Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors

1. PWR Nickel SCC—Aging Effects

   1. The Effect of Hardening via Long Range Order on the SCC and LTCP Susceptibility of a Nickel-30Chromium Binary Alloy

      • Tyler E. Moss, Catherine M. Brown, George A. Young

      Pages 261-279

   2. PWSCC Initiation of Alloy 600: Effect of Long-Term Thermal Aging and Triaxial Stress

      • Seung Chang Yoo, Kyoung Joon Choi, Seunghyun Kim, Ji-Soo Kim, Byoung Ho Choi, Yun-Jae Kim et al.

      Pages 281-292

   3. Stress Corrosion Cracking Behavior of Alloy 718 Subjected to Various Thermal Mechanical Treatments in Primary Water

      • Mi Wang, Miao Song, Gary S. Was, L. Nelson

      Pages 293-305

   4. Time- and Fluence-to-fracture Studies of Alloy 718 in Reactor

      • C. Joseph Long

      Pages 307-320

   5. Development of Short-Range Order and Intergranular Carbide Precipitation in Alloy 690 TT upon Thermal Ageing

      • Roman Mouginot, Teemu Sarikka, Mikko Heikkilä, Mykola Ivanchenko, Unto Tapper, Ulla Ehrnstén et al.

      Pages 321-334

2. PWR Nickel SCC—Alloy 600 Mechanistic

   1. Front Matter

      Pages 335-335

      PDF

   2. Diffusion Processes as Possible Mechanisms for Cr Depletion at SCC Crack Tip

      • Josiane Nguejio, Jérôme Crépin, Cécilie Duhamel, Fabrice Gaslain, Catherine Guerre, François Jomard et al.

      Pages 337-357

   3. Role of Grain Boundary Cr5B3 Precipitates on Intergranular Attack in Alloy 600

      • Daniel K. Schreiber, Matthew J. Olszta, Karen Kruska, Stephen M. Bruemmer

      Pages 359-374

   4. Advanced Characterization of Oxidation Processes and Grain Boundary Migration in Ni Alloys Exposed to 480 °C Hydrogenated Steam

      • S. Y. Persaud, B. Langelier, A. Eskandari, H. Zhu, G. A. Botton, R. C. Newman

      Pages 375-398

   5. Exploring Nanoscale Precursor Reactions in Alloy 600 in H2/N2–H2O Vapor Using In Situ Analytical Transmission Electron Microscopy

      • M. G. Burke, G. Bertali, F. Scenini, S. J. Haigh, E. Prestat

      Pages 399-407

   6. Electrochemical and Microstructural Characterization of Alloy 600 in Low Pressure H2-Steam

      • L. Volpe, G. Bertali, M. Curioni, M. G. Burke, F. Scenini

      Pages 409-422

   7. Effect of Dissolved Hydrogen on the Crack Growth Rate and Oxide Film Formation at the Crack Tip of Alloy 600 Exposed to Simulated PWR Primary Water

      • Johan Stjärnsäter, Jiaxin Chen, Fredrik Lindberg, Peter Ekström, Pål Efsing

      Pages 423-437

   8. A Mechanistic Study of the Effect of Temperature on Crack Propagation in Alloy 600 Under PWR Primary Water Conditions

      • Zhao Shen, Sergio Lozano-Perez

      Pages 439-454

3. PWR Nickel SCC—Alloy 690 Mechanistic

   1. Front Matter

      Pages 455-455

      PDF

   2. Grain Boundary Damage Evolution and SCC Initiation of Cold-Worked Alloy 690 in Simulated PWR Primary Water

      • Ziqing Zhai, Mychailo Toloczko, Karen Kruska, Daniel Schreiber, Stephen Bruemmer

      Pages 457-483

   3. PWSCC Susceptibility of Alloy 690, 52 and 152

      • Takaharu Maeguchi, Kimihisa Sakima, Kenji Sato, Koji Fujimoto, Yasuto Nagoshi, Kazuya Tsutsumi

      Pages 485-500

   4. Relationship Among Dislocation Density, Local Strain Distribution, and PWSCC Susceptibility of Alloy 690

      • Tae-Young Ahn, Sung-Woo Kim, Seong Sik Hwang, Hong-Pyo Kim

      Pages 501-507

   5. Morphology Evolution of Grain Boundary Carbides Precipitated Near Triple Junctions in Highly Twinned Alloy 690

      • Hui Li, Xirong Liu, Kai Zhang, Wenqing Liu, Shuang Xia

      Pages 509-517

   6. A Mechanistic Study of Stress Corrosion Crack Propagation in Heavily Cold Worked TT Alloy 690 Exposed to Simulated PWR Primary Water

      • Toshio Yonezawa, Masashi Watanabe, Atsushi Hashimoto

      Pages 519-534

   7. Microstructural Study on the Stress Corrosion Cracking of Alloy 690 in Simulated Pressurized Water Reactor Primary Environment

      • Wenjun Kuang, Miao Song, Chad M. Parish, Gary S. Was

      Pages 535-545

This two-volume set represents a collection of papers presented at the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors. The purpose of this conference series is to foster an exchange of ideas about problems and their remedies in water-cooled nuclear power plants of today and the future. Contributions cover problems facing nickel-based alloys, stainless steels, pressure vessel and piping steels, zirconium alloys, and other alloys in water environments of relevance. Components covered include pressure boundary components, reactor vessels and internals, steam generators, fuel cladding, irradiated components, fuel storage containers, and balance of plant components and systems.

• Materials degradation
• Environmental degradation
• Water reactors
• Nuclear power systems
• Corrosion

• Idaho National Laboratory, Idaho Falls, USA

  John H. Jackson

• Naval Nuclear Laboratory, Pittsburgh, USA

  Denise Paraventi

• Chalk River Laboratories, Canadian Nuclear Laboratories, Chalk River, Canada

  Michael Wright

The Minerals, Metals & Materials Society (TMS) is a member-driven international professional society dedicated to fostering the exchange of learning and ideas across the entire range of materials science and engineering, from minerals processing and primary metals production, to basic research and the advanced applications of materials. Included among its nearly 13,000 professional and student members are metallurgical and materials engineers, scientists, researchers, educators, and administrators from more than 70 countries on six continents.

Policies and ethics