• 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.

• Nuclear Engineering and Technology
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• v.25 no.2
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• pp.204-214
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• 1993

Work on fracture mechanics has provided a technical basis for elimination of main coolant loop double ended guillotine breaks from the structural design basis of reactor coolant system. Without main coolant loop pipe breaks, the tributary pipe breaks must be considered as design bases until further fracture mechanics work could eliminate some of these breaks from design consideration. This paper determines the core support barrel shell responses for the 3 inch pressurizer spray line nozzle break which is expected to be the only inlet break remaining in the primary side after leak-before-break evaluation is extended to smaller size pipes in the near future. The responses are compared with those due to 14 inch safety injection nozzle break and main coolant loop pipe break. The results show that, when the leak-before-break concept is applied to the primary side piping systems with a diameter of 10 inches or over, the core support barrel shell responses due to pipe breaks in the primary side are negligible for the faulted condition design.

• 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.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.12 no.1
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• pp.15-27
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• 1999

본 연구에서는 노심지지배럴을 축솜형의 원통형 쉘로 이상화하여, 그의 모드 특성을 고찰하였다. 쉘의 모드 해석은 사용코드인 ANSYS를 이용하였으며, 일반적으로 사용하고 있는 요소인 SHELL61과 SHELL63을 이용하여 해석을 수행하였고 이들의 특성을 비교하였다. 또한 두께에 따른 모드 특성을 검토하여 쉘 요소의 사용 한계를 규정하였다. 한편 구멍이 있는 쉘과 없는 쉘의 모드 특성을 조사하여 구멍 및 그의 위치가 모드 특성에 미치는 영향을 파악하였다. 이들 모든 결과를 실험 및 이론에 의한 결과와 비교하였다.

• 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

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2013.04a
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• pp.581-582
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• 2013