• 방사성폐기물학회지
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• 제17권1호
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• pp.95-106
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• 2019

원전 해체를 준비함에 있어 정성적 또는 정량적 위험도 평가는 필수요소이다. 해체 공정간 발생하는 방사선학적 및 비방사선학적 위험요소는 해체 작업자 및 대중의 안전을 보장하기 위해 사전에 평가되어야 한다. 현재 해체 경험이 많은 미국의 기존 사업자들 및 NRC의 경우 위험의 중대성만 평가하는 결정론적 위험도 평가에 집중하고 있다. 하지만 최근 IAEA는 위험도 매트릭스를 활용한 위험도평가를 결정론적 위험도 평가의 대체안으로 제안하고 있다. 따라서 본 연구에서는 위험도평가에 앞서 해체 공정 별 해체 활동을 Risk Breakdown Structure에 맞추어 정리하였고, 미국 20여개 해체 원전에서 해체 공정별 위험도 평가 시행 중 선정한 해체 활동간 잠재적 사고를 해체 활동에 맞게 체계적으로 정리하였다. 그리고 복합 리스크 매트릭스를 개발 및 활용하여 해체 공정간 방사선학적 및 비방사선학적 위험요소의 위험도를 평가하여 정량적으로 수치화 하였다.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2018년도 추계학술논문요약집
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• pp.245-246
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• 2018

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2018년도 춘계학술논문요약집
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• pp.289-290
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• 2018

• 방사성폐기물학회지
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• 제19권1호
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• pp.51-58
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• 2021

Chemical decontamination of primary systems in a nuclear power plant (NPP) prior to commencing the main decommissioning activities is required to reduce radiation exposure during its process. The entire process is repeated until the desired decontamination factor is obtained. To achieve improved decontamination factors over a shorter time with fewer cycles, the appropriate flow characteristics are required. In addition, to prepare an operating procedure that is adaptable to various conditions and situations, the transient analysis results would be required for operator action and system impact assessment. In this study, the flow characteristics in the steady-state and transient conditions for the chemical decontamination operations of the Kori-1 NPP were analyzed and compared via the MARS-KS code simulation. Loss of residual heat removal (RHR) and steam generator tube rupture (SGTR) simulations were conducted for the postulated abnormal events. Loss of RHR results showed the reactor coolant system (RCS) temperature increase, which can damage the reactor coolant pump (RCP)s by its cavitation. The SGTR results indicated a void formation in the RCS interior by the decrease in pressurizer (PZR) pressure, which can cause surface exposure and tripping of the RCPs unless proper actions are taken before the required pressure limit is achieved.

• 방사성폐기물학회지
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• 제15권1호
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• pp.45-63
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• 2017

한국의 가장 오래된 상업 원전인 고리 1호기가 2017년에 해체가 이루어질 예정이다. 원전 해체 폐기물의 적절한 처리는 효율적인 원전해체에 있어 중요한 역할을 할 것이다. 특히, 저준위 또는 오염되지 않은 금속폐기물의 재활용은 폐기물 발생 저감은 물론 처분장의 공간을 절약하는데 기여할 것이다. 본 논문은 재활용 시스템의 개념설계와 정의된 업무 흐름에서 발생하는 피폭 선량을 평가하는데 그 목적이 있다. 작업의 흐름과 운전 개념을 정립하기 위해 다양한 형태의 다이어그램을 설계하였다. 선량평가에 필요한 시나리오는 개념설계를 기반으로 선정되었으며, RESRAD-RECYCLE을 이용하여 선량을 평가하였다. 이를 통하여, 결정적 시나리오 선별, 핵종 특성 및 핵종 분배가 선량에 미치는 영향을 분석하였다. 더 나아가, 선량분석은 피폭 시나리오에 대한 대체 방안 수립, 필요한 제염 및 방사선방어 프로세스 그리고 허용 방사능 검토의 정보를 제공하는데 사용 될 수 있을 것이다.

• 방사성폐기물학회지
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• 제19권1호
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• pp.61-74
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• 2021

A paradigm shift in the government's energy policy was reflected in its declaration of early closure of old nuclear plants as well as cancellation of plans for the construction of new plants. To this end, unit 1 of Kori Nuclear Power Plant was permanently shut down and is set for decommission. Based on these changes, the off-site transport of spent fuels from nuclear power plants has become a critical issue. The purpose of this study is to develop an optimized method for transportation of spent fuels from Kori Nuclear Power Plant's units 1, 2, 3, and 4 to an assumed interim storage facility by simulating the scenarios using the Flexsim software, which is widely used in logistics and manufacturing applications. The results of the simulation suggest that the optimized transport methods may contribute to the development of delivery schedule of spent fuels in the near future. Furthermore, these methods can be applied to decommissioning plan of nuclear power plants.

• 방사성폐기물학회지
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• 제13권3호
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• pp.187-199
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• 2015

원전해체시장이 본격적으로 도래함에 따라 그에 따른 기술연구가 부각되고 있다. 그러한 기술 중 방사선 제염은 직접적인 원전해체 과정 중 가장 초반에 행해지는 작업으로 현장 근로자의 안전확보 및 폐기물 양 감소를 위해 수행되는 중요한 작업이다. 제염을 통해 폐기물 표면에 존재하는 방사선 물질을 제거하게 되는데 해체에 적용되는 제염기술은 보다 강한 매개체를 사용하거나 개선된 설비를 활용하여 표면층 제거 정도가 일반적인 제염보다 훨씬 크다. 따라서 제염 계획 수립시 다양한 관점에서 분석 방법이 필요하다. 본 연구에서는 제염기술 선정을 위해 고려해야 할 요인을 설명하였으며, 대표적인 제염기술 사례 분석을 통해 실제 기술 수행을 위해 원전 설비 내 제염 아이템 선정 및 제염 장비 활용을 위해 검토해야 할 사항을 제시하였다.

• 방사성폐기물학회지
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• 제19권1호
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• pp.9-27
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• 2021

As an example of research activities in decontamination for decommissioning, new data are presented on the options for corrosion layer dissolution during the decommissioning decontamination, or persulfate regeneration for decontamination solutions re-use. For the management of spent decontamination solutions, new method based on solvent extraction of radionuclides into ionic liquid followed by electrodeposition of the radionuclides has been developed. Fields of applications of composite inorganic-organic absorbers or solid extractants with polyacrylonitrile (PAN) binding matrix for the treatment of liquid radioactive waste are reviewed; a method for americium separation from the boric acid containing NPP evaporator concentrates based on the TODGA-PAN material is discussed in more detail. Performance of a model of radionuclide transport, developed and implemented within the GoldSim programming environment, for the safety studies of the LLW/ILW repository is demonstrated on the specific case of the Richard repository (Czech Republic). Continuation and even broadening of these activities are expected in connection with the approaching end of the lifespan of the first blocks of the Czech NPPs.

• Nuclear Engineering and Technology
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• 제54권2호
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• pp.493-500
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• 2022

The objective of this study was to derive derived concentration guideline levels (DCGLs) reflecting the site-specific characteristics of KRR-1&2. A total of 7 nuclides (H-3, C-14, Co-60, Sr-90, Cs-137, Eu-152, and Eu-154) were selected for DCGLs derivation. Radiation dose at the sites was evaluated with RESRAD-ONSITE program. The dose contribution due to direct external exposure was the highest during the entire evaluation period. Ingestion had the second effect. The DCGLs of Co-60 was derived to be 0.051 Bq/g, and DCGLs of Cs-137 was 0.193 Bq/g. The DCGLs of H-3 showed the highest value of 129 Bq/g. The ratio of DCGLs derived by applying site-specific values and default values ranged from 0.27 to 19.6. For six nuclides excluding H-3, KRR-1&2 sites and the overseas NPP sites showed similar DCGLs. H-3 showed large differences in DCGLs from this study and overseas NPPs. The large difference resulted from input parameter values applied to the sites. In conclusion, it is critical to apply site-specific parameter values reflecting the site characteristics to derive DCGLs for decommissioned site clearance. The result of this study can be used as a reference for nuclide selection and DCGLs derivation reflecting the site characteristics when decommissioning nuclear facilities, including nuclear power plants in Korea.

• 한국시스템다이내믹스연구
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• 제15권1호
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• pp.51-74
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• 2014

Despite the high efficiency of nuclear power plant, people in Korea do not give approvals and supports the facilities because the risk of the accidents and incidents. In particular, the low level of safety culture is a crucial mechanism that damages the robustness of the NPP. By considering the various definitions of safety culture and analyzing the major reasons of incidents, the conceptual safety culture model is made by using Causal Loop Diagramming. For sustaining development of nuclear power, social supports, incentives and organizational learning are needed. It also requires the coordination of work schedules and the expansion of human resource for protecting the rules and procedures in NPP. Decommissioning aging nuclear power plants will prevent a serious accident. In order to promote the safety culture, Korea Hydro & Nuclear Power Corporation should disclose more information to the public and promote the internal and external communications.