Computer Programs
DLC-0038 ZZ-ORYX-E.
last modified: 01-JUL-1979 | catalog | categories | new | search |

DLC-0038 ZZ-ORYX-E.

ZZ ORYX-E, Group Constant Library from ENDF/B Fission Product Data for ORIGEN Calculation

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1. NAME

ZZ-ORYX-E.

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2. COMPUTERS
To submit a request, click below on the link of the version you wish to order. Rules for end-users are available here.
Program name Package id Status Status date
ZZ-ORYX-E DLC-0038/01 Tested 01-JUL-1979

Machines used:

Package ID Orig. computer Test computer
DLC-0038/01 IBM 360 series IBM 360 series
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3. DESCRIPTION

%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%
FORMAT: ORIGEN

 

NUMBER OF GROUPS: 124 energy groups

 

NUCLIDES: H, He, Li, Be, B, C, N, O, F, Ne, Na, Mg, Al, Si, P, S, Cl, Ar, K, Ca, Sc, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zn, Ga, Ge, As, Se, Br, Kr, Rb, Sr, Y, Zr, Nb, Mo, Tc, Ru, Rh, Pd, Ag, Cd, In, Sn, Sb, Te, I, Xe, Cs, Ba, La, Ce, Pr, Nd, Pm, Sm, Eu, Gd, Tb, Dy, Ho, Er, Tm, Yb, Lu, Hf, Ta, W, Re, Os, Ir, Pt, Au, Hg, Tl, Pb, Bi, Po.

 

ORIGIN: ENDF/B-IV

 

WEIGHTING SPECTRUM: Maxwellian (1/E) fission spectrum with a one percent tolerance.

 

%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%
ORYX-E increases the versatility of the program 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.

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4. METHODS

The data is generated from ENDF/B-IV and is formatted for input to the ORIGEN code. Applications include calculations for waste projection, decay heat, nuclear safeguards, and fuel cycle economics.

The data library is generated from the ENDF/B-IV fission product data. The capture cross section of all fission product nuclides for which capture cross section information is given (about 180 nuclides) were processed into 124 energy groups using MINX. Multigroup cross sections were generated at 0 degrees with infinite dilution and one broad thermal group. Fine group data was generated using a Maxwellian (1/E) fission spectrum with a one percent 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

 

Positron decay

17

11

Cross sections (LMFBR)

*1-group (n,γ) cross sections

181

423*

 

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6. 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.

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8. RELATED OR AUXILIARY PROGRAMS

COLAPS - can be used to collapse the fine cross sections by weighting with a user-specified reference spectrum.
UPDATE - will update the ORIGEN fission product data library using the cross sections generated with COLAPS.
Both utilities are included in the data package.

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9. STATUS
Package ID Status date Status
DLC-0038/01 01-JUL-1979 Tested at NEADB
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10. REFERENCES
DLC-0038/01, included references:
- Master Tape List
- Update of Tritium Data Sets on DLC-38, Informal Notes
- CCC-217/ORIGENS Feedback from User
- G.W. Morrison and C.R. Weisbin (September 1975)
- Energy Structure for the 124 Group Cross Sections
- G.W. Morrison, C.R. Weisbin, 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"  Reprint from Trans. Am. Nucl. Soc..19,
487-488 (1975)
C.W. Kee, C.R. Weisbin, R.E. Schenter;
"Processing and Testing of ENDF/B-IV Fission Product and Transmutation
Data"  Reprint from Trans. Am. Nucl. Soc.., 19
398-399 (1974)
- Documentation for CCC-217/ORIGEN Computer Code Package.
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12. PROGRAMMING LANGUAGE(S) USED
Package ID Computer language
DLC-0038/01 FORTRAN-IV
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15. NAME AND ESTABLISHMENT OF AUTHORS

Contributed by: Radiation Safety Information Computational Center
                Oak Ridge National Laboratory
                Oak Ridge, Tennessee, U. S. A.
Developed by:   Oak Ridge National Laboratory, Oak Ridge, Tennessee, USA

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16. MATERIAL AVAILABLE
DLC-0038/01
File name File description Records
DLC0038_01.001 INFORMATION 5
DLC0038_01.002 LIGHT ELEMENT LIBRARY (KEE; 5/78) EBCDIC 3370
DLC0038_01.003 ORFIP-X (124 GRP X-SECTION; 7/75) EBCDIC 8689
DLC0038_01.004 ORFIP-Y (YIELD&DECAY SCHEME; 1/76) EBCDIC 4105
DLC0038_01.005 NUCLIDE LIB. - STRUCTURAL MATERIALS (5/77) 1265
DLC0038_01.006 NUCLIDE LIB. - ACTINIDES & DECAY PROD.(1/76) 505
DLC0038_01.007 SUPPLEMENTARY ACTINIDES LIB. ADDED(11/74) 505
DLC0038_01.008 FISSION PRODUCTS(PACKAGED 7/73,UPDATED 1/76) 2305
DLC0038_01.009 PHOTON LIBRARY - LIGHT ELEMENTS (7/73) 253
DLC0038_01.010 PHOTON LIBRARY - HEAVY ELEMENTS (7/73) 202
DLC0038_01.011 FISSION PRODUCTS (PHOTON) 461
DLC0038_01.012 COLAPS - SOURCE PROGRAM (F4,EBCDIC) 52
DLC0038_01.013 COLAPS - DD CARDS 8
DLC0038_01.014 COLAPS - SAMPLE PROBLEM INPUT DATA 21
DLC0038_01.015 COLAPS - SAMPLE PROBLEM PRINTED OUTPUT 184
DLC0038_01.016 UPDATE - SOURCE PROGRAM (F4,EBCDIC) 43
DLC0038_01.017 UPDATE - DD CARDS 9
DLC0038_01.018 UPDATE - 'ORFIP-Y' WRITTEN BY UPDATE 4105
DLC0038_01.019 UPDATE - PRINTED OUTPUT 822
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17. CATEGORIES
  • Z. Data

Keywords: ENDF/B, cross sections, data, decay, fission products, multigroup.