• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 1991.04a
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• pp.163-166
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• 1991

현재 PWR의 동력로의 값은 대당 10억불이나 되는데, 원래 20년 수명을 예 측하고 설계된 것이나, 이를 두 배로 늘려서 40년이 수명을 가질 수 있겠는 가 하는 문제가 크게 대두되었다. 본 연구는 가장 중요한 구조물인 로심부 지지통의 수명판정조건을 제시하기 위하여 계산한 일부이다. 수명판정을 하 기 위해서는 barrel의 강제진동 응답으로부터 fluctuating stress를 구해야만 한다. 본 연구에서는 modal analysis를 이용하여 변위를 모드함수의 급수전 개의 형태로 표시하고 가진주파수가 barrel의 고유진동수와 일치하는 모드만 을 택하여 fluctuation stress를 구하였다.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2002.10a
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• pp.859-862
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• 2002

The Core Support Barrel (CSB) is a major component of Reactor Internals, and is designed to support and protect the Reactor Core. In this study, Reactor Core, Core Shroud and CSB were simplified to coaxial cylinders and then the offset of Reactor Core & Core Shroud to the dynamic characteristic of CSB was analyzed. For the beam modes, natural frequencies of the cantilevered cylinder are compared with those of the cantilevered beam. And it was found out that shear modulus must be used correctly to convert the shell model to the equivalent beam model. From the dynamic characteristics of the beam model, it was found out that natural frequencies are proportional to the length of Reactor Core & Core Shroud and inversely proportional to the mass. From the comparison with the dynamic characteristics of a beam model and a lumped-mass model it was found out that the size of lumped-mass must be determined considering both the length and the mass of Reactor Core & Core Shroud.

• Transactions of the KSME C: Technology and Education
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• v.5 no.2
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• pp.127-134
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• 2017

The reactor internals shall be secured in safety and structural integrity under various vibrational loading conditions. Thus, U.S. NRC, Regulatory Guide 1.20 requires the evaluation for the reactor internals due to acoustic induced vibration including the response to the reactor coolant pump pressure pulsation. This paper suggests a methodology to develop an analytical model of the core support barrel accounting for the fluid around the structure and to analyze the responses to the pump pulsation loads using acoustic structure interaction analysis. The analysis results were compared with those of US Palo Verde 1 CVAP and showed a good agreement. Thus, it is expected that the suggested methodology could be an efficient way to evaluate the response of the core support barrel to the pump pulsation loads.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2000.06a
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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.

• Nuclear Engineering and Technology
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• v.47 no.2
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• pp.165-175
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• 2015

The application of neutron noise analysis (NNA) to the ex-core neutron detector signal for monitoring the vibration characteristics of a reactor core support barrel (CSB) was investigated. Ex-core flux data were generated by using a nonanalog Monte Carlo neutron transport method in a simulated CSB model where the implicit capture and Russian roulette technique were utilized. First and third order beam and shell modes of CSB vibration were modeled based on parallel processing simulation. A NNA module was developed to analyze the ex-core flux data based on its time variation, normalized power spectral density, normalized cross-power spectral density, coherence, and phase differences. The data were then analyzed with a fuzzy logic module to determine the vibration characteristics. The ex-core neutron signal fluctuation was directly proportional to the CSB's vibration observed at 8Hz and15Hzin the beam mode vibration, and at 8Hz in the shell mode vibration. The coherence result between flux pairs was unity at the vibration peak frequencies. A distinct pattern of phase differences was observed for each of the vibration models. The developed fuzzy logic module demonstrated successful recognition of the vibration frequencies, modes, orders, directions, and phase differences within 0.4 ms for the beam and shell mode vibrations.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.15 no.1
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• pp.59-68
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• 2002

The cylindrical Upper Guide Structure assembly of the reactor intervals wish the Core Support Barrel and the Inner Barrel Assembly is subjected to flow induced loads horizontally which include random pressure fluctuation due to turbulent flow and pump pulsation pressures. The purpose of this papers is to perform random vibration and harmonic response analyses fort flow induced loads. The dynamic response characteristics due to random turbulence and pump pulsation loads were evaluated using the lumped mass beam model. Especially the model considered the annulus effects due to water gaps existing between cylindrical structures such as the Upper Guide Structure Barrel, the Core Support Barrel, and the Inner Barrel Assembly. The effect of the Inner Barrel Assembly inside the Upper Guide Structure assembly was studied. The peak dynamic responses lot each loading condition due to the addition of IBA were affected by the natural frequencies of the structures. Therefore the peak dynamic responses of the structures should be conservatively obtained from evaluation of dynamic analysis for various loading conditions.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 1997.04a
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• pp.217-222
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• 1997

Free vibration analysis of a Core Support Barrel shell structure is studied through experimental and finite element analysis methods. The structure is considered to be a thick shell with the ratio of thickness to radius 3/10. Finite element model is established by solid model with brick elements. Modal analyses are performed with respect to the various ratios of thickness to radius with clamped-free and free-free boundary conditions. Experimental test is done to find out how well the results are agreed with those of analysis. The comparison of the results from experiment and analysis shows a good agreement between them in general.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05d
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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.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2013.10a
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• pp.402-403
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• 2013

• Nuclear Engineering and Technology
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• v.29 no.5
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• pp.399-405
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• 1997

In this study a new phase separated spectral analysis algorithm is proposed to identify CSB vibration mode directly from ex-core neutron signals. Ex-core neutron signals can be decomposed into the global, core support barrel (CSB) beam mode, and CSB shell mode components by the new phase separation algorithm based on the characteristics of Fourier transform. By using the proposed algorithm and the conventional spectral analysis the vibration mode of the CSB and the fuel assembly of Ulchin-1 NPP were identified from measured ex-core neutron signals.