Publications ‒ LRS ‐ EPFL

Avertissement

Veuillez noter que les listes de publications issues d’Infoscience et intégrées sur le site web de l’EPFL, les pages de laboratoires et les pages people sont figées suite au lancement de la nouvelle version de la plateforme. Les gestionnaires de ces pages sont invité.es à recréer leur liste de publications depuis Infoscience. Pour toute assistance, veuillez consulter l’aide infoscience ou contacter le support.

Tool Developments in the OpenMC Code: Correlated Sampling and Transient Fission Matrix Approach Coupled to OpenFOAM

A. Laureau; T. Le Meute; T. J-F. Ligonnet; E. Merle

Nuclear Science And Engineering. 2024-06-17. DOI : 10.1080/00295639.2024.2357422.

Tool Developments in the OpenMC Code: Correlated Sampling and Transient Fission Matrix Approach Coupled to OpenFOAM

Neutron-Gamma Noise Measurements in a Zero-Power Reactor Using Organic Scintillators

Enhancing higher education through hybrid and flipped learning: Experiences from the GRE@T-PIONEeR project

Gamma-ray Spectroscopy in Low-Power Nuclear Research Reactors

Computation of sensitivity coefficients in fixed source simulations with SERPENT2

High-Resolution Modeling Without Computation Slowdown for PETALE in CROCUS

Gamma noise to non-invasively monitor nuclear research reactors

Intra-pin Reaction Rate Measurements in the CROCUS Experimental Reactor

Design of an Open-Loop Pile-Oscillation Program in the CROCUS Reactor

Atomic level insight into irradiation effects in nuclear fuel materials

Innovative thermophysical characterization of an ATF composite cladding: the case of silicon carbide

Preliminary safety analysis of the TRANSMUTEX sub-critical reactor using the GeN-Foam multi-physics solver

Monte Carlo analysis of BWR geometries using a cycle check-up methodology

Analysis of the preliminary campaign for the PETALE program

Extension of the OFFBEAT fuel performance code to finite strains and validation against LOCA experiments

Design of an Instrumented Fuel Rod for Intra-pin Neutron Flux Measurements in the CROCUS Experimental Reactor

Development of a multi-channel gamma-blind fast neutron detector based on wavelength shifting fibers embedded in a ZnS:Ag epoxy mixture

Extension of GeN-Foam to departure from nucleate boiling prediction and validation against the OECD/NRC PSBT benchmark

Extension, validation, and optimization of Serpent/DYN3D/ATHLET code system for SFR applications

Microscopic Digital Image Correlation Analysis of Strain Distributions in Stainless Steel 316L Induced by Multiaxial Cyclic Loading

Extension of GeN-Foam to modeling of boiling water and validation against the OECD/NRC PSBT benchmark

Gamma-Blind Fast Neutron Detection for Spent Nuclear Fuel Characterization

Unmoderated molten salt reactors design optimisation for power stability

Towards coupling conventional with high-fidelity fuel behavior analysis tools

High accuracy measurement of the prompt neutron decay constant in CROCUS using Gamma Noise and bootstrapped uncertainties

Development of a point-kinetics model in OpenFOAM, integration in GeN-Foam, and validation against FFTF experimental data

Characterization of High Harmonic Frequencies in Reactor Noise Experiments Within the CORTEX Project

Investigation and Validation of Unstructured Mesh Methodologies for Modeling Experimental Reactors

On the Capability of Wall-Modeled Large Eddy Simulations to Predict Particle Dispersion in Complex Turbulent Flows

Advanced methodologies for modeling and assessment of SFR safety functions

Towards implementing new isotopes for environmental research:Redetermination of the 32Si half-life

Miniature and Minimalistic Neutron Detectors for Online High-Resolution Experiments in the Zero-Power Reactor CROCUS

Delayed Hydride Cracking in Irradiated and Unirradiated Zircaloy-2 Cladding

GeN-ROM-An OpenFOAM(R)-based multiphysics reduced-order modeling framework for the analysis of Molten Salt Reactors

Multi-physics modeling of VVERs with high fidelity high-resolution codes

On the development of multi-physics tools for nuclear reactor analysis based on OpenFOAM (R): state of the art, lessons learned and perspectives

Validation of spent nuclear fuel decay heat calculations using Polaris, ORIGEN and CASMO5

Data-Driven Predictive Models: Calculational Bias in Characterization of Spent Nuclear Fuel

Interface Tracking Investigation of the Sliding Bubbles Effects on Heat Transfer in the Laminar Regime

Development of a semi-implicit contact methodology for finite volume stress solvers

Modeling the response of a diamond detector in the zero power reactor CROCUS

Development of a novel two-phase flow solver for nuclear reactor analysis: Validation against sodium boiling experiments

Delayed gamma fraction determination in the zero power reactor CROCUS

Uncertainty analyses of spent nuclear fuel decay heat calculations using SCALE modules

Hybrid halide perovskite neutron detectors

Study of a Current-Based Quench Detection Method for CCT Magnets via a Co-Wound Superconducting Sensing Wire

Optimisation of used nuclear fuel canister loading using a neural network and genetic algorithm

Piecewise linear interface-capturing volume-of-fluid method in axisymmetric cylindrical coordinates

Investigation on the effect of eccentricity for fuel disc irradiation tests

Development of a novel two-phase flow solver for nuclear reactor analysis: algorithms, verification and implementation in OpenFOAM

Role of mean stress on fatigue behavior of a 316L austenitic stainless steel in LWR and air environments

Characterization of Austenitic Stainless Steels with Regard to Environmentally Assisted Fatigue in Simulated Light Water Reactor Conditions

Mean Stress Effect on the Fatigue Life of 304L Austenitic Steel in Air and PWR Environments Determined with Strain- and Load-Controlled Experiments

High resolution measurements with miniature neutron scintillators in the SUR-100 zero power reactor

Local and high distance neutron and gamma measurements of fuel rods oscillation experiments

First in-core gamma spectroscopy experiments in a zero power reactor

Design of a 150-miniature detectors 3D core-mapping system for the CROCUS reactor

PISTIL, a reactivity modulation device to probe the transfer function of the nuclear reactor CROCUS

Fundamental study on microlayer dynamics in nucleate boiling

Improved methodology for analysis and design of Molten Salt Reactors

Optimization of multi-group energy structures for diffusion analyses of sodium-cooled fast reactors assisted by simulated annealing-Part II: Methodology application

Data-Driven Reduced-Order Modeling of Convective Heat Transfer in Porous Media

Development of a new method to determine the axial void velocity profile in BWRs from measurements of the in-core neutron noise

Development of a Novel Finite Volume Methodology for Multi-Dimensional Fuel Performance Applications

A Coarse-mesh Methodology for the Analysis of One and Two-phase Nuclear Reactor Thermal-hydraulics in a Multi-physics Context

Structural performance and mechanical properties investigation of spent nuclear fuel rods under static and dynamic bending loads

Optimization of multi-group energy structures for diffusion analyses of sodium-cooled fast reactors assisted by simulated annealing – Part I: Methodology demonstration

Steady-state neutronic measurements and comprehensive numerical analysis for the BME training reactor

Loading optimization for Swiss used nuclear fuel assemblies into final disposal canisters (vol 370, 110897, 2020)

Validation of axial void profile measured by neutron noise techniques in CROCUS

CORE SIM+ simulations of COLIBRI fuel rods oscillation experiments and comparison with measurements

Towards the Validation of Noise Experiments in the CROCUS Reactor Using the TRIPOLI-4 Monte Carlo Code in Analog Mode

Analysis of the first COLIBRI fuel rods oscillation campaign in the CROCUS reactor for the European project CORTEX

Dosimetry modeling and experimental validation for the PETALE program in the CROCUS reactor

Highly localized azimuthal measurements in the CROCUS reactor towards the validaiton of high-fidelity neutronics codes

A Hybrid Approach to Neutron Transport with Thermal-hydraulic Feedback for Reactor Transient Analysis

APT/TEM characterization of the element segregation in ferritic/martensitic steels after irradiation in SINQ

Loading optimization for Swiss used nuclear fuel assemblies into final disposal canisters

Artificial Neural Networks as Surrogate Models for Uncertainty Quantification and Data Assimilation in 2-D/3-D Fuel Performance Studies

Development and verification of a methodology for neutron noise response to fuel assembly vibrations