1. NAME AND TITLE OF DATA LIBRARY
FCXSEC: 22 Neutron, 21 Gamma-Ray Group Cross Section Libraries in ANISN Format for
Nuclear Fuel Cycle Shielding Calculations.
2. NAME AND TITLE OF DATA RETRIEVAL PROGRAMS
BCD2BIN: Converts ASCII data to binary ANISN library.
Oak Ridge National Laboratory, Oak Ridge, Tennessee.
Battelle, Columbus Ohio.
4. HISTORICAL BACKGROUND AND INFORMATION
The FCXSEC library was derived from DLC-41/VITAMIN-C and both microscopic and
macroscopic cross sections are included. Starting with the pseudo-composition-independent
VITAMIN-C cross-section library, composition-dependent fine- (171n-36g) and broad-group (22n-21g)
self-shielded AMPX master, broad-group microscopic ANISN-formatted, and broad-group
macroscopic ANISN-formatted cross-section libraries were generated to be used for nuclear fuel cycle
5. APPLICATION OF THE DATA
The library once used for ANISN calculations for nuclear fuel cycle activities was an in-house,
coupled 27-neutron-20-gamma-ray group, microscopic cross-section library. It had two deficiencies:
1) the source of the cross-section data and the methods for producing the library were unknown and
2) some of the cross-section data for key materials were apparently unknown. To correct these
problems, the FCXSEC libraries were created to 1) be based on "accepted," currently available,
pseudo-problem-independent, cross-section data; 2) include a wide variety of nuclides in a variety of
specified materials and concentrations; 3) have neutron and gamma-ray energy structures compatible
with a broad case of radiation transport calculations required in nuclear fuel cycle studies; 4) be
available as fine-group and broad-group microscopic libraries in AMPX master format and broad-group microscopic and macroscopic libraries in ANISN format; 5) be tested for accuracy by checking
for internal data consistency; and 6) be documented with respect to the sources of the basic cross-section data, the processing of the data, and the results of the first order check of the cross sections.
6. SOURCE AND SCOPE OF DATA
Data are provided for water, stainless steel 304, spent U-fuel, air, spent Th-fuel, Gd2O3, Simonite concrete, type 2a concrete, boron in water, Li, plutonium oxide, Eu, B in boral, W, and zircalloy. Microscopic cross sections for the constituents of the above materials are also contained in the package. A retrieval code for BCD-to-binary conversion is included.
The following isotopes are included in the FCXSEC master cross-section libraries:
1H, 16O, 10B, 11B, Mg, 27Al, Si, 32S, Ca, Fe, K, Ba, D, 6Li, 7Li, 12C, 14N, 238U, 235U, Ni, Cr, F, 31P,
149Sm, 151Eu, 153Eu, Gd, 182W, 183W, 184W, 186W, 234U, 236U, 237Np, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu,
241Am, 243Am, 244Cm, 232Th, 233U, 107Ag, 109Ag, Na, Zrly, Cd, 93Nb, Mo, Pb, Be, Ti, V, 55Mn, 59Co,
Cu, Sn, 101Ta.
7. DISCUSSION OF THE DATA RETRIEVAL PROGRAM
The BCD2BIN program converts the ASCII (card image) data into a binary file for use with
8. DATA FORMAT AND COMPUTER
BCD output; IBM 370/3033, IBM PC 386, IBM PC 486.
9. TYPICAL RUNNING TIME
On a PC 386, the FCXB2B program took about 15 minutes to convert the microscopic cross
sections to binary form. It took about 3 minutes to convert the macroscopic cross sections to binary
W. E. Ford, III, C. C. Webster, B. R. Diggs, R. E. Pevey, and A. G. Croff, "FCXSEC:
Multigroup Cross-Section Libraries for Nuclear Fuel Cycle Shielding Calculations," ORNL/TM-7038
(ENDF-287) (May 1980).
11. CONTENTS OF LIBRARY
Included are the referenced documents and two (1.2MB) DOS diskettes which contain the 22n, 21g
microscopic cross sections in ANISN format, BCD-binary conversion program, and input data for BD-binary program, plus output from the BCD-binary run.
12. DATE OF ABSTRACT
April 1985; April 1991.
KEYWORDS: ANISN FORMAT; COUPLED NEUTRON-GAMMA-RAY CROSS SECTIONS; MICROCOMPUTER; MULTIGROUP CROSS SECTIONS