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

Abstract

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. Thermal behaviour 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.

Uncertainty and Sensitivity Analysis of a Dry Cask for Spent Nuclear Fuel

“…As PCT is the FOM selected for this work, it is the focus of the results reporting. Additional thermal results might be found in [30].…”

“…In this framework, the present paper reports the results of a preliminary uncertainty and sensitivity analysis (UASA) based on a MELCOR model of the HI-STORM 100S dry cask [25], [30]. The analysis focuses on the PCT, being the variable under regulatory surveillance, and on the uncertainties embedded in the values predicted by the code.…”

“…The model for the HI-STORM 100S, better described in [30], is briefly summarized in this section. The MELCOR 2.2 code, which is a fully integrated computer code to model accident progression in LWRs, has been employed.…”

Towards a more realistic MELCOR model for a dry cask for spent nuclear fuel. Part II: application.

“…Starting from the model presented in [14], an improved MELCOR model for the HI-STORM 100S system is implemented. Considering the outcomes of the sensitivity studies in Part I and having in mind the creation of a model as consistent to the reality as possible, modelling choices were made as presented below:…”

“…Considering what said above, the present paper (that is the second of a two-paper series) pursues the need for more realistic results, when evaluating the thermal performance of a dry storage system for Spent Nuclear Fuel (SNF). More precisely, an improved model for the HI-STORM 100S system [13] is set up starting from a previously developed model [14], created employing the MELCOR code [15,16] and verified against a Computational Fluid Dynamic (CFD) model of the same system [17].…”