• Proceedings of the Korean Nuclear Society Conference
• /
• 1997.10a
• /
• pp.707-712
• /
• 1997

This paper presents the current state of knowledge on molten material relocation into the lower plenum. Consequences of movement of material to the lower head are considered with regardt to the potential for reactor pressure vessel failure from both thermal hydraulic and mechanical standpoints. The models are applied to evaluating various in-vessel retention strategies for the Korean Standard power plant (KSNPP) reactor The results are summarized in terms of thermal response of the reactor vessel from the very relevant severe accident management perspective.

• Nuclear Engineering and Technology
• /
• v.26 no.1
• /
• pp.41-53
• /
• 1994

A SGTR accident postulated at Kori unit 1 is simulated with Mihama unit experience, which occurred on February 1991, to evaluate the capability of plant to cope with the transient. The system design and plant conditions of Kori Unit 1 are much similar with those of Mihama Unit 2. Therefore, special concern has been given to evaluate the sequence and the resulting consequence of the postulated SGTR accident at the Kori unit 1 An analysis is peformed as realistically as possible, with following the EOP of Kori unit 1. The result indicates that the leak through tube break terminates within about forty minutes, and the Kori unit 1 may be sufficient to cope with SGTR accident with same type of sequence. However, the reconsideration may be required for the design of Kori unit 1 which disconnects non-safety AC power from off-site power on SI signal generation. It may be pointed out that the content of EOP for SGTR accident is not enough to require operator's proper judgements. An analysis of SGTR accident tested in the LSTF which simulated the SGTR accident at the Mihama Unit 2 is peformed using the RELAP5/MOD3. The results indicates that the code yields in general good agreement with the test, except the break flowrate at the early stage of the event.

• Nuclear Engineering and Technology
• /
• v.31 no.1
• /
• pp.17-28
• /
• 1999

A thermal-hydraulic analysis is conducted on the loss-of-coolant-accident (LOCA) during shutdown operation of YGN Units 3/4. Based on the review of plant-specific characteristics of YGN Units 3/4 in design and operation, a set of analysis cases is determined, and predicted by the RELAP5/MOD3.2 code during LOCA in the hot-standby mode. The evaluated thermal-hydraulic phenomena are blowdown, break flow, inventory distribution, natural circulation, and core thermal response. The difference in thermal-hydraulic behavior of LOCA at shutolown condition from that of LOCA at full power is identified as depressurization rate, the delay in peak natural circulation timing and the loop seal clearing (LSC) timing. In addition, the effect of high pressure safety injection (HPSI) on plant response is also evaluated. The break spectrum analysis shows that the critical break size can be between 1% to 2% of cold leg area, and that the available operator action time for the Sl actuation and the margin in the peak clad temperature (PCT) could be reduced when considering uncertainties of the present RELAP5 calculation.

• Nuclear Engineering and Technology
• /
• v.44 no.6
• /
• pp.611-622
• /
• 2012

"Scaling" plays an important role for safety analyses in the licensing of water cooled nuclear power reactors. Accident analyses, a sub set of safety analyses, is mostly based on nuclear reactor system thermal hydraulics, and therefore based on an adequate experimental data base, and in recent licensing applications, on best estimate computer code calculations. In the field of nuclear reactor technology, only a small set of the needed experiments can be executed at a nuclear power plant; the major part of experiments, either because of economics or because of safety concerns, has to be executed at reduced scale facilities. How to address the scaling issue has been the subject of numerous investigations in the past few decades (a lot of work has been performed in the 80thies and 90thies of the last century), and is still the focus of many scientific studies. The present paper proposes a "roadmap" to scaling. Key elements are the "scaling-pyramid", related "scaling bridges" and a logical path across scaling achievements (which constitute the "scaling puzzle"). The objective is addressing the scaling issue when demonstrating the applicability of the system codes, the "key-to-scaling", in the licensing process of a nuclear power plant. The proposed "road map to scaling" aims at solving the "scaling puzzle", by introducing a unified approach to the problem.

• Journal of the Korean Society of Safety
• /
• v.31 no.1
• /
• pp.126-131
• /
• 2016

Although many studies have been conducted to find solutions to deal with human errors effectively, violations have been rarely studied in depth. The violation is a type of human error when an employee takes an action with intention but does not intend harmful outcomes. Violations have characteristics similar to other types of human errors but it is difficult to understand the intention of an employee from accident reports. The objective of this study is to develop a conceptual model of violation errors for preventing accidents/failures in a nuclear power plant from violation errors. Based on the previous studies, the characteristics of violations were collected in 9 categories and 136 items. They were classified into three-kinds of characteristics (human-related, task-related, organization-related characteristics) to construct conceptual models of routine/situational violations. The representative cases of accidents/failures in a nuclear power plant were analyzed to derive the specific types of routine/situational violation occurrence. Three types of conceptual model for each violation were derived according to whether the basic, human-related, and task-related characteristics are included or not. The conceptual models can be utilized to develop guidelines to support employees preventing routine/situational violations and to develop supportive system in nuclear power plant.

• Journal of Radiation Protection and Research
• /
• v.46 no.4
• /
• pp.178-183
• /
• 2021

Background: This study was carried out to provide environmental transfer parameter values to estimate activity concentrations of these radionuclides in agricultural crops when direct contamination occurred. Materials and Methods: Mass interception fractions (F_Bs) and weathering half-lives (T_ws) of ¹³¹I and radiocesium were calculated using openly available monitoring data obtained after the Fukushima Daiichi Nuclear Power Plant accident. F_B is the ratio between the initial radioactivity concentration of a radionuclide retained by the edible part of the plant (Bq·kg^-1 fresh weight [FW]) and the amount of deposited radionuclide in that area (Bq·m^-2). T_w values can be calculated using activity concentrations of crops decreased with time after the initial contamination. Results and Discussion: Calculated F_B and T_w values for ¹³¹I and radiocesium were mostly obtained for leafy vegetables. The analytical results showed that there was no difference of F_Bs between ¹³¹I and radiocesium by t-test; geometric mean values for leafy vegetables cultivated under outdoor conditions were 0.058 and 0.12 m²·kg^-1 FW, respectively. Geometric mean T_w value of ¹³¹I in leafy vegetables grown under outdoor conditions was 8.6 days, and that of radiocesium was 6.6 days; there was no significant difference between T_w values of these radionuclides by Wilcoxon rank sum test. Conclusion: There was no difference between ¹³¹I and radiocesium for F_Bs and T_ws. By using these factors, we would be able to carry out a rough estimation of the activity concentrations of ¹³¹I and radiocesium in the edible part of leafy crops when a nuclear accident occurred.

• Nuclear Engineering and Technology
• /
• v.54 no.6
• /
• pp.2055-2061
• /
• 2022

In the event of a short-circuit accident on a 4.16 kV non-safety bus, the voltage is temporarily lowered as backflow occurs on the safety bus. In such cases, the undervoltage relay of the safety bus shall not pick up the undervoltage so as not to interfere with the operation of the safety motors. The aim of this study is to develop an adaptive undervoltage protection scheme for the 4.16 kV safety bus considering the faults on the 13.8 kV and 4.16 kV non-safety buses connected to secondary windings of the three winding transformers, UAT and SAT. The result of this study will be the adaptive undervoltage protection scheme for the safety bus of nuclear power plants satisfying functional requirements of the safety related medium voltage motors. The adaptive undervoltage protection scheme can be implemented into an integrated digital protective relay to make user friendly and reliable protection scheme.

• Korean Journal of Remote Sensing
• /
• v.28 no.1
• /
• pp.129-143
• /
• 2012

This paper demonstrates the effectiveness of advanced spatial modeling techniques for environmental monitoring and impact assessment through a case study of Chernobyl nuclear accident occurred in 1986. Land-cover types changed after the accident are analysed by a post classification comparison method using bi-temporal Landsat TM data acquired in 1986 and 1992 near the accident site. Spatial modeling including various kriging algorithms are also applied to analyze the relationships between Cesium concentrations in soil and thyroid cancer incidence rates in Belarus, which was greatly damaged by the accident. The change detection results clearly showed the decrease of croplands and the increase of abandoned lands, and concrete structures were newly built around the nuclear plant to prevent the spread of radioactive contamination. In Belarus, high Cesium concentrations were observed in southern areas with high thyroid cancer risk estimated by Poisson kriging. Geographically weighted regression, which could account for geographic variations of independent variables including Cesium concentrations and distances from the Chernobyl nuclear power plant, was applied to extract the relationships between the independent variables and the thyroid cancer risk. The estimated risk values showed a correlation coefficient value of 0.98 with respect to the thyroid cancer risk values, which implied that the thyroid cancer risk in Belarus was affected by the accident. In conclusion, it is expected that advanced spatial modeling techniques applied in this study would be useful for environmental impact assessment and public health research.

• Nuclear Engineering and Technology
• /
• v.51 no.8
• /
• pp.1939-1950
• /
• 2019

The MELCOR code useful for a plant-specific hydrogen risk analysis has inevitable limitations in prediction of a turbulent flow of a hydrogen mixture. To investigate the accuracy of the hydrogen risk analysis by the MELCOR code, results for the turbulent gas behavior at pipe rupture accident were compared with CFX results which were verified by the American National Standard Institute (ANSI) model. The postulated accident scenario was selected to be surge line failure induced by station blackout of an Optimized Power Reactor 1000 MWe (OPR1000). When the surge line failure occurred, the flow out of the surgeline was strongly turbulent, from which the MELCOR code predicted that a substantial amount of hydrogen could be released. Nevertheless, the results indicated nonflammable mixtures owing to the high steam concentration released before the failure. On the other hand, the CFX code solving the three-dimensional fluid dynamics by incorporating the turbulence closure model predicted that the flammable area continuously existed at the jet interface even in the rising hydrogen mixtures. In conclusion, this study confirmed that the MELCOR code, which has limitations in turbulence analysis, could underestimate the existence of local combustible gas at pipe rupture accident. This clear comparison between two codes can contribute to establishing a guideline for computational hydrogen risk analysis.

• Nuclear Engineering and Technology
• /
• v.55 no.12
• /
• pp.4607-4616
• /
• 2023

We implement machine learning regression models to predict peak pressures of primary and secondary systems, a major safety concern in Loss Of Condenser Vacuum (LOCV) accident. We selected the Multi-dimensional Analysis of Reactor Safety-KINS standard (MARS-KS) code to analyze the LOCV accident, and the reference plant is the Korean Optimized Power Reactor 1000MWe (OPR1000). eXtreme Gradient Boosting (XGBoost) is selected as a machine learning tool. The MARS-KS code is used to generate LOCV accident data and the data is applied to train the machine learning model. Hyperparameter optimization is performed using a simulated annealing. The randomly generated combination of initial conditions within the operating range is put into the input of the XGBoost model to predict the peak pressure. These initial conditions that cause peak pressure with MARS-KS generate the results. After such a process, the error between the predicted value and the code output is calculated. Uncertainty about the machine learning model is also calculated to verify the model accuracy. The machine learning model presented in this paper successfully identifies a combination of initial conditions that produce a more conservative peak pressure than the values calculated with existing methodologies.