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Title: ENDF/B-VI Release 3 Cross Section Library for Use with the MCNP Monte Carlo Code.

Version 00 This continuous energy cross-section data library for MCNP is in ACE format. The present library was satisfactorily tested in thermal and fast criticality benchmarks. For analyses below 20 MeV, MCB63NEA.BOLlB was applied also in cell and core calculations dedicated to the study of the subcritical accelerator driven systems (ADS). This library provides users an additional ENDF/B-VI based, continuous-energy and multi-temperature library for MCNP with an important feature: there is a perfect consistency with the twin library MCJEFF22NEA.BOLIB already released, in terms of nuclear data processing calculation methodology. Both libraries are based on the NJOY-94.66 data processing system. This may be important, in particular, for the users involved in nuclear data validation who have already used the MCJEF22NEA.BOLIB library.
Authors:
;
Publication Date:
OSTI Identifier:
1266943
Report Number(s):
MCB63NEA.BOLIB; 004192MLTPL00
RSICC ID: D00216MNYCP
DOE Contract Number:
AC05-00OR22725
Resource Type:
Software
Software Revision:
00
Software Package Number:
004192
Software CPU:
MLTPL
Source Code Available:
No
Other Software Info:
Owner Installation: ENEA - BOLOGNA Contributors: ENEA - Centro Ricerche, Bologna, Italy, through the OECD Nuclear Energy Agency Data Bank, Issy-Les Molineaux, France. MCB63NEA.BOLIB is a pointwise library for nuclear fission applications produced at the Nuclear Data Centre of ENEA-Bologna. The library was processed in ACE format for the MCNP Monte Carlo transport code with the NJOY-94.66 nuclear data processing system. The library is based on the ENDF/B-VI Release 3 evaluated data file. It contains at present 107 isotopes/natural elements, including fission products, processed for up to eight temperatures: 300 K, 500 K, 560 K, 760 K, 800 K, 1000 K, 1500 K and 2200 K. The processed data include gamma-ray and gas production data when available in the specific ENDF/B-VI Release 3 evaluated data files. Thermal scattering cross sections were processed for some of the most important moderator materials using the thermal scattering matrices S (alpha, beta) at various temperatures, included in the original ENDF/B-VI Release 3 thermal scattering law data file. KEYWORDS: REACTION CROSS SECTIONS; DOSIMETRY CROSS SECTIONS; MCNP FORMAT; FISSION PRODUCT CROSS SECTIONS
Research Org:
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Sponsoring Org:
USDOE
Country of Publication:
United States

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