Modeling Validation Exercises Using the Dry Cask Simulator
- Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
- US Nuclear Regulatory Commission (NRC), Washington, DC (United States)
- Pacific Northwest National Lab. (PNNL), Richland, WA (United States)
- Research Centre for Energy, Environment and Technology (CIEMAT), Madrid (Spain)
- Empresa Nacional del Uranio, S.A., S.M.E. (ENUSA), Madrid (Spain)
- Univ. Politecnica de Madrid (UPM), Madrid (Spain)
The U.S. Department of Energy (DOE) established a need to understand the thermal-hydraulic properties of dry storage systems for commercial spent nuclear fuel (SNF) in response to a shift towards the storage of high-burnup (HBU) fuel (> 45 gigawatt days per metric ton of uranium, or GWd/MTU). This shift raises concerns regarding cladding integrity, which faces increased risk at the higher temperatures within spent fuel assemblies present within HBU fuel compared to low-burnup fuel (≤ 45 GWd/MTU). The dry cask simulator (DCS) was previously built at Sandia National Laboratories (SNL) in Albuquerque, New Mexico to produce validation-quality data that can be used to test the validity of the modeling used to determine cladding temperatures in modern vertical dry casks. These temperatures are critical to evaluating cladding integrity throughout the storage cycle of commercial spent nuclear fuel. In this study, a model validation exercise was carried out using the data obtained from dry cask simulator testing in the vertical, aboveground configuration. Five modeling institutions – Nuclear Regulatory Commission (NRC), Pacific Northwest National Laboratory (PNNL), Centro de Investigaciones Energéticas, MedioAmbientales y Tecnológicas (CIEMAT), and Empresa Nacional del Uranio, S.A., S.M.E. (ENUSA) in collaboration with Universidad Politécnica de Madrid (UPM) – were granted access to the input parameters from SAND2017-13058R, “Materials and Dimensional Reference Handbook for the Boiling Water Reactor Dry Cask Simulator”, and results from the vertical aboveground BWR dry cask simulator tests reported in NUREG/CR-7250, “Thermal-Hydraulic Experiments Using A Dry Cask Simulator”. With this information, each institution was tasked to calculate minimum, average, and maximum fuel axial temperature profiles for the fuel region as well as the axial temperature profiles of the DCS structures. Transverse temperature profiles and air mass flow rates within the dry cask simulator were also calculated. These calculations were done using modeling codes (ANSYS FLUENT, STARCCM+, or COBRA-SFS), each with their own unique combination of modeling assumptions and boundary conditions. For this validation study, four test cases of the vertical, aboveground dry cask simulator were considered, defined by two independent variables – either 0.5 kW or 5 kW fuel assembly decay heat, and either 100 kPa or 800 kPa internal helium pressure. For the results in this report, each model was assigned a model number. Three of the models used porous media model representations of the fuel, two models used explicit fuel representations, and one model used an explicit subchannel representation of the fuel. Even numbers were assigned to explicit fuel models and odd numbers were assigned to porous media models. The plots provided in Chapter 3 of this report show the axial and transverse temperature profiles obtained from the dry cask simulator experiments in the aboveground configuration and the corresponding models used to describe the thermal-hydraulic behavior of this system. The tables provided in Chapter 3 illustrate the closeness of fit of the model data to the experiment data through root mean square (RMS) calculations of the error in peak cladding temperatures (PCTs), average fuel temperatures across six axial levels, transverse temperatures across the PCT locations for the four test cases, and air mass flow rates. The peak cladding temperature is typically the most important target variable for cask performance, and all models capture the PCT within 5% RMS error. Two models show comparable fits to experimental results when considering the combined RMS error of all target variables. Since one uses a porous media representation of the fuel while the other uses an explicit fuel representation, it can be concluded that the porous media fuel representation can achieve modeling calculation results of peak cladding temperatures, average fuel temperatures, transverse temperatures, and air mass flow rates that are comparable to explicit fuel representation modeling results.
- Research Organization:
- Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
- Sponsoring Organization:
- USDOE Office of Nuclear Energy (NE)
- DOE Contract Number:
- AC04-94AL85000; NA0003525
- OSTI ID:
- 1601637
- Report Number(s):
- SAND-2019-6079R; 676217; TRN: US2102851
- Country of Publication:
- United States
- Language:
- English
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