• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1513-1523
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• 2017

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.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2014.10a
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• pp.399-401
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• 2014

The comprehensive vibration assessment program(CVAP) of APR+ nuclear power plant(NPP) is classified as non-prototype, category II with Palo Verde NPP as valid prototype. In this paper, CVAP for first plant of APR+ NPP is proposed. The Control Element Assembly(CEA) shroud of APR+ NPP is different from that of Palo Verde NPP. And the Core Support Barrel(CSB) outer diameter and the flow rate of normal operation of APR+ NPP are larger than those of Palo Verde NPP. Vibration and stress analysis program should be conducted to establish test acceptance criteria. Limited vibration measurement program should be implemented to establish the margin of safety, demonstrate the satisfaction of test acceptance criteria and confirm the similar vibratory response between the APR+ and Palo Verde NPP. Because of the change of normal operation condition, the nominal differences between APR+ and Palo Verde NPP in the structural and hydraulic analysis are studied to determine the measurement locations.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.20 no.1
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• pp.25-31
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• 2024

The core shroud of rector vessel internals (RVI) of OPR1000 and ARP1400 is made of Type 304 stainless steel (SS) by bending and welding process that may induce high deformation and residual stress in manufacturing. This work aims to evaluate the susceptibility of stress corrosion crack (SCC) initiation of bent parts of RVI in high temperature primary water environment. For SCC initiation test, tensile specimens were fabricated from the 90 degree bent plate of Type 304 SS (DT specimen), that is an archived part of a Korean APR1400. After the SCC initiation test, the specimen surface was thoroughly examined by optical and scanning electron microscopy, and compared to the specimen fabricated from the as-received plate of Type 304 SS (AR specimen). The surface observation revealed that SCC initiated on the AR specimen surface in typical intergranular (IG) mode, while SCC on the DT specimen occurred in transgrannular mode as well as IG mode. It was also found that the size and number of SCC on the DT specimen were larger than that on the AR specimen. This was attributable to a strain-hardening during the bending process. To compare the susceptibility of SCC initiation, total crack density (TCD) was calculated from the total crack length divided by the measured area of AR and DT specimens. TCD of DT specimen was 4.6 times higher than AR specimen in average, indicating that higher possibility of degradation of bent parts of RVI for a long-term operation.