Transactions of the Korean Society of Pressure Vessels and Piping (한국압력기기공학회 논문집)

Korean Society of Pressure Vessels and Piping (한국압력기기공학회)
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Preliminary Analysis on IASCC Sensitivity of Core Shroud in Reactor Pressure Vessel

원자로 노심 쉬라우드의 조사유기응력부식균열 민감도 예비 분석

  • Received : 2019.11.21
  • Accepted : 2019.12.18
  • Published : 2019.12.30
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Abstract

This paper presents preliminary analysis and results on IASCC sensitivity of a core shroud in the reactor pressure vessel. First, neutron irradiation flux distribution of the reactor internals was calculated by using the Monte Carlo simulation code, MCNP6.1 and the nuclear data library, ENDF/B-VII.1. Second, based on the neutron irradiation flux distribution, temperature and stress distributions of the core shroud during normal operation were determined by performing finite element analysis using the commercial finite element analysis program, ABAQUS, considering irradiation aging-related degradation mechanisms. Last, IASCC sensitivity of the core shroud was assessed by using the IASCC sensitivity definition of EPRI MRP-211 and the finite element analysis results. As a result of the preliminary analysis, it was found that the point at which the maximum IASCC sensitivity is derived varies over operating time, initially moving from the shroud plate located in the center of the core to the top shroud plate-ring connection brace over operating time. In addition, it was concluded that IASCC will not occur on the core shroud even after 60 years of operation (40EFPYs) because the maximum IASCC sensitivity is less than 0.5.

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References

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