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American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors

召开年：2019
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1.3D Characterization of Corrosion Fatigue Cracks and Crack Tips Generated in Type 304/304L Stainless Steel during Enhanced and Retarded Fatigue Crack Growth

机译：304 / 304L型不锈钢在增强和延迟疲劳裂纹扩展过程中产生的腐蚀疲劳裂纹和裂纹尖端的3D表征
- 作者：B.S. Anglin;B.D. Miller;K.B. Fisher;T.W. Webb
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
2.INVESTIGATING Pb-INDUCED STRESS CORROSION CRACKING OF ALLOY 800 VIA IN-SITU XPS

机译：通过原位XPS研究Pb诱导的合金800的应力腐蚀开裂
- 作者：B. Payne;S.Y. Persaud;J.M. Smith
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
3.In-situ monitoring of the corrosion behavior and impurity enrichment of steam generator tube materials under PWR secondary side thermal hydraulic and heat transfer conditions

机译：在压水堆二次侧热工水力和传热条件下对蒸汽发生管材料的腐蚀行为和杂质富集进行现场监测
- 作者：Caitlin Huotilainen;Essi Jäppinen;Tiina Ikäläinen;Seppo Peltonen;Timo Saario;Konsta Sipilä
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
4.EFFECT OF WELDING DILUTION AND DENDRITE ORIENTATION ON SCC BEHAVIOR OF ALLOY 52M WELDING METAL IN SIMULATED PWR PRIMARY WATER

机译：模拟压水堆原始水中焊接稀释和枝晶取向对52M合金焊接金属应力腐蚀开裂行为的影响
- 作者：Jiarong Ma;Kun Zhang;Tongming Cui;Zhanpeng Lu;Xue Liang;Wenqing Liu;Zhimin Zhong;Guangdong Han
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
5.ASSESSMENT OF THE SCC MITIGATION CAPABILITIES OF THE NOBLE METAL CHEMICAL APPLICATION TECHNOLOGY IN A SIMULATED BWR ENVIRONMENT

机译：模拟压水堆环境中贵金属化学应用技术的SCC缓解能力评估
- 作者：Pascal V. Grundler;Stefan Ritter;Sriharitha Rowthu;Hans-Peter Seifert
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
6.RADIATION-INDUCED FERRITE FORMATION AS A POTENTIAL ISSUE IN PWR AUSTENITIC INTERNALS FOLLOWING PLANT LIFE EXTENSION

机译：植物寿命延长后，压敏奥氏体内部辐射诱发的铁素体形成是可能的问题
- 作者：D. A. Merezhko;M. S. Merezhko;O. P. Maksimkin;M. N. Gussev;F. A. Garner
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
7.SCC INITIATION AND GROWTH IN PWR PRIMARY WATER CONTAINING KOH VS. LIOH

机译：含KOH的压水式原水的SCC起始和生长OH
- 作者：Peter Andresen;Peter Chou
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
8.INSIGHTS INTO PROSPECTIVE FUSION REACTOR COOLING SYSTEMS FROM FISSION REACTOR COOLING CIRCUITS

机译：裂变反应堆冷却电路对预期的熔融反应堆冷却系统的认识
- 作者：T.L. Martin;A.D. Warren;D. Kumar;A. Siberry;R. Springell;R. Holmes;R. Clark;L. Platts;R. Burrows;C. Harrington;M. Gorley;E. Surrey;S. Rowthu;P. Grundler;S. Ritter
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
9.Stress Corrosion Cracking and Fracture Toughness Tests of Irradiated Type 304 Stainless Steel

机译：辐照型304不锈钢的应力腐蚀开裂和断裂韧性测试
- 作者：Y. Chen;B. Alexandreanu;K. Natesan;A. S. Rao
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
10.MORPHOLOGY OF CAVITIES FORMED ON GRAIN BOUNDARY OF CRACKED ALLOY TT690

机译：TT690裂纹合金晶界上的型腔形态
- 作者：Takumi Terachi;Makie Okamoto;Takuyo Yamada;Koji Arioka
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
11.MITIGATION OF PRIMARY WATER STRESS CORROSION CRACKING INITIATION OF ALLOY 600

机译：合金600的初始水应力腐蚀开裂的缓解
- 作者：Sonya Pemberton;Jack Beswick;Mark Chatterton;Jill Meadows;Stuart Medway
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
12.DEVELOPMENT OF RESIDUAL STRESS DRIVEN STRESS CORROSION CRACK INITIATION AND GROWTH SPECIMENS

机译：残余应力驱动应力腐蚀裂纹萌生和生长试样的发展
- 作者：Tyler Moss;John Brockenbrough
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
13.Microstructural Characterization of Proton Irradiated 304L SS at 100°C and 360°C

机译：质子在100°C和360°C辐射304L SS的显微结构表征
- 作者：C.D. Judge;B. Langelier;M.A. Mattucci;Q. Wang;M. Daymond;G.S. Was;J. Smith
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
14.DISTINCT EFFECTS OF IRRADIATION ON THE STRUCTURE AND CHEMICAL REACTIVITY OF SILICATES AND CARBONATES

机译：辐照对硅酸盐和碳酸盐的结构和化学反应性的不同影响
- 作者：Erika Callagon La Plante;Yi-Hsuan Hsiao;Isabella Pignatelli;Aditya Kumar;N. M. Anoop Krishnan;Tandré Oey;Howard Dobbs;Yingtian Yu;Bu Wang;Narayanan Neithalath;Kevin G. Field;Jacob Israelachvili;Yann Le Pape;Mathieu Bauchy;Gaurav Sant
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
15.The influence of metal substrate on CRUD build-up under simulated PWR conditions

机译：在模拟压水堆条件下，金属基材对CRUD堆积的影响
- 作者：Stefano Cassineri;Jonathan Duff;Michele Curioni;Andrew Banks;Fabio Scenini
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
16.Evaluating the Effects of Hydride Embrittlement and Pulse Width on Zircaloy Cladding Transient Performance

机译：评估氢化物脆化和脉冲宽度对锆合金熔覆瞬态性能的影响
- 作者：D. Kamerman;D. Swank;D. Haggard;A. Matthews;L. Emerson;D. Wachs
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
17.INVESTIGATING THE ROLE OF HYDROGEN ON STRESS CORROSION CRACKING BY MICROMECHNAICAL TESTING

机译：用显微技术研究氢在应力腐蚀开裂中的作用
- 作者：Edward Roberts;Edmund Tarleton;Florence Carrette;Sergio Lozano-Perez
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
18.Impact of the startup temperature and primary chemistry program on the oxidation and release of corrosion products from alloy 690 steam generator

机译：启动温度和主要化学程序对690合金蒸汽发生器腐蚀产物的氧化和释放的影响
- 作者：Flambard Julie;Carrette Florence;Monchy-Leroy Carole;Martin Bachet;Hadrien Perron;Andrieu Eric;Laffont Lydia
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
19.ENVIRONMENTAL DEGRADATION RESISTANCE OF ATF FeCrAl CLADDING TUBE SPECIMENS DURING THE FUEL CYCLE

机译：燃料循环过程中ATF FeCrAl包覆试管的耐环境降解性
- 作者：Michael Schuster;Evan J. Dolley;Timothy B. Jurewicz;Raul B. Rebak
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
20.Understanding the mechanisms of surface and crack tip oxidation with valence mapping

机译：用价态图了解表面和裂纹尖端氧化的机理
- 作者：Sergio Lozano-Perez;Zhao Shen;Donghai Du;Lefu Zhang
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
21.CHARACTERIZING AND QUANTIFYING THE AGING OF POLYETHYLENE THIN FILMS USING NOVEL DOPED-FILMS AND GOLD NANOPARTICLE LABELING STRATEGIES TOWARD UNDERSTANDING FAILURE OF CABLING INSULATION

机译：使用新颖的掺杂薄膜和金纳米颗粒标签策略表征和量化聚乙烯薄膜的老化，以了解电缆绝缘的失败
- 作者：Daniel Zoltek;Tana O’Keefe;Faith Murphy;Magdalene B. Jones;Robert C. Duckworth;Brian Hinderliter;Melissa A. Maurer-Jones
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
22.Microstructural characterization of reactor cavity concrete

机译：反应堆空腔混凝土的微观结构表征
- 作者：J. David Arregui-Mena;Elena Tajuelo Rodriguez;Alain B. Giorla;Christa E. Torrence;Philip D. Edmondson;Yann Le Pape;Thomas M. Rosseel
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
23.In Situ SEM Microtensile Testing of Oxidized Grain Boundaries in Alloy 600

机译：合金600中氧化晶粒边界的原位SEM微拉伸测试
- 作者：E.K. Still;H. Vo;K. Lam;J. Kabel;R. Auguste;P. Chou;P. Hosemann
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
24.STRESS CORROSION CRACK INITIATION OF COLD-WORKED ALLOY 800 IN Pb-CONTAINING, ALKALINE ENVIRONMENTS

机译：含铅碱性环境中冷作合金800的应力腐蚀裂纹萌生
- 作者：Jaganathan Ulaganathan;Jared Smith
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
25.Non-Destructively Detecting LWR Structural Material Embrittlement using Transient Grating Spectroscopy

机译：瞬态光栅光谱技术无损检测轻水堆结构材料的脆性
- 作者：Saleem A. Al Dajani;Benjamin R. Dacus;Cody A. Dennett;M. Grace Burke;Kudzanai Mukahiwa;Kuba Anglin;James J. Wall;Thak Sang Byun;Michael P. Short
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
26.REVEALING HOW ALKALI CATIONS AFFECT THE ELECTRICAL DOUBLE LAYER STRUCTURE OF STAINLESS STEEL IN ALKALINE AQUEOUS ENVIRONMENTS

机译：揭示碱性阳离子如何影响碱性水溶液中不锈钢的双电层结构
- 作者：Rachel Guia P. Giron;Xin Chen;Erika Callagon La Plante;Maxim N. Gussev;Keith J. Leonard;Gaurav N.Sant
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》 | 2019年
27.Small Scale Tensile Testing of Grain Boundary Strength of X-750 Alloy

机译：X-750合金晶界强度的小规模拉伸试验
- 作者：Lingfeng He;Daniel Murray;Xiang Liu;Wen Jiang;Mukesh Bachhav;Xianming Bai;Sebastien Teysseyre
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》 | 2019年
28.Effect of UV irradiation on Corrosion Behavior of Zirconium Alloy in High Temperature Water Conditions

机译：高温水条件下紫外线辐射对锆合金腐蚀行为的影响
- 作者：Taeho Kim;Yalong He;Zefeng Yu;Mohamed Elbakshwan;Li He;Adrien Couet
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》 | 2019年
29.LOW TEMPERATURE CRACK PROPAGATION MECHANISM OF ALLOY 182 WELD

机译：182合金的低温裂纹扩展机理。
- 作者：Young Suk Kim;Sung Soo Kim
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》 | 2019年
30.BIAXIAL CREEP OF TEXTURED Zr-Nb ALLOYS

机译：Zr-Nb织构合金的双轴蠕变
- 作者：Pratik Joshi;Micah Tillman;Nilesh Kumar;Mahmut Cinbiz;Korukonda Murty
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》 | 2019年
31.Effect of isothermal aging on Ni-Cr-based commercial alloys

机译：等温时效对Ni-Cr基商用合金的影响
- 作者：Nicholas Aerne;Fei Teng;Julie D. Tucker
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》 | 2019年
32.NONDESTRUCTIVE EVALUATION OF CONCRETE SPECIMENS REPRESENTATIVE OF NUCLEAR POWER PLANTS CONTAINING KNOWN DEFECTS

机译：包含已知缺陷的核电站混凝土标本的非破坏性评估
- 作者：N. Dianne Bull Ezell;Austin Albright;Dan Floyd;Lev Khazanovich
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》 | 2019年
33.Oxide Dispersion Type 304L Stainless Steel by Additive Manufacturing Process

机译：通过增材制造工艺制造的氧化物弥散型304L不锈钢
- 作者：Takahiro Ishizaki;Kinya Aota;Yingjuan Yang;Yusaku Maruno;Atuhiko Onuma
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》 | 2019年
34.Overview of Electric Power Research Institute’s Research on Irradiation of Concrete

机译：电力研究所对混凝土辐射的研究概述
- 作者：Emma L. Wong;James J. Wall
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》 | 2019年
35.THE EFFECT OF IRRADIATION-INDUCED MATRIX BARRIER HARDENING ON IRRADIATION-ASSISTED STRESS CORROSION CRACKING (IASCC) IN AUSTENITIC STEELS AND NICKEL BASED ALLOYS

机译：辐照诱发的基质壁垒硬化对奥氏体钢和镍基合金中辐照辅助应力腐蚀开裂（IASCC）的影响
- 作者：R.G. Faulkner
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》 | 2019年
36.STEAM OXIDATION, BURST AND CRITICAL HEAT FLUX TESTING OF COMMERCIAL FECRAL CLADDING

机译：商业粪便熔覆的蒸汽氧化，爆裂和临界热通量测试
- 作者：B. A. Pint;L. A. Baldesberger;K. A. Kane
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》 | 2019年
37.EFFECT OF GAMMA IRRADIATION ON CREEP PROPERTIES OF CEMENT PASTE ANALOGUES

机译：γ射线辐照对水泥浆模拟物蠕变性能的影响
- 作者：Elena Tajuelo Rodriguez;Yann Le Pape;Thomas M. Rosseel;William A. Hunnicutt
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》 | 2019年
38.IRRADIATION-INDUCED CRACKING OF DUAL-PURPOSE COATINGS ON SiC

机译：辐照引起的SiC双面涂层的开裂
- 作者：Stephen S. Raiman;Peter J. Doyle;Caen Ang;Takaaki Koyanagi;David Carpenter;Kurt A. Terrani;Yutai Katoh
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》 | 2019年
39.WEIBULL AND BOOTSTRAP BASED PROBABILISTIC FATIGUE LIFE MODELING OF STAINLESS STEEL UNDER PWR COOLANT WATER ENVIRONMENT CONDITION

机译：PWR冷却水环境下基于Weibull和Bootstrap的不锈钢疲劳概率寿命模型
- 作者：Jae Phil Park;Subhasish Mohanty;Chi Bum Bahn
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》 | 2019年
40.HIGH TEMPERATURE CYCLIC DEFORMATION BEHAVIOR OF AN ADVANCED AUSTENITIC STAINLESS STEEL (ALLOY 709)

机译：先进的奥氏体不锈钢（合金709）的高温循环变形行为
- 作者：Zeinab Y. Alsmadi;Abdullah S. Alomari;N. Kumar;K.L. Murty
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》 | 2019年
41.IRRADIATION-ASSISTED STRESS CORROSION CRACKING OF TI-STABILIZED AUSTENITIC STAINLESS STEEL

机译：TI稳定奥氏体不锈钢的辐照辅助应力腐蚀开裂
- 作者：Miroslava Ernestova
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》 | 2019年
42.MICROSTRUCTURAL CHARACTERIZATION OF STEAM GENERATOR TUBE FLAKES FROM AN OPERATING PWR AND ITS GALVANIC CORROSION BEHAVIOR

机译：压水堆运行时蒸汽发生器管片的微观结构表征及其电偶腐蚀行为
- 作者：Soon-Hyeok Jeon;Geun Dong Song;Do Haeng Hur
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》 | 2019年
43.Effects of gamma ray irradiation on corrosion of carbon steel at water line

机译：γ射线辐照对水线碳钢腐蚀的影响
- 作者：Tomomichi Ariga;Hiroshi Abe;Yutaka Watanabe
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》 | 2019年
44.Stress Corrosion Cracking of an Austenitic Stainless Steel Pipe Weld

机译：奥氏体不锈钢管焊缝的应力腐蚀开裂
- 作者：Michael R. Ickes
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》 | 2019年
45.Concurrent effects of irradiation and hydrogenation on microstructural and mechanical properties of Zr based fuel cladding alloys

机译：辐射和加氢对Zr基燃料熔覆合金显微组织和力学性能的同时影响
- 作者：Sho Kano;Huilong Yang;John McGrady;Toru Higuchi;Yoshinori Etoh;Hiroaki Abe
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》 | 2019年
46.Environmentally Assisted Cracking and Fracture Toughness of an Irradiated Stainless Steel Weld

机译：辐照不锈钢焊缝的环境辅助开裂和断裂韧性
- 作者：Y. Chen;B. Alexandreanu;C. Xu;Y. Yang;K. Natesan;A. S. Rao
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》 | 2019年
47.EFFECTS OF DRY, WET-DRY CYCLING AND AQUEOUS IMMERSION ON POLYPROPYLENE AND POLYETHYLENE CABLE INSULATION MECHANICAL PROPERTIES

机译：干，湿循环和浸入对聚丙烯和聚乙烯电缆绝缘机械性能的影响
- 作者：Isaiah Salinas;Adam Finke;Brian Hinderliter;Keith B. Lodge;Melissa A. Maurer-Jones;Robert Duckworth
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》 | 2019年
48.3D CREEP RESPONSE OF NUCLEAR CONCRETE

机译：核混凝土的3D蠕变响应
- 作者：Aishwarya Baranikumar;Christa E. Torrence;Zachary Grasley
- 会议名称：《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》 | 2019年