SUMMARY OF THE SESSIONS ON IASCC – Chaired by Ulla Ehrnsten and Anna Hojna

Part I of the IASCC session comprised three talks (before the summary/poster presentation session on Tuesday morning).

Kale Stephenson (University of Michigan, USA) presented the paper “New insights into IASCC initiation in neutron irradiated austenitic stainless steel”, with truly new, novel results on IASCC initiation. The work is done as part of his thesis on the subject. He first summarized briefly the main existing observations, e.g., the good correlation between localized deformation (weighted channel height/total elongation) and % intergranular cracking in different stainless steels. The initiation testing was performed (on both, commercial 304L and high-purity alloys, irradiated to 5 – 45 dpa) as interrupted (incremented) 4-point bending tests, using the end pieces from tensile test specimens. Initiation occurred after receiving a certain channel height, and was always associated with inclusions in commercial 304L stainless steel. Dislocation channeling was observed already at a stress of 0.4σy. The high-purity alloys were less susceptible to IASCC initiation. The conclusions from the work are the slip transmission limits initiation, i.e., localized deformation is key, while grain boundary chemical composition cannot explain IASCC susceptibility.

The second paper presented by Kenichi Takakura (NRA, Japan) had the title “Neutron irradiation effect on SCC growth behavior of low carbon austenitic stainless steel in BWR environment”. C(T) specimens, flat tensile specimens and specimens for hardness measurements were cut from the heat-affected zone (HAZ) of a core shroud mock-up made out of 316L stainless steel and irradiated in the Halden test reactor (to about 0.07 – 0.7 dpa). It was concluded that the SCC growth rates in the current HAZ material were between those of non-irradiated and highly irradiated HAZ stainless steels. Crack growth rates increased with increasing K, neutronfluence and hardness.

The third talk (“Stress corrosion crack growth behavior of strain hardened stainless steel evaluated by FRI mmodel calculation”) was held by Masato Koshiishi (NFD, Japan). He compared the behavior of cold-worked and irradiated stainless steel. According to the results and his modelling work IASCC crack growth rate can be calculated if the Rp0.2 and n are known. Cold-worked and irradiated materials behave differently though. In the discussions it was clarified that the change of εf as a function of irradiation is an assumption, and not based on test data, and that the model can be used up to 1.5 dpa at least.

After the coffee break the session continued with four papers.

Yoshiyuki Kaji (JAEA, Japan) showed in his paper with the title “Study on effect of microstructure on crack propagation behavior in austenitic stainless steels“ crack growth rates (CGRs) in A304SS specimens which were irradiated at the Japan Materials Testing Reactor up to 1 or 1.8 dpa at two dose rates, 10-7 or 10-8 dpa/s. EBSD analysis around the crack-tip at specimen center sections was performed and it revealed practically no dose rate effect. Grain Orientation Spread (GOS) maps of grains adjacent to the cracks in the tip area indicating the degree of crystal lattice distortion were compared. Kernel Average Misorientation (KAM) showed clear tendency of microscopic strain localization to grain boundaries with increased dose.

The paper “Crack growth rate and fracture toughness of CF-8 cast austenitic stainless steel” presented by Appajosula Rao (US NRC / ANL, USA) showed results of 0.08 dpa irradiated and thermally aged (400°C for 6000 – 10000 h) materials. Materials of delta ferrite contents of 13 and 23% were studied. Fatigue and SCC CGRs in low DO high-purity water environment were not affected by the irradiation. The fatigue CGR was enhanced in the water compared to air, more in the low delta ferrite material. Fracture toughness values measured in water decreased by about 50% owing to irradiation as well as ageing. The effect of ferrite content on the combined effect of thermal aging and irradiation was inconclusive.

Yasuhiro Chimi (JAEA, Japan) in his talk with the title “Effects of environmental mitigation and water radiolysis on crack growth in simulated BWR environment in highly irradiated 316L stainless steel” referred to results of 12 dpa 316L specimens tested in BWR NWC followed by phases with HWC. Two specimens showed CGRs in NWC of about 10-8 m/s which consequently decrease to about 10-9 m/s when HWC was applied. The third specimen annealed at 700 °C for 1h gave preliminary results showing a very slow CGR of about 3×10-12 m/s, which is similar to CGRs of non-irradiated specimens. If this could be confirmed it would be a remarkable effect of recovery of radiation damage by annealing.

Yusaku Maruno (Hitachi, Japan) and his co-workers investigated the “Radiation effect and electrochemical property in crevice of Ta-modified stainless steel”. It is expected that over-sized element (e.g., Ti, Zr, Hf, V, Nb, Ta) can reduce the IASCC susceptibility of stainless steels. As a first step to show this, radiation-induced segregation of Ta-modified stainless steel of 1 dpa proton irradiation at 300 °C was studied. The Cr concentration on grain boundaries of Ta-modified material was depleted to approximately 15% and that of Type316L was around 10%. At the same time less Ni and no Si enrichment on grain boundaries could be found. Furthermore, it was confirmed via electrochemical testing that the Ta-modified material showed a higher corrosion resistance than Type 316L in simulated BWR condition and 1 ppm H2SO4 environment. It was thought that Ta would form the Ta oxide especially in oxidative environments and perform as an increasing factor for corrosion resistance.

After the lunch break two more presentations completed the IASCC sessions.

Ernest Eason (Modeling & Computing Services, USA) presented a study on the “Feasibility of using the cyclic steps from IASCC tests to model environmentally-enhanced fatigue crack growth in irradiated austenitic stainless steels“. He referred to the evaluation of laboratory test results with a simple substitution model approach. In the pilot study about 1/3 of the heats used were from A316, 316L, 316Ti and the rest were from A304 and 304L. Corrosion fatigue data were plotted against estimated fatigue CGRs in air according to the Jonson equation. The preliminary model relationship and data plots show dose rate dependences towards very slow air rates.

Koji Fukuya (INSS, Japan) gave the last talk in the IASCC session showing results on “APT analysis of stainless steel welds after long-term aging and irradiation”. He referred on a irradiated flux thimble tube (FTT) weld made of type 304 stainless steel and a thermally aged main circulation piping (MCP) weld of type 316L, which were analyzed by APT. Common micro chemical changes were identified in the ferrite phase: Cr decomposition from matrix and G phase formation. The FTT irradiation condition (324°C, 111,000 h, ~3 dpa, 7×10-9 dpa/s) caused no significant enhancement of Cr decomposition but enhancement of G phase growth, considering Mo effects on thermal aging. A very similar G phase was found in the non-irradiated weld which was aged at 320°C for 92000 h.

Please note that there are five more IASCC related papers presented during the summary/poster presentation sessions:

Justin Hesterberg (University of Michigan, USA): “Post-irradiation annealing of a BWR-irradiated 304L stainless steel”

Zhijie Jiao (University of Michigan, USA): “Post-Irradiation annealing in mitigating IASCC of a proton-irradiated 304 stainless steel”

Stephen Raiman (University of Michigan, USA): “Characterization of 316L stainless steel oxidation following irradiation-corrosion in primary water”

Terumitsu Miura (INSS, Japan): “Characterization of surface oxides formed on Fe ion-irradiated stainless steel in high temperature water”

Peng Wang (University of Michigan, USA): “Oxidation of Zircaloy-Th4 in simulated PWR environments during in-situ proton irradiation”

General conclusions by the working group leaders:

Many talks dealt with IASCC crack growth rate testing. The following rather new ideas or observations can be pointed out: (i) IASCC initiation in neutron irradiated steel could be observed, (ii) annealing at 700°C for 1 h for IASCC mitigation, (iii) Ta-modified stainless steel seems to be more resistant to SCC, and (iv) APT is showing similar Ni-Si-Mo clusters in irradiated as well as in aged welds.