MATRA-LMR-FB assessment with THORS bundle 2B experiments

Abstract

In this study, the ‘THORS (Thermal-Hydraulic Out-of-Reactor Safety) bundle 2B’ test data in ORNL (Oak Ridge National Laboratory, US) were used to supplement a qualification of the MATRA-LMR-FB code. The code was developed to analyze a sub-channel blockage accident in a Sodium cooled Fast Reactor (SFR), as an effort to enhance its reliability for an analysis of a blockage disturbance. The ‘THORS bundle 2B’ test was conducted to investigate the thermal-hydraulic effects of 24% and 45% sub-channel inlet blockages with a 19-pin bundle. The test covered several flow rates at the bundle inlet with different bundle powers. The MATRA-LMR-FB predictions were compared with not only the CFX simulation results but also the test data.As a result, most of the comparative results between the MATRA-LMR-FB predictions and the test data lay within a range of ±15°C. Such differences were not usually derailed much from other predictions found in a literature survey. The code, however, is slightly biased toward an under-prediction, with the most probable difference occurring at around −2 to −4°C. Nevertheless, it was anticipated that the comparison will supplement the applicability of the MATRA-LMR-FB to a partial flow blockage accident in a subassembly of an SFR. The CFX simulation results mostly agreed with the MATRA-LMR-FB predictions.