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. GoluogluOak 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 P roblem-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 CodesUse adjoint-based first order linear perturbation theory

to calculate sensitivities and propagate uncertainties

Operate as automated SCALE analysis sequences

keff 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 keff 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 keff 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 Sn 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=newtCalvert Cliffs fuel assembly (one-fourth)read parmfillmix=5 prtmxsec=no prtbroad=no sn=6 inners=10 outers=200 end parmread materials 1 1 1.0 ! 3.0 enriched fuel, pin location 1 ! end 4 1 0.0 @cladding@ end 5 2 0.0 ! water (background material) ! endend materialsread geom' Fuel rodsubgrid 1 1.2600 1.2600 4 4cylinder 1 0.63 0.63 0.41000 !fuel! endcylinder 4 0.63 0.63 0.4750 !clad! end' Water holesubgrid 4 1.2600 1.2600 4 4cylinder 5 0.63 0.63 0.571 !water hole! endcylinder 4 0.63 0.63 0.613 !guide tube! endarray 0.0 0.0 17 17 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 4 1 1 4 1 1 4 1 1 1 1 1 1 1 1 4 1 1 1 1 1 1 1 1 1 4 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 4 1 1 4 1 1 4 1 1 4 1 1 4 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 4 1 1 4 1 1 4 1 1 4 1 1 4 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 4 1 1 4 1 1 4 1 1 4 1 1 4 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 4 1 1 1 1 1 1 1 1 1 4 1 1 1 1 1 1 1 1 4 1 1 4 1 1 4 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1domain 21.42 21.42 30 30boundary 1 1 1 1end geomend

Other Models

Simple pin cell VENUS-2 MOX benchmark

(1/4 core)

NEWT thermal spectra plotsBWR w/ Gd rods MOX Supercell

STARBUCS FeaturesSTARBUCS 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 10-18 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

SCALEDriverandSTARBUCS

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 keff

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 keff with given fissile amount Redistribute the fissile material and remove/redistribute

other materials

Determine the configuration that yields the specified keff with minimum amount of fissile material Remove/redistribute the fissile and other materials

SMORES Example

Spherical fissile system with 239PuO2, polyethylene, and beryllium

Eight equal-thickness zones Flat fissile material profile

(initial keff = 0.7)

Determine maximum keff for the system

H2O

SMORES Example (Cont.)

239PuO2, Polyethylene, Beryllium Sphere

00.00050.001

0.00150.002

0.00250.003

0.00350.004

0.0045

0 5 10 15 20

Radius (cm)

Vo

lum

e F

ract

ion

239PuO2 - final

239PuO2 - initial

SMORES Example (Cont.)

239PuO2, Polyethylene, Beryllium Sphere

0

0.2

0.4

0.6

0.8

1

0 5 10 15 20

Radius (cm)

Vo

lum

e F

ract

ion

polyethylene - final

beryllium - final

polyethylene - initial

beryllium - initial

When will SCALE 5 be released?

My final answer: When we have funding When it’s ready Sometime in 2003

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