1. NAME AND TITLE OF DATA LIBRARY
VITAMIN-C: 171 Neutron, 36 Gamma-Ray Group Cross Sections in AMPX and CCCC Interface
Formats for Fusion and LMFBR Neutronics.
2. NAME AND TITLE OF DATA RETRIEVAL PROGRAMS
No retrieval program is included in the package.
Oak Ridge National Laboratory, Oak Ridge, Tennessee.
4. HISTORICAL BACKGROUND AND INFORMATION
DLC-41 was designed as a program for the development, generation, validation, and distribution
of a general purpose Processed Multigroup Cross Section Library (PMCSL) for use in controlled
thermonuclear research (CTR) neutronics studies. A single 171 neutron, 36 gamma-ray group cross-section library useful for both CTR and LMFBR neutronics analysis was defined. The master library
was generated using the MINX neutron processor and the gamma-ray processor from the AMPX
system. A preliminary release was made to stimulate implementation and testing at several installations to improve the quality of the libraries which were ultimately distributed as DLC-41/VITAMIN-C
5. APPLICATION OF THE DATA
The United States Energy Research and Development Administration (ERDA) Division of Controlled Thermonuclear Research (DCTR) began in 1974 to sponsor the Radiation Shielding Information Center (RSIC) at Oak Ridge National Laboratory (ORNL) to provide nuclear data and other information to meet the needs of ERDA-DCTR contractors. The VITAMIN-C project was the first special activity of RSIC in its role as the Nuclear Data Center for ERDA-DCTR and involves the generation, packaging, distribution, validation, and maintenance of a general-purpose multigroup cross-section library for neutronics and other radiation transport studies.
DLC-41/VITAMIN-C was reorganized by removing the AMPX modules from the package. The
sample problems are designed to be executed using PSR-117/MARS (or PSR-63/AMPX-II). The
change was made to try to minimize the updating required when a change is made to a retrieval code
needed by a particular DLC.
6. SOURCE AND SCOPE OF DATA
The MINX computer code generates multigroup neutron cross sections with self-shielding factors. For a given nuclide, MINX calculates multigroup energy-averaged microscopic neutron cross sections from evaluated cross section files in ENDF format. Group constants in both the AMPX and CCCC interface formats are produced by a version of MINX which is operational on the IBM 360 system at ORNL. The Bondarenko method (narrow resonance approximation) is employed to create group dependent resonance self-shielding factors to account for temperature and dilution effects.
MINX was developed to calculate multigroup constants with user control over computational
errors. This helps to distinguish uncertainties in the multigroup data arising from numerical
approximations in the averaging procedure from those due to errors in the basic ENDF cross section
H, 2H, 3H, 4He, 6Li, 7Li, 9Be, 10B, 11B, 12C, 14N, 16O, F, 23Na, Mg, 27Al, Si, P, S, K, Ca, Ti, V, Cr,
55Mn, Fe, 59Co, Ni, Cu, 63Cu, 65Cu, 2Zirc, 93Nb, Mo, 107Ag, 109Ag, Cd, Sn, 151Eu, 152Eu, 153Eu, 154Eu,
181Ta, 182W, 183W, 184W, Pb, 232Th, 233U, 234U, 235U, 236U, 238U, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, 241Am.
7. DISCUSSION OF THE DATA RETRIEVAL PROGRAM
8. DATA FORMAT AND COMPUTER
EBCDIC card images; IBM 360/91.
9. TYPICAL RUNNING TIME
R. W. Roussin, C. R. Weisbin, J. E. White, N. M. Greene, R. Q. Wright, and J. B. Wright, "The CTR Processed Multigroup Cross-Section Library for Neutronics Studies," ORNL-RSIC-37 (July 1980).
R. W. Roussin and D. B. Simpson, "Sample Problems for DLC-41/VITAMIN-C," presented at
the RSIC Workshop on Multigroup Cross-Section Handling Codes (March 1978).
11. CONTENTS OF LIBRARY
Included are the referenced documents and one tape cartridge in TAR format or 11 (1.44MB) DOS
diskettes which contain the neutron group cross sections, JCL and input data, and output from
12. DATE OF ABSTRACT
KEYWORDS: AMPX INTERFACE FORMAT; COUPLED NEUTRON-GAMMA-RAY CROSS SECTIONS; CTR PROCESSED CROSS-SECTION LIBRARY; GAMMA-RAY CROSS SECTIONS; GAMMA-RAY PRODUCTION DATA; MULTIGROUP CROSS SECTIONS; MULTIGROUP CROSS SECTIONS BASED ON ENDF/B; NEUTRON CROSS SECTIONS; REACTION CROSS SECTIONS