|Program name||Package id||Status||Status date|
|Package ID||Orig. computer||Test computer|
|DLC-0002/01||IBM 370 series||IBM 370 series|
FORMAT: ANISN, DOT or DTF-4
NUMBER OF GROUPS: 100
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, 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: The nuclides in DLC-2 are those which have been released as category I ENDF/B by the National Neutron Cross Section Center, Brookhaven National Laboratory.
WEIGHTING SPECTRUM: The explicit assumption was made that the flux has the shape of a fission spectrum joined at 0.0674 MeV by a 1/E tail.
Neutron transport calculations can be performed with DLC-2 data. Since the data are intended for use in multigroup discrete-ordinates or Monte Carlo transport codes which treat anisotropic scattering, possible cross section angular expansion is limited only by the options available in the particular code used. Specifically, the retrieval program manipulates DLC-2 such that it conforms to input requirements of the ANISN, DOT, or DTF-4 codes, or any computer code using data in the ANISN or DTF-4 format.
The nuclides in DLC-2 are those which have been released as category 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, 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.
DLC-2 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. The explicit assumption was made that the flux (weighting function) has the shape of a fission spectrum joined 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. DLC-2 consists of fine group constants such as one-dimensional reaction arrays (absorption, fission, etc.), PN elastic scattering matrices, and inelastic and (n,2n) scattering matrices which were generated, combined and written on tape as card images in the ANISN format. The units are barns rather than cm2.
DLC-2 represents a P8 approximation to elastic scattering angular distributions. The data have a 100-group structure with energy group boundaries identical to those in the GAM-2 library, with a group 1 upper boundary energy of 14.92 MeV and a group 99 lower energy of 0.414 eV. The group-to-group transfer matrices reflect only downscatter in energy, and group 100 serves as a thermal group. Cross section values for the thermal group were selected as described in ref. 3. as noted therein, the user should exercise caution in interpreting results for the thermal group.
Using the DLC2RP retrieval program to produce an unformatted tape for use by ANISN, containing elements hydrogen and oxygen for P8 expansion requires approximately 2 minutes on the IBM 360/65.
To compile APRFX-1 and collapse 100 group P1 cross sections to 7 groups using generated spectra requires 35 seconds on the IBM 360/91
DLC2RP will retrieve DLC-2 data from a maximum of 46 data sets and merge them into one data set. The program will then, by input option, edit the data, punch cards in either the ANISN or DTF-4 format, or write an unformatted tape for use by ANISN. The program was written by the authors of the SUPERTOG program.
APRFX-1 collapses the fine group cross sections to a broad group structure according to a flux spectrum either input by the user or calculated by the code. The code will average the fine group cross sections to form either macroscopic or microscopic isotope cross sections and any combination of macroscopic mixtures of these cross sections on the same problem. It also determines the broad group input source and generates averaged neutron velocities for use with transport calculations.
|Package ID||Status date||Status|
|DLC-0002/01||04-SEP-1981||Tested at NEADB|
R.Q. Wright: 'User's Manual for DLC-2 Data Retrieval Program' Informal Notes (July 1972).
R.W. Roussin: 'Comments on ANISN and DTF-4 Format' Informal Notes (1969).
R.Q. Wright: 'Values of Thermal Cross Sections Used for DLC-2D' Informal Notes (1972).
R.Q. Wright, N.M. Greene, J.L. Lucius, and C.W. Craven: 'SUPERTOG, A Program to Generate Fine Group Constants and PN Scattering Matrices from ENDF/B' ORNL-TM-2679 (Sept. 1969).
R.W. Roussin: 'Using ANISN to Reduce the DLC-2 100-Group Cross Section Data to a Smaller Number of Groups' ORNL-TM-3049 (May 1969).
J.B. Wright and R.W. Roussin: 'Comments on Reactions Included in the SUPERTOG Procedure and a Compilation of File 1 Information Listing References for the ENDF/B Data ' Informal Notes (1972).
P.S. Pickard and D.O. Williams: 'Calculated Neutron Energy Spectra for the APRF Reactor' Memo for record AMXRD-BNL (September 1970).
P.S. Pickard: 'APRFX-I, Neutron Cross Section Collapsing Code' Memo for record, AMXRD-BNL (December 1970).
|Package ID||Computer language|
|File name||File description||Records|
|DLC0002_01.008||RETRIEVAL PROGRAM SOURCE||244|
|DLC0002_01.011||OUTPUT OF APRFX-I||869|
Keywords: ENDF/B, absorption, angular distribution, cross sections, data, elastic scattering, fission, inelastic scattering, multigroup, neutron transport theory.