To investigate the mechanical integrity of spent nuclear fuel, the failure behavior of the cladding tube was examined under accident conditions. According to the SNL report, the failure behavior of cladding can be broadly classified into two types. The first is failure due to bending load caused by falling. The second is failure due to pinch load caused by space grid. In this study, mechanical integrity was evaluated through the stress intensity factor applied to the crack in failure behavior due to bending load. Since the exact value of the impact load due to fall was unknown, the load was applied by increasing the value up to 200 G in 20 G increments. The size of the crack is an important input variable, and 300 um was given by referring to the EPRI report, and the elastic modulus, a material property that determines the stress field, was given 75.22 GPa by referring to the FRAPCON code. Since the relationship between the direction of stress and the direction of the crack is also a major variable, simulations were conducted for both cracks perpendicular to and parallel to the stress direction. It was confirmed that at a load of 200 G, when the crack was parallel to the stress direction, stress concentration did not occur and had a very low stress intensity factor 0.01 􀜯􀜲􀜽√􀝉. When perpendicular to the direction of stress, the stress intensity factor showed a value of 1 􀜯􀜲􀜽√􀝉. However, considering that the critical value of the stress intensity factor due to hydride is 5 􀜯􀜲􀜽√􀝉, it can be seen that perpendicular result also ensures the mechanical integrity of the cladding.

Research on the safety of nuclear spent fuel has been heavily experimented and modelled from a mechanical perspective. The issues of corrosion, irradiation creep, hydride and hydrogen embrittlement have been addressed more than two decades since the early 2000s. Among these degradation behavior, hydrogen embrittlement and hydride reorientation have been the most important topics for establishing the integrity of nuclear spent fuel and have been studied in depth. In order to assess the safety of spent nuclear fuel, firstly, it is necessary to establish the safety criteria in all nuclear cycle, i.e., the failure criteria guidelines for nuclear fuel assemblies and nuclear fuel rods, and then examine the safety analysis. The contents of U.S.NRC Regulations, Title 10 General, Chapter 1 Code of Federal Regulation (CFR), Part 50, 71 and 72, describe the safety criteria for the safety assessment of nuclear fuel assemblies and nuclear fuel rods. In this study, technically important points in safety analysis on nuclear fuel are checked through the reference of those NRC regulation. As result, we confirmed that the safety assessment of nuclear fuel after 20 years of interim storage is now being tested by ORNL and PNNL. There are not quantitative criteria related to material safety. However qualitative criteria which is dependent on environmentally condition describe the safety analysis. There is some literature study about DBTT, yield stress, ultimate tensile strength, flexural rigidity data. In FRAPCON code Modelling of yield strength and creep had been established, but radial hydride or hydride reorientation has not considered.

The conventional research trend on spent fuel was safety analysis based on mechanical perspective. Analysis of spent fuel cladding is based on the temperature of cladding and pressure inside cladding. To improve fuel cladding analysis, precise and accurate thermal safety evaluation is required. In this study a database which is about thermal conductivity and emissivity for the thermal modeling was established for a long-term safety analysis of spent fuel. As a result, we confirmed that the thermal conductivity of zirconium hydride was not accounted in conventional model such as FRAPCON and MATPRO. The conductivity of zirconium and its oxide was evaluated only as a function of temperature. However, the behavior of heat conductivity and emissivity is determined by the change of the material properties. The material properties depend on the microstructural characteristic. It can be seen that this conventional approach does not consider the microstructure change behavior according to vacuum drying process or burn-up induced degradation phenomena. To improve the thermal properties of spent nuclear fuel cladding, the measurement experiments of heat conduction and emissivity are required according to spent fuel experience and status such as the number of vacuum drying, cooling rate, burn up, hydrogen concentration and oxidation degree. In previous domestic reports and papers, we found that relative data between thermal properties and spent fuel experience and status does not exist. Recently, in order to understand the failure mechanism of hydrogen embrittlement, many studies have been conducted by accounting and spent fuel experience and status in a mechanical perspective. If microstructure information could be obtained from these studies, the modeling of thermal conductivity and emissivity will be possible indirectly. According to a recent abroad paper, it was confirmed that the thermal conductivity decreased by about 30% due to irradiation damage. The radiation damage effects on thermal conductivity also has not been studied in zirconium oxide and hydride. These un-revealed phenomena will be considered for the thermal safety model of spent fuel.

In order to treat groundwater containing high levels of nitrate, nitrate reduction by nano sized zero-valent iron (nZVI) was studied using batch experiments. Compared to nitrate removal efficiencies at different mass ratios of nitrate/Fe0, the removal efficiency at the mass ratio of 0.02% was the highest(54.59%). To enhance nitrate removal efficiency, surface modification of nZVI was performed using metallic catalysis such as Pd, Ni and Cu. Nitrate removal efficiency by Cu-nZVI (at catalyst/Fe0 mass ratio of 0.1%) was 66.34%. It showed that the removal efficiency of Cu-nZVI was greater than that of the other catalysts. The observed rate constant (kobs) of nitrate reduction by Cu-nZVI was estimated to 0.7501 min-1 at the Cu/Fe mass ratio of 0.1%. On the other hand, TEM images showed that the average particle sizes of synthetic nZVI and Cu-nZVI were 40~60 and 80~100 nm, respectively. The results imply that catalyst effects may be more important than particle size effects in the enhancement of nitrate reduction by nZVI.