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The Advanced Fuel Cycle Initiative Status of Neutronics Modeling

The Advanced Fuel Cycle Initiative Status of Neutronics Modeling. Won Sik Yang Argonne National Laboratory NEAMS Reactor Simulation Workshop May 19, 2009. Status of Neutronics Analyses. Within the current knowledge of physics, theory and governing equations are well known

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The Advanced Fuel Cycle Initiative Status of Neutronics Modeling

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  1. The Advanced Fuel Cycle InitiativeStatus ofNeutronics Modeling Won Sik Yang Argonne National Laboratory NEAMS Reactor Simulation Workshop May 19, 2009

  2. Status of Neutronics Analyses • Within the current knowledge of physics, theory and governing equations are well known • Boltzmann equation for neutron transport • Bateman equation for fuel composition evolution • The coefficients of these equations are determined by nuclear data, geometry, and composition • Nuclear data are for the most part relatively well known for the most commonly used nuclides • But still improved data are required to reduce design uncertainties • Geometry and composition have stochastic uncertainties and are affected by thermal, mechanical, irradiation, and chemical phenomena • These coupled phenomena are not as well described, and they can dominate the analysis errors • The challenge in neutronics analysis is to determine the solution efficiently by taking into account geometric complexity and complicated energy dependence of nuclear data NEAMS Reactor Simulation Workshop

  3. Reaction Rate Traverse Example • Monte Carlo simulation with MCNP5 (INL) • Reaction rate tally uncertainties < 1% • C/E values for U-235 fission rate distribution in CIRANO-2A (Blanket) and CIRANO-2B (Reflector) experiments NEAMS Reactor Simulation Workshop

  4. Negative Reactivity Transients of PHENIX • Four unexpected scrams occurred in 1989 - 1990 due to short negative reactivity transients (200 ms) with the same signal shape • Several potential explanations were given, but not satisfactory • Experiments are planned for PHENIX end-of-life tests for further investigation NEAMS Reactor Simulation Workshop

  5. Generation IV Target Uncertainties • Current and Target Uncertainties for sodium cooled fast reactors NEAMS Reactor Simulation Workshop

  6. Objectives and Requirements • The final objective is to produce an integrated, advanced neutronics code that allows the high fidelity description of a nuclear reactor and simplifies the multi-step design process • Integration with thermal-hydraulics and structural mechanics analyses to account for reactivity feedbacks due to geometry deformation accurately • Required modeling capabilities • Reactivity and power distribution (coupled neutron and gamma heating) • Non-equilibrium and equilibrium fuel cycle analyses • Refueling, fuel shuffling, and ex-core models • Perturbation and sensitivity analyses • Uncertainty analysis and optimization • Transient analysis (coupled with T/H and T/M analyses) • Reactivity coefficients and kinetics parameters • Shielding, decay heat, coolant activation and dose rate calculations, etc. NEAMS Reactor Simulation Workshop

  7. Selected Approaches • Utilize modern computing power and computational techniques • Meshing, domain decomposition strategies, parallel linear solvers, new visualization techniques, etc • Allow uninterrupted applicability to core design work • Phased approach for multi-group cross section generation • Simplified multi-step schemes • Online cross section generation • Adaptive flux solution options from homogenized assembly geometries to fully explicit heterogeneous geometries in serial and parallel environments • Allow the user to smoothly transition from the existing homogenization approaches to the explicit geometry approach • Rapid turn-around time for scoping design calculations • Detailed models for design refinement and benchmarking calculations NEAMS Reactor Simulation Workshop

  8. Adaptive Flux Solution Options Homogenized assembly Homogenized assembly internals Homogenized pin cells Fully explicit assembly • Unified geometrical framework • Unstructured finite element analysis for coupling with structural mechanics and thermal-hydraulics codes NEAMS Reactor Simulation Workshop

  9. Flux Solvers Available in UNIC • PN2ND • Second-order, even-parity transport equation (CG solve) • 1-D, 2-D, 3-D Cartesian with general reflected and vacuum b.c.s • Spherical harmonics combined with Serendipity and Lagrangian FE • SN2ND • Second-order even-parity transport equation (CG solve) • 2-D & 3-D Cartesian with general reflected and vacuum b.c.s • Discrete ordinates combined with Serendipity and Lagrangian FE • MOCFE • First-order transport equation (long characteristics) • 3-D Cartesian with general reflected and vacuum b.c.s • Discrete ordinates combined with Serendipity and Lagrangian FE • NODAL: hybrid finite element method for structured geometries • Will replace nodal diffusion and VARIANT options in DIF3D • Use as an multi-grid preconditioner for other solvers NEAMS Reactor Simulation Workshop

  10. Takeda Benchmark 4 NEAMS Reactor Simulation Workshop

  11. ABTR Whole-Core Calculations NEAMS Reactor Simulation Workshop

  12. ZPPR-15 Critical Experiments Computational Mesh and Example Flux Solutions of ZPPR-15 Critical Experiment NEAMS Reactor Simulation Workshop

  13. 2D OECD/NEA C5G7 Benchmark Thermal Group Flux in Core Thermal Group Flux in Pin Cell NEAMS Reactor Simulation Workshop

  14. Parallel Implementation • The scalability to peta-scale computing resources has been demonstrated • 163,840 cores of BlueGene/P (Argonne) • 131,072 cores of XT5 (ORNL) • Over 75% weak scalability Weak Scaling Study by Angle on BlueGene/P (PHENIX EOL test) NEAMS Reactor Simulation Workshop

  15. Parallel Implementation Weak Scaling Study by Angle on XT5 (PHENIX EOL test) NEAMS Reactor Simulation Workshop

  16. PHENIX End-of-Life Experiments • Participating in the PHENIX end-of-life experiments • Whole-core geometry is required (no symmetry) using homogenized fuel and explicit control rods • Space/angle convergence study completed using over 4 billion DOF on up to 163,840 cores of Blue Gene/P • Energy discretization study is ongoing 600 eV Flux and Radial Mesh NEAMS Reactor Simulation Workshop

  17. ZPR-6 Critical Experiments • Two ZPR-6 critical experiments are targeted for V&V in 2009 (Assemblies 6A and 7) • Explicit fuel plate representation allows direct comparison to legacy homogenization methods • Spatial mesh requirements are large; U-235 plates are 1/16th in thick • Preliminary studies performed on BG/P and Jaguar up to 130,000 processors indicate that over 10 billion DOF will be required to resolve the space-angle-energy mesh NEAMS Reactor Simulation Workshop

  18. ZPR-6 Critical Experiments 14 MeV Flux / Mesh U-235 Plate Power NEAMS Reactor Simulation Workshop

  19. Advanced Multi-group Cross Section Generation Code MC2-3 • A modular version has been integrated into UNIC for on-line generation of multi-group cross sections of each spatial region with given material and temperature distribution • Standalone code to generate ISOTXS datasets for legacy tools • Ultrafine group (2082 groups) transport calculations • Homogeneous mixture, and 1-D slab and cylindrical geometries • Resolved resonance self-shielding with numerical integration of point-wise cross sections using the narrow resonance (NR) approximation • Unresolved resonance self-shielding with the generalized resonance integral method • Elastic scattering transfer matrices obtained with numerical integration of isotopic scattering kernel in ENDF/B data NEAMS Reactor Simulation Workshop

  20. Advanced Multi-group Cross Section Generation Code MC2-3 • 1-D hyperfine group (~100,000) transport capability • Consistent P1 transport calculation for entire resolved resonance energy range (< ~1 MeV) with anisotropic scattering sources • Optionally used for accurate resolved resonance self-shielding and scattering transfer matrix generation • Efficient strategy to generate accurate multi-group cross sections for heterogeneous assembly or full-core calculations is being developed by combining various solution options • 1-D hyperfine group cell calculation • 1-D ultrafine group whole-core calculation (with homogenized regions) • 2-D MOCFE calculation in several hundred groups NEAMS Reactor Simulation Workshop

  21. MC2-3.0 and Coupling with UNIC NEAMS Reactor Simulation Workshop

  22. Reconstructed Pointwise Cross Sections (ENDF/B-VII.0) NEAMS Reactor Simulation Workshop

  23. Hyper-Fine-Group Spectrum Calculation • Inner core composition of ZPR-6/6A NEAMS Reactor Simulation Workshop

  24. Hyper-Fine-Group vs. Ultra-Fine-Group Spectra NEAMS Reactor Simulation Workshop

  25. LANL Criticality Assembly Benchmarks (UFG Calculation) • Multiplication factors are in an excellent agreement within 0.15% ∆ρ by taking into account the anisotropy of inelastic scattering NEAMS Reactor Simulation Workshop

  26. MC2-3 vs. VIM for ZPR-6/7 (Standalone UFG Calculation) NEAMS Reactor Simulation Workshop

  27. ZPPR-15 Critical Experiments Reflector Blanket Outer core Inner core NEAMS Reactor Simulation Workshop

  28. A Realistic View of ZPPR-15 Double Fuel Column Drawer Matrix tube Drawer Void X DEPLETED URANIUM STAINLESS STEEL DEPLETED URANIUM PU-U-MO FUEL Z SODIUM SODIUM SODIUM SODIUM SODIUM DEPLETED URANIUM STEEL BLOCK STAINLESS STEEL DEPLETED URANIUM STAINLESS STEEL STAINLESS STEEL DEPLETED URANIUM SODIUM SODIUM SODIUM SODIUM SODIUM PU-U-MO FUEL DEPLETED URANIUM STAINLESS STEEL DEPLETED URANIUM NEAMS Reactor Simulation Workshop

  29. ZPPR-15 Critical Experiments • Three loading configurations of ZPPR-15 Phase A were analyzed • Loading 15: initial criticality • Loading 16: reference configuration for sodium void worth measurement • Loading 20: configuration with an 18” sodium void in part of inner core NEAMS Reactor Simulation Workshop

  30. Summary • An initial version of new multi-group cross section generation code MC2-3 has been developed • Preliminary tests showed significantly improved performance relative to MC2-2 • Integrated with UNIC for online cross section generation • Consistent thermal feedbacks • Account for spectral transition effects • Second order solvers PN2ND and SN2ND have been improved • SN2ND demonstrated good scalability to >100,000 processors • Working on enhancing the anisotropic scattering iteration • Fixing the load imbalance for reflected boundary conditions • Starting next phase of pre-conditioner development • p-refinement multi-grid and • Algebraic multi-grid beyond that or possibly h-refinement NEAMS Reactor Simulation Workshop

  31. Summary • First order solver MOCFE • Improving parallel performance with Krylov Method • Added more elements to ray tracing capabilities • Adding back projection for parallel • Started NODAL • Implement Krylov solution technique to fix some convergence problems • Eliminate memory problems and 1970s architecture • Will investigate energy parallelization on multi-core machines (8-32 cores) NEAMS Reactor Simulation Workshop

  32. Backup Slides NEAMS Reactor Simulation Workshop

  33. Perturbation Evaluation with MCNP (LANL) • The MCNP perturbation option was used to determine the difference in net neutron production in every fuel assembly as a resulting of reducing • Fuel density by 2%, cladding density by 5%, and coolant density by 50% • While the fuel density reduction showed reasonable results, the clad and coolant density effects still showed significant statistical variations • Observed statistical errors are less than 2% for the fuel density perturbation • However, as large as 41% for the cladding density perturbation and 100% for the coolant density perturbation • Direct perturbation calculations showed even worse results • Relative statistical uncertainties of the re-converged production rates are often above 50%, and in some cases reach 100% • The re-converged calculation ran 50,000 histories per cycle for 160 active cycles, each of which took 1000 minutes on a 2.7-GHz Opteron processor NEAMS Reactor Simulation Workshop

  34. NGNP with 60-degree periodic symmetry Core multiplication factor converges relatively quickly Power distribution converges very slowly Convergence of Assembly Power Distribution • Asymmetric assembly power distribution is observed • Extremely large number of histories would be required for converged pin power distribution NEAMS Reactor Simulation Workshop 34

  35. Comparison of whole core depletions performed by GA, BNL, and ANL MONTEBURNS (MCNP5+ORIGEN2) Simple cubic lattice model CPU time: ~40 hours for 50K and ~100 hours for 100K histories Much larger number of histories are required for converged flux solutions Depletion with Monte Carlo Method • DB-MHR benchmark • Cycle length = 540 EFPD • Total 7 cycles • 6 burn steps per cycle (90 days interval) • 50K and 100K neutron histories per burn step • Note that there are ~3 billion fuel particles NEAMS Reactor Simulation Workshop 35

  36. APPLO2:172-group CP and 28-group MOC calculation CRONOS2: 8-group diffusion calculation (finite element method) CEA: NEPHTIS Verification Results Homogenous Element % difference in fission rate distributions from MCNP4C (3D core) Heterogeneous Element NEAMS Reactor Simulation Workshop 36

  37. Power Distribution of Fuel Block (CR Inserted) NEAMS Reactor Simulation Workshop 37

  38. Effective Multiplication Factors for 2D and 3D VHTRs with Heterogeneous Fuel Compact All Rods In (ARI) All Rods Out (ARO) Operating Rods In (ORI) NEAMS Reactor Simulation Workshop 38

  39. 2D Power Distributions ORI ARO ARI NEAMS Reactor Simulation Workshop 39

  40. 2D Block Power Comparison with MCNP5 ORI ARO ARI NEAMS Reactor Simulation Workshop 40

  41. 3D Flux Distribution for All Rods Out (ARO) Case 7 eV 1 MeV 1 eV 0.13 eV NEAMS Reactor Simulation Workshop 41

  42. 3D Flux Distribution for Operating Control Rods In (ORI) 7 eV 1 MeV 1 eV 0.13 eV NEAMS Reactor Simulation Workshop 42

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