NUCLEAR SCIENCE AND ENGINEERING

Benchmarks of Criticality in Solid-Moderated and Solid-Reflected Cores at Kyoto University Critical Assembly

A Note on the Nomenclature in Neutron Multiplicity Mathematics

Uncertainty Propagation in Neutron Activation Cross-Section Measurement Using Unscented Transformation Method

Review of Hybrid Methods for Deep-Penetration Neutron Transport

Gaussian Process–Based Inverse Uncertainty Quantification for TRACE Physical Model Parameters Using Steady-State PSBT Benchmark

Dynamic Mode Decomposition for Subcritical Metal Systems

A Coupled Multiscale Approach to TREAT LEU Feedback Modeling Using a Binary-Collision Monte-Carlo–Informed Heat Source

Application of Linear Combination Method to Pulsed Neutron Source Measurement at Kyoto University Critical Assembly

Constraint Annealing Method for Solution of Multiconstrained Nuclear Fuel Cycle Optimization Problems

Acceleration of the Power Method with Dynamic Mode Decomposition

Heat Loss Simulations in Rod Bundle Tests

The Asymptotic Diffusion Limit of Numerical Schemes for the SN Transport Equation

Improvement of the VVER-1200 Fuel Cycle by Introducing Thorium with Different Fissile Material in Blanket-Seed Assembly

Coupling Methods for HTR-PM Multicircuit Problem

Integral Experiments on Critical Irradiation of 237Np and 241Am Foils at Kyoto University Critical Assembly

A Multilevel in Space and Energy Solver for 3-D Multigroup Diffusion and Coarse-Mesh Finite Difference Eigenvalue Problems

An Experimental Study of Local Self-Similarity in the Mixing Transition of a Turbulent Free Jet

An Estimation of Volatile Iodine in a Pool at Low pH and High Iodide Concentrations Under Irradiation

A Diffusion Monte Carlo–Based Algorithm for Estimation of Higher Modes of a Reactor

Utilization of Regionwise Even-Parity Discontinuity Factor to Reduce Discretization Error of MOC

Large-Scale Application of the Constraint Annealing Method for Pressurized Water Reactor Core Design Optimization

Estimating Code Biases for Criticality Safety Applications with Few Relevant Benchmarks

Compression of Assembly Power Form Factors for Core Calculations

A Review of Void Drift Models in Subchannel Analysis

Coupled Neutron and Gamma Heating Calculation Based on VARIANT Transport Solutions

Validation and Verification of the Evaluated Electron Data Library in FRENSIE

Key Results from Detailed Nondestructive Examinations of 25 Pressurized Water Reactor High Burnup Spent Nuclear Fuel Rods

A New Approach to Monte Carlo in High Temperature Reactors

Development of the Assembly-Level Monte Carlo Neutron Transport Code M3C for Reactor Physics Calculations

Reactor Kinetics Using Partial Equations

Investigation of Neutron Cross Section for Iron in the ENDF Library with Pulsed Sphere Measurements

Experimental Investigation of Natural Circulation During Air-Ingress Scenario in a Very High Temperature Reactor

Scoping Studies of Dopants for Stabilization of Uranium Nitride Fuel

On CCFL at a PWR Hot-Leg Pipe Geometry and Comparison Between Results in COLLIDER 1/4th- and Kobe 1/15th-Scaled Facilities

Accelerating Monte Carlo Shielding Calculations in TRIPOLI-4 with a Deterministic Adjoint Flux

A Directional Detector Response Function for Anisotropic Detectors

Fuzzy Logics as an Integral Part of Evolutionary Algorithms

Coupling of Neutronics and Thermal-Hydraulic Codes for Simulation of the MNSR Reactor

The Roussopoulos and Schwinger Functionals for Nuclear Systems Involving Imprecisely Known Fluxes and Parameters: Distinctions and Equivalences

Approximate Methods for Inverting Generating Functions from the Pál-Bell Equations for Low Source Problems

Effects of Gamma Irradiation on the Morphological, Physical, and Thermal Properties of B4C/CF/PI/AA6061 Hybrid Composite Laminates

Assessment of nTRACER and PARCS Performance for VVER Configurations

Numerical and Experimental Study on the Air Convective Heat Transfer Characteristics in Large-Scale Rectangular Channel with Asymmetric Heating Considering the Influence of Entry Form

Tantalum, Titanium, and Zirconium Neutron Total Cross-Section Measurements from 0.4 to 25 MeV

Validation of the Interfacial Area Transport Equation Coupled with the Void Transport Equation for Prediction of Flashing Flows

Alternative Adjoint Boltzmann Transport Operators in Alternative Hilbert Spaces in Spherical Geometry: Theory and Paradigm Applications

Modeling for Late-Phase Hydrogen Generation in a Degraded Core