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NEA-1669 ZZ-MCB-ENDF/B-6.8.
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NEA-1669 ZZ-MCB-ENDF/B-6.8.

ZZ MCB-ENDF/B6.8, MCB Continuous-Energy Neutron Cross Section Libraries for Temperatures from 300 to 1800 K.

NAME, COMPUTER, PROBLEM, SOLUTION, CPU, AUXILIARIES, STATUS, REFERENCES, REQUIREMENTS, LANGUAGE, AUTHOR, MATERIAL, CATEGORIES
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1. NAME OR DESIGNATION OF PROGRAM

ZZ-MCB-ENDF/B6.8.

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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-MCB-ENDF/B-6.8 NEA-1669/03 Tested 20-DEC-2004

Machines used:

Package ID Orig. computer Test computer
NEA-1669/03 Linux-based PC,UNIX W.S. Linux-based PC
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3. DESCRIPTION OF PROGRAM OR FUNCTION

MCB-ENDF/B6.8 is a continuous-energy cross section libraries in ACE format suitable for the MCB-1C and MCNP codes. Libraries for various materials were generated at six different temperatures, and cover the energy range up to 150 MeV.

FORMAT: ACE

NUMBER OF GROUPS: Continuous energy

NUCLIDES: Library includes elements with the following atomic numbers:

Z= 1-9, 11-17, 19-29, 31-68, 71-75, 77, 79, 82-83, 90-99, (i.e.: H, He, Li, Be, B, C, N, O, F, Na, Mg, Al, Si, P, S, Cl, K, Ca, Sc, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Ga, Ge, As, Se, Br, Kr, Rb, Sr, Y, Zr, Nb, Mo, Tc, Ru, Rh, Pd, Pd, Ag, Cd, In, Sn, Sb, Te, I, Xe, Cs, Ba, La, Ce, Pr, Nd, Pm, Sm, Eu, Gd, Tb, Dy, Ho, Er, Lu, Hf, Ta, W, Re, Ir, Au, Pb, Bi, Th, Pa, U, Np, Pu, Am, Cm, Bk, Cf, Es).
TEMPERATURES: 300 K, 600 K, 900 K, 1200 K, 1500 K, and 1800 K.

ORIGIN: ENDF/B-6.8

WEIGHTING SPECTRUM:

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4. METHODS

The library was generated using the NJOY processing code. An example of the input data is provided.

 

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6. TYPICAL RUNNING TIME

The ASCII-to-binary conversion can generally be accomplished within a few seconds, even for the large data files.

 

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8. RELATED OR AUXILIARY PROGRAMS

MAKSF, which is distributed with CCC-700/MCNP4C, can be used to convert MCNP Type 1 data to Type 2 (binary). For use with the code MCB-1C the MAKXSF installation must be slightly modified.

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9. STATUS
Package ID Status date Status
NEA-1669/03 20-DEC-2004 Tested at NEADB
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10. REFERENCES

  • J. Cetnar, W. Gudowski and J. Wallenius, "MONTE-CARLO CONTINUOUS ENERGY BURNUP (MCB1C) - THE CODE DESCRIPTION, METHODS AND BENCHMARKS" in preparation for NSE.

  • J. Cetnar, W. Gudowski and J. Wallenius, "MCB: A continuous energy Monte Carlo Burnup simulation code",  In "Actinide and Fission Product Partitioning and Transmutation", EUR 18898 EN, OECD/NEA (1999) 523.

  • J. F. Briesmeister, Ed.
    "MCNP - A General Monte Carlo N-Particle Transport Code, Version 4C," LA-13709-M (April 2000).

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11. HARDWARE REQUIREMENTS

ASCII card images; Unix workstation. The library distribution was prepared on a LINUX system. On a Windows-based PC, one may use utilities such as WinZip(R) to unzip and extract the library files.

 

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12. PROGRAMMING LANGUAGE(S) USED
No specified programming language
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15. NAME AND ESTABLISHMENT OF AUTHORS

Department of Nuclear and Reactor Physics

Royal Institute of Technology

100 44 Stockholm

Sweden

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16. MATERIAL AVAILABLE
NEA-1669/03
Readme.Endfb6.8    Original information file
InstEndf-l  Installation procedure for independent library installation
(outside MCB system)
ENDFB6.8 cross section data library (two files per temperature)
XSDIR files XSDIR file for LINUX and UNIX operating systems
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17. CATEGORIES
  • Z. Data

Keywords: Monte Carlo method, cross sections, data library, fission products, neutrons, reaction.