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DLC-0024 ZZ-SINEX.

ZZ SINEX, 100 Neutron-Group Neutron Reaction Cross-Section Library from ENDF/B by SUPERTOG for ANISN

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1. NAME OR DESIGNATION OF PROGRAM:  ZZ-DLC-24/SINEX. 100-group neutron reaction cross section data generated by SUPERTOG from ENDF/B.
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2. COMPUTERS
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Program name Package id Status Status date
ZZ-SINEX DLC-0024/01 Tested 01-DEC-1974

Machines used:

Package ID Orig. computer Test computer
DLC-0024/01 Many Computers Many Computers
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3. NATURE OF PHYSICAL PROBLEM SOLVED


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  FORMAT: ANISN

NUMBER OF GROUPS: 100 group reaction cross sections for neutron interactions.

  NUCLIDES:
H, D, He, He-3, Li-6, Li-7, Be-9, B-10, B-11, C-12, N-14, O-16, Na-23, Mg, Al-27, Si, Cl, K, Ca, V, Cr, Mn-55, Fe, Co-59, Ni, Cu, Cu-63, Cu-65, Nb, Mo, Ag-107, Ag-109, Xe-135, Cs-133, Sm-149, Eu-151, Eu-153, Gd, Dy-164, Lu-175, Lu-176, Ta-181, Ta-182, W-182, W-183, W-184, W-186, Re-185, Re-187, Au-197, Pb, Th-232, Pa-233, U-234, U-235, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Am-243, and Cm-244.

  ORIGIN: ENDF/B

WEIGHTING SPECTRUM: For the top 99 groups, the explicit assumption  was made that the flux (weighting function) has the shape of a fission spectrum jointed at 0.0674 MeV by a 1/E tail. For the thermal group (group 100), values for all materials except hydrogen  were taken from the Maxwellian average values derived from the ENDF/B data.

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The data can be used in combination with 100 group neutron transport calculations (using, e. g., the DLC-2 library) to determine the spatial distribution of individual reaction rates. In  particular, the retrieval program allows the preparation of dummy materials based on DLC-24 which can be used in the activity calculation option in ANISN to calculate the desired reaction rates.  The library consists of 100 group reaction cross sections for neutron interactions as follows - total, elastic, inelastic, (n,2n), fission, (n,n'alpha), (n,n'3alpha), (n,2nalpha), absorption, (n,n'p), capture, (n,gamma), (n,p), (n,d), (n,t), (n,He3), (n,alpha), (n,2alpha), and nubar. The units are barns, except that nubar is the average number of neutrons per fission event. A table listing the reactions included for each material is found in ref.1.   The nuclides in DLC-24 are those which have been released ascategory I ENDF/B by the National Neutron Cross Section Center, Brookhaven National Laboratory. The library contains data for H, D, He, 3-He, 6-Li, 7-Li, 9-Be, 10-B, 11-B, 12-C, 14-N, 16-O, 23-Na, Mg, 27-Al, Si, Cl, K, Ca, V, Cr, 55-Mn, Fe, 59-Co, Ni, Cu, 63-Cu, 65-Cu, Nb, Mo, 107-Ag, 109-Ag, 135-Xe, 133-Cs, 149-Sm, 151-Eu, 153-Eu, Gd, 164-Dy, 175-Lu, 176-Lu, 181-Ta, 182-Ta, 182-W, 183-W, 184-W, 186-W, 185-Re, 187-Re, 197-Au, Pb, 232-Th, 233-Pa, 234-U, 235-U, 238-U, 238-Pu, 239-Pu, 240-Pu, 241-Pu, 242-Pu, 241-Am, 243-Am, and 244-Cm.
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4. METHOD OF SOLUTION

DLC-24 was generated by SUPERTOG from nuclear data in either point-by-point or parametric representation as specified by ENDF/B. This data is averaged over each specified group width. For the top 99 groups, the explicit assumption was made that the flux (weighting function) has the shape of a fission spectrum jointed at 0.0674 MeV by a 1/E tail. When resonance data were available, resolved and unresolved resonance contributions were calculated, using the infinite dilution approximation. For the thermal group (group 100), values for all materials except hydrogen were taken from the Maxwellian average values derived from the ENDF/B data. The values for hydrogen are more consistent for hydrogen in water. It should be noted that these thermal group values are in some cases different from values used in the latest version of DLC-2.
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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM
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6. TYPICAL RUNNING TIME

To process 124 reactions and produce an unformatted tape for ANISN input requires 10 seconds on the IBM 360/75.
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7. UNUSUAL FEATURES: UNUSUAL FEATURES OF THE PROGRAM
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8. RELATED AND AUXILIARY PROGRAMS

The data retrieval program can be used to list or selectively punch cards or write an unformatted tape in the ANISN cross section formats. The purpose is to arrange the data so they can be read into ANISN as a dummy cross section material which can be used in the ANISN activity calculation to calculate the desired reaction rate distribution in a system of interest.
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9. STATUS
Package ID Status date Status
DLC-0024/01 01-DEC-1974 Tested at NEADB
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10. REFERENCES

- R.Q. Wright and R.W. Roussin:
  'Description of the DLC-24/SINEX 100 Group One-Dimensional Cross
  Sections Based on ENDF/B'
  Informal Notes (February 1973).
- R.Q. Wright:
  'Input Instructions for RESOLVE, a Program for Listing or
  Converting DLC-24/SINEX Data into ANISN Cross Section Input
  Formats'
  Informal Notes (February 1973).
DLC-0024/01, included references:
No printed documentation provided
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11. HARDWARE REQUIREMENTS: MACHINE REQUIREMENTS
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12. PROGRAMMING LANGUAGE(S) USED
No specified programming language
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13. SOFTWARE REQUIREMENTS: OPERATING SYSTEM OR MONITOR UNDER WHICH PROGRAM IS EXECUTED
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14. ANY OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS

The current version of the data library is dated February 1973.
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15. NAME AND ESTABLISHMENT OF AUTHOR

     Oak Ridge National Laboratory
     Oak Ridge, Tennessee, U.S.A.
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16. MATERIAL AVAILABLE
DLC-0024/01
File name File description Records
DLC0024_01.001 CROSS SECTION LIBRARY 9126
DLC0024_01.002 RETRIEVAL PROGRAM (F4) 279
DLC0024_01.003 JCL 4
DLC0024_01.004 SAMPLE PROBLEM DATA 19
DLC0024_01.005 SAMPLE PROBLEM PRINTED OUTPUT 134
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17. CATEGORIES
  • C. Static Design Studies
  • Z. Data.

Keywords: ENDF/B, cross sections, data, multigroup, neutrons, reaction kinetics.