Thermal analysis of HI-STORM 100S dry cask with the MELCOR code

As spent nuclear fuel (SNF) pools are reaching their capacity, dry storage systems have been introduced worldwide as interim step before permanent deep underground disposal. The thermal behavior of the SNF, and of the system containing it, has proven to be essential to ensure fuel rod integrity for the entire storage time. In the effort to achieve a proper characterization, this work investigates the capabilities of the lumped parameter MELCOR 2.2 code to model the thermal evolution of SNF in the scenario of a concrete-based dry cask. Being dry storage systems completely outside MELCOR domain of application, a number of approximations and hypotheses have been necessary, without however jeopardizing the real cask characteristics and behaviour. Results have been discussed to be coherent with what physically expected and verified against estimates previously obtained from a FLUENT calculation on the same system. The code shows the ability to well predict the Peak Cladding Temperature (PCT) and its location, as well as temperature radial profile; however, some discrepancies show up when analysing temperature profiles in the down-comer and in the air gap. For this reason, further assessment studies should be performed to confirm the potential shown in this preliminary work.