• Nuclear Engineering and Technology
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• v.51 no.4
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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.

• The Journal of Engineering Geology
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• v.24 no.4
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• pp.609-613
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• 2014

The Korean Peninsula, located at the southeastern tip of the Eurasian Plate, is known to be tectonically stable, and no critical evidence has yet been found that would override the safety design of nuclear facilities in South Korea. Because a nuclear power plant, like other major social overhead capital facilities, could cause great damage to both the environment and society through an unexpected tectonic event, even one of extremely low probability, like the Fukushima accident, a defense-in-depth safety approach is required in geological and geotechnical site safety evaluation for nuclear projects. This paper introduces the regulatory procedures that are in place to confirm nuclear site safety and site monitoring (e.g., earthquakes and groundwater) systems applied to nuclear facilities in order to reduce inherent uncertainties within the site safety review of geological and seismological issues related with a NPP project.

• Nuclear Engineering and Technology
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• v.49 no.2
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• pp.286-294
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• 2017

The events at the Fukushima Daiichi Nuclear Power Station in March 2011 point out, among other matters, that concurrent accidents at multiple units of a site can occur in reality. Although site risk has been deterministically considered to some extent in nuclear power plant siting and design, potential occurrence of multi-unit accident sequences at a site was not investigated in sufficient detail thus far in the nuclear power community. Therefore, there is considerable worldwide interest and research effort directed toward multi-unit site risk assessment, especially in the countries with high-density nuclear-power-plant sites such as Korea. As the technique of probabilistic safety assessment (PSA) has been successfully applied to evaluate the risk associated with operation of nuclear power plants in the past several decades, the PSA having primarily focused on single-unit risks is now being extended to the multi-unit PSA. In this paper we first characterize the site risk with explicit consideration of the risk associated with spent fuel pools as well as the reactor risks. The status of multi-unit risk assessment is discussed next, followed by a description of the emerging issues relevant to the multi-unit risk evaluation from a practical standpoint.

• Nuclear Engineering and Technology
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• v.54 no.2
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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.

• Nuclear Engineering and Technology
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• v.38 no.6
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• pp.575-584
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• 2006

As the development of total management systems for sites along with site environmental information is becoming standard, the system known as the Site Information and Total Environmental database management System (SITES) has been developed over the last two years. The first result was a database management system for storing data obtained from facilities, and a site characterization in addition to an environmental assessment of a site. The SITES database is designed to be effective and practical for use with facility management and safety assessment in relation to Geographic Information Systems. SITES is a total management program, which includes its database, its data analysis system required for site characterization, a safety assessment modeling system and an environment monitoring system. It can contribute to the institutional management of the facility and to its safety reassessment. SITES is composed of two main modules: the SITES Database module (SDM) and the Monitoring & Assessment (M&A) module [1]. The M&A module is subdivided into two sub-modules: the Safety Assessment System (SAS) and the Site Environmental Monitoring System (SEMS). SAS controls the data (input and output) from the SITES DB for the site safety assessment, whereas SEMS controls the data obtained from the records of the measuring sensors and facilities. The on-line site and environmental monitoring data is managed in SEMS. The present paper introduces the procedure and function of the M&A modules.

• Nuclear Engineering and Technology
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• v.50 no.8
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• pp.1199-1209
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• 2018

After the Fukushima-Daiichi accident in 2011, the multi-unit risk, i.e., the risk due to several nuclear power plants (NPPs) in a site has become an important issue in several countries such as Korea, Canada, and China. However, the multi-unit risk has been discussed for a long time in the nuclear community before the Fukushima-Daiichi nuclear accident occurred. The regulatory authorities around the world and the international organizations had proposed requirements or guidelines to reduce the multi-unit risk. The concerns regarding the multi-unit risk can be summarized in the following three questions: How much the accident of an NPP in a site affects the safety of other NPPs in the same site? What is the total risk of a site with many NPPs? Will the risk of the simultaneous accidents at several NPPs in a site such as the Fukushima Daiichi accident be low enough? The multi-unit risk assessment (MURA) in an integrated framework is a practical approach to obtain the answers for the above questions. Even though there were few studies to assess the multi-unit risk before the Fukushima-Daiichi nuclear accident, there are still several issues to be resolved to perform the complete MURA. This article aims to provide an overview of the multi-unit risk issues and its assessment. We discuss the several critical issues in the current MURA to get useful insights regarding the multi-unit risk with the current state art of probabilistic safety assessment (PSA) technologies. Also, the qualitative answers for the above questions are addressed.

• Nuclear Engineering and Technology
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• v.50 no.8
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• pp.1210-1216
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• 2018

Site or multi-unit (MU) risk assessment has been a major issue in the field of nuclear safety study since the Fukushima accident in 2011. There have been few methods or experiences for MU risk assessment because the Fukushima accident was the first real MU accident and before the accident, there was little expectation of the possibility that an MU accident will occur. In addition to the lack of experience of MU risk assessment, since an MU nuclear power plant site is usually very complex to analyze as a whole, it was considered that a systematic method such as probabilistic safety assessment (PSA) is difficult to apply to MU risk assessment. This paper proposes a new MU risk assessment methodology by using the conventional PSA methodology which is widely used in nuclear power plant risk assessment. The logical failure structure of a site with multiple units is suggested from the definition of site risk, and a decomposition method is applied to identify specific MU failure scenarios.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05b
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• pp.477-482
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• 1997

Safe shutdown earthquakes (SSE) of four existing nuclear power plant sites were evaluated by using a probabilistic method. It turned out that the SSE's of the two sites are smaller than those of the rest. Site-specific response spectra were developed for two sites of which SSE's show a comparatively large difference. The result shows that the site-specific response spectrum of one site is higher by a factor of 1.5 than that of the other. The comparison of uniform hazard spectrum and site-specific response spectrum at one of the two sites shows that the both spectra are consistent with each other.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.3
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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.

• Journal of the Korean Society of Safety
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• v.33 no.6
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• pp.144-156
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• 2018

Level 3 Probabilistic Safety Assessment (PSA) is performed for the risk assessment that calculates radioactive material dispersion to the environment. This risk assessment is performed with a tool of MELCOR Accident Consequence Code System (MACCS2 or WinMACCS). For the off-site consequence analysis of multi-unit nuclear power plant (NPP) accident, the single location (Center Of Mass, COM) method has been usually adopted with the assumption that all the NPPs in the nuclear site are located at the same COM point. It was well known that this COM calculation can lead to underestimated or overestimated radionuclide concentration. In order to overcome this underestimation or overestimation of radionuclide concentrations in the COM method, Multiple Location (ML) method was developed in this study. The radionuclide concentrations for the individual NPPs are separately calculated, and they are summed at every location in the nuclear site by the post-processing of radionuclide concentrations that is based on two-dimensional Gaussian Plume equations. In order to demonstrate the efficiency of the ML method, radionuclide concentrations were calculated for the six-unit NPP site, radionuclide concentrations of the ML method were compared with those by COM method. This comparison was performed for conditions of constant weather, yearly weather in Korea, and four seasons, and the results were discussed. This new ML method (1) improves accuracy of radionuclide concentrations when multi-unit NPP accident occurs, (2) calculates realistic atmospheric dispersion of radionuclides under various weather conditions, and finally (3) supports off-site emergency plan optimization. It is recommended that this new method be applied to the risk assessment of multi-unit NPP accident. This new method drastically improves the accuracy of radionuclide concentrations at the locations adjacent to or very close to NPPs. This ML method has a great strength over the COM method when people live near nuclear site, since it provides accurate radionuclide concentrations or radiation doses.