Zhijie Jiao

Oxidation of a Proton-Irradiated 316 Stainless Steel in Simulated BWR NWC Environment

A proton-irradiated SUS316 stainless steel was exposed to the simulated BWR NWC environment for 70 hours during a constant extension rate tensile test and the resulted oxide film was examined using transmission electron microscopy. The oxide film on both the unirradiated and irradiated parts of the sample consists of an outer layer of hematite particles and an inner layer of (Fe, Cr, Ni)3O4 spinel. Formation of hematite under BWR NWC condition is consistent with the predication by the potential-pH diagram. Both the outer layer and the inner layer of the oxide film show a strong dependence on grain orientation. Some grains exhibit an inner layer thickness of 40–100 nm while some others have barely any oxidation. Persistent damage induced by proton irradiation did not show a strong influence on the oxidation process as the thickness structure and compositions of the oxide film on both the unirradiated and irradiated parts of the sample were very similar.

IASCC Susceptibility of 304L Stainless Steel Irradiated in a BWR and Subjected to Post Irradiation Annealing

Post-irradiation annealing (PIA) was conducted to investigate the cause of irradiation-assisted stress corrosion cracking (IASCC). The effects of PIA on irradiation hardening, dislocation channel formation, and IASCC susceptibility were examined for a 304L stainless steel irradiated to 5.9 dpa in the Barseback-1 reactor (Sweden). The annealing treatments were performed at temperatures in the range 450–600 °C and times ranging from 1–20 h. Longer annealing times and higher temperatures, as represented by iron diffusion distance, resulted in a significant reduction in irradiation hardening. IASCC susceptibility was measured for the as-irradiated and two PIA conditions (500 °C: 1 h and 550 °C: 20 h) via interrupted CERT tests under simulated BWR-NWC conditions. The annealing treatments progressively reduced IASCC susceptibility (as measured by the final intergranular fracture fraction) and dislocation channel density.

Solute Clustering in As-irradiated and Post-irradiation-Annealed 304 Stainless Steel

A commercial purity 304SS was irradiated to 5 dpa Kinchin-Pease (10 dpa full-cascade) using 2 meV protons at 360 °C. Post-irradiation annealing (PIA) was applied to reduce or remove IASCC susceptibility. This paper focuses on the links between irradiation-induced hardening and irradiated microstructures of the as-irradiated and PIA conditions; the irradiated microstructure is assessed by transmission electron microscopy (TEM) and atom probe tomography (APT). Dislocation loops, Ni–Si clusters, and Cu-enriched clusters are present in the as-irradiated condition. When the dislocation loops are removed by PIA, ~40% of the as-irradiated hardness remains and can be rationally attributed to the solute clusters still present in the PIA microstructure. The observations indicate that hardening in the as-irradiated condition is controlled by both dislocation loops and solute clusters and suggest that radiation-induced solute clusters may be important to detailed understanding of IASCC (irradiation-assisted stress corrosion cracking).

Radiation-Induced Precipitates in a Self-ion Irradiated Cold-Worked 316 Austenitic Stainless Steel Used for PWR Baffle-Bolts

5 meV Ni++ and Fe++ ion irradiations were performed to investigate radiation-induced precipitates evolution in a cold-worked 316 austenitic stainless steel at high doses and temperatures. The irradiation conditions were 23 dpa at 380 °C, 130 dpa at 380 °C, 23 dpa at 500 °C, and 15 dpa at 600 °C. TEM selected electron diffraction (SAED), TEM dark-field imaging and energy dispersive spectroscopy (EDS) mapping were used as complementary techniques to determine crystallography, morphology and chemical composition of radiation-induced precipitates. The precipitates were predominantly in form of the Ni–Si rich γ′ phase at all irradiation conditions. The EDS analysis further determined Ni–Si–Mo–P and Ni–Si–Mn rich precipitates after irradiation at 380 and 600 °C, respectively. The precipitates were found close to saturated state between 23 and 130 dpa at 380 °C irradiation conditions. A different effect of higher irradiation temperatures was found between 500 and 600 °C. In case of the irradiation to 23 dpa at 500 °C, the average size of precipitates was similar to irradiations at 380 °C, but the density was lower. However, the precipitates revealed large size and very low density following the irradiation to 15 dpa at 600 °C. The original dislocation network introduced by cold-working was found as dominant sink for intra-granular solute radiation-induced segregation (RIS) and possibly took place as primary nucleation site of radiation-induced precipitates at irradiation temperatures 380 and 500 °C. At the temperature 600 °C, the RIS at dislocation network almost vanished and the main nucleation sites became twin boundaries as more energetically favorable intra-granular sinks.

Microstructural Evolution of Self-Ion Irradiation HT9

Ferritic/martensitic steels are candidates for fast reactors because of their sodium compatibility, superior resistance to corrosion and radiation damage, including swelling, and excellent thermal conductivity and thermal expansion coefficient. One significant limitation of any cladding material is its susceptibility to swelling at high doses. While HT9 has neutron irradiation performance data up to ∼200 dpa, dose requirements for the Traveling Wave Reactor (TWR) may be much higher. Obtaining higher-dose data will take many years, but in the interim, heavy ion irradiation could provide a useful tool toward predicting the swelling trends beyond 200 dpa. In this study, HT9 was irradiated from 440–480°C using 5 MeV Fe++ ions. The samples are compared to a portion of HT9 fuel assembly duct from FFTF, which was characterized after neutron irradiation at 440°C with an accumulated dose of 155 dpa. Comparisons are made of the void size and density using transmission electron microscopy (TEM). The increase in dose from 280 dpa to 375 dpa increased void size, number density and swelling at 440°C, while swelling was generally lower at 480°C for the same helium pre-implantation conditions. Helium generally enhanced the nucleation of voids, as measured by the void density.Copyright

Microstructure and Property Evolution in Advanced Cladding and Duct Materials Under Long-Term Irradiation at Elevated Temperature: Critical Experiments

The in-service degradation of reactor core materials is related to underlying changes in the irradiated microstructure. During reactor operation, structural components and cladding experience displacement of atoms by collisions with neutrons at temperatures at which the radiation-induced defects are mobile, leading to microstructure evolution under irradiation that can degrade material properties. At the doses and temperatures relevant to fast reactor operation, the microstructure evolves by microchemistry changes due to radiation-induced segregation, dislocation loop formation and growth, radiation induced precipitation, destabilization of the existing precipitate structure, as well as the possibility for void formation and growth. These processes do not occur independently; rather, their evolution is highly interlinked. Radiation-induced segregation of Cr and existing chromium carbide coverage in irradiated alloy T91 track each other closely. The radiation-induced precipitation of Ni-Si precipitates and RIS of Ni and Si in alloys T91 and HCM12A are likely related. Neither the evolution of these processes nor their coupling is understood under the conditions required for materials performance in fast reactors (temperature range 300-600°C and doses to 200 dpa and beyond). Further, predictive modeling is not yet possible, as models for microstructure evolution must be developed along with experiments to characterize these key processes and provide tools for extrapolation. To extend the range of operation of nuclear fuel cladding and structural materials in advanced nuclear energy and transmutation systems to that required for the fast reactor, the irradiation-induced evolution of the microstructure, microchemistry, and the associated mechanical properties at relevant temperatures and doses must be understood. This project builds upon joint work at the proposing institutions, under a NERI-C program that is scheduled to end in September, to understand the effects of radiation on these important materials. The objective of this project is to conduct critical experiments to understand the evolution of microstructural and microchemical features (loops, voids, precipitates, and segregation) and mechanical properties (hardening and creep) under high temperature and full dose range radiation, including the effect of differences in the initial material composition and microstructure on the microstructural response, including key questions related to saturation of the microstructure at high doses and temperatures.