DLC-0185 ZZ-BUGLE-96.
ZZ BUGLE-96, Multigroup Coupled Neutron Gamma Cross-Section for LWR Shielding Calculation
NAME OR DESIGNATION OF PROGRAM, COMPUTER, DESCRIPTION OF PROGRAM OR FUNCTION, METHOD OF SOLUTION, RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM, TYPICAL RUNNING TIME, UNUSUAL FEATURES OF THE PROGRAM, RELATED AND AUXILIARY PROGRAMS, STATUS, REFERENCES, MACHINE REQUIREMENTS, LANGUAGE, OPERATING SYSTEM UNDER WHICH PROGRAM IS EXECUTED, OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS, NAME AND ESTABLISHMENT OF AUTHORS, MATERIAL, CATEGORIES
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| Program name | Package id | Status | Status date |
|---|---|---|---|
| ZZ-BUGLE-96 | DLC-0185/01 | Tested | 30-AUG-1996 |
Machines used:
| Package ID | Orig. computer | Test computer |
|---|---|---|
| DLC-0185/01 | Many Computers | VAX under VMS |
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3. DESCRIPTION OF PROGRAM OR FUNCTION
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FORMAT: ANISN
NUMBER OF GROUPS: 47 Neutron, 20 Gamma-Ray
NUCLIDES: Ag, Al, Am, Au, B, Ba, Be, Bi, C, Ca, Cd, Cl, Cm, Co, Cr, Cu, Eu, F, Fe, Ga, H, He, Hf, In, K, Li, Mg, Mn, Mo, N, Na, Nb, Ni, Np, O, P, Pa, Pb, Pu, Re, S, Si, Sn, Ta, Th, Ti, U, V, W, Y, Zr.
ORIGIN: ENDF/B-IV data
WEIGHTING SPECTRUM: The concrete-spectrum-weighted cross sections have been shown to be generally applicable to a wide range of shielding problems. Flux spectra from five specific locations were used, corresponding to:
1) off-center in a BWR core region,
2) off-center in a PWR core region,
3) the downcomer region in a PWR model,
4) within the pressure vessel at a depth of one-fourth the total thickness, and
5) within the concrete shield surrounding a PWR reactor vessel.
%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%
A new multigroup cross-section library based on ENDF/B-VI data has been produced and tested for light water reactor shielding and reactor pressure vessel dosimetry applications. The broad-group library, designated DLC-185/BUGLE-96, is intended to replace the DLC-75/BUGLE-80 and DLC-76/SAILOR libraries, which are both based on ENDF/B-IV data. It also replaces the DLC-175/BUGLE-93 library by correcting some deficiencies and adding several additional data sets. The processing methodology for BUGLE-96 is consistent with ANSI/ANS 6.1.2, since the ENDF data were first processed into a fine-group, pseudo-problem-independent format and then collapsed into the final broad group format. An extensive integral data testing effort was performed to qualify the data and to assess its impact on LWR shielding applications. In general, results using the new data show significant improvements relative to earlier ENDF data.
The BUGLE-96 cross sections are intended for use in LWR shielding and pressure vessel dosimetry applications. The multigroup data have been collapsed, and in some cases self-shielded, using flux spectra typical of PWR and BWR reactor models. Flux spectra from five specific locations within these models were used, corresponding to:
1) off-center in a BWR core region,
2) off-center in a PWR core region,
3) the downcomer region in a PWR model,
4) within the pressure vessel at a depth of one-fourth the total thickness, and
5) within the concrete shield surrounding a PWR reactor vessel. The concrete-spectrum-weighted cross sections have been shown to be generally applicable to a wide range of shielding problems.
%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%
FORMAT: ANISN
NUMBER OF GROUPS: 47 Neutron, 20 Gamma-Ray
NUCLIDES: Ag, Al, Am, Au, B, Ba, Be, Bi, C, Ca, Cd, Cl, Cm, Co, Cr, Cu, Eu, F, Fe, Ga, H, He, Hf, In, K, Li, Mg, Mn, Mo, N, Na, Nb, Ni, Np, O, P, Pa, Pb, Pu, Re, S, Si, Sn, Ta, Th, Ti, U, V, W, Y, Zr.
ORIGIN: ENDF/B-IV data
WEIGHTING SPECTRUM: The concrete-spectrum-weighted cross sections have been shown to be generally applicable to a wide range of shielding problems. Flux spectra from five specific locations were used, corresponding to:
1) off-center in a BWR core region,
2) off-center in a PWR core region,
3) the downcomer region in a PWR model,
4) within the pressure vessel at a depth of one-fourth the total thickness, and
5) within the concrete shield surrounding a PWR reactor vessel.
%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%
A new multigroup cross-section library based on ENDF/B-VI data has been produced and tested for light water reactor shielding and reactor pressure vessel dosimetry applications. The broad-group library, designated DLC-185/BUGLE-96, is intended to replace the DLC-75/BUGLE-80 and DLC-76/SAILOR libraries, which are both based on ENDF/B-IV data. It also replaces the DLC-175/BUGLE-93 library by correcting some deficiencies and adding several additional data sets. The processing methodology for BUGLE-96 is consistent with ANSI/ANS 6.1.2, since the ENDF data were first processed into a fine-group, pseudo-problem-independent format and then collapsed into the final broad group format. An extensive integral data testing effort was performed to qualify the data and to assess its impact on LWR shielding applications. In general, results using the new data show significant improvements relative to earlier ENDF data.
The BUGLE-96 cross sections are intended for use in LWR shielding and pressure vessel dosimetry applications. The multigroup data have been collapsed, and in some cases self-shielded, using flux spectra typical of PWR and BWR reactor models. Flux spectra from five specific locations within these models were used, corresponding to:
1) off-center in a BWR core region,
2) off-center in a PWR core region,
3) the downcomer region in a PWR model,
4) within the pressure vessel at a depth of one-fourth the total thickness, and
5) within the concrete shield surrounding a PWR reactor vessel. The concrete-spectrum-weighted cross sections have been shown to be generally applicable to a wide range of shielding problems.
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4. METHOD OF SOLUTION
BUGLE-96 contains 120 nuclides which have been processed as infinitely dilute and collapsed using an LWR concrete shield spectrum. Additionally, it contains 105 nuclides which have been energy self-shielded and collapsed using LWR-specific material compositions and flux spectra. A separate set of data are included which combines the isotopic BUGLE-96 nuclides into natural elements and provides them with material identifiers which are identical to the original DLC-76/SAILOR library. In addition to the BUGLE-96 and SAILOR-96 data sets, which have been processed without upscatter in the thermal groups, two new data sets are provided which retain the upscatter reactions for groups below 5eV. These data sets are designated as BULGE-96T and SAILOR-96T. Nuclides with Z < 30 (hydrogen through copper) are given in a P7 Legendre expansion while P5 expansion is available for all other nuclides. Several dosimetry and standard response functions are included with the library along with kerma factors for all nuclides. The library was collapsed from the VITAMIN-B6 fine-group library using the AMPX-77 processing code system. VITAMIN-B6 is derived from ENDF/B-VI nuclear data, except for two nuclides (Sn obtained from LENDL and Zirc2 obtained from ENDF/V-IV). The responses and kerma factors were also derived primarily from ENDF/B-VI.
BUGLE-96 contains 120 nuclides which have been processed as infinitely dilute and collapsed using an LWR concrete shield spectrum. Additionally, it contains 105 nuclides which have been energy self-shielded and collapsed using LWR-specific material compositions and flux spectra. A separate set of data are included which combines the isotopic BUGLE-96 nuclides into natural elements and provides them with material identifiers which are identical to the original DLC-76/SAILOR library. In addition to the BUGLE-96 and SAILOR-96 data sets, which have been processed without upscatter in the thermal groups, two new data sets are provided which retain the upscatter reactions for groups below 5eV. These data sets are designated as BULGE-96T and SAILOR-96T. Nuclides with Z < 30 (hydrogen through copper) are given in a P7 Legendre expansion while P5 expansion is available for all other nuclides. Several dosimetry and standard response functions are included with the library along with kerma factors for all nuclides. The library was collapsed from the VITAMIN-B6 fine-group library using the AMPX-77 processing code system. VITAMIN-B6 is derived from ENDF/B-VI nuclear data, except for two nuclides (Sn obtained from LENDL and Zirc2 obtained from ENDF/V-IV). The responses and kerma factors were also derived primarily from ENDF/B-VI.
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DLC-0185/01
NEA-DB ran the ASCII-to-binary conversion program on a VAXStation 3100 M38. Conversion times varied from 30 seconds to 5 minutes, depending on the data file processed.[ top ]
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DLC-0185/01, included references:
- RSIC "READ.ME" (March 28, 1996)- J.E. White et al.:
BUGLE-96: A Revised Multigroup Cross Section Library for LWR
Applications Based on ENDF/B-VI Release 3 (presented at the
American Nuclear Society Radiation Protection & Shielding
Topical Meeting, April 21-25, 1996, Falmouth, MA) (April 1996)
- D.T. Ingersoll et al.:
Production and Testing of the VITAMIN-B6 Fine-Group and the
BUGLE-93 Broad-Group Neutron/Photon Cross-Sections Libraries
Derived from ENDF/B-VI Nuclear Data
ORNL-6795, NUREG/CR-6214 (January 1995)
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DLC-0185/01
NEA-DB installed the conversion program on a VAXStation 3100 M38 running under VAX/VMS version 6.1. The source file was compiled with the VAX Fortran compiler version 6.2-108.[ top ]
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DLC-0185/01
| File name | File description | Records |
|---|---|---|
| DLC0185_01.001 | Information file | 172 |
| DLC0185_01.002 | ANISN identifiers for BUGLE96 and BUGLE96T | 126 |
| DLC0185_01.003 | ANISN id. for files with PWR & BWR weighting | 117 |
| DLC0185_01.004 | Fortran 77 retrieval code | 46 |
| DLC0185_01.005 | Total kerma factor for 120 nuclides (n+phot) | 1135 |
| DLC0185_01.006 | Neutron response function | 995 |
| DLC0185_01.007 | ANISN id for response function | 101 |
| DLC0185_01.008 | Neutron response functions in fine group | 2150 |
| DLC0185_01.009 | Infinitely dilute X-sections for 120 nucl. | 100804 |
| DLC0185_01.010 | Self-shielded X-sections for 10 nuclides | 10240 |
| DLC0185_01.011 | Self-shielded X-sections for 13 nuclides | 13502 |
| DLC0185_01.012 | Self-shielded X-sections for 20 nuclides | 20000 |
| DLC0185_01.013 | Self-shielded X-sections for 17 nuclides | 16903 |
| DLC0185_01.014 | Self-shielded X-sections for 15 nuclides | 14465 |
| DLC0185_01.015 | Self-shielded X-sections for 30 nuclides | 28929 |
| DLC0185_01.016 | Infinitely dilute X-sections for 120 nucl. | 102772 |
| DLC0185_01.017 | Self-shielded X-sections for 10 nuclides | 10442 |
| DLC0185_01.018 | Self-shielded X-sections for 13 nuclides | 13737 |
| DLC0185_01.019 | Self-shielded X-sections for 20 nuclides | 20401 |
| DLC0185_01.020 | Self-shielded X-sections for 17 nuclides | 17198 |
| DLC0185_01.021 | Self-shielded X-sections for 15 nuclides | 14764 |
| DLC0185_01.022 | Self-shielded X-sections for 30 nuclides | 29527 |
| DLC0185_01.023 | ANISN id for the SAILOR-like datasets | 46 |
| DLC0185_01.024 | Same as BUGLE96 with ANISN identifiers | 25475 |
| DLC0185_01.025 | Same as BUGLE96T with ANISN identifiers | 26006 |
Keywords: LWR reactors, data library, multigroup, pressure vessels, pwr reactors, self-shielding, shielding.