Validation of the fast reactor plant dynamics analysis code Super-COPD using FFTF loss of flow without scram test #13

Abstract

The Japan Atomic Energy Agency (JAEA) has been developing a plant dynamics analysis code named Super-COPD for design studies and safety analyses of sodium-cooled fast reactors. Since there are a few validation studies of a thermal–hydraulic model in a plant, including a point kinetics model considering reactivity feedback for anticipated transient without scram (ATWS) events classified into the beyond design basis events (BDBEs), further validation studies under BDBE conditions are needed to improve the prediction accuracy of the model. Therefore, JAEA has participated in the international benchmark analysis of loss of flow without scram (LOFWOS) test #13 performed at the Fast Flux Test Facility (FFTF) as one of passive safety demonstration tests hosted by the International Atomic Energy Agency. In the first step, the transient calculation with major reactivity feedback mechanisms and imposing measured power history was conducted with Super-COPD. The results were compared with the measured data. Several challenges were identified about the thermal–hydraulic model in the core and heat transport systems and the reactivity model in the point kinetics model. Consequently, the transient calculation was performed with modified models considering the radial heat transfer, flow redistribution, and interwrapper flow in the core. The time-dependent change and the second peak of the outlet temperature in the specified assembly were predicted accurately because the core thermal–hydraulics was reproduced well. Furthermore, the assembly bowing reactivity model was developed to improve the net reactivity, thereby resulting in a good agreement with the measured data on the reactor power, temperature, and flow rate in the primary loop. Thus, this paper proves the validation of Super-COPD for the LOFWOS event in the BDBE.