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American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors
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.
INTERFACE CHARACTERIZATION OF CANDIDATE DUAL-PURPOSE BARRIER COATINGS FOR SIC/SIC ACCIDENT TOLERANT FUEL CLADDING
机译:
SIC / SIC耐高温燃料包壳的双键双壁障壁涂层的界面表征
作者:
Joey Kabel
;
Takaaki Koyanagi
;
Yutai Katoh
;
Ryan Schoell
;
Djamel Kaoumi
;
Caen Ang
;
Peter Hosemanna
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
2.
INFLUENCE OF MEAN STRESS AND PWR ENVIRONMENT ON FATIGUE BEHAVIOR OF A 304L SS
机译:
平均应力和压水堆环境对304L不锈钢疲劳性能的影响
作者:
Ziling Peng
;
Gilbert Hénaff
;
Jean-Christophe Le Roux
;
Romain Verlet
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
3.
Influence of Long-Term Thermal Aging on SCC Initiation Susceptibility of L-grade Austenitic Stainless Steel
机译:
长期热时效对L级奥氏体不锈钢SCC萌发敏感性的影响
作者:
K. Kondo
;
S. Aoki
;
Y. Fujimura
;
T. Hirade
;
Y. Kaji
;
S. Yamashita
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
4.
IASCC SUSCEPTIBILITY OF HIGHLY IRRADIATED 316 STAINLESS STEEL IN SIMULATED PWR PRIMARY WATER
机译:
模拟压水堆原水中高辐照度316不锈钢的IASCC敏感性
作者:
Donghai Du
;
Gary S. Was
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
5.
USE OF ON-LINE MONITORING TECHNIQUES FOR EVALUATIONS OF LEAD STRESS CORROSION CRACKING (PBSCC) AND A PBSCC INHIBITOR
机译:
在线监测技术在评估铅应力腐蚀开裂(PBSCC)和PBSCC抑制剂中的应用
作者:
Brent Capell
;
Jared Smith
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
6.
Effect of thermal aging on microstructure and hardness of industrial heats of Alloy 690
机译:
热时效对690合金工业加热组织和硬度的影响
作者:
Caitlin Huotilainen
;
Ulla Ehrnstén
;
Matias Ahonen
;
Hannu Hänninen
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
7.
CHRACTERISTICS OF WORK HARDENED SURFACE LAYER ON AUSTENITIC STAINLESS STEELS AND ITS RELATION TO SCC SUSCEPTIBILITY IN HIGH TEMPERATURE WATER
机译:
奥氏体不锈钢上工作硬化表面层的特征及其与高温水中SCC敏感性的关系
作者:
Hiroshi Abe
;
Yutaka Watanabe
;
Takamichi Miyazaki
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
8.
FRACTURE PROPERTY DEGRADATION IN CAST STAINLESS STEELS AFTER LONG-TERM THERMAL AGING
机译:
长期热老化后铸不锈钢的断裂性能退化
作者:
T.S. Byun
;
T.G. Lach
;
D.A. Collins
;
C. Jang
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
9.
UNDERSTANDING THE ROLE OF GRAIN BOUNDARY MIGRATION ON THE INITIATION STAGES OF PWSCC
机译:
理解谷物边界迁移在PWSCC起始阶段的作用
作者:
L. Volpe
;
M.G. Burke
;
F. Scenini
会议名称:
《》
10.
Combined approach to Medium Voltage Cable and Accessories Aging Management Technique at Nuclear Power Plants
机译:
核电站中压电缆及附件老化管理技术的组合方法
作者:
Raihan Khondker
;
Sarajit Banerjee
;
David Rouison
;
Rick Easterling
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
11.
CABLE AGING AND CM APPROACHES BEYOND THE SHORTCOMINGS OF IAEA NUCLEAR ENERGY SERIES NO. NP-T-3.6
机译:
电缆老化和CM方法在国际原子能机构(IAEA)核能No. NP-T-3.6
作者:
Kenneth T. Gillen
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
12.
EFFECTS OF AGING TEMPERATURE AND ALLOY COMPOSITION ON LONG-RANGE ORDERING IN NICR- FE ALLOYS DURING ISOTHERMAL AGING
机译:
时效温度和合金成分对等温时效后镍铁合金长距离有序化的影响
作者:
Xiangkun Ru
;
Zhanpeng Lu
;
Jiarong Ma
;
Chengdong Yang
;
Qian Yuan
;
Weibao Tang
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
13.
EFFECT OF STRAIN AND STRAIN RATE ON CRACK INITIATION OF 316L STEEL IN THE SIMULATED PWR WATER
机译:
应变和应变速率对模拟压水堆中316L钢开裂的影响
作者:
Anna Hojná
;
Mariia Zimina
;
Lukáš Horák
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
14.
CHARACTERIZATION OF IN-SERVICE THERMAL AGING EFFECTS IN BASE METALS AND WELDS OF THE PRESSURE VESSEL OF A DECOMMISSIONED PWR PRESSURIZER, AFTER 27 YEARS OF OPERATION
机译:
在运行27年后,对退役的PWR加压器的贱金属和压力容器的焊接中的在役热时效进行表征
作者:
Pierre JOLY
;
Lingtao SUN
;
Pål EFSING
;
Jean PaulMASSOUD
;
Frédéric SOMVILLE
;
Robert GERARD
;
Ying Hui AN
;
Jonathan BAILEY
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
15.
Stress Corrosion Cracking of A286 Reactor Coolant Pump Turning Vane Bolts
机译:
A286反应堆冷却剂泵转向叶片螺栓的应力腐蚀开裂
作者:
Michael R. Ickes
;
Andrew M. Ruminski
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
16.
EFFECT OF LOCAL STRAIN AND GND DENSITY ON CRACK INITIATION IN ALLOY 600
机译:
局部应变和GND密度对600合金裂纹萌生的影响
作者:
Naganand Saravanan
;
Phani S Karamched
;
Morgane Le Faucheur
;
Emilien Burger
;
Fabio Scenini
;
Sergio Lozano-Perez
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
17.
OPERATIONAL EXPERIENCE OF SIEMENS/KWU LWR NUCLEAR POWER PLANTS
机译:
西门子/ KWU轻水堆核电站的运行经验
作者:
Renate Kilian
;
Armin Roth
;
Martin Widera
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
18.
CORROSION ISSUES IN FUTURE YEARS: A TSO PERSPECTIVE
机译:
未来几年的腐蚀问题:TSO的观点
作者:
Ian de CURIERES
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
19.
DESIGN AND FABRICATION OF ALLOY A-286 FOR RESISTANCE TO IGSCC IN PWRS
机译:
PWRS抗IGSCC合金A-286的设计与制造
作者:
Stephen Fyfitch
;
Sarah Davidsaver
;
Kyle Amberge
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
20.
EMPIRICAL EQUATIONS FOR TENSILE PROPERTIES AND STRESS-STRAIN CURVES OF NEUTRON IRRADIATED STAINLESS STEELS IN LWR CONDITIONS
机译:
LWR条件下中子辐照不锈钢的拉伸性能和应力-应变曲线的经验方程
作者:
Koji Fukuya
;
Katsuhiko Fujii
;
Yasuhiro Chimi
;
Kuniki Hata
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
21.
Effect of Residual δ-Ferrite on SCC Behavior of 321 Stainless Steel
机译:
残余δ铁素体对321不锈钢SCC行为的影响
作者:
Jiamei Wang
;
Kai Chen
;
Haozhan Su
;
Donghai Du
;
Xianglong Guo
;
Lefu Zhang
;
Zhao Shen
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
22.
PHYSICALLY-BASED MODELING OF ENVIRONMENTAL FATIGUE CRACK GROWTH IN TYPE 304/304L STAINLESS STEEL
机译:
304 / 304L型不锈钢环境疲劳裂纹扩展的基于物理的建模
作者:
B.S. Anglin
;
J.R. Brockenbrough
;
J.A. Savchik
;
C.B. Geller
;
T.W. Webb
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
23.
EFFECT OF FOLLOW-UP POST WELD HEAT TREATMENT ON MICROSTRUCTURE AND PWSCC OF ALLOY 52M WELD METAL IN DISSIMILAR METAL WELD JOINT
机译:
后续焊后热处理对异种金属焊接接头中52M合金的组织和PWSCC的影响
作者:
Jiarong Ma
;
Kun Zhang
;
Tongming Cui
;
Qi Xiong
;
Zhanpeng Lu
;
Junjie Chen
;
Chengdong Yang
;
Maolong Zhang
;
Weibao Tang
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
24.
CHARACTERIZATION OF MICROSTRUCTURE OF ALLOY 52M WELD METAL NEAR THE FUSION BOUNDARY AND OXIDE FILMS FORMED IN SIMULATED PWR PRIMARY WATER
机译:
模拟压水堆原水中熔合边界和氧化物膜附近的52M合金焊接金属的微观组织特征
作者:
Kun Zhang
;
Jiarong Ma
;
Tonging Cui
;
Zhanpeng Lu
;
Fei Ning
;
Yibo Jia
;
Xue Liang
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
25.
ENVIRONMENTALLY-ASSISTED SHORT CRACK FATIGUE TESTING ON AUSTENITIC STAINLESS STEELS
机译:
在奥氏体不锈钢上进行环境辅助的短裂纹疲劳测试
作者:
B. E. Coult
;
A.S Griffiths
;
J. P. Beswick
;
P. J. Gill
;
N. Platts
;
J. M. Smith
;
G. L. Stevens
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
26.
IASCC INITIATION TEST OF NEUTRON IRRADIATED TYPE 316L STAINLESS STEEL IN SIMULATED BWR CONDITION TO EVALUATE THRESHOLD STRESS
机译:
模拟中压水冷条件下中子辐照型316L不锈钢的IASCC初始测试
作者:
Kazuhiro Chatani
;
Shigeaki Tanaka
;
Hitoshi Seto
;
Hiroyuki Nakano
;
Takayuki Kaminaga
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
27.
NANO-MECHANICAL TESTING OF PROTON IRRADIATED 304L SS AT 100°C AND 360°C TO SUPPORT IASCC
机译:
质子辐照的304L SS在100°C和360°C下对IASCC的纳米力学测试
作者:
M.A. Mattucci
;
Q. Wang
;
T. Skippon
;
M.R. Daymond
;
G.S. Was
;
J. Smith
;
C.D. Judge
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
28.
EMPIRICAL EQUATIONS OF CRACK GROWTH RATES BASED ON DATA FITTING OF NEUTRON IRRADIATED STAINLESS STEEL UNDER HIGH TEMPERATURE WATER SIMULATING BOILING WATER REACTOR CORE CONDITIONS
机译:
高温水模拟沸水反应堆堆芯条件下基于中子辐照不锈钢裂纹扩展速率的经验方程
作者:
Shigeki Kasahara
;
Yasuhiro Chimi
;
Kuniki Hata
;
Koji Fukuya
;
Katsuhiko Fujii
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
29.
THE MECHANISTIC UNDERSTANDING ON THE STRESS CORROSION CRACKING GROWTH RATE IN SIMULATED PWR PRIMARY WATER FOR COLD WORKED TT ALLOY 690
机译:
TT 690合金在模拟PWR原水中应力腐蚀开裂速率的机理理解。
作者:
Toshio Yonezawa
;
Masashi Watanabe
;
Atsushi Hashimoto
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
30.
TRIBOCORROSION PHENOMENA IN CO- AND FE-BASED HARDFACING ALLOYS EVALUATED USING PIN-IN-ON-DISC WEAR TESTS IN A SIMULATED PWR ENVIRONMENT
机译:
在模拟压水堆环境中使用针入式圆盘磨损测试评估的基于Co和FE的硬质合金中的三腐蚀现象
作者:
V L Ratia
;
M J Carrington
;
D Zhang
;
J L Daure
;
D G McCartney
;
P H Shipway
;
D A Stewart
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
31.
EVALUATION OF THE CORROSION KINETICS OF SIC WITH AND WITHOUT MITIGATION COATINGS IN LWR CHEMISTRIES
机译:
在轻水堆化学中有无涂层的SIC腐蚀动力学评估
作者:
Peter Doyle
;
Stephen Raiman
;
Caen Ang
;
Yutai Katoh
;
Steven Zinkle
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
32.
The effect of diffusion-induced grain boundary migration ahead of SCC crack tips on propagation
机译:
SCC裂纹尖端前扩散诱导的晶界迁移对扩展的影响
作者:
Zhao Shen
;
Koji Arioka
;
Sergio Lozano-Perez
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
33.
SIMULATING THE SUSCEPTIBILITY TO IGSCC OF COLD WORK 316 AUSTENITIC STAINLESS STEEL EXPOSED TO PRIMARY WATER
机译:
模拟暴露于一次水的冷作316奥氏体不锈钢对IGSCC的敏感性
作者:
Thierry COUVANT
;
Emilien BURGER
;
Claire THAURY
;
Claire RAINASSE
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
34.
CORROSION PERFORMANCE OF SUPERAUSTENTIC STAINLESS STEEL AL-6XN IN HIGH TEMPERATURE WATER
机译:
奥氏体不锈钢AL-6XN在高温水中的腐蚀性能
作者:
M.J. Stiger
;
B.A. Webler
;
J.K. Heuer
;
R.E. Hermer
;
W.C. Moshier
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
35.
UNDERSTANDING THE EFFECT OF STRAIN LOCALIZATION ON CORROSION FATIGUE OF TYPE 304 AUSTENITIC STAINLESS STEELS IN HIGH TEMPERATURE WATER
机译:
理解应变局部化对304型奥氏体不锈钢在高温水中的腐蚀疲劳的影响
作者:
Hanxiao Wang
;
Fabio Scenini
;
João Quinta da Fonseca
;
M. Grace Burke
;
Jill Meadows
;
Norman Platts
;
David Tice
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
36.
INFLUENCE OF HEAT TRANSFER CREVICE CONDITIONS ON STRESS CORROSION CRACKING IN LEAD FAULTED ALKALINE ENVIRONMENTS
机译:
传热条件对含铅断裂碱性环境中应力腐蚀开裂的影响
作者:
Frederick D. Miller
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
37.
MICROSTRUCTURE OF NI AND X-750 IRRADIATED AT LOW AND HIGH TEMPERATURES
机译:
NI和X-750在低温和高温下的显微组织
作者:
W. Li
;
C. Judge
;
L. Walters
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
38.
Examination of Cavity Formation in Cold-Worked Alloy 690
机译:
690冷加工合金中空洞形成的检验
作者:
Matthew Olszta
;
Ziqing Zhai
;
Mychailo Toloczko
;
Stephen Bruemmer
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
39.
INFLUENCE OF THE COMBINATION OF MICROSTRUCTURE AND MECHANICAL FIELDS ON STRESS CORROSION CRACKING INITIATION OF COLD-WORKED AUSTENITIC STAINLESS STEELS
机译:
显微组织和力学场的结合对冷加工奥氏体不锈钢应力腐蚀开裂的影响
作者:
Qi Huang
;
Yann Charles
;
Cécilie Duhamel
;
Monique Gaspérini
;
Jérôme Crépin
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
40.
EFFECT OF MATERIAL CONDITION ON STRESS CORROSION CRACK INITIATION OF COLD-WORKED ALLOY 600 IN SIMULATED PWR PRIMARY WATER
机译:
材料条件对模拟压水堆原始水中冷作合金600应力腐蚀开裂的影响
作者:
Ziqing Zhai
;
Mychailo B. Toloczko
;
Stephen M. Bruemmer
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
41.
Development of a gaseous iodine testing system to determine the I-SCC properties of zirconium alloys
机译:
开发用于确定锆合金I-SCC特性的气态碘测试系统
作者:
Sean M. Hanlon
;
Andrew Phillion
;
Conor Gillen
;
Mark R. Daymond
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
42.
CRACK INITIATION BEHAVIOR OF COLD-WORKED ALLOY 690 IN SIMULATED PWR PRIMARY WATER – ROLE OF STARTING MICROSTRUCTURE, APPLIED STRESS AND COLD WORK ON PRECURSOR DAMAGE EVOLUTION
机译:
模拟压水堆原水中冷作合金690的裂纹萌生行为-微观组织的启动,施加应力和冷作对前驱损伤演化的作用
作者:
Ziqing Zhai
;
Matthew Olszta
;
Mychailo Toloczko
;
Stephen Bruemmer
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
43.
UNDERSTANDING THE EVOLUTION OF DEFORMATION STRUCTURES AT FATIGUE CRACK TIPS AND IMPLICATIONS FOR ENVIRONMENTALLY ENHANCED AND RETARDED FATIGUE CRACK GROWTH
机译:
理解疲劳裂纹尖端变形结构的演变以及对环境的增强和延缓疲劳裂纹增长的意义
作者:
B. D. Miller
;
D. J. Paraventi
;
B. S. Anglin
;
T. W. Webb
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
44.
SCC Behavior at Alloy 52M-182 Weld Overlay Interfaces in a PWR Environment
机译:
PWR环境中52M-182合金堆焊界面处的SCC行为
作者:
Bogdan Alexandreanu
;
Yiren Chen
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
45.
CORROSION BEHAVIOUR OF PHYSICAL VAPOUR DEPOSITED COATINGS ON NUCLEAR COMPONENTS AND Fe-Cr-Al ALLOY UNDER NORMAL OPERATION AND ACCIDENT SCENARIOS
机译:
正常操作和事故情况下物理气相沉积涂层对核成分和Fe-Cr-Al合金的腐蚀行为
作者:
Caitlin Dever
;
Mitchell Mattucci
;
Kevin Daub
;
Suraj Persaud
;
Raul B. Rebak
;
Brian Langelier
;
Heidi Nordin
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
46.
MODELING THE EVOLUTION OF INTERGRANULAR HELIUM BUBBLES IN NICKEL USING THE INCLUDED PHASE MODEL
机译:
包含相模型模拟镍中粒状氦气泡的演化
作者:
Andrew A. Prudil
;
Michael J. Welland
;
Colin D. Judge
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
47.
COUPLING EFFECT OF CHARGED-HYDROGEN AND COLD WORK ON OXIDATION BEHAVIOR OF 316L STAINLESS STEEL IN DEAERATED HIGH TEMPERATURE WATER
机译:
带电的氢和冷作对脱氮高温水中316L不锈钢的氧化行为的影响
作者:
Tongming Cui
;
Jiarong Ma
;
Fei Ning
;
Zhanpeng Lu
;
Kun Zhang
;
Yibo Jia
;
Tetsuo Shoji
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
48.
CHARACTERISTICS OF OXIDE FILMS FORMED ON 309L AND 308L STAINLESS STEELS IN SIMULATED PWR PRIMARY WATER
机译:
模拟压水堆原水中309L和308L不锈钢制成的氧化膜的特征
作者:
Qi Xiong
;
Jiarong Ma
;
Kun Zhang
;
Tongming Cui
;
Zhanpeng Lu
;
Junjie Chen
;
Yibo Jia
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
49.
SCC OF ALLOY 82 WELD METAL
机译:
SCC合金82焊接金属
作者:
Peter L. Andresen
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
50.
EFFECT OF SPECIMEN SIZE AND PERCENT ENGAGEMENT ON SCC BEHAVIOR OF ALLOY 82 IN BWR ENVIRONMENTS
机译:
标本大小和参与度对BWR环境中合金82的SCC行为的影响
作者:
Katsuhiko Kumagai
;
Hiroyuki Nakano
;
Takayuki Kaminaga
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
51.
IRRADIATION ASSISTED STRESS CORROSION CRACKING (IASCC) OF LOW STRENGTH AND HIGH STRENGTH ALLOYS IN LIGHTWATER REACTOR ENVIRONMENTS
机译:
低强度和高强度合金在照明反应器环境中的辐照应力腐蚀开裂(IASCC)
作者:
M. Wang
;
M. Song
;
L. Nelson
;
R. Pathania
;
G. S. Was
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
52.
SPECIMEN SIZE EFFECTS ON THE CRACK GROWTH RATE RESPONSE OF HIGHLY IRRADIATED TYPE 304 STAINLESS STEEL
机译:
标本尺寸对高辐照度304不锈钢的裂纹扩展速率响应的影响
作者:
A. Jenssen
;
J. Stjärnsäter
;
C. Topbasi
;
P. Chou
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
53.
NEXT GENERATION OF NOBLE METAL APPLICATION FOR BOILING WATER REACTORS
机译:
沸腾反应器的下一代贵金属应用
作者:
Joe Giannelli
;
Erica Libra-Sharkey
;
Collin Custer
;
George Inch
;
Andrew Odell
;
Michelle Mura
;
Mike Ford
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
54.
REVIEW OF RADIATION INDUCED DEGRADATION OF CONCRETE STRUCTURE IN COMMERCIAL NUCLEAR POWER PLANTS AROUND REACTOR VESSEL
机译:
反应堆容器中商用核电站辐射结构退化的研究进展
作者:
Bruce Biwer
;
David Ma
;
Yunping Xi
;
Yuxiang Jing
;
Madhumita Sircar
;
Jinsuo Nie
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
55.
USING A MULTI-SCALE APPROACH TO ASSESS THE MECHANICAL PROPERTIES AND DEFORMATION MECHANISMS OF HIGH DOSE INCONEL X-750
机译:
使用多尺度方法评估高剂量Inonell X-750的力学性能和变形机理
作者:
C. Howard
;
C.D. Judge
;
V. Bhakhri
;
Q. Wang
;
M.R. Daymond
;
D. Murray
;
F. Teng
;
T. Skippon
;
M. Mattucci
;
H. Rajakumar
;
C. Mayhew
;
C. Dixon
;
D. Poff
;
G.A. Bickel
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
56.
LOCAL OXIDATION PENETRATION OF HYDROGEN-CHARGED 308L STAINLESS STEEL CLADDING IN DEAERATED PWR PRIMARY WATER
机译:
脱氢压水后原水充氢308L不锈钢包层的局部氧化渗透
作者:
Tongming Cui
;
Fei Ning
;
Jiarong Ma
;
Zhanpeng Lu
;
Kun Zhang
;
Yibo Jia
;
Xue Liang
;
Xiangkun Ru
;
Tetsuo Shoji
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
57.
'Mind the Gap' between old and new nuclear qualified cable: Lessons learned from aging, historical adverse events and other nuclear cable issues
机译:
新旧核电电缆之间的“差距”:从老化,历史不良事件和其他核电电缆问题中学到的经验教训
作者:
Larry Cunningham
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
58.
DEGRADATION EFFECTS OF HYDROGEN AND HIGH-TEMPERATURE WATER ENVIRONMENTS ON THE FRACTURE RESISTANCE OF LOW-ALLOY RPV STEELS
机译:
氢和高温水环境对低合金RPV钢抗断裂性能的影响
作者:
Z. Que
;
H. P. Seifert
;
P. Spätig
;
J. Holzer
;
A. Zhang
;
G. S. Rao
;
S. Ritter
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
59.
EFFECT OF AERATED TRANSIENTS ON OXIDATION AND SCC OF STAINLESS STEELS IN PWR PRIMARY WATER
机译:
充气瞬态对压水式原水中不锈钢的氧化和SCC的影响
作者:
Marc MAISONNEUVE
;
Cécilie DUHAMEL
;
Catherine GUERRE
;
Jérôme CREPIN
;
Ian DE CURIERES
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
60.
An Investigation into the Corrosion Behaviors of Stainless Steel and Ni-based Alloy in Simulated PWR Primary Water Environments
机译:
模拟压水堆一次水环境中不锈钢和镍基合金的腐蚀行为研究
作者:
Che Jung Chang
;
Mei Ya Wang
;
Tsung Kuang Yeh
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
61.
DEVELOPMENT AND VALIDATION OF A NEW EXPERIMENTAL DEVICE FOR STUDIES OF IODINE STRESS CORROSION CRACKING OF ZIRCONIUM ALLOYS
机译:
锆合金碘应力腐蚀裂纹研究新实验装置的研制与验证
作者:
Kamila WILCZYNSKA
;
Matthew BONO
;
David LE BOULCH
;
Marion FREGONESE
;
Valérie CHABRETOU
;
Nathanael MOZZANI
;
Laureline BARBIE
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
62.
NUCLEAR ELECTRICAL PENETRATOR ASSEMBLIES AND CONNECTORS FOR SMALL MODULAR AND OTHER ADVANCED REACTORS
机译:
小型模块化和其他高级反应堆的核动力渗透器组件和连接器
作者:
Pat Kumar
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
63.
EXAMINATION OF CRACKS IN PRESSURE SENSING LINES OF THE FEED WATER SYSTEM AND THE STRATEGY OF NPP GOESGEN FOR REPLACEMENT
机译:
给水系统压力传感线裂纹的检测及NPP Goesgen的更换策略
作者:
Thomas Wermelinger
;
Michael Schinhammer
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
64.
MICROSTRUCTURAL CHARACTERIZATION OF ALLOY 690TT EXPOSED TO Pb- CONTAINING CAUSTIC SOLUTIONS
机译:
含铅苛性溶液的690TT合金的微观结构表征
作者:
G. B. Mazzei
;
J. Duff
;
M. G. Burke
;
F. Scenini
;
G Meredith
;
T. Horner
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
65.
Nuclear concrete microstructure generation for simulating creep
机译:
模拟蠕变的核混凝土微观结构生成
作者:
Christa E. Torrence
;
Aishwarya Baranikumar
;
Zachary Grasley
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
66.
INTER-DIGITAL CAPACITIVE SENSOR FOR EVALUATING CABLE INSULATION THROUGH JACKET
机译:
用于通过护套评估电缆绝缘的数字电容传感器
作者:
S. W. Glass
;
L. S. Fifield
;
A. Sriraman
;
N. Bowler
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
67.
Parameter Sensitivity of Interdigital Sensors for their Design for Cable Insulation Aging Detection
机译:
用于电缆绝缘老化检测的叉指式传感器的参数敏感性
作者:
Md. Nazmul Al-Imran
;
S.W. Glass
;
Leonard S. Fifield
;
Mohammod Ali
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
68.
ELECTRICAL BREAKDOWN STRENGTH AND AC WITHSTAND IN HARVESTED EPR INSULATIONS FOR NUCLEAR POWER PLANTS
机译:
核电站EPR绝缘的电气击穿强度和交流耐受
作者:
Robert Duckworth
;
Alvin Ellis
;
Tam Ha
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
69.
THE DETERMINATION OF THE REACTION RATES, WATER VAPOR PERMEABILITY, AND ACTIVATION ENERGY FOR THERMAL OXIDATION OF LDPE FILMS
机译:
LDPE薄膜热氧化反应速率,水蒸气渗透率和活化能的测定
作者:
Noumon Munir
;
Keith B. Lodge
;
Brian Hinderliter
;
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》
70.
Stress assessment of baffle former bolt of PWR reactor for IASCC
机译:
IASCC压水堆反应堆挡板前螺栓的应力评估
作者:
A.M. Pandit
;
F.J. Blom
;
P.J. Baas
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
71.
Irradiation damage and IASCC of printed 316L for use as fuel cladding
机译:
用作燃料包层的印刷316L的辐照损伤和IASCC
作者:
M. McMurtrey
;
R. O’Brien
;
C. Sun
;
C. Shiau
;
F. Teng
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
72.
FORMULATION OF THE IRRADIATION ASSISTED STRESS CORROSION CRACK GROWTH RATES FOR NEUTRON-IRRADIATED STAINLESS STEELS IN HIGH-TEMPERATURE WATER OF A BOILING WATER REACTOR
机译:
沸水反应堆高温水中中子辐照不锈钢的辐照应力应力腐蚀裂纹扩展速率的公式
作者:
Masato Koshiishi
;
Kazuhiro Chatani
;
Shigeaki Tanaka
;
Hiroyuki Nakano
;
Takayuki Kaminaga
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
73.
IMPACT OF CORROSION OF NICKEL BASED COATINGS FOR FLOW ACCELERATED CORROSION CONTROL ON BWR RADIATION FIELDS
机译:
镍基涂层腐蚀对BWR辐射场的流动加速腐蚀控制的影响
作者:
Alfred J. Jarvis
;
Joseph F. Giannelli
;
Barry Gordon
;
David Segletes
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
74.
EVALUATION OF THE INFLUENCE OF IRRADIATION DOSE RATE ON CRACK GROWTH BEHAVIOR OF IRRADIATION ASSISTED STRESS CORROSION CRACKING FOR NEUTRON-IRRADIATED STAINLESS STEELS IN HIGH-TEMPERATURE WATER OF A BOILING WATER REACTOR
机译:
沸腾反应器中高温对中子辐照不锈钢辐照剂量率对辐照辅助应力腐蚀开裂裂纹扩展行为影响的评价
作者:
Masato Koshiishi
;
Ryoji Obata
;
Hiroyuki Nakano
;
Takayuki Kaminaga
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
75.
Development of Mo-free Low Alloy Steels for Mitigation of Flow-Accelerated Corrosion in Secondary Side of PWRs
机译:
开发用于缓解压水堆次级侧流动加速腐蚀的无钼低合金钢
作者:
Seunghyun Kim
;
Gi Dong Kim
;
Ji Hyun Kim
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
76.
ALKALI-ACTIVATED CONCRETES FOR USE IN NUCLEAR FACILITIES AND HIGH TEMPERATURE ENVIRONMENTS
机译:
碱活化混凝土,用于核设施和高温环境
作者:
Casey Sundberg
;
Mary Christiansen
;
Andrea Schokker
;
Brian Hinderliter
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
77.
EXPERIMENTAL DESIGN TO STUDY THE MECHANICAL PERFORMANCE OF IRRADIATED ACRYLONITRILE BUTADIENE STYRENE FABRICATED VIA FUSED FILAMENT FABRICATION
机译:
通过熔丝制备研究辐照丙烯腈丁二烯苯乙烯共聚物机械性能的实验设计
作者:
Arielle J. Miller
;
Grant Warner
;
Dharmaraj Raghavan
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
78.
TECHNICAL BASIS FOR ASME SECTION XI CODE CASE FOR STRESS CORROSION CRACK GROWTH RATE EVALUATIONS FOR ALLOY 690 AND ASSOCIATED WELDS
机译:
合金690及相关焊缝应力腐蚀裂纹扩展速率评估的ASME第XI代码案例的技术基础
作者:
Warren Bamford
;
Amanda Jenks
;
Ron Janowiak
;
Gary Stevens
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
79.
Comparison of selective oxidation in Ni-based alloys exposed to PWR primary water and Rhines Pack environments
机译:
暴露于PWR原始水和Rhines Pack环境中的镍基合金中选择性氧化的比较
作者:
Karen Kruska
;
Daniel K Schreiber
;
Matthew J Olszta
;
Brian J Riley
;
Stephen M Bruemmer
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
80.
DEVELOPMENT OF PREDICTION METHOD FOR RADIATION HARDENING OF REACTOR INTERNALS
机译:
反应堆内部辐射硬化预测方法的发展
作者:
Hitoshi Seto
;
Yuji Kitsunai
;
Shigeaki Tanaka
;
Ryoji Obata
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
81.
Corrosion Behavior of Candidate Alloys used in Supercritical Water Environment
机译:
在超临界水环境中使用的候选合金的腐蚀行为
作者:
Hsuan-Kan Lin
;
Tsung-Kuang Yeh
;
Mei-Ya Wang
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
82.
Effect of Cold Work on Hydrogen Diffusion in Zr-2.5Nb Alloys at Reactor Temperatures
机译:
反应温度下冷作对Zr-2.5%Nb合金氢扩散的影响
作者:
Heidi Nordin
;
Jaganathan Ulaganathan
;
Sean Hanlon
;
Dylan Broad
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
83.
Hydrogen Assisted Oxidation Behavior of Alloy 600 in High Temperature Air Environment by in-situ H Charging Method
机译:
原位氢充注法在高温空气环境中600合金的氢辅助氧化行为
作者:
Zihao WANG
;
Yoichi TAKEDA
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
84.
ASSESSMENT OF DEFORMATION MECHANISMS IN NEUTRON-IRRADIATED ACCIDENT-TOLERANT FeCrAl ALLOYS VIA IN SITU MECHANICAL TESTING AND TEM ANALYSIS
机译:
通过原位机械测试和TEM分析评估中子辐照耐候FeCrAl合金的变形机理
作者:
M. N. Gussev
;
D. Zhang
;
K. G. Field
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
85.
SCC INITIATION OF ALLOY 82 IN NOMINAL PWR PRIMARY WATER AT 360°C
机译:
在360°C的名义压水原始水中合金82的SCC起始
作者:
Catherine GUERRE
;
Brice BOURDILIAU
;
Emilien BURGER
;
Thierry COUVANT
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
86.
OXIDATION CHARACTERISTICS OF Ni-Cr-Fe ALLOY SYSTEMS IN SIMULATED PWR PRIMARY WATER: EFFECT OF Cr CONTENT
机译:
模拟压水堆原水中Ni-Cr-Fe合金体系的氧化特性:Cr含量的影响
作者:
Hee-Sang Shim
;
Do Haeng Hur
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
87.
EFFECT OF PROTON-INDUCED RADIOLYSIS ON CORROSION OF 316L STAINLESS STEEL IN SIMULATED PWR PRIMARY WATER
机译:
质子诱导的放射性对模拟压水堆原水中316L不锈钢腐蚀的影响
作者:
Rigel D. Hanbury
;
Gary S. Was
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
88.
SPATIAL RESOLUTION OF A CABLE FAULT LOCATION ATTEMPT BY FREQUENCY DOMAIN REFLECTOMETRY
机译:
频率域反射法的电缆故障定位尝试的空间分辨
作者:
Yoshimichi Ohki
;
Naoshi Hirai
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
89.
EFFECT OF MARTENSITE ON SCC INITIATION IN AUSTENTITC STAINLESS STEELS IN SIMULATED PWR PRIMARY WATER ENVIRONMENT
机译:
模拟压水堆原始水环境中马氏体对奥氏体不锈钢应力腐蚀开裂的影响
作者:
Litao Chang
;
M. Grace Burke
;
Jonathan Duff
;
Fabio Scenini
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
90.
POST-IRRADIATION EXAMINATION OF HIGH FLUENCE BAFFLE-FORMER BOLTS RETRIEVED FROM A WESTINGHOUSE TWO-LOOP DOWNFLOW TYPE PWR
机译:
从西屋两环流向下型压水堆取回的高强度隔板形螺栓的辐照后检查
作者:
Xiang Chen
;
Tianyi Chen
;
Chad M. Parish
;
Mikhail A. Sokolov
;
Keith J. Leonard
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
91.
On the Potential Synergies of Helium and Hydrogen on the Nucleation and Stability of Cavity Clusters in Inconel X-750® Irradiated in a High Thermal Neutron Flux Spectra
机译:
高热中子通量光谱中辐照的InconelX-750®中氦和氢的潜在协同作用对腔团簇形核和稳定性的潜在影响
作者:
C.D. Judge
;
H. Rajakumar
;
A. Korinek
;
G.A. Bickel
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
92.
The Stress Intensity Factor Dependence of 304 Stainless Steel SCC growth in Deaerated Water
机译:
脱气水中304不锈钢SCC生长的应力强度因子依赖性
作者:
David Morton
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
93.
STRESS CORROSION CRACKING OF STAINLESS STEEL CLADDING LAYERS IN SIMULATED PWR PRIMARY WATER
机译:
模拟压水堆原水中不锈钢覆层的应力腐蚀开裂
作者:
Qi Xiong
;
Tongming Cui
;
Jiarong Ma
;
Zhanpeng Lu
;
Fei Ning
;
Junjie Chen
;
Kun Zhang
;
Zhiming Zhong
;
Guangdong Han
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
94.
ELECTROCHEMICAL BEHAVIOUR OF DILUTE ELECTROACTIVE SPECIES ON A HEAT-TRANSFER SURFACE UNDER BOILING
机译:
沸腾下传热表面上稀薄带电物质的电化学行为
作者:
Caitlin Dever
;
Jaganathan Ulaganathan
;
Stan Klimas
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
95.
DIFFUSING HYDROGEN EFFECT ON THE OXIDE FILM ON 316L SS IN HIGH TEMPERATURE WATER
机译:
高温水中316L不锈钢上氧化膜的氢扩散作用
作者:
Jiarong Ma
;
Tongming Cui
;
Hao Peng
;
Zhanpeng Lu
;
Junjie Chen
;
Yibo Jiap
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
96.
FRACTURE BEHAVIOR OF OXIDIZED GRAIN BOUNDARY IN NEUTRON-IRRADIATED STAINLESS STEEL
机译:
中子辐照不锈钢中氧化晶粒边界的断裂行为
作者:
Terumitsu Miura
;
Katsuhiko Fujii
;
Koji Fukuya
;
Yuji Kitsunai
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
97.
Characterizing Fatigue Damage in Zr-2.5Nb
机译:
Zr-2.5Nb中疲劳损伤的表征
作者:
H.M. Nordin
;
M. Mattucci
;
A. Phillion
;
T.M. Karlsen
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
98.
Tests on Mock-ups Representative of Rolling Zones of Steam Generator tubes in High Temperature Hydrogenated Water
机译:
在高温氢化水中模拟蒸汽发生器管滚动区的模型试验
作者:
Daniel Brimbal
;
Steve Fyfitch
;
Olivier Calonne
;
Nicolas Huin
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
99.
Corrosion of Fe-12Cr-2Si alloy in High Temperature Steam Environments
机译:
Fe-12Cr-2Si合金在高温蒸汽环境下的腐蚀
作者:
Amanda Leong
;
Jinsuo Zhang
;
Yi Xie
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
100.
Transgranular Stress Corrosion Cracking Mechanism of Alloy 690 in Simulated PWR Primary Water
机译:
模拟压水堆一次水中690合金的晶界应力腐蚀开裂机理
作者:
H.P. Kim
;
S.W.Kim
;
S.H. Lim
;
J.Y.Lee
;
S.H.Cho
;
M.J.Choi
;
S.S.Hwang
;
Y.S.Lim
;
D.J.Kim
会议名称:
《American Nuclear Society;International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors》
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