Operational feedback on the primary circuit of Pressurized Water Reactors (PWR) shows cases of stress corrosion cracking (SCC) affecting cold-worked stainless steel components. Some working steps require water from auxiliary reservoirs to be added to the primary water. Depending on the operator, this water can be either aerated, or deaerated and monitored. These aerated transients may have a detrimental effect on the SCC susceptibility of stainless steels. In this context, the aim of this work is to study the influence of dissolved oxygen on the oxidation and SCC susceptibility of a cold-worked 316L stainless steel in PWR primary water. For this purpose, oxidation and SCC tests were performed in PWR primary water with nominal (hydrogenated ...
International audienceThe sensitivity of precipitation-strengthened A286 austenitic stainless steel ...
Recently intergranular stress corrosion cracking (IGSCC) in austenitic stainless steel 316L of Prima...
L’état de surface des vis de liaison des internes de cuve du circuit primaire des REP (Réacteurs à E...
International audienceExperience on Pressurized Water Reactors (PWR) shows cases of Stress Corrosion...
Les aciers inoxydables austénitiques de type 304L et 316L sont largement employés dans le circuit pr...
International audience304L and 316L SS samples are SCC-tested in PWR water for various conditions: s...
Austenitic stainless steels are widely used in primary circuits of Pressurized Water Reactors (PWR) ...
Internal parts of pressurized water reactor (PWR) vessels are often made of austenitic stainless ste...
Slow strain rate tensile tests (SSRT) and oxidation tests were performed on a 316L stainless steel i...
Alloy 600, a nickel base alloy (containing 15 wt% Cr), is susceptible to intergranular stress corros...
Oxidation and stress corrosion cracking (SCC) of 316L stainless steel were studied in simulated pres...
La majorité des composants internes de la cuve des réacteurs à eau pressurisée (REP) est fabriquée e...
Since the early 1970s certain component parts of primary loops of nuclear pressurised water reactors...
The oxidation of 316 L stainless steel in hydrogenated supercritical water at 600 °C is strongly dep...
International audienceThe sensitivity of precipitation-strengthened A286 austenitic stainless steel ...
Recently intergranular stress corrosion cracking (IGSCC) in austenitic stainless steel 316L of Prima...
L’état de surface des vis de liaison des internes de cuve du circuit primaire des REP (Réacteurs à E...
International audienceExperience on Pressurized Water Reactors (PWR) shows cases of Stress Corrosion...
Les aciers inoxydables austénitiques de type 304L et 316L sont largement employés dans le circuit pr...
International audience304L and 316L SS samples are SCC-tested in PWR water for various conditions: s...
Austenitic stainless steels are widely used in primary circuits of Pressurized Water Reactors (PWR) ...
Internal parts of pressurized water reactor (PWR) vessels are often made of austenitic stainless ste...
Slow strain rate tensile tests (SSRT) and oxidation tests were performed on a 316L stainless steel i...
Alloy 600, a nickel base alloy (containing 15 wt% Cr), is susceptible to intergranular stress corros...
Oxidation and stress corrosion cracking (SCC) of 316L stainless steel were studied in simulated pres...
La majorité des composants internes de la cuve des réacteurs à eau pressurisée (REP) est fabriquée e...
Since the early 1970s certain component parts of primary loops of nuclear pressurised water reactors...
The oxidation of 316 L stainless steel in hydrogenated supercritical water at 600 °C is strongly dep...
International audienceThe sensitivity of precipitation-strengthened A286 austenitic stainless steel ...
Recently intergranular stress corrosion cracking (IGSCC) in austenitic stainless steel 316L of Prima...
L’état de surface des vis de liaison des internes de cuve du circuit primaire des REP (Réacteurs à E...