The Author Gareth Neighbour is a Senior Lecturer in Engineering at the University of Hull and is an accepted international authority on the phenomenon of irradiation creep in nuclear graphite. He has acted on various International Scientific Advisory Committees and was a member of the Organising Committee for the premier international carbon conference, Carbon 2006.

• Part A - Reactor Designs & Operation: The Management of Magnox Graphite Reactor Cores to Underwrite Continued Safe Operation; AGR Core Design, Operation and Safety Functions; An Overview of British Energy's Core Assessment Methodology; UK Regulatory Strategy for Management of Graphite; Ageing in Gas-Cooled, Graphite-Moderated Reactors; Part B - Advances in the Fundamental Knowledge of Graphite Behaviour under Irradiation: The Effects of Thermal Annealing on the Mechanical Properties of PGA Graphite; Development of a Model of Dimensional Change in AGR Graphites Irradiated in Inert Environments; Relating Measurements of Mechanical Properties of Nuclear Graphites to Reactor Conditions. A Review of the Effects of Temperature and Pressure; Flexural Strength of Graphite; Observations of Strain Localisation and Failure in Nuclear Graphite; Evaluation of the Brazilian Disc Test in Fracture Strength Measurements of Nuclear Graphite; Irradiation Damage in Graphite from First Principles; Development of a Code to Predict Fundamental Material and Fracture Properties of Nuclear Graphite; Modelling Graphite Ageing: Black Art or Forensic Science?; A Holistic, Structurally Based, Approach to Modelling Graphite Properties; X-ray Tomographic Observations Applied to Porosity; Models for the Thermal Properties of Oxidised Nuclear Graphite; Part C - Advances in the Understanding of Graphite-Component Behaviour and its Assessment: Crack Propagation Resistance and Damage Mechanisms in Nuclear Graphite; Graphite Irradiation Testing at the HFR Petten; The Development of Measurement Techniques for Mechanical Properties Applicable to Small Reactor Graphite Samples; Techniques for Measuring the Properties of Unirradiated and Irradiated AGR Graphite; Material Modelling of Nuclear Graphite under Irradiation Conditions; Predicting the Stresses and Deformations of Irradiated Graphite Moderator Bricks; Explicit Modelling of Graphite Core Components During a Reactor Thermal Transient; Part D - The Consequences of Graphite Deterioration: Whole-Core Behaviour: PBMR Graphite Core Structures Fitness for Purpose Management Strategy; Seismic Modelling of an AGR Nuclear Reactor Core; Measurement of AGR Graphite Fuel Brick Shrinkage and Channel Distortion; Investigation of Degradation of AGR Graphite Core Geometry Using a Whole Core Scale Model; The Feasibility of Introducing Core Stiffening Devices and Diverse Shutdown Systems at Hinkley Point B / Hunterston B and Hartlepool / Heysham 1; An Integrated Architecture for Graphite Core Data Analysis; Graphite Core Condition Monitoring Through Intelligent Analysis of Fuel Grab Load Trace Data; Application of Whole Core Modelling Methodology to Life Extension of AGR Reactor Graphite Cores; Statistical Models for Bore Cracking in AGR Graphite Cores; Development and Validation of Whole Core Modelling Methodology for AGR Graphite Cores; Method for Determining if Control Rods and Fuel Stringer can be Inserted / Removed from an AGR Core; Part E - Lessons Learned from Management of Ageing Plant: Experience of Ageing Processes in Nuclear Power Reactors; The Importance of Understanding of Basic Principles in the Management of Ageing Plant; An Overview and Summary of the Technical Papers in Parts B, C and D;