3. DESCRIPTION OF PROGRAM OR FUNCTION
The SCALE system was developed for the Nuclear Regulatory Commission to satisfy a need for a standardized method of analysis for the evaluation of nuclear fuel facility and package designs. In its present form, the system has the capability to perform criticality, shielding, radiation source term, spent fuel depletion/decay, and reactor physics analyses using well-established functional modules tailored to the SCALE system. See the developers' website and the SCALE 5 electronic notebook for news on SCALE, updates, and tips on running the code.
SCALE newsletter: http://www.ornl.gov/sci/scale/newsletter.htm
What's New in SCALE5.1? http://rsicc.ornl.gov/rsiccnew/SCALE5.1_WhatsNew.pdf
SCALE website http://www.ornl.gov/sci/scale
electronic notebook: http://rsicc.ornl.gov/rsiccnew/CFDOCS/scale5enotebook.cfm
The CSAS control module contains criticality safety analysis sequences that calculate the neutron multiplication factor for one-dimensional (XSDRNPM-S) and multidimensional (KENO V.a) system models. The CSAS module also has the capability to perform criticality searches (optimum, minimum, or specified values of k-eff) on geometry dimensions or nuclide concentrations in KENO V.a. The CSAS6 control module contains criticality safety analysis sequences using the KENO-VI module for multidimensional models with more complex geometries, including hexagonal arrays. Sequences that provide problem-dependent cross sections for use in stand-alone codes are also available in the CSAS module.
In addition, sensitivity and uncertainty (S/U) analysis capabilities for criticality safety are included in SCALE 5.1. Both 1-D and 3-D sequences plus several auxiliary codes have been developed into a new suite of sensitivity and uncertainty analysis codes called TSUNAMI (Tools for Sensitivity and Uncertainty Analysis Methodology Implementation). TSUNAMI contains a number of codes that were developed primarily to assess the degree of applicability of benchmark experiments for use in criticality code validations. However, the sensitivity and uncertainty data produced by these codes can be used in a wide range of studies. Sensitivity coefficients produced by the TSUNAMI sensitivity analysis sequences predict the relative changes in a system's calculated k-eff value due to changes in the neutron cross-section data. Both TSUNAMI-1D and TSUNAMI-3D fold the sensitivity data with cross-section covariance data to calculate the uncertainty in the calculated k-eff value due to tabulated uncertainties in the cross-section data. The applicability of benchmark experiments to the criticality safety validation of a given application can be assessed using S/U-based integral indices. The TSUNAMI-IP (Indices and Parameters) code utilizes sensitivity data and cross-section covariance data to produce a number of relational integral indices that can be used to assess system similarity.
The SAS2H module uses ORIGEN-S to perform a one-dimensional (1-D) fuel depletion analysis (to characterize spent fuel and/or generate source terms).
Two-dimensional (2-D) spent fuel depletion is available in the TRITON control module. TRITON couples ORIGEN-S depletion calculations with the 2-D flexible mesh discrete ordinates code NEWT. TRITON supports branch calculations that allow calculation of cross sections and their first derivatives with respect to fuel and moderator temperature, moderator density, soluble boron concentration, and control rod insertion, as a function of burnup. These cross sections are stored in a database format that can be retrieved and processed as appropriate for use by core analysis codes. The rigor of the NEWT solution in estimating angular flux distributions combined with the world-recognized accuracy of ORIGEN-S depletion gives TRITON the capability to perform rigorous burnup-dependent physics calculations with few implicit approximations.
Three-dimensional (3-D) Monte Carlo spent fuel depletion is a new capability added in SCALE 5.1 via the TRITON and TRITON6 control modules. TRITON couples ORIGEN-S depletion calculations with KENO V.a, while TRITON6 uses KENO-VI.
ORIGEN-ARP is an automated depletion decay sequence for both Windows and Unix/Linux systems. It includes a Windows graphical user interface (GUI) for ORIGEN-S and ARP (Automated Rapid Processing), which automatically interpolates cross sections on enrichment, burnup, and optionally moderator density using a set of standard basic cross-section libraries for LWR and MOX fuel assembly designs. The interpolated cross sections are passed to ORIGEN-S. Utility codes are provided so users can generate their own ORIGEN-ARP basic cross-section libraries via TRITON or SAS2H.
Other automated criticality safety related sequences include the STARBUCS 3-D burnup credit sequence (combining ORIGEN-ARP with KENO V.a or KENO-VI) and the SMORES 1-D material optimization sequence for criticality safety.
Four shielding analysis sequences are provided. SAS1 analyzes general one-dimensional shielding problems via XSDRNPM-S. SAS3 provides a general procedure for cross-section preparation followed by a shielding analysis using the MORSE-SGC Monte Carlo code. The SAS4 module has been tailored to perform a Monte Carlo shielding analysis for a cask-type geometry. An automated scheme is incorporated in SAS4 to generate Monte Carlo biasing parameters that enable SAS4 to calculate accurate doses with reasonable efficiency. The QADS module analyzes three-dimensional gamma-ray shielding problems via the point kernel code, QAD-CGGP.
DATA LIBRARIES INCLUDED:
SCALE Standard Composition Library
16 group criticality cross sections from Hansen Roach
44 group cross sections based on ENDF/B V
238 group cross sections based on ENDF/B V
238 group cross sections based on ENDF/B VI
27n, 18g coupled cross sections based on ENDF/B IV
22n, 18g coupled cross sections based on Straker Morrison
18 group photon cross sections based on OGRE data
ENDF/B-V continuous energy cross sections for CENTRM
ENDF/B-VI continuous energy cross sections for CENTRM
Albedos and weighting functions for use by KENO
Various cross section, decay, and yield libraries for ORIGEN S
ORIGEN ARP basic cross section libraries:
- Siemens 14x14
- Westinghouse CE 14x14, 16x16
- Westinghouse 14x14, 15x15, 17x17, 17x17 OFA (Optimized Fuel Assembly)
- GE 7x7, 8x8, 9x9, 10x10
- ABB 8x8
- ATRIUM-9 (9x9), ATRIUM-10 (10x10)
- SVEA-64 (8x8), SVEA-100 (10x10)
- VVER-440 flat enrichment (1.6% - 3.6%)
- VVER-440 profiled enrichment, average 3.82%
- VVER-440 profiled enrichment, average 4.25%
- VVER-440 profiled enrichment, average 4.38%
- CANDU 28- and 37-element bundles (available also in DLC-0210)
- AGR (Advanced Gas Cooled Reactor)
- Mixed oxide (MOX) fuel: 8x8, 9x9-1, 9x9-9, 10x10, 14x14, 15x15, 16x16, 17x17, 18x18