Nuclear Science and Engineering

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
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
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
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
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
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
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
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
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
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-YH2-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
Annular Flow Simulation Supported by Iterative In-Memory Mesh Adaptation	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
Experimental and Computational Dose Rate Evaluation Using SN and Monte Carlo Method for a Packaged 241AmBe Neutron Source	3
Analysis of Population Control Techniques for Time-Dependent and Eigenvalue Monte Carlo Neutron Transport Calculations	3
Physics-Informed Neural Network Method and Application to Nuclear Reactor Calculations: A Pilot Study	3
Burnup Performance of CANDLE Burning Reactor Using Sodium Coolant	3
High-Fidelity Simulation of Mixing Phenomena in Large Enclosures	3
Simulated Performance of the Micro-Pocket Fission Detector in the Advanced Test Reactor Critical Facility	3
OpenFOAM-Hybrid: A Morphology Adaptive Multifield Two-Fluid Model	3
Parallel Approximate Ideal Restriction Multigrid for Solving the SN 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
Review of the Fluid Dynamics and Heat Transport Phenomena in Packed Pebble Bed Nuclear Reactors	3
FLASSH 1.0: Thermal Scattering Law Evaluation and Cross-Section Generation for Reactor Physics Applications	3
Stress Profile in Coating Layers of TRISO Fuel Particles in Contact with One Another	3
Neutron Generation Time in Highly-Enriched Uranium Core at Kyoto University Critical Assembly	3
Jet Fragmentation Characteristics During Molten Fuel and Coolant Interactions	3
Post-Neutron Mass Yield Distribution in the Spontaneous Fission of252Cf	3
Direct Comparison of High-Order/Low-Order Transient Methods on the 2D-LRA Benchmark Problem	3
Development of Conceptual Lead Cartridge Design to Perform Irradiation Experiments in VTR	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
Preconceptual Design of Multifunctional Gas-Cooled Cartridge Loop for the Versatile Test Reactor: Instrumentation and Measurement—Part II	3
Blind Benchmark Exercise for Spent Nuclear Fuel Decay Heat	3
Gamma-Induced Degradation Effect of InP HBTs Studied by Keysight Model	3
Post-Neutron Mass Yield Distribution in the Epi-Cadmium Neutron-Induced Fission of 241 Pu	3
Analytical Discrete Ordinates Solution for a 1D Model of Particle Transport in Ducts that Includes Wall Migration	3
Experimental Study of Air-Steam–Mixture Condensation Underneath Containment Vessel Surface	3
Relative Speed Tabulation Method for Efficient Treatment of Resonance Scattering in GPU-Based Monte Carlo Neutron Transport Calculation	3
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
Post-Neutron Mass Yield Distribution in the Thermal Neutron Induced Fission of 233U	2
Frequency Transform Method for Transient Analysis of Nuclear Reactors in Monte Carlo	2
Comparisons of Supercritical Loop Flow and Heat Transfer Behavior Under Uniform and Nonuniform High-Flux Heat Inputs	2
ROM-Based Surrogate Systems Modeling of EBR-II	2
Transformational Challenge Reactor Safety Design and Radionuclide Retention Strategy	2
Transport Calculation of the Multiplicity Moments for Cylinders	2
Application of MELCOR for Simulating Molten Salt Reactor Accident Source Terms	2
Quantification of Deep Neural Network Prediction Uncertainties for VVUQ of Machine Learning Models	2
Mechanism of Fission Neutron Emission: New Experimental Arguments	2
Investigation on the Use of the Monte Carlo Iterative k-Source Scheme for the Study of Neutron Subcritical Multiplication	2
Nuclear Reactor Power Level Model Predictive Control: A Consideration of Coolant Outlet Temperature Relaxation Tracking Method	2
Pressurized Water Reactor Core Power Control Using BAS-RBF-PID Approach During Transient Operation	2
A Nuclear Decay Micropropulsion Technology Based on Spontaneous Alpha Decay	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 PN Approximation for the Transport Equation	2
A Versatile Methodology for Reactor Pressure Vessel Aging Assessments	2
Deep Learning for Multigroup Cross-Section Representation in Two-Step Core Calculations	2
Reactor Core Power Distribution Reconstruction Method by Ex-Core Detectors Based on the Correlation Effect Between Fuel Regions	2
Benchmarking of the NCrystal SANS Plugin for Nanodiamonds	2
MOOSE Reactor Module: An Open-Source Capability for Meshing Nuclear Reactor Geometries	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
Two-Phase Turbulent Kinetic Energy Budget Computation in Co-Current Taylor Bubble Flow	2
Transition Core Modeling for Extended-Enrichment Accident-Tolerant Fuels in Light Water Reactors Using PARCS/Polaris	2
Concrete Modeling for Neutron Transport and Associated Sensitivity Studies Performed at the AMANDE-MIRCOM Facility	2
Unstructured Mesh–Based Neutronics and Thermomechanics Coupled Steady-State Analysis on Advanced Three-Dimensional Fuel Elements with Monte Carlo Code iMC	2
Modeling Reactor Noise due to Rod and Thermal Vibrations with Thermal Feedback Using Stochastic Differential Equations	2
A Newly Developed Suppression Pool Model Based on the ISAA Code	2
The Legendre Polynomial Axial Expansion Method	2
A New Embedded Analysis with Pinwise Discontinuity Factors for Pin Power Reconstruction	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
A New Proof of the Asymptotic Diffusion Limit of the SN Neutron Transport Equation	2
Moment Matching: A New Optimization-Based Sampling Scheme for Uncertainty Quantification of Reactor-Physics Analysis	2
A Nonintrusive Nuclear Data Uncertainty Propagation Study for the ARC Fusion Reactor Design	2
Design Optimization of the Transformational Challenge Reactor Outlet Plenum	2
Nonlinear Elimination Applied to Radiation Diffusion	2
A Multiscale Approach Simulating Generic Pool Boiling	2
State-of-the-Art in Evaluation Approaches for Risk Assessment of Insider Threats to Nuclear Facility Physical Protection Systems	2
Effect of Moderation Condition on Neutron Multiplication Factor Distribution in 1/fβ Random Media	2
Safety Analysis in VVER-1000 Due to Large-Break Loss-of-Coolant Accident and Station Blackout Transient Using RELAP5/SCDAPSIM/MOD3.5	2
The Finite-Element with Discontiguous-Support Method	2
Analyzing APR1400 System Response Under Load Follow Operation Using a Multiphysics Approach	2
Thermal Design and Experimental Verification of Double Helium Gap Conduction Test Facility	2
Supercritical Transient Analysis for Ramp Reactivity Insertion Using Multiregion Integral Kinetics Code	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