RSICC DATA PACKAGE DLC-210
1. NAME AND TITLE OF DATA LIBRARY
CANDULIB-AECL: Burnup-Dependent ORIGEN-S Cross-Section Libraries for CANDU Reactor Fuel Characterization.
Requesters of CCC-545/SCALE4.4a or CCC-702/ORIGEN-ARP who wish to request DLC-210/ CANDU-AECL data libraries will be exempt from the transmittal fee for CANDULIB. Please complete the on-line order form on the RSICC website for the packages you wish and state that you request an exemption for the CANDULIB-AECL package.
2. NAME AND TITLE OF DATA RETRIEVAL PROGRAMS
Atomic Energy of Canada Limited, Sheridan Park Research Community, Toronto, Canada.
4. HISTORICAL BACKGROUND AND INFORMATION
In 1995, updated ORIGEN-S cross-section libraries were created as part of a program to upgrade and standardize the computer codes and nuclear data employed for used fuel characterization. This effort was funded through collaboration between Atomic Energy of Canada Limited and the Canadian Nuclear Power Utilities, under the CANDU Owners Group (COG). The updated cross sections were generated using the WIMS-AECL lattice code and ENDF/B-V and -VI based data to provide cross section consistency between reactor physics codes and used fuel characterization codes.
5. APPLICATION OF THE DATA
The libraries in this data collection are designed for characterizing used fuel from CANDU pressurized heavy water reactors. Two libraries are provided: one for the standard 28-element fuel bundle design, the other for the 37-element fuel bundle design. The libraries were generated for typical reactor operating conditions. The libraries are designed for use with the ORIGEN-S isotope generation and depletion code. The libraries may also be used with the ARP (Automated Rapid Processing) utility of SCALE, which performs interpolation of the library data to create problem-dependent libraries for an ORIGEN-S depletion analysis.
6. SOURCE AND SCOPE OF DATA
The CANDU libraries are updated with cross sections from a variety of different sources. Capture and fission cross sections were obtained from collapsed 89-group ENDF/B-V and -VI data from the WIMS-AECL lattice code. Other reaction cross sections were obtained from the SCALE 27-group ENDF/B-IV data. In all, cross sections for more than 200 important actinides and fission products were updated with burnup-dependent data. The source of the nuclear decay data and cross sections not updated were the base ORIGEN-S libraries distributed with the SCALE 4.2 code package.
7. DISCUSSION OF THE DATA RETRIEVAL PROGRAM
These data are intended for use with ORIGEN-S, which is distributed in the CCC-545/SCALE4.4a package. See the SCALE home page for more information: http://www.cped.ornl.gov/scale/ scale-home.html.
8. DATA FORMAT AND COMPUTER
The libraries are distributed in a card-image format for platform compatibility. They must be converted to binary library format using the OSBICO utility of SCALE. (See example of input included in the README file).
9. TYPICAL RUNNING TIME
a. Documentation available with library
I. C. Gauld, P. A. Carlson, and K. A. Litwin, "Production and Validation of ORIGEN-S Cross-Section Libraries for CANDU Reactor Fuel Studies," AECL report RC-1442, COG-I-95-200 (October 1995). This document is released with the understanding that it has not been edited for widespread distribution.
b. Other useful documentation
"SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation," NUREG/CR-0200, Rev. 6 (ORNL/NUREG/CSD-2/R6), Vols. I, II, and III (December 1999).
11. CONTENTS OF LIBRARY
Included are the referenced document in 10.a and two DS/HD diskettes which include the ASCII data written in a self-extracting, compressed Windows file and in a GNU compressed tar file.
12. DATE OF ABSTRACT
KEYWORDS: BASED ON ENDF/B-V; BASED ON ENDF/B-VI; BURNUP; FISSION PRODUCT YIELDS; NEUTRON CROSS SECTIONS