Integrity Evaluation of a Reactor Pressure Vessel Based on a Sequential Abaqus-FRANC3D Simulation Method

M Annor-Nyarko,Hong Xia,Massimo Zucchetti

doi:10.1155/2021/7035787

Abstract

The safety-risk pressurized thermal shock (PTS) have on a reactor pressure vessel (RPV) is one of the most important studies for the lifetime ageing management of a reactor. Several studies have investigated PTS induced by postulated accidents and other anticipated transients. However, there is no study that analyzes the effect of PTS induced by one of the most frequent anticipated operational occurrences—inadvertent operation of the safety injection system. In this paper, a sequential Abaqus-FRANC3D simulation method is proposed to study the integrity status of an ageing pressurized water reactor subjected to PTS induced by inadvertent actuation of the safety injection system. A sequential thermal-mechanical coupling analysis is first performed using a three-dimensional reactor pressure vessel finite element model (3D-FEM). A linear elastic fracture mechanics submodel with a postulated semielliptical surface crack is then created from the 3D-FEM. Subsequently, the submodel is used to evaluate the stress intensity factors based on the M-integral approach coupled within the proposed simulation method. Finally, the stress intensity factors (SIFs) obtained using the proposed method are compared with the conventional extended finite element method approach, and the result shows a good agreement. The maximal thermomechanical stress concentration was observed at the inlet nozzle-inner wall intersection. In addition, The ASME fracture toughness of the reactor vessel’s steel compared with SIFs show that the PTS event and crack configuration analysed may not pose a risk to the integrity of the RPV. This work serves as a critical reference for the ageing management and fatigue life prediction of reactor pressure vessels.

Highlights

IntroductionE SIS is a critical safety feature in a nuclear power plant that injects cold water into the RCS during loss of coolant accidents (LOCA)
Sequential ermomechanical Coupling Analysis. e sequential thermomechanical analysis using the FE models described in Section 3.4 was performed in Abaqus code following the multistep procedure shown in Figure 4. e thermal loads were determined separately and coupled with the pressure loads to compute the thermomechanical stresses. is sequential coupling analysis was to determine among others the location of the highest stress from the numerical simulation results
One of the most frequently anticipated operational transients in a pressurized water reactor is the inadvertent operation of the safety injection systems. is transient event can induce significant pressurized thermal shocks in the reactor pressure vessel; it is vital to estimate the structural mechanical changes associated with such cyclic loadings

Summary

Introduction

E SIS is a critical safety feature in a nuclear power plant that injects cold water into the RCS during loss of coolant accidents (LOCA). A typical SIS of a double-circuit PWR has four high-pressure safety injection (HPSI) pumps. E inlets of two of the pumps are connected to the refueling water storage tank (RWST) via power operation isolation and check valves. All inlets of the two pumps are connected to the outlets of the shutdown margin pumps via power operation isolation valves. Under the condition of the safety injection, the HPSI pumps enable cold water from the RWST to be pumped into the reactor core [19]. Events leading to the inadvertent injection of colder water directly into the downcomer result in the increase of the core power and pressurizer level. The frequent occurrence of inadvertent operation of SIS during the lifetime of PWRs either by a control system malfunction or by an operator error may cause thermal stresses in the inlet nozzle and inner surface of the RPV wall [4]

Methods

Results

Conclusion