• Journal of Radiation Protection and Research
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• v.9 no.2
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• pp.90-96
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• 1984

Estimated dose rates in spent fuel pool storage with the extended fuel cycle core management were reviewed and compared with design limit after calculation with the aid of DLC-23/CASK(22 n, 18 g) nuclear data and ANISN code. Radioactivity and gamma spectrum within spent fuel assemblies were calculated with ORIGEN code by extended fuel cycle model. In the calculation of dose rate, the fuel pool geometry was assumed to be infinite slab. Also, composition materials and radiation source within assemblies which are being stored in pool storage were assumed to be uniformly distributed throughout all the assemblies. As a result of culculation of dose rate from stored assemblies and waterborne radionuclides in pool water, the calculated dose rates appear to be lower than design basis limit under normal condition as well as abnormal condition.

• Korean Chemical Engineering Research
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• v.50 no.4
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• pp.696-701
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• 2012

Non-aqueous processes have been developed for stable management and reuse of spent fuels. Nowadays, a plan for the management of spent fuel is being sought focusing on a non-aqueous process in Korea. Named as pyroprocessing, it includes an electrolytic reduction process using molten salt at high temperature for the spent fuels, which provides metallic product for a following electro-refining process. The electrolytic reduction process utilizes electrochemical reaction producing Li to convert oxides into metals in high temperature LiCl medium. Various kinds of elements in the spent fuels would be distributed in the system according to their respective reactivity with the reductant, Li, and the medium, LiCl. This study elucidates the reactions of the elements to understand the behavior of composite elements on the spent fuels by thermodynamic calculations. Uranium and transuranic are reduced into their metallic forms while rare-earth oxides, except for Eu, are stable against the reaction at a process temperature. This study also covers the tendency of reactions with respect to the temperature and, finally, estimates radioactivity and heat load on the distributed phases based on the reference spent fuel characteristics.

• Journal of Radiation Protection and Research
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• v.21 no.2
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• pp.117-124
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• 1996

Source intensities in terms of the exposure rates at 1m from the fresh and spent DUPIC fuels, made from standard and extended turnup PWR fuels, were analyzed. Two cases were studied based on the degree of elimination of removable elements. Homogeneous mixture model was applied to get the exposure rate. The exposure rate turned out to be very high and sensitive to Cs elimination during the dry process. About 90% of exposure can be reduced in the case of fresh DUPIC fuel made from 10-year cooled spent PWR fuels if Cs is fully removed during the dry process. The main radiation source in spent fuels is Cs-137. The dry storage of spent DUPIC fuel may need a longer wet storage period and require a further review.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.497-500
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• 1996

한국원자력연구소 부설원자력환경관리센터에서는 사용후핵연료 소외 중간저장 종합관리시설 관련 부지선정, 설계, 건설, 운영, 환경감시 및 폐쇄 제 단계 인허가에 필요한 기술기준들을 1989년부터 1997년까지 개발해 오고 있다. 본 논문에서는 현재까지 이들 기술기준들의 개발현황과 향후 개선점들을 종합적으로 조명하여 보았다.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.415-419
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• 1996

한국원자력연구소 부설원자력환경관리센터에서는 사용후핵연료 소외 중간저장 종합관리시설 관련 부지선정, 설계, 건설, 운영, 환경감시 및 폐쇄 제 단계 인허가에 필요한 기술기준들을 1989년부터 1997년까지 개발해 오고 있다. 본 논문에서는 현재까지 이들 기술기준들의 개발현황과 향후 개선점들을 종합적으로 조명하여 보았다.

• 대한방사선방어학회:학술대회논문집
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• 2011.11a
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• pp.182-183
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• 2011

• Nuclear industry
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• v.7 no.2 s.48
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• pp.84-90
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• 1987

원자력발전의 개발이용 추진에 있어서 앞으로 남아있는 큰 문제의 하나가 고준위(방사성)폐기물의 처리이다. 고준위폐기물은 사용후핵연료의 재처리에 의해서 발생하는데, 재처리하지 않은 사용후핵연료도 이 류로 취급된다. 폐기물이라면 언젠가 처분해야 하는데, 그에 앞서 아직 이용할 수 있는 물질이 포함되어 있지 않은가, 또한 그것을 분리해 냄으로써 실제의 폐기량을 감소시킬 수 있지 않을까? 확인해 볼 필요가 있다.

• Nuclear industry
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• v.4 no.5 s.21
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• pp.42-45
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• 1984

원자력발전소에서 연소된 사용후핵연료를 여러 외국에서는 재처리 또는 처분키로 하고 있다. 그러나 그 이전에 중간저장을 고려하고 있으며 이를 위해 여러 가지 방식이 개발, 실용화되고 있다. 다음은 이들 여러나라의 동향 및 개발, 실용화된 주된 방식의 특징과 장${\cdot}$단점의 개략이다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.1
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• pp.31-39
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• 2013

Nuclear fuel can be damaged under unexpected circumstances in a nuclear reactor. Fuel rod failure can be occurred due to debris fretting or excessive hydriding or PCI (Pellet-to-clad Interaction) etc. It is important to identify the causes of such failed fuel rods for the safe operation of nuclear power plants. If a fuel rod failure occurs during the operation of a nuclear power plant, the coolant water is contaminated by leaked fission products, and in some case the power level of the plant may be lowered or the operation stopped. In addition, all spent fuels must be transferred to a dry storage. But failed fuel can not be transferred to a dry storage. Therefore, the purpose of this study is to develop a system which is capable of inspecting whether the spent fuel in the storage pool is failed or not. The sipping technology is to analyze the leakage of fission products in state of gas and liquid. The failed fuel inspection system with gamma analyzer has successfully demonstrated that the system is enough to find the failed fuel at Wolsong plant.

• Nuclear Engineering and Technology
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• v.25 no.1
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• pp.166-177
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• 1993

A simplified thermal analysis method to evaluate the maximum temperature of the CANDU 37-element fuel bundle within a fuel basket in a given spent fuel dry storage canister has been presented along with the results of sample analyses performed to examine the parametric effects of the ambient conditions on the maximum fuel temperature within a canister. To solve the multi-dimensional heat transfer problem of the complex geometry of rod bundles within a canister where three modes of heat transfer are superimposed, the CANDU spent fuel bundles stored in the dry storage canister are first replaced by equivalent concentric fuel cylinders. The simplified axi-symmetric two-dimensional multi-mode heat transfer problem of the equivalent fuel cylinders is then analyzed with an existing computer code, HEATING5, using additional input data and heat transfer correlations. A comparison between the predicted temperature profile and the mock-up test results shows that the agreement is quite satisfactory.