• Title, Summary, Keyword: Core damage frequency

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A Study on Review-Level Ground Motion For Seismic Margin Assessment (내진여유도 평가를 위한 부석기준지진동(RLGM) 평가 연구)

  • 연관희;이종림
    • Proceedings of the Earthquake Engineering Society of Korea Conference
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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.

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A Risk Informed Approach to Relax AOTs and STIs in Technical Specifications

  • Cho, Sung-Hwan;Park, Byoung-Chul;Seo, Mi-Ro
    • Proceedings of the Korean Nuclear Society Conference
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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

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Seismic Fragility Analysis of Base Isolated Liquid Storage Tank (면진 유체 저장 탱크의 지진취약도 분석)

  • Ahn, Sung-Moon;Choi, In-Kil;Choun, Young-Sun
    • Proceedings of the Earthquake Engineering Society of Korea Conference
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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.

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A Study on the Final Probabilistic Safety Assessment for the Jordan Research and Training Reactor (JRTR 연구용원자로에 대한 최종 확률론적 안전성평가)

  • Lee, Yoon-Hwan
    • Journal of the Korean Society of Safety
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    • v.35 no.3
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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.

Multi-unit Level 1 probabilistic safety assessment: Approaches and their application to a six-unit nuclear power plant site

  • Kim, Dong-San;Han, Sang Hoon;Park, Jin Hee;Lim, Ho-Gon;Kim, Jung Han
    • Nuclear Engineering and Technology
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    • v.50 no.8
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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.

Consistency issues in quantitative safety goals of nuclear power plants in Korea

  • Kim, Ji Suk;Kim, Man Cheol
    • Nuclear Engineering and Technology
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    • v.51 no.7
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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.

Priority Rankings of the System Modifications to Reduce Core Damage Frequency of Wolsong NPP Units 2/3/4

  • Kwon, Jong-Jooh;Kim, Myung-Ki;Seo, Mi-Ro;Hong, Sung-Yull
    • Proceedings of the Korean Nuclear Society Conference
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    • pp.899-905
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    • 1998
  • The analysis priority makings the recommendation to reduce the total core damage frequency (CDF) of Wolsong nuclear Power Plant nits 2/3/4 was Performed in this paper. In order to derive the recommendation, the sensitivity analysis of CDF on which major contributors effect m performed based on the accident quantification results during Level 1 Probabilistic safety assessment (PSA). Priorities were ranked in tile way that compares the CDF reduction rate with efforts required to implement those recommendations using risk matrix

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Seismic Risk Evaluation of Isolated Emergency Diesel Generator System (면진된 비상디젤발전기의 지진위험도 평가)

  • Kim, Min-Kyu;Ohtori, Yasuki;Choun, Young-Sun
    • Proceedings of the Computational Structural Engineering Institute Conference
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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.

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Determination of Performance Indicator Thresholds Based on Typical PSA Results

  • Kang, Dae-Il;Kim, Kil-Yoo;Hwang, Mee-Jung;Sung, Key-Yong
    • Nuclear Engineering and Technology
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    • v.36 no.6
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    • pp.485-496
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    • 2004
  • Typical probabilistic safety assessment (PSA) results were used to estimate the performance indicator (PI) thresholds of unplanned reactor scram (URS) and safety system unavailability (SSU) for Korean nuclear power plants (NPPs). The changes in core damage frequency (${\Delta}$CDFs) of $10^{-6}/yr$, $10^{-5}/yr$, and $10^{-4}/yr$ were adopted as the risk criteria in setting up the PI thresholds. The PI thresholds for the URS were estimated using information pertaining to the initiating event frequencies, the CDF, and the CDF contribution of each initiating event. The PI thresholds of the SSU were estimated using information on the unavailability, the Fussell-Vesely importance, and the CDF.