United States Nuclear Regulatory Commission - Protecting People and the Environment

Validation of Computational Fluid Dynamics Methods Using Prototypic Light Water Reactor Spent Fuel Assembly Thermal-Hydraulic Data (NUREG-2208)

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Publication Information

Manuscript Completed: November 2016
Date Published: March 2017

Prepared by:
G. Zigh
S. Gonzalez

Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

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Abstract

Applicants submit spent nuclear fuel dry storage cask designs to the U.S. Nuclear Regulatory Commission (NRC) for certification under Title 10 of the Code of Federal Regulations (10 CFR) Part 72, "Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste." The NRC staff performs its technical review of these designs in accordance with 10 CFR Part 72 and NUREG-1536, "Standard Review Plan for Spent Fuel Dry Storage Systems at a General License Facility," Revision 1, issued July 2010. To ensure that the cask and fuel material temperatures of the dry cask storage system remain within the allowable limits or criteria for normal, off-normal, and accident conditions, the NRC staff performs a thermal review as part of the technical review.

Recent applications increasingly have used thermal-hydraulic analyses using computational fluid dynamics (CFD) codes (e.g., ANSYS FLUENT) to demonstrate the adequacy of the thermal design. The NRC Office of Nuclear Material Safety and Safeguards asked the NRC Office of Nuclear Regulatory Research to perform validation studies of the FLUENT CFD code to assist it in making regulatory decisions to ensure adequate protection for storage and transportation casks. The validation studies were based on preignition data (separate effect tests) obtained from the following previous studies:

  • NUREG/CR-7143, "Characterization of Thermal-Hydraulic and Ignition Phenomena in Prototypic, Full-Length Boiling Water Reactor Spent Fuel Pool Assemblies After a Postulates Complete Loss-of Coolant Accident," issued March 2013
  • NUREG/CR-7215, "Spent Fuel Pool Project Phase I: Pre-Ignition and Ignition Testing of a Single Commercial 17x17 Pressurized Water Reactor Spent Fuel Assembly under Complete Loss of Coolant Accident Conditions," issued April 2016
  • NUREG/CR-7216, "Spent Fuel Pool Project Phase II: Pre-Ignition and Ignition Testing of a 1x4 Commercial 17x17 Pressurized Water Reactor Spent Fuel Assemblies under Complete Loss of Coolant Accident Conditions," issued April 2016

The fuel assembly experimental data from these studies provided reliable information for various fuel assembly heat loads that can be used to validate the analytical methods. Combined with current methods to determine modeling and application uncertainty, the measured data offered additional confirmation on the adequacy of the applied analytical methods.

The research summarized in this report relates to the CFD validation studies performed for a single test assembly of a full-length commercial 17x17 pressurized-water reactor (PWR) fuel bundle; a 17x17 PWR 1x4 configuration where the center fuel assembly was electrically heated and the four surrounding assemblies were unheated; and, a single test assembly for a full-length commercial 9x9 boiling-water reactor fuel bundle. For each of these configurations, a detailed and porous media model were developed to validate the results based on the experimental data and conducted parametric studies to assess model sensitivity. The grid conversion index method published by the American Society of Mechanical Engineers (ASME), "Standard for Verification and Validation in Computational Fluid Dynamics and Heat Transfer," was used to calculate the discretization uncertainty of the model.

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Page Last Reviewed/Updated Wednesday, March 08, 2017