Computer Programs
NEA-0886 ZZ-AMPX-2/123.
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NEA-0886 ZZ-AMPX-2/123.

ZZ AMPX-2/123, 123-Group Neutron Cross-Section Library from ENDF/B-4 by AMPX-2
ZZ AMPX-2/219, 219-Group Neutron Cross-Section Library from ENDF/B-4 by AMPX-2

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1. DESCRIPTION

ZZ-AMPX-2/123, ZZ-AMPX-2/219.

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2. COMPUTERS
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-AMPX-2/123 NEA-0886/03 Tested 05-JAN-1989

Machines used:

Package ID Orig. computer Test computer
NEA-0886/03 IBM 370 series DEC VAX 8810
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3. DESCRIPTION OF PROBLEM OR FUNCTION

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FORMAT: "data base" for subsequent collapsing into both fine and broad group data in various formats (working and/or weighted ANISN, CCCC, etc.).

 

NUMBER OF GROUPS:

  • AMPX-2/123: 123 group structure

  • AMPX-2/219: 219 group structure

 

NUCLIDES: H, He, Li, Be, B, C, N, O, F, Na, Mg, Al, Si, Cl, K, Ca, Ti, V, Cr, Mn, Fe, Ni, Cu, Kr, Zirc, Mo, Tc, Rh, Ag, Cd, Xe, Sm, Eu, Gd, Dy, Cu, Ta, W, Re, Pb, Th, Pa, U, Np, Pu, Am, Cm.

 

ORIGIN: ENDF/B-IV

 

WEIGHTING SPECTRUM: Most data were generated using a standard flux over three energy ranges (fission - 1/E - Maxwellian) as point-to-fine-group cross sections weighing function.

 

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The AMPX-2 P3 123- and 219- Group Neutron Cross-Section Master Interface Libraries may be considered as "data bases" for subsequent collapsing into both fine and broad group data in various formats (working and/or weighted ANISN, CCCC, etc.). The built-in 123 and 219 group structures have been used to process all available data of ENDF/B-IV.

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4. METHODS

The program AMPX-2 has been used to generate the data. By various executions of the module XLACS-2 (XLACS for bound 1H in some materials) a number of independent libraries were generated which then were combined using the AMPX-2 module AJAX.

 

Most data were generated using a standard flux over three energy ranges (fission - 1/E - Maxwellian) as point-to-fine-group cross sections weighing function. For some structural materials (e.g. Fe, Cr,...) different master data sets were produced using a weighting function fission - 1/E∑T(SS-304) - Maxwellian, and the three parts of the spectrum were joined at properly selected energies.

 

For some nuclides (e.g. 238U and 240Pu) various master data sets have been produced which contain problem-dependent unresolved cross sections characterized by the associated potential scattering cross sections.

 

Some data sets contain P3 thermal scattering matrices, for which ENDF/B File 7 S(α, β) data were used, e.g. MAT=1002 (H bound in H2O), MAT-1004 (D bound in D2O), etc.

 

Some materials have been generated with more than one thermal scattering matrix. Generally, the following temperatures were considered: 450, 600, 750, and 900 K.

 

Also, high-temperature thermal scattering matrices have been generated. Where applicable (i.e. if the ENDF/B-IV evaluation contains resonance information File 2) Doppler broadening in the resolved and unresolved regions was calculated at the given temperature.

 

Where possible for resonance nuclides, the elastic scattering matrices were calculated using MT=2 data from a corresponding NPTXS generated point cross-section library. The use of internally produced elastic scattering data was bypassed in the XLACS-2 module.

 

Regarding the order of the produced thermal scattering matrices, both above-mentioned very light moderators (H and D) and light nuclides (i.e. A < 20) have P3 matrices.

 

For heavier nuclides, only PO thermal scattering matrices were generated and normalized to the free-atom scattering cross sections in the epithermal range (e.g. 238U); for nuclides with resonances in the thermal region (e.g. 235U, 240Pu), the thermal scattering matrices were normalized to the average values of the cross sections in the thermal range.

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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM

Only 123- and 219- group P3 calculations may be performed.

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8. RELATED OR AUXILIARY PROGRAMS

After appropriate further processing, the master data sets were used as library data in the following programs: ANISN, XSDRNM, KENO-4, and MORSE-CG for sufficiently large test cases.

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9. STATUS
Package ID Status date Status
NEA-0886/03 05-JAN-1989 Screened
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10. REFERENCES
NEA-0886/03, included references:
- P.A. Landeyro, F. Siciliano and Tayyab Abbas:
  AMPX-2 123 Multigroup Neutronic Library Validation for Fuel
  Storage Pools. CNEN-RT/ING(80)17  (July 1980)
- F. Siciliano and T. Abbas:
  AMPX-2 123 Group Neutron Cross Section Library Validation for
  4.29WT.% U-235 Enriched UO2 Fuel Storage Under Water.
  CNEN-RT/ING(80)18  (September 1980)
- F. Siciliano and G. Lai:
  A Complete AMPX-2 123 Group Neutron Cross Section Library
  Production and Testing for Criticality Safety Calculations.
  CNEN-RT/ING(82)10  (April 1982)
- F. Siciliano and T. Abbas:
  AMPX-2 123 and 219 Group Neutron Cross Section Libraries
  Production and Validation for Criticality and Safety Studies.
  CNEN-RT/ING(80)20  (October 1980)
- M. Carotenuto, P.A. Landeyro, F. Siciliano and T. Abbas:
  Cross Section Validation for Fuel Storage Pools (Enrichment
  4.29wt%). Transaction ANS, Vol. 39 (Nov. 1981) pp. 530-531
- P.A. Landeyro, F. Siciliano and T. Abbas:
  Cross Section Validation for Fuel Storage Pools. Transaction ANS,
  (June 1981) pp. 361-362
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12. PROGRAMMING LANGUAGE(S) USED
No specified programming language
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15. NAME AND ESTABLISHMENT OF AUTHORS

F. Siciliano

ENEA

Comitato Nazionale per la Ricerca e per lo Sviluppo dell'Energia Nucleare e delle Energie Alternative

Centro COMB Trisaia

75025 POLICORO (Matera), Italy

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16. MATERIAL AVAILABLE
NEA-0886/03
File name File description Records
NEA0886_03.001 INFORMATION FILE 74
NEA0886_03.002 AMPX-2 P3 123 GP. NEUTRON XSEC.MAST.INT.LIB. 462431
NEA0886_03.003 AIM MODULE LOG FILE, BCD-BINARY CONV. 320
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17. CATEGORIES
  • Z. Data

Keywords: criticality, cross sections, data library, group constants, reactor safety, shielding.