The Effect of Backfill Gas Pressure on the Thermal Response of a Dry Cask for Spent Nuclear Fuel

Michela Angelucci (), Salvatore A. Cancemi, Rosa Lo Frano () and Sandro Paci
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Michela Angelucci: Department of Industrial and Civil Engineering (DICI), University of Pisa, Largo Lucio Lazzarino, 56122 Pisa, Italy
Salvatore A. Cancemi: Department of Industrial and Civil Engineering (DICI), University of Pisa, Largo Lucio Lazzarino, 56122 Pisa, Italy
Rosa Lo Frano: Department of Industrial and Civil Engineering (DICI), University of Pisa, Largo Lucio Lazzarino, 56122 Pisa, Italy
Sandro Paci: Department of Industrial and Civil Engineering (DICI), University of Pisa, Largo Lucio Lazzarino, 56122 Pisa, Italy

Energies, 2025, vol. 18, issue 2, 1-13

Abstract: Dry systems are being employed worldwide as interim storage for Spent Nuclear Fuel (SNF). Despite not being designed as permanent repositories, the safe storage of SNF must still be ensured. In this framework, few experimental campaigns have been conducted in the past. However, their limited number has led to the necessity to exploit numerical simulations for the thermal characterization of the system. Since the majority of the degradation mechanisms are temperature-dependent, conducting a thermal analysis of a dry cask is essential to assess the integrity of the system itself, and of the SNF stored within it. In this regard, both heat production and heat removal mechanisms have to be taken into account. On this basis, the present paper addresses the variation in the system heat removal capacity when considering different backfill gas pressures. In particular, the analysis, carried out with the MELCOR code, investigates the thermal response of the ventilated, concrete-based HI-STORM 100S cask, currently employed for spent fuel elements of Light Water Reactors (LWRs), when imposing different initial pressures for the helium backfill gas. Results are reported primarily in terms of maximum temperature of the fuel cladding, which is the variable under regulatory surveillance. In addition, the adherence to the maximum design pressure for the canister is verified by evaluating the helium pressure as the steady state is reached. The analysis seems to suggest that the safe operation of the HI-STORM 100S cask is guaranteed only for a limited range of the initial helium pressure.

Keywords: dry cask; spent nuclear fuel; MELCOR; modeling; thermal analysis; safety (search for similar items in EconPapers)
JEL-codes: Q Q0 Q4 Q40 Q41 Q42 Q43 Q47 Q48 Q49 (search for similar items in EconPapers)
Date: 2025
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