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Thermomechanical Analysis of a Reactor Pressure Vessel under Pressurized Thermal Shock Caused by Inadvertent Actuation of the Safety Injection System. SCIENCE AND TECHNOLOGY OF NUCLEAR INSTALLATIONS 2022. [DOI: 10.1155/2022/5886583] [Citation(s) in RCA: 0] [Impact Index Per Article: 0] [Reference Citation Analysis] [Abstract] [Track Full Text] [Subscribe] [Scholar Register] [Indexed: 11/18/2022]
Abstract
The damage induced pressurized thermal shock (PTS) may pose to a reactor pressure vessel (RPV) is a critical safety requirement assessed as part of the ageing management programme of pressurized water reactors (PWRs). A number of researches have studied PTS initiated mainly by postulated accidents such as loss of coolant accidents (LOCAs). However, investigations on PTS-induced threat on RPV caused by inadvertent actuation of the safety injection, a frequent anticipated transient, have not been thoroughly studied. In this paper, a simplified multistep analysis method is applied to study the thermomechanical status of a two-loop PWR under PTS loads caused by inadvertent actuation of the safety injection system. A direct-coupling thermomechanical analysis is performed using a three-dimensional (3D) RPV finite element model. A 3D finite element submodel (consisting of the highiest stress concentration area in the RPV) and an assumed crack are then used to perform fracture mechanics analysis. Subsequently, the critical integrity parameter-stress intensity factor (SIF) is estimated based on FRANC3D-M-integral method coupled in the multistep simulation. The material fracture toughness of the vessel is computed based on the master curve method with experimental fracture toughness data. The results obtained from the direct coupling stress analysis in comparison with sequential coupling approach demonstrate the effectiveness of the proposed multistep method. Also, comparing SIF results obtained with that calculated based on the conventional virtual crack-closure technique (VCCT) and extended finite element method (XFEM) show good agreement. This study provides a useful basis for future studies on anticipated transient-induced crack propagation and remaining service life prediction of ageing reactor pressure vessels.
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Mora DF, Niffenegger M. A new simulation approach for crack initiation, propagation and arrest in hollow cylinders under thermal shock based on XFEM. NUCLEAR ENGINEERING AND DESIGN 2022. [DOI: 10.1016/j.nucengdes.2021.111582] [Citation(s) in RCA: 2] [Impact Index Per Article: 1.0] [Reference Citation Analysis] [Track Full Text] [Subscribe] [Scholar Register] [Indexed: 11/25/2022]
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Integrity Evaluation of a Reactor Pressure Vessel Based on a Sequential Abaqus-FRANC3D Simulation Method. SCIENCE AND TECHNOLOGY OF NUCLEAR INSTALLATIONS 2021. [DOI: 10.1155/2021/7035787] [Citation(s) in RCA: 1] [Impact Index Per Article: 0.3] [Reference Citation Analysis] [Abstract] [Track Full Text] [Subscribe] [Scholar Register] [Indexed: 11/17/2022]
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.
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Spencer BW, Hoffman WM, Biswas S, Jiang W, Giorla A, Backman MA. Grizzly and BlackBear: Structural Component Aging Simulation Codes. NUCL TECHNOL 2021. [DOI: 10.1080/00295450.2020.1868278] [Citation(s) in RCA: 1] [Impact Index Per Article: 0.3] [Reference Citation Analysis] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 10/21/2022]
Affiliation(s)
- Benjamin W. Spencer
- Idaho National Laboratory, Computational Mechanics & Materials Department, P.O. Box 1625, Idaho Falls, Idaho 83415-3840
| | - William M. Hoffman
- Idaho National Laboratory, Systems Integration Department, P.O. Box 1625, Idaho Falls, Idaho 83415-3840
| | - Sudipta Biswas
- Idaho National Laboratory, Computational Mechanics & Materials Department, P.O. Box 1625, Idaho Falls, Idaho 83415-3840
| | - Wen Jiang
- Idaho National Laboratory, Computational Mechanics & Materials Department, P.O. Box 1625, Idaho Falls, Idaho 83415-3840
| | - Alain Giorla
- Oak Ridge National Laboratory, Fusion and Materials for Nuclear Systems Division, One Bethel Valley Road, Oak Ridge, Tennessee 37831
| | - Marie A. Backman
- University of Tennessee, Department of Nuclear Engineering, 1412 Circle Drive, Knoxville, Tennessee 37996-1410
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