Nuclear Power Plants In Korea Research Articles

Impact of safety regulation on rare event occurrence: The case of nuclear power plants in Korea

Climate change mitigation and energy supply security are the most important issues worldwide. To attain carbon neutrality and restrict global warming to 1.5°C, expediting investments in energy sectors and facilitating the expansion of clean energy is imperative. Nuclear power generation, as a base load, contributes to the energy transition because it is low-carbon and can be deployed on a large scale; however, particular challenges arise due to the risk of accidents, radioactive waste management problems, and public acceptance. Considering the similarity between macroeconomic data and nuclear power generation data from a time-series perspective, this study estimates the frequency of rare event (RE) occurrence in nuclear reactors using macro time-series methodology. We employ the Bayesian Markov Chain Monte Carlo method, developed by Barro and Jin (2021), to estimate the likelihood of RE occurrences for the reactors using the panel data of electricity production (MWh) obtained from the International Atomic Energy Agency's Power Reactor Information System. The findings indicate that the frequency of RE occurrences significantly decreases as nuclear safety regulations strengthen and the regulatory workforce increases. Despite the limitations of being directly utilized as practical indicators for safety performance, observing its similar pattern with the World Association of Nuclear Operators performance indicators provides insights for policy implications in intermediate safety management aimed at preventing major disaster incidents.

Probabilistic typhoon hazard and sensitivity analysis for nuclear power plant sites in Korea using logic tree

There has been an increase in typhoon intensity, and high-intensity typhoons can affect the safety of nuclear power plants (NPPs). Typhoons can damage the structures, systems, and components (SSCs) of NPPs, causing core damage and the release of radioactive materials into the environment. Therefore, design criteria for the effects of natural phenomena are demanded to ensure the safety-critical SSC functions of NPPs. Also, there is a need for safety assessments of typhoons in operational NPPs, which require typhoon hazard analysis. In this study, we derived typhoon hazards for NPPs in Korea using Logic Tree and Monte Carlo simulations. We performed sensitivity analysis on the typhoon hazards based on the distances (up to 100 km) from the NPP sites. Among other NPP sites, the typhoon hazard of the Kori site showed the highest wind speed, and the Hanul site showed the lowest wind speed. The uncertainty of the wind speed increased as the recurrence interval of the typhoon hazard increased. We compared the typhoon hazard in the areas of interest to that within 100 km. The range of the distance deviation for the Hanbit site's typhoon hazard was within μ±σ up to 75 km, while for the other sites, it was within μ±σ up to 100 km.

A study on the integrity and behavior of nuclear fuel in temporary storage

Due to the saturation of spent fuel pools of nuclear power plant in Korea, spent fuel will be stored in dry casks for a considerable period. Since spent fuel must withstand continuous decay heat, radiation, and high internal pressure of the fuel rod in the cask, behavior of spent nuclear fuel is needed to be reviewed.It is very important to evaluate the behavior of nuclear fuel with realistic modeling. In this study, the behavior of spent nuclear fuel after being burned in a reactor for three cycles and subsequently cooled in a spent fuel pool was analyzed by FALCON-1.5.0 which is a computational code for mechanical and thermal analysis of nuclear fuel. For code analysis, three burnup levels of 42,000 MWD/MTU, 50,000 MWD/MTU, and 58,000 MWD/MTU were considered. The hoop stress, strain, and other parameters were analyzed for each burnup level. Furthermore, the cooling periods in the spent fuel pool are set to 5, 10, 15, and 30 years, allowing for an evaluation of nuclear fuel behavior consistent with the burnup levels. Additionally, nuclear fuel behavior with normal and high thickness of oxidation is reviewed as well.The behavior of various nuclear fuels during the temporary storage period is studied with FALCON and the factors affecting the integrity of nuclear fuel are analyzed for up to 50 years of the temporary storage. This evaluation will contribute to demonstrating the safe storage of nuclear fuel in dry storage.

Radiological safety assessment for transportation of reactor pressure vessel during decommissioning of a nuclear power plant in Korea

In Korea, decommissioning of nuclear power plants and transportation of the decommissioning waste are expected to expand in the near future. It is necessary to confirm that radiological risks to the public and workers are not significant through radiological safety assessment. The objective of this study is to assess the radiological safety for transportation of RPV waste, which is a major decommissioning waste with relatively high level of radioactivity. It was assumed that the waste would be transported to the Gyeongju disposal facility by land transportation. First, the source term and transportation method of the RPV waste were determined, and the external dose rates from the waste were calculated using MCNP. Then, transportation scenarios were assumed under both normal and accident conditions. Under the scenarios, radiation doses were calculated using the RADTRAN. Under normal operation scenarios without a transportation accident, assuming 40 shipments per year, the average individual doses for the public ranged from 6.56 × 10−6 to 2.18 × 10−2 mSv yr−1. The maximum individual doses for only a single shipment ranged from 2.43 × 10−6 to 3.14 × 10−1 mSv. For cargo handlers and vehicle crew members, the average doses were 2.26 × 101 mSv yr−1 and 2.95 mSv yr−1, respectively. Under transportation accident scenarios, average individual radiological risks which are product of the radiation doses and the annual accident rates ranged from 1.14 × 10−11 to 1.61 × 10−10 mSv yr−1 by transportation route segment when considering the transportation accident rate. Average individual doses assuming transportation accident occurrence ranged from 2.62 × 10−4 to 1.42 × 10−3 mSv. The maximum individual dose under accident conditions was 7.99 × 10−2 mSv. The calculated doses were below the regulatory limits in Korea. However, relatively high doses were observed for cargo handlers and vehicle crew members because of conservative assumptions. This study results can be used as basic data for the radiological safety assessment for the decommissioning waste transportation in the future.

Dismantling Quantity Estimation for Nuclear Power Plant: Scan-to-BIM versus Conventional Method

This paper presents a case study of using Scan-to-BIM technology for evaluating the quantity and cost of a nuclear power plant during the decommissioning process. The study proposes a process for applying Scan-to-BIM in the dismantling stage and develops a BIM-based VPL (Visual Programming Language) to acquire accurate as-is databases that are crucial for nuclear power plants handling radioactive materials. The proposed process and system were applied to K-1 NPP (Nuclear Power Plant), the first nuclear power plant in Korea, and the results were compared to those obtained using the conventional method. The study found that errors in quantity evaluation using the conventional method ranged from around 10% to over 10 times, and were attributed to various causes such as outdated as-is drawings and limitations in expressing information in drawings. By using the Scan-to-BIM process and system, errors in quantity evaluation were significantly reduced, and the conventional method was found to have underestimated the dismantling cost by 20%. The findings of this study provide important insights for the efficient and accurate decommissioning of nuclear power plants, and demonstrate the effectiveness of the proposed method and system for evaluating the quantity and cost of a nuclear power plant during the decommissioning process.

Derivation of dose constraint for the general public around nuclear power plants with operational data of radioactive effluent releases

For radiation protection optimization, ICRP proposed dose constraint as quantitative value for planned exposure situation based on representative person concept. The objective of this study is to derive dose constraints for the general public around nuclear power plants in Korea by applying representative person concept. The dose constraints for the general public around NPPs were derived through a total of six steps. The steps consisted of setting source terms, setting exposure pathways and scenarios, setting candidate groups for a critical group decision, setting habit data, calculating radiation doses, and proposing dose constraints. Through these steps, the radiation dose distribution of the general public around the NPPs was obtained, and dose constraints were proposed using the dose distribution. Radiation doses to the general public around all the Korea NPP sites ranged 1.63 × 10−2 to 1.32 × 10−1 mSv/y. Using the dose distribution, 0.15 mSv/y, 0.10 mSv/y, and 0.08 mSv/y were proposed as dose constraints. The dose constraint values derived in this study are proposals. Therefore, it is judged that the dose constraints should need furthermore discussion with regulators, licensees, and radiation protection experts considering societal and economic factors for radiation protection. The proposal for dose constraints developed in this study can be used to optimize radiation protection for the general public around the NPPs.

Comparison of existing and new optical fiber-based scintillation detectors for spent-fuel verification equipment

The Wolsung Nuclear Power Plant in Korea is under IAEA safeguards. During IAEA physical inventory verifications, the Optical Fiber radiation Probe System has been used since 2007 as the official instrument for verification of spent fuel. Its low sensitivity requires excessive time and labor to perform the IAEA safeguard activities and increases the operator load for preparing a dangerous and time-consuming secondary inspection. The objective of this paper is to develop a superior new optical fiber-based scintillation detector to deal with such issues. To increase the sensitivity of the instrument, a new optical fiber-based scintillation detector was developed by combining a p-terphenyl scintillator with a new type of optical fiber. A field test was carried out in the CANDU Wolsung Unit 1 to characterize the performance of the new detector. The field test showed that the OFPS using the new optical fiber-based scintillation detector has at least 100-times higher sensitivity and can also measure even very weak signals not measurable by the existing detector. From the experimental results, we can conclude that the newly developed OFPS detector would increase the effectiveness and efficiency of nuclear safeguards for verification of spent fuel in wet storage pools of CANDU reactors.

Radiological impact assessment of radioactive effluents emitted from nuclear power plants in Korea and China under normal operation

In the present study, the effects of normally operating nuclear power plants (NPPs) in Korea and China on the radiation dose to Korea's major cities were quantitatively analyzed and compared. The annual effective dose to the general public, as incurred from NPPs' radioactive effluents, was evaluated. The results showed that some of the NPP sites in China, such as the Shidao Bay NPP and Hongyanhe NPP, made a larger contribution to the effective dose than those in Korea. It was also found that the effective dose to Korea's major cities was more significantly affected by the NPPs in the Shandong Peninsula than by those in the Liaoning Peninsula. As such, the contribution of China's NPPs to the radiation dose to Korea was not simply inversely proportional to their distance from Korea but also varied from location to location, because atmospheric dispersion patterns were significantly affected by given weather conditions. The effective dose to Korea was calculated considering “all” of China's NPPs (both actual and prospective), including those currently in operation, under construction, and to be built. The results showed that the effective dose should be expected to increase to 1.1E-08 Sv/year, about a 500-fold increase from the 2000 level and a 2.6-fold increase from the 2010 level. Finally, the contributions of the NPPs in China and Korea to the collective dose to Korea's seven major cities were determined. In Seoul, in particular, the contribution of China's NPPs was found to account for more than one-third of the total. The annual effective dose to the Korean general public incurred from the radioactive effluents emitted from normally operating NPPs in China and Korea was measured to just a few or tens of nSv. Given that the total worldwide average effective dose from natural radiation is approximately 2.4 mSv, it was judged to be a very small dose.

Comparative analysis and countermeasures of power system open phase conditions in Korea’s standard nuclear power plants

Open Phase Condition (OPC) is a failure in which one (or two) of the three-phases of the power system is maintained open, and the OPC of the nuclear power plant (NPP) may cause station blackout (SBO) and pose a great risk to the integrity of the reactor core. Recently, OPCs have occurred in power systems at NPPs around the world, but many problems have been reported that failures are not detected by the current protection systems. Accordingly, the NPPs in Korea, which have a power system similar to that of overseas NPPs, also need to review the risks to OPCs. In this paper, the occurrence of OPCs was simulated for Korea’s standard NPPs using ETAP, and the current protective relays were analyzed and compared for each scenario. In addition, based on these analysis results, countermeasures against OPCs are proposed in order to secure reactor core integrity. It is considered that this paper can be used as a technical background material necessary for improving the on/off-site power protection system of NPPs and establishing related detailed regulatory guidelines in the future.

Radiological Impact Assessment of Radioactive Effluents Emitted from Nuclear Power Plants in Korea and China Under Normal Operation

Radiological Impact Assessment of Radioactive Effluents Emitted from Nuclear Power Plants in Korea and China Under Normal Operation

Estimation of radiation dose resulting from the recycling of large metal wastes from decommissioning nuclear power plants in Korea

AbstractFollowing the permanent shutdown of Kori Unit 1 in June 2017, the Republic of Korea has been preparing for full‐scale decommissioning work. In addition, the design life of 12 units will expire by 2030. If decommissioning begins without initially extending the lifespan of nuclear reactors, massive amounts of decommissioning wastes will be generated in a short period. The total amount of radioactive wastes generated during the dismantling of a pressurized water reactor is estimated as 6200 tons, and approximately 70% of the total radioactive wastes are classified as metal wastes. Self‐disposal through the decontamination of contaminated metals can contribute to the economic feasibility of decommissioning nuclear power plants because it can reduce the disposal cost of medium‐ and low‐level radioactive wastes. Therefore, this study evaluated the possibility of self‐disposal of steam generators that may occur during future decommissioning. The radioactivity analysis data on transferring the replaced steam generator of Hanul Unit 1 were used as the source term. The decontamination factor was calculated by applying 200 units from 1200 to 2000, and the radiation dose was evaluated using the RESRAD‐RECYCLE code. Consequently, the single‐nuclide concentration and sum of the allowable concentration fraction for mixed radionuclides at a decontamination factor of 1400 were below the regulatory requirements; however, the dose evaluation results exceeded the allowable dose in some scenarios. The decontamination factor was 2000, when the dose evaluation results for all scenarios met the regulatory requirements.

Derivation of site-specific derived concentration guideline levels at Korea Research Reactor-1&2 sites

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.

A Study on Annual Atmospheric Dispersion Factors Between Continuous and Purge Releases of Gaseous Radioactive Effluents

Radioactive materials from nuclear power facilities can be released into the atmosphere through various channels. Recently, the dispersion of radioactive materials has become critical issue in Korea after Kori Unit 1 and Wolsong Unit 1 were permanently shut down. In this study, annual atmospheric dispersion factors were compared based on the continuous release and purge release using the XOQDOQ computer program, a method for calculating atmospheric dispersion factors at commercial nuclear power stations. The meteorological data analyzed in this study was based on the Shin Kori nuclear power meteorological tower which has the largest operating nuclear power plants in Korea, for three years (from 2008 to 2010). The analysis results of the dispersion factor of the radioactive material release obtained using the XOQDOQ program showed that the difference between the continuous release and purge release was within two times. This study will be valuable helpful for revealing the uncertainty of the predictive atmospheric dispersion factor to achieve regulation.

A site risk assessment for internal events: A case study

Site risk assessment (SRA) is to estimate the risk of accidents that are probable to occur on the site. The group of accidents consists of not only the initiating events inducing impact on a single unit, but also the events causing simultaneous malfunctions, such as reactor trip on more than 2 units by common causes, called common-cause initiators (CCIs). These events need to be distinguished at the initiating event analysis and the accident progress should be analyzed separately to conduct the SRA. Furthermore, inter-unit dependencies such as inter-unit common-cause failure (CCF) and unavailability of shared components should be treated appropriately to model the accident sequences with CCI events. In this study, in order to perform the SRA of a reference site, the accident history of nuclear power plants in Korea is investigated to identify single-unit initiators (SUIs) and CCIs for SRA and to calculate the frequency of each initiator. The Level 1 SRA model is developed implementing inter-unit dependencies, unavailability of alternate AC diesel generator and inter-unit CCF. In addition, re-evaluation of failure probability of off-site power recovery is performed. Using these models, site core damage frequency (SCDF) and the site risk indicators identified in this study are calculated to deduce the impact of the multi-unit factors considered in SRA. It is expected that the SRA model developed in this study can be applied to perform the site risk assessment for the other site and contributes to improve the safety from the site risk perspective.

Investigation of the CRUD Layer in Failed PWR Fuel Using an EPMA

This study characterizes a failed discharged fuel rod with 53 000 MWd/tonne U from a nuclear power plant in Korea. Chalk River Unidentified Deposits (CRUD) and the oxide layer were observed using an electron probe micro-analyzer (EPMA, SX-50 R, CAMECA, France) with wavelength dispersive (X-ray) spectroscopy. A normally irradiated cladding specimen was analyzed for comparison with the failed fuel rod. The analysis revealed an oxide layer with a thickness of about 10 μm and double-stratified agglomerates of CRUD species shapes. In contrast, sound fuel rods irradiated under conditions similar to failed fuel showed clusters in which Fe, Ni, and Cr were distributed. The main elements constituting the CRUD material, notably Ni and Fe, were located in the same position. Moreover, the thickness of the oxidized layer of the failed fuel rod was found to be significantly different from the thickness of the sound fuel rod. Consequently, EPMA techniques offer the possibility of identifying and analyzing the CRUD phases and segregations in spent pressurized water reactor fuel. Although phases and segregations are small in terms of the amount expected to be present in background radiation, they nevertheless present a significant analytical challenge.

Study of the Characteristics of Wind over Nuclear Power Plants in Korea

To establish a strategy for public protective action from radioactive leakage in the event of Nuclear Power Plant (NPP) accidents, long-term records of wind data collected at the Korean NPPs were analyzed. Wind characteristics related to the advection and diffusion of radioactive pollutants were examined by analyzing the wind direction, speed, and land-sea breezes for NPPs (Hanbit, Hanul, Wolsong, Kori, and Shin-Kori) in Korea. The study also analyzed the characteristics of calm winds causing the accumulation of radioactive materials. The wind characteristics of each NPP differ depending on the seasonal and daily variabilities; thus, a detailed time-scaled airflow database is required. In addition, the findings, through continuous updates of the airflow database, will contribute to improving preparedness.

Dynamic characteristics of single door electrical cabinet under rocking: Source reconciliation of experimental and numerical findings

Seismic qualifications of electrical equipment, such as cabinet systems, have been emerging as the key area of nuclear power plants in Korea since the 2016 Gyeongju earthquake, including the high-frequency domain. In addition, electrical equipment was sensitive to the high-frequency ground motions during the past earthquake. Therefore, this paper presents the rocking behavior of the electrical cabinet system subjected to Reg. 1.60 and UHS. The high fidelity finite element (FE) model of the cabinet related to the shaking table test data was developed. In particular, the first two global modes of the cabinet from the experimental test were 16 Hz and 24 Hz, respectively. In addition, 30.05 Hz and 37.5 Hz were determined to be the first two local modes in the cabinet. The high fidelity FE model of the cabinet using the ABAQUS platform was extremely reconciled with shaking table tests. As a result, the dynamic properties of the cabinet were sensitive to electrical instruments, such as relays and switchboards, during the shaking table test. In addition, the amplification with respect to the vibration transfer function of the cabinet was observed on the third floor in the cabinet due to localized impact corresponding to the rocking phenomenon of the cabinet under Reg.1.60 and UHS. Overall, the rocking of the cabinet system can be caused by the low-frequency oscillations and higher peak horizontal acceleration.

Calculation approach to building DCGLs of waste treatment facilility for Kori unit 1 decommissioning and comparative analysis of the results

The Derived Concentration Guideline Levels (DCGLs) of remaining buildings in decommissioning site realistically need to be considered because all nuclear power plants in Korea are operated with the adjacent plant in multi-unit site. In addition, a few buildings are expected to be utilized for the decommissioning such as waste treatment facility to reduce the final disposal volume of radioactive waste. Therefore, DCGLs of buildings postulated for the decommissioning of Kori site have been assessed by using RESRAD-BUILD code. As a result, DCGLs were derived, reflecting the site reuse criteria and input prameters based on site conditions and assumptions, and compared and analyzed how they differ from DCGLs of other nuclear plants depending on radionuclides. The occupancy and renovation scenarios were applied and more than half of the target radionuclides were conservatilvely evaluated in the occupancy scenarion, while some were evaluated in the renovation scenario. The approach and methodology presented in this paper, which can be considered at the level of the transition phase, can be utilized for deriving results that reflect reality at the time of license termination.

Evaluation of LBB characteristics of candidate materials for main steam line piping in Korea nuclear power plants

Recently, efforts have been made to apply the leak-before-break (LBB) design concept to the main steam line (MSL) piping in secondary systems of Korea standard nuclear power plants (KSNPs). In this study, the microstructures, mechanical properties, and consequential LBB margins of five candidate materials (SA335 Gr.P11/P22, SA508 Gr.3 Cl.2, SA508 Gr.3 Cl.1, and SA508 Gr.1A steels) were evaluated, and strong/weak points of candidate materials for applying MSL piping were analyzed through the comprehensive analysis including microstructure, mechanical properties, LBB margin, usage history, economics, and so on. SA508 Gr.1A steels, which has been used extensively as a piping material in NPPs and has a low fabrication cost, had a good combination of yield strength and fracture toughness. They had LBB margin of 1.11–1.21. SA508 Gr.3 Cl.1 steel had an excellent combination of yield strength and fracture toughness with an LBB margin of 1.64. SA508 Gr.3 Cl.2 steel had the highest LBB margin (1.84) because it had the highest strength; however, it had low fracture toughness and low upper shelf energy (USE). The effect of yield strength on the LBB margin was considerably higher than the effect of fracture toughness. SA508 Gr.3 Steels have high LBB margin owing to the high yield strength, but they have never been used for piping materials in NPPs before. SA335 Gr.P11 and P22 had the LBB margin of 1.1 and 1.23, respectively. However, the elongation and J-R fracture toughness of SA335 Gr.P11 and P22 decreased significantly at 286 °C owing to dynamic strain aging (DSA).

Fire Protection Regulations for Ensuring Fire Safety during Decommissioning Nuclear Power Plants in Korea

Nuclear power plants (NPPs) in Korea are required to be maintained using a defense in-depth approach to prevent leakage of radioactive substances outside the plant and allow safe shutdown in the event of a fire. Periodic testing must be conducted to ensure that the fire protection facilities perform as required by the laws for various nuclear reactor types. In June 2017, for the first time in Korea, a nuclear plant, Kori Unit 1, was permanently shut down. It was prepared for decommissioning in accordance with the fire protection regulations imposed by the regulatory body. However, a standard protocol is necessary for systematically establishing the fire protection program for decommissioning of NPPs in the future. Therefore, the nuclear legal systems of countries with many operating nuclear power plants, such as the United States, Japan, Canada, and various European countries, were reviewed and guidelines for establishing a fire protection program for decommissioning NPPs was suggested; the fire protection requirements stated by Reg Guide 1.191 (Decommissioning fire protection program for NPPs during decommissioning and permanent shutdown) were used as a model. Suggestions for establishing legal regulations to optimize fire protection programs and secure basic technology for decommissioning NPPs were also made.