TY - GEN
T1 - A broad-group cross-section library based on ENDF/B-VII.0 for fast neutron dosimetry using the CPXSD methodology
AU - Alpan, F. A.
PY - 2012
Y1 - 2012
N2 - A new ENDF/B-VII.0-based coupled 44-neutron, 20-gamma-ray-group cross-section library was developed to investigate the latest evaluated nuclear data file (ENDF) ,in comparison to ENDF/B-VI.3 used in BUGLE-96, as well as to generate an objective-specific library. The objectives selected for this work consisted of dosimetry calculations for in-vessel and ex-vessel reactor locations, iron atom displacement calculations for reactor internals and pressure vessel, and 58Ni(n,γ) calculation that is important for gas generation in the baffle plate. The new library was generated based on the contributon and point-wise cross-section-driven (CPXSD) methodology and was applied to one of the most widely used benchmarks, the Oak Ridge National Laboratory Pool Critical Assembly benchmark problem. In addition to the new library, BUGLE-96 and an ENDF/B-VII.0-based coupled 47-neutron, 20-gamma-ray-group cross-section library was generated and used with both SNLRML and IRDF dosimetry cross sections to compute reaction rates. All reaction rates computed by the multigroup libraries are within ± 20 % of measurement data and meet the U. S. Nuclear Regulatory Commission acceptance criterion for reactor vessel neutron exposure evaluations specified in Regulatory Guide 1.190.
AB - A new ENDF/B-VII.0-based coupled 44-neutron, 20-gamma-ray-group cross-section library was developed to investigate the latest evaluated nuclear data file (ENDF) ,in comparison to ENDF/B-VI.3 used in BUGLE-96, as well as to generate an objective-specific library. The objectives selected for this work consisted of dosimetry calculations for in-vessel and ex-vessel reactor locations, iron atom displacement calculations for reactor internals and pressure vessel, and 58Ni(n,γ) calculation that is important for gas generation in the baffle plate. The new library was generated based on the contributon and point-wise cross-section-driven (CPXSD) methodology and was applied to one of the most widely used benchmarks, the Oak Ridge National Laboratory Pool Critical Assembly benchmark problem. In addition to the new library, BUGLE-96 and an ENDF/B-VII.0-based coupled 47-neutron, 20-gamma-ray-group cross-section library was generated and used with both SNLRML and IRDF dosimetry cross sections to compute reaction rates. All reaction rates computed by the multigroup libraries are within ± 20 % of measurement data and meet the U. S. Nuclear Regulatory Commission acceptance criterion for reactor vessel neutron exposure evaluations specified in Regulatory Guide 1.190.
KW - Cross-section library
KW - ENDF/B-VII.0
KW - Neutron dosimetry
UR - http://www.scopus.com/inward/record.url?scp=84867950991&partnerID=8YFLogxK
U2 - 10.1520/stp155020120042
DO - 10.1520/stp155020120042
M3 - Conference contribution
AN - SCOPUS:84867950991
SN - 9780803175365
T3 - ASTM Special Technical Publication
SP - 548
EP - 560
BT - Reactor Dosimetry
PB - ASTM International
T2 - 14th International Symposium on Reactor Dosimetry
Y2 - 22 May 2011 through 27 May 2011
ER -