Description of a Dry Cask Simulator for Measuring Internal and External Thermal-Hydraulic Performance
- Sandia National Laboratories Albuquerque, NM 87185-0747 (United States)
- Nuclear Regulatory Commission (United States)
The thermal performance of a commercial nuclear spent fuel dry storage cask is evaluated through detailed analytical modeling. These modeling efforts are performed by the vendor to demonstrate the performance as well as regulatory compliance and are independently verified by the Nuclear Regulatory Commission (NRC). The majority of commercial dry storage casks in use today are aboveground. Cooling of the assemblies located inside the sealed canister is achieved by air flowing in the bottom of the cask and exiting out the top. Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating design and performance models. Numerous studies have been previously conducted. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. These vertical, canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and belowground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of aboveground and belowground canistered dry cask systems. Thus the enhanced performance of modern dry storage casks cannot be fully validated using previous studies. The purpose of the investigation described in this report is to produce data sets for the validation of assumptions associated with the thermal-hydraulic simulation of modern dry cask storage systems. Of particular interest are the steady-state cladding temperatures and the effect of elevated helium pressures in sealed canisters in both above and belowground designs. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly will be deployed inside of a representative storage basket and cylindrical pressure vessel that will represent the canister. The arrangement of a single assembly with well-controlled boundary conditions should simplify interpretation of results. Various arrangements of outer concentric ducting will be used to mimic conditions for above and belowground storage configurations of vertical, dry cask systems with canisters. Radial and axial temperature profiles will be measured for a wide range of decay power and helium cask pressures. As of this writing, the apparatus is currently under construction at Sandia National Laboratories in Albuquerque, NM and is expected to be operational starting in April 2016. Pre-test CFD calculations have been performed using best available information. These calculations are being used to assist in test design but will need to be validated and potentially reevaluated. The planned testing will be adjusted to reflect as-built conditions and measured system limitations. Dimensional analyses indicate that the anticipated ranges of relevant dimensionless groups (Reynolds, Modified Rayleigh, and Nusselt numbers) bracket or closely approach prototypic values for both the aboveground and belowground configurations. While designed to match prototypic values, the expected test matrix will include values that exceed currently acceptable values for decay heat, internal helium pressure, and peak cladding temperatures in order to gain more insight into the underlying behavior of the system. (authors)
- OSTI ID:
- 22992029
- Journal Information:
- Transactions of the American Nuclear Society, Vol. 114, Issue 1; Conference: Annual Meeting of the American Nuclear Society, New Orleans, LA (United States), 12-16 Jun 2016; Other Information: Country of input: France; 8 refs.; Available from American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 United States; ISSN 0003-018X
- Country of Publication:
- United States
- Language:
- English
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