Marco T Pigni Distinguished R&D Nuclear Data Scientist Contact PIGNIMT@ORNL.GOV All Publications (α,n) reactions in oxide compounds calculated from the R-matrix theory PROGRESS ON 140,142CE NEUTRON CROSS SECTION RESOLVED RESONANCE REGION EVALUATIONS NCSP Analytical Methods Subtask 3 & NCSP Nuclear Data Subtask 6: AMPX Development and Maintenance & SAMMY Modernization Verification of R-matrix calculations for charged-particle reactions in the resolved resonance region for the 7Be system Status of (α,n)-reaction data for nuclear safeguards Bayesian optimization of generalized data 17,18 O(a,n) Evaluated Cross Sections to Improve National Security Applications Prioritizing Nuclear Data Needs Using Uncertainty Analysis ENDF/B-VIII.0: The 8th Major Release of the Nuclear Reaction Data Library with CIELO-project Cross Sections, New Standards and Thermal Scattering Data CIELO Collaboration Summary Results: International Evaluations of Neutron Reactions on Uranium, Plutonium, Iron, Oxygen and Hydrogen IAEA CIELO Evaluation of Neutron-induced Reactions on 235U and 238U Targets Early applications of the R -matrix SAMMY code for charged-particle induced reactions and related covariances... The CIELO Collaboration: Progress in International Evaluations of Neutron Reactions on Oxygen, Iron, Uranium and Plutonium n+ 235 U resonance parameters and neutron multiplicities in the energy region below 100 eV Generalized Reich-Moore R -matrix approximation Evaluation of the neutron induced reactions on 235 U from 2.25 keV up to 30 MeV Validation of tungsten cross sections in the neutron energy region up to 100 keV Systematic Approach to Nuclear Data Uncertainty Quantification for Nuclear Security Applications... Uncertainty Quantification in (α,n) Neutron Source Calculations for an Oxide Matrix Uncertainty quantification in (α,n) neutron source calculations for an oxide matrix Uncertainty Quantification in (α, n) Neutron Source Calculations for an Oxide Matrix SAMINT: A New Evaluation Tool to Perform Resonance Parameter Data Adjustments based on Integral Experiment Data... Evaluated 182, 183, 184, 186W Neutron Cross Sections and Covariances in the Resolved Resonance Region... Investigation of inconsistent ENDF/B-VII.1 independent and cumulative fission product yields with proposed revisions... ORNL Nuclear Data Evaluation Accomplishments for FY2013 Pagination First page « First Previous page ‹‹ Page 1 Current page 2 Page 3 Next page ›› Last page Last » Key Links Curriculum Vitae ORCID Organizations Fusion and Fission Energy and Science Directorate Nuclear Energy and Fuel Cycle Division Nuclear Data, Criticality Safety, and Radiation Transport Section Nuclear Data Group SCALE
Research Highlight Uncertainty Quantification in (α,n) Neutron Source Calculations in an Oxide Matrix
