• Journal of the Korean Society of Marine Environment & Safety
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• v.27 no.5
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• pp.618-629
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• 2021

The Fukushima nuclear power plant (NPP) accident caused by the East Japan Earthquake in 2011 and the recent increase in the frequency of earthquakes in Korea have caused safety concerns regarding radionuclide exposure. In addition, the Tokyo Electric Power Company (TEPCO) in Japan recently decided to release radionuclide-contaminated water from Fukushima's NPP into the Pacific Ocean, raising public concerns that the possibility of radionuclide contamination through both domestic- and foreign fishery products is increasing. Although many studies have been conducted on the input of artificial radionuclides into the Pacific after the Fukushima NPP accident, studies on the distribution and accumulation of artificial radionuclides in marine products from East Asia are lacking. Therefore, in this study, we attempted to explore recent research on the distribution of artificial radionuclides (e.g., ¹³⁷Cs, ^239+240Pu, ⁹⁰Sr, and etc.) in marine products from Korean seas after the Fukushima NPP accident. In addition, we also discuss future research directions as it is necessary to prepare for likely radiation accidents in the future around Korea associated with the new nuclear facilities planned by 2030 in China and owing to the discharge of radionuclide-contaminated water from the Fukushima NPP.

• Nuclear Engineering and Technology
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• v.52 no.6
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• pp.1231-1242
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• 2020

The Kori Unit 1 will be decommissioned after a permanent shutdown in June 2017. South Korea has a 0.1 mSv/yr exposure limit standard for limited or unlimited site release. This is South Korea's first commercial NPP; therefore, if the containment building is reused as a memorial hall, it will contribute to the improvement of public understanding and enhance the public's acceptance of NPPs. Also, existing Kori Unit 1 nuclear power plant manpower resources can be reused after decommissioning and resident staff and memorial hall visitors can activate nearby commercial areas. Therefore, such a reuse scenario may also prevent an economic recession. The exposure dose was calculated using the following scenarios: worker in the containment building, visitor in the containment building, and worker in buildings other than the containment building. The exposure dose in the buildings was calculated by the RESRAD-BUILD developed by the Argonne National Laboratory (ANL). The preliminary exposure dose and derived concentration guideline level (DCGL) were derived.

• Nuclear Engineering and Technology
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• v.47 no.1
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• pp.1-10
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• 2015

After the Fukushima Daiichi nuclear power plant (NPP) accident, new regulatory requirements were enforced in July 2013 and a backfit was required for all existing nuclear power plants. It is required to take measures to prevent severe accidents and mitigate their radiological consequences. The Regulatory Standard and Research Department, Secretariat of Nuclear Regulation Authority (S/NRA/R) has been conducting numerical studies and experimental studies on relevant severe accident phenomena and countermeasures. This article highlights fission product (FP) release and hydrogen risk as two major areas. Relevant activities in the S/NRA/R are briefly introduced, as follows: 1. For FP release: Identifying the source terms and leak mechanisms is a key issue from the viewpoint of understanding the progression of accident phenomena and planning effective countermeasures that take into account vulnerabilities of containment under severe accident conditions. To resolve these issues, the activities focus on wet well venting, pool scrubbing, iodine chemistry (in-vessel and ex-vessel), containment failure mode, and treatment of radioactive liquid effluent. 2. For hydrogen risk: because of three incidents of hydrogen explosion in reactor buildings, a comprehensive reinforcement of the hydrogen risk management has been a high priority topic. Therefore, the activities in evaluation methods focus on hydrogen generation, hydrogen distribution, and hydrogen combustion.

• Journal of the Korean Society of Radiology
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• v.11 no.1
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• pp.1-8
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• 2017

The U.S. NRC establishes a radiological criteria with regard to restricted or unrestricted use of nuclear plant site after decommissioning in NUREG-1757. According to this, a nuclear plant site can be released in a restricted way or unrestricted way only if a licensee demonstrates that the dose criteria is fulfilled after the site decontamination and remediation. In order to prove compliance with the radiological criteria of site release, LTP(License Termination Plan) must include the site release criteria, site characterization, final survey plan with major radionuclides and DCGL(Derived Concentration Guideline Levels), etc. Based on the decommissioning case of Rancho Seco nuclear power plant in the United States, this paper analyzed a method of setting the DCGL that can be applied to Kori NPP Unit 1 which will be permanently disabled in 2017.

• Proceedings of the Earthquake Engineering Society of Korea Conference
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• 2000.04a
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• pp.3-16
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• 2000

The ultimate safety-goal of nuclear power plants should be targeted at preventing release of nuclear radiation compared to general structures, Accordingly the phases of siting design construction and operation of NPPs are severely regulated by codes of aseismic design so as to assure safety of NPPs. To accomplish this goal strict quality assurace and seismic qualification tests should be conducted for all phases of NPP construction. In addition seismic monitoring systems should be installed and always in operation to provide proper post-earhquake procedures. Besides periodic safety review should be performed during operation along with the seismic margin assessment. In this paper general procedures to secure seismic safety of NPPs are systematically reviewed and additional considerations for improvement are suggested.

• Nuclear Engineering and Technology
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• v.34 no.2
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• pp.176-185
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• 2002

A 4-bundle shift refuelling method of CANDU6 F/H (Fuel Handling) System is analyzed to assess the operational flexibility and capacity of F/H system. The current 8-bundle shift refuelling scheme requires to refuel eight fuel bundles from a single fuel channel, and to refuel 14 fuel channels in a week on average assuming that the reactor is in a steady state. The analysis showed that the 4-bundle shift refuelling method increases F/M (Fuelling Machine) duty cycle and operator load. However, the fuellin’g method change from the 8- to 4-bundle shift refuelling ill not require additional team of operators. A marginal increase in the maintenance cost may be resulted in by the change of fuelling method and the increase of fatigue usage factors requires some components to be replaced during the FM lifetime.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.2
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• pp.247-253
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• 2023

According to NSSC Notice No. 2021-10, safety analysis needs to be introduced in the decommissioning plan. Public and occupational dose analyses should be conducted, specifically for unexpected radiological accidents. Herein, based on the risk matrix and analytic hierarchy process, the method of selecting accident scenarios during the decommissioning of nuclear power plants has been proposed. During decommissioning, the generated spent resin exhibits relatively higher activity than other generated wastes. When accidents occur, the release fraction varies depending on the conditioning method of radioactive waste and type of radioactive nuclides or accidents. Occupational dose analyses for 2 (fire and drop) among 11 accident scenarios have been performed. The radiation doses of the additional exposures caused by the fire and drop accidents are 1.67 and 4.77 mSv, respectively.

• Journal of Radiation Protection and Research
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• v.44 no.3
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• pp.118-126
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• 2019

Background: Government conducts environmental radioactivity surveillance for verification purpose around nuclear facilities based on the Nuclear Safety Law and issues a surveillance report every year. This study aims to evaluate the short and the long-term fluctuation of radionuclides detected above MDC and their origins using concentration ratios between these radionuclides. Materials and Methods: Sample media for verification surveillance are air, rainwater, groundwater, soil, and milk for terrestrial samples, and seawater, marine sediment, fish, and seaweed for marine samples. Gamma-emitting radionuclides including $^{137}Cs$, $^{90}Sr$, Pu, $^3H$, and $^{14}C$ are evaluated in these samples. Results and Discussion: According to the result of the environmental radioactivity verification surveillance in the vicinity of nuclear power facilities in 2017, the anthropogenic radionuclides were not detected in most of the environmental samples except for the detection of a trace level of $^{137}Cs$, $^{90}Sr$, Pu, and $^{131}I$ in some samples. Radioactivity concentration ratios between the anthropogenic radionuclides ($^{137}Cs/^{90}Sr$, $^{137}Cs/^{239+240}Pu$, $^{90}Sr/^{239+240}Pu$) were similar to those reported in the environmental samples, which were affected by the global fallout of the past nuclear weapon test, and Pu atomic ratios ($^{240}Pu/^{239}Pu$) in the terrestrial sample and marine sample showed significant differences due to the different input pathway and the Pu source. Radioactive iodine ($^{131}I$) was detected at the range of < $5.6-190mBq{\cdot}kg-fresh^{-1}$ in the gulfweed and sea trumpet collected from the area of Kori and Wolsong intake and discharge. A high level of $^3H$ was observed in the air (Sangbong: $0.688{\pm}0.841Bq{\cdot}m^{-3}$) and the precipitation (Meteorology Post: $199{\pm}126Bq{\cdot}L^{-1}$) samples of the Wolsong nuclear power plant (NPP). $^3H$ concentration in the precipitation and pine needle samples showed typical variation pattern with the distance and the wind direction from the stack due to the gaseous release of $^3H$ in Wolsong NPP. Conclusion: Except for the detection of a trace level of $^{137}Cs$, $^{90}Sr$, Pu, and $^{131}I$ in some samples, anthropogenic radionuclides were below MDC in most of the environmental samples. Overall, no unusual radionuclides and abnormal concentration were detected in the 2017's surveillance result for verification. This research will be available in the assessment of environment around nuclear facilities in the event of radioactive material release.

• Journal of Radiation Protection and Research
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• v.40 no.4
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• pp.267-276
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• 2015

Several analyses on tritium that is the largest release of gas or liquid radioactive waste from domestic PWR and PHWR NPPs were carried out, such as release comparison, directional frequency of wind and tritium behavior changes in environmental samples. First of all, analysis result showed that tritium released from PHWR was more than ten times as gas and double to three times as liquid in comparison to PWR in 2013. Independent release management in NPP units is needed to precisely control and analyze tritium, since there were 2 units of some NPPs having the same amount of release during analysis. In analysis on frequency of wind direction, average range showed 1.7 to 11.5% by 16-point compass. In case of analysis on sampling points by wind direction, Result showed most of the sampling points are right in places. However, There are some areas needed to examine. In analysis on tritium concentration changes in environmental samples, tritium concentration near NPPs was higher than one far away from NPPs. In case of environmental samples far from PWR, a trace of tritium occur. While, tritium concentration near NPPs was more than or equal to one further from PHWR. In conclusion, tritium occurs considerably in PHWR and is lower than standard in samples. but, it is still detected. Therefore, it is needed to strengthen control in system in NPPs and to consistently monitor tritium in environment.

• Nuclear Engineering and Technology
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• v.50 no.8
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• pp.1234-1245
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• 2018

The risk of multi-unit nuclear power plants (NPPs) at a site has received considerable critical attention recently. However, current probabilistic safety assessment (PSA) procedures and computer code do not support multi-unit PSA because the traditional PSA structure is mostly used for the quantification of single-unit NPP risk. In this study, the main purpose is to develop a multi-unit Level 2 PSA method and apply it to full-power operating six-unit OPR1000. Multi-unit Level 2 PSA method consists of three steps: (1) development of single-unit Level 2 PSA; (2) extracting the mapping data from plant damage state to source term category; and (3) combining multi-unit Level 1 PSA results and mapping fractions. By applying developed multi-unit Level 2 PSA method into six-unit OPR1000, site containment failure probabilities in case of loss of ultimate heat sink, loss of off-site power, tsunami, and seismic event were quantified.