Ian Gauld Distinguished R&D Staff Contact GAULDIC@ORNL.GOV All Publications Validation and Testing of ENDF/B-VII Decay Data Approach for Validating Actinide and Fission Product Compositions for Burnup Credit Criticality Safety Analyses Validation and Testing of ENDF-B-VII Decay Data A New Approach to Fork Measurements Data Analysis by RADAR-CRISP and ORIGEN Integration Validation and Testing of ENDF/B-VII Decay Data Impact of Nuclear Data Uncertainties on Calculated Spent Fuel Nuclide Inventories and Advanced NDA Instrument Response Validation of ORIGEN for LWR Used Fuel Decay Heat Analysis with SCALE Comparison of Burnup Credit Uncertainty Quantification Methods... OVERVIEW OF SCALE 6.2 SCALE Uncertainty Quantification Methodology for Criticality Safety Analysis of Used Nuclear Fuel Study of Fukushima Dai-ichi Nuclear Power Station Unit 4 Spent Fuel Pool Validation of new depletion capabilities and ENDF/B-VII data libraries in Scale Utilizing NGSI Spent Fuel Sensitivity Libraries to Estimate Model Uncertainties Creating NDA “working standards” through high-fidelity spent fuel modeling... Enhancements in SCALE 6.1... An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses-Isotopic Composition Predictions (NUREG/CR-7108, ORNL/TM-2011/509) Validation of New Depletion Capabilities and ENDF/B-VII Data Libraries in SCALE An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses-Isotopic Composition Predic... Scale 6.1 Enhancements for Nuclear Criticality Safety... GAMMA DOSE RATE NEAR A NEW (252)Cf BRACHYTHERAPY SOURCE ... Modular ORIGEN-S for Multi-Physics Code Systems Isotopic Depletion and Decay Methods and Analysis Capabilities in SCALE... Development of Technical Basis for Burnup Credit Regulatory Guidance in the United States Isotopic Depletion and Decay Methods and Analysis Capabilities in SCALE Proposed Revision of the Decay Heat Standard ANSI/ANS-5.1-2005... Pagination First page « First Previous page ‹‹ … Page 2 Current page 3 Page 4 Next page ›› Last page Last » Key Links ORCID
Research Highlight ORNL and international collaborators deliver world-leading spent nuclear fuel database
Research Highlight Uncertainty Quantification in (α,n) Neutron Source Calculations in an Oxide Matrix
