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

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 cl...

Full description

Saved in:
Bibliographic Details
Published in:Materials 2023-03, Vol.16 (7), p.2585
Main Authors: Margolin, Boris, Pirogova, Natalia, Sorokin, Alexander, Kokhonov, Vasiliy, Dub, Alexey, Safonov, Ivan
Format: Article
Language:English
Subjects:
Autoclaves
Corrosion
Corrosion resistance
Ferritic stainless steels
Fuel cycles
Martensitic stainless steels
Materials selection
Nuclear safety
Primary circuits
Radiation
Radiation damage
Reactors
Stainless steel
Steel
Steel, High strength
Stress corrosion cracking
Thermodynamic efficiency
Threshold stress
Zirconium alloys
Citations: Items that this one cites
Items that cite this one
Online Access:Get full text
Tags: Add Tag
No Tags, Be the first to tag this record!
Description
Summary: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.
ISSN:1996-1944
1996-1944
DOI:10.3390/ma16072585