VALIDATION OF COOLANT THERMAL RESPONSE IN A TRANSIENT FINITE DIFFERENCE THERMAL TRANSPORT MODEL WITH APPLICATIONS TO FAST SPECTRUM REACTORS

Abstract

Transient behavior in nuclear reactors is important in accidents and with reactivity control systems that are driven by thermal feedback. Here, we describe a transient finite difference model for a pin cell system. The fidelity of the model is shown by validation against the thermocouple measurements of the CABRI BI1 experiment and the Safety Analysis System-Sodium Fast Reactor model of the experiment. In the BI1 experiment, a sodium-cooled mixed oxide fuel pin was subject to a loss of flow transient to coolant boiling within a sodium test loop positioned in the center of the CABRI research reactor. Comparisons to the initial steady-state coolant temperature profile, coolant temperature profile at twenty seconds into the transient, and at four axial locations within the coolant show agreement of the simple model with the experimental results better than or similar to those of the Safety Analysis System-Sodium Fast Reactor model. The model can be used to determine the thermal response times of coolant in fast reactors currently operating or in the design phase when subject to loss of flow accidents or other transients. Here, we investigate the difference in coolant thermal response for metal fueled and mixed oxide fueled sodium fast reactors when subject to transient overpower and loss of flow events. Additionally, we determine the effect of pin pitch on outlet coolant temperatures during the overpower event. Finally, we return to the CABRI experiment and show the importance of porosity in fuel temperature calculations.