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International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors

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1.ROLE OF CAVITY FORMATION ON CRACK GROWTH OF COLD-WORKED CARBON STEEL, TT 690 AND MA 600 IN HIGH TEMPERATURE WATER

机译：腔形成对高温水中冷加工碳钢，TT 690和MA 600裂纹的作用
- 作者：Koji Arioka;Tomoki Miyamoto;Takuyo Yamada
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Cavity formation;
- IGSCC;
- Creep;
- TT690;
- MA600;
- Carbon steel;
2.INITIATION OF PWSCC OF WELD ALLOY 182

机译：焊接合金PWSCC的启动182
- 作者：Thierry Couvant;Francis Vaillant
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Alloy 182;
- stress corrosion cracking;
- initiation;
3.HYDRIDE BEHAVIOR IN ZIRCALOY-4 DURING THERMOMECHANICAL CYCLING

机译：热机械循环期间锆石氢化物行为
- 作者：Kimberly Colas;Arthur Motta;Mark R. Daymond
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Zircaloy;
- hydrides;
- thermomechanical cycling;
4.RECENT INSIGHTS ON THE PARAMETRIC DEPENDENCE OF IRRADIATION CREEP OF AUSTENITIC STAINLESS STEELS

机译：奥氏体不锈钢照射蠕变参数依赖性的最新见解
- 作者：F. A. Garner;E. R. Gilbert;V. S. Neustroev
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- austenitic steel;
- neutron irradiation;
- irradiation creep;
- void swelling;
- stress state;
5.SCC PROPERTIES OF MODIFIED ALLOY 718 IN BWR PLANT

机译：BWR植物改性合金718的SCC性能
- 作者：Yoshinori Katayama;Motoji Tsubota;Yoshiaki Saito
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Alloy 718;
- Alloy X-750;
- stress corrosion cracking;
- jet pump beam;
- uni-axial constant load test;
- fatigue test;
- spring properties;
6.GRAIN BOUNDARY OXIDATION AND STRESS CORROSION CRACKING IN NICKEL-BASE ALLOYS STRAINED IN SUPERCRITICAL WATER

机译：超临界水中应变镍基合金中的晶界氧化与应力腐蚀裂纹
- 作者：Tyler Moss;Matthew J. Olszta;William Grant
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Nickel Alloys;
- Stress Corrosion Cracking;
- Intergranular Oxidation;
- Supercritical Water;
7.DEFORMATION MICROSTRUCTURES OF 30 dpa AISI 304 STAINLESS STEEL AFTER MONOTONIC TENSILE AND CONSTANT LOAD AUTOCLAVE TESTING

机译：单调拉伸恒定载荷高压釜检测后30dPA AISI 304不锈钢的变形微观结构
- 作者：Wade Karlsen;Janne Pakarinen;Aki Toivonen
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- stainless steel;
- irradiation assisted stress corrosion cracking;
- TEM;
8.Characterization of Type 304L Stainless Steel: Comparison of ASTM A262 Practice A and Analytical Electron Microscopy Techniques

机译：304L不锈钢型特征：ASTM A262实践A和分析电子显微镜技术的比较
- 作者：B.D. Miller;M.G. Burke
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
9.Effect of Thermal Aging on SCC, Material Properties and Fracture Toughness of Stainless Steel Weld Metals

机译：热老化对不锈钢焊接金属的SCC，材料性能和断裂韧性的影响
- 作者：T. Lucas;R. G. Ballinger;H. Hanninen
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- thermal aging;
- stainless steel;
- weld;
- stress corrosion cracking;
- fracture toughness;
- environmental effect;
- spinodal decomposition;
10.SCC BEHAVIOR OF ALLOY 152 WELD IN A PWR ENVIRONMENT

机译：合金152焊缝在PWR环境中的SCC行为
- 作者：Bogdan Alexandreanu;Yiren Chen;Ken Natesan
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Stress Corrosion Cracking;
- Alloy 152 weld;
11.PWR FUEL DEPOSIT ANALYSIS AT A BW PLANT WITH A 24 MONTH FUEL CYCLE

机译：P＆W工厂的PWR燃料沉积分析，24个月燃料循环
- 作者：Mike G. Pop;Larry S. Lamanna;Richard Harne
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- PWR Fuel Deposits;
- PWR Deposit Modeling;
- PWR Deposit Analysis;
12.SCC OF HIGH CR ALLOYS IN BWR ENVIRONMENTS

机译：BWR环境中高Cr合金的SCC
- 作者：Peter L. Andresen
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Stress corrosion cracking;
- BWR;
- high chromium weld metals;
13.X-RAY PHOTOELECTRON STUDY OF THE OXIDES FORMED ON NICKEL METAL AND NICKEL-CHROMIUM 20 ALLOY SURFACES UNDER REDUCING AND OXIDIZING POTENTIALS IN BASIC, NEUTRAL AND ACIDIC SOLUTIONS

机译：在碱性，中性和酸性溶液中，在镍金属和镍 - 铬20％合金表面上形成的氧化物的X射线光电子的研究
- 作者：Brad P. Payne;Peter G. Keech;N. Stewart Mclntyre
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- XPS;
- Nickel corrosion;
- Nickel-chromium alloy corrosion;
14.HIGH RESOLUTION ELECTRON MICROSCOPY STUDY ON OXIDEFILMS FORMED ON NICKEL-BASE ALLOYS X-750,182 AND82 IN SIMULATED HIGH FLOW VELOCITY BWR WATERCONDITIONS

机译：镍基合金X-750,182和82在模拟高流量速度BWR滑水条件上形成的高分辨率电子显微镜研究
- 作者：Jiaxin Chen;Fredrik Lindberg;Lyubov Belova
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- nickel-base;
- alloys;
- Inconel;
- corrosion;
- microstructure;
- TEM;
- FIB;
- oxide films;
- BWR;
- Alloy X-750;
- Alloy 182;
- Alloy 82;
15.SCC CRACK GROWTH RATE OF ALLOY 82 IN PWR PRIMARY WATER CONDITIONS - EFFECT OF A THERMAL TREATMENT -

机译：PWR初级水条件合金82的SCC裂纹生长速率 - 热处理的效果 -
- 作者：C. Guerre;C. Duhamel;M. Sennour
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Stress corrosion cracking;
- Alloy 82;
- PWR primary water;
16.OXIDE FILM CHARACTERIZATION ALONG CRACK PATHS IN STAINLESS STEEL IN AERATED AND DEAERATED WATER ENVIRONMENTS

机译：沿着空气脱气水环境中不锈钢裂纹路径氧化膜特征
- 作者：Elaine West;David Morton;Nathan Lewis
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Stainless Steel;
- SCC;
- Oxidation;
- Crack Tip;
17.QUANTITATIVE MICRO-NANO (QMN) APPROACH TO SCC MECHANISM AND PREDICTION-STARTING A THIRD MEETING

机译：SCC机制的定量微纳米（QMN）方法和预测启动第三次会议
- 作者：Roger W.Staehle
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Stress corrosion cracking;
- prediction;
- mechanism;
- Fe-Cr-Ni alloys;
- nuclear power;
- Quantitative Micro-Nano;
18.Evaluation of the Susceptibility to SCC Initiation of Alloy 690 in Simulated PWR Primary Water

机译：对模拟PWR初级水中合金690的SCC启动易感性的评价
- 作者：Kazuya TSUTSUMI;Thierry COUVANT
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Alloy690;
- stress corrosion cracking;
- initiation;
- ageing;
- cold rolling;
19.3D ATOM-PROBE CHARACTERIZATION OF STRESS AND COLD-WORK IN STRESS CORROSION CRACKING OF 304 STAINLESS STEEL

机译：3D原子探测应力和冷加工应力腐蚀裂纹304不锈钢的表征
- 作者：K Kruska;S Lozano-Perez;D W Saxey
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Atom-probe;
- TEM;
- Stress Corrosion Cracking;
- Austenic Stainless Steel;
- PWR;
- Cold-Work;
20.WEAR OF ZIRCALOY-4 GRID STRAPS DUE TO FRETTING AND PERIODIC IMPACTING WITH RV INTERNALS BAFFLE PLATES

机译：Zircaloy-4网格带的磨损由于RV内部挡板板的烦恼和周期性影响
- 作者：Sarah Davidsaver;Stephen Fyfitch;Brian Friend
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Baffle Plates;
- Fuel/Baffle Interaction;
- Grid Strap Degradation;
21.STRESS CORROSION CRACK GROWTH OF ALLOY 52M IN SIMULATED PWR PRIMARY WATER

机译：模拟PWR初级水中合金52M的应力腐蚀裂纹生长
- 作者：M.B. Toloczko;M.J. Olszta;S.M. Bruemmer
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- alloy 52M;
- PWSCC;
- crack growth rate;
22.EFFECT OF DISSOLVED HYDROGEN, SURFACE CONDITIONS AND COMPOSITION ON THE ELECTRONIC PROPERTIES OF THE OXIDE FILMS FORMED ON NICKEL-BASE ALLOYS IN PWR PRIMARY WATER

机译：溶解氢，表面条件和组成对PWR初级水镍基合金形成的氧化膜的电子性质
- 作者：A. Loucif;J.-P. Petit;Y. Wouters
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- nickel-base alloys;
- primary water;
- oxide;
- hydrogen;
- electronic properties;
- photoelectrochemistry;
23.STRESS CORROSION CRACK GROWTH RATE TESTING OF NOVEL COMPOSITE ARREST SPECIMENS

机译：新型复合逮捕标本的应力腐蚀裂纹增长速率试验
- 作者：David S. Morton;John V. Mullen;Eric Plesko
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
24.EFFECT OF CHLORIDE ON ENVIRONMENTALLY ASSISTED CRACKING OF LOW ALLOY STEELS IN OXYGENATED HIGH- TEMPERATURE WATER - GENERAL CORROSION

机译：氯化物对氧化高温水中低合金钢环保抗裂的影响
- 作者：Matthias Herbst;Armin Roth;Erika Nowak
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Low-Alloy Steel;
- Chloride;
- Oxide Layer;
- EAC;
- general corrosion;
- crack initiation;
25.CLUSTER DYNAMICS PREDICTION OF THE MICROSTRUCTURE EVOLUTION OF 300-SERIES AUSTENITIC STAINLESS STEEL UNDER IRRADIATION: INFLUENCE OF HELIUM

机译：辐照下300系奥氏体不锈钢微观结构演化的簇动力学预测：氦气的影响
- 作者：M. Zouari;L. Founder;A. Barbu
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Helium;
- Cavity;
- Austenitic stainless steel;
- Irradiation;
26.NRC RESEARCH ACTIVITIES ON ENVIRONMENTALLY-ASSISTED FATIGUE

机译：NRC关于环境辅助疲劳的研究活动
- 作者：Gary L. Stevens;Robert L. Tregoning
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- environmental fatigue;
- environmentally-assisted fatigue;
- fatigue;
27.USE OF THE AREVA BWR CRUD MODEL TO STUDY HIGH ZINC OPERATION AT A US PLANT

机译：使用AREVA BWR CRUD模型在美国植物中研究高锌操作
- 作者：Mike G. Pop;Larry Lamanna;Merl Bell
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- BWR Fuel Crud;
- Zinc Injection;
- BWR Crud Modeling;
28.EFFECT OF DH CONCENTRATION ON CRUD DEPOSITION ON HEATED ZIRCALOY-4 IN SIMULATED PWR PRIMARY WATER

机译：DH浓度对模拟PWL初级水中加热锆型-4的CRUD沉积的影响
- 作者：Hirotaka Kawamura
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Corrosion product;
- Crud;
- DH concentration;
- Zircaloy-4;
- PWR primary coolant;
29.ELECTROCHEMICAL STUDIES OF STEAM GENERATOR TUBE DEGRADATION IN THE PRESENCE OF THIOSULPHATE

机译：硫代硫酸盐存在下蒸汽发生器管劣化的电化学研究
- 作者：Lisheng Chi;Yucheng Lu
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- steam generator tubing degradation;
- thiosulphate;
- electrochemical method;
30.STRESS CORROSION CRACKING OF ALLOY 600 IN PWR PRIMARY WATER : INFLUENCE OF CHROMIUM, HYDROGEN AND OXYGEN DIFFUSION

机译：PWR初级水中合金600的应力腐蚀开裂：铬，氢气扩散的影响
- 作者：C. Guerre;P. Laghoutaris;J. Chene
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Stress corrosion cracking;
- Alloy 600;
- PWR primary water;
31.OXIDATION OF A PROTON-IRRADIATED 316 STAINLESS STEEL IN SIMULATED BWR NWC ENVIRONMENT

机译：模拟BWR NWC环境中质子辐照的316不锈钢的氧化
- 作者：Zhijie Jiao;Gary Was
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Corrosion;
- oxidation rate;
- stainless steel;
- BWR;
32.THE EFFECT OF TEMPERATURE ON THE CRACK GROWTH RATE OF STAINLESS STEEL AND NI-ALLOYS IN SIMULATED BWR ENVIRONMENT

机译：温度对模拟BWR环境中不锈钢和Ni合金裂纹生长速率的影响
- 作者：Johan Stjarnsater;Anders Jenssen;Christer Jansson
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Stress corrosion cracking;
- temperature dependence;
- crack growth rate;
- stainless steel;
- Alloy 182;
33.Effect of Environment and Prestrain on IASCC of Austenitic Stainless Steels

机译：环境对奥氏体不锈钢IASCC的影响
- 作者：W. Lai;Z. Jiao;G. S. Was
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- IASCC;
- Austenitic stainless steel;
34.STRESS CORROSION CRACK INITIATION SUSCEPTIBILITY OF IRRADIATED AUSTENITIC STAINLESS STEELS

机译：辐照奥氏体不锈钢的应力腐蚀裂纹启动易感性
- 作者：Kale J Stephenson;Yugo Ashida;Jeremy T Busby
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- IASCC;
- neutron irradiation;
- stainless steel;
- CERT;
- BWR NWC;
35.MICROSTRUCTURE AND SCC OF ALLOY X-750

机译：合金X-750的微观结构和SCC
- 作者：Peter L. Andresen;Juan Flores-Preciado;Martin M. Morra
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- stress corrosion cracking;
- BWR;
- high temperature water;
- Alloy X-750;
- microstructure;
- water chemistry;
36.INFLUENCE OF LOCALIZED PLASTICITY ON IASCC SENSITIVITY OF AUSTENITIC STAINLESS STEELS UNDER PWR PRIMARY WATER.

机译：PWR初级水下局部塑性对奥氏体不锈钢敏感性敏感性的影响。
- 作者：Sarata Cisse;Benoit Tanguy;Lydia Laffont
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
37.NRC/EPRI Welding Residual Stress Validation Program - Phase III

机译：NRC / EPRI焊接残余应力验证程序 - 阶段III
- 作者：M Kerr;LF Fredette;HJ Rathbun
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Welding;
- Residual Stress;
- Stress Corrosion Cracking;
38.INFLUENCE OF PRIMARY WATER CHEMISTRY ON OXIDES FORMED ON ALLOY 600 AND ALLOY 690

机译：初级水化学对合金600和合金690形成的氧化物的影响
- 作者：Thomas M. Devine
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Intergranular stress corrosion cracking;
- surface films;
- Alloy 600;
- Alloy 690;
39.CONTAINMENT LINER CORROSION

机译：遏制衬垫腐蚀
- 作者：Darrell Dunn;April Pulvirenti;Paul Klein
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- containment;
- localized corrosion;
- concrete;
- steel liner;
40.EVALUATION OF THE OXYGEN DIFFUSION COEFFICIENT IN NICKEL-BASE ALLOYS

机译：镍基合金氧扩散系数的评价
- 作者：Hyo On Nam;Jae Young Yoon;Ji Hyun Kim
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- PWSCC;
- oxygen diffusivity;
- nickel-base alloy;
41.Deformation Mode and Microstructure on Stress Corrosion Cracking Path and Kinetics in High Temperature Water Environments

机译：高温水环境中应力腐蚀开裂路径和动力学的变形模式和微观结构
- 作者：Zhanpeng Lu;Tetsuo Shoji;Seiya Yamazaki
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Stress corrosion cracking;
- high temperature water;
- deformation mode;
- microstructure;
- grain boundary;
42.MODELLING MATERIAL EFFECTS IN FLOW-ACCELERATED CORROSION

机译：流动加速腐蚀中的材料效应
- 作者：P. Phromwong;Derek Lister;S. Uchida
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Flow-accelerated Corrosion;
- Carbon Steel;
- Chromium Content.;
43.BEHAVIOR OF STRESS CORROSION CRACKING FOR TYPE 316LSTAINLESS STEEL WITH CONTROLLED DISTRIBUTION OF SURFACE WORK HARDENED LAYER IN SIMULATED BOILING WATER REACTORS ENVIRONMENT

机译：316L不锈钢应力腐蚀裂缝的行为，采用模拟沸水反应器环境下表面工作硬化层的控制分布
- 作者：Yasufumi Miura;Yuichi Miyahara;Masaru Sato
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Boiling water reactor;
- Type 316L stainless steel;
- Stress corrosion cracking;
44.Crack growth testing on Cold Worked Alloy 690 in Primary Water Environment

机译：初级水环境中冷加工合金690裂纹增长试验
- 作者：David R. Tice;Stuart L. Medway;Norman Platts
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Alloy 690;
- Cold Work;
- IGSCC;
- Crack Growth;
45.CYCLIC AND SCC BEHAVIOR OF ALLOY 690 HAZ IN A PWR ENVIRONMENT

机译：PWR环境中合金690 HAZ的循环和SCC行为
- 作者：Bogdan Alexandreanu;Yiren Chen;Ken Natesan
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Stress Corrosion Cracking;
- Alloy 690 HAZ;
46.INFLUENCE OF CHLORIDE IONS AS CONTAMINANTS ON THE CORROSION BEHAVIOR OF ALLOY 718 IN POOL WATER OF NUCLEAR POWER PLANTS

机译：氯离子作为污染物对核电站池水中合金718腐蚀行为的影响
- 作者：Jonathan Hugues;Eric Andrieu;Christine Blanc
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Corrosion;
- Alloy 718;
- Boric Acid;
- Chloride Ions;
47.Development of a method for studying the influence of stress state on the iodine-induced stress corrosion cracking of zirconium alloys

机译：一种研究一种研究应力状态对锆合金碘造成应力腐蚀裂纹的影响
- 作者：Nathanael Mozzani;Quentin Auzoux;David Le Boulch
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- "zirconium alloys;
- stress corrosion cracking;
- iodine;
- viscoplastic;
- anisotropic;
- model";
48.CURRENT NRC PERSPECTIVES CONCERNING PRIMARY WATER STRESS CORROSION CRACKING

机译：关于初级水分腐蚀裂纹的当前NRC透视
- 作者：David Alley;Darrell Dunn
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Nuclear Reactors;
- Nickel Alloys;
- Primary Water Stress Corrosion Cracking;
49.IN-PILE TESTS FOR IASCC GROWTH BEHAVIOR OF IRRADIATED 316L STAINLESS STEEL UNDER SIMULATED BWR CONDITION IN JMTR

机译：在JMTR中模拟BWR条件下辐照316L不锈钢的IASCC生长行为的拟桩试验
- 作者：Yasuhiro Chimi;Shigeki Kasahara;Hideo Ise
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- IASCC;
- In-pile Test;
- Crack Growth;
- Irradiated Stainless Steel;
- BWR;
- JMTR;
- ECP;
50.DEGRADATION OF GRAIN BOUNDARY STRENGTH BY OXIDATION IN ALLOY 600

机译：用合金600氧化氧化晶界强度的降解
- 作者：Katsuhiko Fujii;Terumitsu Miura;Hiromasa Nishioka
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
51.IRRADIATION CREEP AND IRRADIATION STRESS RELAXATION OF 316 AND 304L STAINLESS STEELS IN THERMAL AND FAST NEUTRON SPECTRUM REACTORS

机译：在热和快节中子谱反应器中照射蠕变和照射应力松弛316和304L不锈钢
- 作者：John Paul Foster;Torill M. Karlsen
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
52.STRESS CORROSION CRACKING BEHAVIOR OF DISSIMILAR METAL WELDMENTS IN HIGH TEMPERATURE WATER ENVIRONMENTS

机译：高温水环境中不同金属焊接的应力腐蚀开裂行为
- 作者：J. Y. Huang;M. F. Chiang;R. C. Kuo
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Corrosion fatigue;
- SCC;
- residual stress;
- size effect;
- dissimilar metal weldment;
53.EFFECTS OF TEMPERATURE AND CORROSION POTENTIAL ON SCC

机译：温度和腐蚀电位对SCC的影响
- 作者：Peter L. Andresen;Russell A. Seeman
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Stress corrosion cracking;
- BWR;
- catalysis;
- mitigation;
- structural material;
- start up;
54.Overview of NRC Proactive Management of Materials Degradation (PMMD) Program

机译：材料退化（PMMD）计划的NRC主动管理概述
- 作者：C. E. (Gene) Carpenter Jr.;Amy Hull;Greg Oberson
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
55.THE EFFECT OF GRAIN SIZE ON IGSCC IN SS 316L IN SIMULATED BWR ENVIRONMENT

机译：模拟BWR环境中SS 316L中IGSCC晶粒大小对IGSCC的影响
- 作者：Johan Stjarnsater;Bengt Bengtsson;Bjorn Forssgren
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Stress corrosion cracking;
- grain size;
- crack growth rate;
- stainless steel;
56.STRAIN PATH EFFECT ON IGSCC INITIATION AND OXIDATION OF ALLOY 182 EXPOSED TO PWR PRIMARY WATER

机译：对PWW初级水的IGSCC 182氧化对IGSCC引发和氧化的应变路径效应
- 作者：Thierry Couvant;Laurent Legras;Thierry Ghys
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Alloy 182;
- stress corrosion cracking;
- initiation;
- oxidation;
- strain path;
57.Irradiation assisted stress corrosion cracking of stainless steels in a PWR environment (A combined approach)

机译：PWR环境中不锈钢辅助应力腐蚀开裂（一种组合方法）
- 作者：Morgane Le Millier;Olivier Calonne;Jerome Crepin
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- IASCC;
- 304L stainless steel;
- PWR;
- deformation fields;
58.Slow Strain Rate Tensile Tests of Irradiated Austenitic Stainless Steels in Simulated PWR Environment

机译：模拟PWR环境中辐照奥氏体不锈钢的慢应变速率拉伸试验
- 作者：Y. Chen;B. Alexandreanu;W. K. Soppet
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- IASCC;
- stainless steels;
- PWR;
- slow strain rate tensile;
59.Characterizing environmental degradation in PWRs by 3D FIB sequential sectioning

机译：3D FIB顺序切片在PWRS中表征环境劣化
- 作者：Sergio Lozano-Perez;Na Ni;Karen Kruska
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- FIB;
- 3D;
- Oxidation;
- Stainless Steel;
- Zirconium;
60.STRUCTURE AND THERMODYNAMICAL PROPERTIES OF ZIRCONIUM HYDRIDES FROM FIRST-PRINCIPLE

机译：锆氢化锆的结构和热力学性质从第一原理
- 作者：Jakob Blomqvist;Johan Olofsson;Anna-Maria Alvarez
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Hydride induced embrittlement;
- Zirconium hydrides;
- DFT;
61.ON THE MICROSTRUCTURE OF ALLOY 600 SCC CRACKS OBSERVED BY TEM ON PWR SG PULLED TUBES AND ON LABORATORY SPECIMENS

机译：在PWR SG拉管和实验室标本下TEM观察到的合金600 SCC裂缝的微观结构
- 作者：L. Legras;O. de Bouvier;F. Delabrouille
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Alloy 600;
- SCC;
- Lead;
- Secondary Side Corrosion;
62.FLOW ACCELERATED CORROSION OF CARBON STEEL IN THE FEED WATER SYSTEM OF PWR PLANTS - BEHAVIOUR OF WELDS AND WELD ASSEMBLIES

机译：PWR植物饲料水系统中碳钢的流动加速腐蚀 - 焊接和焊接组件的行为
- 作者：C. Mansour;E.M. Pavageau;A. Faucher
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Flow accelerated corrosion;
- Welds;
- Weld assemblies;
- Chromium content;
63.Electron Microscopy Characterizations and Atom Probe Tomography of Intergranular Attack in Alloy 600 Exposed to PWR Primary Water

机译：合金600中晶体攻击的电子显微镜特征和原子探测层析成像暴露于PWR初级水
- 作者：Matthew J. Olszta;Daniel K. Schreiber;Larry E. Thomas
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- alloy 600;
- intergranular stress corrosion cracking;
- corrosion;
- intergranular attack;
- grain boundary;
- internal oxidation;
- internal sulfidation;
64.THE EFFECT OF COLD WORK ON MICROSTRUCTURE AND SCC SUSCEPTIBILITY IN SIMULATED BWR ENVIRONMENT FOR NON-SENSITIZED AUSTENITIC STAINLESS STEELS

机译：非敏化奥氏体不锈钢模拟BWR环境的微观结构和SCC敏感性的效果
- 作者：Yohei Sakakibara;Guen Nakayama
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Austenitic stainless steel;
- Cold work;
- CSL boundary;
- EBSD;
- plastic strain;
- CBB test;
- grain boundary;
65.EFFECT OF HOT CRACKS ON EAC CRACK INITIATION AND GROWTH IN NICKEL-BASE ALLOY WELD METALS

机译：热裂纹对镍基合金焊接金属EAC裂纹启动和生长的影响
- 作者：Hannu Hanninen;Aki Toivonen;Anssi Brederholm
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Nickel-base alloys;
- weld metal;
- hot cracking;
- environment-assisted cracking;
- electron microscopy;
- Laves phase;
66.Grain boundary oxidation and embrittlement prior to crack initiation in Alloy 600 in PWR primary water

机译：PWR初级水中合金600裂纹开采前晶界氧化与脆化
- 作者：L. Fournier;O. Calonne;P. Combrade
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Nickel Base Alloys;
- PWSCC;
- Initiation;
- Oxidation;
- Grain Boundary;
67.INTERACTION OF MICROSTRUCTURE, COMPOSITION, AND COLD WORK ON THE STRESS CORROSION CRACKING OF ALLOY 82 WELD METAL

机译：微观结构，组成和冷加工对合金82焊缝金属应力腐蚀开裂的相互作用
- 作者：D.J. Paraventi;W.C. Moshier
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Alloy 82;
- Alloy 600;
- SCC;
- Cold work;
- Welds;
68.A PRELIMINARY HYBRID MODEL OF IRRADIATION-ASSISTED STRESS CORROSION CRACKING OF 300 SERIES STAINLESS STEELS IN PWR PRIMARY ENVIRONMENTS

机译：PWR初级环境中300系列不锈钢辐射辅助应力腐蚀裂纹的初步混合模型
- 作者：E. D. Eason;G. Ilevbare;R. Pathania
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- IASCC;
- PWR;
- primary water;
- hybrid;
- model;
- stainless steel;
- strain rate;
69.STRESS CORROSION CRACKING AND CRACK TIP CHARACTERIZATION OF ALLOY X-750 IN BOILING WATER REACTOR ENVIRONMENTS

机译：沸水反应器环境中合金X-750的应力腐蚀裂纹和裂纹尖端
- 作者：J.P. Gibbs;R.G. Ballinger;J.H. Jackson
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
70.AREVA FUEL CONDITION INDEX FOR A PRESSURIZED WATER REACTOR

机译：加压水反应堆的紫花燃料条件指标
- 作者：Mike G. Pop;Merl Bell;Brian Lockamon
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Fuel Performance;
- PWR Fuel Deposits;
- PWR Deposit Modeling;
71.BALANCE OF PLANT CORROSION ISSUES IN AGING NUCLEAR POWER PLANTS

机译：老化核电站植物腐蚀问题的平衡
- 作者：George Licina;Dilip Dedhia
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Balance of plant;
- corrosion;
- nuclear plants;
- service water systems;
- general corrosion;
- pitting;
- MIC;
72.ENVIRONMENTALLY ASSISTED CRACK GROWTH IN COLD WORKED ALLOY 690TT IN PRIMARY WATER AT LOW AND HIGHTEMPERATURES

机译：低于初级水的初级水中冷加工合金690TT的环保裂纹增长
- 作者：Qunjia Peng;Tetsuo Shoji;Juan Hou
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Environmentally assisted cracking;
- Alloy 690TT;
- Primary Water;
- Dissolved hydrogen;
- Cold working;
73.PREDICTING CHROMIUM DEPLETION OF NICKEL BASE ALLOYS

机译：预测镍基合金的铬耗尽
- 作者：Youfa Yin;Feng Zhu;Roy Faulkner
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Modelling;
- Nickel base alloys;
- Chromium depletion;
74.INFLUENCE OF BULK AND SURFACE COLD WORK ON CRACKINITIATION AND CRACK GROWTH OF AUSTENITIC STAINLESS STEELS UNDER SIMULATED BWR ENVIRONMENT

机译：散装与表面冷工作对模拟BWR环境下奥氏体不锈钢裂缝和裂纹生长的影响
- 作者：Bastian Devrient;Renate Kilian;Karin Kuster
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Crack Initiation;
- Crack Growth;
- Austenitic Stainless Steel;
- BWR;
75.ONE DIMENSIONAL COLD ROLLING EFFECTS ON STRESS CORROSION CRACK GROWTH IN ALLOY 690 TUBING AND PLATE MATERIALS

机译：一维冷轧效果对合金690管和板材的应力腐蚀裂纹增长
- 作者：Mychailo B. Toloczko;Matthew J. Olszta;Stephen M. Bruemmer
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- alloy 690;
- PWSCC;
- CRDM;
- plate;
- cold rolling;
- crack growth rate;
76.Crack Growth Rates of Irradiated Commercial Stainless Steels in BWR and PWR Environments

机译：BWR和PWR环境中辐照商业不锈钢裂纹增长率
- 作者：Anders Jenssen;Johan Stjarnsater;Raj Pathania
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- IASCC;
- Crack growth rates;
- BWR;
- PWR;
77.EFFECT OF NITROGEN ADDITION IN 304 L STAINLESS STEEL ON THE IGSCC CRACK GROWTH RATE IN SIMULATED BWR ENVIRONMENT

机译：304L不锈钢中氮施加对模拟BWR环境IGSCC裂纹生长速率的影响
- 作者：S. Roychowdhury;V. Kain;R.C. Prasad
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Stainless Steel;
- Non-sensitised and strain hardened;
- IGSCC;
- Crack Growth Rate;
- BWR Environment;
78.Penetrative Internal Oxidation from Alloy 690 Surfaces and Stress Corrosion Crack Walls during Exposure to PWR Primary Water

机译：在暴露于PWR初级水中的合金690表面和应力腐蚀裂缝壁的渗透内氧化
- 作者：Matthew J. Olszta;Daniel K. Schreiber;Larry E. Thomas
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Alloy 690;
- penetrative oxidation;
- SCC;
79.NON-LINEAR DYNAMICS OF THE MORPHOLOGY AT THE OXIDE /METAL INTERFACE OF AUSTENITIC STEELS IN SIMULATED LIGHTWATER REACTOR ENVIRONMENTS AND ITS IMPLICATIONS FORSCC INITIATION

机译：模拟闪水反应器环境中奥氏体钢氧化物/金属界面形态的非线性动力学及其Forscc发起的影响
- 作者：Yoichi Takeda;Takayuki Sato;Daisuke Yamauchi
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Stress corrosion cracking;
- Surface oxidation;
- Crack initiation;
80.High-Resolution Characterizations of Grain Boundary Damage and Stress Corrosion Cracks in Cold-Rolled Alloy 690

机译：冷轧合金690中晶界损伤和应力腐蚀裂纹的高分辨率特征
- 作者：S. M. Bruemmer;M. J. Olszta;M. B. Toloczko
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- alloy 690;
- grain boundaries;
- cold work;
- stress corrosion cracking;
- PWR primary water;
81.In-situ and Ex-situ Oxide Characterization by Synchrotron X-ray (SPring-8) in Non-sensitized 316 Stainless Steel and High Temperature Water Combination

机译：通过同步X射线（Spring-8）在非敏化316不锈钢和高温水合组合中的原位和前型氧化物表征
- 作者：Toshio Yonezawa;Masashi Watanabe;Takahisa Shobu
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
82.Conditions for Long Term Operation of Nuclear Power Plants in Sweden

机译：瑞典核电站长期运作的条件
- 作者：Peter Ekstrom;Karen Gott;Bjorn Brickstad
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
83.Stress Corrosion Cracking Behavior of Type 304 Stainless Steel Irradiated under Different Neutron Dose Rates at JMTR

机译：在JMTR下不同中子剂量率照射304型不锈钢的应力腐蚀开裂行为
- 作者：Yoshiyuki Kaji;Keietsu Kondo;Yoshitera Aoyagi
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Stress corrosion cracking;
- Type 304 stainless steel;
- Neutron dose rate;
- JMTR;
84.ROLE OF SLIP BEHAVIOR IN THE IRRADIATION ASSISTED STRESS CORROSION CRACKING IN AUSTENTTIC STEELS

机译：防滑行为在奥塞泰钢辐射辅助应力腐蚀开裂中的作用
- 作者：M.D. McMurtrey;G.S. Was
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- IASCC;
- stainless steel;
- localized deformation;
- shear stress;
85.SCC OF ALLOY 690 AND ITS WELD METALS

机译：合金690的SCC及其焊接金属
- 作者：Peter L. Andresen;Martin M. Morra;Kawaljit Ahluwalia
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Stress corrosion cracking;
- crack growth;
- Alloy 690;
- Alloy 52/152 weld metals;
- cold work;
- high temperature water;
- microstructure;
86.FATIGUE LIMIT AND HYSTERESIS BEHAVIOR OF TYPE 304L STAINLESS STEEL IN AIR AND PWR WATER, AT 150°C AND 300°C

机译：空气和PWR水中304L不锈钢型疲劳极限和滞后行为，150°C和300°C
- 作者：H. D. Solomon;C. Amzallag;A.J. Vallee
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- fatigue limit;
- 304L SS;
- PWR water;
- staircase method;
- cycle frequency;
87.THE KEY FACTORS AFFECTING CRACK GROWTH BEHAVIOR OF NEUTRON-IRRADIATED AUSTENITIC ALLOYS

机译：影响中子辐照奥氏体合金裂纹生长行为的关键因素
- 作者：Yugo Ashida;Alexander Flick;Peter L. Andresen
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- LASCC;
- Neutron Irradiated;
- Stainless Steel;
- Crack Growth;
88.EFFECTS OF MATERIAL COMPOSITION ON CORROSION FATIGUE CRACK GROWTH OF AUSTENITIC STAINLESS STEELS IN HIGH TEMPERATUREWATER

机译：材料组成对高温水下奥氏体不锈钢腐蚀疲劳裂纹生长的影响
- 作者：Norman Platts;David Tice;Kevin Mottershead
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Stainless Steel;
- Fatigue;
- Crack Growth;
- Sulfur;
89.Effect of Static Load Hold Periods on the Corrosion Fatigue Behavior of Austenitic Stainless Steels in Simulated BWR Environments

机译：静载保持周期对模拟BWR环境中奥氏体不锈钢腐蚀行为的影响
- 作者：H.P. Seifert;S. Ritter;H. Leber
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Corrosion fatigue;
- environmental assisted fatigue;
- austenitic stainless steel;
- light water reactor;
- static load hold time;
90.IRRADIATION ASSISTED STRESS CORROSION CRACKING OF AUSTENITIC STAINLESS STEEL WWER REACTOR CORE INTERNALS

机译：奥氏体不锈钢WWER反应器核心内部辐照辅助应力腐蚀开裂
- 作者：Anna Hojna;Miroslava Ernestova;Ossi Hietanen
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Ti-stabilized austenitic stainless steel;
- neutron irradiation;
- SSRT;
- Crack Growth Rate;
- IASCC;
- WWER;
91.Electrochemical study of pre- and post-transition corrosion of Zr alloys in PWR coolant

机译：PWR冷却剂ZR合金前后腐蚀的电化学研究
- 作者：Jan Macak;Radek Novotny;Petr Sajdl
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Zirconium alloys;
- Corrosion;
- Oxide layer;
- Impedance spectroscopy;
92.Crack Growth Behavior of Irradiated Type 316 SS in Low Dissolved Oxygen Environment

机译：低溶解氧环境下照射型316 SS的裂纹生长行为
- 作者：Y. Chen;B. Alexandreanu;Y. Yang
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- IASCC;
- Stainless steels;
- BWR;
- HWC;
- Crack growth rate;
93.RESEARCH AND EVALUATION OF LOW TEMPERATURE CRACK PROPAGATION OF Ni BASE ALLOYS IN ACTUAL PLANTS

机译：实际植物中Ni碱基合金低温裂纹繁殖的研究与评价
- 作者：Kimihisa Sakima;Harutaka Suzuki;Hideki Fujiwara
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- LTCP;
- Hydrogen;
- Ni Base Alloy;
- Inconel;
- Weld Metal;
- Fracture Toughness;
94.CYCLIC AND SCC BEHAVIOR OF ALLOY 52M/182 WELD OVERLAY IN A PWR ENVIRONMENT

机译：PWW环境中合金52m / 182焊接覆盖层的循环和SCC行为
- 作者：Bogdan Alexandreanu;Yiren Chen;Ken Natesan
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Stress Corrosion Cracking;
- Alloy 52M;
- Weld overlay;
95.STRESS CORROSION CRACKING BEHAVIOR NEAR THE FUSION BOUNDARY OF DISSIMILAR WELD JOINT WITH ALLOY 182-A533B LOW ALLOY STEEL

机译：用合金182-A533B低合金钢的不同焊接接头融合边界附近的应力腐蚀开裂行为
- 作者：Hiroshi Abe;Makoto Ishizawa;Yutaka Watanabe
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Stress Corrosion Cracking;
- Low Alloy Steel;
- Dissimilar Weld;
- Heat Affected Zone;
96.AN INVESTIGATION INTO STRESS CORROSION CRACKING OF DISSIMILAR METAL WELDS WITH 304L STAINLESS STEEL AND ALLOY 82 IN HIGH TEMPERATURE PURE WATER

机译：用304L不锈钢和合金82在高温纯水中对不同金属焊缝应力腐蚀裂纹的研究
- 作者：Tsung-Kuang Yeh;Guan-Ru Huang;Chuen-Homg Tsai
- 会议名称：《International Conference on Environmental degradation of Materials in Nuclear Power Systems-Water Reactors》 | 2012年
- Dissimilar Metal Weld;
- Stress Corrosion Cracking;
- Stainless Steel;
- SSRT;