3-D Core Design of the TRU-Incinerating Thorium RBWR Using Accident Tolerant Cladding (in session "Poster Session 1")
3D Modeling of Fuel Handling System for PFBR Operator Training Simulator (in session "6.1 CFD and 3D Modeling")
3D SIMULATION IN THE PLEIADES SOFTWARE ENVIRONMENT FOR SODIUM FAST REACTOR FUEL PIN BEHAVIOR UNDER IRRADIATION (in session "5.10 Fuel Modeling and Simulation")
1992-2017: 25 years of success story for the Development of Minor Actinides Partitioning Processes (in session "4.1 Fuel Cycle Overview")
"Peculiarities of behavior of Coated Particle fuel in the core of Fast Gas Reactor BGR-1000" (in session "Poster Session 1")
'EURATOM SUCCESS STORIES’ IN FACILITATING PAN-EUROPEAN E&T COLLABORATIVE EFFORTS (in session "8.1 Professional Development and Knowledge Management - I")
(U,Pu)O2-x MOX pellet for Astrid reactor project (in session "Poster Session 1")
A comprehensive study of the dissolution of spent SFR MOX fuel in boiling nitric acid (the PHENIX NESTOR-3 case) (in session "4.2 Reprocessing and Partitioning")
A Concept of VVER-SCP reactor with fast neutron spectrum and self-provision by secondary fuel (in session "1.8 INNOVATIVE REACTOR DESIGNS")
A Conceptual design of engineering-scale plant applied the simplified MA-bearing fuel fabrication process (in session "Poster Session 1")
A Demand Driven Way of Thinking Nuclear Development – Neutron Physical Feasibility of a Reactor Directly Operating SNF from LWR (in session "Poster Session 1")
A High Density Uranium Zirconium Carbonitride LEU Fuel for Application in Fast Reactors (in session "Poster Session 1")
A Mechanistic Source Term Calculation for a Metal Fuel Sodium Fast Reactor (in session "Poster Session 1")
A Preliminary Study of P&T Scenario on a Sustainable Energy System in China (in session "Poster Session 1")
A proposal for a pan-European E&T programme supporting the development and deployment of ALFRED (in session "8.1 Professional Development and Knowledge Management - I")
A safe and competitive lead-cooled small modular fast reactor concept for a short-term deployment (in session "Panel 2: Small and Medium sized fast reactors")
Actual Status of the Development of Multigroup XS Libraries for the Gas-cooled Fast Reactor in Slovakia (in session "Poster Session 2")
Advanced Coupling Methodology for Thermal-hydraulic calculations (in session "Poster Session 2")
Advanced Design Features of MOX Fuelled Future Indian SFRs (in session "1.1 SFR DESIGN & DEVELOPMENT - 1")
Advanced Energy Conversion for Sodium-Cooled Fast Reactors (in session "Poster Session 1")
ADVANCED FLOW-SHEET FOR PARTITIONING OF TRIVALENT ACTINIDES FROM FAST REACTOR HIGH ACTIVE WASTE (in session "4.3 Partitioning and Sustainability")
Advanced Reactor PSA Methodologies for System Reliability Analysis and Source Term Assessment (in session "Poster Session 1")
Advanced sodium-cooled fast reactor development regarding GIF safety design criteria (in session "1.1 SFR DESIGN & DEVELOPMENT - 1")
Advances in the Development of the SAS4A Code Metallic Fuel Models for the Analysis of PGSFR Postulated Severe Accidents (in session "3.2 Core Disruptive Accident")
ALLEGRO Core Neutron Physics Studies (in session "1.4 CORE AND DESIGN FEATURES - 1")
Americium Retention During Metallic Fuel Fabrication (in session "Poster Session 1")
An Assessment of Fission Product Scrubbing in Sodium Pools Following a Core Damage Event in a Sodium Cooled Fast Reactor (in session "Poster Session 1")
An assessment of transient over-power accident in the PGSFR (in session "3.2 Core Disruptive Accident")
Analyses of unprotected transients in GFR (ALLEGRO) and SFR reactors supporting the group constant generation methodology (in session "Poster Session 2")
Analysis of experimental data on fission gas release and swelling in mononitride fuel irradiated in BR-10 reactor (in session "5.2 Advanced Fast Reactor Fuel Development II")
Analysis of Irradiation Ability of China Experimental Fast Reactor (in session "Poster Session 2")
Analysis of the BFS-115-1 experiments (in session "6.4 Neutronics – 2")
Analysis of the Characteristics of the Fast Breeder Reactor with Metallic Fuel (in session "1.2 SFR DESIGN & DEVELOPMENT - 2")
Analysis of the EBR-II SHRT-45R neutronics benchmark with ERANOS-2.0 (in session "Poster Session 2")
Analysis of the SVBR-100 nuclear fuel cycle by means of the advanced nuclear fuel cycle assessment methodology (ATTR) (in session "7.4 Fuel Cycle Analysis")
ANALYSIS OF VARIOUS APPROXIMATIONS IN NEUTRONIC CALCULATIONS OF TRANSIENT IN FAST REACTORS (in session "Poster Session 2")
APPLICATION OF PHYSICAL MODELING WHEN CALIBRATING HIGH RANGE ELECTROMAGNETIC FLOWMETERS (in session "Poster Session 1")
Application of CFD simulation to validate the BREST-OD-300 primary circuit design (in session "Poster Session 2")
Application of Heterogeneous Fuel Assemblies in the Core of Modular Fast Sodium Reactor (in session "Poster Session 1")
Applications of the DNS CONV-3D Сode for Simulations of Liquid Metal Flows (in session "6.1 CFD and 3D Modeling")
ARRANGEMENT OF THE BN-600 REACTOR CORE REFUELING AT TRANSITION TO THE INCREASED FUEL BURN-UP (in session "2.3 Decommissioning of Fast Reactors and Radioactive Waste Management")
Assessment of a nuclear energy system based on the integral indicator of sustainable development (in session "7.1 Sustainability of Fast Reactors")
Assessment of accuracy from the use of point kinetics when analyzing transition processes in high power fast reactor (in session "Poster Session 1")
Assessment of Creep Damage Evaluation Methods for Grade 91 Steel in the ASME and JSME Nuclear Codes (in session "Poster Session 2")
Assessment of the anticipated improvement of the environmental footprint of future nuclear energy systems (in session "4.1 Fuel Cycle Overview")
Assessment of the reactivity effects of Gas cooled Fast Reactor (in session "Poster Session 1")
ASTRID - An original and efficient project organization (in session "7.3 Non Proliferation Aspects of Fast Reactors")
ASTRID French SFR: Progress in Sodium Gas Heat Exchanger development (in session "5.5 Large Component Technology I")
ASTRID FUEL HANDLING ROUTE FOR THE BASIC DESIGN (in session "1.3 SYSTEM DESIGN AND VALIDATION")
ASTRID hot cells (in session "Poster Session 1")
ASTRID reactor: design overview and main innovative options for Basic Design (in session "Poster Session 1")
Autonomous Reactivity Control (in session "3.6 Safety Analysis")
Basic principles for lifetime and structural integrity assessment of BN-600 and BN-800 fast reactors components with regard for material degradation (in session "5.8 Structural Materials")
Basic Visualization Experiments on Eutectic Reaction of Boron Carbide and Stainless Steel under Sodium-Cooled Fast Reactor Conditions (in session "Poster Session 2")
Benchmark Analysis of EBR-II SHRT45R using MARS-LMR (in session "Poster Session 2")
Benchmark Between EDF And IPPE On The Behavior Of Low Sodium Void Reactivity Effect Sodium Fast Reactor During An Unprotected Loss Of Flow Accident (in session "Poster Session 1")
Benchmark Evaluation of Dounreay Prototype Fast Reactor Minor Actinide Depletion Measurements (in session "6.3 Neutronics - 1")
BERKUT – Best Estimate Code for Modelling of Fast Reactor Fuel Rod Behavior under Normal and Accidental Conditions (in session "5.10 Fuel Modeling and Simulation")
BISON for Metallic Fuels Modeling (in session "Poster Session 1")
BN-800 core with MOX fuel (in session "1.4 CORE AND DESIGN FEATURES - 1")
BOR-60 REACTOR OPERATIONAL EXPERIENCE AND EXPERIMENTAL CAPABILITIES (in session "6.9 Research Reactors")
BREST-OD-300 REACTOR FACILITY. DEVELOPMENT STAGES AND JUSTIFICATION (in session "1.5 LFR DESIGN & DEVELOPMENT")
Burnup Analysis for BN-600 Reactor Core fueled with MOX fuel and Minor Actinides (in session "Poster Session 2")
CALCULATION AND EXPERIMENTAL ANALYSIS OF NEUTRONIC PARAMETERS OF THE BN-800 REACTOR CORE AT THE STAGE OF REACHING FIRST CRITICALITY FOLLOWED BY RATED POWER TESTING (in session "6.9 Research Reactors")
Calculation and Experimental Data Analysis of Neutron Spatial/Energy Distribution in the BOR-60 Blanket (in session "Poster Session 2")
CALCULATION OF NEUTRONIC PARAMETERS IN SUPPORT OF A BOR-60 EXPERIMENTAL FA WITH MODERATING ELEMENTS (in session "Poster Session 2")
CFD investigation of thermal-hydraulic characteristics in a SFR fuel assembly (in session "Poster Session 2")
CFD Simulation of Corium / Materials Interaction for Severe Accidents (in session "Poster Session 1")
CHALLENGES DURING CONSTRUCTION OF SODIUM PIPING SYSTEMS FOR 500MWe PROTOTYPE FAST BREEDER REACTOR (in session "Poster Session 1")
Challenges During Manufacture of Reactor Components of PFBR (in session "5.9 Large Component Technology II")
CHALLENGES IN THE FABRICATION AND RECYCLING OF MIXED CARBIDE FUEL (in session "Poster Session 1")
Change in Mechanical Properties of Spent Fast Reactor Claddings (in session "Poster Session 1")
Characterization of LBE Non-isothermal Natural Circulation by Experiments with HELIOS Test Loop and Numerical Analyses (in session "Poster Session 2")
Chemical compatibility with liquid sodium after in service solicitations: feedback on stainless steel in French sodium Fast reactor after 35 years of operation (in session "5.6 Liquid Metal Technologies")
Chugging boiling in low-void SFR core: new phenomenology of unprotected loss of flow (in session "Poster Session 1")
CIRCE-ICE EXPERIMENTAL ACTIVITY IN SUPPORT OF LMFR DESIGN (in session "6.8 Experimental Facilities")
CLEAR-S: A Large Pool-type Components and Thermo-hydraulic Integrated Test Facility for China Lead based reactor (in session "6.8 Experimental Facilities")
Closed fuel cycle technologies based on fast reactors as the corner stone for sustainable development of nuclear power (in session "Plenary Session 26 June")
Closing Remarks by Conference General Chair (in session "Closing Session")
Closing Remarks by Conference General Co-Chair (in session "Closing Session")
Closing Up Nuclear Fuel Cycle In a Two-Component System with Thermal And Fast Neutron Reactors (in session "7.3 Non Proliferation Aspects of Fast Reactors")
Code Qualification Plan for an Advanced Austenitic Stainless Steel, Alloy 709, for Sodium Fast Reactor Structural Applications (in session "Poster Session 1")
Codes of New Generation Developed for Breakthrough Project (in session "6.6 Coupled Calculations")
COMPARATIVE ANALYSIS OF ELECTRICITY GENERATION FUEL COST COMPONENT AT NPPs WITH WWER AND BN-TYPE REACTOR FACILITIES (in session "7.2 Economics of Fast Reactors")
Comparative analysis of nuclear energy lexicon (in session "Poster Session 2")
Comparison of fast reactors performance in the closed U-Pu and Th-U cycle (in session "4.1 Fuel Cycle Overview")
Comparison of Innovative Nuclear Energy Systems Based on Selected Key Indicators and Their Weighing Factors (in session "7.1 Sustainability of Fast Reactors")
Complex discussion of inherent safety fast reactors start-up with enriched uranium concept (strategical, economical aspects, problems of neutron physics etc.). R&D program proposal (in session "Poster Session 1")
Compliance of Korean SFR Safety Design Approaches with Generation-IV Safety Design Criteria (Korea, R. of) (in session "Panel 1: Development and Standardization of Safety Design Criteria (SDC) and Guidelines (SDG) for Sodium Cooled Fast Reactors")
COMPONENT HANDLING SYSTEM : PFBR AND BEYOND (in session "1.3 SYSTEM DESIGN AND VALIDATION")
Computational Analysis Code Development for core and primary system thermal hydraulic design of SFR (in session "Poster Session 2")
Computational Analysis Code Development for Emergency Heat Removal of Pool-style Fast Reactors (in session "Poster Session 2")
Computational investigation of nuclear waste incineration efficiency in a subcritical molten salt driven by 50-100 MeV protons (in session "Poster Session 1")
Computational modeling of flow blockage in fuel subassemblies and molten material relocation in sodium cooled fast reactors (in session "3.2 Core Disruptive Accident")
Computational modelling of inter-wrapper flow and primary system temperature evolution in FBTR under extended Station Blackout (in session "Poster Session 1")
Concept of multifunctional fast neutron research reactor (MBIR) core with metal (U-Pu-Zr)-fuel (in session "Poster Session 2")
Conceptual design of fuel and radial shielding sub-assemblies for ASTRID (in session "5.1 Advanced Fast Reactor Fuel Development I")
Concluding Report on the Technical Sessions (in session "Closing Session")
Conclusions of a Benchmark Study on the EBR-II SHRT-45R Experiment (in session "6.11 IAEA Benchmark on EBR-II Shutdown Heat Removal Tests")
Concurrent Trends in Indian Fast Reactor Fuel Reprocessing Programme (in session "4.1 Fuel Cycle Overview")
Considerations on GEN IV safety goals and how to implement them in future Sodium-cooled Fast Reactors (France) (in session "Panel 1: Development and Standardization of Safety Design Criteria (SDC) and Guidelines (SDG) for Sodium Cooled Fast Reactors")
Controlling FCCI with Pd in metallic fuel (in session "Poster Session 1")
CORE CONDITON MONITORING IN ADVANCED COMMERCIAL SODIUM BN-1200 (in session "1.6 CORE AND DESIGN FEATURES - 2")
Corrosion behavior of tube steel for BREST-OD-300 steam generator (in session "Poster Session 1")
CORROSION OF 12X18H10T STEEL IN Ce-, Nd- AND U-CONTAINING MOLTEN LiCl-KCl EUTECTIC (in session "Poster Session 1")
Coupled calculations for the fast reactors safety justification with the EUCLID/V1 integrated computer code (in session "6.6 Coupled Calculations")
Creep resistance and fracture toughness of recently-developed optimized Grade 92 and its weldments for advanced fast reactors (in session "5.4 Advanced Fast Reactor Cladding Development II")
Current status and future view of the fast reactor cycle technology development in Japan (in session "Plenary Session 27 June")
Current status and progression of GERMINAL fuel performance code for SFR oxide fuel pins (in session "5.10 Fuel Modeling and Simulation")
Current status of GIF collaborations on sodium-cooled fast reactor system (in session "1.1 SFR DESIGN & DEVELOPMENT - 1")
Current Status of Next Generation Fast Reactor Core & Fuel Design and Related R&D in Japan (in session "7.1 Sustainability of Fast Reactors")
Current Thermal Hydraulic Activities on Sodium-cooled Fast Reactors in Japan (in session "3.6 Safety Analysis")
DECAY HEAT REMOVAL SYSTEM IN THE SECONDARY CIRCUIT OF THE SODIUM-COOLED FAST REACTOR AND EVALUATION OF ITS CAPACITY (in session "Poster Session 1")
Decay-heat removal in accidents in fast reactors with liquid metal coolant (in session "Poster Session 1")
Density of sodium along the Liquid-Vapor Coexistence Curve, including the Critical Point (in session "6.2 Thermal Hydraulics Calculations and Experiments")
Dependability of the fission chambers for the neutron flux monitoring system of the French GEN-IV SFR (in session "2.3 Decommissioning of Fast Reactors and Radioactive Waste Management")
DEPENDENCE OF INTERMEDIATE HEAT EXCHANGER LIFE ON PRIMARY SODIUM HEATING RATE DURING POWER RAISING (in session "Poster Session 2")
Design and Development of Stroke Limiting Device for Control & Safety Rod Drive Mechanisms (CSRDMs) of future FBRs (in session "Poster Session 1")
Design and Fabrication of Closed Loop Systems (CLS) for the Fast Flux Test Facility (FFTF) (in session "Poster Session 2")
Design Evolutions of the Molten Salt Fast Reactor (in session "1.7 ADS AND OTHER REACTOR DESIGNS")
Design modifications of Instrumentation & Control System of future FBRs (in session "2.2 Commissioning and Operating Experience of Fast Reactors II")
Design of a nitride-fueled lead fast reactor for MA transmutation (in session "1.8 INNOVATIVE REACTOR DESIGNS")
Design of Sleeve Valve mechanism for Primary Sodium Pump of future FBR (in session "5.9 Large Component Technology II")
Design Safety Limits for Transients in a Metal Fuelled Reactor (in session "3.6 Safety Analysis")
DESIGN VALIDATION OF PFBR FUEL SUBASSEMBLY TRANSPORTATION CASK WITH MOCKUP TRIAL RUN (in session "Poster Session 2")
Detailed engineering neutron codes for calculations of fast breeder reactors (in session "6.9 Research Reactors")
Detection and analysis of fuel cladding damages using gamma ray spectroscopy (in session "Poster Session 2")
DETERMINISTIC SAFETY ANALYSIS OF REACTOR BREST-OD-300 (in session "3.5 General Safety Approach")
Developing an open-source multi-physics tool for simulating advanced nuclear reactors (in session "YGE Panel")
Development and Applications of Nuclear Design and Safety Assessment Program SuperMC for Fast Reactor (in session "Poster Session 2")
Development and Demonstration of Ultrasonic Under-Sodium Viewing System for SFRs (in session "5.6 Liquid Metal Technologies")
Development and Deployment of Knowledge Management Portal for Fast Breeder Reactors (in session "8.1 Professional Development and Knowledge Management - I")
Development and Validation of EBRDYN code by Benchmark Analysis of EBR-II SHRT-17 Test (in session "Poster Session 2")
Development and Validation of Multi-scale Thermal-Hydraulics Calculation Schemes for SFR Applications at CEA (in session "6.2 Thermal Hydraulics Calculations and Experiments")
Development experience for experimental reactor facility cooled vith evaporating liquid metals (in session "Poster Session 2")
Development of core and structural materials for fast reactors (in session "5.3 Advanced Fast Reactor Cladding Development I")
Development of Electromagnetic Devices for Sodium Cooled Fast Reactor Application (in session "5.1 Advanced Fast Reactor Fuel Development I")
Development of Fast Reactors in the USSR and the Russian Federation; Malfunctions and Incidents in the Course of their Operation and Solution of Problems. (in session "Poster Session 1")
Development of Flow Identification Technology for the PGSFR Thermal Fluidic Design Validation (in session "6.9 Research Reactors")
Development of innovating Na leak detector on pipes (in session "Poster Session 2")
Development of innovative fast reactor nitride fuel in Russian Federation: state-of-art. (in session "5.1 Advanced Fast Reactor Fuel Development I")
Development of Research Nuclear Power Facility with MBIR Multi-Purpose Fast Neutron Research Reactor (in session "Poster Session 2")
Development of Reverse Flow Blockage Device for Primary Sodium Pumps of Fast Breeder Reactor (in session "YGE Panel")
Development of Safety Design Criteria for the Lead-cooled Fast Reactor (in session "3.5 General Safety Approach")
Development of Safety, Irradiation, and Reliability Databases based on Past U.S. SFR Testing and Operational Experiences (in session "Poster Session 2")
Development of Smart Component Based Framework for Dynamic Reliability Analysis of Nuclear Safety Systems (in session "3.3 Probabilistic Safety Assessment")
Development of steam-water cycle chemistry for steam generator of research reactor MBIR (in session "5.7 Chemistry Related Technology")
Development of the built-in primary sodium purification system for the (in session "5.5 Large Component Technology I")
Development of the new generation power unit with the BN-1200 reactor (in session "1.1 SFR DESIGN & DEVELOPMENT - 1")
Development of the U.S. Sodium Component Reliability Database (in session "Poster Session 1")
Development of Tri-iso-Amyl Phospahte (TiAP) based solvent extraction process as an alternate method for the processing of metallic alloy fuels (U-Pu-Zr and UZr) (in session "YGE Panel")
Development of Ultra Sub-size Tensile Specimen for Evaluation of Tensile Properties of Irradiated Materials (in session "Poster Session 1")
Development of under sodium viewer for next generation sodium-cooled fast reactor (in session "2.1 Commissioning and Operating Experience of Fast Reactors I")
Dynamic probabilistic risk assessment at a design stage for a sodium fast reactor. (in session "3.3 Probabilistic Safety Assessment")
DYNAMIC TEST OF EXTRACTION PROCESS FOR AMERICIUM PARTITIONING FROM THE PUREX RAFFINATE (in session "4.3 Partitioning and Sustainability")
EBR-II Passive Safety Demonstration Tests Benchmark Analyses (in session "6.11 IAEA Benchmark on EBR-II Shutdown Heat Removal Tests")
EBR-II SHRT-17 and SHRT-45R Benchmark Analyses (in session "Poster Session 2")
ECOLOGICAL ASPECTS OF THE USE OF FAST REACTORS IN A CLOSED NUCLEAR FUEL CYCLE UNDER THE “PRORYV” PROJECT (in session "Poster Session 1")
Eddy current flowrate and local ultrasonic velocity measurements in liquid sodium (in session "6.7 Experimental Thermal Hydraulics")
EFFECT OF INLET TEMPERATURE AND OPERATING LINEAR HEAT RATING (LHR) ON THE MAXIMUM ACHIEVABLE BURNUP OF MK-1 CARBIDE FUEL IN FBTR (in session "6.2 Thermal Hydraulics Calculations and Experiments")
Effects of Oxygen Partial Pressure During Sintering at Laboratory and Industrial Scales on FR MOX Fuels (in session "5.2 Advanced Fast Reactor Fuel Development II")
ELECTRICAL CONDUCTIVITY OF MOLTEN LiCl-KCl EUTECTIC WITH COMPONENTS OF SPENT NUCLEAR FUEL (in session "Poster Session 1")
Eligibility of Small Molten Salt Fast Reactor (S-MSFR) (in session "Panel 2: Small and Medium sized fast reactors")
Equipment cost estimation for pilot demonstration lead-cooled fast-neutron reactor BREST-OD-300 (in session "7.2 Economics of Fast Reactors")
ESFR-SMART: new Horizon-2020 project on SFR safety (in session "Poster Session 1")
Evaluation of Anticipated Transient without Scram for SM-SFR using SAS4A/SASSYS-1 (in session "Poster Session 1")
EVALUATION OF COBALT FREE COATINGS AS HARDFACING MATERIAL CANDIDATES IN SODIUM FAST REACTOR (in session "Poster Session 1")
Evaluation of data and model uncertainties and their effect on the fuel assembly temperature field of the ALFRED Lead-cooled Fast Reactor (in session "Poster Session 2")
Evaluation of irradiation-induced point defects migration during neutron irradiation in modified 316 stainless steel (in session "Poster Session 1")
Evaluation of multiple primary coolant leakages accidents in Monju with consideration of passive safety features (in session "3.4 Sodium leak/fire and other safety issues")
Evaluation of the OECD/NEA/SFR-UAM Neutronics Reactivity Feedback and Uncertainty Benchmarks (in session "Poster Session 2")
Evaluation of βeff measurements from BERENICE programme with TRIPOLI4® and uncertainties quantification (in session "6.5 Uncertainty Analysis and Tools")
Evaluation results of BN-1200 compliance with the requirements of GENERATION IV and INPRO (in session "7.1 Sustainability of Fast Reactors")
Evolution of the collective radiation dose from the nuclear reactors through the 2nd to the 4th generation. (in session "3.5 General Safety Approach")
Examination of ChS-68 Steel Used as a BN-600 Reactor Cladding Material (in session "Poster Session 1")
Examination of Fast Reactor Materials and Structural Elements at JSC “INM” Premises (in session "5.3 Advanced Fast Reactor Cladding Development I")
EXPERIENCE AND APPLICABILITY OF HIGH DENSE METAL URANIUM IN ADVANCED BN-REACTORS (in session "5.10 Fuel Modeling and Simulation")
EXPERIENCE OF COMMISSIONING OF BN-800 CORE DIAGNOSTIC SYSTEM (SDRU) (in session "Poster Session 2")
EXPERIENCE OF COMMISSIONING OF THE SECTORAL MONITORING TIGHTNESS SYSTEM OF FUEL ELEMENTS CLADDINGS (SSKGO) OF RF BN-600, RF BN-800 (in session "2.2 Commissioning and Operating Experience of Fast Reactors II")
Experience on MOX fuel fabrication for fast reactor at PFPF (in session "Poster Session 1")
Experiences during construction & Commissioning of electrical power Generation and Evacuation systems in PFBR (in session "Poster Session 2")
Experimental investigations of velocity and temperature fields, stratification phenomena in a integral water model of fast reactor in the steady state forced circulation (in session "6.2 Thermal Hydraulics Calculations and Experiments")
Experimental qualification of rotatable plug seals for Sodium Fast Reactor on a large scale test stand (in session "5.5 Large Component Technology I")
EXPERIMENTAL SEISMIC QUALIFICATION OF DIVERSE SAFETY ROD AND ITS DRIVE MECHANISM OF PROTOTYPE FAST BREEDER REACTOR (in session "1.3 SYSTEM DESIGN AND VALIDATION")
Expriment and Analysis of Flow distribution of MOX Assembly (in session "Poster Session 2")
Extending the grid plate life - Incorporation of lower axial shield for FBTR (in session "Poster Session 1")
Extension to Heavy Liquid Metal coolants of the validation database of the ANTEO+ sub-channel code (in session "6.2 Thermal Hydraulics Calculations and Experiments")
External Assessment of the U.S. Sodium-Bonded Spent Fuel Treatment Program (in session "4.3 Partitioning and Sustainability")
Fabrication and Evaluation of Advanced Cladding Tube for PGSFR (in session "5.8 Structural Materials")
Fabrication Characteristics of Injection-cast Metallic Fuels (in session "5.1 Advanced Fast Reactor Fuel Development I")
Fabrication process of NpO2 pellets (in session "Poster Session 1")
FACILITY FOR ADVANCED FUELS THROUGH THE SOL-GEL METHOD (in session "Poster Session 1")
FALCON advancements towards the implementation of the ALFRED Project (in session "7.3 Non Proliferation Aspects of Fast Reactors")
Fast Neutron Reactors, Fuel Cycles and Problem of Nuclear Non-Proliferation (in session "7.4 Fuel Cycle Analysis")
Fast Reactor Development and International Cooperation (by Honorary General Chair) (in session "Opening Session")
Fast reactor systems in the German P&T and related studies (in session "7.4 Fuel Cycle Analysis")
Fast Reactors - The Belgian Regulatory Approach (in session "Poster Session 1")
Fast Reactors and Nuclear Cogeneration: A Market and Economic Analysis (in session "7.2 Economics of Fast Reactors")
FASTER Test Reactor Preconceptual Design (in session "1.2 SFR DESIGN & DEVELOPMENT - 2")
Feasibility and Challenges for Self-sustainable Long-Life SMR without Refueling (in session "Panel 2: Small and Medium sized fast reactors")
Feasibility of Burning Wave Fast Reactor Concept with Rotational Fuel Shuffling (in session "1.1 SFR DESIGN & DEVELOPMENT - 1")
Feasibility of MA Transmutation by (MA, Zr)Hx in Radial Blanket Region of Fast Reactor and Plan of Technology Development (in session "Poster Session 1")
FEATURES OF THE NUCLEAR FUEL CYCLE SYSTEMS BASED ON JOINT OPERATION OF FAST AND THERMAL REACTORS (in session "Poster Session 1")
Features of the physics of the MBIR reactor core (in session "Poster Session 2")
Features of the time dependence of the intensity of delayed neutrons in the range of 0.02 s in the fission 235U by thermal and fast neutrons. (in session "Poster Session 2")
Final Results and Lessons Learned from EBR-II SHRT-17 Benchmark Simulations (in session "6.11 IAEA Benchmark on EBR-II Shutdown Heat Removal Tests")
First assessment of a digestion method applied to recover plutonium from refractory residues after dissolving spent SFR MOX fuel in nitric acid (in session "4.2 Reprocessing and Partitioning")
Fission product and swelling behaviour in FBTR mixed carbide fuel (in session "5.2 Advanced Fast Reactor Fuel Development II")
FRACTURE STRAIN AND FRACTURE TOUGHNESS PREDICTION FOR IRRADIATED AUSTENITIC STEELS OVER WIDE RANGE OF TEMPERATURES TAKING INTO ACCOUNT THE EFFECT OF SWELLING AND THERMAL AGEING (in session "5.3 Advanced Fast Reactor Cladding Development I")
Fuel Cladding Chemical Interaction Tests of Irradiated Metallic Fuel (in session "5.1 Advanced Fast Reactor Fuel Development I")
Fuel cycle studies of Generation IV fast reactors with the SITON v2.0 code and the FITXS burn-up scheme (in session "4.3 Partitioning and Sustainability")
Fuel Melting Margin Assessment of Fast Reactor Oxide Fuel Pins using a Statistical Approach (in session "5.2 Advanced Fast Reactor Fuel Development II")
Full-fledged affination extractive-crystallizng platform for technology validation of the fast reactor spent fuel reprocessing on fast neutrons – the results of first experiments (in session "Poster Session 1")
Fundamental Approaches to High-power Fast Reactor Core Development (in session "1.6 CORE AND DESIGN FEATURES - 2")
GEN IV Education and Training Initiative via Public Webinars (in session "8.1 Professional Development and Knowledge Management - I")
Group Discussion (in session "YGE Workshop")
Group Presentation (in session "YGE Workshop")
Heat transfer and temperature non-uniformities in pin bundles with heavy liquid metal coolant at various spacing ways (in session "6.7 Experimental Thermal Hydraulics")
Heat Transfer Performance Test for a Sodium-to-Air Heat Exchanger with an Inclined Finned-Tube Banks (in session "5.5 Large Component Technology I")
Helium Recovery from Guard Vessel Atmosphere of the ALLEGRO Reactor (in session "5.7 Chemistry Related Technology")
High temperature design and evaluation of forced draft sodium-to-air heat exchanger in PGSFR (in session "Poster Session 1")
Hot test of technique separation of americium and curium (in session "4.3 Partitioning and Sustainability")
How the Next Generation of People will shape the Next Generation of Nuclear (in session "YGE Panel")
How to take into account the fleet composition in order to evaluate Fast Breeder Competitiveness (in session "7.2 Economics of Fast Reactors")
Hydraulic Design and Evaluation of the PHTS Mechanical Pump of PGSFR (in session "Poster Session 2")
IAEA activities in the area of Nuclear Power Reactor Fuel Engineering (in session "5.3 Advanced Fast Reactor Cladding Development I")
IAEA NAPRO Coordinated Research Project: Physical Properties of Sodium Overview of the Reference Database and Preliminary Analysis Results (in session "8.2 Professional Development and Knowledge Management - II")
IAEA NEUTRONICS BENCHMARK FOR EBR-II SHRT-45R (in session "6.11 IAEA Benchmark on EBR-II Shutdown Heat Removal Tests")
IAEA’s Coordinated Research Project on EBR-II Shutdown Heat Removal Tests: An Overview (in session "6.11 IAEA Benchmark on EBR-II Shutdown Heat Removal Tests")
IAEA’s Fast Reactors Knowledge Portals and Catalogues (in session "8.2 Professional Development and Knowledge Management - II")
Identification of important phenomena under sodium fire accidents based on PIRT process (in session "3.4 Sodium leak/fire and other safety issues")
Impact of an accidental control rod withdrawal on the ALFRED core: tridimensional neutronic and thermal-hydraulic analyses (in session "3.7 Core Disruptive Accident Prevention")
Impact of nuclear data uncertainties on the reactivity coefficients of ALFRED (in session "Poster Session 2")
Impact of the irradiation of an ASTRID-type core during an ULOF with SIMMER-III (in session "Poster Session 1")
IMPLEMENTATION STATUS OF CONTAIN-LMR SODIUM CHEMISTRY MODELS INTO MELCOR 2.1 (in session "Poster Session 1")
Improving inherent safety BN-800 by the use of fuel assembly with (U, Pu)C microfuel. (in session "Poster Session 1")
Indian Fast Reactor Programme : Status and R&D Achievements (in session "Plenary Session 27 June")
Industrial Exploitation of Testing Ground for Treatment of Radwaste of Alkaline Coolants under Decommissioning of Fast Research Reactors (in session "2.3 Decommissioning of Fast Reactors and Radioactive Waste Management")
Innovative cold trap filtration technologies for reliable and economical exploitation of lead-bismuth eutectic cooled systems (in session "YGE Panel")
Innovative TRU Burning Fast Reactor Cycle Using Uranium-free TRU Metal Fuel - Core Design Progress - (in session "1.8 INNOVATIVE REACTOR DESIGNS")
INPRO: Fast Reactors and Enhanced Nuclear Energy Sustainability (in session "Plenary Session 28 June")
INSERTION RELIABILITY STUDIES FOR THE RBC-TYPE CONTROL RODS IN ASTRID (in session "Poster Session 1")
Inspection specifications leading to extended ASTRID Design rules (in session "Poster Session 2")
INTEGRAL EXPERIMENTS WITH MINOR ACTINIDES AT THE BFS CRITICAL FACILITIES: STATE-OF-THE-ART SURVEY, REEVALUATION AND APPLICATION (in session "6.3 Neutronics - 1")
INTEGRATED R&D TO VALIDATE INNOVATIVE EMERGENCY HEAT REMOVAL SYSTEM FOR BN-1200 REACTOR (in session "Poster Session 2")
International research center based on MBIR reactor – cornerstone for Generation 4 technologies development (in session "6.3 Neutronics - 1")
Introduction to the Workshop (in session "YGE Workshop")
Introduction to the YGE Panel (in session "YGE Panel")
Investigation of Radiation-Induced Swelling of EK-164 Steel, an Advanced Material for BN-600 and BN-800 Claddings (in session "Poster Session 1")
Investigation of steel corrosion products mass transfer in sodium (in session "Poster Session 1")
Investigation of the homogenization effect in sodium void reactivity in PGSFR (in session "Poster Session 2")
Investigations in a substantiation of high-temperature nuclear energy technology with fast-neutron reactor cooled by sodium for manufacture of hydrogen and other innovative applications (in session "Poster Session 1")
Isothermal transformation austenite-ferrite in a P92 steel (in session "Poster Session 1")
Justification of arrangement, parameters, and irradiation capabilities of the MBIR reactor core at the initial stage of operation (in session "6.9 Research Reactors")
Key features of design, manufacturing and implementation of laboratory and industrial equipment for Mixed Uranium – Plutonium Oxide (MOX) and Nitride fuel pellets fabrication in Russia (in session "Poster Session 1")
Knowledge Transfer and Management during long outage periods (in session "YGE Workshop")
Knowledge Transfer and Management for an active fleet of fast reactors (in session "YGE Workshop")
Knowledge Transfer and Management with interrupted development (in session "YGE Workshop")
Knowledge Transfer to Young Generation and Technical Reconstruction of BFS Complex (in session "YGE Workshop")
Learning from 1970 and 1980-Era Sodium Fire Experiments (in session "3.4 Sodium leak/fire and other safety issues")
LES-SIMULATION OF HEAT TRANSFER IN A TURBULENT PIPE (in session "Poster Session 2")
Lessons and strategies from PFBR to Future Fast Breeder Reactors (in session "1.2 SFR DESIGN & DEVELOPMENT - 2")
Lessons Learned from Fast Flux Test Facility Experience (in session "Poster Session 2")
LOGOS CFD software application for the analysis of liquid metal coolants in the fuel rod bundles geometries (in session "Poster Session 2")
Low-void-effect sodium-cooled core: Uncertainty of local sodium void reactivity as a result of nuclear data uncertainties (in session "Poster Session 2")
Main operation procedures for ASTRID gas power conversion system (in session "1.3 SYSTEM DESIGN AND VALIDATION")
Main outcomes from the JASMIN project: development of ASTEC-Na for severe accident simulation in Na cooled fast reactors (in session "Poster Session 1")
Main R&D objectives and results for under-sodium inspection carriers – Example of the ASTRID matting exceptional inspection carrier. (in session "2.1 Commissioning and Operating Experience of Fast Reactors I")
Manufacture, Installation and Adjustment of the BN-800 Reactor Plant Equipment (in session "2.1 Commissioning and Operating Experience of Fast Reactors I")
Mass Transfer Simulation Model for Justification Sodium Purification System Characteristics (in session "5.6 Liquid Metal Technologies")
Materials corrosion in Fast Reactor environment (in session "5.7 Chemistry Related Technology")
Mathematical modeling of the mononitride nuclear fuel production processes (in session "Poster Session 1")
Mechanical and Thermal Properties of (U,Pu)O2-x (in session "Poster Session 1")
Mechanical Design Evaluation of Fuel Assembly for PGSFR (in session "Poster Session 1")
Metal fuel for fast reactors, a new concept (in session "5.1 Advanced Fast Reactor Fuel Development I")
Methodical uncertainty of criticality precise calculations for fast lead reactor (in session "Poster Session 2")
Methods of controlling concentration of oxygen dissolved in heavy liquid metal coolants (lead and lead-bismuth) of nuclear reactors and test facilities (in session "5.7 Chemistry Related Technology")
Minimisation of Reactivity Margin for Equilibrium Core of Liquid Metal Cooled Fast Reactors (in session "3.7 Core Disruptive Accident Prevention")
Model validation of the ASTERIA-FBR code related to core expansion phase based on THINA experimental results (in session "Poster Session 1")
Modeling of hydrodynamic processes at a large leak of water into sodium in the fast reactor coolant circuit (in session "Poster Session 1")
Modeling of Lanthanide Transport in Metallic Fuels: Recent Progresses (in session "Poster Session 1")
Modeling of Phenix End-of-Life control rod withdrawal tests with the Serpent-DYN3D code system (in session "Poster Session 2")
Modeling of Processes in Austenitic Steel Produced Under Irradiation in Fast Reactors and Possibilities of Model Practical Application (in session "5.4 Advanced Fast Reactor Cladding Development II")
Modeling technologies of fuel cycles (in session "Poster Session 1")
Modelling and Simulation of Heat Transport System and Steam Power Transition System of CEFR (in session "6.1 CFD and 3D Modeling")
More precise definitions of the perturbation theory formulas for reactivity effects calculations (in session "Poster Session 2")
Multiscale computer modeling of nuclear fuel properties at radiation and thermal impacts (in session "Poster Session 1")
NACIE-UP: a HLM loop facility for natural circulation experiments (in session "6.7 Experimental Thermal Hydraulics")
Neutronic evaluation of a GFR of 100 MWt with reprocessed fuel and thorium using SCALE 6.0 and MCNPX (in session "6.4 Neutronics – 2")
Neutronic Self-sustainability of a Breed-and-Burn Fast Reactor Using Super-Simple Fuel Recycling (in session "Poster Session 1")
Neutronics Experimental Verification for ADS with China Lead-based Zero Power Reactor (in session "6.6 Coupled Calculations")
New catalog on (U,Pu)O2 properties for fast reactors and first measurements on irradiated and non-irradiated fuels within the ESNII+ project (in session "5.2 Advanced Fast Reactor Fuel Development II")
NEW NEUTRONIC CALCULATION CODES BASED ON DISCRETE ORDINATES METHOD USING METHODS OF FINITE DIFFERENCES AND FINITE ELEMENTS (in session "Poster Session 2")
New results on the continuous cooling behavior of an ASTM A335 P92 steel (in session "Poster Session 1")
Nuclear Reactor Modelling and Simulation Toolkit (NuReMoST) –Numerical Reactor Model Configuration System with Interface to Simulation Codes (in session "Poster Session 2")
Numerical Analysis of EBR-II Shutdown Heat Removal Test-17 using 1D Plant Dynamic Analysis Code coupled with 3D CFD Code (in session "Poster Session 2")
Numerical and Experimental Investigations of Tube-to-Tube Interaction of Air Heat Exchangers of PFBR under Seismic Excitations (in session "Poster Session 1")
Numerical Investigation of Sodium Spray Combustion Test with SPHINCS code (in session "Poster Session 1")
Numerical Simulation Method of Thermal Hydraulics in Wire-wrapped Fuel Pin Bundle of Sodium-cooled Fast Reactor (in session "6.1 CFD and 3D Modeling")
Numerical simulation of hydraulics and heat transfer in the BREST-OD-300 LFR fuel assembly (in session "6.1 CFD and 3D Modeling")
Numerical – experimental research in justification of fire (sodium) safety of sodium cooled fast reactors (in session "3.4 Sodium leak/fire and other safety issues")
Objectives and Status of the OECD/NEA sub-group on Uncertainty Analysis in Modelling (UAM) for Design, Operation and Safety Analysis of SFRs (SFR-UAM) (in session "6.5 Uncertainty Analysis and Tools")
On the feasibility of Breed-and-Burn fuel cycles in Molten Salt Reactors (in session "Poster Session 1")
On the possibility of using various types of fuel in the MBIR reactor core (in session "Poster Session 1")
On the rational design of fuel assemblies for reactor facilities from the standpoint of providing vibration strength (in session "6.8 Experimental Facilities")
On-site nuclear fuel cycle of “BREST” reactors (in session "Poster Session 1")
Opening Address by Director General, IAEA (by video message) (in session "Opening Session")
Opening Address by Director General, ROSATOM (by video message) (in session "Opening Session")
OPERABILITY VALIDATION OF FUEL PINS WITH CLADDINGS MADE OF EK164-ID STEEL IN THE BN-600 REACTOR (in session "5.4 Advanced Fast Reactor Cladding Development II")
OPERATING EXPERIENCE OF FBTR (in session "2.2 Commissioning and Operating Experience of Fast Reactors II")
Optimization of Passive Safety Devices FAST and SAFE for Sodium-cooled Fast Reactors (in session "3.7 Core Disruptive Accident Prevention")
Optimization of the thermomechanical treatment to achieve a homogeneous microstructure in a 14Cr ODS steel (in session "Poster Session 1")
OSCAR-Na validation against sodium loop experiments (in session "Poster Session 1")
Overview of Experiments for Physics of Fast Reactors from the International Handbooks of Evaluated Criticality Safety Benchmark Experiments and Evaluated Reactor Physics Benchmark Experiments (in session "Poster Session 2")
Overview of GEN-IV International Forum Activities. Status and Prospects of Fast Reactors (in session "Plenary Session 28 June")
Overview of NEA Activities Related to Fast Reactors (in session "Plenary Session 28 June")
Overview of the IAEA Activities in the Field of Fast Reactor Technology Development: Current State and Future Vision (in session "8.2 Professional Development and Knowledge Management - II")
Overview of the international cooperation and collaboration activities initiated and performed under the Technical Working Group on Fast Reactors in last 50 years (in session "Poster Session 2")
Overview of the Nuclear Energy Agency Scientific Activities on Advanced Fuel Cycles (in session "4.1 Fuel Cycle Overview")
Overview of the U.S. DOE fast reactor fuel development program (in session "Poster Session 1")
Overview of U.S. Fast Reactor Technology R&D Program (in session "1.1 SFR DESIGN & DEVELOPMENT - 1")
Panel Discussion (in session "Panel 1: Development and Standardization of Safety Design Criteria (SDC) and Guidelines (SDG) for Sodium Cooled Fast Reactors")
Participation of Mexico in the OECD/NEA SFR Benchmark using the Monte Carlo code Serpent (in session "Poster Session 2")
Passive Complementary Safety Devices for ASTRID severe accident prevention (in session "Poster Session 1")
Passive Safety Testing at the Fast Flux Test Facility Relevant to New LMR Designs (in session "Poster Session 2")
Passive Shutdown Systems for Liquid Metal-Cooled Fast Reactors (in session "Poster Session 1")
Performance Analysis of Various Thorium Fuel Options for the Sodium Cooled Fast Reactor (in session "7.4 Fuel Cycle Analysis")
Performance and sustainability assessment of nuclear energy deployment scenarios with fast reactors: advanced tools and application (in session "7.1 Sustainability of Fast Reactors")
Performance evaluation of ferroboron shielding material after irradiation in FBTR (in session "5.8 Structural Materials")
PERFORMANCE EVALUATION OF TIN OXIDE BASED SENSOR FOR MONIOTORING TRACE LEVELS OF H2 IN ARGON COVER GAS PLENUM OF FBTR (in session "Poster Session 2")
PERSONNAL TRAINING FOR THE "PRORYV" PROJECT AT THE SEVERSK TECHNOLOGICAL INSTITUTE OF NRNU MEPHI (in session "8.2 Professional Development and Knowledge Management - II")
Physical and technical basics of the concept of a competitive gas cooled fast reactor facility with the core based on coated fuel microparticles (in session "1.7 ADS AND OTHER REACTOR DESIGNS")
Physical start-up test of China Experimental Fast Reactor (in session "6.4 Neutronics – 2")
Physics Investigation of a Supercritical CO2–cooled Micro-Modular Reactor (MMR) for Autonomous Load-Follow Operation (in session "1.4 CORE AND DESIGN FEATURES - 1")
PLINIUS-2: a new corium facility and programs to support the safety demonstration of the ASTRID mitigation provisions under Severe Accident Conditions (in session "6.8 Experimental Facilities")
Possibility studies of a boiling water cooled traveling wave reactor (in session "Poster Session 1")
POSTREACTOR STATE OF THE STANDARD AND EXPERIMENTAL BN-600 FUEL KINDS (in session "Poster Session 1")
Potential Capabilities in Transmutation of Minor Actinides of the BOR-60 Reactor and MBIR Reactor under Construction (in session "Poster Session 2")
Precipitate phases in a weldment of P92 steel (in session "Poster Session 1")
PREDICTION OF CREEP-RUPTURE PROPERTIES FOR AUSTENITIC STAINLESS STEELS UNDERGONE NEUTRON IRRADIATION AT DIFFERENT TEMPERATURES (in session "5.8 Structural Materials")
Preliminary Design of Zero Power Reactor for CEFR MOX Core (in session "Poster Session 2")
Preliminary Inspection of Spent Fast Reactor Fuel Claddings (in session "5.4 Advanced Fast Reactor Cladding Development II")
Preliminary Safety Performance Assessment of ESFR CONF-2 Sphere-pac‐Fueled Core (in session "Poster Session 1")
Preliminary transient analyses of SEALER (in session "Poster Session 1")
Primary Analysis on The Nuclear Energy Development Scenario base on the U-Pu Multicycling with PWR, FR and CNFC in China (in session "7.4 Fuel Cycle Analysis")
Probabilistic Safety Analysis of NPP with BREST-OD-300 reactor (in session "Poster Session 1")
PROBABILISTIC SAFETY ANALYSIS RESULTS FOR BN REACTOR POWER UNITS (in session "3.3 Probabilistic Safety Assessment")
PROBLEMS OF CALCULATION MODELLING OF NITRIDE FUEL PERFORMANCE: DRAKON CODE (in session "5.10 Fuel Modeling and Simulation")
Progress in the ASTRID Gas Power Conversion System development (in session "1.3 SYSTEM DESIGN AND VALIDATION")
Progress of Design and related Researches of Sodium-cooled Fast Reactor in Japan (in session "1.2 SFR DESIGN & DEVELOPMENT - 2")
Proposal of Basic Principles of Maintenance Management for Prototype Reactors (in session "2.2 Commissioning and Operating Experience of Fast Reactors II")
Providing the competitiveness of nuclear energy in the implementation of PRORYV project (in session "7.2 Economics of Fast Reactors")
Pu recycling capabilities of ASTRID reactor (in session "4.2 Reprocessing and Partitioning")
Pyrochemical recycling of the nitride SNF of fast neutron reactors in molten salts as a part of the short-circuited nuclear fuel cycle (in session "4.3 Partitioning and Sustainability")
Quantitative Evaluation of the Post Disassembly Energetics of a Hypothetical Core Disruptive Accident in a Sodium Cooled Fast Reactor (in session "3.2 Core Disruptive Accident")
R&D status on in-sodium ultrasonic transducers for ASTRID inspection (in session "2.2 Commissioning and Operating Experience of Fast Reactors II")
RECENT ACTIVITIES OF THE SAFETY AND OPERATION PROJECT OF THE SODIUM-COOLED FAST REACTOR IN THE GENERATION IV INTERNATIONAL FORUM (in session "3.1 Safety Program")
Recent and Potential Advances of the HGPT methodology (in session "6.5 Uncertainty Analysis and Tools")
Recent suppling of 316L(N) stainless steel products for ASTRID (in session "5.8 Structural Materials")
Remote detection of raised radioactivity in emission from Beloyarsk nuclear power plant (in session "Poster Session 1")
Report on Panel 1: Safety Design Guidelines (in session "Closing Session")
Report on Panel 2: Small and Medium sized fast reactors (in session "Closing Session")
Report on the Young Generation Event (in session "Closing Session")
Reprocessing of fast reactors mixed U-Pu used nuclear fuel: studies and industrial test (in session "4.1 Fuel Cycle Overview")
Research and Development on Simulator of Fast Reactor in China (in session "6.6 Coupled Calculations")
Research and Pilot Fast Neutron Reactors in Russian Federation as the Ground for Development of Worldwide Commercial Technologies (in session "Plenary Session 27 June")
Research on modeling and simulation of the primary coolant system for China Experimental Fast Reactor (in session "Poster Session 2")
Research, development and deployment of fast reactors and related fuel cycle in China (in session "Plenary Session 26 June")
Resting Bottom Fast Reactor (in session "Poster Session 1")
Results of monitoring, using high-resolution neutron diffraction, of radiation-induced damages in claddings of fuel pins after their performance in the reactor BN-600 as a ground for prolongation of their life expectancy (in session "5.3 Advanced Fast Reactor Cladding Development I")
Results of old and program of new experiments on the small-sized fast multiplying systems with HEU / LEU fuel for receiving the benchmark data on criticality (in session "Poster Session 2")
Review of Transient Testing of Fast Reactor Fuels in the Transient REActor Test Facility (TREAT) (in session "6.10 Other issues of code development and application")
ROUZ CODE: CFD APPROACH FOR ASSESSMENT OF RADIATION SITUATION DURING ATMOSPHERE RADIOACTIVITY RELEASES WITHIN AN INDUSTRIAL SITE (in session "Poster Session 1")
Russian Companies’ involvement in CEFR RP (China) construction (in session "Poster Session 2")
Russian SFR Safety Requirements and Approaches and Their Correspondence to Generation-IV SFR Safety Design Criteria (Russia) (in session "Panel 1: Development and Standardization of Safety Design Criteria (SDC) and Guidelines (SDG) for Sodium Cooled Fast Reactors")
Safety Assurance for BN-1200 Power Unit During Accidents (in session "3.1 Safety Program")
Safety assurance of the new generation of the Russian fast liquid metal reactors (in session "3.5 General Safety Approach")
Safety criteria for future Indian SFRs (India) (in session "Panel 1: Development and Standardization of Safety Design Criteria (SDC) and Guidelines (SDG) for Sodium Cooled Fast Reactors")
Safety Upgradation of Fast Breeder Test Reactor (in session "2.1 Commissioning and Operating Experience of Fast Reactors I")
Scoping Analysis of STELLA-2 using MARS-LMR (in session "Poster Session 2")
SEALER: a small lead-cooled reactor for power production in the Canadian Arctic (in session "1.8 INNOVATIVE REACTOR DESIGNS")
SEISMIC SLOSHING EFFECTS IN LEAD-COOLED FAST REACTORS (in session "3.4 Sodium leak/fire and other safety issues")
SELECTION OF A LAYOUT FOR THE BN-800 REACTOR HYBRID CORE (in session "1.4 CORE AND DESIGN FEATURES - 1")
SELECTION OF CARRIER SALT FOR MOLTEN SALT FAST REACTOR (in session "Poster Session 1")
Sensitivity and Uncertainty Analysis in Best-Estimate modeling for PGSFR Under ULOF Transient (in session "6.5 Uncertainty Analysis and Tools")
Sensitivity studies of SFR unprotected transients with global neutronic feedback coefficients (in session "3.6 Safety Analysis")
Sensors of content of oxygen dissolved in heavy liquid metal coolants (in session "Poster Session 1")
SFR INHERENT SAFETY FEATURES AND CRITERIA ANALYSIS (in session "3.5 General Safety Approach")
SIBYLLA CODE: ASSESSMENT OF WATER BODIES CONTAMINATION AND DOSES RECEIVED BY POPULATION DUE TO RADIOACTIVITY DISCHARGES INTO THE HYDROSPHERE (in session "Poster Session 1")
SIMMER ANALYSES OF THE EBR-II SHUTDOWN HEAT REMOVAL TESTS (in session "Poster Session 2")
Simplification, the atout of LFR-AS-200 (in session "1.5 LFR DESIGN & DEVELOPMENT")
Simulating circulating-fuel fast reactors with the coupled TRACE-PARCS code (in session "6.6 Coupled Calculations")
Small fast reactors for arctic regions (in session "Panel 2: Small and Medium sized fast reactors")
SOCRAT-BN integral code for safety analysis of NPP with sodium cooled fast reactors: development and plant applications (in session "3.6 Safety Analysis")
Sodium compatibility of Recently-Developed Optimized Grade 92 and its Weldments for Advanced Fast Reactors (in session "Poster Session 1")
Sodium testing of fast reactor components (in session "6.8 Experimental Facilities")
Solution of the OECD/NEA SFR Benchmark with the Mexican neutron diffusion code AZNHEX (in session "6.4 Neutronics – 2")
Source Term Estimation for Radioactivity Release under Severe Accident Scenarios in Sodium cooled Fast Reactors (in session "3.2 Core Disruptive Accident")
Specific features of BN-1200 core in case of use of nitride or MOX fuel (in session "1.6 CORE AND DESIGN FEATURES - 2")
Stability Analysis of a Liquid Metal Cooled Fast Reactor (in session "6.4 Neutronics – 2")
Stability and bifurcation analysis of sodium boiling in a GEN IV SFR reactor core (in session "YGE Panel")
Stainless Steels Corrosion in Sodium Fast Reactor: Feedback from Risks during Maintenance Operations (SCC in Caustic Solution and Intergranular Corrosion by Acid Solution) (in session "5.6 Liquid Metal Technologies")
STATISTICAL INVESTIGATION OF RADIATION-INDUCED POROSITY IN BN FUEL CLADDINGS USING SCANNING ELECTRON MICROSCOPY (in session "Poster Session 1")
Status and perspectives of industrial supply chain for Fast Reactors (in session "7.3 Non Proliferation Aspects of Fast Reactors")
Status of ASTRID Nuclear Island Design and Future Trends (in session "1.2 SFR DESIGN & DEVELOPMENT - 2")
Status of Generation-IV Lead Fast Reactor Activities (in session "1.5 LFR DESIGN & DEVELOPMENT")
Status of severe accident studies at the end of the conceptual design: feedback on mitigation features (in session "3.2 Core Disruptive Accident")
Status of Sodium Cooled Fast Reactor Development Program in Korea (in session "Plenary Session 27 June")
Status of the French Fast Reactor Programme (in session "Plenary Session 26 June")
Steady State Modelling and Validation of Once Through Steam Generator (in session "6.1 CFD and 3D Modeling")
Strategies of maintaining appropriate technology of heavy liquid metal coolants in advanced nuclear power plants (in session "5.7 Chemistry Related Technology")
Strategy and R&D status of China Lead-based Reactor (in session "1.5 LFR DESIGN & DEVELOPMENT")
Structural Design and Evaluation of a Steam Generator in PGSFR (in session "Poster Session 1")
Study about the transient characteristics of the unprotected loss of flow accident in a metal fuel sodium cooled fast reactor based on the SAS4A code (in session "Poster Session 2")
Study for Accelerator-driven System in J-PARC/JAEA (in session "1.7 ADS AND OTHER REACTOR DESIGNS")
Study of isolation valve for Sodium Fast Reactor (in session "Poster Session 1")
Study of the austenitization process in a P91 steel (in session "Poster Session 1")
Study on Safety Design Concept for future Sodium-cooled Fast Reactors in Japan (in session "3.1 Safety Program")
Study on the limits of confinement leakage rates of pool-type sodium-cooled fast reactor (in session "Poster Session 1")
Study on the sensitivity analysis of the installed capacity and the high-level waste generation based on closed nuclear fuel cycle (in session "Poster Session 1")
Superphenix dismantling - Status and lessons learned (in session "2.3 Decommissioning of Fast Reactors and Radioactive Waste Management")
SVBR Project: status and possible development (in session "7.4 Fuel Cycle Analysis")
SVBR-100 as a possible option for developing countries, why? (in session "Panel 2: Small and Medium sized fast reactors")
Synergetic mechanism of high temperature radiation embrittlement of austenitic steels under long term neutron irradiation at high temperatures (in session "Poster Session 1")
System of Codes and Nuclear Data for Neutronics Calculations of Fast Reactors and Uncertainty Estimation (in session "6.5 Uncertainty Analysis and Tools")
System of coordinated calculation benchmarks for a fast reactor with sodium coolant in closed fuel cycle (in session "Poster Session 2")
TEM CHARACTERIZATION OF A SWELLING-RESISTANT AUSTENITIC STEEL IRRADIATED AT HIGH TEMPERATURE (>600°C) IN THE PHENIX FAST REACTOR (in session "5.8 Structural Materials")
Testing and Qualification of shielded flasks for handling sodium wetted large sized components of PFBR (in session "Poster Session 1")
Testing and Qualification of Trailing Cable system for Prototype Fast Breeder Reactor (in session "2.1 Commissioning and Operating Experience of Fast Reactors I")
Testing of electrochemical hydrogen meter in a sodium facility in Cadarache (in session "5.6 Liquid Metal Technologies")
The actinide oxides preparation by thermal denitration (in session "4.2 Reprocessing and Partitioning")
The ALLEGRO experimental Gas Cooled Fast Reactor Project (in session "1.7 ADS AND OTHER REACTOR DESIGNS")
The APOLLO3 scientific tool for SFR neutronic characterization: current achievements and perspectives (in session "6.3 Neutronics - 1")
The approaches to the radiation characteristics of structural elements of the core determination during operation and decommissioning for BN-type reactors (in session "Poster Session 2")
The ASTRID core at the end of the conceptual design phase (in session "1.6 CORE AND DESIGN FEATURES - 2")
The behavior features of fuel elements with nitride fuel - theory and experiment (in session "Poster Session 1")
THE CODE ROM FOR ASSESSMENT OF RADIATION SITUATION ON A REGIONAL SCALE DURING ATMOSPHERE RADIOACTIVITY RELEASES (in session "Poster Session 1")
The Commercial Potential of the Dual-Component Nuclear Power System (in session "Poster Session 1")
The Computer model for the economic assessment of NPP pilot demonstration energy complex with BREST-OD-300 reactor (REM Proryv Project) (in session "Poster Session 1")
The concept of 50-300 MWe modular-transportable nuclear power plant with sodium coolant and a gas turbine (in session "Poster Session 1")
The Conditioning and Chemistry Programme for MYRRHA (in session "5.7 Chemistry Related Technology")
The core of the LFR-AS-200: robustness for safety (in session "Poster Session 1")
The development of a computer code for predicting fast reactor oxide fuel element thermal and mechanical behavior (FIBER-Oxide) (in session "5.9 Large Component Technology II")
The DRESDYN project: A new facility for thermohydraulic studies with liquid sodium (in session "6.8 Experimental Facilities")
The Effect of Proton Irradiation on the Corrosion Behaviors of Ferritic/Martensitic Steel (in session "Poster Session 1")
The European Commission contribution to the development of safe and sustainable fast reactor systems (in session "Plenary Session 28 June")
The evolution of the primary system design of the MYRRHA facility (in session "1.7 ADS AND OTHER REACTOR DESIGNS")
The GIF Proliferation Resistance and Physical Protection (PR&PP) Evaluation Methodology: Status, Applications and Outlook (in session "7.3 Non Proliferation Aspects of Fast Reactors")
The IAEA Coordinated Research Project on Sodium Properties and Safe Operation of Experimental Facilities in Support of the Development and Deployment of Sodium-cooled Fast Reactors (NAPRO) (in session "5.2 Advanced Fast Reactor Fuel Development II")
The influence of porosity on thermal conductivity of low-density uranium oxide. (in session "Poster Session 1")
The lead-cooled fast reactor transition to equilibrium operating conditions (in session "Poster Session 1")
The method of calculating tritium content in various technological media of BN-type reactors (in session "Poster Session 1")
The optimization of core characteristics of fast molten salt reactor based on neutron-physical and thermal-hydraulic calculations and the analysis of fuel cycle closure options (in session "Poster Session 1")
The relative yields and half-lives of precursors of delayed neutrons in the fission 241Am by fast neutrons. (in session "Poster Session 2")
The Role of the IAEA in Fast Reactor Development and Knowledge Transfer (in session "YGE Workshop")
The Safety Design Criteria Development and Summary of Its Update for the Generation-IV SFR Systems (USA/Japan/GIF) (in session "Panel 1: Development and Standardization of Safety Design Criteria (SDC) and Guidelines (SDG) for Sodium Cooled Fast Reactors")
The Safety Design Guideline Development for Generation-IV SFR Systems (Japan/GIF) (in session "Panel 1: Development and Standardization of Safety Design Criteria (SDC) and Guidelines (SDG) for Sodium Cooled Fast Reactors")
The SAIGA experimental program to support the ASTRID Core Assessment in Severe Accident Conditions (in session "3.1 Safety Program")
the simulation of reactor physics for China Experimental Fast Reactor (in session "Poster Session 2")
The Status of Safety Research in the Field of Sodium-cooled Fast Reactors in Japan (in session "3.1 Safety Program")
THE STUDY OF THERMAL-HYDRAULIC PROCESSES IN THE STEAM GENERATOR OF THE BREST-OD-300 REACTOR FACILITY (in session "Poster Session 2")
The study of U-232 accumulation in reprocessed uranium for fast reactor fuel cycle (in session "Poster Session 1")
The U.S. Knowledge Preservation Program for Fast Flux Test Facility Data (in session "Poster Session 2")
The UO2– MeO2 (Me = Th, Pu, Zr) cathode crystalline deposits formation during the melts electrolysis. (in session "Poster Session 1")
The way of nitride fuel producing by high voltage electrodischarge compaction (in session "Poster Session 1")
Thermal and elastic properties of CexTh1-xO2 mixed oxides: a self-consistent thermodynamic approach (in session "Poster Session 1")
Thermal Annealing Effect on Recovery of Corrosion Properties of EP-450 Steel Irradiated IN BN-600 Reactor to High Damage Doses (in session "Poster Session 1")
Thermal conductivity of non-stoichiometric (Pu0.928Am0.072)O2-x (in session "Poster Session 1")
Thermal design of double helium gas gap conduction test facility (in session "Poster Session 2")
Thermal Hydraulic Investigation of EBR-II Instrumented Subassemblies during SHRT-17 and SHRT-45R Tests (in session "6.11 IAEA Benchmark on EBR-II Shutdown Heat Removal Tests")
Thermal hydraulic investigation of sodium fire and hydrogen production in top shield enclosure of an FBR following a core disruptive accident (in session "Poster Session 1")
Thermal Hydraulic Study of Steam Generator of PGSFR (in session "6.2 Thermal Hydraulics Calculations and Experiments")
Thermal-hydraulic experiments supporting the MYRRHA fuel assembly (in session "6.7 Experimental Thermal Hydraulics")
Thermal-hydraulics and Decay Heat Removal in GFR ALLEGRO (in session "3.6 Safety Analysis")
Thermodynamics and separation factor of lanthanides and actinides in system “liquid metal-molten salt” (in session "Poster Session 1")
Topical issues of training of specialists for fast nuclear power engineering and the closed nuclear fuel cycle (in session "8.1 Professional Development and Knowledge Management - I")
Towards a new approach for structural materials of Lead Fast Reactors (in session "Poster Session 1")
Tradeoff Study of Advanced Transmutation Fuels in Sodium-cooled Fast Reactors (in session "Poster Session 1")
TRANSMUTATION TRAJECTORY ANALYSIS IN THE MODELLING OF LFR FUEL CYCLE (in session "Poster Session 1")
U.S. Sodium Fast Reactor Codes and Methods: Current Capabilities and Path Forward (in session "6.10 Other issues of code development and application")
Uncertainty Analysis of Kinetic Parameters for Design, Operation and Safety Analysis of SFRs (in session "Poster Session 2")
Uncertainty Quantification of EBR-II Loss of Heat Sink Simulations with SAS4A/SASSYS-1 and DAKOTA (in session "Poster Session 2")
USDOE NEAMS Program and SHARP Multi-Physics ToolKit for High-Fidelity SFR Core Design and Analysis (in session "6.10 Other issues of code development and application")
Use of ion irradiations to help design of advanced austenitic steels (in session "Poster Session 1")
Using of computer code GEFEST800 at the initial stage of NPP operation with BN-800 (in session "Poster Session 2")
USSR and Russian fast reactor operation through the example of the BN600 reactor operating experience and peculiarities of the new generation BN800 reactor power unit commissioning (in session "2.1 Commissioning and Operating Experience of Fast Reactors I")
V&V STATUS OF CFD CODES APPLIED TO BN REACTORS (in session "Poster Session 2")
Validation of Advanced Metallic Fuel Models of SAS4A using TREAT M-Series Overpower Test Simulations (in session "6.10 Other issues of code development and application")
Validation of the evaluated fission product yields data from the fast neutron induced fission of 235U, 238U, 239Pu (in session "Poster Session 2")
Verification of the neutron diffusion code AZNHEX by means of the Serpent-DYN3D and Serpent-PARCS solution of the OECD/NEA SFR Benchmark (in session "6.3 Neutronics - 1")
VOIDING OF ELSY PRIMARY SYSTEM DURING STEAM GENERATOR LEAKAGE (in session "Poster Session 1")
Welcome Note by Conference General Chair (in session "Opening Session")
Welcome Note by Conference General Co-Chair (in session "Opening Session")
Welcome Note by Deputy Presidential Envoy in the Ural Federation District (in session "Opening Session")
Workshop Conclusion (in session "YGE Workshop")
X-RAY DIFFRACTION STRUCTURAL ANALYSIS OF STRUCTURAL AND FUEL MATERIALS FOR BN-600 REACTOR (in session "Poster Session 1")
“ASTRID safety design: Radiological confinement improvements compared to previous SFRs” (in session "3.1 Safety Program")
Include materials from selected contributions
