Session: MF-29-01 Mechanical Properties of Nuclear Graphite and their Implementation in Codes and Standards (Joint with CS-15)

Paper Number: 106362

106362 - Graphite Material Model Calibration for Conceptual Design and Core Configuration Studies 

Nuclear graphite is a candidate moderator material for many high temperature advanced reactor concepts. However, the material properties of graphite used in this application evolves during service, principally as a function of fast neutron dose, irradiation temperature and, if relevant, radiolytic oxidation (weight loss). One of the most significant threats to the moderator structure continuing to meet its design requirements is graphite dimensional change which causes significant deformation in the components through life. 
In general, as graphite is irradiated, the dimensional change will initially cause it to shrink until it reaches turnaround, further irradiation will then cause it to begin to re-expand. Furthermore, irradiation causes evolution of graphite’s thermal properties, where both the thermal conductivity and thermal expansion coefficient evolve significantly. These can result in substantial thermal strains during normal operation, but particularly during transients.
As the cumulative dose received by graphite varies within each brick and through the core, the rates of property evolution vary temporally. These result in differential dimensional change and thermal strains contributing to the generation of internal stresses. In a brittle material such as graphite, these stresses would quickly reach critical values were they not relieved by creep processes. Irradiation creep in graphite appears to exhibit both recoverable and irreversible process and must be accounted for in component analyses
Finite element analysis is typically required to predict the properties, stresses and distortions of graphite components through life. In turn, this requires a material model and constitutive law which can describe the evolution in properties and resulting anisotropic material behaviour.
This paper describes such a model which has been implemented using a user material (UMAT) subroutine in Simulia’s Abaqus/Standard. The material model used in this study is based on the published work contributing to the EDF Energy Integrated Methodology (EIM) which is used to predict the behaviour of Gilsocarbon components within the Advanced Gas-cooled Reactors (AGRs). This framework represents a series of engineering descriptions of graphite material property evolution linked by a common physically consistent framework. It aims to predict the evolution of graphite material properties in response to long term reactor service as a function of irradiation dose, temperature, oxidation and stress. The UMAT then assembles this information in tensor form to provide a three-dimensional visco-elastic constitutive law suitable for use in finite element calculations. 
A number of candidate graphite grades are available for new reactor concepts. The property evolution data for these is sometimes sparse and variable. However, it has been shown possible to calibrate the proposed UMAT to wide range of these graphite grades. Moreover, the integrated nature of the model, in which all property changes are hypothesised to be driven by common microstructural evolution, allow all properties to be calibrated simultaneously reducing overall uncertainty.
Whilst this approach is used in the UK to support continued operation of the AGRs, it is also judged appropriate to inform the pre-conceptual design stage of new reactor core designs. Being able to rapidly assess different core configuration options and their compliance with the guidance in ASME Section III Division 5 is a valuable tool in the early stages of design, particularly when data is sparse. This paper proposes an approach to this problem together with a robust approach to sensitivity studies aimed at ensuring that a particular core design is tolerant to material evolution uncertainty. This may enable design work to proceed in parallel to the collection of detailed material evolution information significantly derisking and reducing the time to complete the system design and assurance process.

Presenting Author: Amanda Young Frazer-Nash Consultancy

Presenting Author Biography: Amanda is a consultant at Frazer-Nash Consultancy working in the physical systems modelling team. She has worked on nuclear Graphite for 5 years modelling the UK AGRs and new molten salt reactor concept designs.

Authors:

Amanda Young Frazer-Nash Consultancy
Graeme Horne Frazer-Nash Consultancy
Richard Gray Frazer-Nash Consultancy
Daniel Kent Frazer-Nash Consultancy
Mark Joyce Frazer-Nash Consultancy