RSIC DATA LIBRARY DLC-038

1. NAME AND TITLE OF DATA LIBRARY 
ORYX-E: ORIGEN Yields and Cross Sections--Nuclear Transmutation and Decay Data from 
ENDF/B-IV. 

2. NAME AND TITLE OF DATA RETRIEVAL PROGRAMS 
COLAPS: Group Collapsing Code. 
UPDATE: ORIGEN Fission Product data Library Update Code. 

3. CONTRIBUTOR 
Oak Ridge National Laboratory, Oak Ridge, Tennessee. 

4. HISTORICAL BACKGROUND AND INFORMATION 
ORYX-E increases the versatility of CCC-217/ORIGEN, the isotope generation and depletion
code package, by providing basic cross section and decay information for light element, 
fission-product, and actinide nuclides. This data library package results from data 
compiled for ORNL Chemical Technology Division's work with ORIGEN and from a 2-year 
effort of the Cross Section Evaluation Working Group (CSEWG) Fission Product Task Force.

5. APPLICATION OF THE DATA 
The data is generated from ENDF/B-IV and is formatted for input to the ORIGEN (CCC-217) 
code. Applications include calculations for waste projection, decay heat, nuclear 
safeguards, and fuel cycle economics. 

6. SOURCE AND SCOPE OF DATA 
The data library is generated from the ENDF/B-IV fission product data which includes 
decay and cross-section information as described in the following table. The capture 
cross-section information of all fission product nuclides for which capture 
cross-sectioninformation is given (~180 nuclides) was processed into 124 energy groups 
using MINX. Multigroup cross sections were generated at 0 with infinite dilution and 
one broad thermal group. Fine group data was generated using a Maxwellian 1/E fission 
spectrum with a 1% tolerance. 
Table 1. Comparison of Original ORIGEN and ENDF/B-IV Fission Product Data Library 
 ENDF/B-IV ORIGEN 
Number of nuclides  825  461  
Radioactive  712  338  
First excited state  117  83  
Second excited state  7  --  
Delayed neutron precursors  57  --  
Alpha decay  6  0  
Positron decay  17  11  
Cross sections (LMFBR)  181  423*  
*1-group (n,gamma) cross sections. 

7. DISCUSSION OF THE DATA RETRIEVAL PROGRAM 
A utility routine, COLAPS, can be used to collapse the fine cross sections by weighting 
with a user-specified reference spectrum. 
A second utility routine, UPDATE, will update the ORIGEN fission product data library 
using the cross sections generated with COLAPS. Both utility codes are included in the 
data package. 

8. DATA FORMAT AND COMPUTER 
EBCDIC card images; IBM 360. 

9. TYPICAL RUNNING TIME 
To collapse the cross sections, update the ORIGEN data tape, and run an ORIGEN sample 
problem took about one minute on the IBM 360/91 (CPU) for about 10 time steps. 

10. REFERENCES 
a. Included in the documentation: 
G. W. Morrison, C. R. Weisbin, and C. W. Kee, "Decay Heat Analysis for an LMFBR Fuel 
Assembly Using ENDF/B-IV Data," Proceedings of Conference on Nuclear Cross Sections and 
Technology, Washington, D. C. (March 1975), to be published. 
C. W. Kee,"A Revised Light Element Library for the ORIGEN Code," ORNL-TM-4896 (May 1975).
G. W. Morrison, C. R. Weisbin, and C. W. Kee, "Projected CRBRP Spent Fuel Characteristics
and Their Impact on NDA Techniques," Trans. Am. Nucl. Soc., 19, (1975) 487-488. 
C. W. Kee, C. R. Weisbin, and R E. Schenter, "Processing and Testing of ENDF/B-IV Fission
Product and Transmutation Data," Trans. Am. Nucl. Soc., 19 (1974) 398-399. 
b. Background information: 
CCC-217/ORIGEN. 

11. CONTENTS OF LIBRARY 
Included are the referenced documents (10.a) and one (1.2MB) DOS diskette which contains 
the light element, nuclide, fission products, and photon libraries and the source and 
sample data for COLAPS and UPDATE, plus output from COLAPS and UPDATE. 

12. DATE OF ABSTRACT 
September 1975; updated January 1976, March 1985. 

KEYWORDS: MULTIGROUP CROSS SECTIONS; MULTIGROUP CROSS SECTIONS BASED ON ENDF/B; NEUTRON 
CROSS SECTIONS; RADIOACTIVE DECAY SPECTRA; REACTION CROSS SECTIONS