• 한국지진공학회:학술대회논문집
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• 한국지진공학회 2000년도 춘계 학술발표회 논문집 Proceedings of EESK Conference-Spring
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• pp.97-104
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• 2000

Evaluating a Review-Level Ground Motion is a key to efficiently perform Seismic Margin Assessment of nuclear power plants whose purpose is to determine a ground motion level for which a plant has high-confidence-of-a-low-probability of seismic-induced core damage and to identify any weaker-link components. In this study a method to obtain RLGMs is reviewed which is recommended by Electric Power Research Institute and implemented to be applied to Limerick site in eastern and central U. S as a case study. This method provides reasonable and site-specific RLGMs as minimum required plant HCLPF for SMA that meet a target mean seismic core-damage frequency based on seismic hazard results and generic values of uncertainty and randomness parameters of the core-damage fragility curves. In addition high-frequency RLGM is justifiably modified to reflect the increased seismic capacity of high-frequency components and spatial variation and incoherence of input ground motion on a basemat of large structures by establishing a method to obtain high0-frequency reduction factors according to EPRI guidelines.

• 한국지진공학회:학술대회논문집
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• 한국지진공학회 2005년도 학술발표회 논문집
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• pp.453-460
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• 2005

In this study, the seismic fragility analysis of a base isolated condensate storage tank installed in the nuclear power plant. The condensate storage tank is safety related structure in a nuclear power plant. The failure of this tank affect significantly to the core damage frequency of the nuclear power plants. The seismic analysis of the liquid storage tank was performed by the simple calculation method and the dynamic time storage analysis method. The convective and impulsive fluid mass is modeled as added masses proposed by several researchers. To evaluate the effectiveness of the isolation system, the comparison of HCLPF and core damage frequencies in non-isolated and isolated cases are carried out. It can be found from the results that the seismic isolation system increases the seismic capacity of a condensate storage tank and decreases the core damage frequency significantly.

• 한국안전학회지
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• 제35권3호
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• pp.86-95
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• 2020

This paper describes the work and the results of the final Probabilistic Safety Assessment (PSA) for the Jordan Research and Training Reactor (JRTR). This final PSA was undertaken to assess the level of safety for the design of a research reactor and to evaluate whether it is probabilistically safe to operate and reliable to use. The scope of the PSA described here is a Level 1 PSA, which addresses the risks associated with core damage. After reviewing the documents and its conceptual design, nine typical initiating events were selected regarding internal events during the normal operation of the reactor. AIMS-PSA (Version 1.2c) was used for the accident quantification, and FTREX was used as the quantification engine. 1.0E-15/yr of the cutoff value was used to deliminate the non-effective Minimal Cut Sets (MCSs) when quantifying the JRTR PSA model. As a result, the final result indicates a point estimate of 2.02E-07/yr for the overall Core Damage Frequency (CDF) attributable to internal initiating events in the core damage state for the JRTR. A Loss of Primary Cooling System Flow (LOPCS) is the dominant contributor to the total CDF by a single initiating event (9.96E-08/yr), and provides 49.4% of the CDF. General Transients (GTRNs) are the second largest contributor, and provide 32.9% (6.65E-08/yr) of the CDF.

• 한국안전학회지
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• 제37권4호
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• pp.129-138
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• 2022

A new method for injecting cooling water into the Korean research reactor (KRR) in the event of beam tube rupture is proposed in this paper. Moreover, the research evaluates the risk to the reactor core in terms of core damage frequency (CDF). The proposed method maintains the cooling water in the chimney at a certain level in the tank to prevent nuclear fuel damage solely by gravitational coolant feeding from the emergency water supply system (EWSS). This technique does not require sump recirculation operations described in the current procedure for resolving beam tube accidents. The reduction in the risk to the core in the event of beam tube rupture that can be achieved by the proposed change in the cooling water injection design is quantified as follows. 1) The total CDF of the KRR for the proposed design change is approximately 4.17E-06/yr, which is 8.4% lower than the CDF of the current design (4.55E-06/yr). 2) The CDF for beam tube rupture is 7.10E-08/yr, which represents an 84.1% decrease compared with that of the current design (4.49E-07/yr). In addition to this quantitative reduction in risk, the modified cooling water injection design maintains a supply of pure coolant to the EWSS tank. This means that the reactor does not require decontamination after an accident. Thermal hydraulic analysis proves that the water level in the reactor pool does not cause damage to the nuclear fuel cladding after beam tube rupture. This is because the amount of water in the chimney can be regulated by the EWSS function. The EWSS supplies emergency water to the reactor core to compensate for the evaporation of coolant in the core, thus allowing water to cover the fuel assemblies in the reactor core over a sufficient amount of time.

• Nuclear Engineering and Technology
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• 제50권8호
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• pp.1217-1233
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• 2018

Following a surge of interest in multi-unit risk in the last few years, many recent studies have suggested methods for multi-unit probabilistic safety assessment (MUPSA) and addressed several related aspects. Most of the existing studies though focused on two-unit nuclear power plant (NPP) sites or used rather simplified probabilistic safety assessment (PSA) models to demonstrate the proposed approaches. When considering an NPP site with three or more units, some approaches are inapplicable or yield very conservative results. Since the number of such sites is increasing, there is a strong need to develop and validate practical approaches to the related MUPSA. This article provides several detailed approaches that are applicable to multi-unit Level 1 PSA for sites with up to six or more reactor units. To validate the approaches, a multi-unit Level 1 PSA model is developed and the site core damage frequency is estimated for each of four representative multi-unit initiators, as well as for the case of a simultaneous occurrence of independent single-unit initiators in multiple units. For this purpose, an NPP site with six identical OPR-1000 units is considered, with full-scale Level 1 PSA models for a specific OPR-1000 plant used as the base single-unit models.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 추계학술발표회요약집
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• pp.429.1-429
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• 2002

• 한국안전학회지
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• 제36권3호
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• pp.66-73
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• 2021

This report documents the results of an at-power internal events Level 1 Probabilistic Safety Assessment (PSA) for a Korea research reactor (KRR). The aim of the study is to determine the accident sequences, construct an internal level 1 PSA model, and estimate the core damage frequency (CDF). The accident quantification is performed using the AIMS-PSA software version 1.2c along with a fault tree reliability evaluation expert (FTREX) quantification engine. The KRR PSA model is quantified using a cut-off value of 1.0E-15/yr to eliminate the non-effective minimal cut sets (MCSs). The final result indicates a point estimate of 4.55E-06/yr for the overall CDF attributable to internal initiating events in the core damage state for the KRR. Loss of Electric Power (LOEP) is the predominant contributor to the total CDF via a single initiating event (3.68E-6/yr), providing 80.9% of the CDF. The second largest contributor is the beam tube loss of coolant accident (LOCA), which accounts for 9.9% (4.49E-07/yr) of the CDF.

• Nuclear Engineering and Technology
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• 제51권7호
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• pp.1758-1764
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• 2019

As the safety level of nuclear power plants (NPPs) relates to the safety of individuals, society, and the environment, it is important to establish NPP safety goals. In Korea, two quantitative health objectives and one large release frequency (LRF) criterion were formally set as quantitative safety goals for NPPs by the Nuclear Safety and Security Commission in 2016. The risks of prompt and cancer fatalities from NPPs should be less than 0.1% of the overall risk, and the frequency of nuclear accidents releasing more than 100 TBq of Cs-137 should not exceed 1E-06 per reactor year. This paper reviews the hierarchical structure of safety goals in Korea, its relationship with those of other countries, and the relationships among safety goals and subsidiary criteria like core damage frequency and large early release frequency. By analyzing the effect of the release of 100 TBq of Cs-137 via consequence analysis codes in eight different accident scenarios, it was shown that meeting the LRF criterion results in negligible prompt fatalities in the surrounding area. Hence, the LRF criterion dominates the safety goals for Korean NPPs. Safety goals must be consistent with national policy, international standards, and the goals of other counties.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 2007년도 정기 학술대회 논문집
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• pp.217-222
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• 2007

An Emergency Diesel Generator (EDG) is one of the safety related equipments of a Nuclear Power Plant. The seismic capacity of an EDG in nuclear power plants influences the seismic safety of the plants significantly. A recent study showed that the increase of the seismic capacity of the EDG could reduce the core damage frequency (CDF) remarkably. It is known that the major failure mode of the EDG is a concrete coning failure due to a pulling out of the anchor bolts. The use of base isolators instead of anchor bolts can increase the seismic capacity of the EDG without any major problems. This study introduces a seismic risk analysis method and presents sample results about the seismically isolated and conventional EDG system.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.887-892
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• 1998

A risk informed approach to relax AOTs and STIs of RPS/ESPAS in technical specifications of kori units 3, 4 was performed in this paper. With the proposed AOTs and STIs, system unavailabilities and core damage frequency were quantified using PSA model. The results shoe that core damage frequency is slightly increased by extending AOTs and STIs but negligible. As considering the benefits such as reduction of plant transients and man power for test and maintenance, the relaxation of AOTs/STIs of RPS/ESFAS is justified