The fuel rod analysis program SCANAIR has been developed by IRSN (Institut de Radioprotection et de Sûreté Nucléaire) for analysis of reactivity initiated accidents (RIA) in light water reactors. The Swedish Radiation Safety Authority (SSM) has access to SCANAIR in exchange for annual contributions for its further development. This ensures a possibility for SSM to do own analysis of fuel rods in reactivity initiated accidents. The development and administration of the program is done by Quantum Technologies AB on assignment from SSM.
SSM’s development of SCANAIR is primarily focused on the thermohydraulic models, with the aim of improving the analytical capabilities for fuel in boiler water reactors. In a previous work, an improved thermohydraulic model in SCANAIR was proposed and implemented in an SSM-specifc version of SCANAIR V_7_6. The present work is the 2020 contribution to SCANAIR development and contains a validation against tests in the Power Burst Facility (PBF).
This current project has shown that the two-phase coolant channel module developed by Quantum Technologies AB is able to model cladding-to-coolant heat transfer under RIAs. The work includes more in depth analysis of coolant properties and fuel rod gas pressure, from which it is concluded that the pellet-cladding gap conductance has a surprisingly strong efect. The work has also included a parametric study that investigate the efect of modelling aspects of the flm boiling heat transfer.
Knowledge of what is happening in a fuel rod during an event and how it is implemented in analytical tools is essential to SSM for the supervision of nuclear power plants. The participation in the development of SCANAIR also enables SSM to actively be a part of the large eforts that are made internationally with testing, understanding and improving the tools for analysis of reactivity initiated accidents.
Need for further research
Continued work on developing SCANAIR’s analysis capabilities is planned in cooperation with IRSN. This current work suggests that a deeper understanding of axial gas transport and internal gas pressure equalisation is motivated. On the longer time scale there will be new tests in the CABRI reactor with thermohydraulic conditions closer to those of current nuclear power plants. These tests will be valuable to validate the improved thermohydraulic model against and to continue model development from.