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.

3. CONTRIBUTORS

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 shielding calculations.

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 ANISN-ORNL.

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 form.

10. REFERENCE

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.