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.

To dry storage of spent nuclear fuel withdrawn the wet storage, all moisture inside the dry storage container must be removed to ensure the long-term integrity and retrievability. Substantial amounts of residual water in dry storage container may have potential impacts on the fuel, cladding, and other components in the dry storage system, such as fuel degradation and cladding corrosion, embrittlement, and breaching. The drying could perform as a vacuum drying process or a forced helium dehydration process. In NUREG-1536, the evacuation of most water contained within the canister is recommended a pressure of 0.4 kPa (3 torr) to be held in the canister for at least 30 minutes while isolated from active vacuum pumping as a measure of sufficient dryness in the canister. Monitoring the moisture content in gas removed from the canister is considered as a means of evaluating adequate dryness. Dew point monitoring and special techniques could be used to evaluate this adequacy. Various studies are continuing for quantitative evaluation of residual moisture inside the dry storage system. Andrawes proposed a methodology for determining trace water contents in gaseous mixtures, utilizing gas chromatography together with a helium ionization source. A microwave plasma source and emission spectrometry were utilized to determine trace amounts of bound water in solid samples using peak areas of atomic oxygen (O) and hydrogen (H) emissions. Bryans measured the gas samples taken from the High Burn-Up Demonstration Cask at three intervals: 5 hours, 5 days, and 12 days after the completion of drying and backfilling in the North Anna power Station. To measure water content, a Vaisala humidity probe was used. Final results indicated that the cask gas water content built up over 12 days to a value of 17,400 ppmv ±10%, equivalent to approximately 100 g of water within the entire cask gas phase. Tahiyats also proposed a methodology that involves a direct current (dc) driven plasma discharge and optical emission spectroscopy for detecting and quantifying water vapor in a flowing gas stream under both trace and high water vapor loading conditions. For detecting water vapor concentration, the emission from H at 656.2 nm was employed. The H emission is the red visible spectral line generated by a hydrogen atom when an electron falls from the third lowest to the second lowest energy level, this suggests that the normalized H intensity can be used as a marker for water vapor detection and quantification. Several of the attempts are continuing to quantify water contents in dry storage system. Lessons learned by Case studies would be provided insights into how to improve future measurements.

In order to construct and operate the dry storage systems, it is essential to confirm the safety of the systems through safety analysis. If the dry storage cask is damaged due to an accident, a large amount of radioactive material may be leaked to the outside and cause radiation exposure to surrounding workers and nearby public, so the effect thereof should be evaluated. Many input parameter are required in the confinement evaluation for accident condition, and in this study, the change in the confinement evaluation result according to the change of major input parameter is to be studied. In this study, we selected fractions of radioactive materials available for release from spent fuel, cooling time, and distance to exclusive area boundary as the major input parameter. In general, the release fraction suggested by NUREG-1536 has been used, but NUREG-2224 provides the fraction for high burn-up spent fuel in fire and impact accident conditions, unlike NUREG-1536 which provide a single value. In the case of the distance to exclusive area boundary, 100 to 800 m was considered, and in the case of the cooling time, 10 to 50 years was considered in this study. In order to compare the dose change by the parameter, we set up the hypothetical storage system. A storage cask of the system contain 21 PWR spent fuel assemblies with an initial enrichment of 4.5wt%, burnup of 45,000 MWD/MTU. During the accident condition, it is assumed that the cask is leaked at 1.0×10−7cm3·sec−1. Since the main dose criterion for accident conditions is 50 mSv of effective dose, effective doses are calculated in this study. In an accident condition, transuranic particulate contribute most of the doses, so the doses are determined according to the fraction for the particulate. Therefore, it was confirmed that the dose was almost the same as the fraction for the accident conditions in NUREG-1536 and the fraction for the impact accident conditions in NUREG-2224 is 3×10−5, but the dose was also 100 times higher as the fraction for the fire accident conditions in NUREG-2224 is 3×10−3. In the case of the cooling time, it was confirmed that the dose change according to the cooling time was not significant because the dose contribution of transuranic elements having very long half-life was very large. In the case of the distance, it was confirmed that the dose decreased exponentially as the atmospheric dispersion factor decreased exponentially with the distance.

The spent fuel dry storage canister is generally made of austenitic stainless-steel and has the role of an important barrier to encapsulate spent fuels and radioactive materials. The canister on the dry storage system has several welding lines in the wall and lid, which have high residual tensile stresses after welding procedure. Interaction between stainless steel and chloride environment from a sea results in an aged-related degradation phenomenon causing chloride induced stress corrosion cracking (CISCC) in the dry storage system. A pending issue to the interim storage of spent fuel awaiting repository disposal is their susceptibility to CISCC of stainless steel canisters. The available mitigation technology should be studied sufficiently to prevent the degradation phenomenon. This paper assesses stress-based mitigation to control residual tensile stress practically applicable to the atmospheric CISCC for the aging management of the stainless steel canisters. There are major components, that is, elevated tensile stress, susceptible material and corrosive environment that must be simultaneously present for CISCC degradation to occur. Surface stress improvement can effectively mitigate the potential for CISCC of the canister external surfaces. The potential deleterious effect of the additional work is negated by the presence of compressive residual stress, which removes the tensile stress needed for CISCC to occur. Surface stress improvement methods such as shock-based peening, shot peening and low plasticity burnishing can be applied for surface stress improvement of the canisters. Stress relaxation processes and advanced welding methods such as laser beam welding and friction stir welding can be also available to mitigate the susceptibility to CISCC. As the result assessing the stress-based mitigation technologies, promising candidate methods could be selected to reduce the residual tensile stresses and to control an aged-related degradation condition causing CISCC in the spent fuel dry storage canister.

Concrete structures of spent nuclear fuel interim storage facility should maintain their shielding ability and structural integrity during normal, off-normal and accident conditions. The concrete structures may deteriorate if the interim storage facility operates for more than several decades. Even if deterioration occurs, the concrete structures must maintain its unique functions (shielding and structural integrity). Therefore, it is necessary to establish an analysis methodology that can evaluate whether the deteriorated concrete structure maintains its integrity under not only normal or off-normal condition but also accident condition. In accident conditions such as tip over and aircraft collision, both static material properties and dynamic properties of the concrete are required to evaluate the structural integrity of the concrete structures. Unlike the calculated damage results for the static deformation of the concrete structure, it is very difficult to accurately estimate the damage values of the degraded concrete structures where an aircraft collides at a high strain rate. Therefore, the present authors have a plan to establish a database of the dynamic material properties of deteriorated concrete and implement to a Finite Element Analysis model. Prior to that, dynamic increase factors described in a few technical specifications were investigated. The dynamic increase factor represents the ratio of the dynamic to static strength and is normally reported as function of strain rate. In ACI-349, only the strain rate is used as a variable in the empirical formula obtained from the test results of specified concrete strengths of 28 to 42 MPa. The maximum value of dynamic increase factor is limited to 1.25 in the axial direction and 1.10 in the shear direction. On the other hand, in the case of the CEB model, static strength is included as variables in addition to the strain rate, and a constitutive equation in which the slope changes from the strain rate of 30 /s is proposed. As plotting the two dynamic increase factor models, in the case of ACI, it is drawn as a single line, but in the case of CEB, it is plotted as multiple lines depending on the static strength. The test methods and specimen sizes of the previously performed tests, which measured the concrete dynamic properties, were also investigated. When the strain rate is less than 10 /s, hydraulic or drop hammer machines were generally used and the length of the specimens was more than twice the diameter in most cases. However, in the case of Split Hopkinson Pressure Bar tests, the small size specimens are preferred to minimize the inertia effect, so the specimens were small and the length was less than twice the diameter. We will construct the dynamic properties DB with our planned deteriorate concrete specimen test, and also include the dynamic property data already built in the previous studies.

Currently, the interim storage pools of spent fuels in South Korea are expected to become saturated from 2024. It is required to prepare an operation plan of a domestic dry storage facility during a long-term period, with the researches on safety evaluation methods. This study modified the FRAPCON code to predict the spent fuel integrity evaluation such as the axial cladding temperature, the hoop stress and hydrogen distribution in dry storage. The cladding temperature in dry storage was calculated using the COBRA-SFS code with the burnup information which was calculated using the FRAPCON code. The hoop stress was calculated using the ideal gas equation with spent fuel information such as rod internal pressure. Numerical analysis method was used to calculate the degree of hydrogen diffusion according to the hydrogen concentration and temperature distribution during a dry storage period. Before 50 years of dry storage, the cladding temperature and hoop stress decreased rapidly. However, after 50 years, they decreased gradually and the cladding temperature was below 400 K. The initial temperature distribution and hydrogen concentration showed a parabolic line, but hydrogen was transferred by the hydrogen concentration and temperature gradient over time.