• Nuclear Engineering and Technology
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• 제49권2호
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• pp.442-449
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• 2017

Risk management of the Fukushima Daiichi Nuclear Power Station decommissioning is a great challenge. In the present study, a risk management framework has been developed for the decommissioning work. It is applied to fuel assembly retrieval from Unit 3 spent fuel pool. Whole retrieval work is divided into three phases: preparation, retrieval, and transportation and storage. First of all, the end point has been established and the success path has been developed. Then, possible threats, which are internal/external and technical/societal/management, are identified and selected. "What can go wrong?" is a question about the failure scenario. The likelihoods and consequences for each scenario are roughly estimated. The whole decommissioning project will continue for several decades, i.e., long-term perspective is important. What should be emphasized is that we do not always have enough knowledge and experience of this kind. It is expected that the decommissioning can make steady and good progress in support of the proposed risk management framework. Thus, risk assessment and management are required, and the process needs to be updated in accordance with the most recent information and knowledge on the decommissioning works.

• 원자력산업
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• 제29권5호
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• pp.52-58
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• 2009

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 추계학술발표회요약집
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• pp.341-341
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• 2002

• 방사성폐기물학회지
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• 제9권3호
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• pp.191-198
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• 2011

국내 외 수많은 수송 건식저장 시스템의 임계해석은 사용후핵연료내에 초우라늄물질(transuranic) 및 핵분열생성물(fission products) 계산의 불확실성을 이유로, 신연료로 가정된 가상연료를 적용하여 평가해왔다. 그러나 과도한 임계 여유도에 따른 경제적 손실이 크기 때문에 최근 들어 연소도이득(Burnup Credit, BUC)이 반영된 수송 건식저장 시스템의 설계 및 상용화가 추진되고 있다. 이러한 BUC 기술은 기존 임계해석 시요구되는 상수화된 불확실도와 달리 초기 농축도와 연소도 구간에 따라 상이한 불확실도를 갖게 된다. 이에 본 연구에서는 '국내 원전의 제한사항이 반영된 26다발 SNF 장전 BUC 적용 용기'(이하 BK 26 Cask)를 대상으로 관련 기술표준 및 설계요건에서 요구되는 불확실도를 평가하여 농축도 및 연소도의 함수로 계산하였다. 본 연구결과는 추후 BK 26 Cask 국내 사용후핵연료의 장전 수용률 분석의 기반자료로 활용된다.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2018년도 추계학술논문요약집
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• pp.31-32
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• 2018

• 방사성폐기물학회지
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• 제21권3호
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• pp.397-410
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• 2023

This study examined the efficacy of various chlorinating agents in partitioning light water reactor spent fuel, with the aim of optimizing the chlorination process. Through thermodynamic equilibrium calculations, we assessed the outcomes of employing MgCl₂, NH₄Cl, and Cl₂ as chlorinating agents. A comparison was drawn between using a single agent and a sequential approach involving all three agents (MgCl₂, NH₄Cl, and Cl₂). Following heat treatment, the utilization of MgCl₂ as the sole chlorinating agent resulted in a moderate separation. Specifically, this method yielded a solid separation with 96.9% mass retention, 31.7% radioactivity, and 44.2% decay heat, relative to the initial spent fuel. In contrast, the sequential application of the chlorinating agents following heat treatment led to a final solid separation characterized by 93.1% mass retention, 5.1% radioactivity, and 15.4% decay heat, relative to the original spent fuel. The findings underscore the potential effectiveness of a sequential chlorination strategy for partitioning spent fuel. This approach holds promise as a standalone technique or as a complementary process alongside other partitioning processes such as pyroprocessing. Overall, our findings contribute to the advancement of spent fuel management strategies.

• Nuclear Engineering and Technology
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• 제46권4호
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• pp.547-556
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• 2014

Korea expects a shortage in storage capacity for spent fuels at reactor sites. Therefore, a need for more metal and/or concrete casks for storage systems is anticipated for either the reactor site or away from the reactor for interim storage. For the purpose of interim storage and transportation, a dual purpose metal cask that can load 21 spent fuel assemblies is being developed by Korea Radioactive Waste Management Corporation (KRMC) in Korea. At first the gamma and neutron flux for the design basis fuel were determined assuming in-core environment (the temperature, pressure, etc. of the moderator, boron, cladding, $UO_2$ pellets) in which the design basis fuel is loaded, as input data. The evaluation simulated burnup up to 45,000 MWD/MTU and decay during ten years of cooling using the SAS2H/OGIGEN-S module of the SCALE5.1 system. The results from the source term evaluation were used as input data for the final shielding evaluation utilizing the MCNP Code, which yielded the effective dose rate. The design of the cask is based on the safety requirements for normal storage conditions under 10 CFR Part 72. A radiation shielding analysis of the metal storage cask optimized for loading 21 design basis fuels was performed for two cases; one for a single cask and the other for a $2{\times}10$ cask array. For the single cask, dose rates at the external surface of the metal cask, 1m and 2m away from the cask surface, were evaluated. For the $2{\times}10$ cask array, dose rates at the center point of the array and at the center of the casks' height were evaluated. The results of the shielding analysis for the single cask show that dose rates were considerably higher at the lower side (from the bottom of the cask to the bottom of the neutron shielding) of the cask, at over 2mSv/hr at the external surface of the cask. However, this is not considered to be a significant issue since additional shielding will be installed at the storage facility. The shielding analysis results for the $2{\times}10$ cask array showed exponential decrease with distance off the sources. The controlled area boundary was calculated to be approximately 280m from the array, with a dose rate of 25mrem/yr. Actual dose rates within the controlled area boundary will be lower than 25mrem/yr, due to the decay of radioactivity of spent fuel in storage.

• Nuclear Engineering and Technology
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• 제49권3호
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• pp.549-555
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• 2017

A lead slowing-down spectrometer (LSDS) system is under development to analyze isotopic fissile content that is applicable to spent fuel and recycled material. The source neutron mechanism for efficient and effective generation was also determined. The source neutron interacts with a lead medium and produces continuous neutron energy, and this energy generates dominant fission at each fissile, below the unresolved resonance region. From the relationship between the induced fissile fission and the fast fission neutron detection, a mathematical assay model for an isotopic fissile material was set up. The assay model can be expanded for all fissile materials. The correction factor for self-shielding was defined in the fuel assay area. The corrected fission signature provides well-defined fission properties with an increase in the fissile content. The assay procedure was also established. The assay energy range is very important to take into account the prominent fission structure of each fissile material. Fission detection occurred according to the change of the Pu239 weight percent (wt%), but the content of U235 and Pu241 was fixed at 1 wt%. The assay result was obtained with 2~3% uncertainty for Pu239, depending on the amount of Pu239 in the fuel. The results show that LSDS is a very powerful technique to assay the isotopic fissile content in spent fuel and recycled materials for the reuse of fissile materials. Additionally, a LSDS is applicable during the optimum design of spent fuel storage facilities and their management. The isotopic fissile content assay will increase the transparency and credibility of spent fuel storage.

• 방사성폐기물학회지
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• 제16권1호
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• pp.65-82
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• 2018

With the significant increase in spent ion-exchange resin generation, to meet the requirements of Waste Acceptance Criteria (WAC) of the Wolsong disposal facility in Korea, blending is considered as a method for enhancing disposal options for intermediate level waste from nuclear reactors. A mass balance formula approach was used to enable blending process with an appropriate mixing ratio. As a result, it is estimated around 44.3% of high activity spent resins can be blended with the overall volume of low activity spent resins at a 1:7.18 conservative blending ratio. In contrast, the reduction of high activity spent resins is considered a positive solution in reducing the amount of spent resins stored. In an economic study, the blending process has been proven to lower the disposal cost by 10% compared to current APR1400 treatment. Prior to commencing use of this blending method in Korea, coordinated discussion, and safety and health assessment should be undertaken to investigate the feasibility of fitting this blending method to national policy as a means of waste predisposal processing and management in the future.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 춘계학술발표회요약집
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• pp.346-346
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• 1999