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

Abstract

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:
Research Org.:
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Sponsoring Org.:
USDOE
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
Country of Publication:
United States

Citation Formats

MASSIMO,, and PESCARINI,. ENDF/B-VI Release 3 Cross Section Library for Use with the MCNP Monte Carlo Code.. Computer software. Vers. 00. USDOE. 16 Dec. 2003. Web.
MASSIMO,, & PESCARINI,. (2003, December 16). ENDF/B-VI Release 3 Cross Section Library for Use with the MCNP Monte Carlo Code. (Version 00) [Computer software].
MASSIMO,, and PESCARINI,. ENDF/B-VI Release 3 Cross Section Library for Use with the MCNP Monte Carlo Code.. Computer software. Version 00. December 16, 2003.
@misc{osti_1266943,
title = {ENDF/B-VI Release 3 Cross Section Library for Use with the MCNP Monte Carlo Code., Version 00},
author = {MASSIMO, and PESCARINI,},
abstractNote = {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.},
doi = {},
year = {Tue Dec 16 00:00:00 EST 2003},
month = {Tue Dec 16 00:00:00 EST 2003},
note =
}

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  • A revised multigroup cross-section library based ON ENDF/B-VI Release 3 has been produced for light water reactor shielding and reactor pressure vessel dosimetry applications. This new broad-group library, which is designated BUGLE-96, represents an improvement over the BUGLE-93 library released in February 1994 and is expected to replace te BUGLE-93 data. The cross-section processing methodology is the same as that used for producing BUGLE-93 and is consistent with ANSI/ANS 6.1.2. As an added feature, cross-section sets having upscatter data for four thermal neutron groups are included in the BUGLE-96 package available from the Radiation Shielding Information Center. The upscattering datamore » should improve the application of this library to the calculation of more accurate thermal fluences, although more computer time will be required. The incorporation of feedback from users has resulted in a data library that addresses a wider spectrum of user needs.« less
  • A comparison is made of results obtained from neutron transmissions analysis of RPV performed by MCNP with ENDF/B-VI and JENDL-3.1 iron data. At first, a one-dimensional discrete ordinates transport calculation using VITAMIN-C fine-group library based on ENDF/B-IV was performed for a cylindrical model of a PWR to generate the source spectrum at the front of the RPV. And then, the transmission of neutrons through RPV was calculated by MCNP with the moderated fission spectrum incident on the vessel face. For these ENDF/B-IV, -VI and JENDL-3.1 iron data were processed into continuous energy point data form by NJOY91.91. The fast neutronmore » fluxes and dosimeter reaction rates through RPV using each iron data were intercompared.« less
  • The purpose of this study is to compare the results of using either the reference cross-section data set ENDF/B-V or ENDF/B-VI RACER vectorized Monte Carlo calculations on several fast critical experiments. Seven benchmark cores were chosen that span a range of [sup 235]U enrichments and neutron leakage fractions. These include Godiva, Flat-Top-25, Zero-Power Reactor (ZPR)-Ill 6F, Vera-1 B, ZPR-III 12, ZPR- III 12, and Zebra-2.
  • The purpose of this study is to investigate the eigenvalue sensitivity to changes in ENDF/B-V and ENDF/B-VI cross section data sets by comparing RACER vectorized Monte Carlo calculations for several thermal and intermediate spectrum critical experiments. Nineteen Oak Ridge and Rocky Flats thermal solution benchmark critical assemblies that span a range of hydrogen-to-{sup 235}U (H/U) concentrations (2052 to 27.1) and above-thermal neutron leakage fractions (0.555 to 0.011) were analyzed. In addition, three intermediate spectrum critical assemblies (UH3-UR, UH3-NI, and HISS-HUG) were studied.
  • Version 01 This continuous energy cross-section data library for MCNP is based on the JEF-2.2 evaluated nuclear data library (ACE format). The present library was satisfactorily tested in thermal and fast criticality benchmarks. For analyses below 20 MeV, MCJEF22NEA.BOLlB was applied also in cell and core calculations dedicated to the study of the subcritical accelerator driven systems (ADS).

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