• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.381-382
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• 2005

원자력발전소에서 발생하는 사용후연료 집합체를 운반하기 위한 수송용기는 고준위 방사성물질의 위험으로부터 인간과 환경을 보호하기 위하여 안전성이 철저하게 보장되어야만 한다. 원자력법과 IAEA 안전수송규정 등 국내외의 관련규정에 의하면 사용후연료 수송용기는 정상운반조건은 물론 수송 도중 발생할 수 있는 운반사고조건에서 B(U)F형 운반용기에 대한 기술기준을 만족시키어 어떠한 경우에도 방사선차폐, 임계, 격납, 열 및 구조적 건전성을 유지하여 방사성물질을 누출시키지 않아야 한다고 규정하고 있다. 본 논문은 한수원(주)에서 개발하여 현재 사용하고 있는 경수로형 사용후연료 수송용기(KN-12 수송용기)의 격납계통에 대한 건전성을 확인하기 위하여 해석에 의한 격납평가 및 수송용기의 운영 중 수행하는 누설시험 등의 누설평가방법에 대하여 기술하였으며, 또한, 매 운반 시 측정한 실제 누설률을 제시하고 분석하였다.

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

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2006.11a
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• pp.87-88
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• 2006

사용후 핵연료용 수송용기의 설계 안전평가에서는 이제까지 용기에 수납되는 연료는 미조사, 즉 신연료라 가정해서 보수적으로 임계안전설계를 수행하여 왔다. 이것은 연소에 따른 연료내의 핵연료 물질의 감손 및 생성의 의한 반응도의 변동을 계산 평가하는 것이나 또는 연소로 인해 생성되는 중성자 흡수 핵종의 조성 및 함유량 등을 정확히 계산 평가하는 것이 복잡해서 곤란했던 것으로 그 요인을 들 수 있다. 사용 후 핵연료를 신 연료로 가정하는 등의 불합리성을 해소하고, 안전성을 잃지 않고 사용 후 핵연료 운반용기 들의 경제성을 추구하는 기운이 높아지고, 관련 연구가 적극적으로 진척되게 되었다. 그 결과 연소에 따른 연료내의 핵연료 물질의 감손 생성과 핵분열 생성물 등에 의한 반응도의 저하, 즉 중성자 실효 증배율의 저하를 고려한 것을 사용 후 핵연료용 캐스크 설계 안전평가에 취할 수 있게 되었다. 연소도 크레디트를 채용함으로서 사용후 핵연료내의 핵연료물질량은 실제로 존재하는 양을 사용하는 것이 되므로 초기 농축도가 높은 고연소도 연료에서 그 효과가 보다 크게 될 것이다. 이것은 연소도 크레디트 채용에 따라 연료 바스켓의 중성자흡수제 사용량 감소가 가능해져 사용 캐스크의 수를 줄일 수 있어 경제성 향상이 기대되고 아울러 그이 취급 횟수 및 수송횟수가 감소됨에 따라 안전성의 향상도 기대된다.

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

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.19 no.2
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• pp.69-74
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• 2023

The paper presents the results investigating the effect of aircraft engine impact location on the intended function evaluation results of spent nuclear fuel transport cask. As a result of the investigation, it is found that the structural integrity is maintained as the maximum accumulated equivalent plastic strain is below the acceptable criterion regardless of the collision location. It is identified that when the aircraft engine collided with the upper part of the transport cask without considering impact limiter the containment performance is weakened compared to when the aircraft engine collided with the central part.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2011.10a
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• pp.39-40
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• 2011

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2010.05a
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• pp.335-336
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• 2010

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2010.05a
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• pp.343-344
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• 2010

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2016.05a
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• pp.97-98
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• 2016

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.3
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• pp.191-198
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• 2011

In general, conventional criticality analyses for spent fuel transport/dry storage systems have been performed based on assumption of fresh fuel concerning the potential uncertainties from number density calculation of Transuranic and Fission Products in spent fuel. However, because of economic loss due to the excessive criticality margin, recently the design of transport/dry storage systems with Burnup Credit(BUC) application has been actively developed. The uncertainties in criticality analyses on transport/storage systems with BUC technique show strong dependance upon initial enrichment and burnup rate, whereas those in the conventional criticality evaluation based on fresh fuel assumption do not show such a dependance. In this study, regulatory-required uncertainties of the criticality analyses for BK 26 Cask, which is conceptually designed spent fuel transport cask with BUC corresponding to the limiting circumstances on nuclear power plants in Korea, are evaluated as a function of initial enrichment and burnup rate. Results of this study will be used as basic data for spent fuel loading curve of BK 26 Cask.