Helium effects on irradiation assisted stress corrosion cracking susceptibility of 316L austenitic stainless steel

The formation and growth of cracks by irradiation assisted stress corrosion cracking (IASCC) in light water reactor internals is a critical issue for a safe long-term operation of nuclear power plants. The IASCC susceptibility at relatively low dose is dominated by conventional mechanisms such as radiation-induced segregation and radiation hardening. However, the ageing of the nuclear fleet combined with the increase of their life-span reveals other mechanisms that could play an important role on IASCC susceptibility. Recent studies show that a huge amount of helium (He) can be accumulated in reactor internal components of pressurized water reactors (PWR) after long-term operation. This occurrence could significantly increase the IASCC susceptibility at high doses. The main objective was to investigate the He effects on IASCC susceptibility of the solution-annealed (SA) and 20% cold-worked (CW) austenitic steel 316L. SA and CW miniaturized tensile samples and a SA plate were homogeneously implanted up to 1000 appm He. He was implanted at ~45 MeV and 300°C at the cyclotron in CEMHTI/CNRS (France). Slow strain rate tests (SSRT) in air and in high-temperature water were then carried out on SA, CW, post-implantation annealed (PIA) and as-implanted samples; instrumented nanoindentation tests were performed at room temperature on non-implanted and implanted specimens; and scanning electron microscope and transmission electron microscope were employed for fractography and microstructural characterization. The results of SSRTs in high-temperature air and in hydrogenated high-temperature water showed that homogenized implanted He up to 1000 appm corresponding to a displacement damage of about 0.16 dpa (displacement per atom) does not produce intergranular cracking and IASCC. However, the deformation microstructure of non-implanted SA/CW is characterized by high dislocation density arranged in cell walls separated by relatively dislocation-free regions, while the as-implanted SA samples (> 0.05 dpa) exhibit a more planar deformation microstructure. Characterization of as-implanted CW and PIA samples did not show any significant implantation effect on the deformation microstructure. PIA of the He implanted plate was carried out from 650 to 1000°C for 1h to increase the helium grain boundary coverage, and to reproduce the observed microstructure of the replaced reactor internal components that showed IASCC and use the results for the PIA of tensile samples. The transmission electron microscope investigations showed that an increase of the annealing temperature causes an average bubble size increase and density decrease. The average He bubble size and distribution in the grain interior and on the grain boundary were similar in the range of temperatures studied. In both cases, bubbles grew by the Ostwald ripening mechanism. No preferential He build-up took place on the grain boundaries. The increase of yield stress produced by He bubbles was calculated with three distinct dislocation-localized obstacle and compared to the tensile test results. The bubble strength estimated for these models was used to assess their validity and to compare the results to published data. Finally, nano-indentation tests in SA, CW, as-implanted SA and PIA (650 to 850°C for 1h) samples did not show He effects on grain boundaries strength. However, the average hardness of the grain interior increased with the He implantation and decreased increasing the PIA temperature.

Spätig, Philippe
Chen, Jia-Chao
Lausanne, EPFL
Other identifiers:
urn: urn:nbn:ch:bel-epfl-thesis8268-6

 Record created 2018-01-09, last modified 2018-03-17

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