An Overview of What’s New in SCALE 5 S. M. Bowman, D. F. Hollenbach, M. D. DeHart, B. T. Rearden, I. C. Gauld, and S. Goluoglu Oak Ridge National Laboratory American Nuclear Society 2002 Winter Meeting

New Modules in SCALE 5  CENTRM: Continuous energy flux solution  NITAWL-III: Compatible with ENDF/B-VI  TSUNAMI: Sensitivity/uncertainty  NEWT: 2-D flexible mesh  STARBUCS: Burnup credit sequence  SMORES: 1-D material optimization  So many codes, so little time…

CENTRM/PMC  CENTRM (Continuous Energy Transport Module)  1-D discrete ordinates code  Problem-dependent pointwise continuous energy flux spectra at discrete spatial intervals for each unit cell  Processes all resolved resonances in a mixture together  PMC (Pointwise Multigroup Converter)  Collapses pointwise continuous energy cross-sections for each nuclide into a set of problem dependent multigroup cross sections  Separate CENTRM/PMC input files are created for each unit cell + one for all mixtures not in a unit cell

CENTRM/PMC (Cont.)  Eliminate many of the limitations inherent in the Nordheim Integral Treatment used by NITAWL  Overlapping resonances  Multiple fissile materials in unit cell  Assumed flux profile  Process discrete level inelastic cross-section data  Explicitly model rings in a fuel pin for spatial effect on the flux and cross sections

CENTRM/PMC (Cont.)  Problem-dependent multigroup cross sections with accuracy of continuous energy cross sections  ENDF/B-V continuous energy cross-section data files for CENTRM in SCALE 5  Correspond to ENDF/B-V 238- and 44 ‑ group libraries  ENDF/B-VI continuous energy data for CENTRM under development for later release

SCALE 5 Criticality Sequence with CENTRM

NITAWL-III  Can process multi-pole data, compatible with ENDF/B-VI cross-section data  ENDF/B-VI multigroup library under development  Processes cross-section data in the resolved resonance range for each nuclide individually  Still limited to one fuel mixture per unit cell

Sensitivity/Uncertainty Codes  Use adjoint-based first order linear perturbation theory to calculate sensitivities and propagate uncertainties  Operate as automated SCALE analysis sequences  k eff sensitivities to group-wise cross-section data are automatically generated for every reaction/nuclide/region (sensitivity profile)  Group-wise sensitivity coefficients are written to data file for further analysis and plotting  Other responses besides k eff can be added

TSUNAMI ( Tools for Sensitivity/UNcertainty Analysis Methodology Implementation)  1-D deterministic transport (XSDRNPM)  3-D Monte Carlo transport (KENO V.a)  Produce sensitivity coefficients that represent the % change in k eff per % change in cross- section data  Based on multigroup perturbation theory  Accounts for effect of perturbations in resonance processing of cross-section data

Sensitivity Profiles Can Be Plotted to Highlight Similarities/Differences

Benefits of S/U Methodology  Improved understanding of physics, identify parameters and regions of importance  Validation: Establish similarity of experiments to system of interest  Provides estimate of bias and uncertainty, including basis for interpolation and extrapolation beyond experiment range  Identify experimental needs  Optimize experiment design to best reduce bias and uncertainty in validation

NEWT Flexible Mesh S n  NEW Transport algorithm  2-D discrete ordinates neutron transport code with flexible mesh capabilities  Provides spatial and angular rigor necessary for advanced LWR fuel and MOX fuel designs  Simple input concept based on SCALE user interface  Grid generation is automated  Generalized geometry capabilities, not limited to lattice configurations

PWR 17x17 Lattice =newt Calvert Cliffs fuel assembly (one-fourth) read parm fillmix=5 prtmxsec=no prtbroad=no sn=6 inners=10 outers=200 end parm read materials ! 3.0 enriched fuel, pin location 1 ! end end ! water (background material) ! end end materials read geom ' Fuel rod subgrid cylinder !fuel! end cylinder !clad! end ' Water hole subgrid cylinder !water hole! end cylinder !guide tube! end array domain boundary end geom end

Other Models  Simple pin cell  VENUS-2 MOX benchmark (1/4 core)

NEWT thermal spectra plots BWR w/ Gd rodsMOX Supercell

STARBUCS Features  STARBUCS: Standardized Analysis of Reactivity for Burnup Credit using SCALE  Integrated depletion analysis, cross-section processing and Monte Carlo criticality safety calculations for a 3-D system  Uses existing, well-established modules in the SCALE system  STARBUCS creates input, executes codes, and performs all data transfer functions

STARBUCS Features (cont.)  Depletion analysis methodology  Uses the ORIGEN-ARP sequence  Cross-sections for depletion in ORIGEN-S obtained by interpolation of an existing ARP library  Interpolation on enrichment, burnup, moderator density  The analysis is extremely fast and accurate  Criticality safety analysis  KENO V.a or KENO VI  Multigroup, 3-D analysis capability

STARBUCS Features (cont.)  Flexible, easy-to-use sequence, uses input similar to existing SCALE modules  Standard composition data used to define all materials in the problem (depletion and non-fuel)  Depletion analysis input based on SAS2H-like input formats  Any existing KENO V.a or KENO-VI input file (e.g., fresh fuel) can be used directly, with only minor changes

STARBUCS Features (cont.)  Designed to simulate many of the important burnup credit phenomena identified in ISG-8, e.g.,  Axial and horizontal burnup variations  Analyses can be performed for nuclide subgroups, i.e., evaluation of fission product margin  Isotopic correction factors may be applied  Sequence designed for, but is not restricted to, analysis of spent fuel casks  Automatic loading curve generation under development

Data Flow in a BUC Analysis  Spent fuel compositions for each spatial region (typically regions)  separate burnup calculation for each region  interpolation on compositions unreliable  Extract nuclides for analysis  Treatment of isotopic uncertainties - apply bias and/or uncertainty correction factors (if applicable)  Resonance processing of multigroup cross sections  Criticality calculation

SCALE Driver and STARBUCS Input ARP CSASI (BONAMI / NITAWL / ICE) End All regions complete? NO YES ORIGEN-S Spent fuel depletion and decay (repeat for all regions) WAX KENO V.a or KENO-VI Resonance cross-section processing (repeat for all regions) All regions complete? NO Combine cross sections for all regions Criticality calculation STARBUCS Burnup Credit Sequence for SCALE 5

SMORES  SCALE Material Optimization and REplacement Sequence  Performs automated 1-D optimization for criticality safety analysis

SMORES Methodology  Prepare problem-dependent cross sections  BONAMI/NITAWL-III, or  BONAMI/CENTRM/PMC  ICE creates a self-shielded macroscopic cross section library  XSDRNPM 1-D calculation of forward and adjoint fluxes and k eff

SMORES Method (Cont.)  Calculate the effectiveness functions and perform the optimization  SWIF: First-order linear perturbation theory  Determine the configuration that results in the highest k eff with given fissile amount  Redistribute the fissile material and remove/redistribute other materials  Determine the configuration that yields the specified k eff with minimum amount of fissile material  Remove/redistribute the fissile and other materials

SMORES Example  Spherical fissile system with 239 PuO2, polyethylene, and beryllium  Eight equal-thickness zones  Flat fissile material profile (initial k eff = 0.7)  Determine maximum k eff for the system H2OH2O

SMORES Example (Cont.)

When will SCALE 5 be released?  My final answer:  When we have funding  When it’s ready  Sometime in 2003  You will be among the first to know if you join the SCALE News List