Matthew A Jessee Senior R&D Staff, Power Reactor Modeling, Acting Group Lead Contact jesseema@ornl.gov | 865.241.1503 All Publications Coupled neutronics and species transport simulation of the Molten Salt Reactor Experiment... Parameters of Neutron Diffusion Equation for Temperature Effect of MSRE Heat and Mass Transfer Coefficients in the Molten Salt Reactor Experiment Modeling Enhancements and Demonstration of Shift Capabilities for PBR and MSR Benchmark Calculations for Peach Bottom Unit 2 and Hatch Unit 1 Using the SCALE-6.3.0/Polaris-PARCS v3.4.2 Code Package Benchmark Calculations for BEAVRS and Watt Bar Unit 1 Using the SCALE-6.3.0/Polaris-PARCS v3.4.2 Code Package... Multiphysics Mass Accountancy Calculation for Noble Metals Tracking in MSRE Loop Modeling, Performance Assessment, and Nodal Data Analysis of TRISO-Fueled Systems with Shift Two-step neutronics calculations with Shift and Griffin for advanced reactor systems... Dancoff-Based Wigner-Seitz Approximation for the Subgroup Resonance Self-Shielding in the VERA Neutronics Simulator MPACT VERA Enhancements for Cross Section Shielding and Geometry Capabilities Cell Dancoff-Based Embedded Self-Shielding Capability for Doubly Heterogeneous Particulate Fuels in SCALE/Polaris Transient convective delayed neutron precursors of 235U for the Molten Salt Reactor Experiment... Revisit of the Dancoff-Based Wigner-Seitz Approximation for Pointwise and Multigroup Resonance Self-Shielding Calculations in SCALE Methods and Usability Enhancements in Shift for Non-LWR Applications... Application of Markov Chain Monte Carlo Methods for Uncertainty Quantification in Inverse Transport Problems... Lattice Physics Calculations Using the Embedded Self-Shielding Method in Polaris, Part I: Methods and Implementation Lattice physics calculations using the embedded self-shielding method in polaris, Part II: Benchmark assessment... Applications of VERA for Improvement in NRC Methods... SCALE Code System SCALE Lattice Physics Code Assessments of Accident Tolerant Fuel Development of Perturbed MPACT Multigroup Libraries and the Perturbation Methodology for Subgroup Data Initial Application of TSUNAMI for Validation of Advanced Fuel Systems Application of Markov Chain Monte Carlo for Uncertainty Quantification in Quantitative Imaging Problems Preliminary TSUNAMI Assessment of the Impact of Accident Tolerant Fuel Concepts on Reactor Physics Validation... Pagination Current page 1 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 Modeling and Simulation Development and Deployment Section Power Reactor Modeling Group SCALE
