• 한국신뢰성학회지:신뢰성응용연구
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• 제18권3호
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• pp.271-279
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• 2018

Purpose: Probabilistic safety analysis was performed to enhance the safety and reliability of nuclear power plants because traditional deterministic approach has limitations in predicting the risk of failure by crack growth. The study introduces a probabilistic approach to establish a basis for probabilistic safety assessment of passive components. Methods: For probabilistic modeling of fatigue crack growth rate (FCGR), various FCGR tests were performed either under constant load amplitude or constant ${\Delta}K$ conditions by using heat treated X-750 at low temperature with adequate cathodic polarization. Bayesian inference was employed to update uncertainties of the FCGR model using additional information obtained from constant ${\Delta}K$ tests. Results: Four steps of Bayesian parameter updating were performed using constant ${\Delta}K$ test results. The standard deviation of the final posterior distribution was decreased by a factor of 10 comparing with that of the prior distribution. Conclusion: The method for developing a probabilistic crack growth model has been designed and demonstrated, in the paper. Alloy X-750 has been used for corrosion fatigue crack growth experiments and modeling. The uncertainties of parameters in the FCGR model were successfully reduced using the Bayesian inference whenever the updating was performed.

• Nuclear Engineering and Technology
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• 제41권4호
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• pp.391-402
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• 2009

This paper introduces the development of safety assessment technology in Korea, focusing on the activities of the Korea Atomic Energy Research Institute in the areas of system thermal hydraulics, severe accidents and probabilistic safety assessment. In the 1970s and 1980s, safety analysis codes and methodologies were introduced from the United States, France, Canada and other developed countries along with technology related to the construction and operation of nuclear power plants. The main focus was on understanding and utilizing computer codes that were sourced from abroad up to the early 1990s, when efforts to develop domestic safety analysis codes and methodologies became active. Remarkable achievements have been made over the last 15 years in the development and application of safety analysis technologies. In addition, significant experimental work has been performed to verify the safety characteristics of reactors and fuels as well as to support the development and validation of analysis methods.

• 한국압력기기공학회 논문집
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• 제13권1호
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• pp.1-15
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• 2017

After Kyeongju earthquake occurred in September 12, 2016, the seismic safety of nuclear power plants became important issue in our country. The seismic safety of nuclear power plant against beyond design basis earthquake became very important to secure the public safety. In this paper, the current status of the seismic safety assessment methodology is reviewed and some aspects for the reliability improvement of the seismic safety assessment results are proposed. Seismic margin analysis and probabilistic seismic safety assessment have been used for the seismic safety evaluation of a nuclear power pant. The basic procedure and the related issues and proposals for the probabilistic seismic safety assessment are investigated.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2084-2093
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• 2022

Probabilistic safety assessment is widely used to quantify the risks of nuclear power plants and their uncertainties. When the lognormal distribution describes the uncertainties of basic events, the uncertainty of the top event in a fault tree is approximated with the sum of lognormal random variables after minimal cutsets are obtained, and rare-event approximation is applied. As handling complicated analytic expressions for the sum of lognormal random variables is challenging, several approximation methods, especially Monte Carlo simulation, are widely used in practice for uncertainty analysis. In this study, a theoretical approach for analyzing the sum of lognormal random variables using an efficient numerical integration method is proposed for uncertainty analysis in probability safety assessments. The change of variables from correlated random variables with a complicated region of integration to independent random variables with a unit hypercube region of integration is applied to obtain an efficient numerical integration. The theoretical advantages of the proposed method over other approximation methods are shown through a benchmark problem. The proposed method provides an accurate and efficient approach to calculate the uncertainty of the top event in probabilistic safety assessment when the uncertainties of basic events are described with lognormal random variables.

• 한국안전학회지
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• 제22권4호
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• pp.7-12
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• 2007

In order to evaluate the integrity of Zr-2.5Nb pressure tubes, probabilistic fracture mechanics(PFM) approach was employed. Failure assessment diagram(FAD), plastic collapses, and critical crack lengths(CCL) were used for evaluating the failure probability as failure criteria. The Kr-FAD as failure assessment diagram was used because fracture of pressure tubes occurred in brittle manner due to hydrogen embrittlement of material by deuterium fluence. The probabilistic integrity evaluation observed AECL procedures and used fracture toughness parameters of EPRI and recently announced theory. In conclusion, the probabilistic approach using the Kr-FAD made it possible to determine major failure criterion in the pressure tube integrity evaluation.

• Nuclear Engineering and Technology
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• 제50권8호
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• pp.1234-1245
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• 2018

The risk of multi-unit nuclear power plants (NPPs) at a site has received considerable critical attention recently. However, current probabilistic safety assessment (PSA) procedures and computer code do not support multi-unit PSA because the traditional PSA structure is mostly used for the quantification of single-unit NPP risk. In this study, the main purpose is to develop a multi-unit Level 2 PSA method and apply it to full-power operating six-unit OPR1000. Multi-unit Level 2 PSA method consists of three steps: (1) development of single-unit Level 2 PSA; (2) extracting the mapping data from plant damage state to source term category; and (3) combining multi-unit Level 1 PSA results and mapping fractions. By applying developed multi-unit Level 2 PSA method into six-unit OPR1000, site containment failure probabilities in case of loss of ultimate heat sink, loss of off-site power, tsunami, and seismic event were quantified.

• Nuclear Engineering and Technology
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• 제53권4호
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• pp.1146-1156
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• 2021

Single-unit probabilistic safety assessment (SUPSA) has complex Boolean logic equations for accident sequences. Multi-unit probabilistic safety assessment (MUPSA) model is developed by revising and combining SUPSA models in order to reflect plant state combinations (PSCs). These PSCs represent combinations of core damage and non-core damage states of nuclear power plants (NPPs). Since all these Boolean logic equations have complemented gates (not gates), it is not easy to generate exact Boolean solutions. Delete-term approximation method (DTAM) has been widely applied for generating approximate minimal cut sets (MCSs) from the complex Boolean logic equations with complemented gates. By applying DTAM, approximate conditional core damage probability (CCDP) has been calculated in SUPSA and MUPSA. It was found that CCDP calculated by DTAM was overestimated when complemented gates have non-rare events. Especially, the CCDP overestimation drastically increases if seismic SUPSA or MUPSA has complemented gates with many non-rare events. The objective of this study is to suggest a new quantification method named probability subtraction method (PSM) that replaces DTAM. The PSM calculates accurate CCDP even when SUPSA or MUPSA has complemented gates with many non-rare events. In this paper, the PSM is explained, and the accuracy of the PSM is validated by its applications to a few MUPSAs.

• 전산구조공학
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• 제2권4호
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• pp.89-98
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• 1989

구조해석에 사용되는 변수들이 갖고 있는 통계적 특성을 고려하기 위해 기존의 방법에서는 경험에 입각한 안전계수를 사용하여, 변수가 갖고 있는 불확실성을 정성적으로 취급하여 구조물의 안전성을 점검하여 왔다. 그러나, 최근 확률이론에 입각한 신뢰성이론을 적용하여 구조물의 안전성을 보다 정량적으로 파악하여 충분한 경험과 실적이 부족한 새로운 형태의 구조물의 안전성 점검에도 활용하려는 시도가 많이 이루어지고 있다. 이러한 추세에 따라, 본 연구에서는 기존의 유한요소법에 확률론적 수법을 가미한 확률 유한요소법을 개발하여, 구조해석에 사용되는 변수들이 갖고 있는 불확실량들이 구조해석의 최종결과에 어떤 영향을 미치는가를 확률적으로 처리하여, 구조물의 안전성을 보다 합리적으로 평가하도록 하였다.

• 한국지진공학회논문집
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• 제25권2호
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• pp.53-58
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• 2021

Nuclear power plant's safety against seismic events is evaluated as risk values by probabilistic seismic safety assessment. The risk values vary by the seismic failure correlation between the structures, systems, and components (SSCs). However, most probabilistic seismic safety assessments idealized the seismic failure correlation between the SSCs as entirely dependent or independent. Such a consideration results in an inaccurate assessment result not reflecting real physical phenomenon. A nuclear power plant's seismic risk should be calculated with the appropriate seismic failure correlation coefficient between the SSCs for a reasonable outcome. An accident scenario that has an enormous impact on a nuclear power plant's seismic risk was selected. Moreover, the probabilistic seismic response analyses of a nuclear power plant were performed to derive appropriate seismic failure correlations between SSCs. Based on the analysis results, the seismic failure correlation coefficient between SSCs was derived, and the seismic fragility curve and core damage frequency of the loss of essential power event were calculated. Results were compared with the seismic fragility and core damage frequency of assuming the seismic failure correlations between SSCs were independent and entirely dependent.

• 전기학회논문지
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• 제58권3호
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• pp.444-449
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• 2009

This paper introduces a method to establish performance criteria of the in-house power supply system in nuclear power plants. The performance criteria of the system is presented in terms of the number of function failures and amount of the out-of-service time that can be allowed commensurate with the probabilistic safety assessment results of the nuclear power plants. To obtain the performance criteria such as reliability and availability, the functions of the system were analyzed and probabilistic assessment results were utilized. This method provides quantitative guidelines in selecting and monitoring system functions to determine an adequate level of maintenance quality in order to ensure the probabilistic goals for the safety of the nuclear power plants.