DLC-0154 ZZ-ANSLV.
ZZ ANSLV, Multigroup Cross Sections Library for ANS Reactor Design Studies
NAME OR DESIGNATION OF PROGRAM, COMPUTER, DESCRIPTION OF PROGRAM OR FUNCTION, METHOD OF SOLUTION, RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM, CPU, 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-ANSLV | DLC-0154/01 | Tested | 24-NOV-2000 |
Machines used:
| Package ID | Orig. computer | Test computer |
|---|---|---|
| DLC-0154/01 | Many Computers | PC Pentium III 500,Linux-based PC |
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3. DESCRIPTION OF PROGRAM OR FUNCTION
%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%
FORMAT: AMPX Master Interface Library format
NUMBER OF GROUPS: Fine Group (99 energy groups) General Purpose Neutron Library
MATERIALS: H, He, Be, B, Graphite, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Kr, Zr, Mo, Tc, Ru, Ag, Cd, Cs, Ce, Pr, Pm, Sm, Eu, Hf, Ta, U, C, F, Cu, Sn, Pb, Rh, I, Xe, Nd, Th, Np, Pu, Am, Cm, Bk, Cf, Es, MAFP, WAFP
TEMPERATURES:
ORIGIN: ENDF/B-V
FORMAT: AMPX Master Interface Library format
NUMBER OF GROUPS: Broad Group (39 energy groups) General Purpose Neutron Library
MATERIALS: H, He, Be, B, Graphite, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Kr, Zr, Mo, Tc, Ru, Ag, Cd, Cs, Ce, Pr, Pm, Sm, Eu, Hf, Ta, U, C, F, Cu, Sn, Pb, Rh, I, Xe, Nd, Th, Np, Pu, Am, Cm, Bk, Cf, Es, MAFP, WAFP
TEMPERATURES:
ORIGIN: ENDF/B-V
FORMAT: AMPX Master Interface Library format
NUMBER OF GROUPS: Gamma-Ray Interaction (GRI) Library in 44-groups
MATERIALS: H, He, Be, B, C, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zr, Mo, Ag, Cd, Xe, Sm, Eu, Hf, Ta, Ir, Pb, Th, U, Pu
TEMPERATURES:
ORIGIN: ENDF/B-V; LENDL-V evaluations for 12 materials
FORMAT: AMPX Master Interface Library format
NUMBER OF GROUPS: Coupled Library containing (CNG) 99-group neutron and 44-group gamma-ray data
MATERIALS: H, Be, B, C, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zr, Mo, Ag, Cd, Eu, Hf, Ta, Pb, Th, U, Pu
TEMPERATURES:
ORIGIN: ENDF/B-V
FORMAT: AMPX Master Interface Library format
NUMBER OF GROUPS: Coupled neutron-gamma (CNG) Library containing 39-group, and 44-group gamma-ray data
MATERIALS: H, Be, B, C, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zr, Mo, Ag, Cd, Eu, Hf, Ta, Pb, Th, U, Pu,
TEMPERATURES:
ORIGIN: ENDF/B-V
WEIGHTING SPECTRUM: Maxwellian 300K + 1/(E*sigma-total) + fission spectrum4 types of boundaries have been used depending isotope and library type (see report)
%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%
Pseudo-problem-independent, multigroup cross section libraries were generated to support the Advanced Neutron source (ANS) reactor design studies. The ANS was a proposed reactor which would be fueled with highly enriched uranium and cooled with heavy water. The libraries, designated ANSL-V (Advanced Neutron Source Cross Section Libraries based on ENDF/B-V) are data based in AMPX master format. Although the ANS project was canceled, the libraries are being released because they may be utilized in other applications.
ANSL-V data are to be used for the subsequent generation of problem-dependent fine- and/or broad-group cross sections for a wide range of applications, such as core and shield analysis, activation analyses after irradiation of certain elements in the reactor environment, and safety analyses. Problem-dependent cross sections can be derived from the ANSL-V data with AMPX modules included in PSR-0315/AMPX-77, PSR-0352/SCAMPI, or CCC-0545/SCALE-4.3 packages. The derived data libraries in either ANISN or AMPX working library format can be used with codes such as KENO, ANISN, XSDRNPM, DORT, TORT, MORSE.
%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%
FORMAT: AMPX Master Interface Library format
NUMBER OF GROUPS: Fine Group (99 energy groups) General Purpose Neutron Library
MATERIALS: H, He, Be, B, Graphite, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Kr, Zr, Mo, Tc, Ru, Ag, Cd, Cs, Ce, Pr, Pm, Sm, Eu, Hf, Ta, U, C, F, Cu, Sn, Pb, Rh, I, Xe, Nd, Th, Np, Pu, Am, Cm, Bk, Cf, Es, MAFP, WAFP
TEMPERATURES:
ORIGIN: ENDF/B-V
FORMAT: AMPX Master Interface Library format
NUMBER OF GROUPS: Broad Group (39 energy groups) General Purpose Neutron Library
MATERIALS: H, He, Be, B, Graphite, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Kr, Zr, Mo, Tc, Ru, Ag, Cd, Cs, Ce, Pr, Pm, Sm, Eu, Hf, Ta, U, C, F, Cu, Sn, Pb, Rh, I, Xe, Nd, Th, Np, Pu, Am, Cm, Bk, Cf, Es, MAFP, WAFP
TEMPERATURES:
ORIGIN: ENDF/B-V
FORMAT: AMPX Master Interface Library format
NUMBER OF GROUPS: Gamma-Ray Interaction (GRI) Library in 44-groups
MATERIALS: H, He, Be, B, C, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zr, Mo, Ag, Cd, Xe, Sm, Eu, Hf, Ta, Ir, Pb, Th, U, Pu
TEMPERATURES:
ORIGIN: ENDF/B-V; LENDL-V evaluations for 12 materials
FORMAT: AMPX Master Interface Library format
NUMBER OF GROUPS: Coupled Library containing (CNG) 99-group neutron and 44-group gamma-ray data
MATERIALS: H, Be, B, C, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zr, Mo, Ag, Cd, Eu, Hf, Ta, Pb, Th, U, Pu
TEMPERATURES:
ORIGIN: ENDF/B-V
FORMAT: AMPX Master Interface Library format
NUMBER OF GROUPS: Coupled neutron-gamma (CNG) Library containing 39-group, and 44-group gamma-ray data
MATERIALS: H, Be, B, C, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zr, Mo, Ag, Cd, Eu, Hf, Ta, Pb, Th, U, Pu,
TEMPERATURES:
ORIGIN: ENDF/B-V
WEIGHTING SPECTRUM: Maxwellian 300K + 1/(E*sigma-total) + fission spectrum4 types of boundaries have been used depending isotope and library type (see report)
%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%
Pseudo-problem-independent, multigroup cross section libraries were generated to support the Advanced Neutron source (ANS) reactor design studies. The ANS was a proposed reactor which would be fueled with highly enriched uranium and cooled with heavy water. The libraries, designated ANSL-V (Advanced Neutron Source Cross Section Libraries based on ENDF/B-V) are data based in AMPX master format. Although the ANS project was canceled, the libraries are being released because they may be utilized in other applications.
ANSL-V data are to be used for the subsequent generation of problem-dependent fine- and/or broad-group cross sections for a wide range of applications, such as core and shield analysis, activation analyses after irradiation of certain elements in the reactor environment, and safety analyses. Problem-dependent cross sections can be derived from the ANSL-V data with AMPX modules included in PSR-0315/AMPX-77, PSR-0352/SCAMPI, or CCC-0545/SCALE-4.3 packages. The derived data libraries in either ANISN or AMPX working library format can be used with codes such as KENO, ANISN, XSDRNPM, DORT, TORT, MORSE.
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4. METHOD OF SOLUTION
ANSL-V consists of the following fine and broad groups:
1. Fine Group (99 energy groups) General Purpose Neutron Library (FGGPN);
2. Broad Group (39 energy groups) General Purpose Neutron Library (BGGPN);
3. Gamma-Ray Interaction (GRI) Library containing data in 44-group gamma-ray structure.
4. Coupled Library containing (CNG) 99-group neutron and 44-group gamma-ray data.
5. Coupled neutron-gamma (CNG) Library containing 39-group, and 44-group gamma-ray data.
Neutron and secondary gamma-ray production data in the ANSL-V library were generated primarily from evaluations in the ENDF/B-V General Purpose Library. Where evaluations for specified materials were not available in the ENDF/B-V library, ANSL-V GPN and SGRP data sets were generated from evaluations from other ENDF-formatted libraries. Gamma-ray interaction data sets were generated from evaluations in the ENDF/B-V Photon Interaction Library (DLC-0099/ZZ-HUGO). Because some important materials in the ENDF/B-V library do not have gamma-ray production files in the evaluation, both neutron and gamma-ray data from the LENDL-V evaluations were used for 12 materials. Even though both the neutron and gamma-ray evaluations for Sn were taken from LENDL-V, they have different identifiers. In support of local projects, ORNL extended the Sn neutron cross sections down to 10**-5 eV and changed the mat number to 8850, while the gamma-ray identifier was unchanged from 7850.
ANSL-V consists of the following fine and broad groups:
1. Fine Group (99 energy groups) General Purpose Neutron Library (FGGPN);
2. Broad Group (39 energy groups) General Purpose Neutron Library (BGGPN);
3. Gamma-Ray Interaction (GRI) Library containing data in 44-group gamma-ray structure.
4. Coupled Library containing (CNG) 99-group neutron and 44-group gamma-ray data.
5. Coupled neutron-gamma (CNG) Library containing 39-group, and 44-group gamma-ray data.
Neutron and secondary gamma-ray production data in the ANSL-V library were generated primarily from evaluations in the ENDF/B-V General Purpose Library. Where evaluations for specified materials were not available in the ENDF/B-V library, ANSL-V GPN and SGRP data sets were generated from evaluations from other ENDF-formatted libraries. Gamma-ray interaction data sets were generated from evaluations in the ENDF/B-V Photon Interaction Library (DLC-0099/ZZ-HUGO). Because some important materials in the ENDF/B-V library do not have gamma-ray production files in the evaluation, both neutron and gamma-ray data from the LENDL-V evaluations were used for 12 materials. Even though both the neutron and gamma-ray evaluations for Sn were taken from LENDL-V, they have different identifiers. In support of local projects, ORNL extended the Sn neutron cross sections down to 10**-5 eV and changed the mat number to 8850, while the gamma-ray identifier was unchanged from 7850.
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8. RELATED AND AUXILIARY PROGRAMS
AIM: Program to convert from BCD to binary mode; available in PSR-0315/AMPX-77.
The AIM module of PSR-0315/AMPX-77, PSR-0352/SCAMPI, or CCC-0545/SCALE-4.3 can be used for mode conversion of the data. Some other AMPX utility modules, which are included in these packages may also be used. Note that previous versions of AMPX and SCALE will not work because of the AMPX master library format changes.
AIM: Program to convert from BCD to binary mode; available in PSR-0315/AMPX-77.
The AIM module of PSR-0315/AMPX-77, PSR-0352/SCAMPI, or CCC-0545/SCALE-4.3 can be used for mode conversion of the data. Some other AMPX utility modules, which are included in these packages may also be used. Note that previous versions of AMPX and SCALE will not work because of the AMPX master library format changes.
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DLC-0154/01, included references:
- R.Q. Wright et al.:ANSL-V: ENDF/B-V Based Multigroup Cross-Section Libraries for
Advanced Neutron Source Reactor Studies Supplement 1
ORNL/6618/s1 (August 1995)
- W.E. Ford III et al.:
ANSL-V: ENDF/B-V Based Multigroup Cross-Section Libraries for
Advanced Neutron Source Reactor Studies Supplement 1
ORNL/6618 (September 1990)
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DLC-0154/01
miscellaneous mag tapeANSLV.EXE Sefl-extracting DOS file MISTPmiscellaneous mag tapeD154TAR1.Z Unix compressed tar file MISTP
test-case data mag tape39n.bcd Broad group (39 energy groups) DATTP
test-case data mag tape39n44g.bcd Coupled lib-39 n,44 g energy gr.DATTP
test-case data mag tape44g.bcd 44 groupe gamma-ray interaction libDATTP
test-case data mag tape99n.bcd Fine group (99 energy groups) DATTP
test-case data mag tape99n44g.bcd Coupled lib-99 n,44 g energy gr.DATTP
test-case data mag tapeaim39n.inp Sample input to aim & rade DATTP
miscellaneous mag tapeD154TAR1.LIS List of files MISTP
miscellaneous mag tapeDLC154_0.LIS List of files MISTP
miscellaneous mag tapeREADME.CD Information file MISTP
miscellaneous mag tapeREADME.RSI Information file MISTP
report ORNL-6618/s1 (August 1995) REPPT
report ORNL-6618 (September 1990) REPPT
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- B. Spectrum Calculations, Generation of Group Constants and Cell Problems
- Z. Data.
Keywords: activation analysis, data library, group constants, heavy water reactor, multigroup, reactor safety, shielding.