Validation of a thermo-fluid-structure coupling approach for RPV creep failure analysis against FOREVER-EC2 experiment
- 1. Division of Nuclear Power Safety, Royal Institute of Technology (KTH), Roslagstullsbacken 21, 106 91 Stockholm (Sweden)
Description
The failure of reactor pressure vessel (RPV) during a severe accident of light water reactors is a thermal fluid-structure interaction (FSI) problem which involves melt pool heat transfer and creep deformation of the RPV. The present study is intended to explore a reliable coupling approach of thermo-fluid-structure analyses which will not only be able to reflect the transient thermal FSI feature, but also apply the advanced models and computational platforms to melt pool convection and structural mechanics, so as to improve simulation fidelity. For this purpose, the multi-physics platform of ANSYS encompassing Fluent and Structural capabilities was employed to simulate the fluid dynamics and structural mechanics in a coupled manner. In particular, the FOREVER-EC2 experiment was chosen to validate the coupling approach. The natural convection in melt pool was modeled with the SST turbulence model with a well-resolved boundary layer, while the creep deformation for the vessel made of 16MND5 steel was analyzed with a new three-stage creep model (modified theta projection model). A utility tool was introduced to transfer the transient thermal loads from Fluent to Structural which minimizes the user effort in performing the coupled analysis. The validation work demonstrated the well-posed capability of the coupling approach for prediction of the key parameters of interest, including temperature profile, total displacement of vessel bottom point and the evolution of wall thickness profile in the experiment.
Availability note (English)
Available from http://dx.doi.org/10.1016/j.anucene.2019.06.067Additional details
Identifiers
- DOI
- 10.1016/j.anucene.2019.06.067;
- PII
- S0306454919303858;
Publishing Information
- Journal Title
- Annals of Nuclear Energy (Oxford)
- Journal Volume
- 133
- Journal Page Range
- p. 637-648
- ISSN
- 0306-4549
- CODEN
- ANENDJ
INIS
- Country of Publication
- United Kingdom
- Country of Input or Organization
- International Atomic Energy Agency (IAEA)
- INIS RN
- 51008017
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Numerical Data
- Descriptors DEI
- BOUNDARY LAYERS; COMPUTERIZED SIMULATION; CREEP; DEFORMATION; EXPERIMENTAL DATA; FAILURES; FLUID MECHANICS; FLUID-STRUCTURE INTERACTIONS; MELTING; NATURAL CONVECTION; PRESSURE VESSELS; SEVERE ACCIDENTS; STEELS; TURBULENCE; WATER COOLED REACTORS; WATER MODERATED REACTORS
- Descriptors DEC
- ACCIDENTS; ALLOYS; BEYOND-DESIGN-BASIS ACCIDENTS; CARBON ADDITIONS; CONTAINERS; CONVECTION; DATA; ENERGY TRANSFER; HEAT TRANSFER; INFORMATION; IRON ALLOYS; IRON BASE ALLOYS; LAYERS; MASS TRANSFER; MECHANICAL PROPERTIES; MECHANICS; NUMERICAL DATA; PHASE TRANSFORMATIONS; REACTORS; SIMULATION; TRANSITION ELEMENT ALLOYS
Optional Information
- Notes
- © 2019 Elsevier Ltd. All rights reserved.