This theme aims to assess the thermal hydraulic safety analysis codes by validating their predictions against experimental data obtained from integral effect test facilities such as the ATLAS facility at KAERI, in order to identify and validate the dominant thermal hydraulic phenomena occurring in NPPs under accident scenarios, including rare events. The main drive for selecting these topics is that thermal hydraulic studies on, beyond design based accident for the advance pressurized water reactor APR1400, is one of the immediate necessities that the NPP program needs to develop in order to guarantee safe operation of these power plants.
Project: Benchmarking and Verification & Validation (V&V) of Lumped Parameter and Computational Fluid Dynamics (CFD) codes
Principal Investigator: Dr. Ho Joon Yoon
All countries, in which nuclear power plants (NPPs) are being operated including the UAE, are required to maintain technical competence in thermal-hydraulics for nuclear reactor safety evaluations. These assessments are usually carried out using system-scale safety analysis codes. It is evident that before embarking on the NPP licensing practice using these codes, a “must do” validation process against experimental data is required. This will be achieved through:
- Using the experimental database, generated from integral effect test facilities, to validate predictive capability and accuracy of computer codes and models.
- Carry out an assessment of codes currently in use for thermal-hydraulic safety analyses in order to improve the local technical competence in thermal-hydraulics for nuclear reactor safety (NRS) evaluations.
Project: Codes Development and Validation for the Predictions of Rare Events
Principal Investigator: Dr. Yacine Addad
Despite all accident prevention measures adopted in nuclear power plants (NPPs), some accident scenarios with very low probabilities may occur, as proven by the Fukushima-Daiichi’s accident, resulting in severe accidents (SA) with core melting and plant damage. Under such condition, the core is overheated and a melt pool might form in the lower plenum of the reactor vessel. Therefore, a good understandings of the associated failure mechanisms effects on the lower head vessel and a thermal load to the outer reactor vessel under the IVR-ERVC condition (In-Vessel corium Retention through External Reactor Vessel Cooling) are necessary. This will be achieved through:
- Examine the ability of existing numerical approaches to correctly predict the chronological phenomena occurring during corium/steel interactions.
- Improve exiting numerical algorithms to make CFD approach more computational cost effective, hence achieving a shorter and more affordable physical simulation time.
