RSICC DATA PACKAGE DLC-226
1. NAME AND TITLE OF DATA LIBRARY
ENDF/B-VII.0-ACE: Neutron Cross Section Library in NJOY ACE Format
Based on ENDF/B-VII.0.
2. NAME AND TITLE OF DATA RETRIEVAL PROGRAMS
MAKXSF: Preparer of Cross Sections from LANL MCNP package (not included here.)
3. CONTRIBUTOR
National Nuclear Data Center at Brookhaven National Laboratory, Upton, New York.
4. HISTORICAL BACKGROUND AND INFORMATION
The DVD contains ACE-format files for 392 materials in the ENDF/B-VII.0 neutron reaction sublibrary (a complete set except 253Es) and all 20 materials in the thermal neutron scattering sublibrary. These files were generated by the NNDC with the processing code NJOY-99.161 as a part of the Phase 1 testing (data verification) of the ENDF/B-VII.0 library.
Distribution of ENDF/B-VII.0-ACE is limited, and requests are subject to approval. Browse to this link to order the data http://rsicc.ornl.gov/rsiccnew/CFDOCS/ENDFrequest_form1.cfm.
Note that ENDF/B-VII based MCNPDATA libraries from LANL are available in the current CCC-740/MCNP code and data package. These data have been tested and may be more suitable for your applications. The MCNP package can be requested using the online order forms on this website.
5. APPLICATION OF THE DATA
This continuous energy cross-section data library in ACE format is for shielding and criticality applications done with MCNP. Simple calculations using the Godiva model were performed to ensure that the ACE-format files could be used by the MCNP5 code. However, no strict Quality Assurance or benchmark tests were conducted on these files. Due to the preliminary nature of these files, NNDC does not provide any kind of support and/or technical assistance to end-users; neither does it guarantee the accuracy of the processed data.
6. SOURCE AND SCOPE OF DATA
The neutron sublibrary was processed at the temperature of 300o Kelvin, while different temperatures, ranging from 19o to 296o Kelvin for individual materials, were used for the thermal neutron scattering sublibrary. The resulting ACE format files were tested in simple neutronics calculations with the Monte Carlo transport code MCNP5. The files are organized in the following directory structure:
Neutron
§ NJOY_Input_300K - the general input deck used to process all materials, and
§ ACE_Files - ACE-format (ASCII) files for all materials along with the cumulative xsdir file.
Thermal_Scattering
§ NJOY_Inputs - input decks for each material processed, and
§ ACE_Files - ACE-format (ASCII) files for each material along with the xsdir file.
7. DISCUSSION OF THE DATA RETRIEVAL PROGRAM
The MAKXSF code which prepares MCNP cross-section libraries is not included in this package but is distributed with the MCNP5/MCNPX (CCC-730) and MCNP-4C2 (CCC-0701) code packages.
8. DATA FORMAT AND COMPUTER
RSICC ID is D00226MNYCP01. The ASCII data files are distributed in 2 formats for the convenience of users: a Windows self-extracting compressed file and a GNU compressed Unix tar file. Expanding either file will create a subdirectory called "Dlc226" that contains all the data. The expanded package requires approximately 1.7 GB of disk space.
9. TYPICAL RUNNING TIME
Run times vary depending on a number of factors.
10. REFERENCE
No published report is available for this library.
11. CONTENTS OF LIBRARY
The ASCII data libraries and xsdir file are transmitted on a DVD in a WinZIP file and in a unix gzip file.
12. DATE OF ABSTRACT
Data released December 2006, revised January 2007; abstract prepared June 2007 and revised Nov. 2009.
KEYWORDS: MCNP FORMAT; NEUTRON CROSS SECTIONS