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  1. Spatial decomposition of structured grids for nuclear reactor simulations

    Spatial decomposition methods based on graph partitioning are developed and implemented in the high fidelity neutron transport code MPACT. These graph-based spatial decomposition methods are more general than previous decomposition methods and typically provide better load balance and reduced runtimes due to their improved parallel efficiency. Correlations are drawn between simulation runtime and the balance of the partition for 2D simulations. Comparisons are made using partition balance metrics for different decomposition schemes in 2D and 3D simulations. Finally, for typical ranges of subdomains, graph-based partitioning methods offer significant reductions to runtimes. However, for highly decomposed problems, these graph-based methods maymore » decrease convergence rates, thus reducing parallel efficiency compared to older methods.« less
  2. Investigation of rotating mode behavior in BWR out-of-phase limit cycle oscillations – Part 1: Reduced order model

    Previous neutronic/thermal-hydraulic (TH) coupled numerical simulations using full-core TRACE/PARCS and SIMULATE-3K BWR models have shown evidence of a specific “rotating mode” behavior (steady rotation of the symmetry line, i.e. constant phase shift of approximately 90° between the first two azimuthal modes) in out-of-phase limit cycle oscillations, regardless of initial conditions and even if the first two azimuthal modes have different natural frequencies. This suggests a nonlinear coupling between these modes; otherwise, the phase shift between these modes would change at a constant rate during the limit cycle. The goal of the present work is to gain further insights on themore » rotating mode behavior using a simplified mathematical model which contains all of the important physics for this application while providing sufficient flexibility and simplicity to allow for in-depth understanding of the underlying phenomena. This was accomplished using a multi-channel, multi-modal reduced-order model, using a modification of the fixed pressure drop boundary condition to simulate channel coupling via the inlet and outlet plena, in order to destabilize the out-of-phase mode over the in-phase mode. Examination of the time-dependent solution of the nonlinear system showed a clear preference for rotating mode behavior in the four-channel model under stand-alone TH conditions and for conditions with weak neutronic feedback. Furthermore, when neutronic feedback was strengthened (i.e., larger reactivity feedback coefficients), the side-to-side mode (stationary symmetry line) was favored instead. Additional analyses using higher-fidelity numerical modeling, as well as a physical explanation for the rotating behavior seen in both sets of analyses, will be provided in a companion paper (“Part 2”).« less
  3. Investigation of rotating mode behavior in BWR out-of-phase limit cycle oscillations – Part 2: TRACE/PARCS model and physical explanation

    Previous neutronic/thermal-hydraulic (TH) coupled numerical simulations using full-core TRACE/PARCS and SIMULATE-3K boiling water reactor (BWR) models have shown evidence of a specific “rotating mode” behavior (steady rotation of the symmetry line, i.e. constant phase shift of approximately 90° between the first two azimuthal modes) in BWR out-of-phase limit cycle oscillations, regardless of initial conditions and even if the first two azimuthal modes have different natural frequencies. This suggests a nonlinear coupling between these modes; otherwise, the phase shift between these modes would change at a constant rate during the limit cycle. The previous paper (“Part 1”) presented a series ofmore » results to examine this rotating behavior with a reduced-order model. The goal of the present study is to provide additional analyses of the predicted rotating mode behavior using higher-fidelity numerical modeling, as well as a physical explanation for why this mode is favored over side-to-side or other oscillatory behaviors from a TH perspective. Results are presented using TRACE and TRACE/PARCS for a small number of parallel channels, which confirmed that the conclusions developed from the reduced-order model remain applicable when applying a full two-fluid, six-equation, finite-volume modeling approach. From these results, a physical explanation has been put forth to explain why the rotating symmetry line behavior is preferred from a TH standpoint, demonstrating that predominantly out-of-phase unstable systems are most unstable when the variation in the total inlet flow rate is minimized (which minimizes the effective single-phase to two-phase pressure drop ratio) and that the rotating mode is the most successful in minimizing this total flow rate variation as compared with the side-to-side case or any other oscillation pattern. The conclusion is that the rotating mode will be favored for any out-of-phase unstable system of parallel channels with no neutronic feedback or relatively weak neutronic feedback. Here, previous analyses have indicated that systems with sufficiently strong neutronic coupling may favor the side-to-side oscillation mode over the rotating mode; this topic is left as a subject of future investigation.« less
  4. A probabilistic model-based diagnostic framework for nuclear engineering systems

  5. A 2-D/1-D transverse leakage approximation based on azimuthal, Fourier moments

    Here, the MPACT code being developed collaboratively by Oak Ridge National Laboratory and the University of Michigan is the primary deterministic neutron transport solver within the Virtual Environment for Reactor Applications Core Simulator (VERA-CS). In MPACT, the two-dimensional (2-D)/one-dimensional (1-D) scheme is the most commonly used method for solving neutron transport-based three-dimensional nuclear reactor core physics problems. Several axial solvers in this scheme assume isotropic transverse leakages, but work with the axial SN solver has extended these leakages to include both polar and azimuthal dependence. However, explicit angular representation can be burdensome for run-time and memory requirements. The work heremore » alleviates this burden by assuming that the azimuthal dependence of the angular flux and transverse leakages are represented by a Fourier series expansion. At the heart of this is a new axial SN solver that takes in a Fourier expanded radial transverse leakage and generates the angular fluxes used to construct the axial transverse leakages used in the 2-D-Method of Characteristics calculations.« less
  6. A simulation of I2S-LWR selected transients

  7. VERA Core Simulator methodology for pressurized water reactor cycle depletion

    This paper describes the methodology developed and implemented in the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) to perform high-fidelity, pressurized water reactor (PWR), multicycle, core physics calculations. Depletion of the core with pin-resolved power and nuclide detail is a significant advance in the state of the art for reactor analysis, providing the level of detail necessary to address the problems of the U.S. Department of Energy Nuclear Reactor Simulation Hub, the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS has three main components: the neutronics solver MPACT, the thermal-hydraulic (T-H) solver COBRA-TF (CTF), and the nuclidemore » transmutation solver ORIGEN. This paper focuses on MPACT and provides an overview of the resonance self-shielding methods, macroscopic-cross-section calculation, two-dimensional/one-dimensional (2-D/1-D) transport, nuclide depletion, T-H feedback, and other supporting methods representing a minimal set of the capabilities needed to simulate high-fidelity models of a commercial nuclear reactor. Results are presented from the simulation of a model of the first cycle of Watts Bar Unit 1. The simulation is within 16 parts per million boron (ppmB) reactivity for all state points compared to cycle measurements, with an average reactivity bias of <5 ppmB for the entire cycle. Comparisons to cycle 1 flux map data are also provided, and the average 2-D root-mean-square (rms) error during cycle 1 is 1.07%. To demonstrate the multicycle capability, a state point at beginning of cycle (BOC) 2 was also simulated and compared to plant data. The comparison of the cycle 2 BOC state has a reactivity difference of +3 ppmB from measurement, and the 2-D rms of the comparison in the flux maps is 1.77%. Lastly, these results provide confidence in VERA-CS’s capability to perform high-fidelity calculations for practical PWR reactor problems.« less
  8. Stability and accuracy of 3D neutron transport simulations using the 2D/1D method in MPACT

    We derived a consistent “2D/1D” neutron transport method from the 3D Boltzmann transport equation, to calculate fuel-pin-resolved neutron fluxes for realistic full-core Pressurized Water Reactor (PWR) problems. The 2D/1D method employs the Method of Characteristics to discretize the radial variables and a lower order transport solution to discretize the axial variable. Our paper describes the theory of the 2D/1D method and its implementation in the MPACT code, which has become the whole-core deterministic neutron transport solver for the Consortium for Advanced Simulations of Light Water Reactors (CASL) core simulator VERA-CS. We also performed several applications on both leadership-class and industry-classmore » computing clusters. Results are presented for whole-core solutions of the Watts Bar Nuclear Power Station Unit 1 and compared to both continuous-energy Monte Carlo results and plant data.« less
  9. The Modeling of Advanced BWR Fuel Designs with the NRC Fuel Depletion Codes PARCS/PATHS

    The PATHS (PARCS Advanced Thermal Hydraulic Solver) code was developed at the University of Michigan in support of U.S. Nuclear Regulatory Commission research to solve the steady-state, two-phase, thermal-hydraulic equations for a boiling water reactor (BWR) and to provide thermal-hydraulic feedback for BWR depletion calculations with the neutronics code PARCS (Purdue Advanced Reactor Core Simulator). The simplified solution methodology, including a three-equation drift flux formulation and an optimized iteration scheme, yields very fast run times in comparison to conventional thermal-hydraulic systems codes used in the industry, while still retaining sufficient accuracy for applications such as BWR depletion calculations. Lastly, themore » capability to model advanced BWR fuel designs with part-length fuel rods and heterogeneous axial channel flow geometry has been implemented in PATHS, and the code has been validated against previously benchmarked advanced core simulators as well as BWR plant and experimental data. We describe the modifications to the codes and the results of the validation in this paper.« less
  10. Implementation of the transient fixed-source problem in the neutron transport code PROTEUS-MOC

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