As regulations on carbon emissions increase, the interest in renewable energy is also increasing. However, the efficiency of renewable energy generation is highly low and has limitations in replacing existing energy consumption. In terms of this view, nuclear power generation is highlighted because it has the advantage of not emitting carbon. And accordingly, the amount of spent nuclear fuel is going to increase naturally in the future. Therefore, it will be important to obtain the reliability of containers for transporting safely and storing spent nuclear fuel. In this study, a method for verifying the integrity and airtightness of a metal cask for the safe transportation and storage of spent nuclear fuel was studied. Non-destructive testing, thermal stability, leakage stability, and neutron shielding were demonstrated, and as a result, suitable quality for loading spent nuclear fuel could be obtained. Furthermore, it is meaningful in that it has secured manufacturing technology that can be directly applied to industrial field by verifying actual products.

Given the domestic situation, all nuclear power plants are located at the seaside, where interim storage sites are also likely to be located and maritime transportation is considered inevitable. Currently, Korea does not have an independently developed maritime transportation risk assessment code, and no research has been conducted to evaluate the release rate of radioactive waste from a submerged transportation cask in the sea. Therefore, secure technology is necessary to assess the impact of immersion accidents and establish a regulatory framework to assess, mitigate, and prevent maritime transportation accidents causing serious radiological consequences. The flow rate through a gap in a containment boundary should be calculated to determine the accurate release rate of radionuclides. The fluid flow through the micro-scale gap can be evaluated by combining the flow inside and outside the transportation cask. In this study, detailed computational fluid dynamic and simplified models are constructed to evaluate the internal flow in a transportation cask and to capture the flow and heat transfer around the transportation cask in the sea, respectively. In the future, fluid flow through the gap will be evaluated by coupling the models developed in this study.

For safe management of spent nuclear fuels, they should be delivered to repository or waste disposal site. As the amount of spent nuclear fuel transportation is expected to increase in the future due to the provision of an intermediate storage facility, the necessity to secure transportation cask is emerging. In order to secure the spent nuclear fuel transportation cask, it is necessary to analyze the regulatory processes for domestic and foreign spent nuclear fuel transportation cask. In this study, the regulatory processes for domestic and foreign spent nuclear fuel transportation cask was analyzed. In this study, the IAEA, US, and Korea spent nuclear fuel transportation cask regulatory processes were analyzed. The domestic and foreign spent nuclear fuel transportation cask regulatory processes consist of design phase, manufacturing phase, and operation phase. In the design stage, the transport requirements are designed in accordance with the safety requirements of international organizations and countries. The application to be submitted when applying for approval should include a safety analysis report, evidence proving compliance with safety requirements et al. In the manufacturing stage, it is a stage to check whether the safety requirements are satisfied before the first use after manufacturing the transportation cask. Inspections include welding inspection, leakage inspection, shielding inspection, and thermal inspection. In the operation stage, it is a stage of periodically performing inspections for continuous maintenance of the package when the transportation cask is used. The inspection items to be performed are similar to the manufacturing stage and typically include performance inspection of components and leakage inspection. In this study, domestic and foreign spent nuclear fuel transportation cask regulatory processes were analyzed. It was found that the domestic and foreign spent nuclear fuel transportation cask regulatory processes consist of the design phase, the manufacturing phase, and the operation phase. The results of this study can be used as basic data for policy decision-making for the spent nuclear fuel cask.

To rationalize the protection of spent nuclear fuel transport storage cask, we intend to investigate the status of domestic and foreign safety regulations and related technologies to develop sabotage scenarios and analyze the protection performance and radiation impact of transport storage cask. It is essential to conduct an aircraft collision safety evaluation on spent nuclear fuel transportation and storage casks in Korea due to changes in laws and regulations related to nuclear power plant design and demand for enhanced safety. Domestic and foreign research on the protection performance of spent nuclear fuel transport storage cask was based on 9.11 events, and the results of all studies show that the speed of the aircraft and leakage of nuclear materials are insignificant. The Sandia National Laboratory (SNL) calculates Aerosol emissions from spent fuel damage in the event of sabotage and calculates Source Term based on the Durbin-Luna model. In this paper, radiation sensitivity analysis was performed due to damage to the carrier according to the size of the accident, assuming that there was a hole enough to basket from the external shell among the collision scenarios identified for domestic cask models.

본 연구에서는 최근 개발중인 360 다발 장전용량의 중수로 사용후핵연료 운반용기에 대한 설계기준연료의 방사선원항 평가와 용기외부에서의 방사선량률 계산을 수행하였다. 그리고 국·내외 방사선적 안전성평가와 관련한 기술기준 부합여부를 판단하고 결과의 적합성을 제시하였다. 방사선원항으로 작용하는 설계기준연료 선정을 위해 월성원전에서 운영중인 운반 용기 및 두 가지 방식의 건식저장시설에 적용된 설계기준연료의 사양 및 특성을 조사하였다. 각 운반·저장 시스템 별 설계 기준연료의 연소도, 최소 냉각기간 및 중간저장시설로의 운반시점 등을 바탕으로 연소도 7,800 MWD/MTU와 최소 냉각기 간 6년을 설계기준연료로 설정하였다. 설계기준연료의 방사선원항은 SCALE 전산코드의 ORIGEN-ARP모듈을 이용하여 평가하였다. 운반용기의 방사선차폐평가는 MCNP6 전산코드를 이용하였으며, 기술기준에서 요구하는 운반용기 외부에서의 방사선량률 평가를 정상 및 사고조건으로 구분하여 수행하였다. 방사선량률 평가결과, 정상운반조건의 운반용기 표면 및 운반용기 표면 2 m 이격지점에서 계산된 최대 방사선량률은 각각 0.330 mSv·h-1와 0.065 mSv·h-1로 도출되어 선량률 제한치인 2.0 mSv^hr-1와 0.1 mSv^hr-1를 모두 만족하는 결과를 도출하였다. 또한 운반사고조건하 운반용기 표면 1 m 지점에서의 최대 방사선량률은 0.321 mSv·h-1로서 기술기준인 10.0 mSv·h-1 미만으로 평가되어, 대용량 중수로 사용후핵연료 운반용기는 방사선적 안전성을 확보하는 것으로 나타났다.