• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.11a
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• pp.139.1-139.1
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• 2004

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2016.10a
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• pp.23-24
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• 2016

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2012.05a
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• pp.47-48
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• 2012

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05d
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• pp.488-493
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• 1996

보장조치 대상 원자력 시선에 대한 사찰 목적은 평화적 목적으로 사용되기 위한 시설 및 핵물질이 핵무기 생산 등의 비평화적 목적으로 전용되지 않았음을 확인하는 것이다. 이를 위하여 국제원자력기구에서는 보장조치 기준(IAEA Safeguards Criteria : 1991 - 1995)에 따라 적절한 검증 수단을 사용하여 핵물질의 형태 및 양, 시설의 운전기록 등에 대하여 보고된 내용과 실제 상황과의 일치성을 확인하고, 미신고된 핵활동이 없음을 확인하고 있다. 보장조치 측면에서 보면, 중수형원자로(CANDU)는 핵연료의 크기가 작고 운전중에 핵연료를 교체하는 방식(On Load Reactors)을 채택하고 있기 때문에 시설 내에서의 핵물질 이동이 매우 빈번하며, 사용후핵연료의 양 역시 경수형원자로에 비해 매우 많다. 따라서 중수형원자로에 대한 보장조치 사찰은 경수형원자로에 비해 사찰일수(최대허용사찰량 : 중수형원자로 45 인-일/년, 경수형원자로 15 인-일/년)가 훨씬 많고 보장조치 관련 장비 또한 매우 다양하다. 현재 운전 중인 월성 1호기에 이어 건설 중인 월성 2, 3, 4호기의 운전이 시작되면 중수형원자로에 대한 국제원자력기구 및 국가사찰 양이 급격히 늘어날 전망이다. 또한 월성 1호기의 경우 사용후핵연료 저장조의 용량 초과로 인한 건식저장고(Dry Canister)로의 이송이 1992년도부터 매년 실시되고 있으며, 이 기간 중에 이송 대상 핵연료의 검증 및 운반 중 전용을 방지하기 위한 추가적인 사찰이 수행됨으로써 많은 인력과 시간이 투입되고 있다. 또한 국제원자력기구에서 추진하고 있는 보장조치 강화 방안의 일환으로 현재 건설 중인 월성 2, 3, 4호기에 대해서는 월성 1호기에는 적용되지 않은 추가적인 보장조치 관련 장비의 설치가 고려되고 있다. 이에 따라 우리나라에서는 중수형원자로에 대한 국제 원자력기구의 사찰 기준 및 사찰 내용을 분석, 중수형원자로 보장조치 사찰에 대한 개선점을 도출하고, 후속기에 대해서 보다 효율적이고 효과적인 보장조치 방안을 적용토록 하여야 할 것이다.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2015.10a
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• pp.41-42
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• 2015

• 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.

• Nuclear Engineering and Technology
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• v.21 no.3
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• pp.223-230
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• 1989

This paper describes dynamic simulation of the spent fuel storage system which is described by statistical discrete event models. It visualizes flow and queue of system over time, assesses the operational performance of the system acitivies and establishes the system components and streams. It gives information on system organization and operation policy with reference to the design. System was tested and analyzed over a number of critical parameters to establish the optimal system. Workforce schedule and resources with long processing time dominate process. A combination of two workforce shifts a day and two cooling pits gives the optimal solution of storage system. Discrete system simulation is an useful tool to get information on optimal design and operation of the storage system.

• Journal of the Korean Society of Radiology
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• v.17 no.6
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• pp.947-955
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• 2023

This study was conducted to investigate the awareness of spent nuclear fuel among residents in nuclear power plant areas and use it as basic data for establishing a disposal facility for high-level radioactive waste. 204 questionnaires collected online were analyzed using SPSS Window Ver 28.0. To verify differences between groups, t-test and one-way ANOVA were performed. And correlation analysis was conducted to confirm the relationship between variables. As a result, first, risk perception regarding nuclear-related accidents showed statistically significant differences depending on gender and educational level. The position on the construction of a permanent disposal facility for spent nuclear fuel showed a statistically significant difference depending on gender, education, and age, and the perception of the importance of each evaluation standard for establishing a spent nuclear fuel management plan showed a statistically significant difference depending on education and age. In terms of trust in information-providing institutions, trust in the National Assembly was found to be the lowest. Second, the results of the correlation analysis between variables showed that local residents are aware that an alternative to the current disposal of spent nuclear fuel is needed, and that financial support for the construction of a permanent disposal facility is needed. Therefore, in order to build a high-level radioactive waste disposal site, it is believed that it is necessary to increase trust in the government, collect opinions from local residents, and provide economic support.

• Nuclear Engineering and Technology
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• v.22 no.2
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• pp.83-94
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• 1990

The radioactivity in the spent fuel storage pool is calculated to ensure to maintain its concentration below the permissible limit, when the storage capacity of Uljin nuclear power plant unit 1&2 is extended from 9/3 to 32/3 core using consolidated fuels in maximum density rack (MDR). For this evalulation, two models to calculate the spent fuel pool activities on the continuous and intermittent operating its purification system are developed and these results compared, The results of above two cases show that the current water purification system can not guarantee the radioactivity concentration below the design limit, 5$\times$10$^{-4}$ $\mu$Ci/ml, for the extention to 32/3 core. Therefore, it has been concluded that a modification of the current purification system is necessary to extend the spent fuel storage capacity with the above method. The alternative way suggested in this study is to increase the number of cation bed demineralizers.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.3
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• pp.239-245
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• 2010

There are 11,811 LWR spent fuels stored at reactor sites, as of 2009. Source terms based on reference spent fuel which represents entire spent fuels with bounding values in the aspect of source term has been applied to a design of nuclear installations, instead of those which are generated by weighting respective source term for each spent fuel. Simplified regression models to estimate total decay heat, radioactivity, and ingestion hazard index for spent fuel from Westinghouse-type reactors were developed in this study, because it can be used as a fundamental model for weighting source term for respective spent fuel to exclude conservativeness in source terms. It was found that the estimated source terms agreed with calculated value from ORIGEN-ARP within 5%. It was also found that the conservativeness could be excluded if the weight source terms were used as reference source term in the design. Therefore, it is expected that the developed regression model could be widely used in the conceptual design process of nuclear facilities related with storage and disposal of spent nuclear fuel.