Structural Materials for Innovative Nuclear Systems (Smins) Workshop Proceedings - Karisruhe, Germany 4-6, June 2007 (Nuclear Science)

Foreword
 CONTENTS
 Executive summary
 OVERVIEW
 EURATOM RDDD in Innovative Reactor Systems and Fuel Cycles
 SESSION I KEYNOTE LECTURES
 Innovative Reactor Systems and Requirements for Structural Materials
 Ageing Management and Knowledge Base for Safe Long-termOperation of Japanese Light Water Reactors
 Modelling, Microstructures, Radiation Damage and Understandingof Irradiation Data
 SESSION II GAS-COOLED SYSTEMS
 Material Research for VHTR Design Codes
 Evaluation of Nickel-based Materials for VHTR Heat Exchanger An Uncertainty Assessment Methodology for Materials Behaviourin Advanced Fast ReactorsSESSION III LIQUID-METAL-COOLED SYSTEMS
 High-Cr ODS Steels R&D for High Burn-up Fuel Cladding
 Influence of Normalising Temperature on MX PrecipitationBehaviour in High-chromium Steel
 Development of ODS Ferritic-martensitic Steels for InnovativeNuclear Fast Breeder Reactors
 Void Swelling and Defect Processes in Ti-modified SteelsUsing Accelerator Irradiation
 First Results on T91 Claddings with and Without Modified FeCrAlY CoatingsExposed in PbBi Under Varying Conditions SESSION IV FUSION, HIGH-POWER ACCELERATOR TARGETSAND OTHER FISSION REACTOR SYSTEMSAssessing SCC and IASCC of Austenitic Alloys for Applicationto the SCWR Concept
 Oxidation and Stress Corrosion Cracking of Stainless Steels in SCWRs
 Microstructural Influence on the Corrosion Behaviour ofStructural Materials for Nuclear Power Systems
 Influence of Stacking Fault Energy (SFE) on the Deformation Modeof Stainless Steels
 Evolution of Structural Materials in Nuclear Reactors
 SESSION V FUNDAMENTAL RESEARCH Qualification of Structural Materials for Reactor Systems: Synergies inMaterials for Fusion/Fission Reactors and Advanced Fission ReactorsUnderstanding Radiation Damage in FECR Alloys
 A New Computational Method for Studies of 3-D Dislocation-precipitate Interactions in Reactor Steels
 On the Theoretical Development of New Creep Resistant Alloys andTheir Empirical Validation
 Modelling of Nuclear-grade Graphite Using X-ray Tomography Data
 Basic Experiments for Understanding Microstructure Evolution UnderIrradiation: TEM and KMC Characterisation of Irradiated Pure Iron Combined TEM, PAS and SANS Investigation of Neutron-irradiatedPure IronSynchrotron Light Techniques for the Investigation of AdvancedNuclear Reactor Structural Materials
 First-principles Modelling of Advanced Nuclear Fuels: Problemsand Prospects
 POSTER SESSIONS
 High Temperature Corrosion Properties of Alloy 230 in theEnvironment of Gas-cooled Reactors
 Materials Modelling --
A Possible Design Tool for AdvancedNuclear Applications
 Characterisation of Oxidation and Oxide Layers of Nickel-baseSuperalloys at High Temperature