Advancing Thermal-Hydraulic Models and Predictive Tools for Design of SCWR Prototypes

There is a strong interest in further developing Super-Critical Water-cooled Reactor (SCWR) designs among Member States in thus supporting climate change mitigation beyond the current fleet of water cooled reactors. Several conceptual designs of 1000 MWe SCWR have been developed in Canada, China, European Union, Japan and the Russian Federation. In addition, the development process has been extended to super-critical small modular reactors for generating capacity of up to 350 MWe, which can also be adopted as prototypes for the reference SCWR design.
The design process requires advances in key technology areas such as neutronics, fuel, materials, chemistry, thermal-hydraulics, control and safety. Thermal-hydraulics is one of the critical areas in maintaining coolability of the fuel and hence reliable operation of the reference and prototype SCWRs. A significant amount of thermal-hydraulics studies was performed to support design, licensing and operation of the current fleet of nuclear reactors operating at subcritical pressures. However, information and experimental data remain scarce on thermal-hydraulics behaviours of fluids in supercritical regimes.
The objective of this CRP is to establish a coherent body of knowledge about fluids at supercritical pressures and temperatures needed in prototyping the SCWR conceptual designs. This is an ambitious goal, which can be achieved only with a renewed effort in research and development (R&D) that identifies and closes gaps in technology areas as well as enhances knowledge and technological bases relevant to design options. Achieving this goal would also enhance the knowledge basis in similarity to the level of availability in water cooled reactor technologies at subcritical pressures.
The scope of the CRP is to advance predictive tools (such as correlations, system and sub-channel codes) and models supported by computational fluid dynamics tools, based on improved knowledge and understanding of thermal-hydraulics phenomena for the design of SCWR prototypes. Predictive capability of these tools will be assessed against experimental data or benchmarked against other tools. Additionally, these tools could also be applied for licensing and operation of the current fleet of nuclear reactors.
This CRP will bring together experts from Member States with water cooled reactor technologies and those with interests in SCWR technologies. The associated research activities will foster national excellence and international cooperation, promote sharing of newly developed knowledge, and contribute to capacity building in developing countries. Participating in this CRP would enhance the R&D capability and infrastructure (such as system and subchannel codes), as well as facilitate training of highly qualified personnel for the nuclear industry in both developed and developing Member States.