A tensile test is performed to obtain the mechanical property data of the spent fuel cladding. In general, the elastic modulus, elongation, yield stress, tensile stress, etc. are obtained by axial tensile test of cladding attaching an extensometer. However, due to the limitation in the number of specimens for spent nuclear fuel that can be made, the ring tensile test (RTT) whose required length of the specimen is short is mainly performed. In the case of RTT, an extensometer or strain gauge cannot be attached because the gauge part of the specimen is formed around the cladding and is short. In addition, since a load is applied in the radial direction of the cladding, a curved portion of the circular cladding is spread out and becomes straight, and then the cladding is tensioned. For this reason, it is difficult to obtain the stress-strain curve directly from the RTT results. Isight, which is used to identify the optimization design parameters, was used to build an optimization process that minimizes the difference between the RTT and the analysis to estimate the material property. For this, the elastic modulus, plastic strain, and the radius of the RTT jig were taken as fixed variables. As variables, isotropic hardening data and plastic stress were taken. The objective function was taken as the minimization of the area difference of the load-displacement curve obtained from the tests and analysis, of the difference in the magnitude of the maximum reaction force, and of the difference in the location where the maximum reaction force occurred. Optimization workflow was configured in the following order. First, using the calculator component, plastic stress design variables were created. Next, ABAQUS was placed to perform analysis using design variables, and the reaction force or displacement was calculated. After that, the reaction force was calculated considering the 1/4 symmetry condition using the script component. After that, the data matching component performed quantitative comparison of test and analysis data. Finally, by utilizing the exploration component, the plastic stress design variable that minimizes the difference in the objective function was obtained by automatically changing six optimization algorithms. In this paper, the constructed optimization process and the obtained plastic stress by applying it to the SUS316 RTT results are briefly described. The established optimization process can be utilized to obtain mechanical property from the results of the cladding RTT of spent nuclear fuel or new material.

To minimize the short-term thermal load on the repository facility, heat generating nuclides such as Cs-137 and Sr-90 should be separated from the spent nuclear fuel for efficiency of repository facility. In particular, Sr-90 must be separated because it generates high heat during the decay process. Recently, Korea Atomic Energy Research Institute (KEARI) is developing a waste burden minimization technology to reduce the environmental burden caused by the disposal of spent nuclear fuel and maximize the utilization of the disposal facility. The technology includes a nuclide management process that can maximize disposal efficiency by selectively separating and collecting major nuclides such as Cs, Sr, I, TRU/RE, and Tc/Se from spent nuclear fuel. Among the major nuclides, Sr nuclides dissolve in chloride phase during the chlorination process of spent nuclear fuel and recovered in the form of carbonate or oxide via reactive distillation. In this process, Ba nuclides are also recovered along with Sr nuclides due to their chemical similarity. In this study, we prepared group II nuclide ceramic waste form, Ba(x)Sr(1-x)TiO3 (x=0, 0.25, 0.5, 0.75, 1), using the solid-state reaction method by considering the various ratio of Sr/Ba nuclides generated from nuclide management process. The established waste form fabrication process was able to produce a stable waste form regardless of the ratio of Sr/Ba nuclides. To evaluate the stability of group II waste form, physicochemical properties such as leaching and thermal properties were evaluated. Also, the radiological properties of the Ba(x)Sr(1-x)TiO3 waste forms with various Sr/Ba ratios were evaluated, and the estimation of centerline temperature was carried out using the experimental thermal property data. These results provided fundamental data for long-term storage and management of group II nuclides waste form.