# Presentation: 128344

## 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 (**_**C**_**ontinuous **_**En**_**ergy **_**Tr**_**ansport **_**M**_**odule) **
**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 (**_**P**_**ointwise **_**M**_**ultigroup **_**C**_**onverter) **
**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

- 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 Sn

_**NEW**_** **_**T**_**ransport 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
-   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

**Simple pin cell**

**VENUS-2 MOX benchmark (1/4 core)**

## NEWT thermal spectra plots

- BWR w/ Gd rods

- MOX Supercell

## STARBUCS Features

**STARBUCS:  ****St****andardized ****A****nalysis of ****R****eactivity for ****Bu****rnup ****C****redit using ****S****CALE**
**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**

## STARBUCS Burnup Credit Sequence for SCALE 5

**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

_**S**_**CALE **_**M**_**aterial **_**O**_**ptimization and **_**RE**_**placement **_**S**_**equence**
**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**

- H2O

## SMORES Example (Cont.)

## SMORES Example (Cont.)

## When will SCALE 5 be released?

**My final answer:**
**When we have funding**
**When it’s ready**
**Sometime in 2003**
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**www.ornl.gov/scale**