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

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