Thermal-hydraulic system evaluation of a natural circulation SMR using accident tolerant fuel

Abstract

A large focus of research in nuclear engineering is on improving the safety of light water reactors (LWRs). This research is driven by high-level nuclear accidents such as the events at Fukushima in March of 2011, among others. The two major factors which contributed to the severity of this accident were the failure of active cooling systems and oxidation of the cladding material which led to hydrogen explosions. Passive safety systems and new reactor concepts are considered to improve passive safety of LWRs. Furthermore, the use of accident tolerant fuel (ATF) is considered to reduce oxidation within the core. A passive LWR will be selected with ATF claddings being applied to the system to evaluate the thermal-hydraulic effects, with the goal of improving safety. Natural circulation small modular reactors (SMRs) are new reactor concepts designed around passive safety. The naturally circulating primary coolant is only driven by buoyancy forces which improves the passive safety of the reactor for both steady state operation and accident transients. One such reactor called the NuScale SMR is chosen for the current study. A model of the NuScale SMR was developed using ASYST 3.4, a modified RELAP5/MOD3 code. Specifications of the NuScale SMR was taken from the NuScale Final Safety Analysis Report (FSAR) made available by the U.S. Nuclear Regulatory Commission (NRC), as well as previous studies reported in literature. The model was verified for steady state operation and one worst-case scenario accident transient. The two ATF cladding materials selected for this study was FeCrAl alloy and a SiC composite. Both materials offer improvements in oxidation resistance and failure temperature. FeCrAl has favourable thermophysical properties while having a large neutron absorption cross section which negatively impacts the neutron economy. On the other hand, SiC has a small neutron absorption cross section and a very high melting temperature while having less favourable thermophysical properties. Results showed that the use of ATF cladding materials in the NuScale SMR model had little effect on the steady state operation of the reactor. Small changes were seen in the peak cladding temperature (PCT) and peak fuel centreline temperature (PFCT) when using FeCrAl and SiC. The worst-case scenario accident transient involves the inadvertent opening of one reactor vent valve (RVV) with subsequent failure of the emergency core cooling system (ECCS) and decay heat removal system (DHRS). In the current model this led to failure of the Zr-alloy cladding after 6.42 h. This grace time was increased with the use of ATF cladding, with FeCrAl and SiC providing an increase of 0.5 h and 4.35 h, respectively.