• Proceedings of the Computational Structural Engineering Institute Conference
• /
• 2007.04a
• /
• pp.217-222
• /
• 2007

An Emergency Diesel Generator (EDG) is one of the safety related equipments of a Nuclear Power Plant. The seismic capacity of an EDG in nuclear power plants influences the seismic safety of the plants significantly. A recent study showed that the increase of the seismic capacity of the EDG could reduce the core damage frequency (CDF) remarkably. It is known that the major failure mode of the EDG is a concrete coning failure due to a pulling out of the anchor bolts. The use of base isolators instead of anchor bolts can increase the seismic capacity of the EDG without any major problems. This study introduces a seismic risk analysis method and presents sample results about the seismically isolated and conventional EDG system.

• Nuclear Engineering and Technology
• /
• v.54 no.9
• /
• pp.3324-3335
• /
• 2022

Several countries show interest in the Generation-IV power reactor innovative small module (PRISM), including: Canada, Japan, Korea, Saudi Arabia and the United Kingdom. Generation IV International Forum (GIF) has recommended the utilizing of probabilistic safety assessment (PSA) in evaluating the safety of Generation-IV reactors. This paper reviews the PSA performed for PRISM using SAPHIRE 7.27 code. This work shows that the core damage frequency (CDF) of PRISM for a single module is estimated by 8.5E-8/year which is lower than the Generation-IV target that is 1E-6 core damage per year. The social risk of PRISM (likelihood of latent cancer fatality) with evacuation is estimated by 9.0E-12/year which is much lower than the basic safety objective (BSO) that is 1E-7/year. The social risk without evacuation is estimated by 1.2E- 11/year which is also much lower than the BSO. For the individual risk (likelihood of prompt fatality), it is concluded that it can be considered negligible with evacuation (1.0E-13/year). Assuming no evacuation, the individual risk is 2.7E-10/year which is again much lower than the BSO. In comparison with other PSAs performed for similar sodium fast reactors (SFRs), it shows that PRISM concept has the lowest CDF.

• Nuclear Engineering and Technology
• /
• v.50 no.8
• /
• 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
• /
• v.39 no.3
• /
• pp.215-220
• /
• 2007

The existing flooding Probabilistic Safety Analysis(PSA) was updated to reflect the Korean plant specific operating experience data into the flooding frequency to improve the PSA quality. Both the Nuclear Power Experience(NPE) database and the Korea Nuclear Pipe Failure Database(NuPIPE) databases were used in this study, and from these databases, only the Pressurized Water Reactor(PWR) data were used for the flooding frequencies of the flooding areas in the primary auxiliary building. With these databases and a Bayesian method, the flooding frequencies for the flooding areas were estimated. Subsequently, the Core Damage Frequency(CDF) for the flooding PSA of the Ulchin(UCN) unit 3 and 4 plants based on the Korean Standard Nuclear Power Plant(KSNP) internal full-power PSA model was recalculated. The evaluation results showed that sixteen flooding events are potentially significant according to the screening criterion, while there were two flooding events exceeding the screening criterion of the existing UCN 3 and 4 flooding PSA. The result was compared with two kinds of cases: (1) the flooding frequency and CDF from the method of the existing flooding PSA with the PWR and Boiled Water Reactor(BWR) data of the NPE database and the Maximum Likelihood Estimate(MLE) method and (2) the flooding frequency and CDF with the NPE database(PWR and BWR data), NuPIPE database, and a Bayesian method. From the comparison, a difference in CDF results was revealed more clearly between the CDF from this study and case (2) than between case (1) and case (2). That is, the number of flooding events exceeding the screen criterion further increased when only the PWR data were used for the primary auxiliary building than when the Korean specific data were used.

• Nuclear Engineering and Technology
• /
• v.25 no.4
• /
• pp.515-528
• /
• 1993

In a Passive PWR, the successful actuation of Automatic Depressurization System (ADS) is essentially required so that no core damage is occurred following small LOCA. But it has been shown in the previous studies that Core Damage Frequency (CDF) from small LOCA is significantly caused by unavailability of ADS. In this study, the design vulnerabilities impacting the ADS unavailability have been identified and the design improvement items have been proposed through the system reliability assessment using the fault tree methodology The impacts on CDF according to the change of system unavailability have also been analyzed. In addition, small LOCA simulation using RELAP5/MOD3 code has been performed to show the thermal-hydraulic feasibility of the suggested design enhancements.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.9 no.1
• /
• pp.15-19
• /
• 2013

The auxiliary feedwater is an important to remove the heat from the reactor core when the main feedwater system is unavailable. In most initiating events in Probabilistic Safety Assessment(PSA), the operaton of this system is required to mitigate the accidents. For one of domestic nuclear power plants, a design change of a turbine-driven auxiliary feedwater pump(TD-AFWP), pipe, and valves in the auxiliary system is implemented due to the aging related deterioration by long time operation. This change includes the replacement of the TD-AFWP, the relocation of some valves for improving the system availability, a new cross-tie line, and the installation of manual valves for maintenance. The design modification affects the PSA because the system is critical to mitigate the accidents. In this paper, the safety effect of the change of the auxiliary feedwater system is assessed with regard to the PSA view point. The results demonstrate that this change can supply the auxiliary feedwater from the TD-AFWP in the accident with the motor-driven auxiliary feedwater pump(MD-AFWP) unavailable due to test or maintenance. In addition, the change of MOV's normal position from "close" to "open" can deliver the water to steam generator in the loss of offsite power(LOOP) event. Therefore, it is confirmed that the design change of the auxiliary feedwater system reduces the total core damage frequency(CDF).

• Proceedings of the Korean Nuclear Society Conference
• /
• 1998.05a
• /
• pp.862-868
• /
• 1998

Korean Next Generation Reactor(KNGR) is a standardized evolutionary Advanced Light Water Reactor design under development Korea Power Company(KEPCO). It incorporates design enhncements such as active and passive advanced design features(ADFs) to increase the plant safety. A Preliminary level 1 Probabilistic Safety Assessment(PSA) has been performed for KNGR to examine the effect of these safety features. The preliminary PSA result shows that it meets the KNGR safety goal on core damage frequency(CDF). The result of this safety assessment shows that the four-train safety systems, and the ADFs such as Passive Secondary Cooling System (PSCS) contributes greatly to the reduction the CDF. Furthermore, several design changes are made or proposed for detailed review based on the PSA insights.

• Journal of the Korean Society of Safety
• /
• v.27 no.5
• /
• pp.224-228
• /
• 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.

• Proceedings of the KSME Conference
• /
• 2007.05b
• /
• pp.3464-3466
• /
• 2007

Power uprate is the process of increasing the maximum power level at which a commercial nuclear power plant may operate. Power uprate applications(113 units) for NPPs(Nuclear Power Plants) were recently approved in the United States. Utilities have been using power uprates since the 1970s as a way of increasing the power output of their nuclear plants. To increase the power output of a reactor, typically more highly enriched uranium fuel and/or more fresh fuel is used. This enables the reactor to produce more thermal energy and therefore more steam, driving a turbine generator to produce electricity. In this paper, the propriety of power uprate is explained through the review on the power uprate method and the changes of the physical parameters due to power uprate. The analysis results showed that the CDF(Core Damage Frequency) and LERF(Large Early Release Frequency) are affected in the current probabilistic safety assessment (PSA) model.

• Nuclear Engineering and Technology
• /
• v.53 no.2
• /
• pp.498-508
• /
• 2021

Micro reactors are increasingly being considered for utilization as distributed power sources. Hence, the probabilistic safety assessment (PSA) of a direct supercritical-CO₂-cooled fast reactor, called micro modular reactor (MMR), was performed in this study; this reactor was developed using innovative design concepts. It adopted a modular design and passive safety systems to minimize site constraints. As the MMR is in its conceptual design phase, design weaknesses and valuable safety insights could be identified during PSA. Level 1 internal event PSA was carried out involving literature survey, system characterization, identification of initiating events, transient analyses, development of event trees and fault trees, and quantification. The initiating events and scenarios significantly contributing to core damage frequency (CDF) were determined to identify design weaknesses in MMR. The most significant initiating event category contributing to CDF was the transients with the power conversion system initially available category, owing to its relatively high occurrence frequency. Further, an importance analysis revealed that the safety of MMR can be significantly improved by improving the reliability of reactor trip and passive decay heat removal system operation. The findings presented in this paper are expected to contribute toward future applications of PSA for assessing unconventional nuclear reactors in their conceptual design phases.