Sodium cooled nuclear reactors are being considered for the next generation of nuclear power plants to provide clean electricity to the world. To prove the safety of these reactors, the fuel must be shown to safely handle large neutron fluxes that cause a spike in the thermal output of the fuel. To prove that the fuel is capable of withstanding these events, transient experiments are performed to show the response of the fuel to these events. This data is also used to verify and validate software packages to be used in the design of the nuclear reactors. In this work a computational fluid dynamics benchmark analysis is performed to predict the thermal response of the HOP 1-6A experiment performed at TREAT. HOP 1-6A is an historic experiment performed at TREAT to determine the safety of wire-wrapped nuclear fuel to be used in the Fast Flux Test Facility. The goal of the benchmark exercise is to show that the software and methodology used are appropriate tools to simulate sodium cooled experiments similar to the one analyzed. When possible, this work also aims to provide insight into improvements into future designs of sodium cooled experiments.