Nassia Tzelepi

National Nuclear Laboratory, UK | NNL · Reactor Operations Support

Coated fuel particles, most commonly tri-structural isotropic (TRISO), are intended for application in several designs of advanced nuclear reactors. A complete understanding of the residual stresses and local properties of these particles through their entire lifecycle is required to inform fuel element manufacturing, reactor operation, accident sc...

Thermal oxidation of nuclear graphite components is highly undesirable because it can cause structural and property degradation that negatively affect a reactor's intended operation. In this work, the body of knowledge of nuclear graphite oxidation is highlighted, including when O2, H2O, and/or CO2 are the oxidant. Oxidation conditions relevant to...

The Young's modulus, Poisson's ratio and flexural strength of 218 beams of radiolytically oxidised nuclear graphite have been determined by combining electronic speckle pattern interferometry, ESPI, and inverse finite element modelling techniques. The graphite was extracted from three operational advanced gas cooled reactor stations where it had be...

Graphite is the moderator material of several Generation IV nuclear reactor concepts, as well as the British Advanced Gas-cooled Reactors (AGR). Porosity can heavily influence the material properties, mechanical irradiation response, and neutron induced shrinkage or swelling of nuclear-grade graphite. Due to the sub-micron size of several types of...

Herein, the first study is presented using 4D synchrotron X-ray microtomography to capture all stages of crack development in neutron irradiated and radiolytically oxidised nuclear graphite. Employing a novel loading setup, specimens of Gilsocarbon graphite, both unirradiated and irradiated at 301 °C to 19.7 x 1020 neutrons/cm2 (~2.6 displacements/...

As the Advanced Gas-cooled Reactors approach the end of generation, research and characterisation are required to support the decommissioning strategy. Radiological data for AGR graphite are negligible and the radiological inventory of the AGR core and other graphite components rely on activation modelling. This is the first study of C-14 activity...

The mechanism of irradiation creep and its effect on the internal stresses of reactor core graphite components are not well understood. Historical studies on Materials Test Reactor creep specimens have shown that thermal annealing can lead to recovery of creep strain. By selecting appropriate sets of trepanned graphite specimens for the AGR cores,...

Nuclear power generation in the United Kingdom is based principally on graphite-moderated gas-cooled reactors. One of the principal long-lived radionuclides produced during graphite irradiation is radiocarbon (C-14). Graphite from some Magnox reactors is known to contain carbonaceous surface deposits, which have been shown to contain significantly...

In this study, the oxidation characteristics of nuclear grade graphites IG-110, IG-430, Type A(MA), Type B (MB), PGA, ATR-2E, G347A and G458A in air environments at temperatures ranging from 700 to 1600 °C were investigated. The oxidation rates and activation energies were determined based on mass loss measurements in a series of oxidation tests. S...

The United Kingdom Magnox reactors have all ceased generation and await decommissioning. This fleet of reactors contains 50,000–60,000 tonnes of irradiated graphite categorised as Intermediate Level Waste, which will require long-term storage/disposal either in a proposed Geological Disposal Facility (England and Wales) or in an on-site near surfac...

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 specimen...

Graphite is used in the cores of gas-cooled reactors as both the neutron moderator and a structural material, and traditional and novel graphite materials are being studied worldwide for applications in Generation IV reactors. In this study, the oxidation characteristics of petroleum-based IG-110 and pitch-based IG-430 graphite pellets in helium an...

Property measurement of graphite extracted from operational nuclear reactors is vital for forward prediction of the graphite's properties and integrity as part of continued operation and plant life extension assessments. Limited core monitoring material is available, so robust and efficient measurement techniques are required for samples with const...

The mechanisms behind the changes of the polycrystalline graphite properties due to temperature variation and irradiation are still not fully understood. Computational models are widely used to study and predict the changes, but there are no commonly accepted validation methodologies to build confidence in these models. In the current study, a nove...

Microcracks in neutron-irradiated nuclear grade graphite have been examined in detail for the first time using a combination of transmission electron microscopy (TEM), electron energy loss spectroscopy (EELS), energy dispersive X-ray (EDX), and energy filtered TEM (EFTEM). Filler particles from both unirradiated Pile Grade A (PGA) and three irradia...

A micromechanistic model has been applied to understand the relation between the properties of individual graphite crystals and the properties of polycrystalline graphite. The proposed model accounts for irradiation-induced creep as well as microcracking in both intragranular (parallel to basal planes) and intergranular positions. The two types of...

One of the key issues that must be addressed when a sampling and measurement program underwrites the performance of full-size graphite components is the understanding of the influence of material type and specimen size on the measured property. Because of sampling constraints, specimen sizes (geometry and/or volume) may not be compliant with measur...

An original model is proposed that predicts the response of a polycrystalline graphite agglomerate to temperature change and irradiation. The model explicitly considers the graphite crystal anisotropy and the development of cracks along the interfaces between differently oriented grains. Using only graphite single crystal data, the model yields pre...

Gas-cooled, graphite-moderated nuclear reactors suffer ageing and degradation to the graphite during service posing a threat to the functionality of the core, and potentially, the safe operation of the reactor. Thus, the importance of modelling and measuring reactor core graphite properties and performance increases especially as continued use beyo...

Gas-cooled, graphite-moderated nuclear reactors suffer ageing and degradation to the graphite during service posing a threat to the functionality of the core, and potentially, the safe operation of the reactor. Thus, the importance of modelling and measuring reactor core graphite properties and performance increases especially as continued use beyo...

The effect of thermal oxidation on PGA graphite samples was investigated by X-ray tomography. This technique allowed an investigation of the evolution of microstructure and local density to be carried out as a function of weight loss. The results show a localised oxidation process which progressively reveals the large coke particles with increasing...