Preliminary transient analysis of the Autonomous Reactivity Control system for fast reactors

IMPLEMENTATION OF AN AUTONOMOUS REACTIVITY CONTROL SYSTEM IN A SMALL LEAD-COOLED FAST REACTOR

This paper describes the design, implementation and characterisation of an Autonomous Reactivity Control (ARC) system in a small modular lead-cooled fast reactor. The aim of this work was to demonstrate the applicability of the ARC system and to study its dynamic behaviour during an anticipated transient without scram. A simplified one-dimensional model was developed to calculate the heat transfer within the ARC system, and the reactivity worth as a function of the neutron poison’s insertion into the active core was obtained via static neutronic calculations. By coupling the aforementioned models, the ARC’s time-dependent reactivity was derived as a function of the coolant outlet temperature variation. This model was implemented into the BELLA multi-point dynamics code and transient simulations were run. A control rod ejection accident was studied leading to an unprotected transient overpower scenario, in which 350 pcm reactivity was inserted during one second. It was shown that the ARC system provides a forceful negative reactivity feedback and that steady-state temperatures after the transient were reduced by almost 300 K compared to an identical transient without its action. In this scenario, the ARC system managed to stabilise the coolant outlet temperature at a value 100 K above nominal conditions. The implementation of an ARC system provided the reactor with a passively actuated self-regulating reactivity control system able to insert large amounts of negative reactivity in a short amount of time.

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 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.