Towards a more realistic MELCOR model for a dry cask for spent nuclear fuel. Part I: sensitivity studies

The United States (US) Department Of Energy (DOE) has addressed the thermal analysis of the Spent Nuclear Fuel (SNF) stored within a dry cask system as a matter of high priority. In this regard, it is of utmost importance that simulation tools effectively reproduce the general thermal behavior of the modelled cask, including heat exchange and removal. Temperature distribution in the different components of the system is usually the focus of performed thermal analyses. In particular, attention is paid to the maximum temperature reached in the fuel cladding, namely the Peak Cladding Temperature (PCT). Within this framework, the present paper is the first of a two-paper series aimed at developing a more accurate model for the HI-STORM 100S cask. The dry cask in question is modelled and its behavior is simulated by means of the MELCOR code (version 2.2.18019). Stressing the need for a more realistic model rather than a conservative one, this paper reports the efforts undertaken to evaluate the influence of some specific modelling choices on the PCT. The study of the cask performance is therefore conducted taking into consideration three main factors: the axial power distribution in the Fuel Assembly (FA), the flow losses in the air gap between the internal canister and the external overpack, and the conductivity of the overpack concrete.