Advanced Fuel Cycle Initiative (AFCI)

Advanced Fuel Cycle Initiative (AFCI)
Fundamental Understanding of Ambient and High-Temperature PlasticityGeorgia Institute of Technology
Advanced Elastic/Inelastic Nuclear Data Development ProjectIdaho State University
Heterogeneous Recycling in Fast ReactorsMassachusetts Institute of Technology
Thermodynamic Development of Corrosion Rate Modeling in Iron PhosphateGlassesMissouri University of Science and Technology
Development of Subspace-Based Hybrid Monte Carlo-Deterministic Algorithms for Reactor Physics CalculationsNorth Carolina State University
SiC Schottky Diode Detectors for Measurement of Actinide Concentrations from Alpha Activities in Molten Salt ElectrolyteOhio State University
Simulations of Failure via Three-Dimensional Cracking in Fuel Cladding forOklahoma State University
Improvements to Nuclear Data and Its Uncertainties by Theoretical ModelingRensselaer Polytechnic Institute
Sharp Interface Tracking in Rotating Microflows of Solvent ExtractionState University of New York at Stony Brook
Bulk Nanostructured FCC Steels with Enhanced Radiation ToleranceTexas A&M University
Fuel Performance Experiments and Modeling: Fission Gas Bubble Nucleationand Growth in Alloy Nuclear FuelsTexas A&M University
Computational Design of Advanced Nuclear FuelsUniversity of California
Data Collection Methods For Validation of Advanced Multi-Resolution Fast Reactor SimulationsUniversity of Idaho
Simulations of the Thermodynamic and Diffusion Properties of Actinide Oxide Fuel MaterialsUniversity of Michigan
Adsorptive Separation and Sequestration of Krypton, I and C14 on DiamondUniversity of Missouri, Columbia
High-Fidelity Space-Time Adaptive Multiphysics Simulations in Nuclear EngineeringUniversity of Nevada
Development of Alternative Technetium Waste FormsUniversity of Nevada, Las Vegas
Quantification of UV-Visible and Laser Spectroscopic Techniques for Materials Accountability and Process ControlUniversity of Nevada, Las Vegas
Ab Initio Enhanced Calphad Modeling of Actinide-Rich Nuclear FuelsUniversity of Wisconsin, Madison
Advanced Mesh-Enabled Monte Carlo Capability for Multi-Physics ReactorUniversity of Wisconsin, Madison
Development of Diffusion Barrier Coatings and Deposition Technologies for Mitigating Fuel Cladding Chemical Interactions (FCCI)University of Wisconsin, Madison
Thermal Properties of LiCl-KCl Molten Salt for Nuclear Waste SeparationUniversity of Wisconsin, Madison

Next Generation Nuclear Plant (NGNP)/Generation IV Nuclear Systems

Next Generation Nuclear Plant (NGNP)/Generation IV Nuclear Systems
Irradiation Creep in GraphiteBoise State University
Modeling the Stress Strain Relationships and Predicting Failure ProbabilitiesFor Graphite Core ComponentsCleveland State University
TRISO-Coated Fuel Durability Under Extreme ConditionsColorado School of Mines
An Innovative and Advanced Coupled Neutron Transport and Thermal Hydraulic Method (Tool) for the Design, Analysis and Optimization of VHTR/NGNP Prismatic ReactorsGeorgia Institute of Technology
Removal of 14C from Irradiated Graphite for Graphite Recycle and Waste Volume ReductionIdaho State University
Millimeter-Wave Thermal Analysis Development and Application to Gen IVReactor MaterialsMassachusetts Institute of Technology
Accurate Development of Thermal Neutron Scattering Cross Section LibrariesNorth Carolina State University
Microscale Heat Conduction Models and Doppler FeedbackNorth Carolina State University
Multiaxial Creep-Fatigue and Creep-Ratcheting Failures of Grade 91 and Haynes 230 Alloys Toward Addressing Design Issues of Gen IV Nuclear Power PlantsNorth Carolina State University
Optimizing Neutron Thermal Scattering Effects in Very High Temperature ReactorsNorth Carolina State University
Understanding Creep Mechanisms in Graphite with Experiments, MultiscaleSimulations, and ModelingNorth Carolina State University
Verification & Validation of High-Order Short-Characteristics-Based Deterministic Transport Methodology on Unstructured GridsNorth Carolina State University
Investigation of Countercurrent Helium–Air Flows in Air-ingress Accidents for VHTRsOhio State University
Testing of Performance of Optical Fibers Under Irradiation in Intense Radiation Fields, When Subjected to Very High TemperaturesOhio State University
Non Destructive Thermal Analysis and In Situ Investigation of Creep Mechanism of Graphite and Ceramic Composites using Phase-sensitive THz Imaging & Nonlinear Resonant Ultrasonic SpectroscopyRensselaer Polytechnic Institute
A Distributed Fiber Optic Sensor Network for Online 3-D Temperature and Neutron Fluence Mapping in a VHTR EnvironmentTexas A&M University
CFD Model Development and Validation for High Temperature Gas Cooled Reactor Cavity Cooling System (RCCS) ApplicationsTexas A&M University
Investigation on the Core Bypass Flow in a Very High Temperature ReactorTexas A&M University
Study of Air ingress across the duct during the accident conditionsTexas A&M University
Verification of the CENTRM Module for Adaptation of the SCALE Code to NGNP Prismatic and PBR Core DesignsUniversity of Arizona
Integral and Separate Effects Tests for Thermal Hydraulics Code Validationfor Liquid-Salt Cooled Nuclear ReactorsUniversity of California, Berkeley
Mechanisms Governing the Creep Behavior of High Temperature Alloys for Generation IV Nuclear Energy SystemsUniversity of Cincinnati
ALD Produced B2O3, Al2O3 and TiO2 Coatings on Gd2O3 Burnable Poison NanoparticlesUniversity of Colorado, Boulder
Experimental Study and Computational Simulations of Key Pebble Bed Thermomechanics Issues for Design and SafetyUniversity of Idaho
Prediction and Monitoring Systems of Creep-Fracture Behavior of 9Cr-1Mo Steels for Reactor Pressure VesselsUniversity of Idaho
Understanding Fundamental Material Degradation Processes in High Temperature Aggressive Chemomechanical EnvironmentsUniversity of Illinois, Urbana-Champaign
Corrosion and Creep of Candidate Alloys in High Temperature Helium and Steam Environments for the NGNPUniversity of Michigan
Creation of a Full-Core HTR Benchmark with the Fort St. Vrain Initial Core and Validation of the DHF Method with Helios for NGNP ConfigurationsUniversity of Michigan
Multi- Scale Multi-physics Methods Development for the Calculation of Hot-Spots in the NGNPUniversity of Michigan
Fission Product Sorptivity in GraphiteUniversity of Missouri, Columbia
Identifying and Understanding Environment-Induced Crack Propagation Behavior in Ni-Based Superalloy INCONEL 617University of Nevada, Las Vegas
Graphite Oxidation Simulation in HTR Accident ConditionsUniversity of New Mexico
Tritium Sequestration in Gen IV NGNP Gas Stream via Proton-Conducting Ceramic PumpsUniversity of South Carolina
Assessment of Embrittlement of VHTR Structural Alloys in Impure Helium EnvironmentsUniversity of Wisconsin, Madison
Experimental Studies of NGNP Reactor Cavity Cooling System with WaterUniversity of Wisconsin, Madison
Liquid Salt Heat Exchanger Technology for VHTR Based ApplicationsUniversity of Wisconsin, Madison
Materials, Turbomachinery and Heat Exchangers for Supercritical CO2 SystemsUniversity of Wisconsin, Madison
Modeling Fission Product Sorption in Graphite StructuresUniversity of Wisconsin, Madison
Effect of Post-Weld Heat Treatment on Creep Rupture Properties of Grade 91 Steel Heavy Section WeldsUtah State University

Investigator-Initiated Research (IIR)

Light Water Reactor Sustainability (LWRS)

Light Water Reactor Sustainability (LWRS)
Advanced Models of LWR Pressure Vessel Embrittlement for Low Flux-HighFluence ConditionsUniversity of California, Santa Barbara