NUCLEAR TECHNOLOGY

TREAT Restart Project

Fluoride-Salt-Cooled High-Temperature Test Reactor Thermal-Hydraulic Licensing and Uncertainty Propagation Analysis

A Second Generation Multiphase-CFD Framework Toward Predictive Modeling of DNB

Conceptual Design of a Pilot-Scale Pyroprocessing Facility

Safety Strategy and Update of the TREAT Facility Safety Basis

Transient Reactor Test Facility Advanced Transient Shapes

FHR, HTGR, and MSR Pebble-Bed Reactors with Multiple Pebble Sizes for Fuel Management and Coolant Cleanup

Response Matrix Method–Based Importance Solver and Variance Reduction Scheme in the Serpent 2 Monte Carlo Code

Adsorption and Desorption of Tritium in Nuclear Graphite at 700°C: A Gas Chromatographic Study Using Hydrogen

Transient Reactor Test (TREAT) Facility Initial Approach to Restart Criticality Following Extended Standby Operation

Transient Reactor Test (TREAT) Facility Restart Heat Balance Calibration and Rod Worth Measurements

Enhancing Electric Grid, Critical Infrastructure, and Societal Resilience with Resilient Nuclear Power Plants (rNPPs)

Tritium Transport Analysis in a 2-MW Liquid-Fueled Molten Salt Experimental Reactor with the Code TMSR-TTAC

Multiscaled Experimental Investigations of Corrosion and Precipitation Processes After Loss-of-Coolant Accidents in Pressurized Water Reactors

Fluoride-Salt-Cooled High-Temperature Reactor (FHR) Using British Advanced Gas-Cooled Reactor (AGR) Refueling Technology and Decay Heat Removal Systems That Prevent Salt Freezing

SHAKE-RoverD Framework for Nuclear Power Plants: A Streamlined Approach for Seismic Risk Assessment

Rod Bundle Heat Transfer Thermal-Hydraulic Program

Restarting the Transient Reactor Test (TREAT) Facility Reactor for Nuclear Transient Testing Science: A Special Issue of Nuclear Technology

Design and Characterization of a Scintillator-Based Position-Sensitive Detector for Muon Imaging

A New Two-Step Methodology for CANDU Analysis with the APEC-Assisted Nodal Equivalence Theory

The CASUALIDAD Method for Uncertainty Evaluation of Best-Estimate System Thermal-Hydraulic Calculations

Assessment of the Subchannel Code CTF for Single- and Two-Phase Flows

Design and Optimization of a Fast Neutron Radiography System Based on a High-Intensity D-T Fusion Neutron Generator

A Study of a Dose Constraint for Members of the Public Living Around Nuclear Power Plants in the United States

Multigroup Cross-Section Generation with the OpenMC Monte Carlo Particle Transport Code

Automated Identification of Causal Relationships in Nuclear Power Plant Event Reports

Minor Actinide Transmutation in Supercritical-CO2-Cooled and Sodium-Cooled Fast Reactors with Low Burnup Reactivity Swings

Electronic Stopping Powers of Formvar and Mylar Polymeric Materials for Heavy Ions: LSS Modified Theory and Monte Carlo Simulation

Thermophysical Properties of Stainless Steel Containing 5 mass % B4C in the Solid Phase

Quantile Regressions Applied to Pressurized Water Reactor CHF Correlation Safety Criteria

Design, Fabrication, and Testing of Electrolytic Cell for Minimizing Simulated Active Waste

Sensitivity Analysis of LB-LOCA in Response to Proposed 10 CFR 50.46c New Rulemaking

Machine-Learning Analysis of Moisture Carryover in Boiling Water Reactors

Foreword: Special issue on Restarting the Transient Reactor Test Facility

Modeling of Aerosol Fission Product Scrubbing in Experiments and in Integral Severe Accident Scenarios

Implementation of a Spacer Grid Rod Thermal-Hydraulic Reconstruction (ROTHCON) Capability into the Thermal-Hydraulic Subchannel Code CTF

Uncertainty Analysis of the Conditional Exceedance Probability Calculation for a Probabilistically Significant SBO Sequence

Methods of Data Collection in Nuclear Power Plants

Commentary: Nuclear Energy for Economic Variable Electricity: Replacing the Role of Fossil Fuels

Comparison of the Subchannel Code CTF to Steady-State and Transient Heat Transfer Experiments at Intermediate Pressures in Water

Evaluation of Non-Oxide Fuel for Fission-Based Nuclear Reactors on Spacecraft

A Data Model for Nuclear Power Plant Work Packages

Development of Unstructured Mesh-Based Numerical Method for Sodium-Water Reaction Phenomenon

The Effects of Plutonium Content and Self-Irradiation on Thermal Conductivity of Mixed Oxide Fuel

A Mutual Information–Based Experimental Design Framework to Use High-Fidelity Nuclear Reactor Codes to Calibrate Low-Fidelity Codes

Extraction of 137Cs from Acidic Feed by Centrifugal Contactors Using a Solution of Calix[4]arene-bis-1,2-benzo-crown-6 in Phenyltrifluoromethyl Sulphone

GPT-Free Sensitivity Analysis for Monte Carlo Models