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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

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An Overview of What’s New in SCALE 5

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An overview of what s new in scale 5 l.jpg

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 l.jpg

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 l.jpg

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 l.jpg

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


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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


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SCALE 5 Criticality Sequence with CENTRM


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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


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Sensitivity/Uncertainty Codes

  • Use 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


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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


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Sensitivity Profiles Can Be Plotted to Highlight Similarities/Differences


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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


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NEWT Flexible Mesh Sn

  • NEWTransport 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 l.jpg

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

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) ! end

end materials

read geom

' Fuel rod

subgrid 1 1.2600 1.2600 4 4

cylinder 1 0.63 0.63 0.41000 !fuel! end

cylinder 4 0.63 0.63 0.4750 !clad! end

' Water hole

subgrid 4 1.2600 1.2600 4 4

cylinder 5 0.63 0.63 0.571 !water hole! end

cylinder 4 0.63 0.63 0.613 !guide tube! end

array 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 1

domain 21.42 21.42 30 30

boundary 1 1 1 1

end geom

end


Other models l.jpg

Other Models

  • Simple pin cell

  • VENUS-2 MOX benchmark (1/4 core)


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NEWT thermal spectra plots

BWR w/ Gd rods

MOX Supercell


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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


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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


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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


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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 l.jpg

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


Slide21 l.jpg

STARBUCS Burnup Credit Sequence for SCALE 5

SCALE

Driver

and

STARBUCS

Input

ARP

Spent fuel depletion and decay (repeat for all regions)

NO

All regions complete?

ORIGEN-S

YES

CSASI

(BONAMI / NITAWL / ICE)

Resonance cross-section

processing

(repeat for all regions)

NO

All regions complete?

Combine cross sections for all regions

WAX

KENO V.a or KENO-VI

Criticality calculation

End


Smores l.jpg

SMORES

  • SCALE Material Optimization and REplacement Sequence

  • Performs automated 1-D optimization for criticality safety analysis


Smores methodology l.jpg

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 l.jpg

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 l.jpg

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 l.jpg

SMORES Example (Cont.)


Smores example cont27 l.jpg

SMORES Example (Cont.)


When will scale 5 be released l.jpg

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 E-mail List

    www.ornl.gov/scale


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