New ENDF/B-VII.0 Evaluations of Neutron Cross Sections for 32 Fission Products

Conference · Mon Apr 23 00:00:00 EDT 2007

DOI:https://doi.org/10.1051/ndata:07542· OSTI ID:909978

Kim, Hyeong Il ^[1]; Lee, Young-Ouk ^[1]; Herman, Mike ^[2]; Mughabghab, Said F. ^[2]; Oblozinsky, Pavel ^[2]; Rochman, Dimitri ^[2]

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1. Korea Atomic Energy Research Institute (KAERI), Daejeon (Korea, Republic of)
2. Brookhaven National Lab. (BNL), Upton, NY (United States)

Neutron cross sections for fission products play important role not only in the design of extended burnup core and fast reactors, but also in the study of the backend fuel cycle and the criticality analysis of spent fuel. New evaluations in both the resonance and fast neutron regions were performed by the KAERI-BNL collaboration for 32 fission products. These were ⁹⁵Mo, ¹⁰¹Ru, ¹⁰³Rh, ¹⁰⁵Pd, ¹⁰⁹Ag, ¹³¹Xe, ¹³³Cs, ¹⁴¹Pr, and complete isotope chains of ^142-148,150Nd, ^{144,147,148-154}Sm, and ^{156,158,160-164}Dy. The evaluations cover a large amount of reaction channels, including all those needed for neutronics calculations. Also, they cover the entire energy range, from 10^-5 eV to 20 MeV, including the thermal, resolved, and unresolved resonance regions, and the fast neutron region.

• Research Organization: Brookhaven National Lab. (BNL), Upton, NY (United States); Korea Atomic Energy Research Institute (KAERI), Daejeon (Korea, Republic of)
• Sponsoring Organization: USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP); Ministry of Science and Technology (MoST), (Korea, Republic of)
• DOE Contract Number: AC02-98CH10886
• OSTI ID: 909978
• Report Number(s): BNL-78105-2007-CP; R&D Project: 05055; KB0301041; TRN: US0704059
• Resource Relation: Conference: International Conference on Nuclear Data for Science and Technology (ND2007), Evaluations in the Resonance, Nice (France), 22-27 Apr 2007; Related Information: http://www.nd2007.org/programme.html
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
97 MATHEMATICS AND COMPUTING
BURNUP
CHAINS
CRITICALITY
CROSS SECTIONS
DESIGN
ENERGY RANGE
FAST NEUTRONS
FAST REACTORS
FISSION PRODUCTS
FUEL CYCLE
NEUTRONS
RESONANCE
SPENT FUELS
Nuclear Criticality Safety Program (NCSP)
Evaluated Nuclear Data File (ENDF)
Fuel Cycle
Spent Fuel
Cross Section Evaluation Working Group (CSEWG)
Monte Carlo N-Particle (MCNP)
NJOY-99.161