Astm stp 1405 2001

STP 1405 Effects of Radiation on Materials: 20th International Symposium Stan T Rosinski, Martin L Grossbeck, Todd R Allen, and Arvind S Kumar, editors ASTM Stock Number: STP1405 ASTM 100 Barr H a r b o r Drive P.O Box C700 West Conshohocken, PA 19428-2959 Printed in the U.S.A Copyright by ASTM Int'l (all rights reserved); Sun Dec 20 18:18:21 EST 2015 Downloaded/printed by University of Washington (University of Washington) pursuant to License Agreement No further reproductions L i b r a r y of Congress ISBN: 0-8031-2878-9 Copyright 2001 AMERICAN SOCIETY FOR TESTING AND MATERIALS, 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 Photocopy Rights Authorization to photocopy items for internal, personal, or educational classroom use, or the internal, personal, or educational classroom use of specific clients, is granted by the American Society for Testing and Materials (ASTM) provided that the appropriate fee is paid to the Copyright Clearance Center, 222 Rosewood Drive, Danvers, MA 01923; Tel: 978-7508400; online: http://www.copyright.com/ Peer Review Policy Each paper published in this volume was evaluated by two peer reviewers and at least one editor The authors addressed all of the reviewers' comments to the satisfaction of both the technical editor(s) and the ASTM Committee on Publications To make technical information available as quickly as possible, the peer-reviewed papers in this publication were prepared "camera-ready" as submitted by the authors 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 maintains the anonymity of the peer reviewers The ASTM Committee on Publications acknowledges with appreciation their dedication and contribution of time and effort on behalf of ASTM Printed in Bridgeport, NJ July 2001 Copyright by ASTM Int'l (all rights reserved); Sun Dec 20 18:18:21 EST 2015 Downloaded/printed by University of Washington (University of Washington) pursuant to License Agreement No further reproductions authorized Dedication ARTHUR L LOWE, JR 1927-1999 Art Lowe's career spanned 48 years, starting in 1951 This included nearly 37 years with the Babcock & Wilcox Company in Lynchburg, VA He had a leadership role in the application of materials for nuclear fuel cladding and internals, the development of the reactor pressure vessel surveillance program, and in the evaluation of radiation effects on materials Art shared his knowledge and experience for over 30 years as an active member of Subcommittee El0.02 on Behavior and Use of Nuclear Structural Materials He has served on numerous other ASTM subcommittees concerned With the testing and evaluation of reactor pressure vessels, fuel cladding, and reactor internals In 1997 Art was presented with the Peter Hedgecock Award in recognition of his dedication to the activities of Committee El0 His efforts led to the development and refinement of numerous standards, to the presentation of many technical papers in the Effects of Radiation on Materials symposia series and other international forums, and to advancement of the materials technology used in nuclear power plants He has been a mentor to many, and a valued advisor to all He will be missed greatly Copyright by ASTM Int'l (all rights reserved); Sun Dec 20 18:18:21 EST 2015 Downloaded/printed by University of Washington (University of Washington) pursuant to License Agreement No further reproductions authorized Foreword This publication, Effects of Radiation on Materials: 20tn International Symposium, contains selected papers presented at the 20th Symposium on Effects of Radiation on Materials, held June, 2000 in Williamsburg, Virginia The symposium was sponsored by ASTM Committee El0 on Nuclear Technology and Applications The symposium chairman was Stan T Rosinski, Electric Power Research Institute Martin L Grossbeck, Oak Ridge National Laboratory, Todd R Allen, Argonne National Laboratory, and Arvind S Kumar, University of Missouri-Rolla served as co-chairmen Copyright by ASTM Int'l (all rights reserved); Sun Dec 20 18:18:21 EST 2015 Downloaded/printed by University of Washington (University of Washington) pursuant to License Agreement No further reproductions authori Contents Overview xi PRESSURE VESSEL S T E E L S ~ E N E R A L Review of Current Recommendations from the Recent IAEA Specialists Meeting on Irradiation Effects on Pressure Vessel Steels and its Mitigation -L M DAVIES, W L SERVER, V LYSSAKOV, AND S T ROSINSKI A Mechanistically-Based Model of Irradiation Damage in Low Alloy Steel Submerged Arc Welds T J WILLIAMS AND D ELLIS Vessel Investigation Program of " C H O O Z A" PWR Reactor after Shutdown -c BRILLAUD, Y GRANDJEAN, AND S SAILLET 28 Development of Reconstitution Technology for Surveillance Specimens in Japan Power Engineering and Inspection Corporation s KATAOKA, N KATO, K TAGUCH1, M YAMAMOTO, AND Y OKA 42 PRESSURE VESSEL S T E E L S - - M A S T E R CURVE APPROACH Master Curve Characterization of Irradiation Embrittlement Using Standard and l/3Sized Precracked Charpy Specimens B.-s LEE,W.-L YANG,M.-Y HUH,S.-H CHI, AND J.-H HONG Radiation Damage Assessment by the Use of Dynamic Toughness Measurements on PreCracked Charpy-V Specimens -E LUCONAND R CHAOUAD1 55 68 Comparison of Transition Temperature Shifts Between Static Fracture Toughness and Charpy-V Impact Properties Due to Irradiation and Post-Irradiation Annealing for Japanese A533B-1 Steels. K ONIZAWAAND M SUZUKI 79 Yield and Toughness Transition Predictions for Irradiated Steels Based on Dislocation Mechanics M WAGENHOFER,H P GUNAWARDANE,ANDM E NATISHAN 97 Master Curve Evaluation of Irradiated Russian VVER Type Reactor Pressure Vessel S t e e l s - - H - W VIEHRIG, J BOEHMART, J DZUGAN, AND H RICHTER 109 Copyright by ASTM Int'l (all rights reserved); Sun Dec 20 18:18:21 EST 2015 Downloaded/printed by University of Washington (University of Washington) pursuant to License Agreement No further reproductions authorized vi CONTENTS Fracture Toughness Characterization of 304L and 316L Austenitic Stainless Steels and Alloy 718 After Irradiation in High-Energy, Mixed Proton/Neutron S p e c t r u m - M A SOKOLOV, J P ROBERTSON, L L SNEAD, D J ALEXANDER, P FERGUSON, M R JAMES, S A MALOY, W SOMMER, G WILLCUTT, AND M R LOUTHAN 125 PRESSURE VESSEL STEELS MICROSTRUCTURE AND MODELING Review of Phosphorous Segregation and Intergranular Embrittlement in Reactor Pressure Vessel Steels c A ENGLISH,S R ORTNER,G GAGE,W L SERVER,AND S T ROSINSKI 151 Modeling of Phosphorous Accumulation on Grain Boundaries in Iron Alloys Under Irradiation v A PECHENKIN, A STEPANOV, AND Y V KONOBEEV 174 Grain Boundary Phosphorous Segregation Under Irradiation and Thermal Aging and Its Effect on the Ductile-to-BmTTLE T R A N S I T I O N - - S SONG, R G FAULKNER, AND P E J 189 FLEWITT An Evaluation of Through-Thickness Changes in Primary Damage Production in Commercial Reactor Pressure V e s s e l s - - - R E STOLLER AND L R GREENWOOD 204 Hardness and Microstructure Changes with Thermal Annealing of Neutron-Irradiated Fe-Cu Alloys H KAWANISHIAND M SUZUKI 218 Effects of Copper Concentration and Neutron Flux on Irradiation Hardening and Microstructure Evolution in Fe-Cu Model Alloys -R KASADA,T KITAO,K MORISHITA,H MATSUI,ANDA KIMURA 237 Effects of Neutron Irradiation and Thermal Annealing on Model Alloys Using Positron Annihilation Techniques -s E CUMBLIDGE, G L CATCHEN, A T MOTTA, G BRAUER, 247 AND J BOHMERT Microstructural Evolution in High Nickel Submerged Arc Welds j M HYDE,D ELLIS, C A ENGLISH, AND T J WILLIAMS 262 PRESSURE VESSEL STEELS MECHANICAL PROPERTIES An Evaluation of the Effect of Radiation Environment on Linde 80 Reactor Vessel Welds M j DEVAN AND W A PAVINICH 291 Reirradiation Response Rate of a High-Copper Reactor Pressure Vessel Weld s K., ISKANDER, R K NANSTAD, C A BALDWIN, D W HEATHERLY, M K MILLER, AND I REMEC 302 Relation Between Resistivity and Mechanical Properties in Heat Affected Zone of Welded Pressure Vessel SteelmR KASADA, T SUZUKI, K ITOH, Y NARUSE, AND A KIMURA 315 Copyright ASTM Int'l and (all rights reserved); Sun Dec 18:18:21 EST 2015 Pressure Vessel FracturebyToughness Tensile Properties of 20 Irradiated Reactor Downloaded/printed Cladding by MaterlalmM G HORSTENANDW P A BELCHER 328 University of Washington (University of Washington) pursuant to License Agreement No further reproductions authorized CONTENTS True C h a r a c t e r i s t i c s of S t r e n g t h a n d Ductility for N e u t r o n - I r r a d i a t e d Metals a n d Alloys -o P MAKSIMKIN AND O u TIVANOVA vii 343 Investigation of Temper Embrittlement in Reactor Pressure Vessel Steels Following T h e r m a l Aging, I r r a d i a t i o n , a n d T h e r m a l A n n e a l i n g - - R K NANSTAD,D E MCCABE, M A SOKOLOV,C A ENGLISH, AND S R ORTNER Composition Effects on the R a d i a t i o n E m b r i t t l e m e n t of Iron Alloys -J BOHMERT, A ULBRICHT,A KRYUKOV,Y NIKOLAEV,AND D ERAK 356 383 AUSTENITIC ALLOYS T h e D e t e r m i n a t i o n of Bias F a c t o r Stress D e p e n d e n c e f r o m E x p e r i m e n t a l Data on I r r a d i a t i o n Creep a n d Stress-Affected Swelling in Austenitic Stainless S t e e l s - Y V KONOBEEV, V A PECHENKIN, AND F A GARNER 401 Swelling a n d M i c r o s t r u c t u r a l Evolution in 316 Stainless Steel H e x a g o n a l Ducts Following L o n g - T e r m I r r a d i a t i o n in E B R - I I - - J I COLE, T R ALLEN, a TSAI, S UKAI, S MIZUTA, N AKASAKA, T DONOMAE, AND T YOSHITAKE 413 R a d i a t i o n - I n d u c e d Segregation a n d Void Swelling in 304 Stainless Steel T R ALLEN, 427 J I COLE, AND E A KENIK Effect of I r r a d i a t i o n E n v i r o n m e n t of Fast R e a c t o r ' s Fuel E l e m e n t s on Void Swelling in P, 443 Ti-Modified 316 Stainless SteeI N AKASAKA,1 YAMAGATA,AND S UKAI The Swelling Dependence of Cold W o r k e d C r - N i - M o - l M n Steel on N e u t r o n I r r a d i a t i o n in T e m p e r a t u r e , Fluence a n d D a m a g e Rate D u r i n g its Use as a C l a d d i n g M a t e r i a l in the BN-600 R e a c t o r - - A v KOZLOV,E A KINEV,S V BRYUSHKOVA, AND A PORTNYKH 457 Tensile P r o p e r t i e s of 12% C o l d - W o r k e d Type 316 Stainless Steel I r r a d i a t e d in E B R - I I U n d e r Lower-Dose-Rate Conditions to High Fluence -T YOSHITAKE, T DONOMAE, S MIZUTA,H TSAI, R V STRAIN, T R ALLEN, AND J I COLE 469 I r r a d i a t i o n C r e e p D e f o r m a t i o n of Modified 316 a n d 15Cr-20Ni Base Austenitic Fuel E l e m e n t s I r r a d i a t e d in F F T F - - A UEH1RA, S UKAI, S MIZUTA, AND R J PUIGH 487 Behavior of Different Austenitic Stainless Steels, Conventional, Reduced Activation (RA) and ODS C h r o m i u m - R i c h F e r r i t i c - M a r t e n s i t i c Steels U n d e r N e u t r o n I r r a d i a t i o n at 325~ in P W R E n v i r o n m e n t - - J - c BRACHET, X AVERTY, P LAMAGNi~RE,a ALAMO, F ROZENBLUM,O RAQUET,AND J.-L BERTIN 500 FERRITIC]MARTENSITIC ALLOYS P o s t - I r r a d i a t i o n D e f o r m a t i o n M i c r o s t r u c t u r e s in Fe-9CrmD S GELLES, M L HAMILTON, AND R SCH,~UBLIN 523 Copyright (allo nrights reserved); EST 2015F e r r i t i c / M a r t e n s i t i c Effect ofby S pASTM e c i m e nInt'l Size Fatigue P r o p e rSun t i e sDec of R20e d18:18:21 u c e d Activation Downloaded/printed by SteelmT H1ROSE, H SAKASEGAWA, A KOHYAMA, Y KATOH, AND H TANIGAWA 535 University of Washington (University of Washington) pursuant to License Agreement No further reproductions au viii CONTENTS Correlation Between Creep Properties and Microstructure of Reduced Activation Ferritic/Martensitic Steels -H SAKASEGAWA, 1" HIROSE, A KOHYAMA, Y KATOH, T HARADA, AND T HASEGAWA 546 Comparison of Thermal Creep and Irradiation Creep of HT9 Pressurized Tubes at Test Temperatures from ~490 ~ to ~ B TOLOCZKO, B R GRAMBAU, F A GARNER, 557 AND K ABE PROTON AND SPALLATION NEUTRON SOURCES Examination of 304L Stainless Steel to 6061-T6 Aluminum Inertia Welded Transition Joints after Irradiation in a Spallation Neutron Spectrum K A DUNN,M g LOUTHAN, JR., J I MICKALONIS, S MALOY, AND M R JAMES 573 Microstrucural Alteration of Structural Alloys by Low Temperature Irradiation with High Energy Protons and Spallation Nentrons -B H SENCER,G M BOND,F A GARNER, S A MALOY, W F SOMMER, AND M R JAMES Retention of Very High Levels of Helium and Hydrogen Generated in Various Structural Alloys by 800 MeV Protons and Spallation Neutrons -B M OLIVER,F A GARNER, S A MALOY,W F SOMMER,P D FERGUSON,AND M R JAMES 588 612 The Influence of High Energy Proton Irradiation on the Corrosion of Materials-S LILLARD,F GAC,M PACIOTTI,P FERGUSON,G WILLCUTT,G CHANDLER,ANDL DAEMEN 631 The Effect of High Energy Protons and Neutrons on the Tensile Properties of Materials Selected for the Target and Blanket Components in the Accelerator Production of Tritium Project s A MALOY, M R JAMES, G J WILLCUTT, W F SOMMER, w R JOHNSON, M R LOUTHAN, JR., M L HAMILTON, AND F A GARNER 644 High-Energy Spallation Neutron Effects on the Tensile Properties of Materials for the Target and Blanket Components for the Accelerator Production of Tritium Project M R JAMES, S A MALOY, W E SOMMER, W R JOHNSON, D A LOHMEIER, 660 AND M L HAMILTON RADIATION DAMAGE FUNDAMENTALS Microstructural Evolution of Reduced Activation and Conventional Martensitic Steels after Thermal Aging and Neutron Irradiation -M.-H MATHON,Y DE CARLAN, C GEOFFROY, X AVERTY, C.-H DE NOVION, AND A ALAMO 674 Dimensional Characteristics of Displacement Cascades in Austenitic Steels under Neutron Irradiation at Cryogenic Temperature -A v ~:OZLOV,I A PORXNYKH, L A SKRYABIN, AND S S LAPIN 694 On the a+T< >v-Phase Boundary in Nickel and in Manganese Containing Stainless Steel AIIoys -w SCH1DLE 704 Copyright by ASTM Int'l (all rights reserved); Sun Dec 20 18:18:21 EST 2015 Downloaded/printed by University of Washington (University of Washington) pursuant to License Agreement No further reproductions authorized CONTENTS ix Effect of Nickel on Irradiation Hardening and Microstructure Evolution of Proton Irradiated Fe-Cu Alloys - SHIBAMOTO, T KITAO, H MATSUI, M HASEGAWA, 722 S YAMAGUCHI, AND A KIMURA OTHER MATERIALS Effect of Final Irradiation Temperature and Frequency of Irradiation Temperature Cycles on Microstructural Evolution of Vanadium Alloys N NITA,K FUKUMOTO, AND H MATSUI 736 Hardening of Vanadium Doped with Nitrogen by Heavy Ion Irradiation and PostIrradiation Annealing T NAGASAKA, H TAKAHASHI,T MUROGA, N YOSHIDA, AND T TANABE 746 Hydrogen and Helium Gas Formation and their Release Kinetics in Tungsten Rods after Irradiation with 800 MeV Protons B M OLIVER,F A GARNER,M L HAMILTON, W F SOMMER, S A MALOY, P D FERGUSON, AND M R JAMES 762 The Influence of Temperature, Fluence, Dose Rate, and Helium Production on Defect Accumulation and Swelling in Silicon C a r b i d e - - KISHIMOTO,Y KATOH, A KOHYAMA,ANDM ANDO 775 Microstuctural Stability of SiC/SiC Composites under Dual-Beam Ion Irradiation Y KATOH,H KISHIMOTO,M ANDO,A KOHYAMA,T SH1BAYAMA,AND H TAKAHASHI 786 Molecular Dynamics Simulation of Radiation Damage Production in Cubic Silicon Carbide -L MALERBA, J M PERLADO, I PASTOR, AND T D DE LA RUBIA 799 Influence of the Reactor and Cyclotron Irradiation on Energy Transformation during Plastic Deformation of Metal Materials -o P MAKSIMKINANDM N GUSEV 813 Irradiation-Induced Amorphization and its Recovery Behavior in Cold-Rolled and Aged Ti-Ni Shape Memory Alloys -A KIMURA,T HIROSE,AND H MATSU1 825 Effect of Mass and Energy on Preferential Amorphization in Polycrystalline Silicon Film During Ion Irradiation M TAKEDA, S OHNUKI, T SUDA, S WATANEBE, H ABE, AND NASH1YAMA 836 Crack Growth Resistance of Irradiated Zr-2,5Nb Pressure Tube Material at Low Hydrogen Levels -P H DAVIES, D D HIMBEAULT, R S W SHEWFELT, AND R R HOSBONS 846 Copyright by ASTM Int'l (all rights reserved); Sun Dec 20 18:18:21 EST 2015 Downloaded/printed by University of Washington (University of Washington) pursuant to License Agreement No further reproductions autho 864 EFFECTSOF RADIATION ON MATERIALS Figure 21 - Comparison of J-R curves at 250~ from standard burst tests sampled from two different axial locations of surveillance tube 508 ( Cl < 0.2 wt ppm, P < 57 wt ppm) T,rriS the irradiation temperature in ~ The difference in fast neutron fluence is relatively small and varied from to 11 x 1025n.m2for material sampled from the center and the inlet end of the pressure tube, respectively 600 :,2 i h L Crack extension, m m strength proved reasonably successful (see Figure 16 in reference 7) However, as results from tubes installed in later reactors have become available, it is now clear that such relationships are sensitive to the initial deformation characteristics of the individual tube, e.g initial yield stress and twinning stress For example, variability in yield stress can arise from variations in the oxygen and niobium concentrations [31] as well as the initial dislocation density [31], grain structure and texture [32] Such variations can result from minor changes in ingot chemistry, fabrication route and extrusion variables [33-35] The sensitivity of Ihe crack growth resistance of irradiated Zr-2.5Nb pressure tube material to the initial deformation properties of the tube at 250~ is due to the similarity of the yield and twinning stresses in the operating temperature regime after irradiation [26] These parameters determine not only the local crack-tip stress for void nucleation, but also the propensity for deformation by dislocation channeling and/or twinning compared with slip, i.e., for strain localization Such strain localization contributes to the failure of ligaments between neighboring voids (stage and crack growth) and controls the onset of slant fracture at the surface (stage 3) Therefore, removal of void nucleation sites is particularly beneficial This is because it increases the effective wall thickness for local ligament failure, and also reduces the extents of stable crack growth and tunneling which orient the crack front for easy development of the "sliding-off" mechanism Work is now underway to extend the database of burst test results and to determine the quantitative relationship between crack growth resistance and CI using material with intermediate levels of CI and low levels of zirconium phosphide The initial deformation characteristics of the burst test material are also being studied with the aim of elucidating the microstructural factors responsible for the variability in results in Figures 19 and 20 This should allow further rationalization of the results obtained from different reactor units as well as the potential for producing further improvements in crack growth resistance Copyright by ASTM Int'l (all rights reserved); Sun Dec 20 18:18:21 EST 2015 Downloaded/printed by University of Washington (University of Washington) pursuant to License Agreement No further reproductions authoriz DAVIES ET AL ON CRACK GROWTH RESISTANCE 865 Summary The key factors controlling the crack growth resistance of irradiated Zr-2.5Nb pressure tube material at 250~ at low levels of hydrogen/deuterium have been reviewed using the current database of small- and large-scale specimen results from different CANDU reactor pressure tubes The review highlights the role of the following factors 1) The mixed-mode nature of the crack growth process in the irradiated, thin-walled material, i.e flat, transition and slant fracture modes 2) The relative proportions and energy-absorbing capacities of the different fracture modes, which is determined by the ease of void nucleation, growth and coalescence ahead of the crack tip 3) The role of primary void nucleating particles, e.g Zr-C1-C (complex carbide), zirconium phosphide and carbide, the beneficial effects of eliminating them For example, the removal of chlorine and Zr-CI-C particles by quadruple vacuum-arc-melting 4) The deformation characteristics of the matrix, e.g yield stress and twinning stress, which control the tendency! for void nucleation, growth, coalescence and strain localization (work-softening) 5) The specimen geometry, e.g crack size, geometry and material effects in influencing the relationship between the J-R curves obtained from different specimens The current results from curved compact specimens, having a low concentration of zirconium phosphide (P < 20 wt ppm), suggest a limiting level of C1 (about wt ppm) above which no further significant degradation in the crack growth resistance occurs Such results require confirmation using rising-pressure burst tests on material with intermediate levels of Cl and low levels of zirconium phosphide The influence of minor variations in tube fabrication on the deformation behavior of Zr-2.5Nb pressure tube material (item above) also requires study This should allow further rationalization of the results obtained from different reactor units as well as the potential for producing further improvements in crack growth resistance Acknowledgments Many individuals contributed to the production of the results reported in this work over a number of years In particular, thanks are due to R Behnke, G R Brady, G R Kasprick, G C Longhurst, A R Reich, J M Smeltzer, A K West, as well as the hot cell staff at the Whiteshell and Chalk River Laboratories for their technical assistance Special thanks go to R R Bawden and J C Owens at Chalk River for retrieval and shipment of surveillance material for testing and to G D Moan for supplying unirradiated archive material for testing Funding of this work by the CANDU Owners Group (COG) under Work Packages 3192, 3305, 6506 and 6536 and contract COG-00136t3 is gratefully acknowledged Copyright by ASTM Int'l (all rights reserved); Sun Dec 20 18:18:21 EST 2015 Downloaded/printed by University of Washington (University of Washington) pursuant to License Agreement No further reproductions authoriz 866 EFFECTSOF RADIATION ON MATERIALS References [1] [2] [3] [4] [5] [6] [7] [8] [9] [10] [11] [12] [13] [14] Cheadle, B A., Coleman, C E., and Licht, H., Nuclear Technology, Vol 57, 1982, pp 413-425 Chow, C K and Simpson, L A., Fracture Mechanics: Eighteenth Symposium, ASTM STP 945, D T Read, R P Read, Eds., American Society for Testing and Materials, West Conshohocken, PA, 1988, pp 419-439 Simpson, L A., Chow, C K., and Davies, P H., CANDU Owners Group Report No COG-89-110-I, AECL, September 1989 Himbeault, D D., and Davies, P H., CANDU Owners Group Report No COG-98-161-I, AECL, January 1999 Davies, P H Shewfelt, R S W and J~vine, A K., Constraint Effects in Fracture: Theory and Applications, ASTM STP 1244, M Kirk and A Bakker, Eds., American Society for Testing and Materials, West Conshohocken, PA, 1995, pp 392-424 Davies, P H and Shewfelt, R S W., Zirconium in the Nuclear Industry: Eleventh International Symposium, ASTM STP 1295, E R Bradley and G P Sabol, Eds., American Society for Testing and Materials, West Conshohocken, PA, 1996, pp 492-517 Davies, P H and Shewfelt, R S W., Zirconium in the Nuclear Industry: Twelfth International Symposium, ASTM STP 1354, G P Sabol and G D Moan, Eds., American Society for Testing and Materials, West Conshohocken, PA, 2000, pp 356-376 Kiefner, J F., Maxey, W A., Eiber, R J and Duffy, A R., Progress in Flaw Growth and Fracture Toughness Testing, ASTM STP 536, American Society for Testing and Materials, West Conshohocken, PA, 1973, pp 461-481 Rice, J R., Journal of Applied Mechanics, Transactions of ASME, Vol 35, 1968, pp379-386 Turner, C E., Size Effects, Mechanical Engineering Publications, London, 1986, pp 25-31 Ernst, H A., Elastic-Plastic Fracture: Second Symposium, Volume L ASTM STP 803, C F Shih and J P Gudas, Eds., American Society for Testing and Materials, 1983, pp 1191-1213 Turner, C E., Fracture Mechanics: Twenty Second Symposium (Volume I), ASTM STP 1131, H A Ernst, A Saxena and D L McDowell, Eds., American Society for Testing and Materials, 1992, pp 71-92 Davies, P H., Fatigue and Fracture Mechanics: Twenty Eighth Volume, ASTM STP 1321, J H Underwood, B D MacDonald and M R Mitchell, Eds., American Society for Testing and Materials, West Conshohocken, PA, 1997, pp 535-561 Chow, C K and Simpson, L A., Fracture Mechanics: Nonlinear Fracture Mechanics- Volume II-Elastic-Plastic Fracture, ASTM STP 995, J D Landes, A Saxena and J G Merkle, Eds., American Society for Testing and Materials, West Conshohocken, PA, 1989, pp 537-562 Copyright by ASTM Int'l (all rights reserved); Sun Dec 20 18:18:21 EST 2015 Downloaded/printed by University of Washington (University of Washington) pursuant to License Agreement No further reproductions authorized DAVIES ET AL ON CRACK GROWTH RESISTANCE [15] [16] [17] [18] [19] [20] [21] [22] [23] [24] [25] [26] [27] [28] [29] 867 Thomason, P F., Ductile Fracture of Metals, Pergamon Press, Oxford, Ch to 4, 1990, pp 1-114 Rltchie, R O., Server, W L., and Wullaert, R A., Metallurgical Transactions A, Vol 10A, 1979, pp 1557-1570 Hutchinson, J W., and Tvergaard, V., Fracture Mechanics; Perspectives and Directions (Twentieth Symposium), ASTM STP 1020, R P Wei and R P Gangloff, Eds., American Society for Testing and Materials, 1989, pp 6183 Shih, C F., Journal of Mechanics and Physics of Solids, Vol 29, 1981, pp 305-326 Dodds, R H., Shih, C F., and Anderson, T., International Journal of Fracture, Vol 64, 1993, pp 101-133 Wallace, A C., Shek, G K., and Lepik, O E., Zirconium in the Nuclear Industry: Eighth International Symposium, ASTM STP 1023, L F Van Swam and C M Eucken, Eds., American Society for Testing and Materials, 1989, pp 66-88 Aitchison, I and Davies, P H., Journal of Nuclear Materials, Vol 203, 1993, pp 206-220 Davies, P H., Hosbons, R R., Griffiths, M and Chow, C K., Zirconium in the Nuclear Industry: Tenth International Symposium, ASTM STP 1245, A M Garde and E R Bradley, Eds., American Society for Testing and Materials West Conshohocken, PA, 1994, pp 135-167 Davies, P H., Aitchison, I., Himbeault, D D., J~'vine, A K and Watters, J F., Fatigue and Fracture of Engineering Materials and Structures, Vol 18, 1995, pp 789-800 Theaker, J R., Choubey, R., Moan, G D., Aldridge, S A., Davies, L., Graham, R A and Coleman, C E., Zirconium in the Nuclear Industry: Tenth International Symposium, ASTM STP 1245, A M Garde and E R Bradley, Eds., American Society for Testing and Materials West Conshohocken, PA, 1994, pp 221-242 Dutton, R., AECL Report No 9930, 1989 April Himbeault, D D., Chow, C K and Puls, M P., Metallurgical and Materials Transactions A, Vol 25A, 1994, pp 135-145 Hosbons, R R., Davies, P H Griffiths, M Sagat, S and Coleman, C E., Zirconium in the Nuclear Industry: Twelfth International Symposium, ASTM STP 1354, G P Sabol and G D Moan, Eds., American Society for Testing and Materials, West Conshohocken, PA, 2000, pp 122-138 Chow, C K., Coleman, C E., Hosbons, R R., Davies, P H., Griffiths, M., and Choubey, R., Zirconium in the Nuclear Industry: Ninth International Symposium, ASTM STP 1132, C M Eucken and A M Grade, Eds., American Society for Testing and Materials, West Conshohocken, PA, 1992, pp 246-275 Sagat, S., Coleman, C E., Griffiths, M., and Wilkins, B J S., Zirconium in the Nuclear Industry: Tenth International Symposium, ASTM STP 1245, A M Garde and E R Bradley, Eds., American Society for Testing and Materials, West Conshohocken, PA, 1994, pp 35-61 Copyright by ASTM Int'l (all rights reserved); Sun Dec 20 18:18:21 EST 2015 Downloaded/printed by University of Washington (University of Washington) pursuant to License Agreement No further reproductions authorized 868 [30] [31] [32] [33] [34] [35] EFFECTS OF RADIATION ON MATERIALS Tabatabai, M and Byrne, T., Ontario Hydro Report No A-NFC-94-34-K, 1994 April Winton, J., Murgatroyd, R A., Watkins, B., and Nichols, R W., Transactions of the Japanese Institute of Metals, Volume Supplement, 1968, pp 630-636 Cheadle, B., and Ells, C E., Transactions of the Metallurgical Society of AIME, Vol 233, 1965, pp 1044-1052 Cheadle, B., Zirconium in the Nuclear Industry, ASTM STP 633, A L Lowe, Jr., G W Parry, Eds., American Society for Testing and Materials West Conshohocken, PA, 1977, pp 457-485 Holt, R A and Aldridge, S A., Journal of Nuclear Materials, Vol 135, 1985, pp 246-259 Choubey, R., Aldridge, S A., Theaker, J R., Cann, C D., and Coleman, C E., Zirconium in the Nuclear Industry: Eleventh International Symposium, ASTM STP 1295, E R Bradley and G P Sabol, Eds., American Society for Testing and Materials West Conshohocken, PA, 1996, pp 657-675 Copyright by ASTM Int'l (all rights reserved); Sun Dec 20 18:18:21 EST 2015 Downloaded/printed by University of Washington (University of Washington) pursuant to License Agreement No further reproductions authorized STP1405-EB/Jul 2001 Audlflor Index A E Abe, H., 836 Abe, K., 557 Akasaka, N., 413, 443 Alamo, A., 500, 674 Alexander, D J., 125 Allen, T R., 413, 427, 469 Ando, M., 775, 786 Averty, X., 500, 674 Ellis, D., 8, 262 English, C A., 151, 262, 356 Erak, D., 383 F Faulkner, R G., 189 Ferguson, P D., 125, 612, 631, 762 Flewitt, P E J., 189 Fukumoto, K., 736 B Baldwin, C A., 302 Belcher, W P A., 328 Bertin, J.-L., 500 Boehmert, J., 109 B6hmert, J., 247, 383 Bond, G M., 588 Brachet, J C., 500 Braner, G., 247 Brillaud, C., 28 Bryushkova, S V., 457 G Gac, F., 631 Gage, G., 151 Garner, F A., 401, 557, 588, 612, 644, 762 Gelles, D S., 523 Geoffroy, G., 674 Grambau, B R., 557 Grandjean, Y., 28 Greenwood, L R., 204 Gunawardane, H P., 97 Gusev, M N., 813 C Catchen, G L., 247 Chandler, G., 631 Chaouadi, R., 68 Chi, S., 55 Cole, J I., 413, 427, 469 Cumblidge, S E., 247 I-I Hamilton, M L., 523, 644, 660, 762 Harada, T., 546 Hasegawa, T., 546, 722 Heatherly, D W., 302 Himbeault, D D., 846 Hirose, T., 535, 546, 825 Hong, J.-H., 55 Horsten, M G., 328 Hosbons, R R., 846 Huh, M Y., 55 Hyde, J M., 262 D Daemen, L., 631 Davies, L M., Davies, P H., 846 De Carlan, Y D., 674 De Nouvion, C.-H., 674 DeVan, M J., 291 Diaz de la Rubia, T., 799 Donomae, T., 413, 469 Dunn, K A., 573 Dzugan, J., 109 Iksander, S K., 302 Itoh, K., 315 869 Copyright by ASTM Int'l (all rights reserved); Sun Dec 20 18:18:21 EST 2015 Downloaded/printed by University of Washington (University of Washington) pursuant to License Agreement No further reproductions authorized 870 EFFECTS OF RADIATION ON MATERIALS N James, M R., 125, 573, 612, 644, 660, 762 Johnson, W R., 644, 660 K Kasada, R., 237, 315 Kataoka, S., 42 Kato, N., 42 Katoh, Y., 535, 546, 775, 786 Kawanishi, H., 218 Kenik, E A., 427 Kimura, A., 237, 315, 722, 825 Kinev, E A., 457 Kishimoto, H., 775, 786 Kitao, T., 237, 722 Kohyama, A., 535, 546, 775, 786 Konobeev, Y V., 174, 401 Kozlov, A V., 457, 694 Kryukov, A., 383 L l_amagn~re, P., 500 Lapin, S S., 694 Lee, B S., 55 Lillard, S., 631 Lohmeier, D A., 660 Louthan, M R., Jr., 125, 573, 644 Lucon, E., 68 Lyssakov, V., Nagasaka, T., 746 Nanstad, R IC, 302, 356 Naruse, Y., 315 Nashiyama, I., 836 Natishan, M E., 97 Nikolaev, Y., 383 Nita, N., 736 O Ohnuki, S., 836 Oka, Y., 42 Oliver, B M., 612, 762 Onizawa, K., 79 Ortner, S R., 151, 356 P Paciotti, M., 631 Pastor, I., 799 Pavinich, W A., 291 Pechenkin, V A., 174, 401 Perlado, I., 799 Portnykh, I A., 457, 694 Puigh, R J., 487 R Raquet, O., 500 Remec, I., 302 Richter, H., 109 Robertson, J P., 125 Rosinski, S T., 3, 151 Rozenblum, F., 500 S M Maksimkin, O P., 343, 813 Malerba, L, 799 Maloy, S A., 125, 573, 588, 612, 644, 660, 762 Mathon, M.-H., 674 Matsui, H., 237, 722, 736, 825 MeCabe, D E., 356 Mickalonis, J I., 573 Miller, M K., 302 Mizuta, S., 413, 469, 487 Morishita, K., 237 Motta, A T., 247 Muroga, T., 746 Saillet, S., 28 Sakasegawa, H., 535, 546 Sch~iublin, R., 523 Schtile, W., 704 Sencer, B H., 588 Server, W L., 3, 151 Shewfelt, R S W., 846 Shibamoto, H., 722 Shibayama, T., 786 Skyrabin, L A., 694 Snead, L L., 125 Sokolov, M A., 125, 356 Sommer, W., 125, 588, 612, 644, 660, 762 Copyright by ASTM Int'l (all rights reserved); Sun Dec 20 18:18:21 EST 2015 Downloaded/printed by University of Washington (University of Washington) pursuant to License Agreement No further reproductions authorized INDEX 871 Song, S., 189 Stepanov, I A., 174 Stoller, R E., 204 Strain, R V., 469 Suda, T., 836 Suzuki, M., 79, 218, 315 T Taguchi, K., 42 Takahashi, H., 746, 786 Takeda, M., 836 Tanabe, T., 746 Tanigawa, H., 535 Tivanova, O V., 343 Tolockzo, M B., 557 Tsai, H., 413, 469 U Uehira, A., 487 Ukai, S., 413, 443, 487 Ulbricht, A., 383 Viehrig, H W., 109 W Wagenhofer, M., 97 Watanabe, S., 836 Willcutt, G., 125, 631, 644 Williams, T J., 8, 262 u Yamagata, I., 443 Yamaguchi, S., 722 Yamamoto, M., 42 Yang, W J., 55 Yoshida, N., 746 Yoshitake, T., 413, 469 Copyright by ASTM Int'l (all rights reserved); Sun Dec 20 18:18:21 EST 2015 Downloaded/printed by University of Washington (University of Washington) pursuant to License Agreement No further reproductions authorized STP1405-EB/Jul 2001 Subject Index BN-600 reactor, 457 Brittleness, 704 A Absorbed energy, 42 Accelerator Production of Tritium Program, 588, 612 material tensile properties, 644 spallation neutron effects, 660 target/blanket system, 125 transition joints, 573 tungsten rods, 762 Activation, reduced, 500, 535, 546, 674 Aging, thermal, 151, 189, 328, 356, 674 Alexandre radiation, 500 Alloy 718, 125, 588, 631, 660 Alpha-ferrite, 704 Alpha-martensite, 704 Aluminum, 573, 631 Amorphization, irradiation induced, 786, 825, 836 Analytical electron microscopy, 189 Annealing, 302 behavior, 109, 237 isochronal, 722 post-irradiation, 218, 315, 825 post-irradiation isochronal, 247 radiation, hardening, 746 recoveries, 79 thermal, 151, 218, 247, 356 AA~pliedstress, 813 TM standards E 1820, 125 E 1921, 55, 68, 79 Atomic displacement cascades, 204 Atom probe microscopy, 8, 262 Auger analysis, 356 Austenite boundaries, 546 B Babcock and Wilcox Owners Group, 291 C Carbide, 546 Ceramic seals, 631 Charpy, 28 impact response, 302 impact tests, 55, 383 precracked, 79, 109 preeracked Ch arpy-V specamem, 68 shift, V-notch, 109, 302, 356 Chemical vapor infiltration, 786 Chlorine, 846 CHOOZ A reactor, 28 Chromium, 674 depletion, 427 chromium-molybdenum steel, 644 chromium-nickel austenitic stainless steel, 487 chromium-nickelmolybdenum, 343 chromium-nickel-molybdenummanganese steel, 457 chromium-nickel-titanium, 343 chromium-rich martemitic steels, 500 iron-chromium, 523 Cladding, 328, 443, 457 Cleavage fracture, 97 Cleavage initiation fracture toughness tests, 55 Cluster accumulation, 694 Compact tension specimem, 125 Composites, silicon carbide, 786 Constraint, loss of, 68 Copper, 8, 262, 573 concentration, 722 effects on embrittlement, 383 ion irradiation, 746 iron-copper alloys, 218, 237, 722 873 Copyright by ASTM Int'l (all rights reserved); Sun Dec 20 18:18:21 EST 2015 Downloaded/printed by University of Washington (University of Washington) pursuant to License Agreement No further reproductions authorized 874 EFFECTS OF RADIATION ON MATERIALS iron-copper-nickel-phosphorus alloys, 247 precipitate, 315 precipitate dissolution, 237 weld, 302 Corrosion, 573 properties, 500 rates, 631 Crack-extension resistance, 125 Crack growth resistance curve, 846 Creep, 401, 557 compliance, 487 creep rupture test, 546 deformation, 487 swelling, 487 thermal, 557 Cross section method, 722 Cryogenic neutron irradiation, 694 Curve toughness, 125 D Damage, 204 accumulation, 204 assessment, 68 attenuation, 204 dose, 457 estimation, prediction, radiation, 799 production, 836 rate, 457 structures, 722 Defect accumulation, 775 Defect dusters, 247 Defect formation, 694 Defect growth, 736 Deformation, 343, 523 irradiation creep, 487 plastic, 813 Differential scanning calorimetry, 825 Diffusional loss, 612, 762 Diffusion parameters, 174 Dislocation, 413, 523, 704 defects, 247 density, 546 loops, 722 mechanics, 97 Displacement cascades, 204, 694, 799 Displacement damage, 487, 775 Displacement rates, 427 Dissipation of heat, 813 Doppler broadening, 247 Dose dependence, 401, 443 Dose rate, 469 effects, 413, 427, 775 Dual-beam irradiation method, 775, 786 Ductile-brittle transition, 189 region, 79 temperature, 109, 315 Ductility, 343, 469, 588, 660 characteristics, 813 loss of, 328 Duets, hexagonal, 413, 469 E EBR-II reflector duct, 469 Electrochemical impedance spectroscopy, 631 Electron microscopy, 694 Elongation, 588, 644 Embrittlement, 8, 28, 68, 79, 262 composition effects on, 383 ~rarainboundary, 174 dening, 189 intergranular, 151 master curve characterization, 55 radiation, sensitivity, 109 temper, 356 trends, 291 Energy deposition, 631 Energy dissipation and accumulation, 813 Energy transformation, 813 Expenmental Breeder Reactor-II, 413, 427, 469 Extensometer, laser, 535 F Failure mode, 469 Failure stress, 343 Fatigue life characteristics, 535 Ferritic steel, 79, 204, 523, 674 ferritic-martensitic steels, 50, 500, 535, 546, 557 Copyright by ASTM Int'l (all rights reserved); Sun Dec 20 18:18:21 EST 2015 Downloaded/printed by University of Washington (University of Washington) pursuant to License Agreement No further reproductions authorized INDEX 875 toughness transition predictions, 97 transition range, 55 FFFF/MOTA, 443, 487 Field emission gun scanning transmission electron microscopy, 262 Flow tubes, 631 Fluence, 457, 644, 660 critical, 836 effects, silicon carbide, 775 fast neutron, 79 increase, 813 levels, 302, 343, 500 variation, 28 Flux effects, 237 Fractographic examinations, 469, 523 Fracture mechanisms, 535 Fracture resistance, 328 Fracture toughness, 79, 125, 846 dynamic, 68 shift, tests, 55 Fuel pin, 487 G Galvanic corrosion, 573 Gas production cross sections, 612 German/Russian irradiation program, 109 Gibbsian adsorption, 174 Grain boundaries, 151, 356, 546 behavior, 836 composition, 427 phosphorus segregation, 174, 189 Grain evolution, 546 H Hardening, irradiation, 247, 315, 469 iron-copper alloys, 237, 722 vanadium, 746 Hardness, 8, 218, 315 distribution, 42 micro-Vickers test, 736, 825 Rockwell, 247 Heat affected zone, 42, 151, 315, 356 Heat, dissipation of, 813 Heat treatment, post irradiation, 786 Heavy Section Steel Irradiation Program, 302, 356 Helium, 588, 762, 786 production, 612, 775 High flux reactor Petten, 328 Hi-Nicalon Type S, 786 Hoop stress, 401 Hydrogen, 588 concentration, dissolved, 631 evolution, 762 hydrogen/deuterium, 846 measurements, 762 retention, 612 Hydrostatic stress component, 401 I Immersion density, 413 Impact loading rate, 68 Impact toughness, 356 Integrity assessment, 109 Interferometric profilometry, 775 Intergranular fracture, 356 International Atomic Energy Agency, 3, 55 International Working Group on Plant Lifetime Management, Ion mass, 836 Iron alloys, 174, 262, 383 iron-based, 612 iron-based stainless steels, 588 iron-chromium, 523 iron-chromium-manganese, 704 iron-chromium-nickel, 704 iron-copper, 218, 237, 722 iron-copper-nickel, 722 iron-copper-nickelphosphorus, 247 Irradiation environment, effect on embrittlement trends, 291 Copyright by ASTM Int'l (all rights reserved); Sun Dec 20 18:18:21 EST 2015 Downloaded/printed by University of Washington (University of Washington) pursuant to License Agreement No further reproductions authorized 876 EFFECTS OF RADIATION ON MATERIALS fracture toughness transition behavior, 97 Japan Materials Testing VVER steel evaluation, 109 Reactor, 79, 237,315, Master Integrated Reactor Vessel 736, 825 Surveillance Program, 291 Japan Power Engineering and Matrix damage, 247 Inspection Corporation, 42 Mechanical energy, 813 J-integral resistance curve Mechanical strength, 786 toughness tests, 125 Micro-indentation test, 746 J-R curve, 125, 846 Microscopy analytical electron, 189 L atom probe, 8, 262 electron, paRe.ms, 694 I.AI-IET, 762 field gun scanning transmission Laser extensometer, 535 electron, 262 Lateral shift method, 302 transmission electron, 674 Lattice structure, 97 Microstructure, 262, 356, 746, 825 Laves phase, 546, 674 alteration, 588 Life management, power plant, changes, 218, 457, 546 characterization, 775 Linde 80 reactor vessel welds, deformation, 523 291 evolution, 237, 262, 413, 674 Loos iron-copper model alloys, ~ank, 588 722 irradiation, 523 modified stainless steel, Los Alamos Neutron Science 443 Center, 644, 660 vanadium alloys, 736 corrosion tests, 631 recovery, 546 helium and hydrogen retention silicon carbide, 775, 786 tests, 612 stability, silicon carbide mixed proton-neutron particle composites, 786 flux and spectra tests, 588 steel, 97 tritium target/blanket Micro-Vickers hardness test, 736 assembly tests, 573 Models and modeling, 204, 799 tungsten rod irradiation, 762 McLean's, 174 modulus hardening, Zerilli-Armstrong M constitutive, 97 Modulus hardening model, Molecular dynamics, 204 Manganese, 262, 315, 704 simulation, 799 chromium-nickel-molybdenumMolybdenum, 343, 775 manganese steel, 457 chromium-molybdenum Martensite, 813 steel, 644 transformation, 343, 825 chromium-nickel-molybdenumMartensitic steel, 674, 825 manganese steel, 457 chromium, 500 ferritic-martensitic, 500, 535, 546 N Master curve, 28, 79 embrittlement Necking, 343 characterization, 55 Neutron dose rate, Copyright by ASTM Int'l (all rights reserved); Sun Dec 20 18:18:21 EST 2015 Downloaded/printed by University of Washington (University of Washington) pursuant to License Agreement No further reproductions authorized INDEX 877 Neutron fluence, 79 Neutron scattering, small angle, 8, 383 Nickel, 8, 262, 315, 383 alloys, 722 chromium-nickel, 487 chromium-nickel-titanium, 343 chromium-nickel-molybdenummanganese steel, 457 enrichment, 427 iron-copper-nickel-phosphorus alloys, 247 nickel-based Alloy 718, 588, 631, 660 nickel-based structural alloys, 612 phase boundary, 704 submerged arc welds, 262 titanium-nickel, 825 Nitrogen effect, 746 Nondestructive evaluation, 315 Nucleation, 704 Positron annihilation lifetime spectrometry, 237 spectroscopy, 247, 736 Positron lifetime, 247 Power plant life management, Precipitate structures, 413 Precision intefferometric profilometry, 775 Pressure tube, 846 Pressurized tubes, 487, 557 Profilometry, interferometric, 775 Proton/neutron spectrum, 125 R Reconstitution technology, 42 Recovery, 722 Reference temperature, 28, 55, 68, 79, 109 Reflector duct, 469 Reirradiation response rate, 302 Resistivity, 315 Russian VVER reactor, 109 O Optical position sensitive atom probe experiments, 262 Order-disorder transformation, 218 Osiris reactor, 500 P Particle flux, 631 Particle type, 631 Petten reactor, 328 Phase diagram, 704 Phase separation, 674 Phosphorus, 383, 443 accumulation, 174 intergranular segregation, 189 segregation, 151 Pitting, 573 Plastic deformation, 343, 813 Plasticity, 694 Plastic zone width, 42 Point defects, 204, 775, 799 Polycrystalline silicon film, 836 Polymer impregnation and pyrolysis methods, 786 Scatterinjg, 8, 262, 383, 674 Segregation phosphorus, 151, 189, 356 radiation-induced, 174, 427 Shape memory effects, 825 Shutdown, reactor, 28 Silicon, 262 carbide, 775, 799 carbide composites, 786 film, 836 Slant fracture, 846 Slip dislocations, 523 Small angle neutron scattering, 8, 262, 383, 674 Small specimen testing technology, 535 Spallation, 573,588, 612, 631 neutron effects, 660 neutrons, 762 Spectrometry, 762 positron annihilation, 237 Spectroscopy, electrical impedance, 631 Spectroscopy, positron annihiliation, 247, 736 Copyright by ASTM Int'l (all rights reserved); Sun Dec 20 18:18:21 EST 2015 Downloaded/printed by University of Washington (University of Washington) pursuant to License Agreement No further reproductions authorized 878 EFFECTS OF RADIATION ON MATERIALS Steel, 28, 189, 218 A302, 356 A508, 356 A533B, 79, 237, 315, 356 austenitic, 694 cold worked, 457, 469 ferritic, 55, 79, 204 ferritic-martensitic, 50, 535, 557 low alloy, 8, 42 martensitic, 674 martensitic, chromium-rich, 500 stainless, 401, 413, 427, 443 stainless, AL2, 573 stainless, annealed, 660 stainless, austenitic, 125, 343, 401, 427, 487 stainless, manganese containing, 704 stainless, nickel containing, 704 stainless, strip clad deposit, 328 stainless, Type 304, 427 stainless, Type 304L, 125, 588, 644 stainless, Type 316, 413, 469 stainless, Type 316L, 125, 588, 631, 644 Strain rate, 343 Strain, transient, 557 Stress, 401 dependence, linear, 557 Structural integrity, 79 Superlattice, 218, 588 Surface activated joining method, 42 Surveillance, 28, 79 Master Integrated Reactor Vessel Surveillance Program, 291 program, 97 specimen, 42 Swelling, 401, 413, 427, 457 influences on, 775 void, 443, 487 T Temperature cycle effects, 736 Temperature gradient, 443 Temperature influence, silicon carbide, 775 Temper embrittlement, 356 Tensil"e properties, 328, 469, 500, 644 Tensile samples, 660 Tensile specimens, 523 Tensile testing, 644, 660 Tension tests, 469, 660 Tersoff potential, 799 Thermal aging, 151, 189, 328, 356, 674 Thermal anneal heat treatment, 151 Thermal annealing, 218, 247, 356 Thermal condition, end-of-life, 328 Thermal creep, 557 Threshold displacement energy, 799 TIG welding, 535 Titanium chromium-nickel-titanium, 343 titanium-modified stainless steel, 443 titanium-nickel alloys, 825 Toughness transition predictions, 97 Transformation, 825 Transition behavior, 97 Transition joints, welded, 573 Transition range, 55 Transition temperature, 109, 151, 302 ductile-to-brittle, 189, 315 shifts, 79 Transmission electron microscopy, 189, 218, 694 cross sectional, 775 in situ observation, 836 martensitic steel, 674 microstructure examinations, 722 phosphorus intergranular segregation, 189 silicon carbide fiber reinforced composites, 786 stainless steel, 413 Copyright by ASTM Int'l (all rights reserved); Sun Dec 20 18:18:21 EST 2015 Downloaded/printed by University of Washington (University of Washington) pursuant to License Agreement No further reproductions authorized INDEX 879 titanium-nickel alloys, 825 vanadium alloys, 736 TRIM code, 722 Tungsten, 546, 762 source rods, 612 target, 644, 660 Tyranno SA, 786 U Ultimate strength, 660 Upper shelf energy, 302 W Water quality, effect on corrosion rate, 631 Water radiolysis, 631 Weld, 28, 151, 356, 535 copper, 302 inertia welded transition joints, 573 Linde 80, 291 metal, 302 metal, submerged arc, 8, 262, 291, 356 Y V Vacancy clusters, 204 Vanadium, 736, 746 Void, 413, 523 formation, 443 swelling, 401, 487 Volume change, 704 VVER, 109, 383 Yield, 97 strength, 343, 660, 694 stress, 328, 644 Z Zerilli-Armstrong constitutive model, 97 Zirconium phosphide, 846 Copyright by ASTM Int'l (all rights reserved); Sun Dec 20 18:18:21 EST 2015 Downloaded/printed by University of Washington (University of Washington) pursuant to License Agreement No further reproductions authorized