• The Sea:JOURNAL OF THE KOREAN SOCIETY OF OCEANOGRAPHY
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• v.12 no.1
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• pp.50-56
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• 2007

Impact of thermal discharge of nuclear power plant on marine environment is naturally reflected in biogeochemical changes of seawater. Until now, many investigation activities and reports of impact analysis of marine environment were occasionally inadequate to provide sufficient justification in resolving controversial conflict of interests related with thermal discharge of nuclear power plant. From the scrutinized examination of recent 17 reports published in Korea between 2002 and 2004, the most apparent issue was inadequacy of sampling design and absence of significance of many measured parameters. In many cases, lack of statistical interpretation of raw data incurred subjective or ambiguous assessment results, which failed in deducing mutual concurrence between the parties of interest. In overall all, the main reasons for impairing integrity of previous reports seem to be inconsistency of environmental investigation procedures between research institutions, inadequate sampling design at the discretion of institutions and lack of objectivity of impact assessment based on statistical interpretation. Therefore, the primary goal of this study is to (1) discuss the blown issues on previous routine practice of seawater quality assessment procedures, and to (2) provide improved investigation procedures and rational standard criteria strictly based on statistical analysis for the better seawater quality assessment, which could be supported by scientific community, public agency, investigation institution and the other parties of interest.

• Nuclear Engineering and Technology
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• v.52 no.7
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• pp.1396-1408
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• 2020

Filtered containment venting system (FCVS) is one of the severe accident mitigation systems designed to release containment pressurization to maintain its integrity. The thermal-hydraulic behaviors in FCVSs are important since they affect the operation characteristics of the FCVS. In this study, a representative FCVS was modeled by RELAP5/Mod3.3 code, and the Station BlackOut (SBO) was chosen as an accident scenario. The thermal-hydraulic behaviors of an FCVS during long-term operation with two venting strategies (open-and-close strategy, open-and-non-close strategy) and the sensitivity analysis of important parameters were investigated. The results show that the FCVS can operate up to 250 h with a periodic open-and-close strategy during an SBO. Under the combined effects of steam condensation and water evaporation, the solution inventory in the FCVS increases during the venting phase and decreases during the intermission phase, showing a periodic pattern. Under this condition, the appropriate initial water level is 3-4 m; however, it should be adjusted according to the environment temperature. The FCVS can accommodate a decay heat power of 150-260 kW and may need to feed water for a higher decay heat power or drain water for a lower decay heat power during the late phase. The FCVS can function within an opening pressure range from 450 kPa to 500 kPa and a closing pressure range between 250 kPa and 350 kPa. When the open-and-non-close strategy is adopted, the solution inventory increases quickly in the early venting phase due to steam condensation and then decreases gradually due to the evaporation of water; drying-up may occur in the late venting phase. Decreasing the venting pipe diameter and increasing the initial water level can mitigate the evaporation of the scrubbing solution. These results are expected to provide useful references for the design and engineering application of FCVSs.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.27 no.5
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• pp.742-749
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• 2003

Failure of a pipeline due to local wall thinning is getting more attention in the nuclear power plant industry. Although guidelines such as ANSI/ASME B31G and ASME Code Case N597 are still useful fer assessing the integrity of a wall thinned pipeline, there are some limitations in these guidelines. For instance, these guidelines consider only pressure loading and thus neglect bending loading. However, most Pipelines in nuclear power plants are subjected to internal pressure and bending moment due to dead-weight loads and seismic loads. Therefore, an assessment procedure for locally wall thinned pipeline subjected to combined loading is needed. In this paper, three-dimensional finite element(FE) analyses were performed to simulate full-scale pipe tests conducted for various shapes of wall thinned area under internal pressure and bending moment. Maximum moments based on true ultimate stress(${\alpha}$$\sub$u,t/) were obtained from FE results to predict the failure of the pipe. These results were compared with test results, which showed good agreement. Additional finite element analyses were performed to investigate the effect of key parameters, such as wall thinned depth, wall thinned angle and wall thinned length, on maximum moment. Also, the effect of internal pressure on maximum moment was investigated. Change of internal pressure did not show significant effect on the maximum moment.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.39 no.5
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• pp.453-460
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• 2015

Since the Fukushima nuclear accident, several researchers are extensively studying the effect of torsion on the piping systems In nuclear power plants. Piping installations in power plants with a circumferential crack can be operated under combined loading conditions such as bending and torsion. ASME Code provides flaw evaluations for fully plastic fractures using limit load criteria for the structural integrity of the cracked pipes. According to the recent version of Code, combined loadings are provided only for the membrane and bending. Even though actual operating conditions have torsion loading, the methodology for evaluating torsion load is not established. This paper provides the results of limit load analyses by using finite element models for circumferentially cracked pipes under pure bending, pure torsion, and combined bending and torsion with tension. Theoretical limit load solutions based on net-section fully plastic criteria are suggested and verified with the results of finite element analyses.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2013.04a
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• pp.589-595
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• 2013

A reactor vessel internals comprehensive vibration assessment program(RVI CVAP) of an advanced power reactor 1400(APR1400) is being verified on the integrity of RVI for the design life of the plant by performing the non-prototype category-2 type on the US Nuclear Regulatory Commission Guide(NRC RG) 1.20, for which consists of a vibration and stress analysis program, a limited vibration measurement program, an inspection program, and the correlation of these programs. The aim of this paper is to describe the plan for the vibration measurement, test and acceptance criteria portion, and documentation and results of the APR1400 RVI CVAP. We will conduct the limited vibration measurement program of the APR1400 RVI CVAP according to the measurement plan and the vibration measurement testing in this paper.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.20 no.6
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• pp.219-227
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• 2019

The purpose of this study is to investigate the relationship between self - integration and frontal cognitive function in the welfare center users. The data were collected on January 18-30, 2018. 120 people in the M city, has participated in more than 65 years older. data were analyzed using frequency, mean and standard deviation, t-test, ANOVAs, and Pearson's correlation coefficient and multiple regression analysis. The results of this study showed that current satisfaction of welfare center users is statistically different from 'attitude toward life', 'acceptance of lifetime', 'satisfaction of present life' of ego integrity. Self - integration sense showed 28.3% explanatory power in current life. Effect of prefrontal cognitive function in accordance with the partner showed a 16.1% explanatory power This study suggests the necessity of policy and program development for the elderly community life to improve self - integration and frontal cognitive function in old age.

• Journal of Korea Society of Industrial Information Systems
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• v.28 no.6
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• pp.1-10
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• 2023

The nuclear power plant(NPP)'s Instrumentation and Control(I&C) system periodically conducts integrity checks for the maintenance of self-diagnostic function during normal operation. Additionally, it performs functionality and performance checks during planned preventive maintenance periods. However, there is a need for technological development to diagnose failures and prevent accidents in advance. In this paper, we studied methods for estimating the reliability function by utilizing environmental data and self-diagnostic data of the I&C equipment. To obtain failure data, we assumed probability distributions for component features of the I&C equipment and generated virtual failure data. Using this failure data, we estimated the reliability function using representative artificial intelligence(AI) models used in survival analysis(DeepSurve, DeepHit). And we also estimated the reliability function through the Cox regression model of the traditional semi-parametric method. We confirmed the feasibility through the residual lifetime calculations based on environmental and diagnostic data.

• Journal of the Korean Society for Nondestructive Testing
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• v.18 no.4
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• pp.304-312
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• 1998

In order to assure the reliability and integrity of structures such as bridges, Power and petrochemical plants, nondestructive evaluation techniques are recently playing more important roles. Among the various kinds of nondestructive evaluation techniques, ultrasonic technique is one of the most widely used methods for nondestructive inspection of internal defects in structures. For the reliable quantitative evaluation of internal defects from the experimental ultrasonic signals, a numerical analysis of ultrasonic scattering field due to a defect distribution is absolutely required. In this paper, the SH-wave scattering by multi-cavity defects using elastodynamic boundary element method is studied. The effects of shape of defects on transmitted and reflected fields are considered. The interaction of multi-cavity defects in 50-wave scattering is also investigated. Numerical calculation by the boundary element method has been carried out to predict near field solution of scattered fields of ultrasonic SH-wave. The presented results would be useful to improve the sensitivity of flaw defection for inverse analysis and pursue quantitative nondestructive evaluation for inverse problem.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.41 no.2
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• pp.157-164
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• 2017

A parallel slide gate valve (PSGV) is located between the heat recovery steam generator (HRSG) and the steam turbine in a combined cycle power plant (CCPP). It is used to control the flow of steam and runs with repetitive operations such as startups, load changes, and shutdowns during its operation period. Therefore, it is necessary to evaluate the fatigue damage and the structural integrity under a large compressive thermal stress due to the temperature difference through the valve wall thickness during the startup operations. In this paper, the thermal-structural analysis and the fatigue life evaluation of a 16-inch PSGV, which is installed on the HP steam line, is performed according to the fatigue life assessment method described in the ASME B&PVC VIII-2; the method uses the equivalent stress from the elastic stress analysis.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.26 no.5
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• pp.351-358
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• 2013

In the electric power supplying industry, outdoor sealing end (pothead) is used and sometimes it is necessary to check the seismic qualification analysis or test which is intended to demonstrate that the equipment have adequate integrity to withstand stress of the specified seismic event and still performs their function. And since the pothead is mounted on the supporting jig, the avoidance of resonance between the pothead and jig is required. In order to design jig, three types of optimization are performed to get the minimum weight while satisfying the natural frequency constraint using ANSYS. Optimal array, position and thickness of truss members of the jig are obtained through topology, shape and sizing optimization process, respectively. And seismic analysis of the pothead on the jig for given RRS acceleration computes the displacement and stress of the pothead which shows the safety of the pothead. The obtained natural frequency, mass, and member thickness of the jig are compared with those of the reference jig which was used for seismic experimental test. The numerical results of the jig in the research is more optimized than the jig used in the experimental test.