1. NAME AND TITLE OF DATA LIBRARY

FIPDOR: 126 Neutron Group Fission Product Cross Sections.

2. NAME AND TITLE OF DATA RETRIEVAL PROGRAMS

No retrieval program is in the package.

3. CONTRIBUTOR

Oak Ridge National Laboratory, Oak Ridge, Tennessee.

Tennessee Valley Authority, Knoxville, Tennessee.

4. HISTORICAL BACKGROUND AND INFORMATION

The ORIGEN computer code--written in the late 1960s and early 1970s--is used for calculating the buildup and depletion of isotopes in nuclear materials. The code was principally intended for use in generating spent fuel and waste characteristics (composition, thermal power, etc.) that would form the basis for the study and design of fuel reprocessing plants, spent fuel shipping casks, waste treatment and disposal facilities, and waste shipping casks.

In 1975, a program was initiated to update ORIGEN and its associated data bases and reactor models. In June, 1983, LMFBR models for the ORIGEN2 computer code were published.

5. APPLICATION OF THE DATA

The data in DLC-68/FIPDOR can be used to produce revised cross sections for ORIGEN. The 84 neutron group cross sections in AMPX master library format are divided into two files. The first has 180 fission products and the second has 24 structural, actinides, moderators and poisons.

The fundamental objective of this work was to develop LMFBR models based on existing cross section data rather than the arbitrary adjustment of cross sections that typified previous ORIGEN reactor models. Two separate libraries were created: (1) a smaller one containing nuclides whose presence in the reactor would have the greatest effects on the neutron spectrum and depletion characteristics and (2) a larger library containing many nuclides of interest in ORIGEN2 with negligible effects on the neutron spectrum and depletion. Only the smaller library was considered in the subsequent multigroup fuel-depletion calculations.

6. SOURCE AND SCOPE OF DATA

This project involved the gathering and processing of a large amount of diverse data which led to the generation of revised ORIGEN reactor models for uranium- and uranium-plutonium-fueled PWRs and BWRs. The specific types of information developed for PWR-U, PWR-PuU, PWR-PuPu, BWR-U, BWR-PuU, and PWR-PuPu fuels are: 1) 84-energy-group neutron spectra; 2) one-group, burnup-dependent cross sections for the major actinides; 3) one-group, "typical" cross sections for 233 nuclides (including the actinides); 4) new values for the ORIGEN flux parameters THERM, RES, FAST; 5) parameters related to the activation of fuel-assembly structural materials outside the active fuel region; 6) recommended initial heavy-metal compositions of fuel-assembly structural materials; and 7) recommended minor constituent concentrations for both the fuel material and the structural materials.

The data in DLC-68/FIPDOR were generated using the NPTXS and XLACS modules of PSR-63/AMPX-II. The library consists of 126 group neutron cross sections for 181 fission products in the ENDF/B-IV library, as follows:

72Ge, 73Ge, 74Ge, 76Ge, 75As, 76Se, 77Se, 78Se, 80Se, 82Se, 79Br, 81Br, 80Kr, 82Kr, 83Kr, 84Kr, 85Kr, 86Kr, 85Rb, 86Rb, 87Rb, 86Sr, 87Sr, 88Sr, 89Sr, 90Sr, 89Y, 90Y, 91Y, 90Zr, 91Zr, 92Zr, 93Zr, 94Zr, 95Zr, 96Zr, 93Nb, 94Nb, 95Nb, 94Mo, 95Mo, 96Mo, 97Mo, 98Mo, 99Mo, 100Mo, 99Tc, 99Ru, 100Ru, 101Ru, 102Ru, 103Ru, 104Ru, 105Ru, 106Ru, 103Rh, 105Rh, 104Pd, 105Pd, 106Pd, 107Pd, 108Pd, 110Pd, 107Ag, 109Ag, 109Ag, 111Ag, 108Cd, 110Cd, 111Cd, 112Cd, 113Cd, 114Cd, 115MCd, 116Cd, 113In, 115In, 115Sn, 116Sn, 117Sn, 118Sn, 119Sn, 120Sn, 122Sn, 123Sn, 124Sn, 125Sn, 126Sn, 121Sb, 123Sb, 124Sb, 125Sb, 126Sb, 122Te, 123Te, 124Te, 125Te, 126Te, 127MTe, 128Te, 129MTe, 130Te, 132Te, 127I, 129I, 130I, 131I, 135I, 128Xe, 129Xe, 130Xe, 131Xe, 132Xe, 133Xe, 134Xe, 135Xe, 136Xe, 133Cs, 134Cs, 135Cs, 136Cs, 137Cs, 134Ba, 135Ba, 136Ba, 137Ba, 138Ba, 140Ba, 139La, 140La, 140Ce, 141Ce, 142Ce, 143Ce, 144Ce, 141Pr, 142Pr, 143Pr, 142Nd, 143Nd, 144Nd, 145Nd, 146Nd, 147Nd, 148Nd, 150Nd, 147Pm, 148Pm, 148MPm, 149Pm, 151Pm, 147Sm, 148Sm, 149Sm, 150Sm, 151Sm, 152Sm, 153Sm, 154Sm, 151Eu, 152Eu, 153Eu, 154Eu, 155Eu, 156Eu, 157Eu, 154Gd, 155Gd, 156Gd, 157Gd, 158Gd, 160Gd, 159Tb, 160Tb, 160Dy, 161Dy, 162Dy, 163Dy, 164Dy, 165Ho, 166Er, 167Er.

Prior to this work, fission product cross section sets existed only as few-group lumped fission products and as fine-group sets containing a limited number of nuclides and reactions. This is the first 126-group library to include all the reactions of the major fission products. This data is intended mainly for use in fast-reactor analysis as a base fine-group set from which to compute few-group libraries.

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, Informal Notes (October 1979).

A. G. Croff, J. W. McAdoo, and M. A. Bjerke, LMFBR Models for the ORIGEN2 Computer Code, ORNL/TM-7176/R1 (June 1983).

11. CONTENTS OF LIBRARY

Included are the referenced documents and a reel of magnetic tape which contains 126 neutron group cross sections in AMPX format; total records 476,216.

12. DATE OF ABSTRACT

March 1985; updated July 1985.