1. NAME AND TITLE OF DATA LIBRARY
ANSL-V: ENDF/B-V Based Multigroup Cross Section Libraries for Advanced Neutron Source (ANS) Reactor Studies.
2. NAME AND TITLE OF DATA RETRIEVAL PROGRAMS
AIM: Program to Convert the Mode from BCD to Binary; available in PSR- 315/AMPX-77.
Oak Ridge National Laboratory, Oak Ridge, Tennessee.
4. HISTORICAL BACKGROUND AND INFORMATION
Pseudo-problem-independent, multigroup cross section libraries were generated to support the Advanced Neutron Source (ANS) reactor design studies. The ANS was a proposed reactor which would be fueled with highly enriched uranium and cooled with heavy water. The libraries, designated ANSL-V (Advanced Neutron Source Cross Section Libraries based on ENDF/B-V) are data based in AMPX master format. Although the ANS project was cancelled, the libraries are being released because they may be utilized in other applications.
5. APPLICATION OF THE DATA
ANSL-V data are to be used for the subsequent generation of problem-dependent fine- and/or broad-group cross sections for a wide range of applications, such as core and shield analysis, activation analyses after irradiation of certain elements in the reactor environment, and safety analyses. Problem-dependent cross sections can be derived from the ANSL-V data with AMPX modules included in PSR-315/AMPX-77, PSR-352/SCAMPI, or CCC-545/SCALE 4.3 packages.. The derived data libraries in either ANISN or AMPX working library format can be used with codes such as KENO, ANISN, XSDRNPM, DORT, TORT, MORSE.
6. SOURCE AND SCOPE OF DATA
ANSL-V consists of the following fine and broad groups:
1. Fine Group (99 energy groups) General Purpose Neutron Library (FGGPN);
2. Broad Group (39 energy groups) General Purpose Neutron Library, BGGPN;
3. Gamma-Ray Interaction (GRI) Library containing data in 44-group gamma-ray structure.
4. Coupled library containing (CNG) 99-group neutron and 44-group gamma-ray data.
5. Coupled neutron-gamma (CNG) Library containing 39-group, and 44-group gamma-ray data.
Neutron and secondary gamma-ray production data in the ANSL-V library were generated primarily from evaluations in the ENDF/B-V General Purpose Library. Where evaluations for specified materials were not available in the ENDF/B-V library, ANSL-V GPN and SGRP data sets were generated from evaluations from other ENDF-formatted libraries. Gamma-ray interaction data sets were generated from evaluations in the ENDF/B-V Photon Interaction Library (DLC-99/HUGO). Because some important materials in the ENDF/B-V library do not have gamma-ray production files in the evaluation, both neutron and gamma-ray data from the LENDL-V evaluations were used for 12 materials. Even though both the neutron and gamma-ray evaluations for Sn were taken from LENDL-V, they have different identifiers. In support of local projects, ORNL extended the Sn neutron cross sections down to 10-5ev and changed the mat number to 8850, while the gamma-ray identifier was unchanged from 7850.
7. DISCUSSION OF THE DATA RETRIEVAL PROGRAMS
The AIM module of PSR-315/AMPX-77, PSR-352/SCAMPI, or CCC-545/SCALE 4.3 can be used for mode conversion of the data. Some other AMPX utility modules, which are included in these packages may also be used. Note that previous versions of AMPX and SCALE will not work because of the AMPX master library format changes.
8. DATA FORMAT AND COMPUTER
ASCII, Card images; All computers.
9. TYPICAL RUNNING TIME
R. Q. Wright, J. P. Renier, J. A. Bucholz, "ANSL-V: ENDF/B-V Based Multigroup Cross-Section Libraries for Advanced Neutron Source Reactor Studies Supplement 1," ORNL-6618/s1 (August 1995).
W. E. Ford III, et al."ANSL-V: ENDF/B-V Based Multigroup Cross-Section Libraries for Advanced Neutron Source (ANS) Reactor Studies," ORNL-6618 (September 1990).
11. CONTENTS OF PACKAGE
The package contains the references listed above and the data libraries. The data libraries are transmitted on one DC 6150 cartridge tape in tar format.
12. DATE OF ABSTRACT
April 1993, September 1996.
KEYWORDS: AMPX INTERFACE FORMAT; BASED ON ENDF/B-V; COUPLED NEUTRON- GAMMA-RAY CROSS SECTIONS; GAMMA-RAY CROSS SECTIONS; GAMMA-RAY PRODUCTION DATA; MULTIGROUP CROSS SECTIONS; NEUTRON CROSS SECTIONS