• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.13 no.4
• /
• pp.301-313
• /
• 2015

The rock cavern storage for spent fuel has been assessed to apply in Korea with reviewing the state of the art of the technologies for surface storage and rock cavern storage of spent fuel. The technical feasibility and economic aspects of the rock cavern storage of spent fuel were also analyzed. A considerable area of flat land isolated from the exterior are needed to meet the requirement for the site of the surface storage facilities. It may, however, not be easy to secure such areas in the mountainous region of Korea. Instead, the spent fuel storage facilities constructed in the rock cavern moderate their demands for the suitable site. As a result, the rock cavern storage is a promising alternative for the storage of spent fuel in the aspect of natural and social environments. The rock cavern storage of spent fuel has several advantages compared with the surface storage, and there is no significant difference on the viewpoint of economy between the two alternatives. In addition, no great technical difficulties are present to apply the rock cavern storage technologies to the storage of domestic spent fuel.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1995.05b
• /
• pp.727-732
• /
• 1995

사용후핵연료 중간저장 시설의 누수사고시 예상되는 핵연료봉의 온도상승을 SFUEL 컴퓨터 코드 분석결과에 따른 실제 $UO_2$의 산화거동을 실험하였다. 외기 온도 38$^{\circ}C$에서 환기회수가 시간당 0, 1, 2회인 조건에서 저장용기 밑바닥 구멍 크기가 2.54, 5.08, 7.62 cm인 경우의 실험결과 환기회수 0회 바닥구멍 크기 2,54 cm 일 때 약 15시간 후 건전성 상실(0.6% 무게증가)이 일어났으며 환기회수 2회 바닥구멍 크기 7.62 cm 일 때는 약 21시간 이후에 건전성 상실이 일어나 가장 느렸다. 바닥구멍 크기가 증가할수록 공기 순환비의 영향을 크게 받으며, 또 외기 온도가 낮을 수록 공기 순환비의 영향을 크게 받았다.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2011.10a
• /
• pp.85-86
• /
• 2011

• Journal of Radiation Protection and Research
• /
• v.21 no.4
• /
• pp.313-327
• /
• 1996

As an at-reactor(AR) storage method o( spent fuel, there are horizontal concrete module type, metal storage cask type, concrete storage cask type, dual purpose (transportation and storage) cask type and multi-purpose (transportation, storage and disposal) cask type. All other types except multi-purpose one have been already used for AR dry storage of spent fuels after obtaining operation license in various foreign countries. Also the development of multi-purpose type has been continued for operation license. In America, Japan, Germany, Canada, Spain, Switzerland, and Czech Republic, etc., AR dry storage facilities are under operation or on propulsion, and spent fuels are transported to interim storage facility or reprocessing plant after dry storage at reactor temporarily. At Wolsung site, in case of Korea, concrete silo type has already been introduced, and it is believed to be inevitable to store spent fuels at reactor temporarily, considering the reality that storage capacity of spent fuel is approaching to the limit in some nuclear power plants. In this report, the system characteristics, design requirements, technical standards and status of AR storage system, which is suitable for domestic site such as Kori, have been studied. In most cases, the licensed period of storage cask is limited up to 20 years and the integrity of material and maintenance of leaktightness are required during the whole service life.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.17 no.4
• /
• pp.375-387
• /
• 2019

In this study, the radiation dose rates for the design basis fuel of 360 assemblies CANDU spent nuclear fuel transportation cask were evaluated, by measuring radiation source terms for the design basis fuel of a pressurized heavy water reactor. Additionally, radiological safety evaluation was carried out and the validity of the results was determined by radiological technical standards. To select the design basis fuel, which was the radiation source term for the spent fuel transportation cask, the design basis fuels from two spent fuel storage facilities were stored in a spent fuel transportation cask operating in Wolsung NPP. The design basis fuel for each transportation and storage system was based on the burnup of spent fuel, minimum cooling period, and time of transportation to the intermediate storage facility. A burnup of 7,800 MWD/MTU and a minimum cooling period of 6 years were set as the design basis fuel. The radiation source terms of the design basis fuel were evaluated using the ORIGEN-ARP computer module of SCALE computer code. The radiation shielding of the cask was evaluated using the MCNP6 computer code. In addition, the evaluation of the radiation dose rate outside the transport cask required by the technical standard was classified into normal and accident conditions. Thus, the maximum radiation dose rates calculated at the surface of the cask and at a point 2 m from the surface of the cask under normal transportation conditions were respectively 0.330 mSv·h^-1 and 0.065 mSv·h^-1. The maximum radiation dose rate 1 m from the surface of the cask under accident conditions was calculated as 0.321 mSv·h^-1. Thus, it was confirmed that the spent fuel cask of the large capacity heavy water reactor had secured the radiation safety.