• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.4
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• pp.391-402
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• 2017

The Korea Radioactive Waste Agency (KORAD) has developed a concrete cask for the dry storage of spent nuclear fuel that has been generated by domestic light-water reactors. During long-term storage of spent nuclear fuel in concrete casks kept in dry conditions, the integrity of the concrete cask and spent nuclear fuel must be maintained. In addition, the radiation dose rate must not exceed the storage facility's design standards. A suitable shielding design for radiation protection must be in place for the dry storage facilities of spent nuclear fuel under normal and accident conditions. Evaluation results show that the appropriate distance to the annual dose rate of 0.25 mSv for ordinary citizens is approximately 230 m. For a $2{\times}10$ arrangement within storage facilities, rollover accidents are assumed to have occurred while transferring one additional storage cask, with the bottom of the cask facing the controlled area boundary. The dose rates of 12.81 and 1.28 mSv were calculated at 100 m and 230 m from the outermost cask in the $2{\times}10$ arrangement. Therefore, a spent nuclear fuel concrete cask and storage facilities maintain radiological safety if the distance to the appropriately assessed controlled area boundary is ensured. In the future, the results of this study will be useful for the design and operation of nuclear power plant on-site storage or intermediate storage facilities based on the spent fuel management strategy.

• Journal of Radiation Protection and Research
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• v.21 no.4
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• pp.313-327
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• 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)
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• v.4 no.2
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• pp.139-152
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• 2006

A spent fuel storage cask is required to prove the safety of a canister under a hypothetical accidental drop condition. A hypothetical accidental drop condition means that a canister is assumed to be a lee drop on to a pad of the storage cask during loading it into a storage cask. A pad of the storage cask absorbs shock to maintain the structural integrities of a canister under a hypothetical accidental drop condition. In this paper a finite element analysis for various pad structures was carried out to improve the structural integrity of a canister under a hypothetical accidental drop condition. A pad of a storage cask was designed a steel structure with concrete. The 1/4 height of a pad was modified with a structure composed of a steel and a polyurethane foam as a impact limiter. The effect of a shape of a steel structure was studied. The effects of the thickness of a steel structure and the density of a polyurethane foam was also studied.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.436-436
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• 2004

사용후 연료 건식 저장 용기가 낙하, 토네이도, 미사일, 홍수 및 지진으로 인한 사고에 대하여 전복이 발생되었을 때 강체 평면과 충돌에 의한 충돌 하중시의 구조 응력 평가하였다. 이를 위해 저장 용기의 무게 중심이 한계를 넘었을 때의 초기 전복 시작각을 무게 중심을 계산을 통해 구하였다. 상용 코드를 사용하여 전복 응력 해석 수행시 저장 용기의 강체 운동에 의하여 계산 시간이 길어지는 데, 이런 계산 시간을 줄이기 위해 일차 충돌 직전까지의 모델의 속도와 각속도 계산식을 이론적인 방법으로 구하여 해석 초기 조건으로 사용하는 방법에 대하여 제안하였다.(중략)

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.470-475
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• 1996

사용후핵연료 장기 건식저장시 여러가지 저장조건에서 사용후핵연료 피복관 및 사용후핵연료 ($UO_2$)에 대한 장기 건전성을 종합적으로 평가할 수 있는 SIECO 코드를 개발하였다. 건식저장 시스템은 사용후핵연료를 헬륨 및 공기분위기하에서 TN-24P 건식 저장용기에 장기 저장할 경우로 하였으며 피복관의 최대 표면온도는 COBRA-SFS코드를 사용하여 계산하였고, 열유동 해석결과를 바탕으로 SIECO코드를 이용하여 핵연료 연소도 및 냉각기간, 냉각매체에 따른 최대 건식저장 허용온도를 피복관의 열화 및 $UO_2$ 산화의 관점에서 계산하였다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.3
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• pp.245-253
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• 2006

A test of the 1/3 scale model was conducted to verify the tip-over analysis of a dry. concrete storage cask under a hypothetical accident condition. The tip-over analysis was executed using the velocity at each point as the initial conditions of the model just before the impact. The initial velocity was determined from the initial angular velocity, which would make the equivalent kinetic energy to the potential energy. To confirm the structural integrity of the canister, the visual testing and the non-detective testings such as Liquid Penetrant testing and Ultrasonic Testing were conducted. The lid of a storage cask was plastically deformed near the impact point. The structural integrity of storage cask was maintained. To verify the tip-over analysis the strains and the accelerations acquired by the tip-over test were compared with those by the analyses. The results of the analysis were larger than the test results about two times.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2016.05a
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• pp.89-90
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• 2016

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.4
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• pp.455-472
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• 2018

It is important to study how to manage dry storage casks of spent nuclear fuels (SNF), because wet storage spaces for SNF will shortly be at full capacity in the Republic of Korea. The US has operated a dry storage cask system for several decades, and has carried out significant studies into how to successfully manage dry storage cask for SNF. This type of expertise and experience is currently lacking in the Republic of Korea. The degradation of dry casks is an important issue that must be considered. In particular, chloride-induced stress corrosion cracking (CISCC) is known to lead to the release of radioisotopes from canisters. The U.S. Department of Energy, U.S. Nuclear Regulatory Commission, and the Electric Power Research Institute have undertaken research into the CISCC mechanism. In addition, Sandia National Laboratories (SNL) has extensively researched CISCC and how to manage it in dry storage canisters. In this review paper, the probabilistic model proposed by the SNL is analyzed and, based on this model, US-based CISCC research is reviewed in detail. This paper will inform the management of dry cask storage of SNF from light water reactors in austenite stainless steel canisters in the Republic of Korea.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2009.11a
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• pp.475-476
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• 2009

CFD 방법을 이용하여 건식저장 용기내 사용후핵연료 열전달 해석을 수행한 결과 연료봉의 붕괴열에 의한 내부 유체의 자연대류 현상과 상세 핵연료 온도분포를 예측할 수 있음을 확인하였다. 향후에는 다양한 시험조건에서 복사열전달을 포함한 정밀한 CFD 계산을 수행하여 피복관 온도분포의 예측치를 실험결과와 비교할 예정이다.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.11a
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• pp.251-251
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• 2005