30 September - 04 October 2018 in Prague Czech Republic


Conference proceedings (full papers)

Also available all in a zip file or by Paper ID (TopFuel2018-A0xxx-fullpaper.pdf).

Advances in designs, materials and manufacturing

  • Out-of-Pile Verification of TRITON11™ BWR Fuel - A0078
  • Introduction of 300MW Fuel Assembly Spacer Grid Improvement in QinShan Phase-I NPP - A0094
  • Commercial introduction and experience with the advanced high iron cladding HiFi in Boiling Water Reactors (BWRs) - A0101
  • Irradiation Test Under Advanced PWR Conditions in the Halden Reactor and Post-Irradiation Examination of Fuel Rod laddings from Different Zirconium Alloys - A0128
  • Additive Manufacturing Paves the Way to Enhanced Utilization of Fuel Assemblies - A0180
  • Early Progress on Additive Manufacturing of Nuclear Fuel Materials - A0248


  • The Main Principles Of Irradiated Dispersion Type Fuel For Floating Power Unit Behavior - A0006
  • Study Of Modified Zirconium Alloys Claddings After Irradiation - A0007
  • Improvements on nuclear fuel manufacturing for reliable performance in the reactor - A0143
  • Development of Cobalt Adjuster Rod for Co-60 Medical Radioactive Sources Production in China Candu-6 Reactor - A0227


  • FUMAC: IAEA’s Coordinated Research Project on Fuel Modelling in Accident Conditions - A0198
  • IAEA FUMAC Benchmark on KIT Bundle Test CORA-15 - A0202
  • IAEA FUMAC Benchmark on Uncertainty and Sensitivity Analysis for Fuel Rod Code Simulation of the Halden LOCA Test FA-650.10 - A0206
  • IAEA FUMAC Benchmark on the Halden, Studisvik and QUENCH-L1 LOCA tests - A0222
  • IAEA FUMAC Benchmark of fuel performance codes based on LOCA separate-effects cladding tests - A0240


Keynote Session

  • Overview of Accident Tolerant Surface- Modified Fuel Cladding Development for LWRs - A0036
  • Path Towards Industrialisation Of Enhanced Accident Tolerant Fuel - A0141
  • Overview of Westinghouse Lead Accident Tolerant Fuel Program - A0151
  • Cr-coated cladding development at Framatome - A0152
  • The Research on Accident Tolerant Fuel in CGN - A0244


  • Implementation of Westinghouse ATF into PWRs: Fuel Cycle Economics and Operational Flexibility Improvements - A0001
  • The Effects of TRISO Particle Distribution on Thermal Behavior of Fully Ceramic Microencapsulated Fuel - A0021
  • Enhanced Radial Thermal Conductivity of UO2 Fuel Pellets with Molybdenum Microplates - A0060
  • New Insight on Volatile Fission Products (I and Cs) release from high burnup UO2 fuel under LOCA type conditions - A0068
  • Pre-oxidation effect of a zirconium-silicide sputtered surface on boiling performance and oxidation resistance - A0070
  • Severe Accident Evaluations for Conventional PWR Power Plant with SiC Composite Fuel Cladding - A0076
  • Modelling of an accident tolerant fuel design using FEMAXI6 - A0086
  • Demonstration of Engineered Multi-Layered SiC-SiC Cladding With Enhanced Accident Tolerance - A0105
  • Code qualification for traditional LWR fuel - A0114
  • Inspection capabilities and in-pile experience of innovative (EATF) materials at kernkraftwerk Gösgen-Däniken (KKG) - A0178


  • Progress on Japanese Development of Accident Tolerant FeCrAl-ODS Fuel Claddings for BWRs - A0011
  • Fuel Performance Assessment of Enhanced Accident Tolerant Fuel Using Iron-Based Alloys as Cladding - A0029
  • Machining induced fissures in relation microstrucure of uranium silicide fuel pellets - A0042
  • Overcoming sensitization in welds using FeCrAl alloys - A0052
  • Scratch and Fretting Wear Characteristics of Surface Modified Claddings for Accident-Tolerant Fuel - A0063
  • Progress in the Development of High Density Fuels for Enhanced Accident Tolerance - A0090
  • Behavior of Cr-coated M5™ claddings during and after high temperature steam oxidation from 800°C up to 1500°C (LOss-of-Coolant Accident & Design Extension Conditions) - A0100
  • Behavior of Chromium Coated M5 Claddings upon thermal ramp tests under internal pressure (LOss-of-Coolant Accident Conditions) - A0102
  • Status Update on Westinghouse SiC Composite Cladding Fuel Development - A0109
  • U3Si2 Developments in Falcon V1 at PS - A0112
  • Fatigue Behavior of Cold Spray-coated Accident Tolerant Cladding - A0126
  • Out of Pile Test with SiC Cladding Simulating LOCA Conditions - A0155
  • Characterization of thermal properties of SiCf/SiC composites for enhanced Accident Tolerant Fuel cladding - A0207


  • Peculiarities of stainless steels application as ATF in VVERs - A0054
  • Fuel Performance Analysis for enhanced characteristics of the Accident Tolerant Fuel under the Loss-of-Coolant Accident condition - A0075
  • Corrosion behaviour of FeCrAl-ODS steels in nitric acid solutions at several temperatures - A0080
  • Performance Evaluation of Accident Tolerant Fuel Claddings during Severe Accidents of BWRs - A0131
  • Westinghouse-Exelon EnCore® Fuel Lead Test Rod (LTR) Program including Coated Cladding Development and Advanced Pellets - A0145
  • Benefits of Framatome’s e-ATF evolutionary solution: Cr-coated cladding with Cr2O3-doped fuel - A0149
  • Inner surface protection of nuclear fuel cladding, towards a full-length treatment by an optimized DLI-MOCVD coating process - A0220
  • Experimental Behaviour of Chromium Based Coatings - A0233


  • Analysis of Irradiation Matrix for the Japanese FeCrAl-ODS Test Fuel Rods Irradiations at the Halden Reactor using FEMAXI-7 code - A0012
  • Steam oxidation of SiC at temperatures above 1600°C - A0026
  • Assessing the electrochemical behavior of ferritic FeCrAl alloys in high temperature water - A0053
  • Development Status of Microcell UO2 Pellet with Enhanced Thermal Conductivity for ATF - A0062
  • Improvement of Corrosion Resistant Coating for Silicon-carbide Fuel Cladding in Oxygenated High Temperature Water - A0072
  • Welding Technology R&D of Japanese Accident Tolerant FeCrAl-ODS fuel claddings for BWRs (2) - A0073
  • Effects of dissolved oxygen and ion irradiation on the corrosion of FeCrAl-ODS in high-temperature water simulating BWR operating conditions - A0083
  • Modeling and Assessment of EBR-Ii Fuel With the Us NRC’s Fast Fuel Performance Code - A0115
  • Experimental Investigation of Cold-Spray Chromium Coating - A0193
  • Nanocrystalline diamond protects Zr cladding surface against oxygen and hydrogen uptake: Nuclear fuel durability enhancement - A0213
  • Fission Gas Behavior of U3Si2 under LWRs Conditions: Experimental and Computational Study - A0214

Modelling, Analysis and Methods

Modelling I: Coupled codes and analysis

  • Development of fully coupled FRAPTRAN with MARS-KS code system for calculation of fuel behavior during LOCA - A0015
  • Towards a more detailed mesoscale fission product analysis in fuel performance codes: a coupling of the TRANSURANUS and MFPR-F codes - A0038
  • High Burnup Structure formation and growth and fission product release modelling: new simulations in the mechanistic code MFPR-F - A0084
  • Industry Use of CASL tools - A0096
  • Update on Westinghouse Benefits of EnCore® Fuel - A0163

Modelling II: Fuel rod codes

  • Establishment OF Centerline Temperatures in Irradiated Nuclear Fuels - A0031
  • Simulate5 Fuel PIN Model Description and Verification Against Enigma - A0043
  • Analysis of stress applied to fuel cladding with a burst opening under vibration - A0074
  • Expanded Assessment of FRAPCON and FAST for Power Ramp Cases with short hold times and Advanced UO2 fuel with various dopants - A0116
  • Improvements of PCMI Criterion for Anticipated Operational Occurrences - A0122
  • Application of the Transuranus Code to High Burn-Up LOCA Tests in View of 10 CFR 50.46c - A0162

Modelling III: Uncertainty Analysis

  • Progressive Bayesian Calibration of the BISON Fuel Performance Capability - A0023
  • Application of the Poolside Fuel Inspection Results in the Validation of Statistical Fuel Rod Performance Analysis - A0082
  • Analysis of Frapcon-4.0'S Uncertainties Predicting PCMI During Power Ramps - A0097
  • Sensitivity and Uncertainty Analysis of Fuel Performance Assessment of Chromia-Doped Fuel During Large-Break LOCA - A0196

Modelling IV: Fuel performance analysis (1)

  • Simulation of RIA transients on MOX fuel rods with ALCYONE fuel performance code - A0135
  • Fuel Performance Analysis of EnCore Fuel - A0156
  • OECD/NEA benchmark on pellet-clad mechanical interaction modelling with fuel performance codes: impact of number of radial pellet cracks and pellet-clad friction coefficient - A0219

Modelling V: Fuel performance analysis (2)

  • 3D Simulation of power ramps with ALCYONE including fuel thermochemistry and oxygen thermodiffusion - A0229
  • Modeling Out-of-Pile LOCA Tests on High Burnup Fuel Rods. Results of the fourth SCIP Modeling Workshop - A0237
  • Modeling fission gas release and bubble evolution in UO2 for engineering fuel rod analysis - A0241

Modelling VI: Mechanical and CFD Codes

  • Effect of the Characteristic Parameter on the Seismic Margin of Fuel Assembly - A0045
  • PWR fuel rod vibration simulation analysis for estimating grid-to-rod-fretting (GTRF) - A0079
  • Framatome’s State-Of-The-Art CFE Methodologies for Industrial Applications to Nuclear Reactors - A0210
  • A UK Regulatory Perspective on Computational Fluid Dynamics for Nuclear Safety Analysis - A0250

Modelling VII: Advanced Codes and Methods

  • Update on Framatome’S Advanced Solutions as a Service Support to Reactor Lifetime Extension - A0183
  • ARTEMIS / RELAP5 Integrated Transient Analysis Application to Non-LOCA Transients - A0192
  • Usage of Arcadia Code System for Neutronic and Thermal-Hydraulic Core Analyses to Support the Crud Risk Assessment of a Three-Loop Plant - A0194
  • Successful deployment of FRAMATOME advanced PWR Codes and Methods worldwide - A0235
  • Seismic analysis of a full 3D reactor core using multi-physics modeling methodology - A0247


  • Reactivity Initiated Accident Analysis Method Using Multi-Physics Coupled Code System Based on RAST-K V2.0 - A0027
  • An Approach to the Simulation of the Behaviour of Accident Tolerant Fuels - A0032
  • Study on the large deformation module in FRAPTRAN 2.0 - A0057
  • Development of experimental platform for analysis and imaging of fuel pellets heated at high temperature - A0065
  • Extension of the TRANSURANUS fuel performance code for uncertainty/sensitivity analyses and its application to design-based accidents (DBA) - A0067
  • Extended Validation of Engineering Models for Express-Method of Burnup Evaluation of WWER 1000 Fuel Elements - A0071
  • Optimization of Fast Fission Gas Release Model Parameters Using Machine Learning Accelerated Evolutionary Algorithms - A0117
  • Application of constrained Gibbs energy minimization to nuclear fuel thermochemistry - A0127
  • Residual stress/strain analysis in UO2 spent fuel by synchrotron micro-beam X-ray diffraction - A0176

Operation & Experience


  • Performance Capabilities Of The MIR.M1 Reactor For Demonstrating Technical Feasibility Of Enhanced Accident Tolerant Fuel - A0018
  • Experience and Opportunities of JSC “INM” Reactor and Experimental Facilities for Fuel Materials Testing - A0058
  • Experimental and simulation results of Expansion-Due-to-Compression tests with different strain biaxiality ratios on Zircaloy- 4 cladding for RIA situation - A0089
  • Non-Destructive Pressure Measurement Technique for Irradiated Nuclear Fuel Rods - A0095
  • Bow Evaluations to Support Fuel Assembly Design Improvements - A0140
  • Accelerated Irradiation Testing of Miniature Nuclear Fuel and Cladding Specimens - A0159
  • Causes of Increased Corrosion and Hydrogen Uptake of Zircaloy-2 Cladding at High Burnups – A Comparative Study of the Chemical Composition of a 3 Cycle and a 9 Cycle Cladding - A0172
  • Pre-Existing Surface Scratches Promoting Flaking of Shadow Corrosion on BWR Cladding - A0186
  • Nuclear Fuel and Materials Research, Experimental Capabilities, and Continuation of the Halden Reactor Project after the permanent shutdown of the Halden Reactor - A0197
  • Estimation of hydrogen in Zircaloy using multi frequency eddy current - A0211


  • GNF Fuel Reliability and Channel Performance: 2018 Update - A0111
  • N-Reactor Creep Behavior of Zirlo and Optimized Zirlo Cladding - A0123
  • Westinghouse 17X17 RFA Fuel Performance - A0136
  • Oxidation and hydrogen pickup properties of Zircaloy cladding upon deposition of platinum nanoparticles in boiling water reactor environment - A0138


  • Investigation of the Development of Fuel Assembly BOW in Ringhals 3 And 4 - A0185
  • Poolside Inspections at Loviisa NPP - A0205
  • End of life inspection of fuel that had experienced transient dryout in Forsmark 2 - A0217
  • ARGOS - Implementation of Framatome’s Universal Core Monitoring System on the European Market - A0245
  • Post Irradiation Examinations of GAIA Lead Fuel Assemblies - A0251


  • Ultrasonic System for Nuclear Fuel Geometrical Changes Evaluation - A0092
  • Development of digital X-ray radiography system for BWR control blade inspection - A0120
  • Synchrotron X-ray study on Determination of Zirconium Oxide Stoichiometry in Hydrogenated Water - A0201
  • Ring Tensile Test of Reference Zircaloy Cladding Tube as a Proof of Principle for Hotcell Setup - A0254

Transient Fuel Behaviour


  • Comparative high-temperature oxidation tests with Zircaloy-4 in various atmospheres - A0010
  • Effect of an oxide layer on the result of a ring compression test performed on a fuel cladding sample after a simulated LOCA transient - A0040
  • Thermal Resistance Effects of Oxide and Crud Layer to the Safety Analysis - A0046
  • New Insight on Volatile Fission Products (I and Cs) release from high burnup UO2 fuel under LOCA type conditions - A0064
  • Behaviors of High-burnup LWR Fuels with Improved Materials under Design-basis Accident Conditions - A0093
  • Modeling Axial Relocation of Fragmented Fuel during Loss of Coolant Conditions using the Bison Fuel Performance Code - A0113
  • Application of Transient Fuel Rod Performance Code Fraptran For SFP-LOCA Test - A0125
  • Simulation of Loss-of-Coolant Accidents in the CODEX integral test facility - A0133
  • Secondary hydriding experiments and simulation on Zr-1%Nb claddings - A0134


  • Mechanical behavior of as-fabricated Zircaloy-4 claddings under the simulated thermo-mechanical post-DNB conditions of a Reactivity Initiated Accident (RIA) - A0041
  • Anisothermal Behaviour of Unirradiated CWSR Zircaloy-4 Fuel Clads Under RIA Conditions - A0051
  • Evaluation of the consequences of fuel dispersion and interaction with coolant following a cladding failure induced by a RIA - A0081
  • The TREAT Experiment Legacy Supporting LWR Fuel Technology - A0168
  • Simulation of iron-chrome-aluminum alloy cladding under LOCA conditions using the BISON fuel performance code - A0170
  • Dynamics of hydride precipitation during LOCA quench process can significantly preserve cladding’s ductility - A0187
  • Consequences of leaking fuel rod failure during RIA transients - A0208
  • Updated RIA criteria in France - A0209
  • High Temperature Oxidation of Sponge-based E110 Alloy in Air - A0226
  • Research of high-temperature oxidation behavior of E110opt and E110М sponge based zirconium alloys - A0239


  • On Safety Objectives for Candu Fuel in Design Extension Conditions - A0069
  • Feasibility Assessment for Developing an Integral LOCA Testing Capability at the Transient Research Test (Treat) Reactor - A0108
  • Speciation and Release Kinetics of the Fission Products Mo, Cs, Ba And I from Nuclear Fuels in Severe Accident Conditions - A0139

Used fuel: storage, transportation and re-use


  • Post-Irradiation Examinations of High Burnup PWR Fuel Rods - Initial Results - A0033
  • High Burnup Spent Fuel Dry Storage Research Project - A0177
  • Spent Fuel Preparation Before Disposal - A0203


  • A Study to Evaluate the Handling Integrity of Spent Nuclear Fuel for Dry Storage in Korea - A0009
  • Impact of Fuel-Cladding Bonding on the Response of High Burnup Spent Fuel Subjected to Transportation and Handling Accidents - A0118
  • Mechanical Integrity of Used Nuclear Fuel: From Experimental to Numerical Studies - A0129
  • Transport of Irradiated Nuclear Fuel Between Reactor Sites for Further Use - A0218
  • ENUSA Integral Solution to for Intergranular Stress Corrosion Cracking on Early 17x17 PWR Designs - A0224
  • Sipping of Fuel Assemblies - A0225
  • Advanced Vacuum Sipping for Spent Fuel Classification - A0230
  • Oxidation of UO2 in dry and wet atmospheres - A0238
  • Handling, Transport and Program for Post-Irradiation Examination of Special Fuel Rods - A0249


  • A CFD analysis of thermal behavior in passive heat removal system of dry storage cask under different conditions - A0003
  • Development of regulatory requirements for safety information for spent nuclear fuel characteristics evaluation in Korea - A0047
  • Ductility of pre-hydrided Zircaloy-4 cladding after creep deformation - A0049
  • STAR-CCM+ simulation of a spent fuel dry cask external cooling by natural convection - A0056
  • The effect of final heat treatment at fabrication on the terminal solid solubility of hydrogen in Zry-4 - A0077
  • Thermal Analysis Oo QM400 Dry Storage Module Without Thermal Baffles - A0091
  • Quivers for Damaged Fuel Rods – Disposal in Castor® V Casks - A0173
  • Spent Fuel Dry Storage Cast Thermal Modeling Round Robin - A0175
  • Temperature calculations in spent nuclear fuel cask using COBRA-SFS - A0188
  • Development of Smart Material-Based Structural Integrity Monitoring Sensors for Detecting the Fracture Sign in Dry Storage Canisters - A0215
  • Thermal performance evaluation of cylindrical modular type dry storage system for PWR spent nuclear fuel using CFD - A0221

Contact the ENS / TOPFUEL 2018 Secretariat if you have any question:

Tel: + 32 2 505 30 54
Fax: +32 2 502 39 02