1. NAME AND TITLE OF DATA LIBRARY
CASK-81: 22 Neutron, 18 Gamma-Ray Group, P3, Cross Sections for Shipping Cask
2. NAME AND TITLE OF DATA RETRIEVAL PROGRAMS
LIBGEN: Card Image to Binary Converter.
For use with DLC-23G MICRO (PC Version):
STAR14: Generator of 14* Array for Use with PC Versions of GIP or ANISN-W.
DOSE: Converter of ANISN Fluxes to Dose Rates.
Oak Ridge National Laboratory, Oak Ridge, Tennessee.
Science Applications, Inc., Huntsville, Alabama.
Division of Material Licensing, U.S. Atomic Energy Commission, Washington, D. C.
Transnuclear, Inc., White Plains, New York.
4. HISTORICAL BACKGROUND AND INFORMATION
The cross-section data in DLC-23F/CASK-81 were compiled for the purpose of performing
calculations of spent fuel shipping casks. The data were described at the 1972 annual ANS meeting
in Las Vegas and results using the data were also presented. DLC-23G MICRO in FIDO format
contains the same data as DLC-23F formatted on two diskettes to facilitate use on micro computers.
Zirconium data from DLC-41/VITAMIN-C, which was included as file 5 of DLC-23F, replaces the
original LENDL data in the CASK file.
5. APPLICATION OF THE DATA
The data were designed for use in shielding analysis of PWR depleted uranium shipping casks. The results of such an analysis can be found in the packaged documentation.
The data were collapsed from a fine-group structure using a weighting function representative of
a water-uranium mixture. Thus, the application of this data for problems not similar to the shipping
cask type should be done with caution.
6. SOURCE AND SCOPE OF DATA
This library of coupled neutron and gamma-ray cross sections was compiled for several elements that are commonly used for shielding calculations. The coupled, P3, cross sections are given in the ANISN format which permits their usage in the discrete ordinates codes. The data sets from which DLC-23F/CASK-81 was derived are listed in the packaged documentation.
The library contains data for:
1H, 4He, Be, 10B, C, N, O, Na, Mg, Al, Si, K, Ca, Ti, Cr, Mn, Fe, Ni, Cu, Zr, Mo, Sn, Ta, W, Pb,
235U, 238U, 239Pu, 240Pu.
The source for the neutron cross sections was primarily the ENDF/B-II library, although some data were taken from other sources, when necessary. PSR-13/SUPERTOG was used to generate resonance-corrected fine-group cross sections for 104 energy groups from the ENDF/B library. Single level Breit-Wigner or multi-level Breit-Wigner resonance parameters were used by SUPERTOG to generate point cross sections for the resonance nuclides. Approximately 100 points per resolved resonance were used to integrate the point cross sections for the fine groups. In the unresolved resonance region, 81 points per fine group were used for the integration. A 1/E spectral weighting function was used.
The multigroup neutron cross sections in a 22 energy group structure were obtained from the 104 group cross sections by averaging the various elemental cross sections across a fine group flux calculated by ANISN for a uranium-water mixture using fine group cross sections.
The secondary gamma-ray production cross sections were calculated by PSR-11/POPOP4. Gamma-ray transport cross sections were calculated for an 18 group gamma-ray energy structure by the MUG code. The multigroup neutron cross sections, the secondary gamma production cross sections, and the gamma-ray transport cross sections were coupled to form a 40 group set. This is the same 40 group structure as used by Straker for various shielding calculations.
Calculations of the neutron and gamma-ray fluence from several shielding problems have been
performed and some results of these calculations are discussed in the packaged documentation.
7. DISCUSSION OF THE DATA RETRIEVAL PROGRAMS
LIBGEN converts the card image data into a so-called ANISN library for use as input to transport codes (for mainframe computers).
The STAR14 program transfers the desired cross sections from the DLC-23G MICRO/CASK
library to a 14* array for use with the PC version of either CCC-255C/ANISN-W or PSR-229/GIP.
DOSE reads the scalar flux and interval dimensions from the ANISN output to create a dose rate file.
STAR14 and DOSE can be run with the Microsoft 3.10 optimizing compiler.
8. DATA FORMAT COMPUTER
Card images; Mainframes or IBM PC.
9. TYPICAL RUNNING TIME
On the IBM PC AT computer, 1 hour and 38 minutes is required to execute STAR14, GIP,
ANISN-W and DOSE runs which comprise the sample case.
a. Included in the documentation:
R. W. Roussin, "Discussion of CASK Update," Informal note (June 1981).
Robert W. Roussin, "Energy Absorption Cross Sections in Multigroup Libraries," Informal note (April 1974).
R. W. Roussin, "Changes to the DLC-23 Cross Sections," Informal note (March 1974).
R. W. Roussin, "Changes to the Gamma-Ray Production Cross Sections for Some Materials in DLC-23/CASK (Formerly PDS-31)," Informal note (October 1973).
G. W. Morrison, E. A. Straker, and R. H. Odegaarden, "A Coupled Neutron and Gamma Ray Multigroup Cross Section Library for Use in Shielding Calculations," Trans. Am. Nucl. Soc., 15 (1972) 535.
G. W. Morrison, E. A. Straker, and R. H. Odegaarden, "The Use of the MORSE Monte Carlo Code to Solve Shielding and Criticality Problems of Spent Fuel Casks," Trans. Am. Nucl. Soc., 15 (1972) 547 (Slides from the oral presentation also provided).
R. W. Roussin and J. B. Wright, "Contents, Energy Group Structure, and Weight Function Used
for DLC-23/CASK," Informal notes (March 1975 Rev.)
b. Background information
E. A. Straker and M. L. Gritzner, "Neutron and Secondary Gamma-Ray Transport in Infinite
Homogeneous Air," ORNL-4464 (1969).
11. CONTENTS OF LIBRARY
Included are the referenced documents (10.a) and one (1.2MB) DOS diskette containing the data,
LIBGEN conversion program, sample input and output for mainframe use. For the PC version, two
diskettes which contain the data, auxiliary programs, and sample problem input, plus the executable
program files for STAR14 and DOSE.
12. DATE OF ABSTRACT
September 1973; revised November 1973; updated March 1974, March 1975, June 1981, June
KEYWORDS: ANISN FORMAT; COUPLED NEUTRON GAMMA-RAY CROSS SECTIONS; MICROCOMPUTER; MULTIGROUP CROSS SECTIONS