• 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 Earthquake Engineering Society of Korea
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• v.4 no.1
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• pp.51-62
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• 2000

핵연료집합체 검증 프로그램의 일환으로 본 연구에서는 지진과 배과파단이 핵연료집합체의 건선성에 미치는 영향을 검토하였다 원자로 노심의 상세 동적해석을 이용하여 지진 및 배과파단시 핵연료 집합체에 발생하는 전단력 굽힘 모우멘트 및 변위를 계산하였고 또한 집합체를 지지하고 있는 지지격자체의 충격력을 검토하였다 이들 하중에 대한 핵연료집합체의 응력해석을 수행하여 사고조건하에서의 구조적 건전성에 대하여 언급하였고 추후 설계시 고려할 사항을 제시하였다.

• Computational Structural Engineering
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• v.9 no.1
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• pp.47-54
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• 1996

The Control Element Assembly(CEA) shroud is one of the most important components in the reactor vessel internals for the nuclear power plant. Because of the severe modification from its original design the structural integrity of this component has been questioned. In an attempt to resolve this question, the response of the CEA shroud to a random loading in the actual operating condition is calculated analytically and experimentally and compared to the code allowables to show that it is structurally adequate and acceptable for the long term operation.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.164-169
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• 1996

핵연료집합체의 하단고정체는 상단고정체와 안내관을 통해 전달되는 하중들로부터 연료봉을 보호하는 주요부품이다. 하단고정체의 구조적 건전성을 유한요소법으로 평가하기위해 상용프로그램인 PATRAN과 ANSYS 5.1을 사용하였다. 하단고정체의 3차원 global 모델에 대한 응력해석을 수행하였으며 응력집중이 일어나는 유로구멍사이의 ligament 부분에 대해 submodeling 기법을 이용하여 해석의 정확도를 높였다. 본 연구에서 수행한 응력해석결과를 하단고정체 구조강도 시험에서 얻은 시험결과와 비교함으로서 응력해석모델에 대한 신뢰성과 보수성을 확인하였고 영광 3&4호기의 핵연료집합체에 부착된 하단고정체가 설계하중에 대해 충분한 건전성을 유지하고 있음을 증명하였다.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.331.1-331
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• 2001

• Journal of KSNVE
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• v.6 no.6
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• pp.827-834
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• 1996

The effect of pipe breaks in the secondary side is investigated as a part of the fuel assembly qualification program. Using the detailed dynamic analysis of a reactor core, peak responses for the motions induced from pipe breaks are obtained for a detailed core model. The secondary side pipe breaks such as main steam line and economizer feedwater line braksare considered because leak-before-break methodology has provided a technical basis for the elimination of double ended guillotine breaks of all high energy piping systems with a diameter of 10 inches or over in the primary side from the design basis. The dynamic responses such as fuel assembly shear force, bending moment, axial force and displacement, and spacer grid impact loads are carefully investigated. Also, the stress analysis is performed and the effect of the secondary side pipe breaks on the fuel assembly structural integrity under the faulted condition is addressed.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.28 no.11
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• pp.1813-1821
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• 2004

Until now, the 40% of wall thickness criterion, which is generally used for the plugging of steam generator tubes, has been applied only to a single cracked geometry. In the previous study by the authors, a total number of 9 local failure prediction models were introduced to estimate the coalescence load of two collinear through-wall cracks and, then, the reaction force model and plastic zone contact model were selected as the optimum ones. The objective of this study is to estimate the coalescence load of two collinear through-wall cracks in steam generator tube by using the optimum local failure prediction models. In order to investigate the applicability of the optimum local failure prediction models, a series of plastic collapse tests and corresponding finite element analyses for two collinear through-wall cracks in steam generator tube were carried out. Thereby, the applicability of the optimum local failure prediction models was verified and, finally, a coalescence evaluation diagram which can be used to determine whether the adjacent cracks detected by NDE coalesce or not has been developed.

• Journal of the Korean Society for Precision Engineering
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• v.22 no.3 s.168
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• pp.93-100
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• 2005

The top and bottom nozzles of PWR fuel assembly are connected by guide thimbles and an instrumentation tube that are connected with spacer grids. The fuel rods are inserted into the each cell of spacer grids. The loads acting on the fuel assembly are transmitted to the guide thimbles through the flow plate of top nozzle The axial loads applied to the fuel assembly are not equally distributed among the guide thimble due to the geometry of the top nozzle flow plate and spacer grid. In this study, the load concentration factors for the $17\times17$ fuel assembly were calculated. The analytical model fur the calculation of the load concentration factor of top nozzle flow plate was developed using ANSYS 5.6. The finite element analyses were performed using the model composed of top nozzle, guide thimble, and spacer grid. And, the analysis results were compared with the test results.

• Nuclear Engineering and Technology
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• v.21 no.3
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• pp.186-192
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• 1989

As a part of the design of the top end piece(TEP) for the 14$\times$14 reload fuel, various models of top end piece structure were analysed, using the ANSYS code, under fuel assembly shipping and handling load conditions. The 3-dimensional isoparametric elements were used in each model. By rearrangement of slots and holes on the adapter plate, without violating the design requirements, and also by changing the enclosure attachment method used on the adapter plate from pin joints to through-weld, the load carving capacity of the adapter plate was greatly strengthened. These concepts were adopted for the design of the 14$\times$14 reload fuel.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.18 no.4 s.70
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• pp.405-416
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• 2005

This study contains the static buckling tests and static buckling analyses for small size grids and full size grids. The buckling tests and finite element analyses were performed to evaluate the buckling characteristics of the spacer grids in a pressurized water reactor fuel assembly and to evaluate the possibility of the prediction lot the buckling strength of spacer grids. The buckling tests were performed for small size grids and full size grids, and the correlations between buckling strength and the number of straps and the correlations between buckling strength and the number of rows are derived based on the test results. The static buckling analyses were performed to identify the effect of the number of rows and the number of columns on the buckling strength of spacer grid by a finite element method using ANSYS program and the results were compared with the buckling test results.