Ian Gauld Distinguished R&D Staff Contact GAULDIC@ORNL.GOV All Publications Uncertainty Quantification in (α,n) Neutron Source Calculations for an Oxide Matrix Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs... Uncertainty Quantification in (α, n) Neutron Source Calculations for an Oxide Matrix U. S. Commercial Spent Nuclear Fuel Assembly Characteristics: 1968 -2013 (NUREG/CR-7227) Re-evaluation of spent nuclear fuel assay data for the Three Mile Island unit 1 reactor and application to code validation U.S. Commercial Spent Nuclear Fuel Assembly Characteristics: 1968-2013 (NUREG/CR-7227) Uncertainty quantification in (α,n) neutron source calculations for an oxide matrix Re-evaluation of Spent Nuclear Fuel Assay Data for the Three Mile Island Unit 1 Reactor and Application to Code Validation Development of ORIGEN libraries for mixed oxide (MOX) fuel assembly designs New Features of the ORIGEN Transmutation Code in SCALE 6.2 Impact of modeling Choices on Inventory and In-Cask Criticality Calculations for Forsmark3 BWR Spent Fuel... Coolant Density and Control Blade History Effects in Extended BWR Burnup Credit... Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems... In-Field Performance Testing of the Fork Detector for Quantitative Spent Fuel Verification Spent Fuel Modeling and Simulation Using ORIGAMI for Advanced NDA Instrument Testing... Modeling and simulation of Hanford B reactor experiments... Investigation of inconsistent ENDF/B-VII.1 independent and cumulative fission product yields with proposed revisions... Extended Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems Impact of Nuclear Data Uncertainties on Calculated Spent Fuel Nuclide Inventories and Advanced NDA Instrument Response... Approach for Validating Actinide and Fission Product Compositions for Burnup Credit Criticality Safety Analyses... OECD NEA Benchmark Database of Spent Nuclear Fuel Isotopic Compositions for World Reactor Designs... Special Nuclear Material Inventory Processes at US Domestic Nuclear Power Plants... Validation of ORIGEN for LWR used fuel decay heat analysis with SCALE Validation and Testing of ENDF/B-VII Decay Data A New Approach to Fork Measurements Data Analysis by RADAR-CRISP and ORIGEN Integration Pagination First page « First Previous page ‹‹ Page 1 Current page 2 Page 3 … Next page ›› Last page Last » Key Links ORCID
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Research Highlight Uncertainty Quantification in (α,n) Neutron Source Calculations in an Oxide Matrix
