Stress assessment of baffle former bolt of PWR reactor for IASCC

摘要

Irradiation Assisted Stress Corrosion Cracking(IASCC) is considered an important time dependentdamage mechanism of the Reactor Pressure Vessel (RPV)internals for long term operation of Pressurized WaterReactors (PWR). This mechanism affects the boltsconnecting the formers and baffles. An evaluation ofstresses in the baffle bolts during normal operation iscarried out using thermo-mechanical Finite ElementAnalysis (FEA). The heat deposition and neutron fluencedata is obtained by reactor physics calculations; effectiveheat transfer coefficients (HTC) are calculated bycomputational fluid dynamics (CFD). The following stepsare followed to understand the IASCC modelling:1Effective HTC are used to obtain temperature of the RPVInternals1.Irradiation assisted creep and swelling models areimplemented2.Global elastic analysis and local plastic analysis isperformed3.Hardening due to irradiation is considered4.Evolution of bolt stresses are compared with MHI resultsThe RPV internal geometry is created based on a PWRdesign. Only 1/8th of the entire RPV internal geometry isconsidered for the CFD and FEA, taking advantage ofsymmetry to reduce calculation time. The analysis issimulated for 40 and 60 years of reactor operation.