Computer Programs

NAME OR DESIGNATION, COMPUTER, DESCRIPTION, METHODS, RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM, CPU, FEATURES, RELATED OR AUXILIARY PROGRAMS, STATUS, REFERENCES, REQUIREMENTS, LANGUAGE, OPERATING SYSTEM, OTHER RESTRICTIONS, NAME AND ESTABLISHMENT OF AUTHORS, MATERIAL, CATEGORIES

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1. NAME OR DESIGNATION: ZZ-WIMS-D/4LIB contains:

1. ZZ-WIMSUT01/61;

2. ZZ-WIMSUT01/64;

3. ZZ-WIMSUT01/69;

4. ZZ-WIMSUT01/76;

5. ZZ-WIMSUT01/79;

Five WIMS-D/4 libraries in 61,64,69,76 and 79 energy groups, which have the same number of fast and thermal groups but different in resonance group, namely 5, 8, 13, 20, 23. with the same format as for the standard WIMS library. The files were processed from ENDF/B6 using a program developed by the library originators.

1. ZZ-WIMSUT01/61;

2. ZZ-WIMSUT01/64;

3. ZZ-WIMSUT01/69;

4. ZZ-WIMSUT01/76;

5. ZZ-WIMSUT01/79;

Five WIMS-D/4 libraries in 61,64,69,76 and 79 energy groups, which have the same number of fast and thermal groups but different in resonance group, namely 5, 8, 13, 20, 23. with the same format as for the standard WIMS library. The files were processed from ENDF/B6 using a program developed by the library originators.

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To submit a request, click below on the link of the version you wish to order. Rules for end-users are
available here.

Program name | Package id | Status | Status date |
---|---|---|---|

ZZ-WIMS-D/4LIB | IAEA1397/01 | Arrived | 16-OCT-2002 |

Machines used:

Package ID | Orig. computer | Test computer |
---|---|---|

IAEA1397/01 | Many Computers |

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3. DESCRIPTION

%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%

FORMAT: WIMS

NUMBER OF GROUPS: 61,64,69,76,79 energy groups

NUCLIDES:

The following nuclides are included:

H-1, H-1 (in H2O), H-1 (in ZrH), H-2, H-2 (in D2O), He-3, He-4, Li-6, Li-7, Be-9, B-10, B-11, C, C (in graphite), N-14, 0-16, F, Na, Al, Si, P-31, S-32, K, Ti, V, Cr, Mn-55, Fe, Co-59, Ni, Cu, Cu-63, Zr, Nb-93, Mo, Mo-92, Mo-94, Mo-95, Mo-96, Mo-97, Mo-98, Mo-99, Mo-100, Ag-107, Ag-109, Cd, In-113, In-115, Sn, Gd, Dy-164, Er-166, Er-167, Lu-176, Hf, Hf-174, Hf-176, Hf-177, Hf-178, Hf-179, Hf-180, Ta-181, Au-197, Pb.

The following fission products are included:

Kr-83, Mo-95, Tc-99, Ru-101, Ru-103, Rh-103, Rh-105, Pd-105, Pd-108, Ag-109, Cd-113, In-115, I-127, I-135, Xe-131, Xe-135, Cs-133, Cs-134, Cs-135, Nd-143, Nd-145, Pm-147, Pm-148g, Pm-148m, Pm-149, Sm-147, Sm-148, Sm-149, Sm-150, Sm-151, Sm-152, Eu-153, Eu-154, Eu-155, Gd-157, Pseudo F.P., Dummy ID = 237 Abs. XS = 1.0E-5 B

The following actinides are included:

Th-232, Pa-233, U-233, U234, U-235, U-236, U-237, U-238, Np-237, Np-238, Np-239, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Am-242g, Am-242m, Am-243, Cm-242, Cm243, Cm-244, !/v absorber ID = 1000.

ORIGIN: ENDF/B-V or IV and JENDL-2(Rev.1); and from ENDL-84, where data was not available in ENDF/B.

Neutron cross section library for thermal reactor design analysis with 61, 64, 69, 76, and 79 energy WIMS groups structure.

Group Energy Structure for resonance region:

WIMSUT61 WIMSUT64 WIMSUT69

9118.00-1942.40 9118.00-3522.40 9118.00-5530.00

1942.40-413.770 3522.40-1344.00 5530.00-3319.10

413.770-88.1450 1344.00-526.000 3319.10-2339.45

88.1450-18.7770 526.000-201.090 2339.45-1425.10

18.7770-4.00000 201.090-76.500 1425.10-906.898

76.5000-28.1270 906.898-367.262

28.1270-10.700 367.262-148.728

10.7000-4.00000 148.728-75.5014

75.5014-48.0520

48.0520-27.7000

27.7000-15.9680

15.9680-9.96800

9.96800-4.00000

WIMSUT76 WIMSUT79

9118.00-6724.90 9118.00-6692.50

6724.90-4960.00 6692.50-4112.20

4960.00-3658.30 4912.20-3605.40

3658.30-2698.20 3605.40-2646.30

2698.20-1912.70 2646.30-1945.00

1912.70-1355.70 1945.00-1423.00

1355.70-961.150 1423.00-1043.00

961.150-681.350 1043.00-783.000

681.350-483.000 783.000-561.000

483.000-340.250 561.000-414.30

340.250-241.200 414.300-304.00

241.200-179.100 304.00-222.910

179.10-126.9600 222.910-161.000

126.960-90.000 161.000-124.000

90.0000-63.8000 124.000-86.500

63.80000-44.800 86.500-53.200

44.80000-28.250 53.200-47.4860

28.25000-13.554 47.486-29.560

13.55400-9.4400 29.560-12.500

.9.440000-4.0000 12.500-10.800

10.800-8.4700

8.4700-4.7900

4.7900-4.0000

WEIGHTING SPECTRUM: Within-group weighting fluxes were computed.

The weighting spectrum is:

4KT>E>0eV ........ (S(E)=E.EXP(-E/(KT))/KT ............Maxwellian

670KeV>E>=4KT......(S(E)=C1/E..........................1/E

10MeV>=E>=670KeV...(S(E)=C2.EXP(-E/Wa).Sinh(E.Wb)^1/2..Fission

KT = 0.025eV; Wa = 9.65E+05; Wb = 2.29E-06

The constants C1 & C2 are defined for continuity of the spectrum function.

%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%

FORMAT: WIMS

NUMBER OF GROUPS: 61,64,69,76,79 energy groups

NUCLIDES:

The following nuclides are included:

H-1, H-1 (in H2O), H-1 (in ZrH), H-2, H-2 (in D2O), He-3, He-4, Li-6, Li-7, Be-9, B-10, B-11, C, C (in graphite), N-14, 0-16, F, Na, Al, Si, P-31, S-32, K, Ti, V, Cr, Mn-55, Fe, Co-59, Ni, Cu, Cu-63, Zr, Nb-93, Mo, Mo-92, Mo-94, Mo-95, Mo-96, Mo-97, Mo-98, Mo-99, Mo-100, Ag-107, Ag-109, Cd, In-113, In-115, Sn, Gd, Dy-164, Er-166, Er-167, Lu-176, Hf, Hf-174, Hf-176, Hf-177, Hf-178, Hf-179, Hf-180, Ta-181, Au-197, Pb.

The following fission products are included:

Kr-83, Mo-95, Tc-99, Ru-101, Ru-103, Rh-103, Rh-105, Pd-105, Pd-108, Ag-109, Cd-113, In-115, I-127, I-135, Xe-131, Xe-135, Cs-133, Cs-134, Cs-135, Nd-143, Nd-145, Pm-147, Pm-148g, Pm-148m, Pm-149, Sm-147, Sm-148, Sm-149, Sm-150, Sm-151, Sm-152, Eu-153, Eu-154, Eu-155, Gd-157, Pseudo F.P., Dummy ID = 237 Abs. XS = 1.0E-5 B

The following actinides are included:

Th-232, Pa-233, U-233, U234, U-235, U-236, U-237, U-238, Np-237, Np-238, Np-239, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Am-242g, Am-242m, Am-243, Cm-242, Cm243, Cm-244, !/v absorber ID = 1000.

ORIGIN: ENDF/B-V or IV and JENDL-2(Rev.1); and from ENDL-84, where data was not available in ENDF/B.

Neutron cross section library for thermal reactor design analysis with 61, 64, 69, 76, and 79 energy WIMS groups structure.

Group Energy Structure for resonance region:

WIMSUT61 WIMSUT64 WIMSUT69

9118.00-1942.40 9118.00-3522.40 9118.00-5530.00

1942.40-413.770 3522.40-1344.00 5530.00-3319.10

413.770-88.1450 1344.00-526.000 3319.10-2339.45

88.1450-18.7770 526.000-201.090 2339.45-1425.10

18.7770-4.00000 201.090-76.500 1425.10-906.898

76.5000-28.1270 906.898-367.262

28.1270-10.700 367.262-148.728

10.7000-4.00000 148.728-75.5014

75.5014-48.0520

48.0520-27.7000

27.7000-15.9680

15.9680-9.96800

9.96800-4.00000

WIMSUT76 WIMSUT79

9118.00-6724.90 9118.00-6692.50

6724.90-4960.00 6692.50-4112.20

4960.00-3658.30 4912.20-3605.40

3658.30-2698.20 3605.40-2646.30

2698.20-1912.70 2646.30-1945.00

1912.70-1355.70 1945.00-1423.00

1355.70-961.150 1423.00-1043.00

961.150-681.350 1043.00-783.000

681.350-483.000 783.000-561.000

483.000-340.250 561.000-414.30

340.250-241.200 414.300-304.00

241.200-179.100 304.00-222.910

179.10-126.9600 222.910-161.000

126.960-90.000 161.000-124.000

90.0000-63.8000 124.000-86.500

63.80000-44.800 86.500-53.200

44.80000-28.250 53.200-47.4860

28.25000-13.554 47.486-29.560

13.55400-9.4400 29.560-12.500

.9.440000-4.0000 12.500-10.800

10.800-8.4700

8.4700-4.7900

4.7900-4.0000

WEIGHTING SPECTRUM: Within-group weighting fluxes were computed.

The weighting spectrum is:

4KT>E>0eV ........ (S(E)=E.EXP(-E/(KT))/KT ............Maxwellian

670KeV>E>=4KT......(S(E)=C1/E..........................1/E

10MeV>=E>=670KeV...(S(E)=C2.EXP(-E/Wa).Sinh(E.Wb)^1/2..Fission

KT = 0.025eV; Wa = 9.65E+05; Wb = 2.29E-06

The constants C1 & C2 are defined for continuity of the spectrum function.

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4. METHODS

The library was produced using preprocessing codes to extract data from ENDF/B-VI or and JENDL-2 (Rev.1); and from where data was not available in ENDF/B6.Transport cross sections were computed using P1 scattering matrix data. Within-group weighting fluxes for actinides were computed on an ultra-fine group basis for accurate intermediate resonance self-shielding. A more explicit representation was adapted for the fission-product chain.

A more extensive representation of the actinide burnup chain including Th cycle was selected.

The library was produced using preprocessing codes to extract data from ENDF/B-VI or and JENDL-2 (Rev.1); and from where data was not available in ENDF/B6.Transport cross sections were computed using P1 scattering matrix data. Within-group weighting fluxes for actinides were computed on an ultra-fine group basis for accurate intermediate resonance self-shielding. A more explicit representation was adapted for the fission-product chain.

A more extensive representation of the actinide burnup chain including Th cycle was selected.

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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM

- The resonance integrals are calculated with GROUPIE, which uses the NR approximation. The difference relative to the definitions used in the WIMS code are acknowledged. The discrepancies in calculating integral spectral parameters originate to a large extent from these definitions.

- The preparation of the scattering matrix is based on an analytical expression for energy loss and average scattering cross sections.This approach ignores the information on energy and angular distributions in the ENDF files.

- The average cosine of scattering should be calculated from the angular distributions, but instead it is taken from older evaluations, in which this parameter was given explicitly.

- The resonance integrals are calculated with GROUPIE, which uses the NR approximation. The difference relative to the definitions used in the WIMS code are acknowledged. The discrepancies in calculating integral spectral parameters originate to a large extent from these definitions.

- The preparation of the scattering matrix is based on an analytical expression for energy loss and average scattering cross sections.This approach ignores the information on energy and angular distributions in the ENDF files.

- The average cosine of scattering should be calculated from the angular distributions, but instead it is taken from older evaluations, in which this parameter was given explicitly.

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10. REFERENCES

- Jung-Do Kim: WIMKAL-88, The 1988 Version of WIMS-KAERI Library

IAEA-NDS-92, Rev. 0 (August 1990)

- Jung-Do Kim, Jong Tai Lee, Choong-Sup Gil and Hark Rho Kim:

Generation and Benchmarking of a 69-group Cross Section Library for Thermal Reactor Applications ENDF/B6

- Jung-Do Kim: WIMKAL-88, The 1988 Version of WIMS-KAERI Library

IAEA-NDS-92, Rev. 0 (August 1990)

- Jung-Do Kim, Jong Tai Lee, Choong-Sup Gil and Hark Rho Kim:

Generation and Benchmarking of a 69-group Cross Section Library for Thermal Reactor Applications ENDF/B6

IAEA1397/01, included references:

- Ali Pazirandeh and Alireza Tabesh:New WIMS Library Generation from ENDF/B6 and Effect of Resonance

Group Structure on Cell Parameters

International Conference on Nuclear Data for Science and Technology

Oct. 7-12, 2001, Tsukuba International Congress Center,Tsukuba, Ibaraki, Japan

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IAEA1397/01

ATP61.LIB LibraryATP64.LIB Library

ATP69.LIB Library

ATP76.LIB Library

ATP79.LIB Library

WIMSLIBCol.doc Describing paper

WIMSUT01.doc Abstract

Keywords: ENDF/B, data library, evaluated data, neutron cross sections, thermal reactors.