DLC-0023 ZZ-DLC-23E/CASK.
ZZ DLC-23F CASK, 40-Group Neutron and Gamma Coupled Cross-Section for PWR Shipping Casks
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, AUTHOR, MATERIAL, CATEGORIES
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| Program name | Package id | Status | Status date |
|---|---|---|---|
| ZZ-CASK/F | DLC-0023/03 | Tested | 27-APR-1983 |
Machines used:
| Package ID | Orig. computer | Test computer |
|---|---|---|
| DLC-0023/03 | IBM 370/168 | IBM 370/168 |
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3. NATURE OF PHYSICAL PROBLEM SOLVED
FORMAT: ANISN, DOT, MORSE
NUMBER OF GROUPS: 40 group set.
NUCLIDES: H, Be, B-10, C, N, O, Na, Mg, Al, Si, K, Ca, Ti, Cr, Mn, Fe, Ni, Cu, Mo, Ta, W, Pb, U-235, U-238, Pu-239, and Pu-240.
ORIGIN: ENDF/B-2 library, although some data were taken from other sources.
WEIGHTING SPECTRUM: Representative of a water-uranium mixture.
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The data were designed for use in shielding analysis of PWR depleted uranium shipping casks. Results of such an analysis are given in ref. 2. The data were collapsed from a fine group structure using a weighting function representative of a water-uranium mixture (3). Thus, the application of this data for problems not similar to the shipping cask type should be done with caution.
The library was compiled for several elements that are commonly used for shielding calculations. The coupled P3 cross sections are given in the ANISN format which permits their usage in the discrete ordinates codes ANISN and DOT as well as the 3-dimensional Monte Carlo code MORSE. The data sets from which DLC-23 are derived are listed in ref.3.
The library contains data for H, Be, 10-B, C, N, O, Na, Mg, Al, Si, K, Ca, Ti, Cr, Mn, Fe, Ni, Cu, Mo, Ta, W, Pb, 235-U, 238-U, 239-Pu, and 240-Pu.
FORMAT: ANISN, DOT, MORSE
NUMBER OF GROUPS: 40 group set.
NUCLIDES: H, Be, B-10, C, N, O, Na, Mg, Al, Si, K, Ca, Ti, Cr, Mn, Fe, Ni, Cu, Mo, Ta, W, Pb, U-235, U-238, Pu-239, and Pu-240.
ORIGIN: ENDF/B-2 library, although some data were taken from other sources.
WEIGHTING SPECTRUM: Representative of a water-uranium mixture.
%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%
The data were designed for use in shielding analysis of PWR depleted uranium shipping casks. Results of such an analysis are given in ref. 2. The data were collapsed from a fine group structure using a weighting function representative of a water-uranium mixture (3). Thus, the application of this data for problems not similar to the shipping cask type should be done with caution.
The library was compiled for several elements that are commonly used for shielding calculations. The coupled P3 cross sections are given in the ANISN format which permits their usage in the discrete ordinates codes ANISN and DOT as well as the 3-dimensional Monte Carlo code MORSE. The data sets from which DLC-23 are derived are listed in ref.3.
The library contains data for H, Be, 10-B, C, N, O, Na, Mg, Al, Si, K, Ca, Ti, Cr, Mn, Fe, Ni, Cu, Mo, Ta, W, Pb, 235-U, 238-U, 239-Pu, and 240-Pu.
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4. METHOD OF SOLUTION
The source for the neutron cross sections was primarily the ENDF/B-2 library, although some data were taken from other sources, when necessary, as indicated in ref.3.
The SUPERTOG code was used to generate resonance corrected fine group cross sections for 104 energy groups from ENDF/B library.
Single-level Breit-Wigner or multi-level Breit-Wigner resonance parameters were used by SUPERTOG to generate point cross sections for the resonance nuclides. Approximately 100 points per resolved resonance were used to integrate the point cross sections for the fine groups. In the unresolved resonance region 81 points per fine group were used for the integration. A 1/E spectral weighting function was used.
The multigroup neutron cross sections in a 22 energy group structure were obtained from the 104 group cross sections by averaging the various elemental cross sections across a fine group flux calculated by ANISN for a uranium-water mixture using fine group cross sections. This weighting function is given in ref.3.
The secondary gamma-ray production cross sections were calculated for an 18 group gamma energy structure by the MUG code. The multigroup neutron cross sections, the secondary gamma production cross sections, and the gamma-ray transport cross sections were coupled to form a 40 group set. This is the same 40 group structure as used by Straker for various shielding calculations and it is tabulated in ref.3.
Calculations of the neutron and gamma-ray fluence from several shielding problems have been performed and some results of these calculations are discussed in ref.2.
The source for the neutron cross sections was primarily the ENDF/B-2 library, although some data were taken from other sources, when necessary, as indicated in ref.3.
The SUPERTOG code was used to generate resonance corrected fine group cross sections for 104 energy groups from ENDF/B library.
Single-level Breit-Wigner or multi-level Breit-Wigner resonance parameters were used by SUPERTOG to generate point cross sections for the resonance nuclides. Approximately 100 points per resolved resonance were used to integrate the point cross sections for the fine groups. In the unresolved resonance region 81 points per fine group were used for the integration. A 1/E spectral weighting function was used.
The multigroup neutron cross sections in a 22 energy group structure were obtained from the 104 group cross sections by averaging the various elemental cross sections across a fine group flux calculated by ANISN for a uranium-water mixture using fine group cross sections. This weighting function is given in ref.3.
The secondary gamma-ray production cross sections were calculated for an 18 group gamma energy structure by the MUG code. The multigroup neutron cross sections, the secondary gamma production cross sections, and the gamma-ray transport cross sections were coupled to form a 40 group set. This is the same 40 group structure as used by Straker for various shielding calculations and it is tabulated in ref.3.
Calculations of the neutron and gamma-ray fluence from several shielding problems have been performed and some results of these calculations are discussed in ref.2.
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10. REFERENCES
- G.W. Morrison, E.A. Straker, and R.H. Odegaarden:
'A Coupled Neutron and Gamma-Ray Multigroup Cross Section Library for Use in Shielding Calculations'
Trans. Am. Nuc. Soc., 15, 535 (1972).
- G.W. Morrison, E.A. Straker, and R.H. Odegaarden:
'The Use of the MORSE Monte Carlo Code to Solve Shielding and Criticality Problems of Spent Fuel Casks'
Trans. Am. Nucl. Soc., 15, 547 (1972).
- R.W. Roussin and J.B. Wright:
'Contents, Energy Group Structure, and Weighting Function Used for DLC-23/Cask'
Informal Notes (1972).
- G.W. Morrison, E.A. Straker, and R.H. Odegaarden:
'A Coupled Neutron and Gamma-Ray Multigroup Cross Section Library for Use in Shielding Calculations'
Trans. Am. Nuc. Soc., 15, 535 (1972).
- G.W. Morrison, E.A. Straker, and R.H. Odegaarden:
'The Use of the MORSE Monte Carlo Code to Solve Shielding and Criticality Problems of Spent Fuel Casks'
Trans. Am. Nucl. Soc., 15, 547 (1972).
- R.W. Roussin and J.B. Wright:
'Contents, Energy Group Structure, and Weighting Function Used for DLC-23/Cask'
Informal Notes (1972).
DLC-0023/03, included references:
- CASK-81 - 22 Neutron, 18 Gamma-Ray Group, P3, Cross Sectionsfor Shipping Cask Analysis. ORNL DLC-23 Updated (June 16, 1981).
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DLC-0023/03
| File name | File description | Records |
|---|---|---|
| DLC0023_03.003 | ZZ-DLC-23E/CASK INFORMATION FILE | 45 |
| DLC0023_03.004 | DATA LIBRARY FOR ALL MATERIALS | 6735 |
| DLC0023_03.005 | AUXILIARY PROGRAM SOURCE | 140 |
| DLC0023_03.006 | JCL TO RUN AUXILIARY PROGRAM | 11 |
| DLC0023_03.007 | INPUT FOR TEST CASE | 1 |
| DLC0023_03.008 | DATA LIBRARY FOR ZR | 268 |
| DLC0023_03.009 | OUTPUT OF TEST CASE | 4 |
Keywords: containers, cross sections, data, gamma radiation, multigroup, neutrons, shielding.