• Journal of Nuclear Fuel Cycle and Waste Technology
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• 제1권1호
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• pp.1-7
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• 2013

The safe management of radioactive waste is a national task required for sustainable generation of nuclear power and for energy self-reliance in Korea. Since the initial introduction of nuclear power to Korea in 1978, rapid growth in nuclear power has been achieved. This large nuclear power generation program has produced a significant amount of radioactive waste, both low- and intermediate-level waste (LILW) and spent nuclear fuel (SNF); and the amount of waste is steadily growing. For the management of LILW, the Wolsong LILW Disposal Center, which has a final waste disposal capacity of 800,000 drums, is under construction, and is expected to be completed by June 2014. Korean policy about how to manage the SNF has not yet been decided. In 2004, the Atomic Energy Commission decided that a national policy for SNF management should be established considering both technological development and public consensus. Currently, SNF is being stored at reactor sites under the responsibility of plant operator. The at-reactor SNF storage capacity will run out starting in 2024. In this paper, the fundamental principles and steps for implementation of a Korean policy for national radioactive waste management are introduced. Korean practices and prospects regarding radioactive waste management are also summarized, with a focus on strategy for policy-making on SNF management.

• 한국소성가공학회:학술대회논문집
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• 한국소성가공학회 2003년도 춘계학술대회논문집
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• pp.236-239
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• 2003

The spent fuel slitting device is an equipment developed in order to feed UO$_2$pellet to the dry pulverizing/mixing device. In this study, we have compared and analyzed the handling method of the slitting and that of the pellet and hull, processing time, separating time for 20kgHM, the number of blades, on the existing slitting device using in DUPIC, and spent fuel management technology research and test facility. Also, we have compared and analyzed about an advantage and weak point, designing and producing, processing, establishment, operation, maintenance about the vertical and horizontal slitting device. Based on these results, we have developed the vertical slitting device. By using the results, we have enhanced the slitting processing time(over 40%)in comparison with DUPIC device, and it will is effectively applied to available data for designing and producing of the hot test facility.

• 방사성폐기물학회지
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• 제4권1호
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• pp.51-58
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• 2006

고리 원전 사용후핵연료 저장조의 저장용량을 확보하기 위하여 2002년부터 사용후핵연료 운반용기를 이용하여 400다발 이상의 PWR 사용후핵연료 집합체를 원전부지 내에 수송, 저장하였다. 이를 위하여 KN-12 운반용기, 관련장비 및 수송차량으로 구성되는 수송시스템을 구성하였다. KN-12 운반용기는 국내 원자력법 및 IAEA의 수송규정에 따라 설계, 제작되고, 정부로부터 인허가를 획득하였으며, 취급장비 역시 관련규정에 따라 구비하였다. 수송 저장작업은 2 대의 운반용기를 동시에 투입하여 수행하였으며, 모든 작업공정에 대하여 엄격한 품질관리 및 방사선 안전관리를 수행하여 수송 안전성을 확보하고 신뢰도를 제고하였다.

• 방사성폐기물학회지
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• 제5권4호
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• pp.339-350
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• 2007

전세계 주요 원자력선진국들의 사용후핵연료 처리에 대한 기술 및 정책현황을 알아보고 향후 우리나라의 연구방향을 제시해 보았다. 재처리 정책을 가진 소위 핵연료주기 국가들은 최근 선진핵 연료주기기술에 기초한 새로운 사용후핵 연료 관리정책을 발표하였다. 그 정책은 사용후핵연료 내에 함유된 우라늄 또는 초우란 원소들을 재순환하고 고독성의 방사성 물질 및 장반감기를 가진 물질들을 소멸하거나 단반감기 원소로 변환하는데 초점을 맞추고 있다. 이러한 정책은 원자력의 자원 활용성을 높일 뿐만 아니라, 영구 처분할 고준위폐기물의 양을 감소시켜 궁극적으로 원자력의 지속가능성을 높여 준다. PUREX 방법에 기초한 습식재처리를 우선순위로 선택한 대부분의 국가들은 이 습식방법이 건식방법에 비해 실용화에 앞서 있음을 그 선택 의 근거로 든다. 그러나 습식방법은 건식에 비해 핵확산저항성 측면에서 더욱 민감하다. 왜냐하면 이 습식방법은 약간의 공정수정에 의해 순수 플루토늄을 회수 할 수 있기 때문이다. 반면에 아직까지 실용화 단계까지는 도달해 있지 않지만 고온 용융염을 사용하는 Pyroprocess와 같은 건식처리 방법은 순수한 플루토늄을 회수 할 수 없어서 핵비확산성 측면에서 유리하며, 제4세대 원자로로 고려되는 고속로의 핵연료주기 등에도 여러 가지 이점을 가지고 있다. 따라서 우리나라의 경우 현재 이 Pyroprocess에 대한 연구가 활발히 진행되고 있다.

• 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.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2003년도 춘계학술대회 논문집
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• pp.1651-1654
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• 2003

The process equipment for handling high level radioactive materials, such as spent nuclear fuel, is operated within a sealed facility, called a hot cell, due to high radioactivity. Thus, this equipment should be maintained and repaired by remotely operated manipulator. In this study, to carry out the sale and effective maintenance of the process equipment installed in the hot cell by a servo type manipulator, a collision free motion planning method of the manipulator using virtual prototyping technology is suggested. To do this, the parts are modelled in 3-D graphics, assembled, and kinematics are assigned and the virtual workcell is implemented in the graphical environment which is the same as the real environment. The method proposed in this paper is to find the optimal path of the manipulator using the function of the collision detection in the graphic simulator. The proposed path planning method and this graphic simulator of manipulator can be effectively used in designing of the maintenance processes for the hot cell equipment and enhancing the reliability of the spent fuel management.

• 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.

• Corrosion Science and Technology
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• 제11권6호
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• pp.213-217
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• 2012

1) 선진 핵연료주기시설 관련 규제기술과 관련하여 인허가 안전심사의 경험이 없으며, 선진 핵연료주기시설 인허가를 위한 규제체계 및 안전성 평가방법 등의 개발이 필요한 단계이며 관련기기와 제반 공정에서 재료의 내식성을 평가하는 기준마련이 시급하다. 2) 선진 핵주기시설 관련 국내 기술수준을 분석하였고 그 핵심 공정인 전해환원, 전해정련, 전해제련공정의 실험변수를 조사하고 평가 필요항목을 정리하였다. 3) 전해환원과 전해정련공정의 경우 Hot-cell 내에 수분 및 산소가 일정 수준 이하로 유지되는 경우, 재료의 부식은 고려하지 않아도 되나 우라늄 잉곳 제조 공정에서 수냉 코일을 사용하게 되는 경우 물에 의한 부식을 고려해야 한다. 4) 전해 제련공정의 경우 LCC, RAR, Cd 증류공정에서 플랜지의 O-ring을 보호하기 위해 수냉 코일을 사용하게 되는 경우 물에 의한 부식을 고려해야 한다.

• 대한환경공학회지
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• 제37권5호
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• pp.312-323
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• 2015

본 논문은 국내 외 사용후핵연료 및 방사성폐기물 관리 현안 분석을 바탕으로 향후 나아갈 방향을 제시한다. 원자력 발전을 앞서 이용해 온 미국 사례를 중심으로 다양한 국가들의 처분장 확보 및 실패 사례와 최근의 관리 정책 기조를 정리하였다. 아울러, 원전 해체에 따른 고선량 방사성폐기물, 핵안보 사안 그리고 핵연료 전주기 관점에서 평가한 경제성 기반 정책 수립의 필요성을 논하였다. 사용후핵연료 및 방사성폐기물 관리의 핵심 사안을 세부적으로 중간저장, 영구처분 그리고 재처리로 분류하고 기술 검토와 인허가 체제 구축 및 연구 추진 방향성에 대한 정책 제언을 담았다.

• Nuclear Engineering and Technology
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• 제55권4호
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• pp.1439-1447
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• 2023

The fuel cycle characteristics of the IVG.1M reactor were studied within the framework of the research reactor conversion program to modernize the IVG.1M reactor. Optimum use of the nuclear fuel and reactor was achieved through routine methods which included partial fuel reloading combined with scheduled maintenance operations. Since, the additional problem in planning the fuel cycle of the IVG.1M reactor was the poisoning of the beryllium parts of the core, reflector, and control system. An assessment of the residual power and composition of spent fuel is necessary for the selection and justification of the technology for its subsequent management. Computational studies were performed using the MCNP6.1 program and the neutronics model of the IVG.1M reactor. The proposed scheme of annual partial fuel reloading allows for maintaining a high reactor reactivity margin, stabilizing it within 2-4 β_eff for 20 years, and achieving a burnup of 9.9-10.8 MW × day/kg U in the steady state mode of fuel reloading. Spent fuel immediately after unloading from the reactor can be placed in a transport packaging cask for shipping or safely stored in dry storage at the research reactor site.