PSI - Issue 37

M. Annor-Nyarko et al. / Procedia Structural Integrity 37 (2022) 225–232 Author name / Structural Integrity Procedia 00 (2019) 000 – 000

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Nomenclature AOO Anticipated operational occurrence ASME PTS Pressurized thermal shock PWR Pressurized water reactors RPV Reactor pressure vessel SIF Stress intensity factor VCCT Virtual crack-closure technique β Isobaric cubic expansion coefficient g Gravity ∆T Temperature change ρ Density μ Viscosity Prandtl numbers D Hydraulic diameter k Thermal conductivity of water λ Thermal conductivity Mean thermal expansion coefficient

American Society of Mechanical Engineers

α E C T

Elastic modulus Specific heat capacity Material temperature

1. Introduction The service life extension of a nuclear power plant (NPP) is largely based on the integrity and material ageing management of its highly coupled structures, systems and components. A major safety requirement in NPP operations is to maintain the structural integrity of the irreplaceable reactor pressure vessel (RPV), which technically determines the feasible lifetime of the reactor (Trampus, 2018). The RPV is also required to have a large integrity margin to prevent failure due to its function as a physical barrier to the release of radioactivity under all reactor operations (Odette et al., 2019). Furthermore, severe anticipated operational occurrence (AOO) or postulated accident (PA) events may generate load concentrations in an ageing reactor pressure vessel which may significantly increase the initiation and growth of crack-like defects. A critical safety-risk to an ageing RPV is the pressurized thermal shock (PTS) phenomenon, which is characterized by a combination of high temperature and internal pressure resulting from the emergency coolant injected during an AOO or PA events (Thamaraiselvi and Vishnuvardhan, 2020). Therefore, fracture mechanic analyses of RPV subjected to all potential PTS loadings are essential in the overall safety assessment of ageing reactors requiring life extension. Results of such RPV assessments inform operators on the development of ageing and plant life management strategies that may prevent catastrophic failures. Several studies on the integrity evaluation of RPV subjected to PTS loadings are mostly initiated by postulated accident scenarios (Thamaraiselvi and Vishnuvardhan, 2020). Mora D.F. et al. has performed a fracture mechanics analysis of a PWR under Large Break Loss of Coolant Accident (LBLOCA) PTS using XFEM and TRACE system code (Mora et al., 2019). Murtaza and Hyder has also investigated the fracture mechanics analysis of the set-in nozzle of a RPV based on small break loss of coolant accident (SBLOCA) and Rancho-Seco transient events (Murtaza and Hyder, 2018). Similarly, the fracture mechanics analysis for structural integrity assessment of PWR pressure vessel under Loss of Coolant Accident (LOCA) has been evaluated by Huang K.R. et al.(Huang et al., 2016). However, the fracture mechanics analysis of an ageing RPV subjected to PTS from anticipated operational transients are rarely studied, despite their comparatively high frequency in pressurized water reactors (PWR) (Miranda, 2018; NEA/IAEA, 2010, 2020). In this study, a proposed simplified Abaqus-FRANC3D co-simulation method is used to perform the fracture mechanics analysis of an ageing double-circuit PWR RPV under the most frequent AOO PTS event inadvertent operation of safety injection system.