1. NAME AND TITLE OF DATA LIBRARY
PUCOR: 84 Group Neutron Cross Sections for Uranium-Plutonium Cycle LWR and PWR Models in AMPX Master Library Format.
2. NAME AND TITLE OF DATA RETRIEVAL PROGRAMS
No retrieval program is in the package.
Oak Ridge National Laboratory, Oak Ridge, 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.
5. APPLICATION OF THE DATA
The data in DLC-67/PUCOR 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 that ORIGEN would predict the correct spent fuel compositions without having to resort to the adjustment of cross sections, as in previous ORIGEN reactor models.
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 84 neutron group cross sections in AMPX format (24 materials) include:
H, Fe, Cr, Ni, O, N, C, 10B, 241Pu, 239Pu, 235U, Mn, Nb, 233U, 233Pa, Mo, 240Pu, 11B, Na, 2Zr, Pure Scatterer, Pure l/v Abs., Sn, 238U.
7. DISCUSSION OF THE DATA RETRIEVAL PROGRAM
8. DATA FORMAT AND COMPUTER
BCD output; IBM 360/91.
9. TYPICAL RUNNING TIME
Robert W. Roussin, "Informal Notes" (September 1979).
A. G. Croff, M. A. Bjerke, G. W. Morrison, and L. M. Petrie, " Revised Uranium-Plutonium Cycle PWR and BWR Models for the ORIGEN Computer Code," ORNL/TM-6051 (September 1978).
11. CONTENTS OF LIBRARY
Included are the referenced documents and 1 tape cartridge which contains the 84 neutron group cross sections in AMPX format (fission products) and 84 neutron group cross sections in AMPX format (actinides, structural materials, moderators and poisons).
12. DATE OF ABSTRACT
KEYWORDS: AMPX INTERFACE FORMAT; MULTIGROUP CROSS SECTIONS