A novel method of manufacturing a heavy integrated support ring in fast reactor (in session "4.2 Structural, Novel, and Large Components Materials")
A statistical design method for steady state creep applied to Grade 91 components (in session "6.6 Fuel Performance and Material Modelling")
A Status of Experimental Program to Achieve In-Vessel Retention during Core Disruptive Accidents of Sodium-Cooled Fast Reactors (in session "2.4 Severe Accidents")
A Study on the Development of a Procedure Complexity Evaluation and Optimization for Operating Procedures of China Experimental Fast Reactor (in session "Poster Session")
Activities of the GIF Safety and Operation Project of Sodium-Cooled Fast Reactor Systems (in session "2.1 General Safety Approach")
Administrative Remarks (in session "Opening Session")
ADS for Energy Production and 233U breeding in HEU-Thorium Oxide system (in session "Poster Session")
Advanced Flow-Sheet for Partitioning of Trivalent Actinides from Fast Reactor High Active Waste (in session "3.3 Reprocessing, Partitioning, and Transmutation")
Advanced in-situ Calibration and Probe Release Mechanism for PFBR SG Inspection System (PSGIS) (in session "8.1 SFR Commissioning, Operation, and Decommissioning")
Advanced Reactor Experiments for Sodium Fast Reactor Fuels (ARES) Project: Transient Irradiation Experiments for Metallic and MOX Fuels (in session "3.2 Development of innovative fuels: design and properties irradiation")
Aerosol module for modeling of the fission product behavior in FR cooling circuits and NPP compartments (in session "6.4 Simulation Tools for Safety Analysis")
ALFRED DHR system scaling verification and numerical pre-test analysis (in session "Poster Session")
ALFRED FLOW BOCKAGE ANALYSIS (in session "6.2 Thermal Hydraulics")
ALFRED High priority R&D Needs (in session "5.3 Experimental Programs II")
An e-learning tool on Fast Reactors and their Fuel Cycles (in session "9.1 Education, Profesional Development, and Knowledge Management")
AN EXPERIMENTAL STUDY ON SECONDARY SODIUM SYSTEM BASED DECAY HEAT REMOVAL CIRCUIT OF A SODIUM COOLED FAST REACTOR (in session "5.2 Experimental Programs I")
Analysis of Fuel Burnup and Safety Parameters of Gas Cooled Fast Breeder Reactors (in session "3.4 Advanced Fuel Development")
Analysis of sodium fire accident after upgrade of ventilation system of primary loop’s corridor (in session "Poster Session")
Analysis of the natural circulation capacity of decay heat removal system in pool-type sodium-cooled fast reactor (in session "6.4 Simulation Tools for Safety Analysis")
ANALYSIS OF THE SGTR ACCIDENT FOR SAFETY JUSTIFICATION OF TWO-CIRCUIT LEAD COOLED REACTOR. (in session "2.2 Safety Design and Analysis")
Application of a Risk-Informed Performance-Based Approach for the Authorization of the Versatile Test Reactor (in session "2.1 General Safety Approach")
APPLICATION OF DIGITAL TWIN OF FAST REACTOR PLANT FOR CONTROL SYSTEM ALGORITHM TESTING (in session "6.5 Integrated Analysis and Digitalization")
Application of Model Based System Engineering in Design of Digital Fast Reactor Nuclear Power Plant (in session "Poster Session")
Application of the practical elimination concept within the framework of the ESFR-SMART project to improve the intrinsic safety of the sodium-cooled fast reactor (in session "2.1 General Safety Approach")
Approach for ALFRED licensing in Romania (in session "2.1 General Safety Approach")
APPROACHES TO FORM THE BN 1200 CORE START LOADING USING MOX-FUEL AND MNUP-FUEL (in session "6.5 Integrated Analysis and Digitalization")
Assay of Waste drum based on Passive Neutron Counting Technique (in session "Poster Session")
Basis for the Safety Approach (BSA) for Design & Assessment of Generation IV Nuclear Systems (in session "2.1 General Safety Approach")
BLIND PHASE RESULTS FOR TRANSIENT SIMULATIONS OF THE FFTF LOSS OF FLOW WITHOUT SCRAM TEST #13 (in session "Special Session: IAEA Coordinated Research Projects")
Blind-Phase Results of the FFTF Neutronic Benchmark (in session "Special Session: IAEA Coordinated Research Projects")
BOR-60 REACTOR OPERATING EXPERIENCE, WORK ON IMPROVING SAFETY AND EXTENDING LIFETIME (in session "8.1 SFR Commissioning, Operation, and Decommissioning")
Calculation of the materials activation with BPSD code (in session "Poster Session")
CEFR Physical Start-Up Tests: the Core Specifications and Experiments (in session "Special Session: IAEA Coordinated Research Projects")
CFD ANALYSES OF THE ALFRED HOT PLENUM (in session "Poster Session")
CFD Simulations on a hexagonal 61-pin wire-wrapped fuel bundle with STARCCM+ and comparison with experimental data. (in session "6.2 Thermal Hydraulics")
Chair of Conference International Advisory Committee Closing Remarks (in session "Closing Session")
Characterization of the Molten Chloride Fast Reactor fuel cycle options (in session "Poster Session")
China Key Note (in session "Plenary 1. Keynotes from Member States")
CHOICE OF A COOLANT FOR A MODULAR SMALL POWER REACTOR SVBR-100 (in session "1.2 Innovative Design Advances")
Codes of new generation – sustainable platform for numerical modeling of installations in the Proryv project (in session "6.3 Multiscale and Multiphysics Calculations")
Cognitive Information Retrieval Based on Ontological Model of Knowledge Representation (in session "Poster Session")
Commissioning and Operating Experience for Secondary Sodium Systems and is Auxiliaries of PFBR (in session "8.1 SFR Commissioning, Operation, and Decommissioning")
COMPARATIVE ANALYSIS OF CALCULATIONAL AND EXPERIMENTAL DIFFERENCES OF THE NEUTRON-PHYSICAL CHARACTERISTICS OF THE BN-800 REACTOR (in session "Poster Session")
Comparative analysis of minor actinides transmutation in a molten-salt burner reactor based on LiF-NaF-KF and LiF-BeF2 salts (in session "Poster Session")
Comparative multi-criteria analysis of scenarios of the Russian nuclear energy development in the context of uncertainty knowledge about the future (in session "7.1 Sustainability: Economics, Environment, and Proliferation")
Comparison of calculation methods for lead cooled fast reactor reactivity effects (in session "6.1 Neutronics")
Comparisons of Feedback under UTOPA with In Pin Fuel Motion Dynamics in Fast Reactors (in session "2.4 Severe Accidents")
Complex of experimental facilities for design and safety justification of fast reactors with liquid metal coolants (in session "5.1 Experimental Reactors and Facilities")
COMPLEX RADIATION AND HYGIENE STUDIES OF RADIATION IMPACT FACTORS ON PRODUCTION PERSONNEL, MIXED NITRIDE URANIUM-PLUTONIUM FUEL FOR FAST NEUTRON REACTORS (in session "Poster Session")
Computational fluid dynamics study for estimation of dilution for failed fuel location system (in session "6.6 Fuel Performance and Material Modelling")
Computational Studies of Advantages of Lead-Cooled Fast Reactor Core (in session "6.5 Integrated Analysis and Digitalization")
Conceptual Core configuration for increasing Power of Fast Breeder Reactor to 40 MWt (in session "Poster Session")
Conceptual design of ultra-long life hybrid micro modular reactor cooled by potassium heat pipe (in session "1.2 Innovative Design Advances")
Conference Chair Address (in session "Opening Session")
Conference General Co-Chair Closing Remarks (in session "Closing Session")
Controlled thermonuclear fusion: potential role of a joint (Th-U-Pu) nuclear fuel cycle (in session "Poster Session")
Coolant flow monitoring with an Eddy Current Flow Meter at a mock-up of a liquid metal cooled fast reactor (in session "5.2 Experimental Programs I")
Core Design of 100MWe Advanced Nitride-fueled Simplified Liquid Metal Cooled Fast Reactor (in session "1.2 Innovative Design Advances")
CORROSION HYDROGEN MASS TRANSFER IN FAST REACTOR STEAM GENERATORS OF THE SODIUM-WATER TYPE (in session "Poster Session")
Coupled neutronic/thermal-hydraulic simulation of Unprotected Loss of Flow Test at Fast Flux Test Facility (in session "2.3 Accident Analysis")
Creep and Creep-Fatigue Behavior of an Advanced Stainless Steel (Alloy 709) - Application to Sodium-Cooled Fast Reactors (in session "4.2 Structural, Novel, and Large Components Materials")
Creep and Tensile Properties of Indian Advanced Fast Reactor Clad tubes (IFAC-1) for Future FBRs (in session "4.1 Advanced Reactor Cladding and Core Material, Coolants, and Related Chemistry")
CRITICALITY SENSITIVITY ANALYSIS IN RELATION TO EMPTIES OF A FAST REGENERATOR NUCLEAR REACTOR (in session "Poster Session")
CURRENT STATE AND ISSUES OF THE HEAVY LIQUID METAL COOLANT TECHNOLOGY DEVELOPMENT (PB, PB-BI) (in session "Poster Session")
Current status of development of 3D DNS CONV-3D code: one- and two-phase flow models (in session "6.3 Multiscale and Multiphysics Calculations")
DESIGN & ANALYSIS OF A NOVEL ARRANGEMENT FOR COUPLING AND DECOUPLING OF ROTATABLE PLUGS IN PFBR (in session "1.3 System Innovations")
DESIGN & DEVELOPMENT OF CUSTOM SHAPED BACK-UP SEAL IN SILICONE FOR PFBR (in session "Poster Session")
Design of experimental scheme for activation method of China demonstration fast reactor (in session "Poster Session")
Design of metal fuel pin for test irradiation in FBTR and for future reactors. (in session "3.2 Development of innovative fuels: design and properties irradiation")
Design of secondary sodium based decay heat removal system for future fast breeder reactors (in session "1.3 System Innovations")
Design Studies Towards Raising FBTR to Full Power (in session "2.2 Safety Design and Analysis")
Design, Experimental trials and Qualification of explosive welding technique for plugging of degraded PFBR Steam Generator tubes (in session "8.1 SFR Commissioning, Operation, and Decommissioning")
DESIGN, MANUFACTURING AND IN-SODIUM TESTING OF AM350-WELDED DISC BELLOWS FOR FBTR CONTROL ROD DRIVE MECHANISM (in session "5.3 Experimental Programs II")
Design, manufacturing and transportation of high capacity High Level Liquid Waste Storage tanks (in session "Poster Session")
Determination of the metallic and oxide compounds in models based on metallic uranium containing uranium dioxide, metallic neodymium, cerium as well as neodymium and cerium oxides (in session "Poster Session")
Development and Demonstration of Diffusion-type Hydrogen Meters for Sodium-cooled Fast Reactors (in session "4.1 Advanced Reactor Cladding and Core Material, Coolants, and Related Chemistry")
Development of a 15 kg servo manipulator for remote handling applications (in session "Poster Session")
Development of an Artificial Neural Network for predicting spatial interdependencies of reactivity effects in Sodium Fast Reactors (in session "6.5 Integrated Analysis and Digitalization")
Development of Artificial Intelligence through PLC & SCADA to predict process related failure and abnormality in a Reprocessing Plant (in session "Poster Session")
DEVELOPMENT OF BN REACTOR TECHNOLOGY IN RUSSIA (in session "1.1 Overviews and Fundamentals of Fast Reactors")
Development of Burnup Analysis System for rotational and Spiral Fuel Shuffling scheme in Breed-and-Burn Fast Rectors (in session "Poster Session")
DEVELOPMENT OF COOLANT VOIDING MODEL FOR FAST REACTOR CORE (in session "6.2 Thermal Hydraulics")
Development of density control technologies for MOX pellet using dry recycled powders (in session "Poster Session")
Development of in-vessel source term evaluation method for ULOF events in sodium-cooled fast reactor (in session "Poster Session")
Development of Integrated Severe Accident Analysis Code, SPECTRA for Sodium-cooled Fast Reactor (in session "6.4 Simulation Tools for Safety Analysis")
Development of methodology to evaluate mechanical consequences of vapor expansion in SFR severe accident transients: lessons learned from previous France-Japan collaboration and future objectives and milestones (in session "2.4 Severe Accidents")
Development of Multi-level Simulation System for Core Thermal-hydraulics Coupled with Plant Dynamics Analysis - Prediction of Transient Temperature Distribution in a Subassembly under Inter-subassembly Heat Transfer Effect - (in session "6.3 Multiscale and Multiphysics Calculations")
Development of Plasma Nitriding as alternate hardfacing technique for Large components of FBR and Assessment of static In-Sodium Stability of Plasma Nitrided Layer (in session "4.2 Structural, Novel, and Large Components Materials")
Development of SFR core degradation simulation code SIMMER-V and its validation & verification studies (in session "6.3 Multiscale and Multiphysics Calculations")
Development of simplified fuel fabrication technologies for fast reactors (in session "3.2 Development of innovative fuels: design and properties irradiation")
DEVELOPMENT OF SUBMERGED ELECTROMAGNETIC PUMP FOR LIQUID LEAD (in session "Poster Session")
Development of the Simplified Radionuclide Transport (SRT) Code Version 2.0 for Versatile Test Reactor (VTR) Mechanistic Source Term Calculations (in session "Poster Session")
DEVELOPMENT OF THE TECHNICAL APPROACH FOR RESEARCH OF THE SODIUM COOLANT CURRENT IN THE INTEGRAL TYPE REACTOR (in session "Poster Session")
Development of the Versatile Test Reactor (VTR) Probabilistic Risk Assessment (in session "2.2 Safety Design and Analysis")
Development status of commercial SMRs and its experience to China (in session "7.1 Sustainability: Economics, Environment, and Proliferation")
DIGITAL TECHNOLOGIES FOR PROJECT DEVELOPMENT ODEC AND PEC AND DIGITAL TWINS (in session "6.5 Integrated Analysis and Digitalization")
DISTINCTIVE FEATURES OF THE BN-800 CORE IN THE COURSE OF TRANSITION TO COMPLETE MOX-FUEL LOADING (in session "6.5 Integrated Analysis and Digitalization")
Dutch Thermal Hydraulic Design and Safety Support for LMFRs (in session "6.3 Multiscale and Multiphysics Calculations")
Eddy Current Flow Meter flow rate measurements in liquid Sodium at the SUPERFENNEC loop (in session "Poster Session")
Effect of Reactor Technology on Economics of SMR Projects (in session "7.1 Sustainability: Economics, Environment, and Proliferation")
EFFECTIVE FUEL SUPPLY OF TWO-COMPONENT NUCLEAR ENERGY SYSTEM WITH VVER-BN REACTORS (in session "7.1 Sustainability: Economics, Environment, and Proliferation")
ELABORATION OF THE THERMAL-HYDRAULIC CHARACTERISTICS OF THE REACTOR PLANT BASED ON THE OPERATION EXPERIENCE OF THE POWER UNIT WITH BN-800 REACTOR (in session "Poster Session")
Electrical conductivity of multicomponent chloride melts, containing ions of mono-, di-, and trivalent metals (in session "Poster Session")
Electrolytic reduction of the simulated oxide spent nuclear fuel in LiCl-Li2O melt (in session "Poster Session")
ENDF/B-VIII.0 NUCLEAR DATA SENSITIVITY AND UNCERTAINTY (S/U) ANALYSIS OF KEY SAFETY-RELEVANT REACTIVITY COEFFICIENTS FOR THE ALFRED CORE (in session "Poster Session")
Estimation of mean charge on sodium metal aerosol in the argon and nitrogen gas environment during external gamma irradiation (in session "5.2 Experimental Programs I")
European Commission (EC) Key Note (in session "Plenary 2. International Organizations and YGE Winners")
Evaluation of an increase of the power density for the French commercial Sodium Fast Reactor and optimization study at 1100 MWe with the SDDS tool (in session "1.3 System Innovations")
Evaluation of EPDM and Silicone rubber compounds for application in Reprocessing Plant (in session "Poster Session")
EXAMPLES OF AREAS OF NOVELTY IN LIQUID METAL FAST REACTORS TO CONSIDER IN THE REVIEW OF APPLICABILITY OF THE IAEA SAFETY STANDARDS: FISSION PRODUCT RETENTION BARRIERS: DIFFERENCES BETWEEN LIQUID METAL FAST REACTORS AND LIGHT WATER REACTORS (in session "2.1 General Safety Approach")
Experience in Preheating of PFBR Reactor Assembly (in session "8.1 SFR Commissioning, Operation, and Decommissioning")
EXPERIENCE OF OPERATIONAL CHEMICAL CLEANING OF BN-600 STEAM GENERATOR EVAPORATORS FROM CORROSION PRODUCT DEPOSITS (in session "8.1 SFR Commissioning, Operation, and Decommissioning")
Experiment and Numerical Simulations on SFR Core-catcher Safety Analysis after Relocation of Corium (in session "2.4 Severe Accidents")
Experiment-calculated method for determination of prompt neutron lifetime in fast metal cores intended for verification of neutron transfer simulation codes (in session "Poster Session")
Experimental and computational studies of heat exchange for liquid metals boiling in fuel assembly models at accidental conditions (in session "5.2 Experimental Programs I")
Experimental and Numerical Study on Temperature Fluctuation in The Upper Plenum of Fast Reactor (in session "Poster Session")
EXPERIMENTAL CAPABILITIES OF THE RESEARCH REACTOR FACILITY MBIR. MAIN AREAS OF THE RESEARCH PROGRAMME IN THE INTERESTS OF THE GENERATION 4 REACTORS (in session "5.1 Experimental Reactors and Facilities")
Experimental investigation of the fluid-structure interaction effect between adjacent equipment supports in a fast reactor (in session "Poster Session")
Experimental modeling of a fuel element simulator vibration in a coolant flow (in session "6.2 Thermal Hydraulics")
Experimental study on sodium insulation interaction and its effect on structural material (in session "Poster Session")
EXPERIMENTAL TEST FACILITY TO TEST A PROTOTYPE OF THE AIR HEAT EXCHANGER GATE FOR THE ADVANCED BN REACTOR PLANT. DESIGN AND CONSTRUCTION ITEMS (in session "5.3 Experimental Programs II")
EXPORT OF RBN WITH SNCD AND NUCLEAR PROLIFERATION RISKS (in session "Poster Session")
Export Potential and Commercialization Conditions of Fast Reactors Considering Non-Proliferation Items (in session "7.1 Sustainability: Economics, Environment, and Proliferation")
FABRICATION AND PERFORMANCE ASSESMENT OF ODS FECRAL CLADDING TUBE (in session "4.1 Advanced Reactor Cladding and Core Material, Coolants, and Related Chemistry")
Fabrication and reprocessing of mixed uranium-plutonium nitride fuel for reactor BREST (in session "3.3 Reprocessing, Partitioning, and Transmutation")
Fast Reactor Source Term Modeling and Simulation Functional Requirements and Gap Assessment (in session "Poster Session")
FEASIBILITY STUDY OF HETEROGENEOUS TRANSMUTATION OF AMERICIUM IN FAST REACTORS (in session "3.3 Reprocessing, Partitioning, and Transmutation")
First fully adjusted set of parameters for the corrosion product contamination code OSCAR-Na (in session "6.6 Fuel Performance and Material Modelling")
France Key Note (in session "Plenary 1. Keynotes from Member States")
France-Japan Collaboration on the SFR Severe Accident Studies: Outcomes and future work program (in session "2.1 General Safety Approach")
France-Japan Collaboration on Thermodynamic and Kinetic Studies of Core Material Mixture in Severe Accidents of Sodium-Cooled Fast Reactors (in session "5.3 Experimental Programs II")
French-Japanese experimental collaboration on fuel-coolant interactions in sodium-cooled fast reactors (in session "2.4 Severe Accidents")
Fuel cycle closure for high power fast neutron reactor (in session "3.1 Fuel Cycle Scenarios")
Fuel handling Experience of FBTR (in session "8.1 SFR Commissioning, Operation, and Decommissioning")
Gear Test Assembly: First Liquid Metal Component Testing in METL (in session "4.1 Advanced Reactor Cladding and Core Material, Coolants, and Related Chemistry")
GEN IV INTERNATIONAL FORUM WEBINARS INITITATIVE (in session "9.1 Education, Profesional Development, and Knowledge Management")
Generation IV International Forum Key Note (in session "Plenary 2. International Organizations and YGE Winners")
GFR Research and Development Programme in V4 countries (in session "1.1 Overviews and Fundamentals of Fast Reactors")
HEAT TRANSFER CALCULATION AND SERVICE LIFE TIME ESTIMATION OF SUBMERGED ELECTROMAGNETIC PUMP FOR LIQUID LEAD (in session "Poster Session")
HETEROGENEOUS BURNING OF MINOR ACTINIDES IN A FAST REACTOR (in session "3.3 Reprocessing, Partitioning, and Transmutation")
Hybrid high power fast breeder reactor with metallic fuel and additives consisting with lightweight atoms (in session "Poster Session")
IAEA Opening Address (in session "Opening Session")
iMAGINE - a Breakthrough Technology for Closing the Fuel Cycle without Reprocessing (in session "1.3 System Innovations")
Impact of Cladding Material on Neutronic Balance in Breed-and-Burn fast reactors (in session "Poster Session")
Impact of Core Materials on The Fuel cladding Irradiation Damage in Breed-and-Burn Fast Reactors (in session "Poster Session")
Implementation of LFR Experimental Infrastructures in Romania (in session "5.3 Experimental Programs II")
Increase of nuclear power plant hydrogen safety using zirconium accumulator (in session "Poster Session")
India Key Note (in session "Plenary 1. Keynotes from Member States")
INDUSTRIAL ENERGY COMPLEX WITH FAST NEUTRON REACTOR (in session "Poster Session")
Influence of Low Dose Irradiation on Permanent Core Structural Materials of PFBR (in session "4.1 Advanced Reactor Cladding and Core Material, Coolants, and Related Chemistry")
Influence of preheating temperature on delta-ferrite formation and mechanical properties of 12%Cr steel weld metals (in session "Poster Session")
Integral code COMPLEX for radiation safety assessment of reactor and nuclear fuel cycle facilities (in session "Poster Session")
INTEGRATED RADIATION AND HYGIENIC APPROACH TO PRODUCTION SAFETY. ASSESSMENT OF THE IMPACT ON PUBLIC HEALTH (in session "Poster Session")
Integrated thermal hydraulic analysis of Hot and Cold Pools of a liquid sodium cooled 600 MWe fast reactor (in session "Poster Session")
Integrating safety at the first design stages: a new methodology for safety-oriented SFR core design (in session "2.2 Safety Design and Analysis")
Integration of Small Modular Lead Fast Reactor with Energy Storage for load-following operation in high V-RES penetration electricity markets (in session "1.3 System Innovations")
International Atomic Energy Agency (IAEA) Key Note (in session "Plenary 2. International Organizations and YGE Winners")
Investigation of characteristics of fast power reactor with an additional function of large-scale production of plutonium-238 (in session "Poster Session")
Investigation of sodium purification (in session "4.1 Advanced Reactor Cladding and Core Material, Coolants, and Related Chemistry")
Investigation of the anodic processes on the ceramic anode in the oxide-chloride melts (in session "3.3 Reprocessing, Partitioning, and Transmutation")
INVESTIGATION OF THE SOLUBILITY OF ACTINIDE FLUORIDES FOR THE CHOICE OF A SALT SOLVENT FOR A MOLTEN-SALT REACTOR-BURNER OF MINOR ACTINIDES (in session "Poster Session")
Investigation on Human Resources Needs and Competences Building for ALFRED Implementation in Romania (in session "9.1 Education, Profesional Development, and Knowledge Management")
Investigation on natural circulation for decay heat removal in reactor vessel of sodium-cooled fast reactor (in session "5.2 Experimental Programs I")
Irradiation Effects of T91 Ferritic/Martensitic Steel (in session "Poster Session")
Irradiation-Thermo-Mechanical Coupling Analysis and Calculation of Fast Neutron Oxide Fuel Element (in session "Poster Session")
Japan Key Note (in session "Plenary 1. Keynotes from Member States")
JUSTIFICATION OF CRITICAL EXPERIMENTS ON STAND FKBN-2 TO VERIFY NEUTRON-PHYSICAL SOFTWARE FOR CALCULATIONS OF THE MOLTEN-SALT REACTOR (in session "Poster Session")
KEY ASPECTS OF COMPETITIVENESS FOR INDUSTRIAL ENERGY COMPLEX WITH FR AND CLOSED NFC (in session "7.1 Sustainability: Economics, Environment, and Proliferation")
LARGE-SCALE HYDROGEN PRODUCTION; Fast-neutron Reactors Coupled to Thermochemical Copper-Chlorine Hydrogen Plant (in session "Poster Session")
Leak-Before-Break Design of Double-Walled Once-Through Steam Generators for Lead Cooled Fast Reactor (in session "Poster Session")
LFR Design and Technologies Development at ENEA: Status and Perspectives (in session "5.3 Experimental Programs II")
LOW ENRICHMENT NUCLEAR FUEL BASED ON URANIUM-ZIRCONIUM CARBONITRIDE: REACTOR TESTS AND PREPARATION FOR STUDIES AT CRITICAL ASSEMBLIES (in session "3.4 Advanced Fuel Development")
Material Data Acquisition Activities to Develop the Material Strength Standard for Sodium-cooled Fast Reactors (in session "4.2 Structural, Novel, and Large Components Materials")
Mechanisms Engineering Test Loop (METL) Facility (in session "5.1 Experimental Reactors and Facilities")
MECHANISTIC MODELLING OF AEROSOL EVOLUTION IN AN SFR CONTAINMENT FOLLOWING A HYPOTHETICAL SEVERE ACCIDENT (in session "2.3 Accident Analysis")
Mitigation of Sloshing Effects in High level Liquid Waste (HLW) Storage Tank for Nuclear Spent Fuel Applications (in session "Poster Session")
Model validation of the ASTERIA-SFR code related to freezing phenomena of liquid and liquid/particle mixtures based on THEFIS experimental results (in session "Poster Session")
Modeling and Simulation of Source Term for Sodium-Cooled Fast Reactor under Hypothetical Severe Accident: Sodium Fire and Radionuclide Transport in Containment (in session "2.3 Accident Analysis")
Modeling of the coolant region in the ALFRED core in case of thermal expansion (in session "Poster Session")
Modeling the optimal economic structure of a global deploying nuclear power system with fast and thermal reactors in a partially closed nuclear fuel cycle (in session "7.1 Sustainability: Economics, Environment, and Proliferation")
Modelling and Simulation of Source Term for Sodium-Cooled Fast Reactor under Hypothetical Severe Accident: Primary System/Containment System Interface Source Term Estimation (in session "2.3 Accident Analysis")
Modelling of postulated reactivity insertion in a Generation IV Molten Salt Reactor (in session "2.2 Safety Design and Analysis")
Modelling of radionuclide release from primary system during a hypothetical severe accident in an SFR (in session "Poster Session")
Models of the integral EUCLID/V2 code for numerical modeling of different regimes of lead-cooled fast reactor (in session "6.4 Simulation Tools for Safety Analysis")
Models of the integral EUCLID/V2 code for numerical simulation of severe accidents in a sodium-cooled fast reactor with MOX and MNUP fuels (in session "6.4 Simulation Tools for Safety Analysis")
MSR Fuel Cycle and Thermo-Dynamics Simulations (in session "6.6 Fuel Performance and Material Modelling")
Multi-criteria comparison of the efficiency of minor actinides burning in different nuclear reactors based on the INPRO/IAEA KIND approach (in session "3.3 Reprocessing, Partitioning, and Transmutation")
MULTIPURPOSE RESEARCH FACILITY MBIR AND POLY FUNCTIONAL RADIOCHEMICAL COMPLEX (R&D COMPLEX) AS A UNIQUE RESEARCH PLATFORM (in session "5.1 Experimental Reactors and Facilities")
MYRRHA, the Belgian prototype that fascinates the world (in session "1.1 Overviews and Fundamentals of Fast Reactors")
Neutronic Calculation of CEFR Core using Different Nuclear Data Libraries (in session "Poster Session")
Neutronics analysis of CEFR Start-up tests at IGCAR using FARCOB and ERANOS 2.1 Code Systems (in session "6.1 Neutronics")
Neutronics Benchmark of CEFR Start-Up Tests: An IAEA coordinated research project (in session "Special Session: IAEA Coordinated Research Projects")
Neutronics Benchmark of CEFR Start-Up Tests: Reaction Rates and Reactivity Coefficients (in session "Special Session: IAEA Coordinated Research Projects")
Neutronics Benchmark of CEFR Start-Up Tests: Temperature Coefficient, Sodium Void Worth, and Swap Reactivity (in session "Special Session: IAEA Coordinated Research Projects")
New ASTRID SFR - Intermediate Heat Exchanger (IHX) and internal vessel interface system: qualification tests onto a scale 1 representative mock-up (in session "4.2 Structural, Novel, and Large Components Materials")
New Concepts and Methodologies for the Effective Deployment of Gen IV reactors (in session "Poster Session")
New Finite Element Neutron Kinetics Code System FENNECS/ATHLET for Coupled Safety Assessment of (Very) Small and Micro Reactors (in session "Poster Session")
NEXT GENERATION NUCLEAR POWER: RADIOLOGICAL SUSTAINABILITY AND ECOLOGICAL ADVANTAGES (in session "Poster Session")
NON-DESTRUCTIVE METHOD FOR DETERMINING STEEL CORROSION COEFFICIENTS IN LEAD (in session "Poster Session")
Novel Electrical, Electronics and Instrumentation systems for Fast Reactor Fuel Reprocessing Plants (in session "Poster Session")
Novel neutronics design of the MYRRHA core (in session "1.2 Innovative Design Advances")
Nuclear Fuels for Fast Reactors-A Review (in session "3.4 Advanced Fuel Development")
Nuclear Hydrogen and Fast Reactors (in session "Poster Session")
Numerical Investigation of Cellular Convection in the Cover Gas space of Fast Breeder Test Reactor (in session "Poster Session")
Objectives and Status of Neutronics Sub-exercises of the OECD/NEA Benchmark for Uncertainty Analysis in Modelling for Design, Operation and Safety Analysis of SFRs (in session "6.1 Neutronics")
OECD/Nuclear Energy Agency Key Note (in session "Plenary 2. International Organizations and YGE Winners")
ON MEASUREMENT OF OXYGEN CONCENTRATION IN SODIUM BY MEANS OF PLUG INDICATOR (in session "Poster Session")
On substantial slowing down of the kinetics of fast transient processes in fast reactor (in session "Poster Session")
ON THE POSSIBILITY TO CHANGE THE ISOTOPIC COMPOSITION OF PLUTONIUM FROM THE SPENT MOX FUEL OF PWRs IN FAST REACTORS (in session "3.4 Advanced Fuel Development")
Operating Experience of FBTR (in session "8.1 SFR Commissioning, Operation, and Decommissioning")
OPTIMIZATION OF BUILT-IN PRIMARY SODIUM PURIFICATION SYSTEM FOR ADVANCED BN REACTOR PLANT (in session "1.3 System Innovations")
Optimization of Ruthenium concentration in PUREX Process during Fast reactor fuel Reprocessing (in session "Poster Session")
Optimization of the secondary loops on the ESFR SMART project (in session "1.3 System Innovations")
Over three decades of radiological protection experience at Fast Breeder Test Reactor (FBTR) (in session "2.3 Accident Analysis")
Overview of a Sodium Fast Reactor Thermal Hydraulic Test Facility (in session "5.1 Experimental Reactors and Facilities")
Overview of critical experiments with fast metal cores held on assembly machine FKBN-2 (in session "5.2 Experimental Programs I")
Overview of IAEA Fast Reactor Related Technology Development Activities (in session "9.1 Education, Profesional Development, and Knowledge Management")
Overview of the R&D programs led by the past at IRSN on sodium fire (in session "5.1 Experimental Reactors and Facilities")
Overview of the Versatile Test Reactor Safety Analysis (in session "2.2 Safety Design and Analysis")
Overview of U.S. Fast Reactor Technology R&D Program (in session "1.1 Overviews and Fundamentals of Fast Reactors")
Passive Heat Removal System Analysis for the Westinghouse Lead Fast Reactor (in session "6.5 Integrated Analysis and Digitalization")
Perspectives and discussions on the modes and development path of China's commercial closed nuclear fuel cycle (in session "3.1 Fuel Cycle Scenarios")
Physical feasibility of MA transmutation in a two-component nuclear energy system in Russia (in session "3.3 Reprocessing, Partitioning, and Transmutation")
Physical modeling of hydrodynamics and heat exchange in fast reactors with liquid metal coolants (in session "5.1 Experimental Reactors and Facilities")
Phénix Control Rod Withdrawal test analysis using a multiphysics methodology (in session "6.3 Multiscale and Multiphysics Calculations")
Pilot Demonstrational Fast Reactor with Lead Coolant BREST-OD-300 (in session "1.2 Innovative Design Advances")
Possibility of Simulating Natural Circulation in Fast Neutron Reactors Using a Light Water Test Facility (in session "6.2 Thermal Hydraulics")
Postirradiation characterization of AFC metallic fuel alloys concepts. (in session "3.4 Advanced Fuel Development")
Potential Role of Fast Reactors with Heterogeneous Fuel Assembly in Development Nuclear Power Structure (in session "3.1 Fuel Cycle Scenarios")
PRE-DESIGN OF A PASSIVE DECAY HEAT REMOVAL SYSTEM WITH A PHASE CHANGE MATERIAL FOR SMR-SFR (in session "2.2 Safety Design and Analysis")
Preliminary Shielding Analysis for the Versatile Test Reactor (in session "Poster Session")
Preliminary testing of ALFRED DHR System (in session "5.3 Experimental Programs II")
Presentation of the new European project PUMMA devoted to Plutonium management in the whole fuel cycle (in session "3.1 Fuel Cycle Scenarios")
Preserving and transferring knowledge in the field of fast reactor technologies. Experience of the Obninsk Institute of Nuclear Power Engineering MEPhI (in session "9.1 Education, Profesional Development, and Knowledge Management")
PROBLEMS OF DECOMISSIONING FAST REACTORS AND WAYS OF THEIR SOLUTION ON THE BASIS OF THE BR-10 RESEARCH REACTOR (in session "8.1 SFR Commissioning, Operation, and Decommissioning")
Progress in conceptual design of a pool-type sodium-cooled fast reactor in Japan (in session "1.1 Overviews and Fundamentals of Fast Reactors")
Progress in system thermohydraulic code HYDRA-IBRAE/LM models development for fast reactor simulation (in session "6.2 Thermal Hydraulics")
Progress in the Design and R&D for future FBRs (in session "1.1 Overviews and Fundamentals of Fast Reactors")
Project of a multipurpose lead reactor with a hard neutron spectrum (in session "1.2 Innovative Design Advances")
Proposal of a compact core design for the 1000 MWe French commercial Sodium Fast Reactor by means of the SDDS multi-objective optimization tool (in session "1.2 Innovative Design Advances")
R&D on recovery and separation of americium and curium under "Proryv" project (in session "Poster Session")
RADIATION AND HYGIENE ASSESSMENT OF EXTERNAL EXPOSURE FACTORS OF PERSONNEL WORKING AT EXPERIMENTAL FACILITIES IN THE PRODUCTION OF MIXED NITRIDE URANIUM-PLUTONIUM FUEL (in session "3.1 Fuel Cycle Scenarios")
RADIATION-HYGIENIC ASSESSMENT OF INTERNAL EXPOSURE FACTORS OF PERSONNEL WORKING AT EXPERIMENTAL FACILITIES IN THE PRODUCTION OF MIXED NITRIDE URANIUM-PLUTONIUM FUEL (in session "Poster Session")
Reactor Core Viewing System for the pre-commissioning stage inspection of reactor core components of Prototype Fast Breeder Reactor (in session "8.1 SFR Commissioning, Operation, and Decommissioning")
REALISATION OF AN ADJUSTED NUCLEAR DATA LIBRARY BASED ON ENDF/B-VIII.0 NUCLEAR DATA EVALUATIONS FOR THE ALFRED CORE (in session "6.1 Neutronics")
Recent studies on fuel properties and irradiation behaviors of Am/Np-bearing MOX (in session "3.2 Development of innovative fuels: design and properties irradiation")
Recent thermal hydraulic studies of Gas Fast Reactor demonstrator ALLEGRO (in session "6.2 Thermal Hydraulics")
Reference Fuel Options for Generation-IV Sodium-cooled Fast Reactors (in session "3.1 Fuel Cycle Scenarios")
Regulatory Perspectives on Analytical Codes and Methods for Advanced Reactors (in session "6.4 Simulation Tools for Safety Analysis")
Removal of Radiocesium from High-Level Liquid Waste using Inorganic Ion-exchangers (in session "Poster Session")
Reprocessing of nitride and metallic spent nuclear fuel using molten salts (in session "Poster Session")
Republic of Korea Key Note (in session "Plenary 1. Keynotes from Member States")
Research and development of nuclear fuel for fast neutron reactor (in session "3.4 Advanced Fuel Development")
Research on the Impact of Advanced Rule Design System on the Digitization of Reactor Building Model (in session "Poster Session")
RESULTS OF POST-IRRADIATIONS EXAMINATIONS OF MIXED NITRIDE PINS WITH GAS AND LIQUID METAL SUB-LAYERS (in session "3.4 Advanced Fuel Development")
Revealing the dependencies of partitioning americium-241 and uranium using sorption technology based on solid-phase extractant TODGA (in session "Poster Session")
RISK FACTORS OF COMPLEX RADIATION AND NON-RADIATION EFFECTS ON THE HEALTH OF PERSONNEL IN ASSESSING THE IMPACT OF THE PRODUCTION OF INNOVATIVE FUEL FOR FAST REACTORS (in session "Poster Session")
Root Cause Analysis of FBTR Failed Fuel Pin (in session "3.2 Development of innovative fuels: design and properties irradiation")
Russian Federation Key Note (in session "Plenary 1. Keynotes from Member States")
Safety Analysis of Small Modular Sodium Fast Reactors in Anticipated Transients Without Scram Scenarios (in session "2.3 Accident Analysis")
Safety Analysis of the ARC-100 Sodium-Cooled Fast Reactor (in session "2.1 General Safety Approach")
Selection, testing and development of qualification procedure for ALLEGRO gas-cooled fast reactor fuel (in session "3.2 Development of innovative fuels: design and properties irradiation")
Simple Design Comparison of uranium nitride pin cell assembly and matrix fuel assembly for a Lithium Cooled Fast Reactor (in session "Poster Session")
Simulation of FFTF Individual Reactivity Feedback Tests with SAS4A/SASSYS-1 Code (in session "6.3 Multiscale and Multiphysics Calculations")
Simulation of fission gas release in the 3-D fuel performance code OFFBEAT (in session "6.6 Fuel Performance and Material Modelling")
SIMULATION OF THE FAST FLUX TEST FACILITY LOSS-OF-FLOW WITHOUT SCRAM ACCIDENT SCENARIO USING THE SAM COMPUTER CODE (in session "Poster Session")
Simulation of ULOF initiation phase in ESFR-SMART with SIMMER-III (in session "2.4 Severe Accidents")
Sketch Design of Fuel Sub-Assemblies for a SFR-150 MWe (in session "1.3 System Innovations")
SOCRAT-BN INTEGRAL CODE: DEVELOPMENT, VALIDATION AND CURRENT STATUS (in session "6.4 Simulation Tools for Safety Analysis")
Sodium coolant: interaction with its environment and coolant processing (in session "4.1 Advanced Reactor Cladding and Core Material, Coolants, and Related Chemistry")
Software for Simulation of Fast Reactor Operation in a Closed Nuclear Fuel Cycle (SC RTM-2) (in session "Poster Session")
Some results of using partial equations for calculations of transient processes in fast breeder reactors (in session "Poster Session")
Spatial interdependence of safety related effects in ESFR-SMART core (in session "6.1 Neutronics")
SPECIFIC FEATURES OF THE EXPORT OF RUSSIAN TECHNOLOGIES OF FAST REACTORS AND A CLOSED NUCLEAR FUEL CYCLE (in session "7.1 Sustainability: Economics, Environment, and Proliferation")
STATE OF DEVELOPMENT OF LEAD COOLANT TECHNOLOGY COMPONENTS FOR BREST-OD-300 REACTOR (in session "4.2 Structural, Novel, and Large Components Materials")
Status of Generation-IV Lead Fast Reactor Activities (in session "1.1 Overviews and Fundamentals of Fast Reactors")
Study on actinide conversion capabilities of Molten Salt Reactors (MSR) (in session "6.1 Neutronics")
Study on Sodium Fire PSA Methodology for Pool-Type Sodium cooled Fast Reactor (in session "Poster Session")
Study on the Method of Correction of Fast Reactor Power Distribution by MCNP (in session "Poster Session")
Sustainability of nuclear and non-nuclear power generation options under Russian conditions: a comparative evaluation study (in session "Poster Session")
System Safety Assessment of the Generation IV Lead Fast Reactor (in session "2.1 General Safety Approach")
Target Accuracy Requirements and an evidence-based background for MSFR safety assessment (in session "6.4 Simulation Tools for Safety Analysis")
TECHNICAL AND ECONOMICAL FEATURES OF COMMERCIAL SODIUM FAST REACTOR IN FRANCE (in session "7.1 Sustainability: Economics, Environment, and Proliferation")
TECHNOLOGICAL SUPPORT OF THE NON-PROLIFERATION FOR SVBR-100 FUEL CYCLES (in session "7.1 Sustainability: Economics, Environment, and Proliferation")
Tensile testing of sub-sized T91 and 316L steel specimens in liquid lead (in session "4.2 Structural, Novel, and Large Components Materials")
The "ALFRED White Book": a business card of the project (in session "Poster Session")
The Code Complex for Computational Evaluation of Technical Solutions and Optimization of Processes Parameters of CNFC (in session "Poster Session")
The fluid structure interaction of narrow gaps between thin-wall coaxial structures in fast reactors (in session "6.6 Fuel Performance and Material Modelling")
The influence of isotopic composition of plutonium in fast reactor fuel on the reactivity margin (in session "3.1 Fuel Cycle Scenarios")
The initial stage of closing the NFC of two-component nuclear power. Challenges and solutions (in session "3.1 Fuel Cycle Scenarios")
The INPRO project studies on the double-component nuclear power systems with the closed fuel cycle and fast reactors: past and future (in session "7.1 Sustainability: Economics, Environment, and Proliferation")
The neutronic study of the nitride fuel loaded CiADS core (in session "Poster Session")
The role of pyrochemical processing in a NetZero economy in the UK (in session "3.3 Reprocessing, Partitioning, and Transmutation")
The SAIGA in-pile experimental program to qualify the SIMMER calculation tool in SFR Severe Accident Conditions (in session "2.4 Severe Accidents")
The Severe Accident Management of the high-power SFR with loss of the heat removal from the core (in session "2.3 Accident Analysis")
The solution of nuclide kinetic equation for fast reactor in the OpenBPS code with options of choosing calculation method and uncertainties analysis. (in session "6.1 Neutronics")
The Status of the ALFRED Project (in session "1.1 Overviews and Fundamentals of Fast Reactors")
The Versatile Test Reactor (VTR) Approach to Sodium Fire Hazards Analysis and Protection System Methodology (in session "2.3 Accident Analysis")
The Westinghouse Lead Fast Reactor: overview and progress in development (in session "1.2 Innovative Design Advances")
The working capacity analysis of boron carbide after two-year operation as an emergency protection material of the fast reactor (in session "Poster Session")
The δ-ferrite transformation behavior and mechanical properties of 316H weld metal during high temperature service (in session "4.2 Structural, Novel, and Large Components Materials")
Thermal hydraulic assessment of the performance of secondary sodium system based decay heat removal circuit (in session "2.2 Safety Design and Analysis")
Thermal Hydraulic Simulation of Loss of Flow Without Scram Test in FFTF using DYANA-P code (in session "Poster Session")
Thermally conductive liquid-metal sublayer in fuel element (in session "4.1 Advanced Reactor Cladding and Core Material, Coolants, and Related Chemistry")
Thermodynamic simulation of the oxidation processes at the reprocessing of spent nuclear fuel in the LiCl-KCl melt (in session "Poster Session")
THERMOHYDRAULIC TESTS IN JUSTIFICATION OF DESIGN CHARACTERISTICS OF THE BREST-OD-300 RP STEAM GENERATOR (in session "5.2 Experimental Programs I")
Towards design guidelines for fast reactor oxide fuel pins with high Pu content: driving post irradiation examination by benchmarking European fuel performance codes (in session "3.2 Development of innovative fuels: design and properties irradiation")
TRAINING OF NEW GENERATION SPECIALISTS IN THE FIELD OF FAST NEUTRON REACTORS AND NUCLEAR FUEL CYCLE CLOSURE (in session "9.1 Education, Profesional Development, and Knowledge Management")
Transient 3D simulations for the ASTRID reactor: preliminary results for the ULOF initiation phase (in session "2.4 Severe Accidents")
Transmutation efficiency of minor actinides in fast-and thermal-spectrum molten salt reactors (in session "Poster Session")
Transmutation of minor actinides in a fast reactor with uranium-curium fuel (in session "Poster Session")
Treatment of sodium of Superphenix Fast Breeder Reactor (in session "8.1 SFR Commissioning, Operation, and Decommissioning")
Two-Component Energy Industry under Conditions of Closed Nuclear Fuel Cycle: Economic Benefits (in session "Poster Session")
TYPES OF CHEMICAL COMPOUNDS IN THE ASSESSMENT OF RADIATION AND HYGIENIC HAZARDS WHEN WORKING WITH IRRADIATED NITRIDE FUEL (in session "3.4 Advanced Fuel Development")
ULTRASONOSCOPY SYSTEM “VIZUS” FOR SODIUM-COOLED BN-TYPE REACTORS (in session "Poster Session")
United States Key Note (in session "Plenary 1. Keynotes from Member States")
Uranium and mixed uranium-plutonium nitrides thermal stability (in session "3.2 Development of innovative fuels: design and properties irradiation")
Verification and validation of neutronic codes using the start-up fuel load and criticality tests performed in the China Experimental Fast Reactor (in session "Special Session: IAEA Coordinated Research Projects")
Verification and validation of the CEFR Serpent model for the generation of reference solutions and Cross Sections database for the deterministic code AZNHEX (in session "Poster Session")
Verification of SARAX Code for the Transient Analysis of Sodium-cooled Fast Reactor (in session "6.3 Multiscale and Multiphysics Calculations")
Verification of the SPL module of the neutron diffusion code AZNHEX through Neutronics Benchmark of CEFR Start-Up Tests (in session "6.1 Neutronics")
Versatile Test Reactor (VTR) Experimental Capabilities (in session "5.3 Experimental Programs II")
Versatile Test Reactor (VTR) Project Mission and Status (in session "5.1 Experimental Reactors and Facilities")
VERSATILE TEST REACTOR: CONCEPTUAL CORE DESIGN OVERVIEW (in session "5.1 Experimental Reactors and Facilities")
VERSATILE TEST REACTOR: CORE SYSTEM DESIGN REQUIREMENTS TO SUPPORT ADVANCED REACTOR DEVELOPMENT (in session "Poster Session")
YGE Winner: Advanced Functional Materials for Next-Generation Fuel Reprocessing (in session "Plenary 2. International Organizations and YGE Winners")
YGE Winner: Production of Mo-99 isotope in the BN reactor by beryllium blocks (in session "Plenary 2. International Organizations and YGE Winners")
YGE Winner: Small Modular Fast Reactors for the ASEAN Region: Implementation of the TRISO Fuel Particle Concept as a Regional Variant of the Fast Reactor (in session "Plenary 2. International Organizations and YGE Winners")
Мechanistic code BERKUT-U: self-consistent modeling of fuel rods thermomechanical behavior and processes in the fuel of fast breeder reactors (in session "6.6 Fuel Performance and Material Modelling")
Мodeling of water leak into sodium in the BN-600 steam generator (in session "Poster Session")
Include materials from selected contributions
