• 대한전기학회논문지:전력기술부문A
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• 제48권5호
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• pp.508-516
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• 1999

This paper presents a method for constructing composite power system effective load duration curves(CMELDC) at load points by Monte Carlo method. The concept of effective load duration curves(ELDC) in power system planning is useful and important in both HLII. CMELDC can be obtained from convolution integral processing of the probability function of unsupplied power and the load duration curve at each load point. This concept is analogy to the ELEC in HLI. And, the reliability indices (LOLP, EDNS) for composite power system are evaluated using CMELDC. Differences in reliability levels between HLI and HLII come from considering with the uncertainty associated with the outages of the transmission system. It is expected that the CMELDC can be applied usefully to areas such as reliability evaluation, probabilistic production cost simulation and analytical outage cost assessment, etc. in HLII, DC load flow and Monte Carlo method are used for this study. The characteristics and effectiveness of thes methodology are illustrated by a case study of the IEEE RTS.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2006년도 추계학술대회 논문집 전력기술부문
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• pp.27-29
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• 2006

This paper deals with the application of the concept of POM to analysis of power system behavior and describes a practical method of PRA for KEPCO system. This paper presents not only marginal power flow evaluation of KEPCO system in view point of physical and operation mode by using Physical and Operational Margins (POM Ver. 2.2), which is developed by V&R Energy System Research, but also by using Probabilistic Reliability Assessment (PRA Ver.3.1), which is developed by EPRI. The ability of the method to provide insights on root causes, weak points and regional causes and effects was shown. The approach offers fast and accurate determination of bottlenecks in the transmission network and optimal mitigation measures to alleviate the identified violations.

• 조명전기설비학회논문지
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• 제20권9호
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• pp.61-69
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• 2006

The tensile strength of an overhead transmission line's conductor in response to an aging is being assessed in this paper. It is our view that, the decrement in the conductor's tensile strength is a key index that can be used to determine a conductor's end of life and a current limits. This paper describes a probabilistic method of assessing this index for main transmission lines which are responsible for the north bound power flow in the Seoul metropolitan area. Such an assessment can be a useful guide for economic system operation.

• 품질경영학회지
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• 제19권2호
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• pp.96-107
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• 1991

At the design stage of a plant, the plausible causes and pathways of release of hazardous materials are not clearly known. Thus there exist large amount of uncertainties on the consequences resulting from the operation of a fusion plant. In order to better handle such uncertain circumstances, we utilize the Probabilistic Risk Assessment(PRA) for the safety analyses on fusion power plant. In this paper, we concentrate on the tritium release accident. We develop a simple model that describes the process and flow of tritium, by which we figure out the locations of tritium inventory and their vulnerability. We construct event tree models that lead to various levels of tritium release from abnormal initiating events. Branch parameters on the event tree are assessed from the fault tree analysis. Based on the event tree models we construct influence diagram models which are more useful for the parameter updating and analysis. We briefly discuss the parameter updating scheme, and finally develop the methodology to obtain the predictive distribution of consequences resulting from the operating a fusion power plant. We also discuss the way to utilize the results of testing on sub-systems to reduce the uncertain ties on over all system.

• Nuclear Engineering and Technology
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• 제51권6호
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• pp.1540-1553
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• 2019

A nuclear power plant (NPP) is a highly complex system-of-systems as manifested through its internal systems interdependence. The negative impact of such interdependence was demonstrated through the 2011 Fukushima Daiichi nuclear disaster. As such, there is a critical need for new strategies to overcome the limitations of current risk assessment techniques (e.g. the use of static event and fault tree schemes), particularly through simulation of the nonlinear dynamic feedback mechanisms between the different NPP systems/components. As the first and key step towards developing an integrated NPP dynamic probabilistic risk assessment platform that can account for such feedback mechanisms, the current study adopts a system dynamics simulation approach to model the thermal dynamic processes in: the reactor core; the secondary coolant system; and the pressurized water reactor. The reactor core and secondary coolant system parameters used to develop system dynamics models are based on those of the Palo Verde Nuclear Generating Station. These three system dynamics models are subsequently validated, using results from published work, under different system perturbations including the change in reactivity, the steam valve coefficient, the primary coolant flow, and others. Moving forward, the developed system dynamics models can be integrated with other interacting processes within a NPP to form the basis of a dynamic system-level (systemic) risk assessment tool.

• Nuclear Engineering and Technology
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• 제51권2호
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• pp.402-409
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• 2019

When a loss of coolant accident (LOCA) occurs in a nuclear power plant, accident scenarios which can prevent core damage are defined based on break size. Current probabilistic safety assessment evaluates that core damage can be prevented under small-break LOCA (SBLOCA) and steam generator tube rupture (SGTR) with rapid cool down (RCD) strategy when all safety injection systems are unavailable. However, previous research has pointed out a limitation of RCD in terms of initiation time. Therefore, RCD success criteria estimation based on allowable coping time under a SBLOCA or SGTR when all safety injection systems are unavailable was performed based on time-line and thermal-hydraulic analyses. The time line analysis assumed a single emergency operating procedure flow, and the thermal hydraulic analysis utilized MARS-KS code with variables of break size, cooling rate, and operator allowable time. Results show while RCD is possible under SGTR, it is impossible under SBLOCA at the APR1400's current cooling rate limitation of 55 K/hr. A success criteria map for RCD under SBLOCA is suggested without cooling rate limitation.

• 대한전자공학회논문지SD
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• 제38권2호
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• pp.137-144
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• 2001

본 논문은 디지털 순서회로 설계시 상태할당 알고리즘 개발에 관한 연구로, 동적 소비전력을 감소시키기 위하여 상태변수의 변화를 최소로 하는 코드를 할당하여 상태코드가 변화하는 스위칭횟수를 줄이도록 하였다. 상태를 할당하는데는 Markov의 확률함수를 이용하여 hamming거리가 최소가 되도록 상태 천이도에서 각 상태를 연결하는 edge에 weight를 정의한 다음, 가중치를 이용하여 각 상태들간의 연결성을 고려하여 인접한 상태들간에는 가능한 적은 비트 천이를 가지도륵 모든 상태를 반복적으로 찾아 계산하였다. 비트 천이의 정도를 나타내기 위하여 cost 함수로 계산한 결과 순서회로의 종류에 따라 Lakshmikant의 알고리즘보다 최고 57.42%를 감소시킬 수 있었다.

• 품질경영학회지
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• 제26권1호
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• pp.201-212
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• 1998

A methodology of the safety analysis on the fusion power plant is introduced. It starts with the understanding of the physics and engineering of the plant followed by the assessment of the tritium inventory and flow rate. We a, pp.y the probabilistic risk assessment. An event tree that explains the propagation of the accident is constructed and then it is translated in to an influence diagram, that is accident is constructed and then it is translated in to an influence diagram, that is statistically equivalent so far as the parameter updating is concerned. We follow the Bayesian a, pp.oach where model parameters are treated as random variables. We briefly discuss the parameter updating scheme, and finally develop the methodology to obtain the predictive distribution of time to next severe accident.

• Nuclear Engineering and Technology
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• 제36권1호
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• pp.112-120
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• 2004

The purpose of this paper is to analyze the piping failure frequency for the main feedwater system in domestic nuclear power plants(NPPs) for the application to an in-service inspection(ISI), leak before break(LBB) concept, aging management program(AMP), and probabilistic safety analysis(PSA). First, a database was developed for piping failure events in domestic NPPs, and 23 domestic piping failure events were collected. Among the 23 events, 12 locations of wall thinning due to flow accelerated corrosion(FAC) were identified in the main feedwater system in 4 domestic WH 3-loop NPPs. Two types of the piping failure frequency such as the damage frequency and rupture frequency were considered in this study. The damage frequency was calculated from both the plant population data and damage(s) including crack, wall thinning, leak, and/or rupture, while the rupture frequency was estimated by using both the well-known Jeffreys method and a new method considering the degradation due to FAC. The results showed that the damage frequencies based on the number of the base metal piping susceptible to FAC ranged from $1.26{\times}10^{-3}/cr.yr\;to\;3.91{\times}10^{-3}/cr.yr$ for the main feedwater system of domestic WH 3-loop NPPs. The rupture frequencies obtained from the Jeffreys method for the main feedwater system were $1.01{\times}10^{-2}/cr.yr\;and\;4.54{\times}10^{-3}/cr.yr$ for the domestic WH 3-loop NPPs and all the other domestic PWR NPPs respectively, while those from the new method considering the degradation were higher than those from the Jeffreys method by about an order of one.

• Nuclear Engineering and Technology
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• 제48권3호
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• pp.684-701
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• 2016

The Gaussian process model (GPM) is a flexible surrogate model that can be used for nonparametric regression for multivariate problems. A unique feature of the GPM is that a prediction variance is automatically provided with the regression function. In this paper, we estimate the safety margin of a nuclear power plant by performing regression on the output of best-estimate simulations of a large-break loss-of-coolant accident with sampling of safety system configuration, sequence timing, technical specifications, and thermal hydraulic parameter uncertainties. The key aspect of our approach is that the GPM regression is only performed on the dominant input variables, the safety injection flow rate and the delay time for AC powered pumps to start representing sequence timing uncertainty, providing a predictive model for the peak clad temperature during a reflood phase. Other uncertainties are interpreted as contributors to the measurement noise of the code output and are implicitly treated in the GPM in the noise variance term, providing local uncertainty bounds for the peak clad temperature. We discuss the applicability of the foregoing method to reduce the use of conservative assumptions in best estimate plus uncertainty (BEPU) and Level 1 probabilistic safety assessment (PSA) success criteria definitions while dealing with a large number of uncertainties.