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The integrity of graphite components is critical for their fitness for purpose. Since graphite is a quasi-brittle material the dominant mechanism for loss of integrity is cracking, most specifically the interaction and coalescence of micro-cracks into a critically sized flaw. Including mechanistic understanding at the length scale of local features (meso-scale) can help capture the dependence on microstructure of graphites macro-scale integrity. Lattice models are a branch of discrete, local approach models consisting of nodes connected into a lattice through discrete elements, including springs and beams. Element properties allow the construction of a micro-mechanically based material constitutive law, which will generate the expected non-linear quasi-brittle response. This research focuses on the development of the Site-Bond lattice model, which is constructed from a regular tessellation of truncated octahedral cells. The aim of this research is to explore the Site-Bond model with a view to increasing understanding of deformation and fracture behaviour of nuclear graphite at the length scale of micro-structural features. The methodology (choice of element, appropriate meso length-scale, calibration of bond stiffness constants, microstructure mapping) and results, which include studies on fracture energy and damage evolution, are presented through a portfolio of published work.

In addition to its role as moderator within British nuclear reactors, polycrystalline graphite is also a major structural component of the core, enabling access for control rods, coolant gas and fuel. Aging processes, primarily fast neutron irradiation and radiolytic oxidation lead to distortion of the graphite components and property changes which ultimately reduce the material's effectiveness and can lead to component failure.Despite much research into the material, graphite behaviour under irradiation conditions is not fully understood and has resulted in an overestimation of the extent of component failures in Magnox reactors, and a subsequent underestimation of component failures in the following generation Advanced Gas-cooled Reactors (AGRs). A greater understanding of the material is therefore required in order to make more informed evaluations as part of on-going safety cases. Young's modulus is one property which varies as a complex function of radiolytic oxidation and fast neutron irradiation dose; this work investigates investigate the Young's modulus behaviour of nuclear grade graphites through property measurement and microstructural characterisation. Physical properties are dependent on microstructure, which is in turn a result of the manufacturing processes and raw materials used in its fabrication. Because of this, this thesis begins with a microstructural study of AGR graphite artefacts from varying points during the manufacturing process and post-irradiation, utilising X-ray diffraction to observe changes in crystallinity, microscopy to directly observe the microstructure and pycnometry to gauge porosity variations. Increases in crystallinity towards graphitisation are seen, with a subsequent decrease after irradiation; and significant changes are observed from inspection of optical and scanning electron micrographs. Young's modulus property data are obtained using a combination of static and dynamic techniques to accumulate data from a variety of techniques. An experiment designed to track changes to the speed of sound under compressive load was carried out on Magnox and AGR graphite, showing different behaviour between the grades, and variation with irradiation.A final series of tests combine compressive testing with in-situ microscopy to try and better understand the reasons behind this varied in behaviour and relate microstructural changes to graphite behaviour under compressive loading.

Once current nuclear reactor operation ceases in the U.K. there will be an estimated 99,000 tonnes of irradiated nuclear graphite waste which may account for up to 30% of any future UK geological ILW disposal facility [1]. In order to make informed decisions of how best to dispose of such large volumes of irradiated graphite (I-graphite) within the UK nuclear programme, it is necessary to understand the nature and migration of isotopes present within the graphite structure. I-graphite has a combination of short and long term isotopes such as 14C, 3H and 36Cl, how these behave prior to and during disposal is of great concern to scientific and regulatory bodies when evaluating present decommissioning options. Various proposed decontamination and immobilisation treatments within the EU Euroatom FP7 CARBOWASTE program have been explored [2, 3]. Experiments have been carried out on UK irradiated British Experimental Pile Zero and Magnox Wylfa graphite in order to remove isotopic content prior to long term storage and to assess the long term leachability of isotopes. Several leaching conditions have been developed to remove 3H and 14C from the irradiated graphite using oxidising and various acidic environments and show mobility of 3H and 14C. Leaching analysis obtained from this research and differences observed under varying leaching conditions will be discussed. Thermal analysis of the samples pre and post leaching has been performed to quantify and validate the 14C and 3H inventory. Finally the research objectives address differences in leachability in the graphite to that of structural and operational variation of the material. Techniques including X-ray Tomography, Scanning Electron Microscopy, Autoradiography and Raman spectroscopy have been examined and show a significant differences in microstructure, isotope distribution and location depending of irradiation history, temperature and graphite source. Ultimately the suitability of the developed chemical treatments will be discussed as whether chemical treatment is a viable option prior to irradiated graphite long term disposal.

Thesis (PhD)--Stellenbosch University, 2015. / ENGLISH ABSTRACT: This thesis presents a failure evaluation methodology for nuclear graphite components used in high-temperature gas reactors. The failure methodology is aimed at predicting the failure of real parts based on the mechanical testing results of material specimens. The method is a statistical failure methodology for calculating the probability of failure of graphite components, and has been developed and implemented numerically in conjunction with a finite element analysis. Therefore, it can be used on any geometry and load configuration that can be modelled using finite element analysis. The methodology is demonstrated by mechanical testing of NBG-18 nuclear grade graphite specimens with varying geometries under various loading conditions. Some tests were developed as an extension of the material characterisation, specifically engineered to assess the effect of stress concentrations on the failure of NBG-18 components. Two relevant statistical distribution functions, a normal distribution and a twoparameter Weibull distribution are fitted to the experimental material strength data for NBG-18 nuclear graphite. Furthermore, the experimental data are normalised for ease of comparison and combined into one representative data set. The combined data set passes a goodness-of-fit test which implies the mechanism of failure is similar between data sets. A three-parameter Weibull fit to the tensile strength data is only used in order to predict the failure of independent problems according to the statistical failure methodology. The analysis of the experimental results and a discussion of the accuracy of the failure prediction methodology are presented. The data is analysed at median failure load prediction as well as at lower probabilities of failure. This methodology is based on the existence of a “link volume”, a volume of material in a weakest link methodology defined in terms of two grouping criteria. The process for approximating the optimal size of a link volume required for the weakest link failure calculation in NBG-18 nuclear graphite is demonstrated. The influence of the two grouping criteria on the failure load prediction is evaluated. A detailed evaluation of the failure prediction for each test case is performed for all proposed link volumes. From the investigation, recommended link volumes for NBG-18 are given for an accurate or conservative failure prediction. Furthermore, failure prediction of a full-sized specimen test is designed to simulate the failure condition which would be encountered if the reactor is evaluated independently. Three specimens are tested and evaluated against the predicted failure. Failure of the full-size component is predicted realistically but conservatively. The predicted failure using link volume values for the test rig design is 20% conservative. The methodology is based on the Weibull weakest link method which is inherently volume dependent. Consequently, the conservatism shows that the methodology has volume dependency as experienced in the classic Weibull theory but to a far lesser extent. / AFRIKAANSE OPSOMMING: Hierdie tesis beskryf ‘n metode wat gebruik kan word om falings in kern grafiet komponente te voorspel. Hierdie komponente word in hoë temperatuur gas reaktore gebruik. Die falings metodologie beoog om die falings van regte komponente te voorspel wat gebaseer is op meganiese toets resultate van materiaal monsters. Dit is ‘n statistiese falings metodologie wat die waarskynlikheid van faling vir grafiet komponente bereken. Die metode is numeries ontwikkel en geïmplementeer deur middel van die eindige element metode, dus kan die metodologie toegepas word op enige geometrie en belastingsgeval wat dan gemodelleer kan word deur gebruik te maak van eindige element metodes. Die metodologie word gedemonstreer deur gebruik te maak van NBG-18 kern grafiet toets monsters. Sommige van hierdie toetse is ontwikkel as ‘n uitbreiding van die materiaal karakterisering wat spesifiek ontwerp is om die effek van die spannings konsentrasies op die faling van die NBG-18 komponente te evalueer. Twee relevante statistiese verspreiding funksies word gekoppel aan die eksperimentele sterkte data van die NBG-18 kern grafiet, naamlik ‘n normale verspreiding en ‘n twee-parameter Weibull verspreiding. Die data stelle word ook genormaliseer vir gemak van vergelyking en gekombineer in een verteenwoordigende data stel. Die gekombineerde data stel slaag ‘n korrelasie toets wat impliseer dat die meganisme van faling soortgelyk is tussen die data stelle. ‘n Drie-parameter Weibull korrelasie op die trek toets monsters word gebruik vir die statistiese falings metodologie. Die analise van die eksperimentele resultate sowel as ‘n bespreking van die akkuraatheid van die faling voorspelling metodologie word voorgelê. Die data word geanaliseer by gemiddelde faling voorspelling asook by laer voorspellings van falings. Hierdie metode is gebaseer op die bestaan van ‘n “ketting volume” wat die volume van ‘n materiaal wat gebruik word in die swakste ketting voorstel en koppel aan die metodologie. ‘n Metode vir die benadering van die ketting volume word voorgestel en daaropeenvolgend gebruik om die ketting volume te bereken vir NBG-18. ‘n Gedetailleerde evaluasie van die falings voorspelling vir elke toets geval word uitgevoer vir die voorgestelde ketting volumes. Gebaseer op hierdie ondersoek is voorgestelde ketting volumes vir NBG-18 gegee vir beide akkurate en konserwatiewe falings voorspellings. Verder was ‘n volgrootte strukturele toets ontwikkel om dieselfde falings omstandighede te simuleer wat verwag is gedurende normale werking van die reaktor. Drie monsters word getoets en geëvalueer teen die voorspelde faling vir beide die berekende ketting volume groottes. Faling van die volgrootte komponente word realisties asook konserwatief voorspel. Die voorpselling is 20% konserwatief. Die metodologie is gebaseer op die Weibull metode wat inherent volume afhanklik is; gevolglik dui die konserwatisme aan dat die metodologie oor volume afhanklikheid beskik soos ondervind word in die klassieke Weibull teorie, maar tot ‘n baie kleiner mate.

An advanced concept high temperature reactor (HTR) design has been proposed - The ‘U-Battery’, which utilises a unique sealed coolant loop, and is intended to operate with minimal human oversight. In order to reduce the need for moving parts within the design, CO2 has been selected as a candidate coolant, potentially allowing a naturally circulated system. HTR fuel is held within a semi-graphitic fuel-matrix material, and this has not previously been tested within a CO2 environment. Graphite in CO2 is subject to two oxidation reactions, one thermally driven and one radiolytically. As such, the oxidation performance of fuel-matrix material has been tested within CO2 at both high temperatures and under ionising radiation within a sealed-system. Performance has been compared to that of the Gilsocarbon and NBG-18 nuclear graphite grades. Gilsocarbon is the primary graphite grade used within the currently operating AGR fleet within the UK, and as such is known to have acceptable oxidation performance under reactor conditions. NBG-18 is a modern graphite grade, and is a candidate material for use within the U-Battery. Virgin characterisation of all materials was performed, including measurements of bulk mass and volume, skeletal volumes and surface areas. High-resolution optical microscopy has also been performed and pore size distributions inferred from digital image analysis. All results were seen to agree well with literature values, and the variation between samples has been quanti- fied and found to be < 10% between samples of Gilsocarbon, and < 4% for samples of fuel-matrix and NBG-18. Thermal performance of fuel-matrix material was observed between 600 °C – 1200 °C and seen to be broadly comparable to that of the nuclear graphite grades tested. NBG-18 showed surprisingly poor performance at 600°C, with an oxidation rate of 3×10−4%/min, approximately ten times faster than Gilsocarbon in similar conditions, and three times faster than fuel-matrix material. The radiolytic oxidation performance of fuel-matrix material and NBG-18 has been observed by irradiating sealed quartz ampoules. Ampoules were pressurised with CO2 prior to irradiation, and the pressure after 30 days of irradiation was measured and seen to fall by 50%. Radiolytic oxidation, and the subsequent radiolysis of the reaction product, CO, was seen to cause significant carbonaceous deposition on the internal surfaces of the ampoule and throughout the samples. Due to the short irradiation times available in the present study, an investigation of the microporosity within irradiated samples has been carried out, using nitrogen adsorption and small-angle neutron scattering (SANS). Pore size distributions produced from SANS show the closure of microporosity within NBG-18, most likely as a result of low-temperature neutron irradiation.As a result of this work, CO2 is no longer a candidate coolant for use with the U-Battery design, due to the rapid deposition observed following irradiation.

Carbon materials play a fundamental role in the development of fusion reactors, for both the generation of electric power and the production of nuclear materials. It is possible to synthesise graphite and carbon materials from coal. Coal is available in large quantities and could be used for the production of high-purity carbon graphite. However, it contains large quantities of impurities that need to be removed prior to graphitisation/carbonisation. The impurity levels of certain elements in this graphite must be kept at very low levels. Boron, which absorbs neutrons strongly, should be below 500 ppb. Europium and gadolinium, which absorb neutrons and are activated to highly radioactive products, as is cobalt, should be as low as 50 ppb. Lithium transforms to tritium, which leads to the circulating helium becoming radioactive. Other elements, such as calcium, sodium, silicon, thorium and uranium, should not be ignored. The purpose of this study was to lower or remove completely the impurities and trace elements in coal that affect the quality of nuclear-grade graphite. The organic part of Tshikondeni coal was dissolved in a solvent, dimethylformamide (DMF), on addition of sodium hydroxide. The first stage of purification is centrifugation and filtration, which removes most of the impurities. The recovered organic material, known as ‘Refcoal’, may be converted to graphitisable coke. Some elements, significantly boron and cobalt, associate with the organic material in solution and are not sufficiently separated by centrifugation and filtration. Further purification was employed during each process step in the conversion of coal solution into graphite. Different methods of purification were employed in this study. They included chlorination, acid treatment and the ion-exchange or complexation method. Chlorine gas and hexachlorocyclohexane (benzene hexachloride) were used in the chlorination method. Acids such as hydrochloric, hydrofluoric and ascorbic were used in acid treatment. In the ion-exchange method, reagents such as methane, starch, potassium cyanide, ethylene-diaminetetraacetic acid, sodium fluoride, sodium sulphate, ice, glycerol and sodium nitrate were used. All the treated Refcoal was coked at 1 000º C. Pyrolysis was applied in other methods with the aim of volatilising elements that form volatile halides at higher temperatures. Analysis was done for elements such as calcium, cobalt, europium, gadolinium, lithium, sodium, silicon, thorium and uranium, and other elements in the periodic table. Inductively coupled plasma mass spectroscopy and inductively coupled plasma optical emission spectroscopy were used to analyse the concentrations of the trace elements in the coal (treated and untreated) and the coked Refcoal. In inductively coupled plasma mass spectroscopy, microwave digestion and fusion were applied as methods of preparation. However, the instrumentation gave different results for the same sample. The results showed that specific methods work for specific elements. The chlorination method and the acid-treatment method (especially using hydrofluoric acid and hydrochloric acid) gave better purification for most of the trace elements and other elements. Better purification was achieved with elements such as, boron, calcium, europium, gadolinium, lithium, sodium and silicon. All the treatments failed to lower uranium and thorium to the level required for nuclear-grade graphite. However, uranium has a low boron equivalent and does not pose serious problems with respect to nuclear usage. All the methods failed to remove cobalt and this remains a problem. / Dissertation (MSc (Chemistry))--University of Pretoria, 2008. / Chemistry / MSc / unrestricted

The aim of this work has been to apply the techniques of non-destructive testing and evaluation to the graphite fuel channel bricks which form the core of an Advanced Gas-Cooled reactor. Two modes of graphite degradation have been studied: subsurface cracks originating from the keyway corners of the bricks and the reduction in material density caused by radiolytic oxidation. This work has focused on electromagnetic inspection techniques. Brick cracking has been studied using a multi-frequency eddy current technique with the aim of determining quantitative information. In order to accurately control the crack dimensions this work has used radially machined slots as an analogue. Two sensor geometries were studied and it was determined that slots of at least 10 mm through-wall extent could be located. A novel, empirical method of determining the slot size is presented using a brick machined with a series of reference slots. Machined slots originating from a keyway could be sized to within 2 mm using this method. A parametric 3D finite element study was also carried out on this problem. These simulations could distinguish the location of the slots and had some sensitivity to their size, however, the model was found to be overly sensitive to the specific mesh used. Two new contributions to the inverse problem are presented. The first is a minor extension to the usual adjoint problem in which one system now contains a gradiometer. The second is a proposed solution to the ambiguous nature of the inner product required by the sensitivity formulation. This solution has been validated with finite element modelling. Density reduction has been studied via its relationship with electrical conductivity using a technique based on impedance spectroscopy. An inverse eddy current problem has been solved using the regularised Gauss-Newton method to determine the conductivity of the brick over its cross section. The associated forward problem has been solved using the finite element method on a simplified geometry. Tikhonov regularisation has been employed to overcome the ill-posed nature of the inverse problem. This method has been applied to a range of sample and sensor geometries and found to produce excellent results from laboratory data provided the finite element model is well calibrated. Bore or surface conductivity values can be reproduced to better than 1% with the accuracy reducing with distance from the sensor. The sensitivity of the algorithm to the regularisation parameter has been studied using the L-curve method and the effect of two regularisation operators has also been examined. A new method of choosing the regularisation parameter a priori is proposed and tested. Data taken during reactor outages produces physically realistic profiles although the results appear off-set from electrical resistivity values measured using the four-point method. The focus of future work should be to remove this effect which will likely require improvements to the forward model.

NBG–18 graphite is developed by SGL Carbon for the Pebble Bed Modular Reactor Company (PBMR), and is used as the preferred material for the internal graphite core structures of a high–temperature gas–cooled nuclear reactor (HTR). The NBG–18 graphite is manufactured using pitch coke, and is vibrationally molded. To assess the structural behaviour of graphite many destructive techniques have been performed in the past. Though the destructive techniques are easy and in some cases relative inexpensive to perform, these methods lead to waste material and require cumbersome time consuming sample preparations. To overcome this problem numerous non–destructive testing techniques are available such as sonic resonance, resonant inspection, ultrasonic testing, low and multifrequency Eddy current analysis, acoustic emission and impulse excitation techniques. The Hammer Impulse Excitation technique was used as a method in predicting the fatigue life of NBG–18 graphite by focussing on the application of modal frequency analysis of determined natural frequencies. Moreover, the typical fatigue characteristics of NBG–18 graphite were determined across a comprehensive set of load ranges. In order to be able to correlate modal frequency parameters with fatigue life, suitable uniaxial fatigue test specimen geometry needed to be obtained. The uniaxial fatigue test specimens were manufactured from two NBG–18 graphite sample blocks. The relationship between natural frequencies of uniaxial test specimens, fatigue life, sample positioning and sample orientation was investigated for different principle stress ratios. Load ratios R = –oo and R = +2 tested proved to show the highest r–values for the Pearson correlation coefficients investigated. However, there was no significant trend found between the natural frequency and the fatigue life. / Thesis (M.Ing. (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2011.

NBG–18 graphite is developed by SGL Carbon for the Pebble Bed Modular Reactor Company (PBMR), and is used as the preferred material for the internal graphite core structures of a high–temperature gas–cooled nuclear reactor (HTR). The NBG–18 graphite is manufactured using pitch coke, and is vibrationally molded. To assess the structural behaviour of graphite many destructive techniques have been performed in the past. Though the destructive techniques are easy and in some cases relative inexpensive to perform, these methods lead to waste material and require cumbersome time consuming sample preparations. To overcome this problem numerous non–destructive testing techniques are available such as sonic resonance, resonant inspection, ultrasonic testing, low and multifrequency Eddy current analysis, acoustic emission and impulse excitation techniques. The Hammer Impulse Excitation technique was used as a method in predicting the fatigue life of NBG–18 graphite by focussing on the application of modal frequency analysis of determined natural frequencies. Moreover, the typical fatigue characteristics of NBG–18 graphite were determined across a comprehensive set of load ranges. In order to be able to correlate modal frequency parameters with fatigue life, suitable uniaxial fatigue test specimen geometry needed to be obtained. The uniaxial fatigue test specimens were manufactured from two NBG–18 graphite sample blocks. The relationship between natural frequencies of uniaxial test specimens, fatigue life, sample positioning and sample orientation was investigated for different principle stress ratios. Load ratios R = –oo and R = +2 tested proved to show the highest r–values for the Pearson correlation coefficients investigated. However, there was no significant trend found between the natural frequency and the fatigue life. / Thesis (M.Ing. (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2011.