Related references
Note: Only part of the references are listed.Proton irradiation assisted localized corrosion and stress corrosion cracking in 304 nuclear grade stainless steel in simulated primary PWR water
Ping Deng et al.
JOURNAL OF MATERIALS SCIENCE & TECHNOLOGY (2021)
Microstructure, mechanical properties and IASCC susceptibility of stainless steel baffle bolts after 30 years of operation in a PWR
Rik-Wouter Bosch et al.
JOURNAL OF NUCLEAR MATERIALS (2021)
Grain boundary oxidation of proton-irradiated nuclear grade stainless steel in simulated primary water of pressurized water reactor
Ping Deng et al.
SCIENTIFIC REPORTS (2021)
A high-resolution characterization of the initiation of stress corrosion crack in Alloy 690 in simulated pressurized water reactor primary water
Wenjun Kuang et al.
CORROSION SCIENCE (2020)
The effect of grain boundary structure on the intergranular degradation behavior of solution annealed alloy 690 in high temperature, hydrogenated water
Wenjun Kuang et al.
ACTA MATERIALIA (2020)
Microstructural examination of neutron, proton and self-ion irradiation damage in a model Fe9Cr alloy
J. C. Haley et al.
JOURNAL OF NUCLEAR MATERIALS (2020)
Irradiation -assisted stress corrosion cracking of Type 347 and Type 316 steels irradiated in commercial pressurized water reactors *
Michael R. Ickes et al.
JOURNAL OF NUCLEAR MATERIALS (2020)
Microstructural characterization of cold-worked 316 stainless steel flux thimble tubes irradiated up to 100 dpa in a commercial Pressurized Water Reactor
Miao Song et al.
JOURNAL OF NUCLEAR MATERIALS (2020)
An insight into PWR primary water SCC mechanisms by comparing surface and crack oxidation
Zhao Shen et al.
CORROSION SCIENCE (2019)
Irradiation assisted stress corrosion cracking of commercial and advanced alloys for light water reactor core internals
Mi Wang et al.
JOURNAL OF NUCLEAR MATERIALS (2019)
Role of irradiation and irradiation defects on the oxidation first stages of a 316L austenitic stainless steel
M. Boisson et al.
CORROSION SCIENCE (2019)
The effect of post-irradiation annealing on the stress corrosion crack growth rate of neutron-irradiated 304L stainless steel in boiling water reactor environment
Wenjun Kuang et al.
CORROSION SCIENCE (2019)
Quantitative linkage between the stress at dislocation channel - Grain boundary interaction sites and irradiation assisted stress corrosion crack initiation
D. C. Johnson et al.
ACTA MATERIALIA (2019)
Mechanistic understanding of the temperature dependence of crack growth rate in alloy 600 and 316 stainless steel through high-resolution characterization
Zhao Shen et al.
ACTA MATERIALIA (2019)
Insights into the stress corrosion cracking of solution annealed alloy 690 in simulated pressurized water reactor primary water under dynamic straining
Wenjun Kuang et al.
ACTA MATERIALIA (2018)
Oxide inclusions in laser additive manufactured stainless steel and their effects on impact toughness and stress corrosion cracking behavior
Xiaoyuan Lou et al.
JOURNAL OF NUCLEAR MATERIALS (2018)
Opportunities and limitations for ion beams in radiation effects studies: Bridging critical gaps between charged particle and neutron irradiations
S. J. Zinkle et al.
SCRIPTA MATERIALIA (2018)
Effect of irradiation on corrosion of 304 nuclear grade stainless steel in simulated PWR primary water
Ping Deng et al.
CORROSION SCIENCE (2017)
An atom probe tomography study of internal oxidation processes in Alloy 600
B. Langelier et al.
ACTA MATERIALIA (2016)
A mechanistic study of the temperature dependence of the stress corrosion crack growth rate in SUS316 stainless steels exposed to PWR primary water
Martina Meisnar et al.
ACTA MATERIALIA (2016)
Properties of oxide films formed on 316L SS and model alloys with modified Ni, Cr and Si contents in high temperature water
Guangdong Han et al.
CORROSION SCIENCE (2016)
Challenges to the use of ion irradiation for emulating reactor irradiation
Gary S. Was
JOURNAL OF MATERIALS RESEARCH (2015)
Crack initiation testing of thimble tube material under PWR conditions to determine a stress threshold for IASCC
R. W. Bosch et al.
JOURNAL OF NUCLEAR MATERIALS (2015)
Hydrothermal corrosion of SiC in LWR coolant environments in the absence of irradiation
K. A. Terrani et al.
JOURNAL OF NUCLEAR MATERIALS (2015)
Emulation of reactor irradiation damage using ion beams
G. S. Was et al.
SCRIPTA MATERIALIA (2014)
Irradiation-Assisted Stress Corrosion Cracking and Impact on Life Extension
Anna Hojna
CORROSION (2013)
Current understanding of radiation-induced degradation in light water reactor structural materials
Koji Fukuya
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY (2013)
On the use of SRIM for computing radiation damage exposure
R. E. Stoller et al.
NUCLEAR INSTRUMENTS & METHODS IN PHYSICS RESEARCH SECTION B-BEAM INTERACTIONS WITH MATERIALS AND ATOMS (2013)
Nanoscale characterisation of grain boundary oxidation in cold-worked stainless steels
Karen Kruska et al.
CORROSION SCIENCE (2012)
Deformation Structure in Highly Irradiated Stainless Steels
Hiromasa NISHIOKA et al.
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY (2012)
Irradiation-assisted stress corrosion cracking
Gary S. Was et al.
CORROSION REVIEWS (2011)
A review of irradiation effects on LWR core internal materials - IASCC susceptibility and crack growth rates of austenitic stainless steels
O. K. Chopra et al.
JOURNAL OF NUCLEAR MATERIALS (2011)
Localized deformation and IASCC initiation in austenitic stainless steels
Z. Jiao et al.
JOURNAL OF NUCLEAR MATERIALS (2008)
Influence of carbide precipitation and rolling direction on intergranular stress corrosion cracking of austenitic stainless steels in hydrogenated high-temperature water
K. Arioka et al.
CORROSION (2006)
Isolating the effect of radiation-induced segregation in irradiation-assisted stress corrosion cracking of austenitic stainless steels
JT Busby et al.
JOURNAL OF NUCLEAR MATERIALS (2002)
Emulation of neutron irradiation effects with protons: validation of principle
GS Was et al.
JOURNAL OF NUCLEAR MATERIALS (2002)
Sink effect of grain boundary on radiation-induced segregation in austenitic stainless steel
S Watanabe et al.
JOURNAL OF NUCLEAR MATERIALS (2000)
Effects of grain boundary misorientation on solute segregation in thermally sensitized and proton-irradiated 304 stainless steel
TS Duh et al.
JOURNAL OF NUCLEAR MATERIALS (2000)
Effects of impurities on environmentally assisted crack growth of solution-annealed austenitic steels in primary water at 325°C
GF Li et al.
CORROSION (2000)
Share Paper
