• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3741-3758
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• 2021

Outcomes of the FALCON code analysis-related part of the STARS-ENSI Service Project on Evaluation of the new U.S.NRC RIA Fuel Safety Criteria and Application to the Swiss Reactors are presented. Substantial conservatism of the updated safety limits for high-temperature and PCMI cladding failure, as proposed in the NRC Regulatory Guide RG 1.236, is confirmed. Applicability of the updated failure limits to fuel safety analysis in the Swiss PWRs, as applied to standard fuel designs using UO2 fuel pellets and SRA Zry-4 as cladding materials is discussed. Conducting of new integral RIA tests with irradiated samples using doped- and gadolinia fuel pellets to support appropriate fuel safety criteria for RIA events is recommended.

• Journal of the Korean Society of Safety
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• v.37 no.5
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• pp.80-88
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• 2022

The Korean nuclear industry had developed the SPACE (Safety and Performance Analysis Code for nuclear power plants) code, which adopts a two-fluid, three-field model that is comprised of gas, continuous liquid and droplet fields and has the capability to simulate three-dimensional models. According to the revised law by the Nuclear Safety and Security Commission (NSSC) in Korea, the multiple failure accidents that must be considered for the accident management plan of a nuclear power plant was determined based on the lessons learned from the Fukushima accident. Generally, to improve the reliability of the calculation results of a safety analysis code, verification is required for the separate and integral effect experiments. Therefore, the goal of this work is to verify the calculation capability of the SPACE code for multiple failure accidents. For this purpose, an experiment was conducted to simulate a Control Element Drive Mechanism (CEDM) break with a safety injection failure using the ATLAS test facility, which is operated by Korea Atomic Energy Research Institute (KAERI). This experiment focused on the comparison between the experiment results and code calculation results to verify the performance of the SPACE code. The results of the overall system transient response using the SPACE code showed similar trends with the experimental results for parameters such as the system pressure, mass flow rate, and collapsed water level in component. In conclusion, it can be concluded that the SPACE code has sufficient capability to simulate a CEDM break with a safety injection failure accident.

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.1051-1065
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• 2020

Safety analysis of nuclear power plant (NPP) especially in accident conditions is a basic and necessary issue for applications and commercialization of reactors. Many previous researches and development works have been conducted. However, most achievements focused on the safety reliability of primary pressure system vessels. Few literatures studied the structural safety of huge concrete structures surrounding primary pressure system, especially for the fourth generation NPP which allows existing of through cracks. In this paper, structural safety reliability of concrete structures of HTR-PM in accidental double-ended break of hot gas ducts was studied by Exceedance Probability Method. It was calculated by Monte Carlo approaches applying numerical simulations by Abaqus. Damage parameters were proposed and used to define the property of concrete, which can perfectly describe the crack state of concrete structures. Calculation results indicated that functional failure determined by deterministic safety analysis was decided by the crack resistance capability of containment buildings, whereas the bearing capacity of concrete structures possess a high safety margin. The failure probability of concrete structures during an accident of double-ended break of hot gas ducts will be 31.18%. Adding the consideration the contingency occurrence probability of the accident, probability of functional failure is sufficiently low.

• Journal of the Korean Society of Safety
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• v.31 no.5
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• pp.95-101
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• 2016

This study proposes a multi-scale dynamic system reliability analysis of control system as a method of quantitative evaluation of its performance in probabilistic terms. In this second paper, we discuss the control effect of the viscous damper on the seismic performance of the structure-level failure. Since the failure of one structural member does not necessarily cause the collapse of the structural system, we need to consider a set of failure scenarios of the structural system and compute the sum of the failure probabilities of the failure scenarios where the statistical dependence between the failure scenarios should be taken into account. Therefore, this computation requires additional system reliability analysis. As a result, the proposed approach takes a hierarchial framework where the failure probability of a structural member is computed using a lower-scale system reliability with the union set of time-sequential member failures and their statistical dependence, and the failure probability of the structural system is again computed using a higher-scale system reliability with the member failure probabilities obtained by the lower-scale system reliability and their statistical dependence. Numerical results demonstrate that the proposed approach can provide an accurate and stable reliability assessment of the control performance of the viscous damper system on the system failure. Also, the parametric study of damper capacity on the seismic performance has been performed to demonstrate the applicability of the proposed approach through the probabilistic assessment of the seismic performance improvement of the damper system.

• Journal of the Korean Society of Safety
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• v.31 no.6
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• pp.58-66
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• 2016

This study aims to estimate the failure probability of concrete gravity dams for their risk analysis under flood situation. To the end, failure modes of concrete gravity dams and their limit state functions are proposed based on numerous review of domestic and international literatures on the dam failure cases and design standards. Three failure modes are proposed: overturning, sliding, and overstress. Based on the failure modes the limit state functions, the failure probability is assessed for a weir section and a non-weir section of a dam in Korea. As water level is rising from operational condition to extreme flood condition, the failure probability is found to be raised up to the warning condition, especially for overturning mode at the non-weir section. The result can be used to reduce the risk of the dam by random environmental variables under possible flood situation.

• Proceedings of the Korean Institute of Industrial Safety Conference
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• 1998.11a
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• pp.261-266
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• 1998

This paper presents a numerical simulation of the failure behaviour for the pipeline which has external corroded parts along the axis. The numerical study for the pipeline failure analysis is based on an FEM with an elastic-plastic and large-deformation. The predicted failure assessments for the simulated corrosion defects are compared with ANSI/ASME B31G code and a modified B31G method.

• Journal of the Korea Safety Management & Science
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• v.1 no.1
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• pp.79-90
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• 1999

In this paper, we suggested system reliability inequality used by failure rate distribution and developed new theorem-reliability function is increasing function. Also we calculated failure probability of coherent system used by variable transformation. Several examples are illustrated.

• Journal of the Korea Safety Management & Science
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• v.3 no.4
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• pp.91-101
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• 2001

This paper is to develop replacement models under minimal repair with exponential polynomial wavelet failure rate function. Wavelets have good time-frequency localization, fast algorithms and parsimonious representation. Also this study is presented along with numerical examples using sensitivity analysis for exponential polynomial trigonometric failure rate function.

• Journal of the Earthquake Engineering Society of Korea
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• v.23 no.2
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• pp.101-108
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• 2019

The seismic safety of nuclear power plants has always been emphasized by the effects of accidents. In general, the seismic safety evaluation of nuclear power plants carries out a seismic probabilistic safety assessment. The current probabilistic safety assessment assumes that damage to the structure, system, and components (SSCs) occurs independently to each other or perfect dependently to each other. In case of earthquake events, the failure event occurs with the correlation due to the correlation between the seismic response of the SSCs and the seismic performance of the SSCs. In this study, the EEMS (External Event Mensuration System) code is developed which can perform the seismic probabilistic safety assessment considering correlation. The developed code is verified by comparing with the multiplier n, which is for calculating the joint probability of failure, which is proposed by Mankamo. It is analyzed the changes in seismic fragility curves and seismic risks with correlation. As a result, it was confirmed that the seismic fragility curves and seismic risk change according to the failure correlation coefficient. This means that it is important to select an appropriate failure correlation coefficient in order to perform a seismic probabilistic safety assessment. And also, it was confirmed that carrying out the seismic probabilistic safety assessment in consideration of the seismic correlation provides more realistic results, rather than providing conservative or non-conservative results comparing with that damage to the SSCs occurs independently.

• Journal of the Korean Society of Safety
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• v.31 no.3
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• pp.102-108
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• 2016

As a range of accidents happen frequently in our society due to safety insensitivity, researches on individual psychological characteristics related to safety behavior and on safety education are increasing. However, safety behavior research that targets university students as potential employees remains rare. Therefore, this research investigates individual characteristics and psychological factors that affect safety behaviors of university students. Research participants were 131 university students (80 males, 51 females) from the Gyeonggi-do and Chungnam areas. Psychological characteristics (self-regulation, cognitive failure, conscientiousness, and emotional stability) were included as variables that may predict university students' safety behaviors (habits, compliance, mistakes, and violations). The major findings of the study are as follows. Stepwise regression analysis showed conscientiousness was a major predictor explaining safety habits and safety compliance; Cognitive failure and emotional stability were main predictors explaining safety mistakes, and self-regulation was a major predictor explaining safety violations. These findings suggest that development of proper education programs focusing on individual characteristics would allow systematic improvement of safety behaviors of university students.