Due to the saturation of spent fuel pool of nuclear power plant in Korea, temporary storage for spent fuel will be installed, and spent fuel will be stored and managed in dry cask for a considerable period of time. Since spent nuclear fuel must withstand continuous decay heat, radiation and high internal pressure of the fuel rod in the cask, behavior of spent nuclear fuel is needed to be reviewed. Spent nuclear fuel used in Pressurized Water Reactor (PWR) in Korea is stored in a wet storage currently, but it is going to store a temporary dry-storage facility on Kori site. Therefore, it is very important and meaningful to evaluate the behavior of nuclear fuel with realistic modeling. Also, domestic PWR nuclear fuel has various burn-up. In the past, the burn-up of nuclear fuel in light water reactors was low, but in order to increase power generation efficiency, the concentration of uranium was increased and the number of new fuel was increased. Therefore, a large amount of nuclear fuel with burn-up of 45,000 MWD/MTU or higher, generally called high burn-up, is also stored in the spent fuel pool (SFP). Therefore, it is necessary to evaluate by dividing three different burn-up such as, low, medium, and high burn-up. Thus, this study will review the behavior of nuclear fuel at different burn-up during the temporary storage period with FALCON (EPRI), computational code and analyze the factors affecting the integrity of nuclear fuel, including when the temporary storage is extended its additional lifetime. And this evaluation will contribute developing the spent fuel management plan in Korea.

Al-B4C neutron absorbers are currently widely used to maintain the subcriticality of both wet and dry storage facilities of spent nuclear fuel (SNF), thus long-term and high-temperature material integrity of the absorbers has to be guaranteed for the expected operation periods of those facilities. Surface corrosion solely has been the main issue for the absorber performance and safety; however, the possibility of irradiation-assisted degradation has been recently suggested from an investigation on Al-B4C surveillance coupons used in a Korean spent nuclear fuel pool (SFP). Larger radiation damage than expectation was speculated to be induced from 10B(n, α)7Li reactions, which emit about a MeV α-particles and Li ions. In this study, we experimentally emulated the radiation damage accumulated in an Al-B4C neutron absorber utilizing heavy-ion accelerator. The absorber specimens were irradiated with He ions at various estimated system temperatures for a model SNF storage facility (room temperature, 150, 270, and 400°C). Through the in-situ heated ion irradiation, three exponentially increasing level of radiation damages (0.01, 0.1, and 1 dpa or displacement per atom) were achieved to compare differential gas bubble formation at near surface of the absorber, which could cause premature absorber corrosion and subsequential 10B loss in an SNF storage system. An extremely high radiation damage (10 dpa), which is unlikely achievable during a dry storage period, was also emulated through high temperature irradiation (350°C) to further test the radiation resistance of the absorber, conservatively. The irradiated specimens were characterized using HR-TEM and the average size and number density of radiation-induced He bubbles were measured from the obtained bright field (BF) TEM micrographs. Measured helium bubble sizes tend to increase with increasing system (or irradiation) temperature while decrease in their number density. Helium bubbles were found from even the lowest radiation damage specimens (0.01 dpa). Bubble coalescence was significant at grain boundaries and the irradiated specimen morphology was particularly similar with the bubble morphology observed at the interface between aluminum alloy matrix and B4C particle of the surveillance coupons. These characterized irradiated specimens will be used for the corrosion test with high-temperature humid gas to further study the irradiation-assisted degradation mechanism of the absorber in dry SNF storage system.

후쿠시마 원전사고 이후 광역의 방사성 오염부지가 발생되었으며, 이에 대한 제염작업으로 인하여 다량의 제염폐기물이 발 생하였다. 일본에서는 이를 보관하기 위하여 각 지역에 임시저장시설이 운영되고 있으며, 이들 시설들은 피난지시해제가 이루어진 지역의 일반인에 대하여 방사선학적 영향을 미칠 것으로 판단된다. 본 연구에서는 임시저장시설 인근에 거주하 는 일반인의 방사선학적 안전성 확보를 위하여 임시저장시설 특성에 따른 거리별 공간 방사선량률 및 선량제한치를 만족하 는 임시저장시설로부터의 이격거리를 평가하였다. 이를 위해 임시저장시설의 형태 및 크기, 복토 두께 등을 고려하였으며, MCNPX를 이용하여 방사선량률을 평가하였다. 복토에 의한 차폐효과는 두께가 10 cm일 때 68.9%, 30 cm일 때 96.9%, 50 cm 일 때 99.7%로 나타났다. 임시저장시설 형태에 따른 공간 방사선량률은 지상 보관형일 때 가장 높게 나타났으며, 이어서 반 지하 보관형, 지하 보관형일 순으로 나타났다. 임시저장시설 크기에 따른 공간 방사선량률은 5 × 5 × 2 m 시설을 제외한 시 설에 대하여 유사하게 나타났다. 이는 임시저장시설 내 적재된 제염폐기물에 의하여 자기차폐가 이루어지기 때문이다. 최종 적으로 크기가 50 × 50 × 2 m이고, 복토가 없는 임시저장시설의 경우, 지상 보관형의 평가된 이격거리는 14 m(최소농도), 33 m(최빈농도), 57 m(최대농도)이며, 반지하 보관형의 이격거리는 9 m(최소농도), 24 m(최빈농도), 45 m(최대농도), 지하 보관형의 이격거리는 6 m(최소농도), 16 m(최빈농도), 31 m(최대농도)로 나타났다.

크로스도킹이란 창고나 물류센터에 하역된 물품이 저장됨이 없이 도착지별로 재분류되어서 직출하되는 물류 시스템이다. 크로스도킹은 물류비용의 큰 비중을 차지하는 보관비용을 감소시킬 수 있을 뿐만 아니라 고객의 요구에 빠른 대응을 할 수 있다는 장점을 지니고 있다. 크로스도킹이 성공적으로 수행되기 위해서는 창고나 물류센터의 입고에서부터 출고까지의 모든 작업들이 계획적이고 원활하게 수행되어져야만 한다. 본 연구에서는 임시보관 장소를 보유하지 않은 크로스도킹 시스