OOIR: Observatory of International Research

Papers

(The median citation count of Nuclear Science and Engineering is 1. The table below lists those papers that are above that threshold based on CrossRef citation counts [max. 250 papers]. The publications cover those that have been published in the past four years, i.e., from 2020-04-01 to 2024-04-01.)

Article	Citations
Data-Enabled Physics-Informed Machine Learning for Reduced-Order Modeling Digital Twin: Application to Nuclear Reactor Physics	34
SAPIUM: A Generic Framework for a Practical and Transparent Quantification of Thermal-Hydraulic Code Model Input Uncertainty	24
Geant4 Tracks of NaI Cubic Detector Peak Efficiency, Including Coincidence Summing Correction for Rectangular Sources	14
Prediction of Neutronics Parameters Within a Two-Dimensional Reflective PWR Assembly Using Deep Learning	13
Wall-Climbing Robot with Active Sealing for Radiation Safety of Nuclear Power Plants	11
Separate-Effects Tests for Studying Temperature-Gradient-Driven Cracking in UO₂ Pellets	10
A New Era of Nuclear Criticality Experiments: The First 10 Years of Planet Operations at NCERC	9
A New Era of Nuclear Criticality Experiments: The First 10 Years of Godiva IV Operations at NCERC	8
Multigroup Constant Calculation with Static α-Eigenvalue Monte Carlo for Time-Dependent Neutron Transport Simulations	8
Optimization of Beta Radioluminescent Batteries with Different Radioisotopes: A Theoretical Study	8
Th-U Breeding Performances in an Optimized Molten Chloride Salt Fast Reactor	8
Progress Toward Simulating Departure from Nucleate Boiling at High-Pressure Applications with Selected Wall Boiling Closures	8
Applicability of Dynamic Mode Decomposition to Estimate Fundamental Mode Component of Prompt Neutron Decay Constant from Experimental Data	8
Reactor Physics Benchmark of the First Criticality in the Molten Salt Reactor Experiment	7
The Versatile Test Reactor Project: Mission, Requirements, and Description	7
Reduced-Order Modeling of Nuclear Reactor Kinetics Using Proper Generalized Decomposition	7
A New Era of Nuclear Criticality Experiments: The First 10 Years of Radiation Test Object Operations at NCERC	7
Development of a Sodium Fast Reactor Cartridge Loop Testing Capability for the Versatile Test Reactor	7
Prediction of the Power Peaking Factor in a Boron-Free Small Modular Reactor Based on a Support Vector Regression Model and Control Rod Bank Positions	7
Optimal Batch Size Growth for Wielandt Method and Superhistory Method	7
Optimized Separative Power of Hyperspeed Iguassu Gas Centrifuge: Dependence on the Rotor Diameter and Velocity	7
Nonmatching Discontinuous Cartesian Grid Algorithm Based on the Multilevel Octree Architecture for Discrete Ordinates Transport Calculation	6
Numerical Simulation on Asymmetrical Three-Dimensional Thermal and Hydraulic Characteristics of the Primary Sodium Pool Under the Pump Stuck Accident in CEFR	6
Post-Neutron Mass Yield Distribution in the Epi-Cadmium Neutron-Induced Fission of ²⁴¹Am	6
Investigation of the Influence of TeO₂ on the Elastic and Radiation Shielding Capabilities of Phospho-Tellurite Glasses Doped With Sm₂O₃	6

Post-Neutron Mass Yield Distribution in the Epi-Cadmium Neutron-Induced Fission of ²⁴³Am	6
A New Era of Nuclear Criticality Experiments: The First 10 Years of Flattop Operations at NCERC	6
Multiphysics Coupling Methods for Molten Salt Reactor Modeling and Simulation in VERA	6
Reactor Physics Considerations for Use of Yttrium Hydride Moderator	6
Conceptual Design of the Transformational Challenge Reactor	6
Generation of the Thermal Scattering Law of Uranium Dioxide with Ab Initio Lattice Dynamics to Capture Crystal Binding Effects on Neutron Interactions	5
Probabilistic Seismic Demand Model and Seismic Fragility Analysis of NPP Equipment Subjected to High- and Low-Frequency Earthquakes	5
Reactivity Feedback Effect on Supercritical Transient Analysis of Fuel Debris	5
Scaling Analysis of Thermal-Hydraulic Integral Systems: Insights from Practical Applications and Recent Advancements	5
A Code-Agnostic Driver Application for Coupled Neutronics and Thermal-Hydraulic Simulations	5
Uncertainty Quantification of Lead and Bismuth Sample Reactivity Worth at Kyoto University Critical Assembly	5
Post-Neutron Mass Yield Distribution in the Epi-Cadmium Neutron-Induced Fission of²⁴⁵Cm	5
A New Era of Nuclear Criticality Experiments: The First 10 Years of Comet Operations at NCERC	5
Density Wave Instability Verification of CFD Two-Fluid Model	5
Modeling Interface Debonding in Coated Fuel Particles with BISON	5
A Robust, Relaxation-Free Multiphysics Iteration Scheme for CMFD-Accelerated Neutron Transport k-Eigenvalue Calculations—II: Numerical Results	5
Fuel Performance Design Basis for the Versatile Test Reactor	5
Neutronic Simulation of Fuel Assembly Vibrations in a Nuclear Reactor	5
Monte Carlo Criticality Calculation of Random Media Formed by Multimaterials Mixture Under Extreme Disorder	5
Modeling of Safety Basis Events in the VTR	5
Versatile Test Reactor Conceptual Core Design	4
The ICSCREAM Methodology: Identification of Penalizing Configurations in Computer Experiments Using Screening and Metamodel—Applications in Thermal Hydraulics	4
Post-Neutron Mass Yield Distribution in the Epi-Cadmium Neutron-Induced Fission of²²⁹Th	4
Diffusion Synthetic Acceleration for Heterogeneous Domains, Compatible with Voids	4
A Machine Learning Method for the Forensics Attribution of Separated Plutonium	4
Thermal Upscattering Acceleration Schemes for Parallel Transport Sweeps	4
A New Resonance Calculation Method Using Energy Expansion Based on a Reduced Order Model	4
A Linear Prolongating Coarse Mesh Finite Difference Acceleration of Discrete Ordinate Neutron Transport Calculation Based on Discontinuous Galerkin Finite Element Method	4
Enhancing lpCMFD Acceleration with Successive Overrelaxation for Neutron Transport Source Iteration	4
Application of Machine Learning Algorithms to Identify Problematic Nuclear Data	4
Assessment of nTRACER and PARCS Performance for VVER Configurations	4
Feasibility of Sodium-Cooled Breed-and-Burn Reactor with Rotational Fuel Shuffling	4
TRISO SiC Failure Probability for Reactivity Initiated Accidents in High-Temperature Gas-Cooled Reactors	4
Experimental and Numerical Investigation into Temperature Distribution of a Simulated PHWR Coolant Channel Under Heatup Condition	4
Validation and Uncertainty Quantification of Transient Reflood Models Using COBRA-TF and Machine Learning Techniques Based on the NRC/PSU RBHT Benchmark	4
Toward Asymptotic Diffusion Limit Preserving High-Order, Low-Order Method	4
Real-Time Monitoring for Detection of Adversarial Subtle Process Variations	4
Verification and Validation of RAPID Formulations and Algorithms Based on Dosimetry Measurements at the JSI TRIGA Mark-II Reactor	4
Neutron Balance Features in Breed-and-Burn Fast Reactors	4
Evaluation of Yttrium Hydride (δ-YH_2-x) Thermal Neutron Scattering Laws and Thermophysical Properties	4
Triangular Polynomial Expansion Nodal Method for VVER Core Analysis	4
Development of the Versatile Test Reactor Probabilistic Risk Assessment	4
Secondary-Source Core Reload Modeling with VERA	4
Transient Multilevel Scheme with One-Group CMFD Acceleration	4
A Perspective on Data-Driven Coarse Grid Modeling for System-Level Thermal Hydraulics	4
Consistent Transport Transient Solvers of the High-Fidelity Transport Code PROTEUS-MOC	4
Variable Dynamic Mode Decomposition for Estimating Time Eigenvalues in Nuclear Systems	4
Validation of Pin-Resolved Reaction Rates, Kinetics Parameters, and Linear Source MOC in MPACT	4
Methodology for Generating Covariance Data of Thermal Neutron Scattering Cross Sections	4
A Multiscale and Multiphysics PWR Safety Analysis at a Subchannel Scale	4

A Preliminary Study on the Use of the Linear Regression Method for Multigroup Cross-Section Interpretation	4
Post-Neutron Mass Yield Distribution in the Epi-Cadmium Neutron-Induced Fission of ²⁴¹ Pu	3
Physics-Informed Neural Network Method and Application to Nuclear Reactor Calculations: A Pilot Study	3
Experimental Study of Air-Steam–Mixture Condensation Underneath Containment Vessel Surface	3
High-Fidelity Simulation of Mixing Phenomena in Large Enclosures	3
Annular Flow Simulation Supported by Iterative In-Memory Mesh Adaptation	3
OpenFOAM-Hybrid: A Morphology Adaptive Multifield Two-Fluid Model	3
Post-Neutron Mass Yield Distribution in the Epi-Cadmium Neutron-Induced Fission of ²³⁷Np	3
Preliminary Analysis of Unprotected Transients in the VTR	3
Numerical Evaluation for Spacer Vane Effects on Flow and Heat Transfer of Water at Supercritical Pressure in Annular Channel	3
A Robust, Relaxation–Free Multiphysics Iteration Scheme for CMFD–Accelerated Neutron Transport k–Eigenvalue Calculations–I: Theory	3
Scaling, Passive Systems, and the AP-1000	3
Review of the Fluid Dynamics and Heat Transport Phenomena in Packed Pebble Bed Nuclear Reactors	3
Analysis of Population Control Techniques for Time-Dependent and Eigenvalue Monte Carlo Neutron Transport Calculations	3
Stress Profile in Coating Layers of TRISO Fuel Particles in Contact with One Another	3
Burnup Performance of CANDLE Burning Reactor Using Sodium Coolant	3
Jet Fragmentation Characteristics During Molten Fuel and Coolant Interactions	3
Simulated Performance of the Micro-Pocket Fission Detector in the Advanced Test Reactor Critical Facility	3
Direct Comparison of High-Order/Low-Order Transient Methods on the 2D-LRA Benchmark Problem	3
Parallel Approximate Ideal Restriction Multigrid for Solving the S_N Transport Equations	3
Individual Adjustment of Independent Cross-Section Set Based on Bayesian Theory	3
Generalized Equivalence Theory Used with Spatially Linear Sources in the Method of Characteristics for Neutron Transport	3
Effects of Vent Size and Wind on Dispersion of Hydrogen Leaked in a Partially Open Space: Studies by Numerical Analysis	3
An Analytic Benchmark for Neutron Boltzmann Transport with Downscattering—Part I: Flux and Eigenvalue Solutions	3
Metallic Fuel Performance Benchmarks for Versatile Test Reactor Applications	3
FLASSH 1.0: Thermal Scattering Law Evaluation and Cross-Section Generation for Reactor Physics Applications	3
Preconceptual Design of Multifunctional Gas-Cooled Cartridge Loop for the Versatile Test Reactor: Instrumentation and Measurement—Part II	3
Neutron Generation Time in Highly-Enriched Uranium Core at Kyoto University Critical Assembly	3
Gamma-Induced Degradation Effect of InP HBTs Studied by Keysight Model	3
Post-Neutron Mass Yield Distribution in the Spontaneous Fission of²⁵²Cf	3
Analytical Discrete Ordinates Solution for a 1D Model of Particle Transport in Ducts that Includes Wall Migration	3
Development of Conceptual Lead Cartridge Design to Perform Irradiation Experiments in VTR	3
Relative Speed Tabulation Method for Efficient Treatment of Resonance Scattering in GPU-Based Monte Carlo Neutron Transport Calculation	3
Uranium Extraction from Gattar Granite Sample After Leaching Using Nitrate Solution in Presence of Peroxide	3
Coupled Monte Carlo Transport and Conjugate Heat Transfer for Wire-Wrapped Bundles Within the MOOSE Framework	3
Integrated Safety and Security Analysis of Nuclear Power Plants Using Dynamic Event Trees	3
Research on a Monte Carlo Simulation Method of Neutron Coded-Aperture Imaging	3
An Analytic Benchmark for Neutron Boltzmann Transport with Downscattering—Part II: Flux and Eigenvalue Sensitivities to Nuclear Cross Sections and Resonance Parameters	3
Modeling and Estimation of Nuclear Reactor Performance Using Fractional Neutron Point Kinetics with Temperature Effect and Xenon Poisoning	3
Blind Benchmark Exercise for Spent Nuclear Fuel Decay Heat	3
Experimental and Computational Dose Rate Evaluation Using S_N and Monte Carlo Method for a Packaged ²⁴¹AmBe Neutron Source	3
Supercritical Transient Analysis for Ramp Reactivity Insertion Using Multiregion Integral Kinetics Code	2
Thermal Design and Experimental Verification of Double Helium Gap Conduction Test Facility	2
Gradient-Informed Design Optimization of Select Nuclear Systems	2
Compression of Cross-Section Data Size for High-Resolution Core Analysis Using Dimensionality Reduction Technique	2
RANS-Based CFD Calculation for Pressure Drop and Mass Flow Rate Distribution in an MTR Fuel Assembly	2
Frequency-Dependent Discrete Implicit Monte Carlo Scheme for the Radiative Transfer Equation	2
Streaming Effect of Void Reactivity in LWR Critical Experiments with Streaming Channel	2
A High-Assay Low-Enriched Uranium Fuel Transportation Concept	2
Serpent and MCNP Calculations of the Energy Deposition in the Transformational Challenge Reactor	2
Multidual Sensitivity Method in Ray-Tracing Transport Simulations	2
Frequency Transform Method for Transient Analysis of Nuclear Reactors in Monte Carlo	2
Post-Neutron Mass Yield Distribution in the Thermal Neutron Induced Fission of ²³³U	2
ROM-Based Surrogate Systems Modeling of EBR-II	2
Comparisons of Supercritical Loop Flow and Heat Transfer Behavior Under Uniform and Nonuniform High-Flux Heat Inputs	2
Transformational Challenge Reactor Safety Design and Radionuclide Retention Strategy	2
Application of MELCOR for Simulating Molten Salt Reactor Accident Source Terms	2
Transport Calculation of the Multiplicity Moments for Cylinders	2
Mechanism of Fission Neutron Emission: New Experimental Arguments	2
Quantification of Deep Neural Network Prediction Uncertainties for VVUQ of Machine Learning Models	2
Nuclear Reactor Power Level Model Predictive Control: A Consideration of Coolant Outlet Temperature Relaxation Tracking Method	2
Investigation on the Use of the Monte Carlo Iterative k-Source Scheme for the Study of Neutron Subcritical Multiplication	2
A Nuclear Decay Micropropulsion Technology Based on Spontaneous Alpha Decay	2
Pressurized Water Reactor Core Power Control Using BAS-RBF-PID Approach During Transient Operation	2
Study on LOFA and LOHS Accidents with Passive Safety System for Integrated Marine Reactor	2
Design and Control of a Fueled Molten Salt Cartridge Experiment for the Versatile Test Reactor	2
VTR Core Design Analyses Supporting Flexible Operations	2
Development of Sodium Fire Analysis Code Capabilities for Versatile Test Reactor	2
The Time-Dependent Asymptotic P_N Approximation for the Transport Equation	2
A Versatile Methodology for Reactor Pressure Vessel Aging Assessments	2
Reactor Core Power Distribution Reconstruction Method by Ex-Core Detectors Based on the Correlation Effect Between Fuel Regions	2
Deep Learning for Multigroup Cross-Section Representation in Two-Step Core Calculations	2
MOOSE Reactor Module: An Open-Source Capability for Meshing Nuclear Reactor Geometries	2
Benchmarking of the NCrystal SANS Plugin for Nanodiamonds	2
Euler-Euler Model of Bubbly Flow Using Particle-Center-Averaging Method	2
A Dynamic Risk Framework for the Physical Security of Nuclear Power Plants	2
Transition Core Modeling for Extended-Enrichment Accident-Tolerant Fuels in Light Water Reactors Using PARCS/Polaris	2
Two-Phase Turbulent Kinetic Energy Budget Computation in Co-Current Taylor Bubble Flow	2
Unstructured Mesh–Based Neutronics and Thermomechanics Coupled Steady-State Analysis on Advanced Three-Dimensional Fuel Elements with Monte Carlo Code iMC	2
Concrete Modeling for Neutron Transport and Associated Sensitivity Studies Performed at the AMANDE-MIRCOM Facility	2

A Newly Developed Suppression Pool Model Based on the ISAA Code	2
Modeling Reactor Noise due to Rod and Thermal Vibrations with Thermal Feedback Using Stochastic Differential Equations	2
A New Embedded Analysis with Pinwise Discontinuity Factors for Pin Power Reconstruction	2
The Legendre Polynomial Axial Expansion Method	2
Continuous-Energy Time-Dependent Coarse Mesh Transport (COMET) Method for Kinetics Calculations	2
Multilevel-in-Space-and-Energy CMFD in VERA	2
Multiphysics Analysis System for Heat Pipe–Cooled Micro Reactors Employing PRAGMA-OpenFOAM-ANLHTP	2
Post-Neutron Mass Yield Distribution in the Thermal Neutron–Induced Fission of ²³⁹Pu	2
High-Fidelity Neutron Transport Solution of High Temperature Gas-Cooled Reactor by Three-Dimensional Linear Source Method of Characteristics	2
Study of Different Seed Fuels with Thorium in Accelerator-Driven Subcritical System	2
Moment Matching: A New Optimization-Based Sampling Scheme for Uncertainty Quantification of Reactor-Physics Analysis	2
A New Proof of the Asymptotic Diffusion Limit of the S_N Neutron Transport Equation	2
Design Optimization of the Transformational Challenge Reactor Outlet Plenum	2
A Nonintrusive Nuclear Data Uncertainty Propagation Study for the ARC Fusion Reactor Design	2
Nonlinear Elimination Applied to Radiation Diffusion	2
State-of-the-Art in Evaluation Approaches for Risk Assessment of Insider Threats to Nuclear Facility Physical Protection Systems	2
A Multiscale Approach Simulating Generic Pool Boiling	2
Safety Analysis in VVER-1000 Due to Large-Break Loss-of-Coolant Accident and Station Blackout Transient Using RELAP5/SCDAPSIM/MOD3.5	2
Effect of Moderation Condition on Neutron Multiplication Factor Distribution in 1/f^β Random Media	2
Analyzing APR1400 System Response Under Load Follow Operation Using a Multiphysics Approach	2
The Finite-Element with Discontiguous-Support Method	2
Impact of the Melt-Refining Process on the Performance of Sodium-Cooled Rotational Fuel-Shuffling Breed-and-Burn Reactors	1
Evaluating Quantities of Interest Other Than Nuclide Densities in the Bateman Equations	1
Development of Uncertainty Analysis Techniques for the Fission Matrix–Based Neutron Transport Code RAPID	1
Theoretical Derivation of a Unique Combination Number Hidden in the Higher-Order Neutron Correlation Factors Using the Pál-Bell Equation	1
Post-Neutron Mass Yield Distribution in the Thermal Neutron–Induced Fission of ²³⁵U	1
A CFD Modeling Coupled with VOF Method and Solidification Model for Molten Jet Breakup at Low Velocity	1
The NILO-CMFD Method for Iteratively Solving Coupled Neutron Transport–Thermal Hydraulics Problems	1
Design of Experiments to Support VTR Core Design	1
Investigations of Gamma and Neutron Shielding Parameters of Ti-Nb-Fe-Cr Alloys with Varying Cr Concentrations	1
Validation and Optimization of Activity Estimates of the FiR 1 TRIGA Research Reactor Biological Shield Concrete	1
Enhanced Cooling Characteristics of the Cylindrical Cooling Tube Using the Inserted Helical Wire Coil Based on Finite Element Analysis	1
Asymptotic Expansion of the Angular Flux Applied to Discrete-Ordinates Source Iterations in Lattice Depletion Calculations	1
Beta-Ray-Bremsstrahlung Contributions to Short-Lived Delayed Photoneutron Groups in Heavy Water Reactors	1
Study of Stable Stratification in HiRJET Facility With Direct Numerical Simulation	1
Scoping Studies for a Lead-Lithium-Cooled, Minor-Actinide-Burning, Fission-Fusion Hybrid Reactor Design	1
An Improved Dual Asymmetric Penalized Least Squares Baseline Correction Method for High-Noise Spectral Data Analysis	1
Physics-Informed Neural Network with Fourier Features for Radiation Transport in Heterogeneous Media	1
Modification of a Dynamic Monte Carlo Technique to Simplify and Accelerate Transient Analysis with Feedback	1
Extended Applications of Subgrid Representation in the 2D/1D Method	1
Experimental Investigation of the Friction Factor of Wire-Wrapped Bundles at Low Velocities Under Natural Circulation Flow Conditions	1
Solving Burnup Equations by Numerical Inversion of the Laplace Transform Using Padé Rational Approximation	1
On the Numerical Method for Photofission-Based Nuclear Material Isotopic Composition Estimation in Thorium-Uranium Systems	1
Nuclear Data–Induced Uncertainty of Criticality in Solid-Moderated and Solid-Reflected Cores with Highly Enriched Uranium Fuel at Kyoto University Critical Assembly	1
A Multigroup Homogeneous Flux Reconstruction Method Based on the ANOVA-HDMR Decomposition	1
Improved Methods for Calculating the Neutron Diffusion Coefficient of General Pebble Bed	1
Direct Numerical Simulation of Heat Transfer in a 7-Pin Wire-Wrapped Rod Bundle	1
Estimation of Absorbed Dose due to Gas Bremsstrahlung Based on Residual Gas in Electron Storage Rings	1
The HighNESS Project at the European Spallation Source: Current Status and Future Perspectives	1
Analytic Error Analysis of the Partial Derivatives Cross-Section Model—II: Numerical Results	1
Advances in Nuclear Data and Software Development for the HighNESS Project	1
Digital Engineering for Integrated Modeling and Simulation for Building-Piping Systems Through Interoperability Solutions	1
TRISO Burnup-Dependent Failure Analysis of a HTGR Design Basis Accident Using BISON	1
Sensitivity Analysis of Effectiveness of FLEX Strategies Using a CDF-Based Importance Measure Under Accident Conditions	1
Generation of Optimal Weight Values Based on the Recursive Monte Carlo Method for Use in Monte Carlo Deep Penetration Calculations	1
Measurements of Neutron Scattering from a Copper Sample Using a Quasi-Differential Method in the Region from 2 keV to 20 MeV	1
Deceptive Infusion of Data: A Novel Data Masking Paradigm for High-Valued Systems	1
Selected papers from the 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18)	1
Parametric Model-Order Reduction for Radiation Transport Simulations Based on an Affine Decomposition of the Operators	1
Design and Thermal-Hydraulic Performance Analysis of a Low-Temperature Heat Pipe Test Facility	1
On the Regularity Order of the Pointwise Uncollided Angular Flux and Asymptotic Convergence of the Discrete Ordinates Approximation of the Scalar Flux	1
Pressure Drop in Seven-Pin Wire-Wrapped Rod Bundle for the Sodium Cartridge Loop in Versatile Test Reactor	1
Radiolytic Production of Fluorine Gas from MSR Relevant Fluoride Salts	1
Nuclear Criticality Safety Aspects for the Future of HALEU: Evaluating Heterogeneity in Intermediate-Enrichment Uranium Using Critical Benchmark Experiments	1
Introduction of the Adding and Doubling Method for Solving Bateman Equations for Nuclear Fuel Depletion	1
Experimental Analysis of Fast-Transient Flow Boiling in Various Flow Conditions Using Inverse Heat Conduction Calculation Method	1
Variance Reduction and Noise Source Sampling Techniques for Monte Carlo Simulations of Neutron Noise Induced by Mechanical Vibrations	1
A Study of Nuclear Fuel Burnup Wave Development in a Fast Neutron Energy Spectrum Multiplying Medium: Improved Model and Consistent Parametric Approach for Evaluation	1
Improvements in Computational Efficiency for Resonance Calculation Using Energy Spectrum Expansion Method	1
A Unified Framework of Stabilized Finite Element Methods for Solving the Boltzmann Transport Equation	1
¹⁴¹Pr(α,x): New Cross-Section Data With Special Reference to ¹⁴⁰Nd Production for Medicine	1
Density Wave Instability Verification of 1-D Computational Two-Fluid Model	1
Impact Assessment of Radiative Heat Transport in ARC-Class Reactor FLiBe Liquid Immersion Blanket	1
A Residual-Based Subgrid-Scale Method Combined with Spherical Harmonics Angular Approximation for Solving the Boltzmann Transport Equation	1
Phonon Sampling Method for Inelastic Thermal Neutron Scattering Events	1
Development and Validation of Thermal-Mechanical Creep Failure Module for Reactor Pressure Vessel Lower Head	1
Nuclear Data Uncertainty Propagation for the Molten Salt Fast Reactor Design	1
Uncertainty Quantification Approach for the Versatile Test Reactor Core Design	1
Cross Sections for Neutron Production from 6- and 10-MeV Neutrons Incident on ¹⁰B and ¹¹B	1
Sensitivity Coefficient Evaluation of an Accelerator-Driven System Using ROM-Lasso Method	1
Monte Carlo Study of an Electron-Based Neutron Source for Bragg Edge Imaging	1
Acceleration Waves in Cylindrical Shrinking Gas Bubbles	1
Current State of the Design Engineering of the Versatile Test Reactor Plant	1
Benchmark Experiments for Turbulent Mixing in the Scaled-Down Upper Plenum of High-Temperature Gas-Cooled Reactors Under Accident Scenario	1
Radiation Dose Assessment of Tritium Released from the Thorium Molten Salt Reactor	1
Effects of Carbon on Void Nucleation in Self-Ion–Irradiated Pure Iron	1
Response Matrix/Discrete Ordinates Solution of the 1D Fokker-Planck Equation	1
A Critical Review of Heat Pipe Experiments in Nuclear Energy Applications	1
Development of an Activity Transport Model and In Situ Measurements on Deposited Radionuclides in the Primary Coolant Pipelines of the Chinshan Nuclear Power Plant	1
Strategies for Fast Fission Matrix Estimation with Fuel Temperature and Control Rod Feedback	1
Multigroup Neutron Transport Using a Collision-Based Hybrid Method	1
Numerical Investigations of Molten Salt Pump Journal Bearings Under Hydrodynamic Lubrication Conditions for FHRs	1
Preliminary Neutronics Design and Analysis of the Fast Modular Reactor	1
Pumping Options for Versatile Test Reactor Molten Lead In-Pile Test Cartridge	1
High-Fidelity Simulations of Shield Assembly Mixed-Convection Flows with Applications Toward Reduced-Resolution Modeling	1