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
Posters
- 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
AEA FUMAC Project
- 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
EATF
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
Monday
- 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
Tuesday
- 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
Wednesday
- 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
Posters
- 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
Posters
- 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
Monday
- 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
Tuesday
- 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
Wednesday
- 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
Posters
- 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
Monday
- 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
Tuesday
- 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
Posters
- 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
Tuesday
- 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
Wednesday
- 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
Posters
- 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
