Irradiation assisted stress corrosion cracking of controlled purity 304L stainless steels

Abstract

The effect of chromium, phosphorus,, silicon and sulfur on the stress corrosion cracking of 304L stainless steel in CERT tests in high purity water or argon at 288°C following irradiation with 3.4 MeV protons at gl0*C to 1 dpa, has been investigated using ultrahigh pur:ty alloys (UHP) with controlled impurity additions. Grain boundary segregation of phosphorus or silicon due to proton irradiation was quantified using both Auger electron spectroscopy and scanning transmission electron microscopy, and the alloys with impurity element additions were observed to have greater grain boundary chromium depletion and nickel enrichment than the UHP alloy. The UHP alloy suffered severe cracking in CERT tests in water. Less cracking was found after CERT tests of irradiated UHP+ P or UHP+Si alloys, despite greater chromium depletion. This suggests a mitigating effect of phosphorus and silicon at grain boundaries. No cracking was fou-.~ in argon tests, eliminating a purely mechanical embrittlement mechanism, but not eliminating a contribution from radiation hardening. Implanted hydrogen was not a factor in the intergranular cracking found.

The effect of oversized solute additions on the microstructure of 316SS irradiated with 5 MeV Ni++ ions or 3.2 MeV protons

“…This work used Ni-ion and proton irradiation to study the effects of misfit elements, Hf and Pt, on the radiation-induced microstructure in alloys 316SS. Heavy-ion and proton irradiations are used due to the advantages that previous experience demonstrated that these irradiations could effectively produce similar radiationinduced microstructure and microchemistry relevant to LWRs at around 300°C [9][10][11][12][13][14]. The use of heavy-ion and proton irradiation in this work also distinguishes whether the major effect of the solute additions is on cascade processes or isolated point defect kinetics during the development of irradiated microstructure.…”

Alloy Selection for Accident Tolerant Fuel Cladding in Commercial Light Water Reactors

“…[10] Austenitic stainless steel (SS) core internal components are susceptible to irradiation-assisted stress corrosion cracking (IASCC) during service in nuclear power plants' light water reactors. [11][12][13] One of the effects of irradiation is the hardening of the SS due to modifications in the dislocation distribution in the alloy. [14,15] Irradiation also alters the local chemistry of these austenitic alloys, for example, in the vicinity of grain boundaries by a mechanism of radiation-induced segregation.…”

Developing improved structural materials using proton irradiation as a rapid analysis tool.

“…Note the excellent correlation between 1 dpa proton irradiation data and the 1.1 dpa neutron data. Finally, we have consistently observed that proton irradiation, under the same conditions as neutron irradiation, causes IASCC susceptibility in austenitic stainless steel [10]. Figure 2.13…”