| Study on characteristics of water film breakdown on the corrugated plate wall under the horizontal shear of airflow |
15 |
| Direct numerical simulation of reactor two-phase flows enabled by high-performance computing |
13 |
| High-fidelity PIV measurement of cross flow in 5 x 5 rod bundle with mixing vane grids |
13 |
| Judgement of critical state of water film rupture on corrugated plate wall based on SIFT feature selection algorithm and SVM classification method |
13 |
| CFD simulation of swirling flow induced by twist vanes in a rod bundle |
12 |
| Surface wettability and pool boiling Critical Heat Flux of Accident Tolerant Fuel cladding-FeCrAl alloys |
12 |
| Inverse uncertainty quantification using the modular Bayesian approach based on Gaussian process, Part 1: Theory |
12 |
| A review of sub-channel thermal hydraulic codes for nuclear reactor core and future directions |
11 |
| Eulerian modelling of turbulent bubbly flow based on a baseline closure concept |
10 |
| Initial results from safety testing of US AGR-2 irradiation test fuel |
10 |
| A novel multiphase MPS algorithm for modeling crust formation by highly viscous fluid for simulating corium spreading |
10 |
| Inverse uncertainty quantification using the modular Bayesian approach based on Gaussian Process, Part 2: Application to TRACE |
10 |
| The upgraded Cheng and Todreas correlation for pressure drop in hexagonal wire-wrapped rod bundles |
10 |
| Effect of temperature on fretting wear behavior and mechanism of alloy 690 in water |
10 |
| Wire-mesh sensors: A review of methods and uncertainty in multiphase flows relative to other measurement techniques |
10 |
| Heat transfer experiment in a partially (internally) blocked 19-rod bundle with wire spacers cooled by LBE |
9 |
| SARAX: A new code for fast reactor analysis part II: Verification, validation and uncertainty quantification |
9 |
| Study on the self-shielding and temperature influences on the neutron irradiation damage calculations in reactors |
9 |
| Experimental and numerical investigation on the pressure pulsation and instantaneous flow structure in a nuclear reactor coolant pump |
8 |
| Development and verification of PWR-core fuel management calculation code system NECP-Bamboo: Part I Bamboo-Lattice |
8 |
| Molten salt reactor waste and effluent management strategies: A review |
8 |
| Study on safety boundary of flow instability and CHF for parallel channels in motion |
8 |
| Signal response of wire-mesh sensors to an idealized bubbly flow |
8 |
| The path-planning in radioactive environment of nuclear facilities using an improved particle swarm optimization algorithm |
8 |
| A review of methods to predict the effective thermal conductivity of packed pebble beds, with emphasis on the near-wall region |
8 |
| Small modular reactor core design for civil marine propulsion using micro-heterogeneous duplex fuel. Part I: Assembly-level analysis |
8 |
| Seismic capacity and fragility analysis of an ASR-affected nuclear containment vessel structure |
8 |
| Development of an SP3 neutron transport solver for the analysis of the Molten Salt Fast Reactor |
8 |
| Multiphysics modeling of accident tolerant fuel-cladding U3Si2-FeCrAl performance in a light water reactor |
8 |
| Numerical modeling of floating bodies transport for flooding analysis in nuclear reactor building |
7 |
| Graphite dust deposition on HTGR steam generator: Effects of particle-wall and particle-vortex interactions |
7 |
| Interfacial and wall friction factors of swirling annular flow in a vertical pipe |
7 |
| Experimental analysis of stationary and transient scenarios of alfred steam generator bayonet tube in circe-hero facility |
7 |
| IAEA CRP benchmark of ROCOM PTS test case for the use of CFD in reactor design using the CFD-Codes ANSYS CFX and TrioCFD |
7 |
| Preliminary evaluation of U3Si2-FeCrAl fuel performance in light water reactors through a multi-physics coupled way |
7 |
| Experimental investigation of the isothermal flow field across slant 5-tube bundles in helically coiled steam generator geometry using PIV |
7 |
| High fidelity numerical simulations of an infinite wire-wrapped fuel assembly |
7 |
| Validation of BWR spent nuclear fuel isotopic predictions with applications to burnup credit |
7 |
| Modeling the performance of TRISO-based fully ceramic matrix (FCM) fuel in an LWR environment using BISON |
7 |
| A margin missed: The effect of surface oxidation on CHF enhancement in IVR accident scenarios |
7 |
| Fractional order PID control of steam generator water level for nuclear steam supply systems |
7 |
| Extension and application on a pool-type test facility of a system thermal-hydraulic/CFD coupling method for transient flow analyses |
7 |
| Numerical study on the turbulent mixing in channel with Large Eddy Simulation (LES) using spectral element method |
7 |
| Estimation of coping time in pressurized water reactors for near term accident tolerant fuel claddings |
7 |
| Investigation of the flow characteristics in a main nuclear power plant pump with eccentric impeller |
7 |
| Numerical study on the single bubble rising behaviors under rolling conditions |
7 |
| Evaluation of critical heat flux of ATF candidate coating materials in pool boiling |
6 |
| Experimental study of transient phenomena in the three-liquid oxidic-metallic corium pool |
6 |
| Proper orthogonal decomposition and dynamic mode decomposition of jet in channel crossflow |
6 |
| CFD investigation of bypass flow in HTR-PM |
6 |
