Nuclear Thermal-Hydraulic Phenomena: Bases and Challenges

In the domain of reactor transient simulation, the identification of thermal-hydraulic phenomena (THPs) plays a major role. The system codes should model all influential THPs and should be validated against integral effect tests and separate effect tests which cover all influential THPs. The validation and the uncertainty quantification should cover every model related to an influential THP. A list of 116 THPs, recently established, covers all water-cooled reactors and design basis accident (DBA) analyses. It synthesizes more than 30 years of Organisation for Economic Co-operation and Development and International Atomic Energy Agency activities conducted by several safety specialists. A new tentative method to identify THPs was proposed based on two sources of information, the parameter evolutions in transients (depressurization, voiding, refill, heating, ...) and the set of balance equations with source and sink terms for convection, diffusion, interfacial transfers, and wall transfers. This method is just based on Gen-2 pressurized water reactor (PWR) transient analysis at the system, component, and basic process levels.

The comparison of the 116 THP list with the list of the tentative methods is made in this paper for the particular case of GEN-2 PWRs. No major contradiction was found. Both methods identified phenomena at the system, component, and process level. The 116 list better identified special components that require “special models” in system codes. The use of equations identified many more local process THPs, which may help ranking phenomena in a scaling analysis and identifying validation needs. The comparison confirms a potential synergy and complementarity between approach 1, which is based on the 116 THP list, and approach 2, which is based on the tentative method; the outcome of this comparison suggests further efforts to combine them and complement them in a new international collaborative context.

This analysis reports ongoing discussions between members of the Forum and Network of System Thermal Hydraulics Codes in Nuclear Reactor Thermal-Hydraulics network of system code developers.