• 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.

• 한국압력기기공학회 논문집
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• 제14권2호
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• pp.49-58
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• 2018

The regulatory periodic inspection program (PSI) conducted at every overhaul period is the most important process for confirming the safety of nuclear power plants. The PSI for operating nuclear power plants in Korea mainly consist of component level performance check that had been developed based on deterministic approach putting the same degree of importance to all the inspection items. This inspection methodology is likely to be effective for preoperational inspection. However, once the plant is put into service, the PSI must be focused on whether to minimize the risk of accident using defense-in-depth concept and risk insight. The incorporation of defense-in-depth concept and risk insight into the deterministic based safety inspection has not been well studied so far. In this study, two track approaches are proposed to make sure that core damage be avoided: one is to secure success path and the other to block the failure path in a specific event tree of PSA. The investigation shows how to select safety important components and how to set up inspection group to ensure that core damage would not occur for a given initiating event, which results in strengthening defense-in-depth level 3.

• Communications for Statistical Applications and Methods
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• 제8권2호
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• pp.453-464
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• 2001

A commercial nuclear power station contains at least tow emergency diesel generators(EDG) to control the risk of severe core damage during station blackout accidents. Therefore, the reliability of the EDG's to start and load-run on demand must be maintained at a sufficiently high level. Probabilistic safety assessments(PSA) are increasingly being used to quantify the public risk of operating potentially hazardous systems such as nuclear power reactors. In this paper, to perform PSA, we will introduce three different types of data and use Bayes procedure to estimate the error rate of nuclear power plant EDG, and using practical examples, illustrate which method is more reasonable in our situation.

• 한국안전학회지
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• 제27권5호
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• pp.224-228
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• 2012

According to the results of Probabilistic Safety Assessment(PSA) for a Nuclear Power Plant(NPP), an Emergency Power Supply(EPS) system has been considered as one of the most important safety system. Especially, the interests in the reliability of the EPS system have been increased after the severe accidents of Fukushima Daiichi. Firstly, we performed the risk assessment and the importance analysis of the EPS system based on the PSA models of the reference plant, which is the Korean standard NPP type. Considering a portable Diesel Generator(DG) system as the reliability reinforcement of the EPS system, we modified the PSA models and performed the risk impact assessment and the importance analysis. Although the reliability of the potable DG could be about 20% of the reliability of the alternative AC DG, we identified that Core Damage Frequency(CDF) was decreased by at least 4.6%. In addition, the risk impacts due to the unavailability of the EPS system on CDF were decreased.

• Structural Engineering and Mechanics
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• 제52권4호
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• pp.701-721
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• 2014

In the assessment of existing RC buildings, the reliable appraisal of the compressive strength of in-situ concrete is a fundamental step. Unfortunately, the data that can be obtained by the available testing methods are typically affected by a high level of uncertainty. Moreover, in order to derive indications about the degradation and ageing of the materials by on site tests, it is necessary to have the proper terms of comparison, that is to say, to know the reference data measured during the construction phases, that are often unavailable when the building is old. In the cases when such a comparison can be done, the in situ strength values typically turn out to be lower than the reference strength values (tests performed on taken samples during the construction). At this point, it is crucial to discern and quantify the specific effect induced by different factors: ageing of the materials; poor quality of the placement, consolidation or cure of the concrete during the construction phases; damage due to drilling. This paper presents a procedure for correlating the destructive compressive tests and non-destructive tests (ultrasonic pulse velocity tests) with the data documenting the compressive strength tested during the construction phases. The research work is aimed at identifying the factors that induce the difference between the in-situ strength and cubes taken from the concrete casting, and providing, so, useful information for the assessment procedure of the building.

• 한국해양환경ㆍ에너지학회지
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• 제15권1호
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• pp.30-37
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• 2012

본 연구는 우리나라 서해안에서의 조력발전소 건설에 따른 환경영향에 대해 분석하고 현재 중점평가항목(스코핑) 및 제도의 문제점을 진단한 후, 환경영향평가 스코핑 방안에 대해 제시하고자 하였다. 해역이용의 적정성 측면에서 명확한 사업목적과 당위성에 대한 폭넓은 공감대 형성이 중요하고 이해당사자 및 기관의 적극적인 의견수렴이 반드시 이루어져야 할 것이다. 또한 조력발전소 건설에 따른 해양생태계의 부정적인 영향과 어업인과의 갈등을 최소화하기 위하여 스코핑 방안을 다음과 같이 제시하였다. 1) 해수교환율 변화와 유속 변화에 따른 침퇴적 양상 변화, 2) 갯벌 감소로 인한 어류의 산란서식지 파괴, 3) 수산자원 및 어장의 영향과 부유사 확산 등이 중점평가사항(스코핑)으로 적용되어야 할 것이다. 또한, 국민 모두가 환경보전에 대한 의식전환을 통해 자원개발과 환경보전의 조화와 균형을 이루는 능동적 계획 수립과 정책 실현이 될 수 있도록 노력해야 할 것이다.

• 한국화재소방학회논문지
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• 제19권3호
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• pp.20-27
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• 2005

본 논문에서는 화재 모델 CFAST를 사용하여 원자력발전소에 있는 8개 펌프실의 화재 안전 수준을 평가해 보고, 그 결과를 조건부 노심손상확률에 반영하여 분석해 보았다. 분석 대상 구역은 원자력발전소 일차보조건물에 위치한 고압안전주입 펌프실 A/B,저압안전주입 펌프실 A/B, 격납건물 살수 펌프실 A/B 및 모터구동 보조급수 펌프실 A/B 총 8곳이었다. 각 화재구역의 상부층 가스 온도를 분석하였으며, 분석 결과를 바탕으로 상부층에 위치한 케이블의 손상 유무를 파악하였다. 분석 결과에 의하면 총 8개 화재구역의 상부층 케이블은 건전성을 유지하는 것으로 확인되었다. 케이블의 건전성 평가 결과를 바탕으로 각 화재구역의 조건부 노심손상확률을 평가한 결과, 기존의 보수적인 가정 하에 수행된 화재위험도분석 결과보다 약 2배정도 감소하는 보다 현실적이고 불확실성이 감소한 결과를 도출할 수 있었다.

• Steel and Composite Structures
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• 제46권5호
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• pp.621-635
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• 2023

The combination of replaceable and repairable properties in structures has introduced new approach called "Low Damage Design Structures". These structural systems are designed in such a way that through self-centering, primary members and specific connections neither suffer damage nor experience permanent deformations after being exposed to severe earthquakes. The purpose of this study is the seismic assessment of steel moment resisting frames with the aid of rigid rocking cores. To this end, three steel moment resisting frames of 4-, 8-, and 12-story buildings with and without rocking cores were developed. The nonlinear static analysis and incremental dynamic analysis were performed by considering the effects of the vertical and horizontal components of 16 strong ground motions, including far-fault and near-fault arrays. The results reveal that rocking systems benefit from better seismic performance and energy dissipation compared to moment resisting frames and thus structures experience a lower level of damage under higher intensity measures. The analyses show that the interstory drift in structures equipped with stiff rocking cores is more uniform in static and dynamic analyses. A uniform interstory drift distribution leads to a uniform distribution of the bending moment and a reduction in the structure's total weight and future maintenance costs.

• Nuclear Engineering and Technology
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• 제47권2호
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• pp.204-211
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• 2015

Probabilistic safety assessment (PSA) has had a significant role in quantitative decision-making by finding design and operational vulnerabilities and evaluating cost-benefit in improving such weak points. In particular, it has been widely used as the core methodology for risk-informed applications (RIAs). Even though the nature of PSA seeks realistic results, there are still "conservative" aspects. One of the sources for the conservatism is the assumptions of safety analysis and the estimation of failure frequency. Surveillance, diagnosis, and prognosis (SDP), utilizing massive databases and information technology, is worth highlighting in terms of its capability for alleviating the conservatism in conventional PSA. This article provides enabling techniques to solidify a method to provide time- and condition-dependent risks by integrating a conventional PSA model with condition monitoring and prognostics techniques. We will discuss how to integrate the results with frequency of initiating events (IEs) and probability of basic events (BEs). Two illustrative examples will be introduced: (1) how the failure probability of a passive system can be evaluated under different plant conditions and (2) how the IE frequency for a steam generator tube rupture (SGTR) can be updated in terms of operating time. We expect that the proposed model can take a role of annunciator to show the variation of core damage frequency (CDF) depending on operational conditions.

• Nuclear Engineering and Technology
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• 제46권1호
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• pp.11-26
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• 2014

The analyses carried out within the Seismic Probabilistic Risk Assessments (SPRAs) of Nuclear Power Plants (NPPs) are affected by significant aleatory and epistemic uncertainties. These uncertainties have to be represented and quantified coherently with the data, information and knowledge available, to provide reasonable assurance that related decisions can be taken robustly and with confidence. The amount of data, information and knowledge available for seismic risk assessment is typically limited, so that the analysis must strongly rely on expert judgments. In this paper, a Dempster-Shafer Theory (DST) framework for handling uncertainties in NPP SPRAs is proposed and applied to an example case study. The main contributions of this paper are two: (i) applying the complete DST framework to SPRA models, showing how to build the Dempster-Shafer structures of the uncertainty parameters based on industry generic data, and (ii) embedding Bayesian updating based on plant specific data into the framework. The results of the application to a case study show that the approach is feasible and effective in (i) describing and jointly propagating aleatory and epistemic uncertainties in SPRA models and (ii) providing 'conservative' bounds on the safety quantities of interest (i.e. Core Damage Frequency, CDF) that reflect the (limited) state of knowledge of the experts about the system of interest.