ENDF/B-VII.1 Neutron Cross Section Data Testing with Critical Assembly Benchmarks and Reactor Experiments

Journal Article · Thu Dec 01 00:00:00 EST 2011 · Nuclear Data Sheets

DOI:https://doi.org/10.1016/j.nds.2011.11.003· OSTI ID:1034732

Kahler, A C; Herman, M; Kahler, A C; MacFarlane, R E; Mosteller, R D; Kiedrowski, B C; Frankle, S C; Chadwick, M B; McKnight, R D; Lell, R M; Palmiotti, G; Hiruta, H; Herman, M; Arcilla, R; Mughabghab, S F; Sublet, J C; Trkov, A; Trumbull, T H; Dunn, M

The ENDF/B-VII.1 library is the latest revision to the United States Evaluated Nuclear Data File (ENDF). The ENDF library is currently in its seventh generation, with ENDF/B-VII.0 being released in 2006. This revision expands upon that library, including the addition of new evaluated files (was 393 neutron files previously, now 423 including replacement of elemental vanadium and zinc evaluations with isotopic evaluations) and extension or updating of many existing neutron data files. Complete details are provided in the companion paper [M. B. Chadwick et al., 'ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data,' Nuclear Data Sheets, 112, 2887 (2011)]. This paper focuses on how accurately application libraries may be expected to perform in criticality calculations with these data. Continuous energy cross section libraries, suitable for use with the MCNP Monte Carlo transport code, have been generated and applied to a suite of nearly one thousand critical benchmark assemblies defined in the International Criticality Safety Benchmark Evaluation Project's International Handbook of Evaluated Criticality Safety Benchmark Experiments. This suite covers uranium and plutonium fuel systems in a variety of forms such as metallic, oxide or solution, and under a variety of spectral conditions, including unmoderated (i.e., bare), metal reflected and water or other light element reflected. Assembly eigenvalues that were accurately predicted with ENDF/B-VII.0 cross sections such as unmoderated and uranium reflected {sup 235}U and {sup 239}Pu assemblies, HEU solution systems and LEU oxide lattice systems that mimic commercial PWR configurations continue to be accurately calculated with ENDF/B-VII.1 cross sections, and deficiencies in predicted eigenvalues for assemblies containing selected materials, including titanium, manganese, cadmium and tungsten are greatly reduced. Improvements are also confirmed for selected actinide reaction rates such as {sup 236}U, {sup 238,242}Pu and {sup 241,243}Am capture in fast systems. Other deficiencies, such as the overprediction of Pu solution system critical eigenvalues and a decreasing trend in calculated eigenvalue for {sup 233}U fueled systems as a function of Above-Thermal Fission Fraction remain. The comprehensive nature of this critical benchmark suite and the generally accurate calculated eigenvalues obtained with ENDF/B-VII.1 neutron cross sections support the conclusion that this is the most accurate general purpose ENDF/B cross section library yet released to the technical community.

• Research Organization: Brookhaven National Lab. (BNL), Upton, NY (United States)
• Sponsoring Organization: USDOE SC OFFICE OF SCIENCE (SC)
• DOE Contract Number: DE-AC02-98CH10886
• OSTI ID: 1034732
• Report Number(s): BNL-96635-2011-JA; NDTSBA; R&D Project: EST-003-NEFA; KB0301041; TRN: US1200913
• Journal Information: Nuclear Data Sheets, Vol. 112, Issue 12; ISSN 0090-3752
• Country of Publication: United States
• Language: English

Related Subjects

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS
73 NUCLEAR PHYSICS AND RADIATION PHYSICS
ACTINIDES
BENCHMARKS
CADMIUM
CRITICALITY
CROSS SECTIONS
DECAY
EIGENVALUES
FISSION
FISSION PRODUCTS
FUEL SYSTEMS
MANGANESE
NEUTRONS
NUCLEAR DATA COLLECTIONS
OXIDES
PLUTONIUM
REACTION KINETICS
TITANIUM
TUNGSTEN
URANIUM
VANADIUM
ZINC