Dynamic characteristics assessment of reactor vessel internals with fluid-structure interaction
Improvement of numerical analysis methods has been required to solve complicated phenomena that occur in nuclear facilities. Particularly, fluid-structure interaction (FSI) behavior should be resolved for accurate design and evaluation of complex reactor vessel internals (RVIs) submerged in coolant....
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doaj-73d664d90a194a0f982bb171567bec5d2020-11-24T23:58:51ZengElsevierNuclear Engineering and Technology1738-57332017-10-014971513152310.1016/j.net.2017.05.003Dynamic characteristics assessment of reactor vessel internals with fluid-structure interactionSang Yun Je0Yoon-Suk Chang1Sung-Sik Kang2Department of Nuclear Engineering, Kyung Hee University, 1732 Deogyeong-daero, Giheung-gu, Yongin-si, Gyeonggi-do 17104, Republic of KoreaDepartment of Nuclear Engineering, Kyung Hee University, 1732 Deogyeong-daero, Giheung-gu, Yongin-si, Gyeonggi-do 17104, Republic of KoreaKorea Institute of Nuclear Safety, 34 Gwahak-ro, Yuseong-gu, Daejeon-si, 34142, Republic of KoreaImprovement of numerical analysis methods has been required to solve complicated phenomena that occur in nuclear facilities. Particularly, fluid-structure interaction (FSI) behavior should be resolved for accurate design and evaluation of complex reactor vessel internals (RVIs) submerged in coolant. In this study, the FSI effect on dynamic characteristics of RVIs in a typical 1,000 MWe nuclear power plant was investigated. Modal analyses of an integrated assembly were conducted by employing the fluid-structure (F-S) model as well as the traditional added-mass model. Subsequently, structural analyses were carried out using design response spectra combined with modal analysis data. Analysis results from the F-S model led to reductions of both frequency and Tresca stress compared to those values obtained using the added-mass model. Validation of the analysis method with the FSI model was also performed, from which the interface between the upper guide structure plate and the core shroud assembly lug was defined as the critical location of the typical RVIs, while all the relevant stress intensities satisfied the acceptance criteria.http://www.sciencedirect.com/science/article/pii/S173857331730147XFluid-Structure InteractionModal AnalysisReactor Vessel InternalsResponse Spectrum Analysis |
collection |
DOAJ |
language |
English |
format |
Article |
sources |
DOAJ |
author |
Sang Yun Je Yoon-Suk Chang Sung-Sik Kang |
spellingShingle |
Sang Yun Je Yoon-Suk Chang Sung-Sik Kang Dynamic characteristics assessment of reactor vessel internals with fluid-structure interaction Nuclear Engineering and Technology Fluid-Structure Interaction Modal Analysis Reactor Vessel Internals Response Spectrum Analysis |
author_facet |
Sang Yun Je Yoon-Suk Chang Sung-Sik Kang |
author_sort |
Sang Yun Je |
title |
Dynamic characteristics assessment of reactor vessel internals with fluid-structure interaction |
title_short |
Dynamic characteristics assessment of reactor vessel internals with fluid-structure interaction |
title_full |
Dynamic characteristics assessment of reactor vessel internals with fluid-structure interaction |
title_fullStr |
Dynamic characteristics assessment of reactor vessel internals with fluid-structure interaction |
title_full_unstemmed |
Dynamic characteristics assessment of reactor vessel internals with fluid-structure interaction |
title_sort |
dynamic characteristics assessment of reactor vessel internals with fluid-structure interaction |
publisher |
Elsevier |
series |
Nuclear Engineering and Technology |
issn |
1738-5733 |
publishDate |
2017-10-01 |
description |
Improvement of numerical analysis methods has been required to solve complicated phenomena that occur in nuclear facilities. Particularly, fluid-structure interaction (FSI) behavior should be resolved for accurate design and evaluation of complex reactor vessel internals (RVIs) submerged in coolant. In this study, the FSI effect on dynamic characteristics of RVIs in a typical 1,000 MWe nuclear power plant was investigated. Modal analyses of an integrated assembly were conducted by employing the fluid-structure (F-S) model as well as the traditional added-mass model. Subsequently, structural analyses were carried out using design response spectra combined with modal analysis data. Analysis results from the F-S model led to reductions of both frequency and Tresca stress compared to those values obtained using the added-mass model. Validation of the analysis method with the FSI model was also performed, from which the interface between the upper guide structure plate and the core shroud assembly lug was defined as the critical location of the typical RVIs, while all the relevant stress intensities satisfied the acceptance criteria. |
topic |
Fluid-Structure Interaction Modal Analysis Reactor Vessel Internals Response Spectrum Analysis |
url |
http://www.sciencedirect.com/science/article/pii/S173857331730147X |
work_keys_str_mv |
AT sangyunje dynamiccharacteristicsassessmentofreactorvesselinternalswithfluidstructureinteraction AT yoonsukchang dynamiccharacteristicsassessmentofreactorvesselinternalswithfluidstructureinteraction AT sungsikkang dynamiccharacteristicsassessmentofreactorvesselinternalswithfluidstructureinteraction |
_version_ |
1725449431724064768 |