• Journal of the Korean Society of Safety
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• v.15 no.2
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• pp.126-135
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• 2000

Probabilistic Safety Assessment (PSA) is an engineering analysis method to identify possible contributors to the risk from a nuclear power plant and now it has become a standard tool in safety evaluation of nuclear power plants. PSA consists of three phases named as Level 1, 2 and 3. Level 2 PSA, mainly focused in this paper, uses a step-wise approach. At first, plant damage states (PDSs) are defined from the Level 1 PSA results and they are quantified. Containment event tree (CET) is then constructed considering the physico-chemical phenomena in the containment. The quantification of CET can be assisted by a decomposition event tree (DET). Finally, source terms are quantitatively characterized by the containment failure mode. As the main benefit of PSA is to provide insights into plant design, performance and environmental impacts, including the identification of the dominant risk contributors and the comparison of options for reducing risk, this technique is expected to be applied to the industrial safety area.

• Journal of Pharmacopuncture
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• v.3 no.2
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• pp.41-54
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• 2000

This study was undertaken to determine if Plantaginis Semen Herbal Acupuncture(PSA) has a protective effect against glycerol-induced acute renal failure in rats. Rats were dehydrated for 24hr and then injected with 8 ml/kg of $50\%$ glycerol, one-half of dose in each hindlimb muscle. In experiments for PSA effect, rats received 0.1 ml of PSA extraction in both sides of corresponding Shenso($BL_{23}$) of human body for 3 days after injection of glycerol. The experimental group were di vided into the Normal group, the Control group, the PSA group. Glycerol injection decreased glomerular filtration rate and increased urine volume, serum creatinine, BUN level and fractional excretion of glucose, $Na^+$, $K^+$ and $CI^-$. These result show that glycerol injection result in acute renal failure. PSA significantly increased glomerular filtration rate and significantly decreased serum creatinine, BUN level and fractional excretion of glucose, $Na^+$ and $CI^-$ as compared Control group. This suggests that PSA could be used in prevention and treatment of acute renalfailure. However, the precise mechanisms of PSA protection remain to be determined.

• Journal of the Korean Society of Safety
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• v.12 no.4
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• pp.143-152
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• 1997

Probabilistic Safety Assessment(PSA) is an engineering analysis of the possible contributors to the risk from a nuclear power plant. It consist of three phases named as Level 1, 2 and 3. Level 1 PSA mainly focused in this paper is the phase of system analysis which includes the development of accident scenarios and the frequency estimation of each scenario. It covers also the system reliability analysis, component data analysis, and human reliability analysis. PSA have become a standard tool in safety evaluation of nuclear power plants. The main benefit of PSA is to provide insights into plant design, performance and environmental impacts, including the identification of dominant risk contributors and the comparison of options for reducing risk.

• Journal of the Korean Society of Safety
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• v.35 no.1
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• pp.97-106
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• 2020

The objective of this paper is to suggest a new quantification method for multi-unit probabilistic safety assessment (PSA) that removes the overestimation error caused by the existing delete-term approximation (DTA) based quantification method. So far, for the actual plant PSA model quantification, a fault tree with negates have been solved by the DTA method. It is well known that the DTA method induces overestimated core damage frequency (CDF) of nuclear power plant (NPP). If a PSA fault tree has negates and non-rare events, the overestimation in CDF drastically increases. Since multi-unit seismic PSA model has plant level negates and many non-rare events in the fault tree, it should be very carefully quantified in order to avoid CDF overestimation. Multi-unit PSA fault tree has normal gates and negates that represent each NPP status. The NPP status means core damage or non-core damage state of individual NPPs. The non-core damage state of a NPP is modeled in the fault tree by using a negate (a NOT gate). Authors reviewed and compared (1) quantification methods that generate exact or approximate Boolean solutions from a fault tree, (2) DTA method generating approximate Boolean solution by solving negates in a fault tree, and (3) probability calculation methods from the Boolean solutions generated by exact quantification methods or DTA method. Based on the review and comparison, a new intersection removal by probability (IRBP) method is suggested in this study for the multi-unit PSA. If the IRBP method is adopted, multi-unit PSA fault tree can be quantified without the overestimation error that is caused by the direct application of DTA method. That is, the extremely overestimated CDF can be avoided and accurate CDF can be calculated by using the IRBP method. The accuracy of the IRBP method was validated by simple multi-unit PSA models. The necessity of the IRBP method was demonstrated by the actual plant multi-unit seismic PSA models.

• Journal of the Korean Society of Safety
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• v.36 no.2
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• pp.101-110
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• 2021

This paper presents the results of an initiating event analysis as part of a Level 1 probabilistic safety assessment (PSA) for at-power internal events for the Korea Research Reactor (KRR). The PSA methodology is widely used to quantitatively assess the safety of research reactors (RRs) in the domestic nuclear industry. Initiating event frequencies are required to conduct a PSA, and they considerably affect the PSA results. Because there is no domestic database for domestic trip events, the safety of RRs is usually assessed using foreign databases. In this paper, operating experience data from the KRR for trip events were collected and analyzed in order to determine the frequency of specific initiating events. These frequencies were calculated using two approaches according to the event characteristics and data availability: (1) based on KRR operating experience or (2) using generic data.

• Asian Pacific Journal of Cancer Prevention
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• v.15 no.7
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• pp.3087-3091
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• 2014

Purpose: To determine the utility of digital rectal examination (DRE), serum total prostate specific antigen (tPSA) estimation, and transrectal ultrasound (TRUS) for the detection of prostate cancer (PCa) in men with lower urinary tract symptoms (LUTS). Materials and Methods: All patients with abnormal DRE, TRUS, or serum tPSA >4ng/ml, in any combination, underwent TRUS-guided needle biopsy. Eight cores of prostatic tissue were obtained from different areas of the peripheral prostate and examined histopathologically for the nature of the pathology. Results: PCa was detected in 151 (50.3%) patients, remaining 149 (49.7%) showed benign changes with or without active prostatitis. PCa was detected in 13 (56.5%), 9 (19.1%), 26 (28.3%), and 103 (74.6%) of patients with tPSA <4 ng/ml, 4-10 ng/ml, 10-20 ng/ml and >20 ng/ml respectively. Only 13 patients with PCa had abnormal DRE and TRUS with serum PSA <4 ng/ml. The detection rate was highest in patients with tPSA >20 ng/ml. The association between tPSA level and cancer detection was statistically significant (p<0.01). Among 209 patients with abnormal DRE and raised serum PSA, PCa was detected in 128 (61.2%). Conclusions: The incidence of PCa increases with increasing serum level of tPSA. The overall screening and detection rate can be further improved by using DRE, TRUS and TRUS-guided prostate needle biopsies.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.611-616
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• 1997

The Level 1 probabilistic safety assessment(PSA) for Wolsong(WS) 2/3/4 nuclear power plant(NPPs) in design stage is performed using the methodologies being equivalent to PWR PSA. Accident sequence evaluation program(ASEP) human reliability analysis(HRA) procedure and technique for human error rate prediction(THERP) are used in HRA of WS 2/3/4 NPPs PSA. The purpose of this paper is to introduce the procedure and methodology of HRA in WS 2/3/4 NPPs PSA. Also, this paper describes the interim results of importance analysis for human actions modeled in WS 2/3/4 PSA and the findings and recommendations of administrative control of secondary control area from the view of human factors.

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

• Nuclear Engineering and Technology
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• v.55 no.2
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• pp.623-639
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• 2023

The specific configuration of the safety systems in VVER-1000/V320 reactors allows a comprehensive study of the Loss of Coolant Accident (LOCA). In the present paper, a verification of the success criteria of the event trees headers for the medium and large break LOCA sequences is conducted. A detailed TRACEV5P5 thermal-hydraulic model of the reactor has been developed, including all safety systems. When analyzing the results of all sequences, some conservatism is observed in certain specific configurations as the success criterion of some headers is not consistent with the classic PSA level 1. Therefore, new proposals for the LOCA event trees are performed based on a reconfiguration of LOCA break ranges and the use of the expanded event trees approach.

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