A methodology is under development to restore and predict the long-term evolution of the natural barrier comprise the site of radioactive waste disposal for surface geological outcrop, tunnel face and drill core. Considering the condition that the radioactive waste repository should be located in the deep part, the drill core is an important subject that can identify deep geological properties that could not be confirmed near the surface. In this study, we investigated proper age dating methods to construct lithological model of the disposal site with regard to the long-term safety. Also, preliminary age dating locations were selected using the lithological distribution results by depth through geochemical and micro-structural analysis for the deep drill cores excavated around KURT. In the study area, the dikes presumed the Cretaceous were intruded by Jurassic granites. As for the granotoids, U-Pb age dating for zircon, which is resistant to deformation or metamorphism and has loss, is often used. In the case of the dikes, K-Ar and 40Ar/39Ar age dating for the argon captured in the rocks after magmatism is often used. Through U-Pb zircon ages of KURT site granotoids, we expect to solve the clustering problem (granite and granodiorite), which is different from precious chemical analysis (XRF) results and TAS-diagrams. 40Ar/39Ar age dating to be used for the dikes is suitable for the perspective of lithological model of the disposal site. Because, it can compensate for accuracy problems such as sample heterogeneity in K-Ar age dating and is used for volcanic rocks. In the further study, we plan to determine the appropriate sampling locations by the selected age dating methods from the perspective of disposal in this study.

Disposal facilities for radioactive waste shall be sited to provide isolation from the accessible biosphere. The features shall aim to provide this isolation for tens of thousands to a million years after closure. For the safety assessments of repository, the long-term natural evolution and possible events of the site, that can cause disturbances to the facility over the period of interest, should be considered. Geological development processes that the site have been experienced can contribute to understanding and descripting the present-day conditions. Moreover, knowledge of the past is necessary to predict the future evolution of the site. With regard to disposal site, understanding past geological evolution history allows to access the possibility of hazardous events of the site that can cause disturbances to the facility over the period of interest, and to verify the change in the geological environment is within the safe performance range even after the period of interest. In addition, certain parameters that change with the geological evolution can affect the hydrological and geochemical characteristics which are essential to disposal performance. There are various factors in the evolution of the geological environment, but not all are related to disposal safety. The objective of this research is to develop a geological reconstruction method considering factors that should be derived preferentially for the geological characteristics of the disposal site and the evaluation of the long-term safety. As a preliminary study on this, we investigated case studies related to geological reconstruction of overseas disposal research institutes, and reviewed which factors are suitable for the domestic granitoid distribution environment. It is expected that systematic and consistent results will be possible in the future through this methodology.

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.

Laboratory testing to simulate the drying of spent fuel is most often done using a cooling rate of approximately 5°C per hour because there are so many restricted test conditions like R&D project duration limit, budget and temporary electronic supply blackout at laboratory building. However, in a real dry cask storage system, the fuel cools much slower. Early data from KAERI on unirradiated, pre-hydrided cladding has shown that slower cooling may result in more brittle behavior than is currently observed based on these short-term tests. Given the potential safety and future handling implications of failed fuel, it is important to determine if the material properties of spent fuel cladding measured in these laboratory tests are the same as would be observed on fuel that has undergone a much longer, slower cooling, which may provide more time for hydrides to precipitate in the radial direction. KAERI and PNNL have started a collaborative I-NERI R&D project on this topic and each organization will perform tests on unirradiated & irradiated cladding under various hoop stress and cooling rate combinations. Scope of collaborative work is to evaluate long-term cooling (slow cooling rate) on hydride reorientation and subsequent material properties of cladding to determine if past and current research activities on spent nuclear fuel are bounding. The results will be used to direct future testing and help predict cladding performance over a wide range of burnups during extended storage and transportation.

The manufactured nuclear fuel assembly is loaded into the nuclear reactor after the core design, and is finally discharged to the wet storage pool after depletion for 3 cycles. The discharged spent nuclear fuel is transported and stored in a dry storage system at the on-site of the nuclear power plant, which is cooled by natural convection, and undergoes final disposal or reprocessing through an intermediate dry storage facility. In this series of processes, the characteristics of the final product, the spent fuel, vary depending on the environmental conditions, so it is essential to manage each history data to verify the long-term integrity of the spent nuclear fuel. In this paper, safety information on spent nuclear fuel is described in order to establish technical requirements that should be considered in each stage of storage, transport, reprocessing, and disposal of spent nuclear fuel. Comprehensive safety information on spent nuclear fuel is basically calculated from basic information that considers characteristic information that can be obtained through the manufacture and design of nuclear fuel assemblies, operation history in a nuclear reactor, and location history in a wet storage pool. It can be divided into secondary production information (SF Burnup, Nuclide Inventory, etc.) and tertiary integrity-related information obtained through cladding inspection during spent fuel storage. KHNP produces this multi-layered information according to the production stage and manages it through the comprehensive management system of the spent nuclear fuel, and safety information with some errors is not only improved through re-verification but also continuously updated. In this paper, the spent nuclear fuel safety information was derived based on various information calculated in the entire process of being discharged and managed in a wet storage pool, including new fuel manufacturing information and depletion history. Such safety information will be used as basic data for long-term safe management of spent nuclear fuel, and will be continuously produced and managed. In the future, additional discussions will be held on the safety information of the spent nuclear fuel through consultation with KORAD and regulatory agencies.

A long-term cooling effect on hydride reorientation of a cladding tube can affect the integrity of spent nuclear fuel transportation and long-term storage. In this study, experimental setup for investigating the degree of radial reorientation of hydrides in the circumferential direction during the long-term cooling was established. The experimental setup was designed to be simplified since the long-term evaluation requires a long term period such as 12, 18 and 24 months when the cladding tube specimen is gradually cooled down from 400°C to 100°C. For the test, hydrogen-charged specimens of 100 ppm, 200 ppm, and 500 ppm were prepared. The specimen was sealed with fixtures and check valve, and was pressurized up to 90 Mpa. To heat the specimen, a box-type furnace was used while the temperature of the specimen was measured from thermocouples attached to the specimen. After the heat treatment, the long-term cooling was performed by developing temperature control program to investigate several cooling rate conditions of the specimen. As a reference case, microstructure and brittle property of the hydrogen-charged specimens of 100 ppm, 200 ppm, and 500 ppm without the long-term cooling was observed. In the case of the hydrogen content, it was uniformly distributed in circumferential direction although it was non-uniform in the axial direction. In the case of the brittle property, a compression test was performed. For the future work, the microstructure and brittle property of the hydrogencharged specimens after the several long-cooling conditions were investigated. Then, the degree of radial reorientation of hydrides in the circumferential direction during the long-term cooling was studied.

In general, if a nuclear fuel cladding tube is damaged during reactor operation, it is called fuel failure. If the cladding tube is damaged, the function of sealing the nuclear fuel material is lost, and the fission products accumulated inside the nuclear fuel rod may leak into the coolant. The causes are the most damage caused by foreign substances in a coolant such as small iron wires, and GTRF (Gridto- Rod Wear) due to a grid, end-plug welding defect, PCMI (pellet cladding mechanical interaction), and oxidation corrosion damage. In this study, a device of simulating friction damage and debris induced damage between grid-fuel rods, which are the main causes of cladding tube damage, was developed. An air vibrator was installed as a function to induce vibration of the nuclear fuel rod. Sandpaper was installed between the grid and the fuel rod to induce friction between the grid-fuel rods. Saw teeth were installed on the grid to induce damage to foreign substances. It is believed that the simulated damaged nuclear fuel rod can be manufactured through on-study to provide the simulated damaged nuclear fuel rod necessary for the stabilization study of the damaged nuclear fuel rod.

In South Korea, the master plan for high-level radioactive waste management, announced in 2016, suggested the construction and operation of intermediate storage facilities on a permanent disposal site and specified the adoption of dry storage in consideration of the ease of operation and expansion. As of 2021, the government is again reviewing its overarching policy on the back-end fuel cycles, including intermediate storage and permanent disposal. In the case of dry storage facilities, safety evaluation is being conducted using a combination of deterministic and probabilistic approaches, similar to that of nuclear power plants. The two methods are complementary, of which Probabilistic Safety Assessment (PSA) has the advantage of being able to identify key scenarios affecting safety, but its use in storage facilities has not been highlighted so far. However, depending on the spent fuel management phases such as loading, transportation, and storage, it may be not enough to capture effective and efficient safety evaluation only deterministically, and probabilistic methods may contribute to the evaluation of long-term operation or external events such as an earthquake. There have already been cases where PSA has been performed on a part of the nuclear fuel cycle through previous studies. This paper created the safety assessment model based on open sources such as the released EPRI reports, by targeting arbitrary intermediate storage facilities. The model considered the scenarios for loading, transportation, and storage, with human error respectively. It will be able to be modified and improved to fit domestic and specific intermediate storage facilities in the future.

The WHO reported the Covid-19 outbreak infected 486,761,597 people, involving 6,142,735 deaths worldwide as of 1 April 2022. This contagious disease has spread rapidly throughout the world, including Malaysia. Since the outbreak in Malaysia began in March 2020, the Movement Control Order (MCO) has been implemented nationwide, leaving a significant impact on its citizens, non-citizens, as well as refugees. There is some exploitation of refugees, where enforcement officers are targeting them for criminal offences. Stakeholders claimed the Malaysian government did not provide any assistance to refugees during the pandemic, including health care and economy. This article examines Malaysia’s responsibilities as a host country to refugees during the Covid-19 outbreak. The Malaysian government is proposed to continuously support refugees on humanitarian grounds based on the country’s economic development capabilities. This paper will look into the current situation of the Refugees in Malaysia; discuss the challenges that the Refugees in Malaysia are facing; analyse the legal framework governing the status of refugees; and check the responsibility Malaysia should assume as a host country.