Abstract
Nuclear safety relies to a good extent on thoroughly validated codes. However, code predictions are affected by uncertainties that need to be quantified for a more accurate evaluation of safety margins. In this regard, the present paper proposes a preliminary uncertainty and sensitivity analysis of the thermal behavior of a concrete-based dry cask for spent nuclear fuel storage, employing the MELCOR code and a series of MATLAB scripts. As thermal behavior is of utmost importance for the fulfillment of United States Nuclear Regulatory Commission (USNRC) safety requirements, the Peak Cladding Temperature (PCT) has been addressed as the key Figure of Merit (FOM). Variables related to the main heat transfer mechanisms have been selected as input parameters for the uncertainty quantification, whereas heat source and heat sink, namely decay power and external air temperature, have been dealt with in a separate sensitivity analysis. The results show that the selected parameters have a weak influence on the PCT, whereas it is strongly related to the decay power and external air temperature values. In any case, PCT stays below the regulatory threshold even under the considered off-normal conditions.
Highlights
The management of spent nuclear fuel (SNF) is recognized as the focal point of the back end of the nuclear fuel cycle
As peak cladding temperature (PCT) is the Figure of Merit (FOM) selected for this work, it axial location, the PCT is in axial level 11, in the upper half of the fuel, which is consistent is the focus of the results reporting
As for reality the cask would reach the steady state from a peak temperature attained during the axial location, the PCT is in axial level 11, in the upper half of the fuel, which is consistent so called “drying process,” during which the MPC goes through time intervals with no with the nearly flat profile of Decay Heat (DCH) axial distribution and the progressive heat-up of helium helium supporting the convective heat removal from fuel rods [48]
Summary
The management of spent nuclear fuel (SNF) is recognized as the focal point of the back end of the nuclear fuel cycle. Very few experimental tests have been performed: one experimental campaign was conducted by EPRI in 1987 for pressurized water reactor (PWR) spent fuel assemblies [11], while more recently a dry cask simulator (DCS) was built in Sandia. A great effort has been devoted to the thermal analysis dry casks have been employed in several works with computational fluid dynamic (CFD) codes [14,20,21,22,23,24,25]. Sensitivity analysis (UASA) based on a MELCOR model of the HI-STORM 100S dry In this framework, the present paper reports the results of a preliminary uncertainty cask [30]. A reasonable accuracy and a significant reduction of the computational cost make the MELCOR code [31,32] suitable for performing UQ in an easier and faster way
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