This STP contains the papers presented at the 16th International Symposium on Zirconium in the Nuclear Industry held in in Chengdu, Sichuan Province, China, May 9-13, 2010 The first symposium was held in Philadelphia 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 presentation 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 significant 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 Noteworthy is the fact that the first 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 conditions 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 development: 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 systematically utilised: high temperatures studies of phase transformations using ix J_ID: DOI: Date: 18-October-11 Stage: Page: 1086 Total Pages: 17 1086 JAI  STP 1529 ON ZIRCONIUM IN THE NUCLEAR INDUSTRY: 16TH SYMPOSIUM FIG 10—“Failed”-“not failed” map based on the EDC test data plotted in the coordinates “cladding temperature”-“hydrogen in cladding” and the data from the NSRR tests LS-1 and LS-2 located in the map just before the ramp, the outer cladding surface temperature was about 280 C and that temperatures above 600 C could be expected at the inner cladding surface at the end of the in-reactor ramp These temperature intervals can be indicated on the failure map based on the EDC test data (see Fig 7) at a hydrogen concentration of about 300 ppm, which is the concentration defined for the AEB 072-E4-J cladding (Fig 10) According to Fig 10, different temperatures of the cladding wall in the LS-1 and LS-2 tests can explain different behaviours of the cladding in the ramp tests The entire cladding thickness in the LS-2 test was in the “no failure” area, where a ductile behaviour of cladding is expected In the LS-1 test, about half of the cladding thickness (outer part) was at temperatures where the cladding should experience a brittle fracture In general, this is consistent with the fracture surface profile and the fractographical features observed for cladding failure in the LS-1 test (Fig 5) Fractographic examination of the EDC specimens clearly showed the change of fracture surface appearance with an increase of test temperature from 60 to 160 C (Fig 11) Numerous dimples are observed at 160 C, while they could not be found at 60 C Specimens AEB072-E4-H1 and AEB072-E4-H5 were examined by means of metallography, and both showed a slant fracture through the entire cladding wall thickness (Fig 12) However, having the same macrotype of the slant fracture, the relief of the fracture surface appears to be smoother after fracture at 100 C (AEB072-E4-H1) and more irregular after fracture at 190 C (AEB072-E4-H5) Thus, we can see indications of more ductile behaviour (dimples and irregular macrosurface of fracture) at the temperatures close to the BDT ID: kumarva Time: 15:23 I Path: Q:/3b2/STP#/Vol01529/AI-STP#018009 J_ID: DOI: Date: 18-October-11 Stage: Page: 1087 Total Pages: 17 GRIGORIEV ET AL., doi:10.1520/JAI102988 1087 FIG 11—Different types of fractographic features observed at fracture surfaces of the EDC specimens AEB072-E4-H0 and AEB072-E4-H9 tested at different temperatures at the lower shelf of ductility (Table 2) FIG 12—Cross-section metallography of the EDC specimens AEB072-E4-H1 and AEB072-E4-H5 tested at different temperatures (Table 2) ID: kumarva Time: 15:23 I Path: Q:/3b2/STP#/Vol01529/AI-STP#018009 J_ID: DOI: Date: 18-October-11 Stage: Page: 1088 Total Pages: 17 1088 JAI  STP 1529 ON ZIRCONIUM IN THE NUCLEAR INDUSTRY: 16TH SYMPOSIUM Summary A segment of a high burnup fuel rod (63 MWd/kg U rod average) irradiated at the Leibstadt BWR in Switzerland (KKL) has been subjected to simulated RIA tests in the Japanese Atomic Energy Agency, NSRR In order to evaluate the temperature influence on transient fuel behaviour, the pulse irradiation tests were performed at ambient (test LS-1) and operating temperatures (test LS-2), representing cool zero power and hot zero power Two ramp tests carried out under almost identical irradiation conditions resulted in specimen failure in the LS-1 test, while no failure was observed in the LS-2 test An adjacent segment of the same fuel rod has been subjected to high strain rate mechanical testing by means of the EDC test performed at Studsvik The EDC tests performed at temperatures from 60 to 260 C show the existence of a transition temperature where an abrupt increase in the specimen hoop strain at failure occurs Experimental data from the EDC tests mapped against test temperature and hydrogen concentration clearly indicate separate “failure” and “non-failure” regions A simple relationship defining the transition temperature versus hydrogen concentration in the cladding has been obtained To verify a possible effect of RH, which is a specific condition for in-reactor RIA experiments compared to the SH typical of the EDC tests, the RHL tests have been performed on unirradiated hydrided ð 500 ppmÞ Zircaloy plate material The results from the RHL tests when both loading and heating are performed simultaneously and within 50–80 ms clearly indicate that strain to failure is dependent on the instantaneous material temperature and is not affected by the pre-heating history The results from the LS-1 and LS-2 RIA tests located in the failure map based on the EDC test data show good consistency between the EDC and inreactor pulse test data It is concluded that the failure of high burnup fuel under RIA conditions is mainly controlled by temperature and hydrogen concentration and that the EDC test can provide valuable information to predict fuel failure Acknowledgments The writers are grateful to Vattenfall AB, OKG Aktiebolag, Barseback Kraft AB, Westinghouse Electric Sweden AB, and Studsvik AB, which sponsored the development of the EDC test and, together with Kernkraftwerk Leibstadt AG, the testing of irradiated cladding from the KKL fuel rods The writers are grateful also to the Electric Power Research Institute, which initiated and sponsored the development of the RHL test and the testing of unirradiated hydrided Zircaloy plate material Special thanks are due to Dr Ken Yueh for continuing discussions and aid during the RHL development The writers would also like to acknowledge Sousan Abholassani, Didier Gavillet, and Holger Wiese (PSI) as well as Roger Lundstroăm and Soăren Karlsson (Studsvik) for material characterisation ID: kumarva Time: 15:24 I Path: Q:/3b2/STP#/Vol01529/AI-STP#018009 J_ID: DOI: Date: 18-October-11 Stage: Page: 1089 Total Pages: 17 GRIGORIEV ET AL., doi:10.1520/JAI102988 1089 References [1] Ledergerber, G., Abolhassani, S., Limbaăck, M., Lundmark, R., and Magnusson, K.-A., “Characterization of High Burn up Fuel for Safety Related Fuel Testing,” J Nucl Sci Technol., Vol 43, No 9, 2006, pp 1006–1014 [2] Grigoriev, V., Schrire, D., Jakobsson, R., Josefsson, B., and Jonasson, A., “Mechanical Testing of High Burnup PWR Cladding to Simulate RIA,” Proceedings of Annual Meeting on Nuclear Technology 2000: JAHRESTAGUNG KERTECHNIK, Bonn, May 23–25, 2000, INFORUM GmbH, Germany, pp 359362 ISSN 07209207 [3] Taăgtstroăm, P., Limbaăck, M., Dahlbaăck, M., Andersson, T., and Pettersson, H., Effects of Hydrogen Pickup and Second Phase Particle Dissolution on the InReactor Corrosion Performance of BWR Claddings,” Zirconium in the Nuclear Industry: 13th International Symposium, ASTM-STP-1423, 2002, G D Moan and P Rudling, Eds., ASTM International, West Conshohocken, PA., pp 96–118 [4] Ledergerber, G., Kaufmann, W., Ritter, A., and Greiner, D., “Burnup Increase and Power Uprate–Operation History of KKL,” Proceedings Intern LWR Fuel Performance Meeting, San Francisco, CA, Sep 30–Oct 3, 2007, Paper 1036 American Nuclear Society, ANS [5] Hermann, A., Wiese, H., Buăhner, R., Steinemann, M., and Bart, G., Hydrogen Distribution Between Fuel Cladding Metal and Overlying Corrosion Layers,” Proceedings Int Topical Meeting on LWR Fuel Performance, Park City, 2000, American Nuclear Society, ANS, pp 372–384, ISBN: 089448-656 [6] NEA, “Nuclear Fuel Behaviour Under Reactivity-Initiated Accident (RIA) Conditions,” Report NEA No 6847, Nuclear Energy Agency, OECD, Paris, France, 2010, ISBN 978-92-64-99113-2 [7] Sugiyama, T., Umeda, M., Fuketa, T., Sasajima, H., Udagawa, Y., and Nagase, F., “Failure of High Burnup Fuels Under Reactivity-Initiated Accident Conditions,” Ann Nucl Energy, Vol 36, 2009, pp 380–385 [8] Sugiyama, T., Umeda, M., Sasajima, H., Suzuki, M., and Fuketa, T., “Effect of Initial Coolant Temperature on Mechanical Fuel Failure Under Reactivity-Initiated Accident Conditions,” Proceedings of Top Fuel 2009, Paris, France, Sept 6–10, 2009, French Nuclear Energy Society, SFEN, Omnipress, Paper 2086 [9] Suzuki, M., Sugiyama, T., Udagawa, Y., Nagase, F., and Fuketa, T., “Comparative Analysis on Behavior of High Burnup PWR Fuels Pulse-Irradiated in ReactivityInitiated Accident Conditions,” Proceedings of Top Fuel 2009, Paris, France, Sept 6–10, 2009, French Nuclear Energy Scoiety, SFEN, Omnipress, Paper 2082 [10] Yueh, K., “Rapid Heating and Loading Test,” US NRC Archive, ADAMS Accession Number ML090861034, 2009 ID: kumarva Time: 15:24 I Path: Q:/3b2/STP#/Vol01529/AI-STP#018009 1090 ZIRCONIUM IN THE NUCLEAR INDUSTRY: 16TH SYMPOSIUM DISCUSSION Question 1, Joe Rashid, ANATECH:—The data you presented mixes two effects: radial hydrides and temperature Similar simulations but without radial hydrides and at higher strain rates, would be closer to RIA Author’s Response:—In the data presented, the mechanical simulations of RIA (EDC tests) have been performed on irradiated cladding with its existing hydride morphology The wording “radial hydrides” in the paper is only used for the RHL unirradiated samples subjected to uniaxial tension and defines the hydride orientation, which is, relatively to the tension direction, similar to the orientation of radial hydrides in the cladding The RHL tests have been mainly performed on the samples of hydrided plate material with their original hydride orientation (no radial hydrides) Only a limited number of the RHL samples were subjected to hydride re-orientation prior to the testing Both the EDC and the RHL tests were performed at strain rates, which are similar to an in-reactor RIA: depending on test temperature the time to failure varied from 10 ms to 130 ms in the RHL tests and from 20 ms to 90 ms in the EDC tests The LS-1 rod in the NSRR test failed at just under 10 ms Question 2, Ron Adamson, Zr+:—Could you comment on the comparison of the state of stress on the specimen in the EDC test and in the in-reactor RIA test? Does the difference affect any conclusions? Author’s Response:—This question has been under discussion since the first presentation of the EDC technique in 1999 Without doubts the in-reactor stress state of the cladding hardly possible to re-produce in out-of-pile mechanical tests Besides specific details of in-reactor through-wall-thickness stress distribution caused by the presence of temperature gradient in the cladding it is usually claimed as the main difference that the cladding experiences a pure hoop straining in the EDC tests rather than a biaxial loading as in the inreactor RIA Indeed, an axial stress component is expected to be much weaker in the EDC specimen, where an axial tension is only present in the outer part of cladding thickness owing to specimen “ballooning” during the loading However, an effect of stress state biaxiality on the material strain at failure is not so obvious One could expect this effect to be more important in ductile materials, while for brittle failures an absolute value of the tensile stress appears to be the decisive factor Such expectations are supported by the published data on biaxial burst tests performed on unirradiated and on irradiated cladding under different biaxiality ratios (published by U.S Nuclear Regulatory Commission, April 2000, NUREG-IA-0199/ IPSN 01-16/NSI RRC 2241): circumferential elongation of unirradiated cladding is highly sensitive to the biaxiality ratio, while the sensitivity of irradiated cladding to this ratio is very low This observation suggests that the conclusions made in the paper based on the EDC and/or RHL test data should not be greatly affected by the differences in the stress state under discussion In fact, the measured residual strain in the LS-1 test was within the range of strains measured in the low temperature EDC tests on cladding samples from the same fuel rod Question 3, Bo Cheng, EPRI:—Your conclusion is that the RIA issue is only associated with the hydrogen content and the temperature For current generation cladding, cladding hydriding resistance has been largely improved If you can establish a hydrogen limit of