• Nuclear Engineering and Technology
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• v.36 no.1
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• pp.83-103
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• 2004

The methodological framework of the Level 2 PSA appears to be currently standardized in a formalized fashion, but there have been different opinions on the way the sources of uncertainty are characterized and treated. This is primarily because the Level 2 PSA deals with complex phenomenological processes that are deterministic in nature rather than random processes, and there are no probabilistic models characterizing them clearly. As a result, the probabilistic quantification of the Level 2 PSA CET / APET is often subjected to two sources of uncertainty: (a) incomplete modeling of accident pathways or different predictions for the behavior of phenomenological events and (b) expert-to-expert variation in estimating the occurrence probability of phenomenological events. While a clear definition of the two sources of uncertainty involved in the Level 2 PSA makes it possible to treat an uncertainty in a consistent manner, careless application of these different sources of uncertainty may produce different conclusions in the decision-making process. The primary purpose of this paper is to characterize typical sources of uncertainty that would often be addressed in the Level 2 PSA and to provide a formal guidance for quantifying their impacts on the PSA Level 2 risk results. An additional purpose of this paper is to give a formal approach on how to combine random uncertainties addressed in the Level 1 PSA with subjectivistic uncertainties addressed in the Level 2 PSA.

• Radiation Oncology Journal
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• v.17 no.2
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• pp.136-140
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• 1999

Purpose : Prostate specific antigen (PSA) is a useful tumor marker, which is widely used as a diagnostic index and predictor of both treatment and follow-up result in prostate cancer. A prospective analysis was carried out to obtain the period of PSA normalization and the half life of PSA and to analyze the factors influencing the period of PSA normalization. The PSA level was checked before and serially after radical radiotherapy. Materials and Method : Twen쇼 patients with clinically localized prostate cancer who underwent radical external beam radiotherapy were enrolled in this study. Accrual period was from April 1993 to May 1998. Median follow-up period was 20 months. Radiotherapy was given to whole pelvis followed by a boost to prostate. Dose range for the whole pelvis was from 45 Gy to 50 Gy and boost dose to prostate, from 14 Gy to 20 Gy. The post-irradiation PSA normal value was under 3.0 ng/ml. The physical examination and serum PSA level evaluation were performed at 3 month interval in the first one year, and then at every 4 to 6 months. Results : PSA value was normalized in nineteen patients (95%) within 12 months. The mean period of PSA normalization was 5.3 (${\pm}$2.7) months. The half life of PSA Of the nonfailing patients was 2.1 (${\pm}$0.9) month. The nadir PSA level Of the nonfailing Patients waS 0.8 (${\pm}$0.5) ng/ml. The period of PSA normalization had the positive correlation with pretreatment PSA level (R$^{2}$=0.468). The nadir PSA level had no definite positive correlation with the pretreatment PSA level (R$^{2}$=0.075). The half life of serum PSA level also had no definite correlation with pretreatment PSA level (R$^{2}$=0.029). Conclusion :The PSA level was mostly normalized within 8 months (85%). If it has not normalized within 12 months, we should consider the residual disease in prostate or distant metastasis. In 2 patients, the PSA level increased 6 months or 20 months before clinical disease was detected. So the serum PSA level can be used as early diagnostic indicator of treatment failure.

• Nuclear Engineering and Technology
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• v.50 no.8
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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.

• Asian Pacific Journal of Cancer Prevention
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• v.13 no.1
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• pp.221-223
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• 2012

Prostate cancer (CaP) is a common reproductive cancer among men. This study was conducted to correlate the cancer incidence with Gangetic zone and to correlate the tumor marker prostate specific antigen (PSA) level in serum with different age groups and stage of malignancy. Patients suffering from CaP in the pathology unit of Mahavir Cancer Sansthan (Hospital and Research Centre), Patna, Bihar, India were studied from June 2009 to May 2010. PSA level in the serum of CaP patients was estimated by ELISA method. CaP incidence was highly recorded in Gangetic zone than the non-Gangetic zone. Maximum patients were in the 56 - 75 years age group with a marked predominance. Results of PSA examination showed that serum PSA level was not correlating with the age of patient and stage of malignancy. Significantly, elevated level of more than 10 ng/ml of PSA was recorded among the studied cancer patients. In this study, it is concluded that Gangetic zone habitat have high risk of CaP and elevated level of PSA was marked in Bihar, India.

• Nuclear Engineering and Technology
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• v.34 no.4
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• pp.331-343
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• 2002

We performed an importance analysis of In-Service Testing (157) components for Ulchin Unit 3 using the integrated evaluation method for categorizing component safety significance developed in this study. The developed method is basically aimed at having a PSA expert perform an importance analysis using PSA and its related information. The importance analysis using the developed method is initiated by ranking the component importance using quantitative PSA information. The importance analysis of the IST components not modeled in the PSA is performed through the engineering judgment, based on the expertise of PSA, and the quantitative and qualitative information for the 157 components. The PSA scope for importance analysis includes not only Level 1 and 2 internal PSA but also Level 1 external and shutdown/low power operation PSA. The importance analysis results of valves show that 167 (26.55%) of the 629 IST valves are HSSCs and 462 (73.45%) are LSSCs. Those of pumps also show that 28 (70%)of the 40157 pumps are HSSCs and 12 (30%) are LSSCs.

• Nuclear Engineering and Technology
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• v.30 no.1
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• pp.58-74
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• 1998

A PC window-based computer code, CONPAS (CONtainment Performance Analysis System), has been developed to integrate the numerical, graphical, and results-operation aspects of Level 2 probabilistic safety assessments (PSA) for nuclear power plants automatically. As a main logic for accident progression analysis, it employs a concept of the small containment phenomenological event tree (CPET) helpful to trace out visually individual accident progressions and of the detailed supporting event tree (DSET) for its detailed quantification. For the integrated analysis of Level 2 PSA, the code utilizes five distinct, but closely related modules. Its computational feasibility to real PSAs has been assessed through an application to the UCN 3&4 full scope Level 2 PSA. Compared with other existing computer codes for Level 2 PSA, the CONPAS code provides several advanced features: (1) systematic uncertainty analysis / importance analysis / sensitivity analysis, (2) table / graphical display & print, (3) employment of the recent Level 2 PSA technologies, and (4) highly effective user interface. The main purpose of this paper is to introduce the key features of CONPAS code and results of its feasibility study.

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1297-1306
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• 2019

Probabilistic safety assessments (PSA) have been used for several decades to visualize the risk level of commercial nuclear power plants (NPPs). Since the role of a human reliability analysis (HRA) is to provide human error probabilities for safety critical tasks to support PSA, PSA quality is strongly affected by HRA quality. Therefore, it is important to understand the underlying limitations or problems of HRA techniques. For this reason, this study conducted a survey among 14 subject matter experts who represent the HRA community of domestic Korean NPPs. As a result, five significant HRA issues were identified: (1) providing a technical basis for the K-HRA (Korean HRA) method, and developing dedicated HRA methods applicable to (2) diverse external events to support Level 1 PSA, (3) digital environments, (4) mobile equipment, and (5) severe accident management guideline tasks to support Level 2 PSA. In addition, an HRA method to support multi-unit PSA was emphasized because it plays an important role in the evaluation of site risk, which is one of the hottest current issues. It is believed that creating such a catalog of prioritized issues will be a good indication of research direction to improve HRA and therefore PSA quality.

• Journal of Ginseng Research
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• v.2 no.1
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• pp.17-33
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• 1977

Panax Saponin A(PSA) , one of dammarane glycosides of Korean ginseng, was labeled with $^{3}H$ or $^{14}C$ by our Previously reported Procedures of organic synthesis. Tracer studies with $^{3}H$-PSA in wino yielded the following results: 1). Oral and intraperitoneal administration of $^{3}H$-PSA resulted in the rapid appearance and prolonged retention of $^{3}H$-PSA in all organs such as liver, brain, bone marrow and spleen of mice. 2). The amount pi cellular intake of $^{3}H$-PSA was shown to have a certain level of saturation ranging from 0.4mg to 0. 7mg Per 20gm body weight of mice. Administration of $^{3}H$-PSA within the dosage of the saturation point did net give urinary excretion of 3H-PSA. On the contrary, excessive administration of $^{3}H$-PSA resulted in rapid excretion of the substance in the urine of mice.

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.975-983
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• 2020

Since the Fukushima Daiichi accident in 2011, concerns for the safety of multi-unit Nuclear Power Plant (NPP) sites have risen. This is because more than 70% of NPP sites are multi-unit sites that have two or more NPP units and a multi-unit accident occurred for the first time. After this accident, Probability Safety Assessment (PSA) has been considered in many countries as one of the tools to quantitatively assess the safety for multi-unit NPP sites. One of the biggest concerns for a multi-unit accident such as Fukushima is that the consequences (health and economic) will be significantly higher than in the case of a single-unit accident. However, many studies on multi-unit PSA have focused on Level 1 & 2 PSA, and there are many challenges in terms of public acceptance due to various speculations without an engineering background. In this study, two kinds of multi-unit Level 3 PSA for multi-unit site have been carried out. The first case was the estimation of multi-unit risk with conservative assumptions to investigate the margin between multi-unit risk and QHO, and the other was to identify the effect of time delays in releases between NPP units on the same site. Through these two kinds of assessments, we aimed at investigating the level of multi-unit risk and understanding the characteristics of risk in a multiunit context.

• Journal of the Korean Society of Safety
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• v.24 no.5
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• pp.63-68
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• 2009

The urgent VAI method development is required since "The Act of Physical Protection and Radiological Emergency that is established in 2003" requires an evaluation of physical threats in nuclear facilities and an establishment of physical protection in Korea. The VAI methodology is developed to (1) make a sabotage model by reusing existing fire/flooding/pipe break PSA models, (2) calculate MCSs and TEPSs, (3) select the most cost-effective TEPS among many TEPSs, (4) determine the compartments in a selected TEPS as vital areas, and (5) provide protection measures to the vital areas. The developed VAI methodology contains four steps, (1) collecting the internal level 1 PSA model and information, (2) developing the fire/flood/pipe rupture model based on level 1 PSA model, (3) integrating the fire/flood/pipe rupture model into the sabotage model by JSTAR, and (4) calculating MCSs and TEPS. The VAT process is performed through the VIPEX that was developed in KAERI. This methodology serves as a guide to develop a sabotage model by using existing internal and external PSA models. When this methodology is used to identify the vital areas, it provides the most cost-effective method to save the VAI and physical protection costs.