Lead–bismuth eutectic (LBE), a heavy liquid metal, is an ideal candidate coolant material for Generation-IV fast reactors and accelerator-driven systems (ADSs), but LBE is also known to pose a considerable corrosive threat to its container. However, the susceptibility of the candidate container material, 316L stainless steel (SS), to flow-accelerated corrosion (FAC) under turbulent LBE flow, is not well understood. In this study, an LBE loop, referred to as JLBL-1, was used to experimentally study the behavior of 316L SS when subjected to FAC for 3000 h under non-isothermal conditions. An orificed tube specimen, consisting of a straight tube that abruptly narrows and widens at each end, was installed in the loop. The specimen temperature was 450 °C, and a temperature difference between the hottest and coldest legs of the loop was 100 °C. The oxygen concentration in the LBE was lower than 10⁻⁸ wt %. The Reynolds number in the test specimen was approximately 5 × 10⁴. The effects of various hydrodynamic parameters on FAC behavior were studied with the assistance of computational fluid dynamics (CFD) analyses, and then a mass transfer study was performed by integrating a corrosion model into the CFD analyses. The results show that the local turbulence level affects the mass concentration distribution in the near-wall region, and therefore, the mass transfer coefficient across the solid/liquid interface. The corrosion depth was predicted on the basis of the mass transfer coefficient obtained in the numerical simulation and was compared with that obtained in the loop. For the abrupt narrow part, the predicted corrosion depth was comparable with the measured corrosion depth, as was the abrupt wide part after involving the wall roughness effects in the prediction; for the straight tube part, the predicted corrosion depth is about 1.3–3.5 times the average experimental corrosion depth, and the possible reason for this discrepancy was provided.

The ductile properties of irradiated materials are among of the important indicators related to their structural integrity. These properties are generally determined by performing tensile tests on irradiated materials in the irradiation environment. Indentation tests are used for evaluating ductile properties easily and rapidly. Constants in the Swift-type material constitutive equation were identified via inverse analysis using the Kalman filter, such that the numerical experimental results reproduced the indentation test results. Numerical tensile experiments were conducted using stress–strain curves with the identified constants to obtain nominal stress and strain curves. The identified yield stress, work hardening coefficient, and exponent were 200–1000 MPa, 1100–1500 Ma, and 0.5–0.7, respectively. Furthermore, two methods were proposed for evaluating the total elongation. Method I was used to calculate the total elongation based on the relationship between the total and uniform elongations obtained from the tensile tests performed on irradiated materials. Method II was used to determine the total elongation from the ductile failure criterion based on the relationship between the stress and strain states in the tensile specimen model using the numerical tensile experiment and failure strain evaluated from actual tensile experiments. Evaluated minimum total elongation was 10%. The evaluation results for ion-irradiated materials were similar to the tensile test results for irradiated materials.