Cihangir Celik Senior R&D Staff, Nuclear Criticality Safety Contact 865.576.1690 | celikc@ornl.gov All Publications SCALE 6.3 Validation: Radiation Shielding ASSESSMENT OF VALIDATION FOR BURNUP CREDIT CALCULATIONS FOR LEU+ AND HIGH BURNUP FUEL Effect of Decay Time on Criticality Safety Analyses for High Burnup and Extended Enrichment Fuels Nuclear Data–Induced Uncertainties in Criticality Safety Analyses for High-Burnup and Extended Enrichment Fuels Photonuclear Physics in SCALE Using the Oak Ridge Subcritical Assembly for Gamma Ray Measurements IMPACT OF RECENT ENDF NUCLEAR DATA ON BURNUP CREDIT CRITICALITY SAFETY ANALYSES Overview of ORNL SCALE Shielding Analyses for Spent Nuclear Fuel Transportation and Storage Applications... SCALE 6.2.4 Validation: Radiation Shielding Update of the Nuclear Criticality Slide Rule Calculations: Plutonium systems – Delayed Fission Gamma Update of the Nuclear Criticality Slide Rule: Review of the Estimation of the Number of Fissions... Update of the Nuclear Criticality Slide Rule Calculations: Plutonium Systems – Delayed Fission Gamma SCALE Analysis of a Fluoride Salt-Cooled High-Temperature Reactor in Support of Severe Accident Analysis Evaluation of Oak Ridge National Laboratory Health Physics Research Reactor Operation Data for Criticality Accident Alarm System Benchmark Creation Nuclear Data Assessment for Advanced Reactors... Development of a Fast-Spectrum Self-Powered Neutron Detector for Molten Salt Experiments in the Versatile Test Reactor Criticality Accident Alarm System Shielding Benchmark: Integral Experiment Request 498, Critical Engineering Decision 2 Report Overview of the 2020 SCALE 6.2.4 Validation Report for Radiation Shielding Applications* 3D Model Visual Verification and Mesh-Based Data Analysis in Fulcrum Monte Carlo Simulation of Background and Source Measurements with CSG and CAD Geometries SCALE Capabilities for High Temperature Gas-Cooled Reactor Analysis The MAVRIC-Shift Sequence in SCALE for Radiation Transport and Shielding Calculations with Automated Variance Reduction and Parallel Computing Monte Carlo Uncertainty Quantification in UF6 Cylinder Neutron Emissions... A Directional Detector Response Function for Anisotropic Detectors Assessment of SCALE Capabilities for High Temperature Reactor Modeling and Simulation... Pagination Current page 1 Page 2 Next page ›› Last page Last » Key Links ORCID Organizations Fusion and Fission Energy and Science Directorate Nuclear Energy and Fuel Cycle Division Nuclear Data, Criticality Safety, and Radiation Transport Section Nuclear Criticality Safety Group SCALE
