| Modelling of reactor pressure vessel subjected to pressurized thermal shock using 3D-XFEM | |
Mora DF1; Niffenegger M1; Qian GA(钱桂安)1,2 ; Jaros M3; Niceno B3
| |
| Corresponding Author | Mora, Diego F.([email protected]) |
| Source Publication | NUCLEAR ENGINEERING AND DESIGN
 |
| 2019-11-01 | |
| Volume | 353Pages:13 |
| ISSN | 0029-5493 |
| Abstract | We describe the model developed for integrity assessment of a reactor pressure vessel (RPV) subjected to pressurized thermal shock (PTS). The assessment is based on a multi-step simulation scheme, which includes the thermo-hydraulic, thermo-mechanical and fracture mechanics analyses. The thermo-mechanical model uses the three-dimensional finite element method (FEM) to calculate the stress fields of the reactor pressure vessel (RPV) subjected to the thermal load and pressure during large break loss of coolant accident (LBLOCA). In this contribution, the multistep simulation scheme is extended to the analysis of two-phase steam-water flow occurring in the LBLOCA situation and the fracture analysis of postulated cracks under this accident event. It is demonstrated that the principle of the combined one-way coupled system code, CFD and structural mechanics codes can be used for the fracture analysis in case of a transient accident, which involves two-phase flow. The prediction of the temperature field is achieved by using two-phase computational fluid dynamics (CFD) simulation. To perform the stress analysis, an appropriate finite element discretization of the RPV wall is used, which considered the cladding and the ferritic low allow steel. The calculation of the stress intensity factor (SIF) in mode I for hypothetical cracks located in critical positions of the RPV is based on the linear elastic fracture mechanics (LEFM). The fracture mechanics model is based on the node-based submodeling technique in combination with XFEM in Abaqus. This is used to refine the mesh required by the fracture analysis in the region of interest, especially those regions under the cooling plume formed during the LBLOCA. The submodel may contain three types of cracks: axial, circumferential and inclined. The performed integrity assessment compares the stress intensity factor calculated in the deepest point of a surface crack with the material's fracture toughness. Two approaches to extract the stress intensity factor were applied in the present paper, namely, the classical FEM and eXtended FEM or XFEM. The classical FEM and XFEM solutions are compared. Both techniques give similar results. However, XFEM is more convenient as it is mesh independent. |
| Keyword | Pressurized thermal shock Large Break LOCA Stress intensity factor Reactor pressure vessel Fracture mechanics XFEM |
| DOI | 10.1016/j.nucengdes.2019.110237 |
| Indexed By | SCI ; EI |
| Language | 英语 |
| WOS ID | WOS:000487821600025 |
| WOS Keyword | STRESS INTENSITY FACTORS ; EXTENDED FINITE-ELEMENT ; HOLLOW CYLINDERS ; AXIAL CRACKS ; PTS ; SIMULATION ; GROWTH |
| WOS Research Area | Nuclear Science & Technology |
| WOS Subject | Nuclear Science & Technology |
| Funding Project | PROBAB project by the Swiss Federal Nuclear Safety Inspectorate (ENSI)[H-101247] |
| Funding Organization | PROBAB project by the Swiss Federal Nuclear Safety Inspectorate (ENSI) |
| Classification | 二类/Q1 |
| Ranking | 3 |
| Contributor | Mora, Diego F. |
| Citation statistics | |
| Document Type | 期刊论文 |
| Identifier | http://dspace.imech.ac.cn/handle/311007/80667 |
| Collection | 非线性力学国家重点实验室 |
| Affiliation | 1.Paul Scherrer Inst, Nucl Energy & Safety Dept, Struct Integr Grp, Forsch Str 111, CH-5232 Villigen, Switzerland; 2.Chinese Acad Sci, Inst Mech, State Key Lab Nonlinear Mech LNM, Beijing 100190, Peoples R China; 3.Paul Scherrer Inst, Nucl Energy & Safety Dept, Lab Sci Comp & Modelling, Forsch Str 111, CH-5232 Villigen, Switzerland |
| Recommended Citation GB/T 7714 | Mora DF,Niffenegger M,Qian GA,et al. Modelling of reactor pressure vessel subjected to pressurized thermal shock using 3D-XFEM[J]. NUCLEAR ENGINEERING AND DESIGN,2019,353:13.Rp_Au:Mora, Diego F. |
| APA | Mora DF,Niffenegger M,钱桂安,Jaros M,&Niceno B.(2019).Modelling of reactor pressure vessel subjected to pressurized thermal shock using 3D-XFEM.NUCLEAR ENGINEERING AND DESIGN,353,13. |
| MLA | Mora DF,et al."Modelling of reactor pressure vessel subjected to pressurized thermal shock using 3D-XFEM".NUCLEAR ENGINEERING AND DESIGN 353(2019):13. |
| Files in This Item: | Download All | |||||
| File Name/Size | DocType | Version | Access | License | ||
| Jp2019398.pdf（5171KB） | 期刊论文 | 出版稿 | 开放获取 | CC BY-NC-SA | View Download | |
Items in the repository are protected by copyright, with all rights reserved, unless otherwise indicated.
Edit Comment