Postirradiation examination (PIE) of graphite samples trepanned from the UK reactor cores has been carried out for more than 50 years. Due to the nature of the material, there are sample size and geometry restrictions and no standard test methods to cover measurements on this material. Nevertheless, these measurements are used to support the continued operation of the UK reactor cores, and hence a large program of trials is carried out to provide confidence that each method is accurate and reproducible. These trials typically involve a study of size effects using virgin graphite and simulant materials for the irradiated and oxidized graphite, but a corresponding study with irradiated samples is usually not possible. This paper combines the work of two UK studies to investigate the size effect of the PIE test methods on irradiated graphite on the basis of the characterization of graphite used in the advanced gas-cooled reactors (AGRs) and Magnox reactors. The AGR study focused on the static Young's modulus, three-point bend strength of unnotched and notched beams, and the work of fracture. The Magnox study focused on the coefficient of thermal expansion, diametral compression, and flexural strength. The two studies used large irradiated graphite samples from “installed sets” (i.e., precharacterized graphite samples installed in the reactor cores before the start of operation for monitoring purposes). Large Magnox samples that were trepanned from the reactor core after shutdown were also tested. The purpose of these investigations was to relate the graphite measurements normally undertaken on small trepanned samples to property values obtained using standard test methods on irradiated material. The sample selection was such that it covered as wide a range of dose and radiolytic weight loss as possible. This paper outlines the methodology, results, and conclusions for each of the studies and provides some guidelines for similar studies on new graphites.

Keyway root cracking and radiolytic weight loss in graphite moderated nuclear reactors may limit future successful operation of the civil nuclear fleet. A novel means of initiating fracture using internally generated stresses has been developed, which uses bromine intercalation to induce fracture in small graphite samples. Successful crack initiation and propagation has been achieved in 1:10 scale reactor components. In situ X-ray Computed Tomography during bromination has enabled observations of failure in real time, allowing the role of microstructure during fracture and weight loss to be determined. Crack paths predicted by models have been assessed.

The assessment of the post-cracking behaviour of the reactor core components requires knowledge of the fracture properties, such as the fracture toughness, KIC, and work of fracture, γf, of irradiated graphite. The measurement of these properties is a proven technique for linearly elastic materials but application on small nuclear graphite specimens demands further consideration. The purpose of this work is to develop this technique for use on irradiated graphite specimens whose size and geometry are restricted by the reactor core trepanning or Materials Test Reactor experiments. This paper describes the theoretical basis of the method, the work undertaken to prove the measurement technique and demonstrate its applicability to small irradiated and oxidised graphite specimens. Finally, the paper presents work of fracture values of irradiated graphite trepanned from AGR core bricks; for the first time, the relationship between work of fracture and other graphite properties, such as density, modulus and strength is experimentally demonstrated.

SummaryThe complex dynamics of a quarter‐scale model of a graphite nuclear reactor core, representative of the second generation of British advanced gas‐cooled nuclear reactors, is investigated numerically and experimentally. Advanced gas‐cooled nuclear reactor cores are polygonal, multilayer, arrays of graphite bricks, with each brick allowed to rock by design relative to each other in accordance with the boundary conditions. A 35 000 DOF, nonlinear finite element model of the core created by Atkins Nuclear, was analysed on a high performance computing facility at the University of Bristol, and a corresponding 8 t physical model, equipped with 3200 data acquisition channels, was built and tested on the University of Bristol 6‐DOF shaking table. In this paper, the two models are subjected to a series of (1) synthetic earthquake and (2) idealised harmonic input motions. The experimental data are used to compare and verify the two models and explore the dynamics of the core. A kinematic model of the response is also developed based solely on geometric constraints. The results are presented in the form of response maps and graphs. Important conclusions are drawn as to the dynamics and earthquake response of such systems, which inform numerical model validation. It is found that contrary to the case of a small number of rocking blocks that exhibit highly complex response patterns, the behaviour of the model at hand is both smooth and repeatable. An analogy between the response of the core and that of dense granular matter exhibiting particle interlocking and dilatancy is highlighted.

The ageing issues of the Advanced Gas Cooled Reactor (AGR) cores need addressing to maintain their safe and reliable operation, hence the requirement for the computer models of the cores used for the seismic resilience assessments to be conservative and to represent larger percentages of damaged graphite components. The current models have undergone limited experimental validation for high levels of degradation, so there is a need to validate those numerical models and also to enhance the understanding of core dynamics by physical modelling and testing. This paper outlines the feasibility study of a quarter scale model rig of an AGR core developed by the University of Bristol. The damage scenarios to be considered in demonstrating the core seismic tolerability were defined. The principles of scale modelling were put under scrutiny in parallel with several practical aspects of material selection and component design and manufacturing. Several variants of physical models of different size and shape were proposed and their merits with respect to their feasibility and outcomes were discussed. Aspects of instrumentation design are presented together with relevant measurement results. The rig is a viable experimental tool whose outputs can be employed directly in computer model validation. (C) 2017 Elsevier B.V. All rights reserved.

In this paper, crack propagation in Advanced Gas-cooled Reactor (AGR) graphite bricks with ageing properties is studied using the eXtended Finite Element Method (X-FEM). A parametric study for crack propagation, including the influence of different initial crack shapes and propagation criteria, is conducted. The results obtained in the benchmark study show that the crack paths from X-FEM are similar to the experimental ones. The accuracy of the strain energy release rate computation in a heterogeneous material is also evaluated using a finite difference approach. Planar and non-planar 3D crack growth simulations are presented to demonstrate the robustness and the versatility of the method utilized. Finally, this work contributes to the better understanding of crack propagation behaviour in AGR graphite bricks and so contributes to the extension of the AGR plants' lifetimes in the UK by reducing uncertainties.

In this paper, we describe a model of irradiated graphite properties that is based on an understanding of the interconnection of the structure. Simplified conceptual models have enabled us to explain many of the correlations that exist between irradiated graphite material properties and derive mathematical formulations that appear to be of widespread applicability. The principles of the model are illustrated with reference to the irradiation ‘structure term’ of Young’s modulus. It is shown that the currently used definition may be considered to be a combination of three separate processes – pore closure driven densification, increased structural interconnectivity and (latterly) pore generation. It is the structural interconnectivity component that is most closely linked to the changes in other properties such as dimensional change rate and coefficient of thermal expansion, and this relationship is used to direct the mathematical formulations.

In this paper, the authors describe the progress towards development of a holistic model of irradiated graphite properties, which is based on an understanding of the interconnection of the structure. The approach has enabled us to explain many of the correlations that exist between irradiated graphite material properties and to make predictions of the effect of radiolytic oxidation beyond the limits of the database that currently exists. The principles are illustrated with reference to the irradiation "structure term" of Young's modulus. It is shown that the currently used definition is actually a combination of three separate processes - pore closure driven densification, increased structural interconnectivity and (latterly) pore generation. It is the structural interconnectivity component that is most closely linked to the changes in other properties such as dimensional change rate and coefficient of thermal expansion.

Advanced Gas Cooled Reactor (AGR) cores are multi-layered arrays of graphite components whose geometry and mechanical properties change under prolonged exposure to neutron irradiation. The presence of cracked components in the arrays later in their operational life may cause disruption of core geometry with implications for fuel cooling and control rod insertion in the event of a severe, but infrequent, seismic event. These ageing issues need addressing in both the computational and the physical models employed in the seismic resilience assessments. This paper presents a physical model with quarter-sized components of an array representative of those in AGR cores. The model was developed by the University of Bristol to provide experimental validation to computational tools which model high levels of core degradation. This paper outlines the principles of model design and the relevant aspects of rig development. The rig is tested on an earthquake simulator with the purpose to explore the mechanical interactions inside the array and to output acceleration and displacement data at selected locations. Relevant experimental outputs are presented showing dynamic responses of the array columns and top layer response maps. Overall, the model rig is capable of providing experimental evidence for the computational modelling methods, and so makes significant contributions to reducing uncertainties in these methods.

The graphite components of an Advanced Gas Cooled Reactor (AGR) are subject to ageing processes that lead to changes of geometry and mechanical properties. Such changes need addressing in the safety case strategy of the operator, hence the necessity for both the numerical and the physical reactor models to be conservative and to represent high levels of graphite component degradation. This paper presents a quarter scale physical model of a multi-layer array representative of those in AGR cores. The rig was developed by the University of Bristol to support the seismic capabilities of the existing computer core models. The physical model can embed high percentages of doubly cracked bricks in various pattern distributions. Intact and cracked array configurations were subjected to seismic testing on an earthquake simulator. Relevant results of component displacement in the array are presented together with separation data between doubly cracked brick halves that provide evidence of key-keyway disengagement. The outlined experimental output demonstrates that the model rig is capable of providing an enhanced understanding of the mechanical interactions that take place inside the array with relevance for both the nuclear plant operator and the computer modellers.