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

Angelucci, M.; Paci, Sandro; Feria, F.; Herranz, L.E.

doi:10.3390/en15031216

Cited by 3 publications

(3 citation statements)

References 28 publications

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“…The entire system has been modelled with the MELCOR code (version 2.2.18019), that is originally developed for the simulation of Severe Accident (SA) scenarios. The model is described in [13,14]; however, few details will be reported in the following in relation to the performed sensitivity studies.…”

Section: Sensitivity Studiesmentioning

confidence: 99%

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

Angelucci,

Paci

2024

J. Phys.: Conf. Ser.

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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.

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Section: Sensitivity Studiesmentioning

confidence: 99%

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

Angelucci,

Paci

2024

J. Phys.: Conf. Ser.

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“…Following this path, a number of studies were performed, during the past decades, to promote Best-Estimate (BE) analyses rather than conservative ones. In particular, Best Estimate Plus Uncertainty (BEPU) methodologies and their application in different areas, from thermal-hydraulics to Severe Accidents, have been largely investigated [3][4][5][6][7][8][9][10][11]. Furthermore, another option beyond BEPU has been introduced in [12], where a realistic approach with BE computer codes, BE assumptions and BE initial and boundary conditions (with no request for an uncertainty assessment) is foreseen.…”

Section: Introductionmentioning

confidence: 99%

“…The improved model has been obtained exploiting the findings of the sensitivity studies carried out in the Part I [18] of this two-paper series. In particular, the three addressed factors (namely the axial power profile for the Fuel Assembly (FA), the loss coefficients in the air gap between the internal metallic canister and the external concrete-based overpack, and the thermal conductivity of the concrete itself) have been taken into account conjointly in this Part II.…”

Section: Introductionmentioning

confidence: 99%

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

Angelucci,

Paci

2024

J. Phys.: Conf. Ser.

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Nowadays, a great deal of attention is devoted to the development of best-estimate models able to produce more realistic outcomes. This is also the case for system codes, such as MELCOR, that are being mostly used in a conservative way especially when dealing with the licensing process. The above-mentioned need for more realistic results is at the core of this two-paper series related to the creation of a more accurate MELCOR model for the HI-STORM 100S dry cask. The findings obtained from the sensitivity studies carried out in the Part I are leveraged to set up an improved MELCOR model, the characteristics of which are consistent with the typical features of Spent Nuclear Fuel (SNF), and with geometrical and material properties of the cask itself. The addition of an axial power profile in the Fuel Assembly (FA), the better characterization of the flow losses in the air gap between internal metallic canister and external concrete-based overpack, and the choice of an appropriate value for the concrete thermal conductivity, are taken into account conjointly in this Part II. The outcomes from the improved MELCOR simulation are reported mainly in terms of the Peak Cladding Temperature (PCT), being the variable under regulatory surveillance. However, in addition to PCT, calculated temperature profiles are displayed and compared against the ones resulting from the previous model.

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On the behavior of a concrete-based dry cask for spent nuclear fuel in off-normal and accidental conditions

Angelucci,

Herranz,

Paci

2024

Nuclear Engineering and Design

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

Cited by 3 publications

References 28 publications

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

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

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

On the behavior of a concrete-based dry cask for spent nuclear fuel in off-normal and accidental conditions

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