• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2016.05a
• /
• pp.79-80
• /
• 2016

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2012.05a
• /
• pp.25-26
• /
• 2012

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2011.10a
• /
• pp.507-508
• /
• 2011

• Journal of Radiation Protection and Research
• /
• v.38 no.4
• /
• pp.189-193
• /
• 2013

Annual dose on the containment building wall of the interim storage facility at normal condition was calculated to estimate the dose rate transition of the facility of PWR spent nuclear fuel. In this study, source term was generated by ORIGEN-ARP with 4.5 wt% initial enrichment, 45,000 MWd/MTU burnup and 10 years cooling time. Modeling of the storage facility and the containment building and radiation shielding evaluations were conducted by MCNP code depending on the distance between the wall and the facility in the building. In the case of the centralized storage system, the distance required for the annual dose rate limit from 10CFR72 was estimated to be 50 m.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.9 no.2
• /
• pp.93-98
• /
• 2011

Saline corrosion is one of the major degradation mechanisms for stainless steel type 304 (SS304) dry storage cask during the spent fuel interim storage period. Slow strain rate test (SSRT) and neutral salt spray test (NSS) were performed at $85^{\circ}C$ and $200^{\circ}C$ with 0.5 wt% sodium chloride mist sprayed on the cold-rolled SS304 specimens of different degrees of reduction in this study. The weight changes of the NSS specimens tested at $85^{\circ}C$ for 2000 hours differed greatly from those at $200^{\circ}C$. The weight loss of NSS specimens was not significant at $85^{\circ}C$ but the weight gain decreased gradually with increasing the cold-rolled reduction. The yield strength (YS) and ultimate tensile stress (UTS) values obtained from the SSRT tests for lightly cold-rolled specimens in the salt spray environment at $85^{\circ}C$ and $200^{\circ}C$ are slightly lower than in air. But for those with 20% reductions, the specimen strengths were no longer changed by the saline corrosion. The preliminary results demonstrated that the quality and performance of cold-rolled SS304 is acceptable for fabrication of dry storage casks. However, more work on the corrosion behavior of cold-rolled stainless steel in the saline atmosphere is needed to better understand its long-term performance.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2008.11a
• /
• pp.352-353
• /
• 2008

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2012.10a
• /
• pp.449-450
• /
• 2012

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.16 no.1
• /
• pp.110-116
• /
• 2020

As the management plan for domestic spent nuclear fuel is delayed, the storage of the operating nuclear power plant is approaching saturation, and the Kori 1 Unit that has reached its end of operation life is preparing for the dismantling plan. The first stage of dismantling is the transfer of spent nuclear fuel stored in storage at plants. The spent fuel management process leads to temporary storage, interim storage, reprocessing and permanent disposal. In this paper, the technical issues to be considered when transporting spent fuel in this process are summarized. The spent fuels are treated as high-level radioactive waste and strictly managed according to international regulations. A series of integrity tests are performed to demonstrate that spent fuel can be safely stored for decades in a dry environment before being transferred to an intermediate storage facility. The safety of spent fuel transport container must be demonstrated under normal transport conditions and virtual accident conditions. IAEA international standards are commonly applied to the design of transport containers, licensing regulations and transport regulations worldwide. In addition, each country operates a physical protection system to reduce and respond to the threat of radioactive terrorism.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2004.06a
• /
• pp.351-351
• /
• 2004

원자력연구소에서는 PWR 사용후핵연료를 건식 개질함으로써 관리 부피를 줄이고 안전상에 문제를 일으키는 고방사성 핵종인 세슘과 스트론튬 등을 선택적으로 제거하여 방사능 및 냉각부하를 줄일 수 있는 사용후핵연료 차세대관리 공정개발에 대한 연구를 수행하고 있다. 이는 세라믹 형태의 PWR 핵연료를 금속으로 전환시켜 관리하는 방법으로 금속전환체는 PWR 사용후핵연료와 비교하여 체적, 방사능 및 발열량을 약 1/4로 줄일 수 있는 이점이 있다.(중략)

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.14 no.4
• /
• pp.411-422
• /
• 2016

Recently, because the wet pool storage facilities of NPPs in Korea has become saturated, there has been much active R&D on an interim dry storage system using a transportation and storage cask. Generally, the shielding evaluation for the design of a spent fuel transportation and storage cask is performed by the design basis fuel, which selects the most conservative fuel among the fuels to be loaded into the cask. However, the loading of actual spent fuel into the transportation metal cask is not limited to the design basis fuel used in the shielding evaluation; the loading feasibility of actual spent fuel is determined by the shielding evaluation that considers the characteristics of the initial enrichment, the maximum burnup and the minimum cooling period. This study describes a shielding analysis method for determining the minimum cooling period of spent fuel that meets the domestic transportation standard of the dual purpose metal cask. In particular, the spent fuel of 3.0~4.5wt% initial enrichment, which has a large amount of release, was evaluated by segmented shielding calculations for efficient improvement of the results. The shielding evaluation revealed that about 81% of generated spent fuel from the domestic nuclear power plants until 2008 could be transported by the dual purpose metal cask. The results of this study will be helpful in establishing a technical basis for developing operating procedures for transportation of the dual purpose metal cask.