Stress Corrosion Cracking of Cold Worked Austenitic Stainless Steels in Laboratory Primary PWR Environment

Abstract

Austenitic Stainless Steels (ASS) are widespread in primary and auxiliary circuits of Pressurized Water Reactors (PWRs). Moreover, some components suffer stress corrosion cracking (SCC) under neutron irradiation. This degradation could be the result of the increase of hardness and / or the modification of chemical composition at the grain boundary by irradiation. In order to avoid complex and costly corrosion facilities, the effects of radiation hardening on the material are commonly simulated by applying a pre-strain on non-irradiated material prior to stress corrosion cracking tests. The typical features of the cracking process in primary environment at 360°C during CERTs included an initiation stage (composed of a true initiation time and a slow propagation regime leading to a crack depth lower than 50 μm), then a “rapid” propagation stage before mechanical failure. Pre-straining increased significantly CGRs and the mode of pre-straining could strongly modify the crack path. No significant cracking (< 50 μm) was obtained under a pure static loading. A dynamic loading (CERT or cyclic) was required and various thresholds (hardness, elongation, stress) for the occurrence of SCC were determined. An important R&D program is in progress to develop initiation and propagation models for SCC of austenitic SS in primary environment.

Short review: Potential impact of delamination cracks on fracture toughness of structural materials

“…In the last 4 decades, some researcher noticed the existence of delamination cracks and tried to understand the origins, mechanisms and consequences on the fracture toughness, impact toughness and tensile properties. Some delamination cracks are shown on the fracture surfaces of irradiated 08Ch18N10T with elongated δ-ferrite [16,17], the 304 L and the 316 L after exposure in PWR [20] environments and degraded by hydrogen embrittlement [21].…”

Corrosion Behavior of Stainless Steels in Simulated PWR Primary Water—Effect of Chromium Content in Alloys and Dissolved Hydrogen—

“…Many studies have been conducted on nonsensitized stainless steels to investigate the influences of various parameters, such as water chemistry, material composition and stress, on the initiation and growth of SCC and to elucidate SCC mechanisms. [1][2][3][4][5][6][7][8][9][10][11][12] With respect to SCC in hydrogenated high-temperature water relevant to PWR primary water, referred to as low-potential SCC (LPSCC), effects of chromium content in materials and dissolved hydrogen have recently been reported by Arioka et al 12) The chromium content in materials has a beneficial effect on LPSCC susceptibility, whereas dissolved hydrogen accelerates LPSCC based on the constant extension rate technique (CERT) results. Since water chemistry and material composition influence both SCC susceptibility and corrosion behavior, the mechanism of SCC is considered to be inevitably related to corrosion behavior.…”

Corrosion Behavior of Stainless Steels in Simulated PWR Primary Water —Effect of Chromium Content in Alloys and Dissolved Hydrogen—

“…Many studies have been conducted on nonsensitized stainless steels to investigate the influences of various parameters, such as water chemistry, material composition and stress, on the initiation and growth of SCC and to elucidate SCC mechanisms. [1][2][3][4][5][6][7][8][9][10][11][12] With respect to SCC in hydrogenated high-temperature water relevant to PWR primary water, referred to as low-potential SCC (LPSCC), effects of chromium content in materials and dissolved hydrogen have recently been reported by Arioka et al…”