• Nuclear Engineering and Technology
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• 제51권4호
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• pp.1098-1108
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• 2019

Korea's first commercial nuclear power plant at Kori site was permanently shut down in 2017 and is currently in transition stage. Preparatory activities for decommissioning such as historical site assessment, characterization, and dismantling design are being actively carried out for successful D&D (Dismantling and Decontamination) at Kori site. The ultimate goal of decommissioning will be to ensure the safety of workers and residents that may arise during the decommissioning of nuclear facilities and, thereby finally returning the site to its original status in accordance with the release criteria. Upon completion of decommissioning, the resident's safety at a site released will be assessed from the evaluation of dose caused by radionuclides expected to be present or detected at the site. Although the U.S. commercial nuclear power plants with decommissioning experience use different site release criteria, most of them are 0.25 mSv/y. In Korea, both the unrestricted and restricted release criteria have been set to 0.1 mSv/y by the Nuclear Safety and Security Commission. However, since the dose is difficult to measure, measurable concentration guideline levels for residual radionuclides that result in dose equivalent to the site release criteria should be derived. For this derivation, site reuse scenario, selection of potential radionuclides, and systematic methodology should be developed in planning stage of Kori site decommissioning. In this paper, for calculation of a preliminary site-specific Derived Concentration Guideline Levels (DCGLs) for the Nuclear Power Plant site, a novel approach has been developed which can fully reflect practical reuse plans of the Kori site by taking into account multiple site reuse scenarios sequentially, thereby striking a remarkable distinction with conventional approaches which considers only a single site scenario.

• 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.

• 에너지공학
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• 제16권3호
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• pp.93-102
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• 2007

본 논문에서는 해체 시나리오를 정량적 및 정성적 고려사항을 반영하여 평가하기 위하여 계층적분석이론(Analytic Hierarchy Process, AHP)을 이용한 평가모델을 개발하였으며 또한 해체 시나리오의 정량적인 자료산출을 위하여 해체일정, 폐기물량, 방사화 가시화, 해체비용, 작업자 피폭량 등과 같은 해체정보산출모듈을 개발하였다. 그리고 해체공정을 가상환경에서 구현하여 해체절차를 파악하기 위하여 디지털 목업(Digital Mock-Up, DMU)을 개발하였으며 DMU 시스템은 해체정보산출모듈, 해체 DB 및 해체 시나리오 평가 모듈을 통합적으로 관리하도록 개발되었다. 마지막으로 개발된 해체 DMU 시스템과 계층분석과정 모델을 연구로 1호기(Korea Research Reactor-1, KRR-1) thermal column의 플라즈마 절단 시나리오와 nibbler 절단 시나리오에 적용하여 비교 평가하였다.

• 방사성폐기물학회지
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• 제19권3호
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• pp.297-306
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• 2021

In this study, basic strategies for the decommissioning and site remediation of the Fukushima Daiichi Nuclear Power Station (FDNPS) were investigated. Six scenarios were formulated based on two of the three decommissioning strategies of nuclear power plants defined by the International Atomic Energy Agency (IAEA): immediate dismantling and deferred dismantling. A multicriteria decision analysis was performed to analyze the preferences of the options from the viewpoints of the timeframe to complete decommissioning, the resulting waste, the site usability, and the availability of the radioactive waste disposal route. The same six scenarios were applied to both the FDNPS and the nuclear power plants that ceased operation after a normal plant life cycle for comparison. For the FDNPS, the decommissioning project involved fuel debris retrieval, dismantling, and site remediation. The analysis results suggest that the balance between the amount of waste and the time to achieve the end state may be one of the most critical factors to consider when planning the decommissioning and site remediation of the FDNPS.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3559-3570
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• 2023

In the decommissioning of nuclear facilities, Derived Concentration Guideline Level (DCGL) derivation is necessary for the release of the facility after the site remediation, which also needs to be implemented in the stage of establishing a decommissioning planning. In order to derive DCGL, the dose assessment for the receptors can be conducted from residual radioactivity by using RESRAD code. When performing sensitivity analysis on probabilistic parameters, secondary evaluation is performed by assigning a single value for parameters classified as sensitive. However, several options may arise in the handling of nonsensitive parameters. Therefore, we compared the results of the first execution of RESRAD applying probabilistic parameters for each scenario with the results of the second execution applying a single value to sensitive parameters among the probabilistic parameters. In addition, we analyzed the effect of setting options for non-sensitive parameters. As a result, the effect on DCGL were different depending on the application scenario, the target radionuclides, and the input parameter selections. In terms of the overall evaluation period, the DCGL graph of the default option was generally shown as the most conservative except for some radionuclides. However, it will not necessarily be given priority in the aspect of the need to reflect site characteristics. The reason for selecting a probabilistic parameter is the availability of the parameter and the uncertainty of applying a single value. Therefore, as an alternative, it can be consistently applied to distribution as an option for non-sensitive parameters after sensitivity analysis.

• Nuclear Engineering and Technology
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• 제55권11호
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• pp.4181-4190
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• 2023

Nuclear power plants are subjected to various processes during decommissioning, including cutting, decontamination, disposal, and treatment. The cutting of massive bio-shields is a significant step in the decommissioning process. Cutting is performed near the target structure, and during this process, workers are exposed to potential radioactive elements. However, studies considering worker exposure management during such cutting operations are limited. Furthermore, dismantling a nuclear power plant under certain circumstances may result in the unnecessary radiation exposure of workers and an increase in secondary waste generation. In this study, a cutting scenario was formulated considering the bio-shield as a representative structure. The specifications of a standard South Korean radioactive waste disposal drum were used as the basic conditions. Additionally, we explored the hot-to-cold and cold-to-hot methods, with and without the application of polishing during decontamination. For evaluating various scenarios, different cutting time points up to 30 years after permanent shutdown were considered, and cutting speeds of 1-10nullm²/h were applied to account for the variability and uncertainty attributable to the design output and specifications. The obtained results provide fundamental guidelines for establishing cutting methods suitable for large structures.

• Journal of Radiation Protection and Research
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• 제45권2호
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• pp.81-87
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• 2020

Background: Kori unit #1 is permanently shut down after a 40-year lifetime. The Nuclear Safety and Security Commission recommends establishing initial decommissioning plans for all nuclear and radwaste treatment facilities. Therefore, the Korea Atomic Energy Research Institute (KAERI) must establish an initial and final decommissioning plan for radwaste-treatment facilities. Radiation safety assessment, which constitutes one chapter of the decommissioning plan, is important for establishing a decommissioning schedule, a strategy, and cost. It is also a critical issue for the government and public to understand. Materials and Methods: This study provides a method for assessing external radiation dose to workers during decommissioning. An external dose is calculated following each exposure scenario, decommissioning strategy, and working schedule. In this study, exposure dose is evaluated using the deterministic method. Physical characterization of the facility is obtained by both direct measurement and analysis of the drawings, and radiological characterization is analyzed using the annual report of KAERI, which measures the ambient dose every month. Results and Discussion: External doses are calculated at each stage of a decommissioning strategy and found to increase with each successive stage. The maximum external dose was evaluated to be 397.06 man-mSv when working in liquid-waste storage. To satisfy the regulations, working period and manpower must be managed. In this study, average and cumulative exposure doses were calculated for three cases, and the average exposure dose was found to be about 17 mSv/yr in all the cases. Conclusion: For the three cases presented, the average exposure dose is well below the annual maximum effective dose restriction imposed by the international and domestic regulations. Working period and manpower greatly affect the cost and entire decommissioning plan; hence, the chosen option must take account of these factors with due consideration of worker safety.

• 방사성폐기물학회지
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• 제18권3호
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• pp.347-354
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• 2020

Site characterization for decommissioning Kori Unit 1 is ongoing in South Korea after 40 years of successful operation. Kori Unit 1's containment building is assumed to be mostly radioactively contaminated, and therefore radiation exposure management and detailed contamination investigation are required for decommissioning and dismantling it safely. In this study, site-specific Derived Concentration Guideline Levels (DCGLs) were derived using the residual radioactivity risk evaluation tool, RESRAD-BUILD code. A conceptual model of containment building for Kori Unit 1 was set up and limited occupational worker building inspection scenario was applied. Depending on the source location, the maximum contribution source and exposure pathway of each radionuclide were analyzed. The contribution of radionuclides to dose and exposure pathways, by source location, is expected to serve as basic data in the assessment criteria of survey areas and classification of impact areas during further decommissioning and decontamination of sites.

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1781-1788
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• 2022

The containment building Kori Unit 1 may require sequential steps for full decommissioning. This study assumes that the containment building is to be used as an auxiliary building that handles nuclear power systems and materials during decommissioning before conversion into a greenfield. Through the derivation of guidelines and dose evaluation, it was confirmed whether the radiation workers were satisfied with the ALARA decision. The specific modeling of the external radiation exposure was performed based on the facility investigation procedures. The external radiation specific derived concentration guideline levels (DCGLs) for radiation workers in containment building were obtained using the RESRAD-BUILD code and were applied to the VISIPLAN 3D ALARA Planning Tool code to calculate the working dose and check worker safety. The derivation of site-specific and realistic DCGLs and dose evaluation via 3D modeling can contribute to the scenario development for the decommission and remediation of containment building.

• Nuclear Engineering and Technology
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• 제50권4호
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• pp.606-612
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• 2018

Many nuclear power plants in Japan are approaching the end of their planned operational life spans. They must be decommissioned safely in the near future. Using augmented reality (AR), workers can intuitively understand information related to decommissioning work. Three-dimensional (work-site) reconstruction models of dismantling fields are useful for workers to observe the conditions of dismantling field situations without visiting the actual fields. This study, based on AR and work-site reconstruction models, developed and evaluated an information reference system. The evaluation consists of questionnaires and interview surveys administered to six nuclear power plant workers who used this system, along with a scenario. Results highlight the possibility of reducing time and mitigating mistakes in dismantling fields. Results also show the ease of referring to information in dismantling fields. Nevertheless, it is apparently difficult for workers to build reconstruction models of dismantling fields independently.