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Reactors designed by Argonne National Laboratory

Software

SAM: A Modern System Analysis Tool for Advanced Nuclear Reactors

 

The System Analysis Module (SAM) is a modern system analysis tool being developed at Argonne National Laboratory for advanced non-LWR safety analysis.  It aims to provide fast-running, whole-plant transient analyses capability with improved-fidelity for SFR, LFR, and MSR/FHR. SAM takes advantage of advances in physical modeling, numerical methods, and software engineering, to enhance its user experience and usability. It utilizes an object-oriented application framework (MOOSE), and its underlying meshing and finite-element library (libMesh) and linear and non-linear solvers (PETSc), to leverage the modern advanced software environments and numerical methods.

System Thermal-Fluids Modeling

SAM is being developed as a system-level modeling and simulation tool with higher fidelity but yet computationally efficient. As a new code development, the initial effort has been focused on the modeling and simulation capabilities of the heat transfer and single-phase fluid dynamics responses in reactor systems. The transient simulation capabilities of typical reactor accidents have been demonstrated in the transient simulations of the Advanced Burner Test Reactor and validated against the EBR-II benchmark test results. The key features include:

  • Robust and high-order FEM model of single-phase fluid flow and heat transfer;
  • Component-based system modeling;
  • Flexible coupling between fluid and solid components enables a wide range of engineering applications;
  • Enhanced built-in closure models and flexible modeling of fluid properties, friction, and convective heat transfer.
Temperature Distributions in the Simulation of Advanced Burner Test Reactor (ABTR)
Temperature Distributions in the Simulation of Advanced Burner Test Reactor (ABTR)

Reduced-Order Multi-Dimensional Flow Model

Computationally efficient multi-dimensional flow model is under development for thermal mixing and stratification phenomena in large enclosures for safety analysis. An advanced and efficient thermal mixing and stratification modeling capability embedded in a system analysis code is very desirable to improve the accuracy of reactor safety analyses and to reduce modeling uncertainties.

Natural convection flow in a square cavity
Natural convection flow in a square cavity

Flexible Core Modeling

A pseudo 3-D full-core conjugate heat transfer modeling capability has been developed in SAM for efficient and accurate temperature predictions of SFR structures. A multi-channel rod bundle model is developed to account for the temperature differences between the center region and the edge region of the coolant channel in a fuel assembly. The hexagon lattice core can be modeled with automatically-generated 1-D parallel channels representing the subassembly flow, and 2-D duct walls and inter-assembly gaps.

Comparison of average radial wall temperature distributions between SAM and CFD in a 7-assembly demonstration problem
Comparison of average radial wall temperature distributions between SAM and CFD in a 7-assembly demonstration problem

Multi-Physics Multi-Scale Integration

Flexible coupling interfaces have been developed to allow for convenient integration with other advanced or conventional simulation tools for multi-scale and multi-physics modeling capabilities.

Snapshots of Temperature Distribution in the coupled SAM-CFD simulations of ABTR Protected Loss-of-Flow Transient
Snapshots of Temperature Distribution in the coupled SAM-CFD simulations of ABTR Protected Loss-of-Flow Transient

Relevant Publications

  • Rui Hu, "A fully-implicit high-order system thermal-hydraulics model for advanced non-LWR safety analyses," Annals of Nuclear Energy, Vol. 101, 174–181, 2017.
  • T.H. Fanning and R. Hu, “Coupling the System Analysis Module with SAS4A/SASSYS-1,” ANL/NE-16/22, Argonne National Laboratory, September 2016.
  • R. Hu and Y. Yu, “A Computationally Efficient Method for Full-Core Conjugate Heat Transfer Modeling of Sodium Fast Reactors,” Nuclear Engineering and Design, Vol. 308, 182-193, 2016.
  • R. Hu and T. Sumner, “Benchmark Simulations of the Thermal-Hydraulic Responses during EBR-II Inherent Safety Tests using SAM”, Proceedings of ICAPP’16, San Francisco, CA, April 17-20, 2016.
  • R. Hu, “An Advanced One-Dimensional Finite Element Model for Incompressible Thermally Expandable Flow”, Nuclear Technology, Vol. 190, No. 3, 313-322, 2015.
  • R. Hu, J. W. Thomas, E. Munkhzul, T. H. Fanning, “Coupled System and CFD Code Simulation of Thermal Stratification in SFR Protected Loss-Of-Flow Transients,” Proceedings of ICAPP 2014, Charlotte, NC, April 6-9, 2014.

Last Modified: Wed, May 31, 2017 2:24 PM

 

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Rui Hu
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