PhD Thesis Defenses

Thermo-mechanical Assessment of Reactor Pressure Vessels of Light Water Reactors During Severe Accidents


The reactor pressure vessel (RPV) is one of the crucial safety barriers designed to isolate the reactor core, safeguarding against potential radioactive releases into the environment during a severe accident. The assessment of RPV behaviour and its failure is necessary to predict the characteristics of melt release into the reactor pit and succeeding ex-vessel accident progression.

This thesis aims to develop both the model and methodology in the Finite Element Analysis (FEA) of the RPV to predict its structural behaviour during postulated severe accidents. The improvement in the FE model starts from the material models for the material properties of the RPVs. Since the SA533B1 and 16MND5 carbon steels are considered the structural materials of the RPVs relevant to a Nordic Boiling Water Reactor (BWR) and a Pressurized Water Reactor (PWR), respectively, the material models for these two materials are established and subsequently validated against multiple tests. The simulations strictly adhere to the test conditions in the validation process. The observed agreement between the simulation and test results serves as a good foundation for subsequent analysis on the RPV applications.

A thermo-mechanical coupling approach is developed by coupling the ANSYS Mechanical APDL for the structural analysis of RPVs and MELCOR for defining boundary conditions. This approach can efficiently predict the RPV behaviour during accident scenarios, including deformation, stress and strain. Subsequently, the obtained results are subjected to a comprehensive failure analysis of RPVs with three failure criteria, namely melt-through, stress-based, and strain-based failure criteria. In addition, an advanced model in LS-DYNA is introduced to simulate the possible rupture phenomenon of RPVs during failure.

The developed model and methodology are applied in structural analysis of the Nordic BWR and the PWR during severe accidents. The analysis results contribute to:

(i)              A benchmark specification in the EU-IVMR project WP2.4 conducted to investigate the effect of the ablated profile on RPV failure in numerical analysis;

(ii)            The feasibility of In-Vessel Retention (IVR) strategy mitigation analyzed for the RPV of a Nordic BWR in two severe accidents: a Station Blackout (SBO) and an SBO combined with a Loss-of-coolant Accident (SBO+LOCA); and

(iii)          A comprehensive failure analysis for RPVs in a Nordic BWR carried out under the mentioned two severe accidents. The RPV lower plenum model is extended from the two-dimensional case for a standalone vessel wall to a three-dimensional case for a vessel wall with a cluster of IGT structures. This failure analysis aims to investigate the failure mechanism and timing for both vessel wall and IGTs, providing valuable insights into the possible earliest failure mode of RPV lower plenum for different reactor designs and severe accident scenarios.