• Nuclear Engineering and Technology
• /
• 제50권3호
• /
• pp.319-326
• /
• 2018

Since the Fukushima Daiichi nuclear disaster, concern and worry about multiunit accidents have been increasing. Korea has a higher urgency to evaluate its site risk because its number of nuclear power plants (NPPs) and population density are higher than those in other countries. Since the 1980s, technical documents have been published on multiunit probabilistic safety assessment (PSA), but the Fukushima accident accelerated research on multiunit PSA. It is therefore necessary to summarize the present situation and draw implications for further research. This article reviews journal and conference papers on multiunit or site risk evaluation published between 2011 and 2016. The contents of the reviewed literature are classified as research status, initiators, and methodologies representing dependencies, and the insights and conclusions are consolidated. As of 2017, the regulatory authority and nuclear power utility have launched a full-scale project to assess multiunit risk in Korea. This article provides comprehensive reference materials on the necessary enabling technology for subsequent studies of multiunit or site risk assessment.

• Nuclear Engineering and Technology
• /
• 제50권8호
• /
• pp.1372-1386
• /
• 2018

Despite recent advances in multi-hazard analysis, the complexity and inherent nature of such problems make quantification of the landslide effect in a probabilistic safety assessment (PSA) of NPPs challenging. Therefore, in this paper, a practical approach was presented for performing an earthquake-induced landslide PSA for NPPs subject to seismic hazard. To demonstrate the effectiveness of the proposed approach, it was applied to Korean typical NPP in Korea as a numerical example. The assessment result revealed the quantitative probabilistic effects of peripheral slope failure and subsequent run-out effect on the risk of core damage frequency (CDF) of a NPP during the earthquake event. Parametric studies were conducted to demonstrate how parameters for slope, and physical relation between the slope and NPP, changed the CDF risk of the NPP. Finally, based on these results, the effective strategies were suggested to mitigate the CDF risk to the NPP resulting from the vulnerabilities inherent in adjacent slopes. The proposed approach can be expected to provide an effective framework for performing the earthquake-induced landslide PSA and decision support to increase NPP safety.

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

• Nuclear Engineering and Technology
• /
• 제49권2호
• /
• pp.360-372
• /
• 2017

Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general) for the postulated transient event.

• Nuclear Engineering and Technology
• /
• 제53권4호
• /
• pp.1127-1133
• /
• 2021

A Spent Fuel Pool (SFP) is designed to store spent fuel assemblies in the pool. And, a SFP cooling and cleanup system cools the SFP coolant through a heat exchanger which exchanges heat with component cooling water. If the cooling system fails or interfacing pipe (e.g., suction or discharge pipe) breaks, the cooling function may be lost, probably leading to fuel damage. In order to prevent such an incident, it is required to properly cool the spent fuel assemblies in the SFP by either recovering the cooling system or injecting water into the SFP. Probabilistic safety assessment (PSA) is a good tool to assess the SFP risk when an initiating event for the SFP occurs. Since PSA has been focused on reactor-side so far, it is required to study on the framework of PSA approach for SFP and identify the key factors in terms of fuel damage frequency (FDF) through a case study. In this study, therefore, a case study of SFP-PSA on the basis of design information of APR-1400 has been conducted quantitatively, and several sensitivity analyses have been conducted to understand the impact of the key factors on FDF.

• 대한안전경영과학회지
• /
• 제10권4호
• /
• pp.93-104
• /
• 2008

A successful fire risk assessment is depends on identification of risk, the analytical process of potential risk, on estimation of likelihood and the width and depth of consequence. Take the influence on enterprise into consideration, Fire risk assessment could carry out along the evaluation of the risk importance, the risk level and the risk acceptance. A large part of the limitation of choosing the risk assessment techniques impose restrictions on expense and time. If it is unnecessary high level risk assessment or Probabilistic Risk Assessment of buildings, in compliance with the Relative Ranking Method, Fire risk indexing and assessing is possible. As working-level technique, AHP method is useful with practical technique.

• 환경영향평가
• /
• 제31권5호
• /
• pp.321-333
• /
• 2022

본 연구에서는 대산 석유화학 산업단지에서 배출되는 BTEX (benzene toluene, ethylbenzene, and xylene)의 농도 및 분포 특성을 조사하여 지역주민에 대한 잠재적 위해성을 파악하였다. 산업단지 인근 지역주민들은 다양한 매체(공기, 물, 토양), 특히 공기를 통해 화학물질에 노출될 수 있다. 이 연구는 결정론적 및 확률론적 위해성 평가 접근 방식을 모두 사용하여 흡입에 의한 인체 건강 위험을 평가하였다. 결정론적 위해성 평가 결과 모든 지점에 대해 비발암 위해도의 유해지수(HI) 1.0보다 훨씬 낮은 결과가 나타났다. 그러나 발암 위해성 평가 결과, 산업단지 내에 위치한 A 지점에서 벤젠에 대한 초과발암위해도는 2.28×10^-6로 기준치인 1.0×10^-6을 약간 상회하는 것으로 나타났다. 또한, 해당 지점에 대한 확률론적 위해성 평가 결과, 보수적 기준인 1.0×10^-6을 초과하는 Percentile은 45.3%로 나타났으며, 민감도 분석 결과 노출시간(ET)가 결과에 미치는 영향이 가장 크다고 판단되었다. 인체 위해성 평가 결과, 에틸벤젠, 톨루엔, 자일렌에 대해서는 인체에 위해한 영향이 적은 것으로 판단되었으나, 벤젠은 초과발암위해도 기준(1.0×10^-6)을 초과하는 것으로 나타났다. 산업단지에서 공기 중 VOCs에 대한 광범위한 모니터링을 통해 이러한 잠재적 위험을 평가하기 위한 추가적인 연구가 필요하다.

• 한국안전학회지
• /
• 제18권2호
• /
• pp.98-103
• /
• 2003

Based on ASME probabilistic risk assessment (PRA) and NEI PRA peer review guidance, we evaluate a human reliability analysis (HRA) in probabilistic safety assessment (PSA) for Korea standard nuclear power plants, Ulchin Unit 3&4, to improve it performed at under design. The HRA for Ulchin Unit 3&4 is assessed as higher than Grade I based on ASME PRA standard and as higher than Grade 2 based on NEI PRA peer review guidance. The major items to be improved identified through the evaluation process are the documentation, the systematic human reliability analysis, the participitation of operators in the works and review of HRA. We suggest the guidance on the identification and qualitative screening analysis for pre-accident human errors and solve some items to be improved using the suggested guidance.

• 한국방재학회 논문집
• /
• 제9권6호
• /
• pp.57-68
• /
• 2009

최근 인적재난 발생의 불확실성에 대한 유연한 대처를 위해 확률 통계적 재난위험성 평가 및 위험관리 기술에 대한 필요성이 고조되고 있어 관련기술을 인적재난에 적용하기 위한 연구를 수행하였다. 먼저 재난위험성 평가 기법의 실효성, 경제성 및 지속가능한 시스템 구현을 위한 선제조건을 검토하였다. 이로부터 재난의 피해규모-발생확률 분포함수의 이론적 검토를 통해 확률 통계적인 재난위험 지표를 도출하고 재난안전(위험)도 평가에 활용함으로서 보다 간편한 정량적 재난위험도 평가기법을 개발하였기에 이를 소개한다. 또한 이를 활용하여 우리나라와 일본의 확률 통계적인 화재 안전유지 성능을 비교 분석하고 그 결과를 안전지수로 제시하였다. 향후 기존의 재난위험 평가기술과 융화 발전시켜 국내실정에 맞는 미래 재난 추정 및 예측 모델의 최적화 방안을 마련함으로써 지속적인 위험도 분석결과에 기반을 둔 합리적인 통합재난관리 방안 마련이 가능 할 것으로 기대된다.

• Nuclear Engineering and Technology
• /
• 제49권2호
• /
• pp.313-326
• /
• 2017

The systematic management of plant risk is crucial for enhancing the safety of nuclear power plants and for designing new nuclear power plants. Accident sequence precursor (ASP) analysis may be able to provide risk significance of operational experience by using probabilistic risk assessment to evaluate an operational event quantitatively in terms of its impact on core damage. In this study, an ASP methodology for two operation mode, full power and low power/shutdown operation, has been developed and applied to significant accident precursors that may occur during the operation of nuclear power plants. Two operational events, loss of feedwater and steam generator tube rupture, are identified as ASPs. Therefore, the ASP methodology developed in this study may contribute to identifying plant risk significance as well as to enhancing the safety of nuclear power plants by applying this methodology systematically.