neams-mcnp - The University of...
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Form 836 (7/06) LA-UR- Approved for public release; distribution is unlimited. Los Alamos National Laboratory, an affirmative action/equal opportunity employer, is operated by the Los Alamos National Security, LLC for the National Nuclear Security Administration of the U.S. Department of Energy under contract DE-AC52-06NA25396. By acceptance of this article, the publisher recognizes that the U.S. Government retains a nonexclusive, royalty-free license to publish or reproduce the published form of this contribution, or to allow others to do so, for U.S. Government purposes. Los Alamos National Laboratory requests that the publisher identify this article as work performed under the auspices of the U.S. Department of Energy. Los Alamos National Laboratory strongly supports academic freedom and a researcher’s right to publish; as an institution, however, the Laboratory does not endorse the viewpoint of a publication or guarantee its technical correctness. Title: Author(s): Intended for: 09-03055 MCNP Monte Carlo & Advanced Reactor Simulations Forrest Brown NEAMS Reactor Simulation Workshop ANL, 19 May 2009
Transcript of neams-mcnp - The University of...
Form 836 (7/06)
LA-UR- Approved for public release; distribution is unlimited.
Los Alamos National Laboratory, an affirmative action/equal opportunity employer, is operated by the Los Alamos National Security, LLC for the National Nuclear Security Administration of the U.S. Department of Energy under contract DE-AC52-06NA25396. By acceptance of this article, the publisher recognizes that the U.S. Government retains a nonexclusive, royalty-free license to publish or reproduce the published form of this contribution, or to allow others to do so, for U.S. Government purposes. Los Alamos National Laboratory requests that the publisher identify this article as work performed under the auspices of the U.S. Department of Energy. Los Alamos National Laboratory strongly supports academic freedom and a researcher’s right to publish; as an institution, however, the Laboratory does not endorse the viewpoint of a publication or guarantee its technical correctness.
Title:
Author(s):
Intended for:
09-03055
MCNP Monte Carlo & Advanced Reactor Simulations
Forrest Brown
NEAMS Reactor Simulation WorkshopANL, 19 May 2009
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Monte Carlo CodesX-3-MCC, LANL
NEAMS Reactor Simulation WorkshopArgonne National LaboratoryMay 19, 2009
MCNP Monte Carlo&
Advanced Reactor Simulations
Forrest B. Brown [email protected]
Los Alamos National Laboratory, Los Alamos, NM, USA
LA-UR-09-03055
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Monte Carlo CodesX-3-MCC, LANLAbstract
MCNP Monte Carlo & Advanced Reactor SImulations
Forrest B. Brown (LANL)
The MCNP Monte Carlo code is widely used in studies of advanced reactorconcepts, either directly as a main-line design tool or indirectly as part of theverification/validation process. MCNP is routinely used to calculate k-effective anddetailed distributions of power and reaction rates. MCNP provides highly accurateresults, using continuous-energy physics, ENDF/B-VII nuclear data, and explicit 3Dconstructive solid geometry. There are over 10,000 MCNP users world-wide.
MCNP has many special features for criticality calculations of reactors, and hasbeen coupled to burnup and thermal/hydraulic codes for multi-physicsapplications. MCNP users have requested a number of enhancements for modelingadvanced reactor concepts. These requests are directly aligned with proposedNEAMS activities.
LA-UR-09-03055
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Monte Carlo CodesX-3-MCC, LANLMCNP Monte Carlo for Reactor Applications
• MCNP Monte Carlo strengths– General & accurate 3D constructive solid geometry– Direct use of best cross-section data (ENDF/B-VII)– Continuous-energy neutron/photon transport & physics– Runs everywhere
• Windows / Mac / Linux / Unix• netbook / laptop / office / cluster / terascale• cluster & multicore parallel computing (MPI+threads)
– Examples on next few slides …..
• Over 10,000 MCNP users world-wide– Extensive documentation and V&V
• MCNP is used in nearly every study of advanced reactor concepts– Directly as main-line neutronics package, or– Indirectly for V&V of approximate neutronics– MCNP is used in several multi-physics efforts, with burnup & CFD
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Monte Carlo CodesX-3-MCC, LANLExamples - Reactor Analysis with MCNP
MITresearch reactor
Pictures frommcnp plotter
ATR PWR(1/4 of geometry)
VHTR with TRISO fuel
• Accurate & explicit modeling at multiple levels• Accurate continuous-energy physics & data
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Monte Carlo CodesX-3-MCC, LANL
3D geometry
Example - TRIGA reactor model
Fast Flux Thermal Flux
DiffusionTheoryCodes
MCNP5Analysis
Radial Power DensityFrom MCNP5 Analysis (from Luka Snoj, Jozef Stefan Inst.)
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Monte Carlo CodesX-3-MCC, LANLExample - PWR model
Whole-core Thermal & Total Flux from MCNP5 Analysis
Assembly Thermal & Fast Flux from MCNP5 Analysis
(from Luka Snoj, Jozef Stefan Inst.)
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Monte Carlo CodesX-3-MCC, LANLExamples - Multi-physics with MCNP
• From multi-physics R&D at University of Michigan
• Also, MCNP5 coupled to RELAP5-3D/ATHENA for 3D models of VHTR with burnup & T/H feedback
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Power Density in an inner fuel cell
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Monte Carlo CodesX-3-MCC, LANLMCNP Features for Reactor Analysis
• Criticality calculations - K-effective & alpha• Coupled neutron-photon calculations, with photo-neutrons
• Plots of geometry, cross-sections, tallies, & convergence data• 3D constructive solid geometry, with quadric surfaces & hierarchical embedding
• Continuous-energy physics & data• S( , ) thermal scattering treatment
• Option for multigroup data
• Utilities for adjusting cross-section temperatures• Perturbation theory, using differential operator approach
• Validation suite for criticality, based on ICSBEP benchmark experiments
• Analytic verification suite for criticality, based on exact solutions
• Stochastic geometry option, for modeling random TRISO fuel
• Shannon entropy to assess convergence of power distribution
• Mesh tallies, superimposed on 3D geometry, for 3D power maps, etc.
• Adjoint-weighted reaction tallies [new & unique]
• Adjoint-weighted reactor kinetics parameters [new & unique]
• Dominance ratio calculations [new & unique]
• Wielandt acceleration [new & unique]
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Monte Carlo CodesX-3-MCC, LANLRequested Features for Reactor Analysis
MCNP users have asked for improvements for reactor calculations:
• Include resonance scattering effects into free-gas scattering model
• Depletion calculations with fission products are limited to ~15K depletable regions.Users would like to do 100K+ depletable regions.
• For multiphysics coupling, on-the-fly Doppler broadening would permit nearly continuousvariation in region temperatures.
• Additional features for sensitivity/uncertainty analyses of xsec data
• Remove the limit of 99,999 materials &/or cells
• Coupling to CAD geometry, with robust tracking through gaps/overlaps
• Mesh geometry embedded in 3D models
• And more ……
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Monte Carlo CodesX-3-MCC, LANLReality
• DOE Nuclear Criticality Safety Program is the only MCNP5 support– $0 from DP, $0 from NA– The only near-term enhancements will be those requested by NCSP
Note: DP is providing some support for MCNP6 development (including merger of high-energyfeatures from MCNPX), but release date is uncertain. Features for reactor analysis have low/zeropriority.
• Need MCNP5 support from NEAMS– Either directly, or
– As collaborator/partner on other proposals that use MCNP
– Would help NEAMS users with both design & V&V efforts
– NEAMS needs advanced features in MCNP for reactor analysis
– Frameworks (eg, UNIC) should be designed to accommodate multiple plug-in neutronics modules (eg, diffusion, Sn, MCNP, etc.)
• Fast, approximate modules for routine use
• MC module for hard problems & V&V
