• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.899-905
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• 1998

The analysis priority makings the recommendation to reduce the total core damage frequency (CDF) of Wolsong nuclear Power Plant nits 2/3/4 was Performed in this paper. In order to derive the recommendation, the sensitivity analysis of CDF on which major contributors effect m performed based on the accident quantification results during Level 1 Probabilistic safety assessment (PSA). Priorities were ranked in tile way that compares the CDF reduction rate with efforts required to implement those recommendations using risk matrix

• Journal of the Korean Society of Safety
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• v.27 no.5
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• pp.219-223
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• 2012

For an off-site consequence analysis at nuclear power plant, MELCOR Accident Consequence Code System(MACCS) II code is widely used as a software tool. In this study, the algorithm of web-based off-site consequence analysis program(OSCAP) using the MACCS II code was developed for an Integrated Leak Rate Test (ILRT) interval extension and Level 3 probabilistic safety assessment(PSA), and verification and validation(V&V) of the program was performed. The main input data for the MACCS II code are meteorological, population distribution and source term information. However, it requires lots of time and efforts to generate the main input data for an off-site consequence analysis using the MACCS II code. For example, the meteorological data are collected from each nuclear power site in real time, but the formats of the raw data collected are different from each site. To reduce the efforts and time for risk assessments, the web-based OSCAP has an automatic processing module which converts the format of the raw data collected from each site to the input data format of the MACCS II code. The program also provides an automatic function of converting the latest population data from Statistics Korea, the National Statistical Office, to the population distribution input data format of the MACCS II code. For the source term data, the program includes the release fraction of each source term category resulting from modular accident analysis program(MAAP) code analysis and the core inventory data from ORIGEN. These analysis results of each plant in Korea are stored in a database module of the web-based OSCAP, so the user can select the defaulted source term data of each plant without handling source term input data.

• Corrosion Science and Technology
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• v.14 no.1
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• pp.1-11
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• 2015

Local wall thinning and integrity degradation caused by several mechanisms, such as flow accelerated corrosion (FAC), cavitation, flashing and/or liquid drop impingements, are a main concern in carbon steel piping systems of nuclear power plant in terms of safety and operability. Thinned pipe management program (TPMP) had been developed and optimized to reduce the possibility of unplanned shutdown and/or power reduction due to pipe failure caused by wall thinning in the secondary side piping system. This program also consists of several technical elements such as prediction of wear rate for each component, prioritization of components for inspection, thickness measurement, calculation of actual wear and wear rate for each component. Decision making is associated with replacement or continuous service for thinned pipe components. Establishment of long-term strategy based on diagnosis of plant condition regarding overall wall thinning is also essential part of the program. Prediction models of wall thinning caused by FAC had been established for 24 operating nuclear plants. Long term strategies to manage the thinned pipe component were prepared and applied to each unit, which was reflecting plant specific design, operation, and inspection history, so that the structural integrity of piping system can be maintained. An alternative integrity assessment criterion and a computer program for thinned piping items were developed for the first time in the world, which was directly applicable to the secondary piping system of nuclear power plant. The thinned pipe management program is applied to all domestic nuclear power plants as a standard procedure form so that it contributes to preventing an accident caused by FAC.

• Nuclear Engineering and Technology
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• v.45 no.6
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• pp.791-794
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• 2013

Since the Fukushima accident in 2011, the importance of the electrical systems in nuclear power plants (NPPs) has been emphasized. The result has been that NPP regulators are enhancing their monitoring of loss of offsite power (LOOP) events. Korea Hydro & Nuclear Power Co. (KHNP) is reviewing the status and issues related to LOOPs, and is attempting to establish specific countermeasures to prevent LOOPs, because they can have severe consequences in the complicated maintenance schedule during an outage. A starting point for preventing LOOPs is the control of the loss of voltage (LOV)-initiating components. In order to reflect this in the risk assessment program, an LOV monitor is being developed for use during plant outages.

• Korean Journal of Environmental Biology
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• v.29 no.4
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• pp.373-378
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• 2011

Our aim was to investigate the genotoxicity of ambient air in the Krak$\acute{o}$w area after Fukushima Nuclear Power Plant (NPP) accident and compare with results from Chernobyl fallout. For the detection of ambient air genotoxicity the technique for screening gene mutation frequency in somatic cells of the $Tradescantia$ stamen hairs ($Trad$-SH assay) was used. Since 11th of March 2011 (Fukushima NPP accident), several pots containing at least 15 shoots of bioindicating plants were exposed to ambient air at 2 sites in the Krak$\acute{o}$w surrounding area, one in the city center, and about 100 pots in a control site (in the glasshouse of the Institute of Nuclear Physics) Continuous screening of mutations was performed. Progenies of 371,090 cells exposed were analyzed. Mutation frequency obtained in the first 10 days has shown a mean control level (GMF*100=$0.06{\pm}0.01$). At scoring period related to influence of a potential Fukushima fallout, a significant increase of gene mutation frequencies above the control level was observed at each site in the range, 0.10~0.33 depending on the location, (mean value for all sites GMF*100=$0.19{\pm}0.05$) that was associated with a strong expression of toxic effects. In the reported studies following the Chernobyl NPP accident monitoring $in$ $situ$ of the ambient air genotoxicity was performed in the period since April $29^{th}$ till June $3^{rd}$ 1986 also with Trad-SH bioindicator. In general, mutation frequency increases due to Chernobyl fallout(GMF*100=$0.43{\pm}0.02$) were corresponding to fluctuation of radioactivity in the air reported from physical measures, and to published reports about increase in chromosome aberration levels. Although, recent data obtained from monitoring of the ambient air quality in the Krak$\acute{o}$w and surroundings are lower when compared to results reported after Chernobyl NPP accident, though results express a significant increase above the control level and also are corresponding with increased air radioactivity reported from physical measurements. Statistically significant in comparison to control increase in gene mutation rates and more prolonged than that after Chernobyl fallout increase of GMF was observed during the period following the Fukushima NPP failure.

• Nuclear Engineering and Technology
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• v.49 no.2
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• pp.286-294
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• 2017

The events at the Fukushima Daiichi Nuclear Power Station in March 2011 point out, among other matters, that concurrent accidents at multiple units of a site can occur in reality. Although site risk has been deterministically considered to some extent in nuclear power plant siting and design, potential occurrence of multi-unit accident sequences at a site was not investigated in sufficient detail thus far in the nuclear power community. Therefore, there is considerable worldwide interest and research effort directed toward multi-unit site risk assessment, especially in the countries with high-density nuclear-power-plant sites such as Korea. As the technique of probabilistic safety assessment (PSA) has been successfully applied to evaluate the risk associated with operation of nuclear power plants in the past several decades, the PSA having primarily focused on single-unit risks is now being extended to the multi-unit PSA. In this paper we first characterize the site risk with explicit consideration of the risk associated with spent fuel pools as well as the reactor risks. The status of multi-unit risk assessment is discussed next, followed by a description of the emerging issues relevant to the multi-unit risk evaluation from a practical standpoint.

• Journal of the Korean Society of Safety
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• v.25 no.2
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• pp.24-28
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• 2010

The nuclear power Plant onsite AC electrical power sources are required to supply power to the engineering safety facility buses if the offsite power source is lost. Typically, Diesel Generators are used as the onsite power source. The 125 VAC buses are part of the onsite Class 1E AC and DC electrical power distribution system. The DC power distribution system ensure the availability of DC electrical power for system required to shutdown the reactor and maintain it in a safety condition after an anticipated operational occurrence or a postulated Design Base Accident. Recently, onsite DC power supply system trip occurs the loss of system function. To obtain the performance such as reliability and availability, we analyzed the cause of battery charger trip and described the improvement of DC power supply system reliability. Finally, we provide reliability performance criteria of charger in order to ensure the probabilistic goals for the safety of the nuclear power plants.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.24 no.3
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• pp.343-351
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• 2011

Nuclear containment building is the last barrier for being secure from any nuclear power plant accident. Therefore, it is very important to understand the ultimate capacity of nuclear containment building to loads associated with severe accidents. LOCA (loss of coolant accident) is considered as the basic accidental load and CANDU-type containment building is considered as a target structure in order to conduct the numerical analysis for the structural safety of a containment building. The CANDU-type containment building is a prestressed concrete shell structure which has the dome and the cylindrical wall and is reinforced with bonded tendons. In this paper, the evaluation of ultimate internal pressure capacity was carried out by nonlinear analysis of a prestressed concrete containment building using 3-dimensional structural analysis system.

• Journal of the Earthquake Engineering Society of Korea
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• v.25 no.2
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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.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4534-4550
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• 2022

Integrated severe accident codes should be capable of simulating not only specific physical phenomena but also entire plant behaviors, and in a sufficiently fast time. However, significant uncertainty may exist owing to the numerous parametric models and interactions among the various phenomena. The primary objectives of this study are to present best-practice uncertainty and sensitivity analysis results regarding the evolutions of severe accidents (SAs) and fission product source terms and to determine the effects of mitigation measures on them, as expected during a short-term station blackout (STSBO) of a reference pressurized water reactor (optimized power reactor (OPR)1000). Three reference scenarios related to the STSBO accident are considered: one base and two mitigation scenarios, and the impacts of dedicated severe accident mitigation (SAM) actions on the results of interest are analyzed (such as flammable gas generation). The uncertainties are quantified based on a random set of Monte Carlo samples per case scenario. The relative importance values of the uncertain input parameters to the results of interest are quantitatively evaluated through a relevant sensitivity/importance analysis.