Korean Nuclear Power Plants Research Articles

Design challenges and response plans for intelligent decision support systems for Korean nuclear power plants under normal and abnormal conditions

In nuclear power plants (NPPs), an intelligent decision support system (IDSS) aids the decision-making process of main control room operators. It achieves this by monitoring and diagnosing conditions, predicting progress, and providing preventive advice during both normal and abnormal operations. Despite ongoing research in Korea, few IDSS for NPPs have been effectively applied and verified. Furthermore, established designs and validation guidelines for IDSSs for Korean NPPs remain lacking. To address this gap, this study systematically identifies problems and corresponding countermeasures for applying artificial intelligence-based design and validation technologies, while complying with licensing regulatory standards. The aim is to develop a practical and effective IDSS. This study identifies design challenges in the practical application of fundamental IDSS technology, outlining necessary design and validation technologies. Additionally, it suggests technology-specific response plans for the identified design challenges.

Do procedural fairness and conflict perception affect the acceptance of the temporary nuclear spent fuel storage facility? Evidence from South Korea

This study uses data collected from the residents near South Korean nuclear power plants to explore the complex psychological factors that may influence the local acceptance of the expansion of temporary nuclear spent fuel (NSF) storage facilities in South Korea. Employing structural equation modeling, we examine the roles of perceived procedural fairness and perceived local conflicts in the traditional model of trust, knowledge, and perceptions of risks and benefits. Our findings demonstrate that perceived procedural fairness significantly influences trust, which further mediates the relationship between perceived local conflicts and benefit and risk perceptions, ultimately shaping local acceptance of temporary NSF storage facilities. Our findings underscore the importance of perceived procedural fairness and trust in shaping local acceptance and highlight perceived local conflicts’ role in amplifying the perceptions of benefits and risks. These findings inform new policy implications for the local acceptance of expanding temporal NSF storage facilities.

Radionuclides deposition on sea surface following hypothetical accidental releases at Korean nuclear power plant

Radionuclides deposition on sea surface following hypothetical accidental releases at Korean nuclear power plant

High-radiation-exposure work in Korean pressurized water reactors

Owing to strict radiation safety management in Korean nuclear power plants (NPPs), most radiation workers receive very low radiation doses, even lower than the annual dose limit for the general public. However, the occupational dose distribution indicates that some Korean NPP workers receive a relatively higher dose than the average dose. This inequity in radiation exposure could be reduced by providing customized radiation protection measures, such as dose constraints, to workers receiving relatively higher doses. In this study, dose normalization was performed to identify the highest radiation exposure work in Korean pressurized water reactors (PWRs). The results show that most of the occupational exposure in Korean PWRs occurs during the planned maintenance period. Finally, the three highest radiation exposure tasks in Korean PWRs were identified: nozzle dam installation and removal, eddy current testing, and man-way opening and closing.

A methodology to designate radiation‐controlled areas in decommissioning nuclear power plants

AbstractKori Unit 1, which was permanently shut down in 2017, is expected to be the first decommissioned commercial nuclear power plant (NPP) in Korea. Operating NPPs designate radiation‐controlled areas (RCAs) to protect workers from radiological hazards and provide appropriate measures. Decommissioning RCAs (DRCAs) in NPPs being decommissioned should also be designated by considering the radiological characteristics of workplaces and exposure routes of decommissioning workers, such as external and internal exposure. However, the criteria for the DRCAs in decommissioning Kori Unit 1 are presently the same as those during normal operation, according to the public‐hearing material of the final decommissioning plan for Kori Unit 1. This study analyzed criteria for RCAs in all Korean NPPs to propose criteria for DRCAs in decommissioning NPPs. Analysis results show that RCAs are classified into four, five, six, and eight zones in Korean pressurized water reactors with respective external radiation dose rates. The RCAs in Korean pressurized heavy water reactors are classified into six zones with external dose rates and derived air concentrations of tritium. This study proposes three classifications of DRCAs to keep occupational exposure as low as reasonably achievable while considering potential exposure routes, that is, external exposure, airborne contamination, and surface contamination areas. First, the external exposure areas are classified into several zones according to the levels of external radiation dose rates. Second, the airborne contamination areas are designated to prevent internal exposure to workers from the inhalation of radioactive material in the air during cutting and demolition. Last, the surface contamination areas are designated to minimize the skin contamination of workers as radioactive dust in the air is deposited on the surface of facilities and equipment over time.

Multi-Hazard Risk Assessment of South Korean Nuclear Power Plants

Multi-hazard risk assessments of potential earthquake-triggered tsunamis with the positive contribution of climate change-related sea level rise (SLR) are performed for Hanul and Shin-Hanul (originally the Uljin), Kori and Shin-Kori, and Wolsong and Shin-Wolsong Nuclear Power Plants (NPPs) throughout the 21st century. Logic tree approach is used to construct probabilistic tsunami hazard model for evaluating the inundation levels at the coast of each NPP. Hypothetical earthquake sources are generated throughout the East Sea (Sea of Japan) using Monte Carlo Simulations. Epistemic uncertainty of the sea level rises and aleatory variability of the tsunami hazards is considered for stochastic multi-hazard assessment. Tsunami simulations are performed using revised bathymetric levels based on SLR projections of different Shared Socioeconomic Pathways (SSP) revealed by the Intergovernmental Panel on Climate Change (IPCC). Tsunami hazard curves are presented to determine the level of the adverse effects of SLR on tsunami inundations for the projected years. Environmental risk assessment is conducted by evaluating multi-hazard curves. The results show that the effect of SLR will be extremely significant on tsunami inundation levels, especially for the worst-case scenario estimated by IPCC. Depending on the multi-hazard risk assessments, Hanul NPP is the only site to survive against the worst-case scenario. Wolsong and Shin Wolsong NPP is under moderate multi-hazard risk. Whereas, Kori and Shin-Kori NPP may encounter a destructive multi-hazard environmental risk according to the analysis conducted in this study.

Loading pattern design and economic evaluation for 24-month cycle operation of OPR-1000 in Korea

Due to the tightened regulatory environment since the Fukushima accident, the capacity factor of Korean nuclear power plants has been declining since 2011. To overcome this circumstance, a shift from 18-month to 24-month cycle operation is being considered in Korea. Therefore, in this study, loading patterns(LPs) for 24-month cycle operation of the Korean standard nuclear power plant(OPR-1000) are suggested and economic evaluations are performed. A single-zone LP with 89 fresh fuels was evaluated to be optimal for 24-month operation of OPR-1000 in terms of economic gain. The 24-month operation of OPR-1000 with this LP gives a profit of 7.073 million dollars per year compared to 18-month operation.

Evaluation of aerosol retention inside a steam generator under steam generator tube rupture accident conditions

The necessity has arisen to conduct risk evaluations of the steam generator tube rupture (SGTR) accident for domestic operating nuclear power plants in the Republic of Korea in accordance with a recent amendment to the Nuclear Safety Act. To evaluate the aerosol-type radionuclides released to the environment during the accident and to prepare mitigation strategies, tests conducted in experimental facilities that faithfully reflect the properties of domestic plants are essential. In this work, one such experimental facility was installed at the Korea Atomic Energy Research Institute with scaling work considering the geometrical shapes of a domestic Korean plant and the major aerosol removal mechanisms inside the steam generators. A preliminary test was first conducted to understand the amount of aerosol removal inside the pipes of the facility with analytical methods. Then, aerosol retention tests were performed in both dry and flooded conditions. Decontamination factors were evaluated in each test, and the results were analyzed using filter measurements, an electrical low-pressure impactor, and deposition mass in the tubes. The test results are believed to provide a good starting point to improve the related accident management strategies to increase the safety of domestic Korean nuclear power plants.

Assessment of plant response capability in the event of loss of electrical power and loss of ultimate heat sink

This paper presents a stress test method developed for Korean nuclear power plant and assessment results of plant response capability of OPR1000 in an effort to enhance the safety against extreme natural events exceeding the design basis of the plant. Multi-barrier accident coping strategy (MACST) guidelines have been developed to accomplish specific tasks in support of Emergency Operating Procedures (EOP) when the Extended Loss of AC Power (ELAP) and/or Loss of Ultimate Heat Sink (LUHS) is in progress as a consequence of beyond design basis natural events. Provisions to secure critical safety functions are discussed with cliff edge time evaluation at each step of plant response to ELAP and LUHS. It is concluded that the plant can maintain the safe shutdown state through the natural circulation cooldown by steam generator and the electric power and can obtain additional capability and redundancy with off-site equipment to provide coping capability for an indefinite period.

Derivation of a new dose constraint applicable to radioactive discharges from Korean nuclear power plants through retrospective dose assessment

A new methodology to derive a dose constraint for radioactive effluent from a unit of nuclear power plant (NPP) through retrospective assessment was developed to reflect operational flexibility in line with international standards. The new dose constraint can retain the safety margin between the offsite dose and the past dose constraints. As case studies, the new approach was applied to 24 Korean NPPs to address the limitations of the existing seven dose constraints that do not fully comply with current international radiation protection standards. Therefore, an effective dose constraint for Korean NPPs was proposed as no less than 0.15 mSv/y, which is comparable to the international practices and previous studies (0.05–0.3 mSv/y). Although the lower bound of the equivalent dose constraint was calculated as 0.17 mSv/y, it is not proposed in this study since the compliance with the derived effective dose constraint can prevent accompanied equivalent doses to any organs from exceeding equivalent dose limits. The new framework and the case studies are expected to contribute toward and support the revision of existing dose constraints for radioactive effluent from NPPs, ensuring better compliance with the current international safety standards as well as reflect the operational flexibility in practice.

Traffic management for large-scale evacuation with public transportation and calculation of appropriate operating ratio

In 2013, the International Atomic Energy Agency (IAEA) changed the recommended maximum range of the Emergency Planning Zone (EPZ) to 30 km, and the Kori Nuclear Power Plant in Republic of Korea has also expanded the EPZ to 30 km, following the recommendation. As a result, metropolitan cities with a high population density are contained within the EPZ, and evacuating millions of people should be considered if the 30 km range of evacuation is to take place. This study proposes an evacuation plan using buses (public transportation) to transport people outside of the EPZ, quickly and efficiently. To verify the appropriate mode share ratio of buses that can guarantee the right of vulnerable road users and reduce traffic congestion, a model was built simulating the Kori Nuclear Power Plant in Ulsan Metropolitan City. The scenarios were established by changing the mode share ratio of buses and passenger cars by 10%. Considering a large-scale network analysis at the city level, a cell transmission model was applied to calculate the evacuation time in each scenario. The result shows that the optimal mode share ratio of buses is 40%, with a total evacuation time of 132 min, considering feasible bus fleets in Ulsan Metropolitan City.

Simplified Assessment Methodology of Radionuclide Contamination for Subsurface Soil

Determining whether to release a site after decommissioning a nuclear facility should be preceded by an environmental impact assessment of the exposure radiation dose according to the radionuclides in the soil. Currently, in Korea, various evaluation methodologies and decommissioning technologies are being studied for the first decommissioning of nuclear power plants, starting with Kori Nuclear Power Plant Unit (Kori-1), which is based on the “Multi-Agency Radiation Survey and Site Investigation Manual MARSSIM” developed in the United States. The scope and evaluation targets of deep soil may differ depending on the purpose, but it has been confirmed that the International Atomic Energy Agency and the U.S. Nuclear Regulatory Commission are targeting subsurface soil. MARSSIM outlines the need for an evaluation of this subsurface soil but does not suggest specific methods. In NUREG-1757, which complements MARSSIM, it is confirmed that subsurface soil specifically means a soil layer that is 15 to 30 cm deep in the surface layer. In the current study, using the previously verified computational code RESidual RADioactivity (RESRAD)-ONSITE, a methodology for summation is proposed to evaluate the impact of subsurface soil more flexibly and realistically while minimizing the exposure dose evaluation procedure. When using RESRAD-ONSITE according to this evaluation methodology, it was confirmed that it is possible to respond to changes in the depths of various soil layers. In addition, it was also confirmed that this methodology is adaptable to the contamination of nuclides, such as 60Co, 137Cs, 152Eu, and 154Eu, which are expected to be major nuclides when decommissioning nuclear facilities.

Performance testing of a FastScan whole body counter using an artificial neural network

In Korea, all nuclear power plants (NPPs) participate in annual performance tests including in vivo measurements using the FastScan, a stand type whole body counter (WBC), manufactured by Canberra. In 2018, all Korean NPPs satisfied the testing criterion, the root mean square error (RMSE) ≤ 0.25, for the whole body configuration, but three NPPs which participated in an additional lung configuration test in the fission and activation product category did not meet the criterion. Due to the low resolution of the FastScan NaI(Tl) detectors, the conventional peak analysis (PA) method of the FastScan did not show sufficient performance to meet the criterion in the presence of interfering radioisotopes (RIs), 134Cs and 137Cs. In this study, we developed an artificial neural network (ANN) to improve the performance of the FastScan in the lung configuration. All of the RMSE values derived by the ANN satisfied the criterion, even though the photopeaks of 134Cs and 137Cs interfered with those of the analytes or the analyte photopeaks were located in a low-energy region below 300 keV. Since the ANN performed better than the PA method, it would be expected to be a promising approach to improve the accuracy and precision of in vivo FastScan measurement for the lung configuration.

Separate and integral effect tests of aerosol retention in steam generator during tube rupture accident

A steam generator tube rupture accompanying a core damage may cause the fission product to be released to environment bypassing the containment. In such an accident, the steam generator is the major path of the radioactive aerosol release.AEOLUS facility, the scaled-down model of Korean type steam generator, was built to examine the aerosol removal in the steam generator during the steam generator tube rupture accident. Integral and separate effect tests were performed with the facility for the dry and flooded conditions, and the decontamination factors were presented for different tube configurations and submergences.The dry test results were compared with the existing test results and with the analyses to investigate the aerosol retention physics by the tube bundle, with respect to the particle size and the bundle geometry. In the flooded tests, the effect of submergence were shown and the retention in the jet injection region were presented with respect to the Stokes number.The test results are planned to be used to constitute the aerosol retention model, specifically applicable for the analysis of the steam generator tube rupture accident in Korean nuclear power plants to evaluate realistic fission product behavior.

Cause Analysis for Wall Thinning of Small-Bore Piping in Nuclear Power Plant by ToSPACE, FLUENT and Theoretical Evaluation

It has been known that FAC, LDIE, cavitation and flashing are the damage mechanisms that can cause the pipe thickness of the secondary system of nuclear power plants thinner. Severe wall thinning was found in the MSR drain pipes at a Korean nuclear power plant a decade ago, and all the affected pipes were replaced with low alloy steel with higher chromium contents. Therefore, this study was conducted to reduce the possibility of similar thinning cases that may occur in the future by identifying the exact cause of thinning. ToSPACE and FLUENT codes and theoretical evaluation method were applied to analyze the causes of thinning. ToSPACE and FLUENT analyses and theoretical evaluation including all the operating conditions show a relatively large pressure drop and a pressure lower than the saturated vapor pressure in common at the end of the pipe entering the condenser. This means that flashing occurs at the end of the pipe under all operating conditions, and the effect can be greater than that of other parts. As a result, since severe wall thinning occurred at the end of the pipeline entering the condenser, it was evaluated that flashing by the high-velocity two-phase fluid was the direct cause of the wall thinning in the MSR drain pipes. The results of this study will contribute to establishing appropriate countermeasures in the event of pipe wall thinning in the future.

Characteristics of the Vertical and Horizontal Response Spectra of Earthquakes in the Jeju Island Region

In this study, we evaluated the response spectra of 24 earthquake series, which includes 15 from the Kumamoto earthquake series and 9 from the Pohang earthquake series, and explored the effects of earthquake magnitude on the resonance frequencies of structures and buildings. Furthermore, the observations of this study were compared with the design response spectra, Regulatory Guide 1.60 (The United States Nuclear Regulatory Commission, 1973) for Korean nuclear power plants, and with the Korean Building Code (MOLIT, 2016, hereinafter referred to as KBC 2016) for general structures and buildings. The response spectra, after normalization with reference to the peak ground acceleration (PGA), were derived using a total of 423 horizontal and vertical accelerations. It was observed that the shapes of the horizontal and vertical response spectra were strongly dependent on the magnitude of the earthquake and the resonance frequency. Given the strong dependence of the response on the magnitude, it is suggested to consider magnitude > ML ~ 6.0 when establishing design response spectra. Compared to inland areas, a fairly higher amplitude of response at significantly lower frequency ranges could be attributed to the local geological environment of Jeju Island, which was formed by a surface volcano eruption and the distribution of unconsolidated Pleistocene marine sediments in the Jeju area. It is necessary to study the characteristic influence of layers with low shear wave velocity distributed in the Jeju region on seismic responses more rigorously while considering the frequency band and amplitudes at the surface of Jeju. The resonance frequencies of general low-rise and mid-rise buildings by the brief formula and those by design response spectra both suggested by KBC 2016 were overlapped, and these indicated that the seismic hazard could be much higher on general buildings in the Jeju region than in inland areas. Lastly, it is necessary to make the design standard criteria for Reg. Guide 1.60 and KBC 2016 more conservative in the lower frequency range of higher than 0.6 Hz and 2.0–6.0 Hz, respectively, which is significantly lower than those of the inland area, and to establish improved design response spectra with site-specific seismic design standards by referencing large amounts of qualitative data from the region around the Korean Peninsula.

Application of a combined safety approach for the evaluation of safety margin during a Loss of Condenser Vacuum event

A combined safety approach, which uses a best-estimate computer code and adopts conservative assumptions for safety systems availability, is developed and applied to the safety margin evaluation for the Loss of Condenser Vacuum (LOCV) of the 1000 MWe Korean Nuclear Power Plant. The Multi-dimensional Analysis of Reactor Safety-KINS standard (MARS-KS) code is selected as a best-estimate code and the PAPIRUS program is used to obtain different initial operational conditions through random sampling of control variables. During an LOCV event, fuel integrity is not threatened by the increase in Departure from Nuclear Boiling Ratio (DNBR). However, the high pressure in the primary coolant system and the secondary system might affect the system integrity. Thus, the peak pressure becomes a major safety concern. Transient analyses are performed for 124 cases of different initial conditions and the most conservative case, which results in the highest system pressure is selected. It is found the suggested methodology gives similar peak pressures when compared to those predicted from existing methodologies. The proposed approach is expected to minimize the time and efforts required to identify the conservative plant conditions in the existing conservative safety methodologies.

Analysis of Systems Thinking Safety by Artificial Intelligence (AI) Algorithm in South Korean Nuclear Power Plants (NPPs)

The artificial intelligence (AI) is applied to the safety analysis in the South Korean nuclear power plants (NPPs). The reinforcement learning (RL) is one of promising skills in the wise manipulations for the nuclear safety analysis where the reward is a critical factor to make the modelling. In the simulations, Y-axis means the relative value which shows the quantity of the accident possibility. The highest value is 4.0 in 46.25th year in which the values are increasing gradually. Otherwise, the values in the case with Agent gradually decrease. The highest value is near initial stage, which means the operation in NPPs is comparatively unstable. In the result, the values in the AI based controller graphs are higher than those of the other one. The RL algorithm is expressed by the Agent in this modelling, which is the most important factor in the AI-based operation in NPPs.

Seismic capacity re-evaluation of the 480V motor control center of South Korea NPPs using earthquake experience and experiment data

The recent seismic events that occurred in South Korea have increased the interest in the re-evaluation of the seismic capacity of nuclear power plant (NPP) equipment, which is often conservatively estimated. To date, various approaches—including the Bayesian method proposed by the United States (US) Electric Power Research Institute —have been developed to quantify the seismic capacity of NPP equipment. Among these, the Bayesian approach has advantages in accounting for both prior knowledge and new information to update the probabilistic distribution of seismic capacity. However, data availability and region-specific issues exist in applying this Bayesian approach to Korean NPP equipment. Therefore, this paper proposes to construct an earthquake experience database by combining available earthquake records at Korean NPP sites and the general location of equipment within NPPs. Also, for the better representation of the seismic demand of Korean earthquake datasets, which have distinct seismic characteristics from those of the US at a high-frequency range, a broadband frequency range optimization is suggested. The proposed data construction and seismic demand optimization method for seismic capacity re-evaluation are demonstrated and tested on a 480 V motor control center of a South Korea NPP.

New methodologies to derive discharge limits considering operational flexibility of radioactive effluents from Korean nuclear power plants based on historical discharge data

The new methodologies to derive discharge limits considering operational flexibility according to international safety standards were developed to help reduce the environmental releases of radioactive effluents from nuclear power plants (NPPs). To overcome the limitations of the two existing methods to set up discharge limits assuming a specific statistical distribution of the effluent discharge, two modified equations were newly proposed to directly derive a particular discharge limits corresponding to the target ‘compliance probability’ based on the actual annual discharge data for a specific NPP and radionuclide groups. By applying these to the actual yearly discharge data of 14 Korean NPPs for 7 radionuclide groups for the past 20 years, the applicability of two new methodologies to actual cases was demonstrated. The ‘characteristic value’ with approximately a 90% compliance probability for each Korean NPP and radionuclide group was proposed based on the results. The new approaches for setting up the discharge limits and the characteristic values developed in this study are expected to be effectively utilized to foster operator's efforts to progressively reduce the environmental releases of radioactive effluents of NPPs relative to the previous discharge data considering operational flexibilities.