ResearchArticle IntegrityEvaluationofaReactorPressureVesselBasedona Sequential Abaqus-FRANC3D Simulation Method M. Annor-Nyarko 1,2,3 andHongXia 1,2 1 KeyLaboratoryofNuclearSafetyandAdvancedNuclearEnergyTechnology,MinistryofIndustryandInformationTechnology, Harbin Engineering University, Harbin 150001, China 2 Fundamental Science on Nuclear Safety and Simulation Technology Laboratory, Harbin Engineering University, Harbin 150001, China 3 Nuclear Installations Directorate, Nuclear Regulatory Authority, Kwabenya, Ghana Correspondence should be addressed to Hong Xia; xiahong@hrbeu.edu.cn Received 16 May 2021; Revised 16 July 2021; Accepted 23 August 2021; Published 7 September 2021 Academic Editor: Massimo Zucchetti Copyright © 2021 M. Annor-Nyarko and Hong Xia. is is an open access article distributed under the Creative Commons Attribution License, which permits unrestricted use, distribution, and reproduction in any medium, provided the original work is properly cited. e 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. e maximal thermomechanical stress concentration was observed at the inlet nozzle- inner wall intersection. In addition, e 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. is work serves as a critical reference for the ageing management and fatigue life prediction of reactor pressure vessels. 1.Introduction e effect of ageing degradation mechanisms in nuclear power plants (NPPs) may result in a substantial loss of plant availability and costly part replacement [1]. In addition, the re-licensing regime of ageing NPPs is largely premised on the integrity assessment of critical components such as the reactor pressure vessel (RPV). e structural integrity of the RPV is a key safety priority in the operation of ageing pressurized water reactor (PWR) NPPs since it technically determines the feasible lifetime of the reactor [2]. Fur- thermore, under some NPP transient conditions, a small defect of the size of the nondestructive testing limit in an ageing RPV may rapidly grow leading to damage or failure. erefore, structural integrity analysis of vital components such as RPV subjected to potential transient loading from anticipated operational occurrences (AOOs) or postulated accidents (PAs) is essential to guarantee the safety of the whole NPP. In addition, integrity assessment results inform operators about the development of predictive maintenance and ageing management strategies that may prevent cata- strophic failures of the RPV equipment. A major risk to ageing RPVs is the exposure to pres- surized thermal shock (PTS) induced by temperature and internal pressure loads from emergency cooling water triggered by AOOs or PA events [3]. e PTS initiating Hindawi Science and Technology of Nuclear Installations Volume 2021, Article ID 7035787, 12 pages https://doi.org/10.1155/2021/7035787