• 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.

• 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.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2019년도 춘계학술논문요약집
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• pp.387-388
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• 2019

• 방사선산업학회지
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• 제17권4호
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• pp.509-518
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• 2023

The Kori Unit 1 nuclear power plant, which is planned to be dismantled after permanent shutdown, is expected to generate a large amount of various types of radioactive waste during the dismantling process. For the disposal of Very-low-level waste, which is expected to account for the largest amount of generation, the Korea Radioactive waste Agency (KORAD) is in the process of detailed design to build a 3-phase landfill disposal facility in Gyeongju. In addition, a large container is being developed to efficiently dispose of metal and concrete waste, which are mainly generated as Very low-level waste of decommissioning. In this study, based on the design characteristics of the 3-phase landfill disposal facility and the large container under development, radiation exposure dose evaluation was performed considering the normal and accident scenarios of radiation workers during operation. The direct exposure dose evaluation of workers during normal operation was performed using the MCNP computer program, and the internal and external exposure dose evaluation due to damage to the decommissioning waste package during a drop accident was performed based on the evaluation method of ICRP. For the assumed scenario, the exposure dose of worker was calculated to determine whether the exposure dose standards in the domestic nuclear safety act were satisfied. As a result of the evaluation, it was confirmed that the result was quite low, and the result that satisfied the standard limit was confirmed, and the radiational disposal suitability for the 3-phase landfill disposal facility of the large container for dismantled radioactive waste, which is currently under development, was confirmed.

• 방사성폐기물학회지
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• 제19권2호
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• pp.243-253
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• 2021

In the dismantling process of a reactor coolant system (RCS) piping, a radiation protection plan should be established to minimize the radiation exposure doses of dismantling workers. Hence, it is necessary to estimate the individual effective dose in the RCS piping dismantling process when decommissioning a nuclear power plant. In this study, the radiation exposure doses of the dismantling workers at different positions was estimated using the MicroShield dose assessment program based on the NUREG/CR-1595 report. The individual effective dose, which is the sum of the effective dose to each tissue considering the working time, was used to estimate the radiation exposure dose. The estimations of the simulation results for all RCS piping dismantling tasks satisfied the dose limits prescribed by the ICRP-60 report. In dismantling the RCS piping of the Kori-1 or Wolsong-1 units in South Korea, the estimation and reduction method for the radiation exposure dose, and the simulated results of this study can be used to implement the radiation safety for optimal dismantling by providing information on the radiation exposure doses of the dismantling workers.

• Journal of Radiation Protection and Research
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• 제48권1호
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• pp.20-27
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• 2023

Background: As the number of nuclear facilities nearing their pre-determined design life increases, demand is increasing for technology and infrastructure related to the decommissioning and decontamination (D&D) process. It is necessary to consider the nature of the dismantling environment constantly changing and the worker doing new tasks. A method was studied that can calculate the effect of learning and the change in work time on the work process, according to the learning-forgetting curve model (LFCM). Materials and Methods: The LFCM was analyzed, and input values and scenarios were analyzed for substitution into the D&D process of a nuclear facility. Results and Discussion: The effectiveness and efficiency of the training were analyzed. It was calculated that skilled workers can receive a 16.9% less collective radiation dose than workers with only basic training. Conclusion: Using these research methods and models, it was possible to calculate the change in the efficiency of workers performing new tasks in the D&D process and the corresponding reduction in the work time and collective dose.

• 방사성폐기물학회지
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• 제20권1호
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• pp.99-110
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• 2022

In decommissioning a nuclear power plant, numerous concrete structures need to be demolished and decontaminated. Although concrete decontamination technologies have been developed globally, concrete cutting remains problematic due to the secondary waste production and dispersion risk from concrete scabbling. To minimize workers' radiation exposure and secondary waste in dismantling and decontaminating concrete structures, the following conceptual designs were developed. A micro-blast type scabbling technology using explosive materials and a multi-dimensional contamination measurement and artificial intelligence (AI) mapping technology capable of identifying the contamination status of concrete surfaces. Trials revealed that this technology has several merits, including nuclide identification of more than 5 nuclides, radioactivity measurement capability of 0.1-10⁷ Bq·g^-1, 1.5 kg robot weight for easy handling, 10 cm robot self-running capability, 100% detonator performance, decontamination factor (DF) of 100 and 8,000 cm²·hr^-1 decontamination speed, better than that of TWI (7,500 cm²·hr^-1). Hence, the micro-blast type scabbling technology is a suitable method for concrete decontamination. As the Korean explosives industry is well developed and robot and mapping systems are supported by government research and development, this scabbling technology can efficiently aid the Korean decommissioning industry.

• 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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• 제15권2호
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• pp.181-189
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• 2017

방사성금속폐기물의 관리 옵션들을 안전성, 경제성, 작업자 피폭, 부피 감용 등의 선별 기준을 적용하여 비교 평가하였다. 원전 해체로부터 발생하는 금속폐기물의 관리 옵션에는 무구속 방출, 제한적 재사용, 그리고 직접 처분이 있다. 고려된 각각의 옵션들은 금속폐기물의 절단과 용융에 의한 부피감용을 수반한다. AHP기법을 적용하여 각 옵션들의 순위를 부여하였다. 방사성금속폐기물을 용융하여 금속 잉곳을 제조한 후 제한적 재이용 또는 무구속 방출하는 방안이 가장 효율적인 옵션으로 도출되었다.

• Nuclear Engineering and Technology
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• 제55권1호
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• pp.238-247
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• 2023

The present work represents a technical guideline for decommissioning a disused teletherapy machine model Theratron-780 and contains category one ⁶⁰Co radioactive source. The first section predicts the dose rate from the source in case of normal and radiological emergency situations via FLUKA-MC simulation code. Moreover, the dose assessment for the occupational during the whole process is calculated and compared to the measured values. A suggested cordoned area for safety and security in a radiological emergency is simulated. The second section lists the whole process's technical procedures, including (preview, dismantle, securing, transport and storage) of the disused teletherapy machine. Results show that the maximum obtained accumulated dose for occupational were found to be 24.5 ± 4.9 μSv in the dismantle and securing process in addition to 3.5 ± 1.8 μSv during loading on the transport vehicle and unloading at the storage facility. It was found that the measured accumulated dose for workers is in good agreement with the estimated one by uncertainty not exceeding 5% in normal operating conditions.