• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.343-356
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• 2022

As a part of round robin analysis to develop a finite element elastic-plastic seismic analysis procedure for nuclear safety class 1 components, a series of parametric analyses was carried out on the simulated pressurizer surge line system model to investigate sensitivity of the analysis results to finite element analysis variables. The analysis on the surge line system model considered dynamic effect due to the seismic load corresponding to PGA 0.6 g and elastic-plastic material behavior based on the Chaboche combined hardening model. From the parametric analysis results, it was found that strains such as accumulated equivalent plastic strain and equivalent plastic strain are more sensitive to the analysis variables than von Mises effect stress. The parametric analysis results also identified that finite element density and ovalization option in the elbow elements have more significant effect on the analysis results than the other variables.

• Nuclear industry
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• v.3 no.4 s.14
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• pp.8-13
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• 1983

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.266.1-266
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• 2002

• Nuclear Engineering and Technology
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• v.49 no.2
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• pp.387-394
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• 2017

Seismic probabilistic risk assessment (SPRA) requires a large number of simulations to evaluate the seismic vulnerability of structural and nonstructural components in nuclear power plants. The effect of structural modeling and analysis assumptions on dynamic analysis of 3D and simplified 2D stick models of auxiliary buildings and the attached nonstructural components is investigated. Dynamic characteristics and seismic performance of building models are also evaluated, as well as the computational accuracy of the models. The presented results provide a better understanding of the dynamic behavior and seismic performance of auxiliary buildings. The results also help to quantify the impact of uncertainties associated with modeling and analysis of simplified numerical models of structural and nonstructural components subjected to seismic shaking on the predicted seismic failure probabilities of these systems.

• Journal of Power System Engineering
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• v.20 no.5
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• pp.52-59
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• 2016

Environmental Qualification has been almost developed except those of Non-Material Sub-components for valves and pumps though the time has only passed about 10years since EQ test launch of Korea. However EQ test has been performed by a few of engineers under the conditions that experience of EQ test is insufficient and EQ system is not developed completely. In recent years, Strengthen Nuclear Safety Regulation is being done Strictly Nuclear safety components Verification Procedure for Non-Material Sub-components, but the reports contain only performance test results, not Enviro nmental test methods relating to real Aging Degradation. In this Study, there were developed to performance systematically research to acquire EQ technology for five specimens of the Non-Material Sub-components in the Nuclear Power Plant.

• Nuclear Engineering and Technology
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• v.30 no.6
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• pp.609-616
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• 1998

Cast duplex stainless steels are extensively used for primary pressure boundary components. These components are, however, embrittled due to the precipitation of $\alpha$' phase by spinodal decomposition and other processes when exposed to reactor operating temperature for a design lifetime or life extension conditions. This report presents a procedure for estimating the current condition and the residual life of safety-related stainless steel components by using ANL database and correlations. The database of Charpy impact energy suggests that CF-8M grade is the most susceptible to thermal aging and CF-3 grade is the least. Thus, the integrity of CF-8M alleys may be degraded seriously and the degree of deterioration may exceed acceptance limit after several years of service in the nuclear reactors.

• Proceedings of the KSME Conference
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• 2004.11a
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• pp.102-107
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• 2004

In general, major components of nuclear power plant have been evaluated based on 2-dimensional design codes conservatively. However, more exact assessment is necessary for continued operation beyond the design life. In this paper, 3-dimensional stress and fatigue analyses reflecting full geometry and monitored operating condition of reactor pressure vessel have been carried out. The analyses results showed that conservatism of current 2-dimensional evaluation based on design transient. Therefore, it is anticipated that the schemes developed from this research such as 3-dimensional finite element modeling, stress analysis and fatigue analysis related techniques can be utilized as fundamental tools for exact lifetime evaluation and license renewal of major nuclear components.

• Journal of the Korean Society of Systems Engineering
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• v.16 no.1
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• pp.84-97
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• 2020

This paper aims to develop an optimized inventory management method for safety critical Instrument and Control (I&C) components. In this regard, the paper focuses on estimating the consumption rate of I&C components using demand forecasting methods. The target component for this paper is the Foxboro SPEC-200 controller. This component was chosen because it has highest consumption rate among the safety critical I&C components in Korean OPR-1000 NPPs. Three analytical methods were chosen in order to develop the demand forecasting methods; Poisson, Generalized Linear Model (GLM) and Bootstrapping. The results show that the GLM gives better accuracy than the other analytical methods. This is because the GLM considers the maintenance level of the component by discriminating between corrective and preventive.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1168-1172
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• 2018

This study went beyond making an indicator simply based on theoretical arguments, and explored a wide spectrum of different types of perceptions about energy safety to make a concept of energy safety for the Korean society. The energy safety schemata of people can be divided into three types. Type1 is concern about multi-level risks-responsibility-centric, type2 is concern about security and personal burden-expertise-centric, and type3 is concern about health and personal burden-responsibility-centric. Questions were designed on the basis of the characteristics, differences and commonalities of the three types of perceptions, explored through the Q methodology, and Koreans' perception of nuclear safety was examined. Based on the results of this research the following components of trust in nuclear safety were derived, risk perception, responsibility, honesty, expertise and procedural justification. The items for specifically evaluating them were developed, and factor analysis was conducted, and as a result, the validity of each item was proven. The components of the nuclear safety trust indicator do not exist independently, but influence each other continuously through interactions. For this reason, rather than focusing on any one of them, laws and systems must be improved first so that they can move together in one big frame.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05b
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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.