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Abstract: Iron-based austenitic alloys are often observed to develop magnetic properties during irradiation, possibly associated with the radiation-induced acceleration of ferrite phase formation. Some of the parametric sensitivities of this phenomenon have been addressed using a series of alloys irradiated in the BOR-60 reactor at 593K. An increase in magnetic phase amount for all alloys was observed over the 0–12dpa dose range. However, magnetic phase (ferrite according to TEM results) did not appear to continuously increase at higher doses (above 12dpa) but did tend to saturate. The formation of a magnetic phase in austenitic stainless steels during irradiation at 593K appeared to be sensitive to alloy composition. It was found that silicon and manganese accelerated ferrite accumulation in the dose range of 0–12dpa, whereas carbon and probably molybdenum resisted it. Also, an increase in grain size resisted ferrite formation, but cold work was found to stimulate it. [Copyright &y& Elsevier]

Abstract: Casting of austenitic stainless steels offers the possibility of directly producing large and/or relatively complex structures, such as the first wall shield modules or the divertor cassette for the ITER, which may lead to simpler component fabrication and major cost savings. Past efforts to use cast steel for these large components were unsuccessful due to lower than acceptable strength in the test components. To improve and validate cast stainless steel as a substitute for wrought stainless steel for shield module applications, a series of test cast steels based on the commercially available CF3M specification have been designed and fabricated. These modifications utilize combinations of Mn and N, which result in significant increases in strength, fracture toughness, and impact properties. These mechanical performance improvements have been achieved without any loss of irradiation performance, corrosion performance, or weldability. [Copyright &y& Elsevier]

Abstract: Zirconium or hafnium additions to austenitic stainless steels caused a reduction in grain boundary Cr depletion after proton irradiations for up to 3dpa at 400°C and 1dpa at 500°C. The predictions of a radiation-induced segregation (RIS) model were also consistent with experiments in showing greater effectiveness of Zr relative to Hf due to a larger binding energy. However, the experiments showed that the effectiveness of the solute additions disappeared above 3dpa at 400°C and above 1dpa at 500°C. The loss of solute effectiveness with increasing dose is attributed to a reduction in the amount of oversized solute from the matrix due to growth of carbide precipitates. Atom probe tomography measurements indicated a reduction in amount of oversized solute in solution as a function of irradiation dose. The observations were supported by diffusion analysis suggesting that significant solute diffusion by the vacancy flux to precipitate surfaces occurs on the time scales of proton irradiations. With a decrease in available solute in solution, improved agreement between the predictions of the RIS model and measurements were consistent with the solute-vacancy trapping process, as the mechanism for enhanced recombination and suppression of RIS. [Copyright &y& Elsevier]

Post-irradiation annealing was used to help identify the role of radiation-induced segregation (RIS) in irradiation-assisted stress corrosion cracking (IASCC) by preferentially removing dislocation loop damage from proton-irradiated austenitic stainless steels while leaving the RIS of major and minor alloying elements largely unchanged. The goal of this study is to better understand the underlying mechanisms of IASCC. Simulations of post-irradiation annealing of RIS and dislocation loop microstructure predicted that dislocation loops would be removed preferentially over RIS due to both thermodynamic and kinetic considerations. To verify the simulation predictions, a series of post-irradiation annealing experiments were performed. Both a high purity 304L (HP-304L) and a commercial purity 304 (CP-304) stainless steel alloy were irradiated with 3.2 MeV protons at 360 °C to doses of 1.0 and 2.5 dpa. Following irradiation, post-irradiation anneals were performed at temperatures ranging from 400 to 650 °C for times between 45 and 90 min. Grain boundary composition was measured using scanning transmission electron microscopy with energy-dispersive spectrometry in both as-irradiated and annealed samples. The dislocation loop population and radiation-induced hardness were also measured in as-irradiated and annealed specimens. At all annealing temperatures above 500 °C, the hardness and dislocation densities decreased with increasing annealing time or temperature much faster than RIS. Annealing at 600 °C for 90 min removed virtually all dislocation loops while leaving RIS virtually unchanged. Cracking susceptibility in the CP-304 alloy was mitigated rapidly during post-irradiation annealing, faster than RIS, dislocation loop density or hardening. That the cracking susceptibility changed while the grain boundary chromium composition remained essentially unchanged indicates that Cr depletion is not the primary determinator for IASCC susceptibility. For the same reason, the visible dislocation microstructure and radiation-induced hardening are also not sufficient to cause IASCC alone. [Copyright &y& Elsevier]

Abstract: Grain boundary engineering (GBE), primarily implemented by thermomechanical processing, is an effective and economical method of enhancing the properties of polycrystalline materials. Among the factors affecting grain boundary character distribution, literature data showed definitive effect of grain size and texture. GBE is more effective for austenitic stainless steels and Ni-base alloys compared to other structural materials of nuclear reactors, such as refractory metals, ferritic and ferritic–martensitic steels, and Zr alloys. GBE has shown beneficial effects on improving the strength, creep strength, and resistance to stress corrosion cracking and oxidation of austenitic stainless steels and Ni-base alloys. [Copyright &y& Elsevier]

The objective of this work is to explore possible thermodynamic correlations between the degradation modes of austenitic and ferritic–martensitic alloys observed in high temperature He–CO–CO2 environments with oxygen and carbon chemical potentials equivalent to that in a liquid sodium environment containing 2–5molppm oxygen and 0.02–0.2molppm carbon at temperatures 500–700°C. Two He–CO–CO2 environments (Pco/Pco2 =1320, Pco=1980molppm, and Pco/Pco2 =9, Pco=13.5molppm) were selected to test alloys NF616 and 316L at 700 and 850°C. Upon exposure to He environments at 850°C, 316L samples exhibited thick surface Cr2O3 scales and substantial internal oxidation; however at 700°C no significant internal oxidation was observed. NF616 samples exhibited relatively thinner surface Cr2O3 scales compared to 316L samples at both temperatures. NF616 samples exposed to liquid sodium at 700°C and He–Pco/Pco2 =9 at 850°C showed decarburization. No surface oxide formation was observed on the sample exposed to the Na environment. Results obtained from He exposure experiments provide insight into what may occur during long exposure times in a sodium environment. [ABSTRACT FROM AUTHOR]