last modified: 30-AUG-1996 | catalog | categories | new | search |

DLC-0185 ZZ-BUGLE-96.

ZZ BUGLE-96, Multigroup Coupled Neutron Gamma Cross-Section for LWR Shielding Calculation

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1. NAME OR DESIGNATION OF PROGRAM:  ZZ-BUGLE-96.
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2. COMPUTERS
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-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.

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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.
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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM:
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6. TYPICAL RUNNING TIME:  Not applicable.
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.
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7. UNUSUAL FEATURES OF THE PROGRAM:
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8. RELATED AND AUXILIARY PROGRAMS:  BCBN: Convert ANISN card-image data to binary format.
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9. STATUS
Package ID Status date Status
DLC-0185/01 30-AUG-1996 Tested at NEADB
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10. REFERENCES:
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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11. MACHINE REQUIREMENTS:
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12. PROGRAMMING LANGUAGE(S) USED
Package ID Computer language
DLC-0185/01 FORTRAN-77
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13. OPERATING SYSTEM UNDER WHICH PROGRAM IS EXECUTED:
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.
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14. OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS:
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15. NAME AND ESTABLISHMENT OF AUTHORS

Contributed by: Radiation Safety Information Computational Center
                Oak Ridge National Laboratory
                Oak Ridge, Tennessee, U. S. A.

Developed by:   Oak Ridge National Laboratory
                Oak Ridge, Tennessee, U. S. A.
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16. MATERIAL AVAILABLE
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
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17. CATEGORIES
  • J. Gamma Heating and Shield Design
  • Z. Data.

Keywords: LWR reactors, data library, multigroup, pressure vessels, pwr reactors, self-shielding, shielding.