• 한국건축시공학회:학술대회논문집
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• 한국건축시공학회 2016년도 춘계 학술논문 발표대회
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• pp.229-230
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• 2016

The Nuclear Power Plant construction industry has the related to Information-integration field. In this study, the end user developed an Information Management System early in the project, and developed a management structure that systematically integrates and interfaces with information in each life-cycle phase. Particularly this paper related to the construction information of component.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 춘계학술발표회논문집(2)
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• pp.635-640
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• 1995

Two techniques for the verification of the phase separation in ferrite phase of primary pressure bounary component materials, the primary cause of thermal aging embrittlement, are presented. Data base of room-temperature Charpy V-notch impact energy during reactor service was estimated as a measure of the degree of embrittlement. The serviceable period of CF-3 and CF-8 alloys as the primary pressure boundary components may be acceptably extended for 60 years of lifetime. However, the integrity of CF-8M alloys can be degraded seriously after several years of service in the nuclear reactor.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2467-2474
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• 2022

Nuclear Power Plants (NPPs) are the most protected facilities among all critical infrastructures (CIs). In addition to physical security, cyber security becomes a significant concern for NPPs since swift digitalization and overreliance on computer-based systems in the facility operations transformed NPPs into targets for cyber/physical attacks. Despite technical competencies, humans are still the central component of a resilient NPP to develop an effective nuclear security culture. Turkey is one of the newcomers in the nuclear energy industry, and Turkish Akkuyu NPP has a unique model owned by an international consortium. Since Turkey has limited experience in nuclear energy industry, specific multinational and multicultural characteristics of Turkish Akkuyu NPP also requires further research in terms of the Facility's prospective nuclear security. Yet, the link between "national cultures" and "nuclear security" is underestimated in nuclear security studies. By relying on Hofstede's national culture framework, our research aims to address this gap and explore possible implications of cross-national cultural differences on nuclear security. To cope with security challenges in the age of hybrid threats, we propose a security management model which addresses the need for cyber-physical security integration to cultivate a robust nuclear security culture in a multicultural working environment.

• Nuclear Engineering and Technology
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• 제53권8호
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• pp.2445-2453
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• 2021

Since the nuclear data forms a vital component in reactor physics computations, the nuclear community needs processing codes as tools for translating the Evaluated Nuclear Data Files (ENDF) to simulate nuclear-related problems such as an ACE format that is used for MCNP. Errors, inaccuracies or discrepancies in library processing may lead to a calculation that disagrees with the experimentally measured benchmark. This paper provides an overview of the processing and preparation of ENDF/B-VIII.0 incident neutron data with NECP-Atlas and NJOY codes for implementation in the MCNP code. The resulting libraries are statistically inter-compared and tested by conducting benchmark calculations, as the mutualcomparison is a source of strong feedback for further improvements in processing procedures. The database of the benchmark experiments is based on a selection taken from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP handbook) and those proposed by Russell D. Mosteller. In general, there is quite good agreement between the NECP-Atlas1.2 and NJOY21^(1.0.0.json) results with no substantial differences, if the correct input parameters are used.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.153-158
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• 1997

This paper describes a formal safety analysis technique which is demonstrated by performing empirical formal safety analysis with the case study of beamline hutch door Interlock system that is developed by using PLC(Programmable Logic Controller) systems at the Pohang Accelerator Laboratory. In order to perform formal safety analysis, we have built the Z formal specifications representation from user requirement written in ambiguous natural language and target PLC ladder logic, respectively. We have also studied the effective method to express typical PLC timer component by using specific Z formal notation which is supported by temporal history. We present a formal proof technique specifying and verifying that the hazardous states are not introduced into ladder logic in the PLC-based safety critical system.

• 한국압력기기공학회 논문집
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• 제8권1호
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• pp.25-32
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• 2012

Korea Hydro & Nuclear Power Co. (KHNP) has developed and implemented the maintenance effectiveness monitoring (MR) programs for the operating nuclear power plants. The MR program is developed by reflecting design characteristics of the operating nuclear power plants to monitor the plant performance for improving the safety and reliability. Recently, KHNP has built a new nuclear power plant, and developed the MR program to establish the advanced maintenance system by reflecting unique design characteristics based on the OPR1000 standard model. So, the MR program developed in this study has another characteristics in comparison with the OPR1000 standard model, and we will verify the suitability of the MR program through evaluating initial performance of the plant. The safety and reliability of the new plant will be improved by developing and implementing the MR program.

• Nuclear Engineering and Technology
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• 제45권3호
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• pp.277-294
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• 2013

In the early 1980s, attention focused on the possibility that pressurized thermal shock (PTS) events could challenge the integrity of a nuclear reactor pressure vessel (RPV) because operational experience suggested that overcooling events, while not common, did occur, and because the results of in-reactor materials surveillance programs showed that RPV steels and welds, particularly those having high copper content, experience a loss of toughness with time due to neutron irradiation embrittlement. These recognitions motivated analysis of PTS and the development of toughness limits for safe operation. It is now widely recognized that state of knowledge and data limitations from this time necessitated conservative treatment of several key parameters and models used in the probabilistic calculations that provided the technical of the PTS Rule, 10 CFR 50.61. To remove the unnecessary burden imposed by these conservatisms, and to improve the NRC's efficiency in processing exemption and license exemption requests, the NRC undertook the PTS re-evaluation project. This paper provides a synopsis of the results of that project, and the resulting Alternate PTS rule, 10 CFR 50.61a.

• Nuclear Engineering and Technology
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• 제51권1호
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• pp.146-154
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• 2019

An improved principal component analysis (PCA) method is applied for sensor fault detection and isolation (FDI) in a nuclear power plant (NPP) in this paper. Data pre-processing and false alarm reducing methods are combined with general PCA method to improve the model performance in practice. In data pre-processing, singular points and random fluctuations in the original data are eliminated with various techniques respectively. In fault detecting, a statistics-based method is proposed to reduce the false alarms of $T^2$ and Q statistics. Finally, the effects of the proposed data pre-processing and false alarm reducing techniques are evaluated with sensor measurements from a real NPP. They are proved to be greatly beneficial to the improvement on the reliability and stability of PCA model. Meanwhile various sensor faults are imposed to normal measurements to test the FDI ability of the PCA model. Simulation results show that the proposed PCA model presents favorable performance on the FDI of sensors no matter with major or small failures.

• 한국정보처리학회:학술대회논문집
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• 한국정보처리학회 2010년도 추계학술발표대회
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• pp.465-465
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• 2010

A controlled medical vocabulary is a vital component of medical information management because it enables computers to use information meaningfully and different institutions to share the medical data. There are currently many standard medical vocabularies - SNOMED-CT, ICD-10, UMLS, GALEN, MED, etc, but none is universally accepted as an optimal controlled medical vocabulary for application to medical information system. Moreover, it is difficult to settle the well-designed local data dictionary consisting of controlled medical vocabularies for the individual hospital information system (HIS). One of the major reasons is the local terminology with poor contents have been used in the hospital. Thus, as a trial, the local controlled vocabulary referencing system has being constructed in a limited medical field - nuclear medicine. We selected practical nuclear medicine terms from interpretation reports and electronic medical records, and removed ambiguity and redundancy, mapping the selected terms to standard medical vocabularies. Relationship and hierarchy structure between terms have being made, referring to standard medical vocabularies. Further studies may be warranted.

• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2670-2677
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• 2023

In the nuclear safety studies, a new trend refers to the evaluation of uncertainties as a mandatory component of best-estimate safety analysis which is a modern and technically consistent approach being known as BEPU (Best Estimate Plus Uncertainty). The major objectives of this study consist in performing a study of uncertainties/sensitivities of the major analysis results for a generic CANDU 6 Nuclear Power Plant during Station Blackout (SBO) progression to understand and characterize the sources of uncertainties and their effects on the key figure-of-merits (FOMs) predictions in severe accidents (SA). The FOMs of interest are hydrogen mass generation and event timings such as the first fuel channel failure time, beginning of the core disassembly time, core collapse time and calandria vessel failure time. The outcomes of the study, will allow an improvement of capabilities and expertise to perform uncertainty and sensitivity analysis with severe accident codes for CANDU 6 Nuclear Power Plant.