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ZZ VITAMINB7/BUGLEB7, Broad-Grp, Fine-Grp, Coupled N/Gamma Cross-Sec Lib derived from ENDF/B-VII.0 Nuclear Data

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Program name Package id Status Status date
ZZ-VITAMINB7/BUGLEB7 DLC-0245/02 Arrived 26-MAR-2012

Machines used:

Package ID Orig. computer Test computer
DLC-0245/02 Many Computers
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NUMBER OF GROUPS: 199 neutron and 42 gamma-ray groups

NUCLIDES: H-1; H-1(H2O); H-1(CH2); H-1(ZrH); H-1(benzine); H-1(liq CH4); H-1(solid CH4); H-1(ortho H); H-1(para H); H-2; H-2(D2O); H-2(ortho D); H-2(para D); H-3; He-3; He-4; Li-6; Li-7; Be-7; Be-9; Be-9(Be metal); Be-9(BeO); B-10; B-11; C; C(benzine); C(graphite); N-14; N-15; O-16; O-16(BeO); O-17; F-19; Na-22; Na-23; Mg; Mg-24; Mg-25; Al-27; Si; Si-28; Si-29; Si-30; P-31; S; S-32; S-33; S-34; S-36; Cl; Cl-35; Cl-37; Ar-36; Ar-38; Ar-40; K; K-39; K-40; K-41; Ca; Ca-40; Ca-42; Ca-43; Ca-44; Ca-46; Ca-48; Sc-45; Ti; Ti-46; Ti-47; Ti-48; Ti-49; Ti-50; V; Cr-50; Cr-52; Cr-53; Cr-54; Mn-55; Fe-54; Fe-56; Fe-57; Fe-58; Co-58m; Ni-58; Ni-59; Ni-60; Ni-61; Ni-62; Ni-64; Cu63; Cu65; Zn; Ga; Ga-69; Ga-71; Ge-70; Ge-72; Ge-73; Ge-74; Ge-76; As-74; As-75; Se-74; Se-76; Se-77; Se-78; Se-79; Se-80; Se-82; Br-79; Br-81; Kr-78; Kr-80; Kr-82; Kr-83; Kr-84; Kr-85; Kr-86; Rb-85; Rb-86; Rb-87; Sr-84; Sr-86; Sr-87; Sr-88; Sr-89; Sr-90; Y-89; Y-90; Y-91; Zr; Zr-90; Zr-91; Zr-92; Zr-93; Zr-94; Zr-95; Zr-96; Nb-93; Nb-94; Nb-95; Mo; Mo-92; Mo-94; Mo-95; Mo-96; Mo-97; Mo-98; Mo-99; Mo-100; Tc-99; Ru-96; Ru-98; Ru-99; Ru-100; Ru-101; Ru-102; Ru-103; Ru-104; Ru-105; Ru-106; Rh-103; Rh-105; Pd-102; Pd-104; Pd-105; Pd-106; Pd-107; Pd-108; Pd-110; Ag-107; Ag-109; Ag-110m; Ag-111; Cd; Cd-106; Cd-108; Cd-110; Cd-111; Cd-112; Cd-113; Cd-114; Cd-115m; Cd-116; In; In-113; In-115; Sn; Sn-112; Sn-113; Sn-114; Sn-115; Sn-116; Sn-117; Sn-118; Sn-119; Sn-120; Sn-122; Sn-123; Sn-124; Sn-125; Sn-126; Sb-121; Sb-123; Sb-124; Sb-125; Sb-126; Te-120; Te-122; Te-123; Te-124; Te-125; Te-126; Te-127m; Te-128; Te-129m; Te-130; Te-132; I-127; I-129; I-130; I-131; I-135; Xe-123; Xe-124; Xe-126; Xe-128; Xe-129; Xe-130; Xe-131; Xe-132; Xe-133; Xe-134; Xe-135; Xe-136; Cs-133; Cs-134; Cs-135; Cs-136; Cs-137; Ba-130; Ba-132; Ba-133; Ba-134; Ba-135; Ba-136; Ba-137; Ba-138; Ba-140; La-138; La-139; La-140; Ce-136; Ce-138; Ce-139; Ce-140; Ce-141; Ce-142; Ce-143; Ce-144; Pr-141; Pr-142; Pr-143; Nd-142; Nd-143; Nd-144; Nd-145; Nd-146; Nd-147; Nd-148; Nd-150; Pm-147; Pm-148; Pm-148m; Pm-149; Pm-151; Sm-144; Sm-147; Sm-148; Sm-149; Sm-150; Sm-151; Sm-152; Sm-153; Sm-154; Eu-151; Eu-152; Eu-153; Eu-154; Eu-155; Eu-156; Eu-157; Gd-152; Gd-153; Gd-154; Gd-155; Gd-156; Gd-157; Gd-158; Gd-160; Tb-159; Tb-160; Dy-156; Dy-158; Dy-160; Dy-161; Dy-162; Dy-163; Dy-164; Ho-165; Ho-166m; Er-162; Er-164; Er-166; Er-167; Er-168; Er-170; Lu-175; Lu-176; Hf-174; Hf-176; Hf-177; Hf-178; Hf-179; Hf-180; Ta-181; Ta-182; W-182; W-183; W-184; W-186; Re-185; Re-187; Ir-191; Ir-193; Au-197; Hg-196; Hg-198; Hg-199; Hg-200; Hg-201; Hg-202; Hg-204; Pb-204; Pb-206; Pb-207; Pb-208; Bi-209; Ra-223; Ra-224; Ra-225; Ra-226; Ac-225; Ac-226; Ac-227; Th-227; Th-228; Th-229; Th-230; Th-232; Th-233; Th-234; Pa-231; Pa-232; Pa-233; U-232; U-233; U-234; U-235; U-236; U-237; U-238; U-239; U-240; U-241; Np-235; Np-236; Np-237; Np-238; Np-239; Pu-236; Pu-237; Pu-238; Pu-239; Pu-240; Pu-241; Pu-242; Pu-243; Pu-244; Pu-246; Am-241; Am-242; Am-242m; Am-243; Am-244; Am-244m; Cm-241; Cm-242; Cm-243; Cm-244; Cm-245; Cm-246; Cm-247; Cm-248; Cm-249; Cm-250; Bk-249; Bk-250; Cf-249; Cf-250; Cf-251; Cf-252; Cf-253; Cf-254; Es-253; Es-254; Es-255; Fm-255


From 10**-5 ev to 0.125 eV   -> Maxwellian Thermal Spectrum
From 0.125 eV to 820.8 keV   -> "1/E" Slowing-Down Spectrum
From 820.8 keV to 20.0 MeV   -> Fission Spectrum



NUMBER OF GROUPS: 47 Neutron, 20 Gamma-Ray groups

NUCLIDES: Ag-107; Ag-109; Al-27; Am-241; Am-242; Am-242m; Am-243; Au-197; B-10; B-11; Ba-138; Be-9; Be-9(Thermal); Bi-209; C; C(Graphite); Ca; Cd-Nat; Cl-Nat; Cm-241; Cm-242; Cm-243; Cm-244; Cm-245; Cm-246; Cm-247; Cm-248; Co-59; Cr-50; Cr-52; Cr-53; Cr-54; Cu-63; Cu-65; Eu-151; Eu-152; Eu-153; Eu-154; Eu-155; F-19; Fe-54; Fe-56; Fe-57; Fe-58; Ga; H-1(H2O); H-1(CH2); H-2(D2O); H-3; He-3; He-4; Hf-174; Hf-176; Hf-177; Hf-178; Hf-179; Hf-180; In-Nat; K; Li-6; Li-7; Mg; Mn-55; Mo; N-14; N-15; Na-23; Nb-93; Ni-58; Ni-60; Ni-61; Ni-62; Ni-64; Np-237; Np-238; Np-239; O-16; O-17; P-31; Pa-231; Pa-233; Pb-206; Pb-207; Pb-208; Pu-236; Pu-237; Pu-238; Pu-239; Pu-240; Pu-241; Pu-242; Pu-243; Pu-244; Re-185; Re-187; S; S-32; Si; Sn-Nat; Ta-181; Ta-182; Th-230; Th-232; Ti; U-232; U-233; U-234; U-235; U-236; U-237; U-238; V; W-Nat; W-182; W-183; W-184; W-186; Y-89; Zr; Zr(Zirc-2)


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.

The weighting spectra were generated using the 1D XSDRNPM discrete-ordinates transport code in SCALE.


The fine-group VITAMIN-B6 and broad-group BUGLE-96 coupled cross-section libraries, which are based on ENDF/B-VI.3, have been successfully used for light water reactor (LWR) shielding applications since 1996. The new VITAMIN-B7 and BUGLE-B7 libraries, which were developed for the same type of applications, are based on ENDF/B-VII.0. In addition to using the most recent ENDF release, the new VITAMIN-B7 library provides cross-sections for a substantially increased set of nuclides (393 nuclides and 24 thermal moderators) compared to VITAMIN-B6. The BUGLE-B7 library, which was group collapsed from VITAMIN-B7, maintains the same nuclide ordering as BUGLE-96 to provide compatibility with transport calculations that have used BUGLE-96. Consistent with BUGLE-96, BUGLE-B7 provides data with upscattering included, as well as data using the ANISN "upscatter correction" for calculations that do not apply outer iterations.

The processing methodology used to develop the VITAMIN-B7 and BUGLE-B7 libraries is consistent with ANSI/ANS 6.1.2. The ENDF data were first processed (using the AMPX code system) into the fine-group, pseudo-problem-independent VITAMIN-B7 library, and then group collapsed into the BUGLE-B7 format using representative weighting spectra from important regions of LWR models. The fine-group and broad-group libraries were extensively verified to confirm proper processing of the data. Validation of the libraries was accomplished using a range of benchmark analyses, including pressure vessel dosimetry benchmarks, which have been previously evaluated with VITAMIN-B6 and BUGLE-96. In general, results with the new libraries are in good agreement with experimental data and with calculational results using VITAMIN-B6 and BUGLE-96.

The primary application of the VITAMIN-B7 and BUGLE-B7 libraries is for LWR shielding applications, including pressure vessel fluence calculations. It is expected that the full range of applications will be similar that of previous multigroup cross section development efforts using the VITAMIN concept (generation of fine-group, pseudo problem-independent data). Previous VITAMIN libraries have proven to be very effective for fusion reactor neutronics, LMFBR core physics analysis, radiation effects of nuclear weapons, and light water reactor shielding and dosimetry.
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BUGLE-B7 is designed to be a "drop-in" replacement for BUGLE-96. Thus, it does not contain all nuclides included in VITAMIN-B7. Like BUGLE-96, the new library 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. Other than carbon, vanadium, and zinc, all elemental data in BUGLE-B7 was created by combining the isotopic evaluations from ENDF into elemental data. In addition to the BUGLE-B7 data sets, which were processed using the "ANISN upscatter" approximation to remove the upscatter cross sections in the thermal groups, data sets are provided which retain the upscatter reactions for groups below 5 eV. These data sets are designated as BUGLE-B7T. Nuclides with Z up to 30 (hydrogen through zinc) are given in a P7 Legendre expansion, while P5 data is available for all other nuclides. Several dosimetry and standard response functions are included with the library along with neutron and gamma kerma factors for all nuclides.

The energy group structure of the VITAMIN-B7 library is identical to that of the VITAMIN-B6 library, with 199 neutron groups and 42 gamma groups. The neutron energy group structure has an upper energy limit of 19.64 MeV. The thermal neutron energy range (i.e., the range of groups that include upscatter) has an upper energy boundary of 5.043 eV and includes 36 groups. The neutron groups typically have uniform lethargy widths ranging from 0.025 to 0.25 for energies above 1.445 eV, with additional boundaries to resolve resonance minima that are important for shielding calculations. The gamma energy group structure has an upper energy limit of 30 MeV. The energy range from 14 MeV to 30 MeV is covered by two groups. From 14 MeV to 8 MeV, the group widths are uniform at 2 MeV. From 8 MeV to 11 MeV, there are group boundaries every 500 keV, with some additional boundaries in the vicinity of important source energies. Below 1 MeV the group widths decrease in a nearly monotonic fashion from 200 keV to 10 keV. The most narrow photon group is a 2-keV group that spans the positron annihilation photon energy of 511 keV.
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AMPX-6 Version1.0 (available from RSICC, package id. is P00562MNYCP00).

The AMPX-6 processing system contains more than 100 distinct modules that can be used to perform a wide range of nuclear data processing functions. The specific modules needed to process a cross-section evaluation and generate a pointwise or multigroup cross-section library depend upon the data specified in the nuclide evaluation. For example, a resonance nuclide will require more modules to be executed than a non-resonance nuclide. If the evaluation is a thermal moderator material with S(?,?) data, the AMPX-6 execution sequence will be different because the final cross-section data will be a mixture of data from one evaluation for energies above the thermal cutoff and from another evaluation in the thermal range. The general procedure to generate multigroup neutron cross-section data is outlined in the documentation. The flow chart to produce the coupled part of the library is depicted in the documentation. The ENDF/B-VII.0 cross-section evaluations were downloaded from the National Nuclear Data Center (NNDC) web page. The thermal evaluations for U(UO2) and O(UO2) were excluded because some inconsistencies in the ENDF evaluations prevent the AMPX Y12 module from processing the data.
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Package ID Status date Status
DLC-0245/02 26-MAR-2012 Masterfiled Arrived
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Background References:
J. E. White, et al., Production and Testing of the Revised VITAMIN-B6 Fine-Group and the BUGLE-96 Broad-Group Neutron/Photon Cross-Section Libraries Derived from ENDF/B-VI.3 Nuclear Data, Oak Ridge National Laboratory draft report ORNL-6795, R1, NUREG/CR-6214, Revision 1 (January 1995).

D. T. Ingersoll, et al., Production and Testing of the VITAMIN-B6 Fine-Group and the BUGLE-93 Broad-Group Neutron/Photon Cross-Section Libraries Derived from ENDF/B-VI Nuclear Data, Oak Ridge National Laboratory report ORNL-6795, NUREG/CR-6214 (January 1995).

J. E. White, et al., VITAMIN-B6: A Fine-Group Cross Section Library Based on ENDF/B-VI for Radiation Transport Applications, pp. 733-736 in Proceedings of the International Conference on Nuclear Data for Science and Technology, Gatlinburg, Tennessee, (May 1994).

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).
DLC-0245/02, included references:
- J. M. Risner, et al.:
Production and Testing of the VITAMIN-B7 Fine-Group and BUGLE-B7 Broad-Group
Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear
Data, Oak Ridge National Laboratory, Oak Ridge, TN 37831-6170, report
NUREG/CR-7045; ORNL/TM-2011/12 (2011)
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No specified programming language
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Programs to convert between the different formats are provided in the AMPX-6 code package.
Data are provided in binary as well as in card image format. Binary data are given in big-endian as well as in little-endian format.
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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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vitaminb7.bcd           - directory containing BCD files for all nuclides in
                          Vitamin B7
VitaminB7_bigEndian     - binary distribution of Vitamin B7
                          in big endian format
VitaminB7_littleEndian  - binary distribution of Vitamin B7
                          in little endian format
Bugle7.bcd              - Bugle7 library without upscatter in BCD format
Bugle7T.bcd             - Bugle7 library with upscatter in BCD format
Bugle7_littleEndian     - Bugle7 library without upscatter in
                          binary little endian format
Bugle7T_littleEndian    - Bugle7 library with upscatter in
                          binary little endian format
Bugle7_bigEndian        - Bugle7 library without upscatter in
                          binary big endian format
Bugle7T_bigEndian       - Bugle7 library with upscatter in
                          binary big endian format
LibraryGeneration       - Input files used to make the library
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  • J. Gamma Heating and Shield Design
  • Z. Data

Keywords: ENDF/B-VII.0, cross sections, data library, detector responses, gamma-ray, multigroup, neutron, neutron cross sections.