• 한국안전학회지
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• 제35권1호
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• pp.97-106
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• 2020

The objective of this paper is to suggest a new quantification method for multi-unit probabilistic safety assessment (PSA) that removes the overestimation error caused by the existing delete-term approximation (DTA) based quantification method. So far, for the actual plant PSA model quantification, a fault tree with negates have been solved by the DTA method. It is well known that the DTA method induces overestimated core damage frequency (CDF) of nuclear power plant (NPP). If a PSA fault tree has negates and non-rare events, the overestimation in CDF drastically increases. Since multi-unit seismic PSA model has plant level negates and many non-rare events in the fault tree, it should be very carefully quantified in order to avoid CDF overestimation. Multi-unit PSA fault tree has normal gates and negates that represent each NPP status. The NPP status means core damage or non-core damage state of individual NPPs. The non-core damage state of a NPP is modeled in the fault tree by using a negate (a NOT gate). Authors reviewed and compared (1) quantification methods that generate exact or approximate Boolean solutions from a fault tree, (2) DTA method generating approximate Boolean solution by solving negates in a fault tree, and (3) probability calculation methods from the Boolean solutions generated by exact quantification methods or DTA method. Based on the review and comparison, a new intersection removal by probability (IRBP) method is suggested in this study for the multi-unit PSA. If the IRBP method is adopted, multi-unit PSA fault tree can be quantified without the overestimation error that is caused by the direct application of DTA method. That is, the extremely overestimated CDF can be avoided and accurate CDF can be calculated by using the IRBP method. The accuracy of the IRBP method was validated by simple multi-unit PSA models. The necessity of the IRBP method was demonstrated by the actual plant multi-unit seismic PSA models.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 춘계학술발표회
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• pp.741-742
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• 2005

• Nuclear Engineering and Technology
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• 제44권5호
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• pp.437-452
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• 2012

The Tohoku earthquake (Mw9.0) occurred on March 11, 2011 and caused a large tsunami. The Fukushima Dai-ichi NPP (F1-NPP) were overwhelmed by the tsunami and core damage occurred. This paper describes the overview of F1-NPP accident and the usability of tsunami PRA at Tohoku earthquake. The paper makes reference to the following current issues: influence on seismic hazard of gigantic aftershocks and triggered earthquakes, concepts for evaluating core damage frequency considering common cause failure with correlation coefficient against seismic event at multi units and sites, and concepts of "seismic-tsunami PSA" considering a combination of seismic motion and tsunami effects.

• 한국안전학회지
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• 제31권6호
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• pp.129-134
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• 2016

Standard design of APR+(advanced power reactor plus) was certified at 2014 by Korea regulatory body. Based on the experience gained from OPR1000 and APR1400, the APR1400 was being developed as a 1,500MWe class reactor using Korean technologies for design code, reactor coolant pump, and man-machine interface system. APR+ has been basically designed to have the seismic design basis of safe shutdown earthquake (SSE) 0.3g, a 4-train safety concept based on N+2 design philosophy, and a passive auxiliary feedwater system (PAFS). Also, safety issues on the Fukushima-type accidents have been extensively reviewed and applied to enhance APR+ safety. APR+ provides higher reliability and safety against tsunami and earthquake. The purpose of this paper is to implement probabilistic safety assessment considering these design features and to analyze sensitivity of core damage frequency for large loss of coolant accident of APR+.

• 한국압력기기공학회 논문집
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• 제9권1호
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• pp.15-19
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• 2013

The auxiliary feedwater is an important to remove the heat from the reactor core when the main feedwater system is unavailable. In most initiating events in Probabilistic Safety Assessment(PSA), the operaton of this system is required to mitigate the accidents. For one of domestic nuclear power plants, a design change of a turbine-driven auxiliary feedwater pump(TD-AFWP), pipe, and valves in the auxiliary system is implemented due to the aging related deterioration by long time operation. This change includes the replacement of the TD-AFWP, the relocation of some valves for improving the system availability, a new cross-tie line, and the installation of manual valves for maintenance. The design modification affects the PSA because the system is critical to mitigate the accidents. In this paper, the safety effect of the change of the auxiliary feedwater system is assessed with regard to the PSA view point. The results demonstrate that this change can supply the auxiliary feedwater from the TD-AFWP in the accident with the motor-driven auxiliary feedwater pump(MD-AFWP) unavailable due to test or maintenance. In addition, the change of MOV's normal position from "close" to "open" can deliver the water to steam generator in the loss of offsite power(LOOP) event. Therefore, it is confirmed that the design change of the auxiliary feedwater system reduces the total core damage frequency(CDF).

• 한국안전학회지
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• 제27권5호
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• pp.229-234
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• 2012

Operator's action time is evaluated from MAAP4 analysis used in conventional probabilistic safety assessment(PSA) of a nuclear power plant. MAAP4 code which was developed for severe accident analysis is too conservative to perform a realistic PSA. A best-estimate code such as RELAP5/MOD3, MARS has been used to reduce the conservatism of thermal hydraulic analysis. In this study, operator's action time of core cooling recovery operation is evaluated by using the MARS code, which its Fussell-Vessely(F-V) value was evaluated as highly important in a small break loss of coolant(SBLOCA) event and loss of component cooling water(LOCCW) event in previous PSA. The main conclusions were elicited : (1) MARS analysis provides larger time window for operator's action time than MAAP4 analysis and gives the more realistic time window in PSA (2) Sufficient operator's action time can reduce human error probability and core damage frequency in PSA.

• 한국정밀공학회지
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• 제30권2호
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• pp.231-235
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• 2013

The purpose of this paper is to compare with estimation of equivalent fatigue load in time domain and frequency domain and estimate the fatigue life of structure with multi-axial vibration loading. The fatigue analysis with two methods is implemented with various signals like random, sinusoidal signals. Also an equivalent fatigue life estimated by rainflow cycle counting in time domain is compared with results estimated with probability density function of each signal in frequency domain. In case of frequency domain, equivalent fatigue life can estimate through Dirlik's method with probability density function. And the work proposed in this paper compared the fatigue damage accumulated under uni-axial loading to that induced by multi-axial loading. The comparison is preformed for a simple cantilever beam, which is exposed to vibrations of several directions. For verification of estimation performance of fatigue life, results are compared to those of FEM analysis (ANSYS).

• 전기학회논문지
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• 제62권3호
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• pp.300-305
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• 2013

LOOP(Loss of Offsite Power) and SBO(Station Blackout) events have been occurring in nuclear power plants should be reviewed and be controlled on important electrical equipments by professional engineer to prevent and to safety improvement from safety assessment and reliability analysis report. LOOP and SBO occasionally happened by internal or external causes. This paper contained that LOOP frequency in the United States NPPs and in the domestic NPPs have compared and analyzed data by the past lessons and probabilistic statistics. Additionally will be installed MG(Mobile Generator) according to the lessons of Fukushima nuclear accident in Japan, which CDF(Core Damage Frequency) and LOOP frequency have reconsidered. And this paper proposed to reduce reliability criteria using PSA(Probabilistic Safety Analysis).

• Nuclear Engineering and Technology
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• 제25권4호
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• pp.515-528
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• 1993

피동형 원자력 발전소의 설계 특성상 소형 냉각재상실사고시 노심손상이 발생되지 않기 위해서는 자동감압계통의 성공적인 작동이 필수적으로 요구된다. 그러나 기수행된 연구들에서 자동감압계통의 비신뢰도가 소형 냉각재상실사고로부터 기인되는 노심손상빈도에 상당 부분을 기여하고 있음을 알 수 있다. 본 연구에서는 자동감압계통의 불능도에 기여하는 계통의 취약점을 파악함과 함께 계통의 신뢰도를 증대시키기 위한 설계개선 방안들을 제시하고 각 방안에 대한 신뢰도 분석과 함께 열수력학적 타당성 여부를 보기 위한 소형 냉각재상실사고 모의가 RELAP5/MOD3 전산 코드를 사용하여 수행되었다. 신뢰도 분석은 고장수목 기법을 이용하여 수행되어졌다.

• Nuclear Engineering and Technology
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• 제51권5호
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• pp.1307-1313
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• 2019

Recent analysis results with realistic assumptions provide the variability of operator allowable time for the initiation of aggressive cooldown under small break loss of coolant accident or steam generator tube rupture with total failure of high pressure safety injection. We investigated how plant risk may vary depending on the variability of operators' failure probability of timely initiation of aggressive cooldown. Using a probabilistic safety assessment model of a nuclear power plant, we showed that plant risks had a linear relation with the failure probability of aggressive cooldown and could be reduced by up to 10% as aggressive cooldown is more reliably performed. For individual accident management, we found that core damage potential could be gradually reduced by up to 40.49% and 63.84% after a small break loss of coolant accident or a steam generator tube rupture, respectively. Based on the importance of timely initiation of aggressive cooldown by main control room operators within the success criteria, implications for improvement of emergency operating procedures are discussed. We recommend conducting further detailed analyses of aggressive cooldown, commensurate with its importance in reducing risks in nuclear power plants.