DLC-0024 ZZ-SINEX.
ZZ SINEX, 100 Neutron-Group Neutron Reaction Cross-Section Library from ENDF/B by SUPERTOG for ANISN
NAME OR DESIGNATION OF PROGRAM, COMPUTER, NATURE OF PHYSICAL PROBLEM SOLVED, METHOD OF SOLUTION, RESTRICTIONS, TYPICAL RUNNING TIME, FEATURES, RELATED AND AUXILIARY PROGRAMS, STATUS, REFERENCES, REQUIREMENTS, LANGUAGE, OPERATING SYSTEM, ANY OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS, NAME AND ESTABLISHMENT OF AUTHOR, MATERIAL, CATEGORIES
[ top ]
[ top ]
To submit a request, click below on the link of the version you wish to order.
Only liaison officers are authorised to submit online requests. Rules for requesters are
available here.
| 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 |
[ top ]
3. NATURE OF PHYSICAL PROBLEM SOLVED
%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%
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.
%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%
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.
%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%
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.
%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%
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.
[ top ]
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.
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.
[ top ]
[ top ]
[ top ]
[ top ]
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.
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.
[ top ]
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).
- 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[ top ]
[ top ]
[ top ]
[ top ]
[ top ]
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 |
Keywords: ENDF/B, cross sections, data, multigroup, neutrons, reaction kinetics.