• Nuclear Engineering and Technology
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• 제54권4호
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• pp.1516-1525
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• 2022

Interfacing the output of severe accident analysis with the input of radiological consequence analysis is an important and mandatory procedure at the beginning of Level 3 PSA. Such interfacing between the severe accident analysis code MELCOR and MACCS, one of the most commonly used consequence analysis codes, is relatively tractable since they share the same chemical groups, and the related interfacing software, MelMACCS, has already been developed. However, the linking between MAAP, another frequently used code for severe accident analyses, and MACCS has difficulties because MAAP employs a different chemical grouping method than MACCS historically did. More specifically, MAAP groups by chemical compound, while MACCS groups by chemical element. An appropriate interfacing method between MAAP and MACCS has therefore long been requested by users. This study suggests a way of extracting relevant information from MAAP results and providing proper source term information to MACCS by an appropriate treatment. Various parameters are covered in terms of magnitude and manner of release in this study, and special treatment is made for a bypass scenario. It is expected that the suggested approach will provide an important contribution as a guide to interface MAAP and MACCS when performing radiological consequence analyses.

• Nuclear Engineering and Technology
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• 제55권1호
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• pp.119-130
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• 2023

Among the various elements of probabilistic safety assessment (PSA), human failure events (HFEs) and their dependencies are major contributors to the quantification of risk of a nuclear power plant. Currently, the dependency among HFEs is reflected using a post-processing method in PSA, wherein several drawbacks, such as limited propagation of minimal cutsets through the fault tree and improper truncation of minimal cutsets exist. In this paper, we propose a method to model the HFE dependency directly in a fault tree using the if-then-else logic. The proposed method proved to be equivalent to the conventional post-processing method while addressing the drawbacks of the latter. We also developed a software tool to facilitate the implementation of the proposed method considering the need for modeling the dependency between multiple HFEs. We applied the proposed method to a specific case to demonstrate the drawbacks of the conventional post-processing method and the advantages of the proposed method. When applied appropriately under specific conditions, the direct fault-tree modeling of HFE dependency enhances the accuracy of the risk quantification and facilitates the analysis of minimal cutsets.

• Nuclear Engineering and Technology
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• 제44권5호
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• pp.471-482
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• 2012

To assess the risk of nuclear power plant operation and to determine the risk impact of digital systems, there is a need to quantitatively assess the reliability of the digital systems in a justifiable manner. The Probabilistic Risk Analysis (PRA) is a tool which can reveal shortcomings of the NPP design in general and PRA analysts have not had sufficient guiding principles in modelling particular digital components malfunctions. Currently digital I&C systems are mostly analyzed simply and conventionally in PRA, based on failure mode and effects analysis and fault tree modelling. More dynamic approaches are still in the trial stage and can be difficult to apply in full scale PRA-models. As basic events CPU failures, application software failures and common cause failures (CCF) between identical components are modelled.The primary goal is to model dependencies. However, it is not clear which failure modes or system parts CCF:s should be postulated for. A clear distinction can be made between the treatment of protection and control systems. There is a general consensus that protection systems shall be included in PRA, while control systems can be treated in a limited manner. OECD/NEA CSNI Working Group on Risk Assessment (WGRisk) has set up a task group, called DIGREL, to develop taxonomy of failure modes of digital components for the purposes of PRA. The taxonomy is aimed to be the basis of future modelling and quantification efforts. It will also help to define a structure for data collection and to review PRA studies.

• 한국안전학회지
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• 제27권5호
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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.

• 융합정보논문지
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• 제9권2호
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• pp.34-42
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• 2019

본 연구에서는 자동제어시스템이 리액터 운전에 미치는 영향을 분석해 보았다. Propylene 리액터 공정은 복잡하게 구성되어 일반적으로 효율이 낮고, 생산성이 떨어지는 문제로 인해 경제적 손실이 많다. 본 연구에서는 장애요인을 보완하여 운전효율을 높이고, 생산성을 향상시키는 목적으로 연구 모델을 제시하였다. 기존 공정의 구성을 분석하여 하드웨어 시스템 및 소프트웨어 시스템을 독창성 있는 모델로 보완하였다. 설비의 구성을 연간 60 만톤/Year 프로플린을 생산하는 리액터 8기를 기준으로 적용하였다. 연구 모델 적용결과 운전 효율이 높아졌고, 안정성 91%과 더불어 생산량이 90~95% 으로 증가하였다. 향후 연구는 생산성 100% 향상을 위한 연구모델을 제시할 예정이다. 추가적으로 프로플린 부산물인 수소 생산공정인 PSA 시스템의 안정성과 생산성 향상 방안에 대해 연구하고자 한다.

• 한국정보처리학회:학술대회논문집
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• 한국정보처리학회 2008년도 춘계학술발표대회
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• pp.257-260
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• 2008

디지털 컴퓨터와 정보처리기술의 급속한 발전과 함께 산업계 전반적으로 아날로그 기술은 쇠퇴하고 디지털 기술로 전환되고 있다. 심지어 안전-필수 기능을 담당하는 원자력발전소의 계측제어시스템에서도 제한적으로 디지털 기술을 채택하여 사용하기 시작했다. 지금까지 소프트웨어의 신뢰도의 정량화에 대한 연구는 많이 이루어져 왔으나 소프트웨어가 가지는 특수성 때문에 연구결과에 대해 전문가들의 동의를 얻지 못하고 있는 상태이다. 원자력발전소에서는 확률적 안전성 평가(PSA)를 수행할 때 소프트웨어의 실패에 기인한 위험은 무시하고 있다. 하지만, 소프트웨어를 기반으로 한 디지털 시스템의 사용이 점점 늘어남에 따라 소프트웨어 신뢰도에 대한 정량화가 점점 더 요구되고 있다. 본 연구에서는 소프트웨어의 실패모드를 정의하고 해당 실패모드에 의해 사고가 발생할 확률을 베이지안 통계이론을 이용하여 정량화하였다.

• Asian Pacific Journal of Cancer Prevention
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• 제15권19호
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• pp.8177-8182
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• 2014

Background: Blocking angiogenesis by targeting vascular endothelial growth factor (VEGF) signaling pathway to inhibit tumor growth has proven to be successful in treating a variety of different metastatic tumor types, including kidney, colon, ovarian, and lung cancers, but its role in castration-resistant prostate cancer (CRPC) is still unknown. We here aimed to determine the efficacy and toxicities of anti-VEGF agents in patients with CRPC. Materials and Methods: The databases of PubMed, Web of Science and abstracts presented at the American Society of Clinical Oncology up to March 31, 2014 were searched for relevant articles. Pooled estimates of the objective response rate (ORR) and prostate-specific antigen (PSA) response rate (decline ${\geq}50%$) were calculated using the Comprehensive Meta-Analysis (version 2.2.064) software. Median weighted progression-free survival (PFS) and overall survival (OS) time for anti-VEGF monotherapy and anti-VEGF-based doublets were compared by two-sided Student's t test. Results: A total of 3,841 patients from 19 prospective studies (4 randomized controlled trials and 15 prospective nonrandomized cohort studies) were included for analysis. The pooled ORR was 12.4% with a higher response rate of 26.4% (95%CI, 13.6-44.9%) for anti-VEGF-based combinations vs. 6.7% (95%CI, 3.5-12.7%) for anti-VEGF alone (p=0.004). Similarly, the pooled PSA response rate was 32.4% with a higher PSA response rate of 52.8% (95%CI: 40.2-65.1%) for anti-VEGF-based combinations vs. 7.3% (95%CI, 3.6-14.2%) for anti-VEGF alone (p<0.001). Median PFS and OS were 6.9 and 22.1 months with weighted median PFS of 5.6 vs. 6.9 months (p<0.001) and weighted median OS of 13.1 vs. 22.1 months (p<0.001) for anti-VEGF monotherapy vs. anti-VEGF-based doublets. Conclusions: With available evidence, this pooled analysis indicates that anti-VEGF monotherapy has a modest effect in patients with CRPC, and clinical benefits gained from anti-VEGF-based doublets appear greater than anti-VEGF monotherapy.

• 한국안전학회지
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• 제26권4호
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• pp.87-95
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• 2011

The urgent VAI(Vital Area Identification) 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 KAERI(Korea Atomic Energy Research Institute) has developed the VAI methodology and VAI software called as VIPEX(Vital area Identification Package EXpert) for identifying the vital areas. This study is to demonstrate the applicability of KAERI's VAI methodology to a hypothetical facility, and to identify the importance of information of cable and piping runs when identifying the vital areas. It is necessarily needed to consider cable and piping runs to determine the accurate and realistic TEPS(Top Event Prevention Set). If the information of cable and piping runs of a nuclear power plant is not considered when determining the TEPSs, it is absolutely impossible to acquire the complete TEPSs, and the results could be distorted by missing it. The VIPEX and FTREX(Fault Tree Reliability Evaluation eXpert) properly calculate MCSs and TEPSs using the fault tree model, and provide the most cost-effective method to save the VAI and physical protection costs.