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.

3. CONTRIBUTOR

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 and DLC-53/VITAMIN-4C.

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 data.

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

Not applicable.

8. DATA FORMAT AND COMPUTER

EBCDIC card images; IBM 360/91.

9. TYPICAL RUNNING TIME

Not applicable.

10. REFERENCES

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 problems.

12. DATE OF ABSTRACT

December 1984.