• 한국소음진동공학회논문집
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• 제21권8호
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• pp.708-713
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• 2011

U.S. nuclear regulatory commission(NRC) regulatory guide(RG) 1.20 requires a comprehensive vibration assessment program(CVAP) for use in verifying the structural integrity of reactor vessel internals(RVI) for flow-induced vibrations prior to commercial operation. The CVAP program consist of vibration and fatigue analysis, a vibration measurement program, an inspection program, and a correlation of their results. One of the main purposes of the analysis program is to select measurement locations, however measurement locations can not be determined by only analysis results, therefore we developed selection criteria of measurement locations for advanced power reactor 1400(APR1400) RVI CVAP, It will be used to select measurement locations and instrument types for APR1400 RVI CVAP.

• Nuclear Engineering and Technology
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• 제48권3호
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• pp.758-764
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• 2016

Specimens of stainless steel reactor internals were irradiated with 240 keV protons and 6 MeV Xe ions at room temperature. Nanoindentation constant stiffness measurement tests were carried out to study the hardness variations. An irradiation hardening effect was observed in proton- and Xe-irradiated specimens and more irradiation damage causes a larger hardness increment. The Nix-Gao model was used to extract the bulk-equivalent hardness of irradiation-damaged region and critical indentation depth. A different hardening level under H and Xe irradiation was obtained and the discrepancies of displacement damage rate and ion species may be the probable reasons. It was observed that the hardness of Xe-irradiated specimens saturate at about 2 displacement/atom (dpa), whereas in the case of proton irradiation, the saturation hardness may be more than 7 dpa. This discrepancy may be due to the different damage distributions.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2000년도 춘계학술대회논문집
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• pp.679-686
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• 2000

Reactor Noise is defined as the fluctuations of measured instrumentation signals during full-power operation of reactor which have informations on reactor system dynamics such as neutron kinetics, thermal-hydraulics, and structural dynamics. Reactor noise analyses of ex-core neutron detector signals have been performed to monitor the vibration modes of reactor internals such as fuel assembly and Core Support Barrel in Yonggwang 3&4 Nuclear Power Plant. A real time mode separation technique have been developed and applied for the analyses. It has been found that the first vibration mode frequency of the fuel assembly was around 2.5 Hz, the beam and shell mode frequencies of CSB(Core Support Barrel) 8 Hz and 14.5 Hz, respectively. Also the analyses data base have been constructed for the continuous monitoring and diagnose of the reactor internals.

• 대한간호학회지
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• 제29권6호
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• pp.1434-1444
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• 1999

The purpose of this study was to investigate the influence of job characteristics on the nurses' the moderating effect of locus of control on the job satisfaction and organizational commitment and relationship between job characteristics and attitude. The sample for this study consisted of 594 nurses from 8 university hospitals. Factor analysis, Cronbach's alpha analysis, multiple regression analysis and hierarchical multiple regression analysis were used for the statistical methods. The results of this study were found that (1) autonomy among 5 core job characteristics showed positive influence on job satisfaction, (2) task significance and autonomy among 5 core job characteristics had positive influence on organizational commitment, (3) the internals of locus of control moderated the effect of job characteristics on nurses' job satisfaction, and (4) internals and externals of locus of control moderated the effect of job characteristics on nurses' organizational commitment.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2011년도 추계학술대회 논문집
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• pp.212-218
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• 2011

APR1400 RVI CVAP using the non-prototype category is being conducted to verify integrity of the RVI design and to secure the CVAP technology. The measurement programs are to confirm vibration analysis results for reactor internals during preoperational and initial startup testing and to detemine the safety margin. One of the important basis for the measurement programs is test acceptance criteria. Therefore, this paper is on establishment of response instrumentation test acceptance criteria for APR1400 RVI CVAP. The established acceptance criteria show that the stress criteria of APR1400 RVI are more conservative values than those of the valid prototype plant(Palo Verde unit 1) and, the displacement criteria of the IBA and the UGS were established to 0.03 in and 0.01 in, respectively.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2014년도 추계학술대회 논문집
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• pp.102-105
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• 2014

In accordance with U.S. Nuclear Regulatory Commission regulatory guide(NRC RG) 1.20(Rev.3), we are writing a comprehensive vibration assessment program(CVAP) design control document(DCD) and a technical report for U.S. NRC design certification(DC) of an Advanced Power Reactor 1400(APR1400) nuclear power plant(NPP). CVAP of an APR1400 NPP for U.S. NRC DC is classified as a non-prototype category 1 type. Therefore, CVAP DCD of reactor vessel internals(RVI) and steam generator internals(SGI) consist of analysis and full inspection program. However, piping system of primary and secondary system will be described as measurement program.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2012년도 추계학술대회 논문집
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• pp.300-305
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• 2012

The vibration and stress analysis program of comprehensive vibration assessment program (CVAP) is to verify theoretically the structural integrity of reactor vessel internals (RVI) and to provide the basis for selecting the locations monitored in measurement and inspection programs. This paper covers the verification of the vibration and stress analysis methodology of APR1400 RVI CVAP. The analysis methodology was developed to use 3-dimensional hydraulic and structural models with ANSYS and CFX. To validate the methodology, the hydraulic loads and structural reponses of OPR1000 were predicted and compared with the calculated and measured data in the OPR1000 RVI CVAP. Since the results predicted with this methodology were close to the measured values considerably, it was confirmed that the analysis methodology was developed properly.

• 한국압력기기공학회 논문집
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• 제4권2호
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• pp.57-61
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• 2008

During the period of retubing work for the licensing renewal, the fuel channels, calandria tubes and feeders of CANDU Reactors will be replaced, and calandria visual examination will be performed. This period is a unique opportunity to inspect the inside of the calandria. The visual inspection for the calandria vessel and its internals of Wolsong NPP is scheduled for confirming the calandria integrity. The first visual inspection for the calandria is planned in Pt. Lepreau led by AECL. The visual inspection for Wolsong NPP, led by NETEC(Nuclear Engineering & Technology Institute) of KHNP, will employ 3D laser scanner and 3D CAD Mock-up for the first time in the world, in addition to a conventional video camera. The inspection system is composed of a robot with the 3D laser scanner, a video camera and a hardness meter.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(4)
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• pp.283-288
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• 1996

Core support barrel (CSB) assembly is one of the most important reactor internals structures supporting and protecting the nuclear core during normal operation and faulted events. For Yonggwang 3 and 4 (YGN 3&4), the adequacy of the analytical response prediction of reactor internals for flow induced vibration was demonstrated through the comprehensive vibration assessment program (CVAP) performed during hot functional test. Besides, the vibration characteristics of the CSB of operating nuclear power plant can be examined via the excore neutron noise monitoring signal. In this paper data from YGN 4 analyses, CVAP, and neutron noise monitoring system are compared and evaluated. In general, the results are comparable each other and conservative enough to ensure sufficient design margin and structural integrity. Further investigations on the modelling and analyses procedure are recommended to utilize the experimental results to the maximum extent. And collection of the neutron noise data is desired to serve as a baseline information.

• Nuclear Engineering and Technology
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• 제53권8호
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• pp.2716-2727
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• 2021

The core shroud is one of the most important internal components of the reactor vessel internals because it meets the neutron fluence directly emitted by the nuclear fuel. In particular, dynamic effects for an earthquake should be evaluated with respect to the neutron irradiation flux. As a prerequisite to this study, simplified and detailed finite element models are developed for the core shroud using the ANSYS Design Parametric Language. Using the El Centro earthquake, seismic analyses are performed for the simplified and detailed core shroud models. Modal characteristics are obtained and their results are used for a time history analysis. Response spectrum analyses are also performed to access the degree of seismic excitation. The results of these analyses are compared to investigate the response characteristics between the simplified and detailed core shroud models from the time history and response spectrum analyses.