MCNP Monte Carlo & Advanced Reactor Simulations. Forrest Brown. NEAMS Reactor Simulation Workshop ANL, 19 May Title: Author(s): Intended for:

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1 LA-UR Approved for public release; distribution is unlimited. Title: MCNP Monte Carlo & Advanced Reactor Simulations Author(s): Forrest Brown Intended for: NEAMS Reactor Simulation Workshop ANL, 19 May 2009 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-AC52-06NA 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. Form 836 (7/06)

2 NEAMS Reactor Simulation Workshop Argonne National Laboratory May 19, 2009 LA-UR MCNP Monte Carlo & Advanced Reactor Simulations Forrest B. Brown Los Alamos National Laboratory, Los Alamos, NM, USA 1

3 Abstract LA-UR MCNP Monte Carlo & Advanced Reactor SImulations Forrest B. Brown (LANL) The MCNP Monte Carlo code is widely used in studies of advanced reactor concepts, either directly as a main-line design tool or indirectly as part of the verification/validation process. MCNP is routinely used to calculate k-effective and detailed distributions of power and reaction rates. MCNP provides highly accurate results, using continuous-energy physics, ENDF/B-VII nuclear data, and explicit 3D constructive solid geometry. There are over 10,000 MCNP users world-wide. MCNP has many special features for criticality calculations of reactors, and has been coupled to burnup and thermal/hydraulic codes for multi-physics applications. MCNP users have requested a number of enhancements for modeling advanced reactor concepts. These requests are directly aligned with proposed NEAMS activities. 2

4 MCNP 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 3

5 Examples - Reactor Analysis with MCNP MIT research reactor ATR PWR (1/4 of geometry) VHTR with TRISO fuel Pictures from mcnp plotter Accurate & explicit modeling at multiple levels Accurate continuous-energy physics & data 4

6 Example - TRIGA reactor model 3D geometry Fast Flux Thermal Flux Diffusion Theory Codes MCNP5 Analysis Radial Power Density From MCNP5 Analysis (from Luka Snoj, Jozef Stefan Inst.) 5

7 Example - PWR model Whole-core Thermal & Total Flux from MCNP5 Analysis Assembly Thermal & Fast Flux from MCNP5 Analysis (from Luka Snoj, Jozef Stefan Inst.) 6

8 Examples - Multi-physics with MCNP From multi-physics R&D at University of Michigan McStar DeCART/STAR Norm. Power Cell Number (top to bottom) Power Density in an inner fuel cell McSTAR DeCART/STAR power (w/cc) MCNP STAR-CD plane # (bottom to top) Also, MCNP5 coupled to RELAP5-3D/ATHENA for 3D models of VHTR with burnup & T/H feedback 7

9 MCNP 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] 8

10 Requested 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 continuous variation 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 9

11 Reality 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-energy features from MCNPX), but release date is uncertain. Features for reactor analysis have low/zero priority. 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 plugin neutronics modules (eg, diffusion, Sn, MCNP, etc.) Fast, approximate modules for routine use MC module for hard problems & V&V 10