1. NAME AND TITLE OF DATA LIBRARY
BUGLE-93: Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications.
2. NAME AND TITLE OF DATA RETRIEVAL PROGRAMS
BCBN: Convert ANISN card-image data to binary format.
Oak Ridge National Laboratory, Oak Ridge, Tennessee.
4. HISTORICAL BACKGROUND AND INFORMATION
A new multigroup cross-section library based on ENDF/B-VI data has been produced and tested for light water reactor shielding and reactor pressure vessel dosimetry applications. The broad-group library, designated BUGLE-93, is intended to replace the BUGLE-80 and SAILOR libraries, which are both based on ENDF/B-IV data. The processing methodology for BUGLE-93 is consistent with ANSI/ANS 6.1.2, since the ENDF data were first processed into a fine-group, pseudo-problem-independent format and then collapsed into the final broadgroup format. An extensive integral data testing effort was performed to qualify the data and to assess its impact on LWR shielding applications. In general, results using the new data show significant improvements relative to earlier ENDF data.
5. APPLICATION OF THE DATA
The BUGLE-93 cross sections are intended for use in LWR shielding and pressure vessel dosimetry applications. The multigroup data have been collapsed, and in some cases self-shielded, using flux spectra typical of PWR and BWR reactor models. Flux spectra from five specific locations within these models were used, corresponding to: (1) off-center in a BWR core region, (2) off-center in a PWR core region, (3) the downcomer region in a PWR model, (4) within the pressure vessel at a depth of one-fourth the total thickness, and (5) within the concrete shield surrounding a PWR reactor vessel. The concrete-spectrum-weighted cross sections have been shown to be generally applicable to a wide range of shielding problems.
6. SOURCE AND SCOPE OF DATA
BUGLE-93 contains 120 nuclides which have been processed as infinitely dilute and collapsed using an LWR concrete shield spectrum. Additionally, it contains 105 nuclides which have been energy self-shielded and collapsed using LWR-specific material compositions and flux spectra. Nuclides with Z < 30 (hydrogen through copper) are given in a P7 Legendre expansion while P5 expansion is available for all other nuclides. Several dosimetry and standard response functions are included with the library along with kerma factors for all nuclides. The library was collapsed from the VITAMIN-B6 fine-group library using the AMPX-77 processing code system. VITAMIN-B6 is derived from ENDF/B-VI nuclear data, except for two nuclides (Sn obtained from LENDL and Zirc2 obtained from ENDF/B-IV). The responses and kerma factors were also derived primarily from ENDF/B-VI.
Attached tables provide information on file contents.
7. DISCUSSION OF THE DATA RETRIEVAL PROGRAM
The BCBN Fortran program is included to read formatted ANISN records and write them as unformatted records. No input is required.
8. DATA FORMAT AND COMPUTER
Card images in ANISN format; all computers (D00175/ALLCP/00).
9. TYPICAL RUNNING TIME
D. T. Ingersoll, J. E. White, R. Q. Wright, H. T. Hunter, C. O. Slater, N. M. Greene, R. E. MacFarlane, R. W. Roussin, "Production and Testing of the VITAMIN-B6 Fine-Group and the BUGLE-93 Broad-Group Neutron/Photon Cross-Section Libraries Derived from ENDF/B-VI Nuclear Data," ORNL-6795 (Draft 4/94).
11. CONTENTS OF LIBRARY
Included are the referenced document and 5 DS/HD 5.25-in. (1.2 MB) diskettes written in self-extracting compressed DOS files which contain the data files and BCBN retrieval program.
12. DATE OF ABSTRACT
February 1994, April 1994, May 1994.
KEYWORDS: ANISN FORMAT; BENCHMARK PROBLEM CROSS SECTIONS; CONCRETE CROSS SECTIONS; COUPLED NEUTRON-GAMMA-RAY CROSS SECTIONS; MULTIGROUP CROSS SECTIONS; MULTIGROUP CROSS SECTIONS BASED ON ENDF/B; NEUTRON CROSS SECTIONS; KERMA FACTORS; DETECTOR RESPONSE