Analysis of mechanisms inducing corrosion cracking of irradiated austenitic steels and development of a model for prediction of crack initiation

Марголин, Б. З.; Sorokin, A. A.; Pirogova, N. E.; Toivonen, Aki; Sefta, Faiza; Pokor, C.

doi:10.1016/j.engfailanal.2019.104235

Cited by 11 publications

(4 citation statements)

References 28 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…Radiation-hardened steel in some cases is subjected to the creep process when it is loaded after irradiation [ 19 ]. In addition, radiation hardening leads to an increase in the portion of creep strain due to grain boundary (GB) sliding.…”

Section: Discussionmentioning

confidence: 99%

“…Localized deformation, the creep process, and GB sliding, in the general case, can lead to SCC since these processes result in the passivating film rupturing [ 19 ].…”

Section: Discussionmentioning

confidence: 99%

“…According to [ 19 , 31 , 32 , 33 ], irradiated austenitic stainless chromium-nickel steels are susceptible to SCC in the WWER primary circuit coolant environment at stresses above 0.3–0.4-σ Y (σ Y is the yield strength of irradiated material). Thus, to determine the threshold stress, it is necessary to carry out autoclave tests in the simulated primary circuit coolant environment of SWCR at stresses in the range of 0.3∙σ Y ÷0.7∙σ Y in order to confirm the corrosive microcrack nucleation at stresses above the threshold and the absence of microcracks at stresses below the threshold.…”

Section: Load Calculation For Disc Specimens In the Loading Devicementioning

confidence: 99%

“…Most of the data relates to the SCC of irradiated stainless steel in PWR and BWR environments [ 19 , 20 , 21 , 22 , 23 , 24 , 25 , 26 , 27 , 28 , 29 ]. According to these investigations, the main mechanisms of SSCs are GBs chromium depletion, hardening, localized deformation, GBs weakening, and creep-induced GBs sliding.…”

Section: Introductionmentioning

confidence: 99%

See 3 more Smart Citations

Investigation of Stress Corrosion Cracking Resistance of Irradiated 12Cr Ferritic-Martensitic Stainless Steel in Supercritical Water Environment

et al. 2023

Self Cite

View full text Add to dashboard Cite

The supercritical water-cooled reactors (SWCR) belong to Generation IV of reactors. These reactors have a number of advantages over currently operating WWERs and PWRs. These advantages include higher thermal efficiency, a more simplified unit design, and the possibility of incorporating it into a closed fuel cycle. It is therefore necessary to identify candidate materials for the SWCR and validate the safety and effectiveness of their use. 12Cr ferritic–martensitic (F/M) stainless steel is considered a candidate material for SWCR internals. Radiation embrittlement and corrosion cracking in the primary circuit coolant environment are the main mechanisms of F/M steels degradation during SWCR operation. Here, the stress corrosion cracking (SCC) in supercritical water at 390 and 550 °C of 12Cr F/M steel irradiated by neutrons to 12 dpa is investigated. Autoclave tests of specially designed disk specimens in supercritical water were performed. The tests were carried out under different constant load (CL), temperature 450 °C, and pressure in autoclave 25 MPa. The threshold stress, below which the SCC initiation of irradiated 12Cr F/M steel does not occur, was determined.

show abstract

Section: Discussionmentioning

confidence: 99%

“…Localized deformation, the creep process, and GB sliding, in the general case, can lead to SCC since these processes result in the passivating film rupturing [ 19 ].…”

Section: Discussionmentioning

confidence: 99%