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Currently in the UK, composite blends of Portland cement (PC) and blastfurnace slag (up to 90%) are commonly used for the encapsulation of low and intermediate level wastes. The high volume replacement of PC is considered necessary in order to to reduce the high heat generation resulting from cement hydration in 500 litre waste packages. While suited to the majority of waste streams, the high pH environment in such systems (usually around 12.5 -13), will cause the corrosion of certain waste metals such as aluminium. Since aluminium is only passive between pH4 - 8.5, the use of an alternative low-pH cement system could serve to reduce/inhibit the corrosion. However, before such cements can be considered, two main research problems must be addressed, as follows: (1) quantitative evaluation of alternative cement systems based on their corrosion performance with aluminium; (2) high heat generation due to the rapid rate of hydration. The research in this thesis was thus divided into two strands, as follows: (1) The design and development of a novel, scientifically robust testing facility for the quantitative monitoring of aluminium corrosion in cement pastes; (2) the development of novel cement composites based on weakly alkaline calcium sulfoaluminate (CSA) cement for the encapsulation of aluminium from nuclear wastes. The output from this research is considered to be of interest to the UK nuclear industry.

This report investigates the detection of small leaks from the primary system of a Nuclear Pressurised Water Reactor. Leak rates of 12 g/s are invariably difficult to detect and locate. The typical leak indicators in a nuclear reactor control room are a drop in pressure and level from the pressuriser, and the air sampler detecting particulate matter. However, in both cases the leak is normally quite substantial by the time any parameters or values are obviously outside the normal operating conditions. Therefore, a small leak could go undetected for a significant amount of time. As part of the reactor safety studies, it is important to have more information about small leaks. Due to the lack of small leak data, the solution was to construct a high pressure water rig producing temperatures and pressures close to those experienced in the primary circuit, these being 200ºC and 100 bar respectively. Pressure is maintained by a vane water pump and heating is achieved by passing a high current through a small diameter, thin walled pipe. To reproduce different size cracks, various size carburettor jets are used. The water on exiting this crack, flashes to steam and immediately meets metallic pipe lagging, which is typical of most primary systems. With the typical crack scenario recreated it is now important to add sensors that will detect conditions associated with a small leak. These sensors are either mounted on or around the lagging material. The parameters that are monitored include vibrations, acoustics, thermal variations, moisture change, air flow and pressure adjustment leaving a predetermined outlet. The sensor outputs are pre-processed and the nonlinear data are applied to an artificial neural network, whereas the other data are applied to a digital logic system. The results showed that with 13 different leak rates, separated by only 1.4 g/s the ANN was able to correctly differentiate and identify different leak sizes with a certainty of over 97%. The results from all the analysis are further presented graphically through an Operator Advisory System. This informs the operator of the predicted leak size and location. All of the available sensor data relevant to the leak can be viewed and location of the leak is presented by a three dimensional model of the reactor system.

This thesis presents a complete theoretical framework for turbulence and transport in tokamak plasmas. The fundamental scale separations present in plasma turbulence are codified as an asymptotic expansion in the ratio of the gyroradius to the equilibrium scale length. Proceeding order-by- order in this expansion, a framework for plasma turbulence is developed. It comprises an instantaneous equilibrium, the fluctuations driven by gra- dients in the equilibrium quantities, and the transport-timescale evolu- tion of mean profiles of these quantities driven by the fluctuations. The equilibrium distribution functions are local Maxwellians with each flux surface rotating toroidally as a rigid body. Large-scale deviations of the distribution function from a Maxwellian are given by neoclassical theory. The fluctuations are determined by the high-flow gyrokinetic equation, from which we derive the governing principle for gyrokinetic turbulence in tokamaks: the conservation and local cascade of free energy. Transport equations for the evolution of the mean density, temperature and flow ve- locity profiles are derived. These transport equations show how the neo- classical corrections and the fluctuations act back upon the mean profiles through fluxes and heating. This framework is further developed by exploiting the scale separation between ions and the electrons. The gyrokinetic equation is expanded in powers of the electron to ion mass ratio, which provides a rigorous method for deriving the electron response to ion-scale turbulence. We prove that such turbulence cannot change the magnetic topology, and ar- gue that, therefore, the magnetic field lies on fluctuating flux surfaces. These flux surfaces are used to construct magnetic coordinates, and in these coordinates a closed system of equations for the electron response is derived. All fast electron timescales have been eliminated from these equations. Simplified transport equations for electrons in this limit are also derived.

This work focuses on diffusion and advection through cementitious media, the work arises from two research contracts undertaken at Loughborough University: Experiments to Demonstrate Chemical Containment funded by UK NDA and the SKIN project, funded by the European Atomic Energy Community's Seventh Framework Programme. Diffusion will be one of the most significant mechanisms controlling any radionuclide migration from a nuclear waste, deep geological disposal facility. Advection may also occur, particularly as the immediate post closure groundwater rebound and equilibration proceeds but is expected to be limited by effective siting and management during the operational phase of the facility. In this work advection is investigated at laboratory scale as a possible shorter timescale technique for providing insight into the much slower process of diffusion. Radial techniques for diffusion and advection have been developed and the developmental process is presented in some detail. Both techniques use a cylindrical sample geometry that allows the radionuclide of interest to be introduced into a core drilled through the centre of the test material. For diffusion the core is sealed and submerged in a container of receiving solution which is sampled and analysed as the radionuclide diffuses into it. For advection, a cell has been designed that allows inflow via the central core to pass through the sample in a radial manner and be collected as it exits from the outer surface. The radionuclide of interest can be injected directly into the central core without significant disturbance to the advective flow. Minor improvements continue to be made but both techniques have provided good quality, reproducible results. The majority of the work is concentrated on a potential cemetitious backfill known as NRVB (Nirex Reference Vault Backfill) this is a high porosity, high calcium carbonate content cementitious material. The radioisotopes used in this work are 3H (in tritiated water), 137Cs, 125I, 90Sr, 45Ca, 63Ni, 152Eu, 241Am along with U and Th salts. In addition the effect of cellulose degradation products (CDP) on radioisotope mobility was investigated by manufacturing solutions where paper tissues were degraded in water, at 80°C, in the absence of air and at high pH due to the presence of the components of NRVB. All diffusion experiments were carried out under a nitrogen atmosphere. All advection experiments were undertaken using an eluent reservoir pressurised with nitrogen where the system remained closed up to the point of final sample collection. Results for tritiated water and the monovalent ions of Cs and I were produced on a timescale of weeks to months for both diffusion and advection. The divalent ions of Sr, Ca and Ni produced results on a timescale of months to years. Variations of the experiments were undertaken using the CDP solutions. The effects of CDP were much more apparent at radiotracer concentration than the much higher radiotracer with non-active carrier, concentration. In the presence of CDP Cs, I and Ni were found to migrate more quickly; Sr and Ca were found to migrate more slowly. Additional Sr experiments were undertaken at elevated ionic strength to evaluate the effect of the higher dissolved solids content of the CDP solutions. Some of the results for HTO, Cs, I and Sr have been modelled using a simple numerical representation of the system in GoldSim to estimate effective diffusivity and partition coefficient. The diffusion model successfully produced outputs that were comparable to literature values. The advection model is not yet producing good matches with the observed data but it continues to be developed and more processes will be added as new results become available. Autoradiography has been used to visualise the radionuclide migration and several images are reproduced that show the fate of the radiotracers retained on the NRVB at the end of the experiments. As the experimental programme progressed it was clear that results could not be produced in a suitable timescale for Eu, Am U and Th. These experiments have been retained and will be monitored every six months until either diffusion is detected or the volume of receiving liquid is inadequate to ensure the NRVB is saturated.

As the stocks of fossil fuels are rapidly depleting the world has turned to other forms or electricity generation including nuclear power. The production of electricity via nuclear power already supplies a large amount of the world's population and is becoming increasing more viable as the concern of global warming also becoming progressively more apparent. Thorium dioxide fuel has been widely researched and investigated as a potential replacement to uranium dioxide for many years as it has many advantages over the current uranium dioxide fuel. Due to the hazards of working with radioactive materials in the laboratory, computational work has become a popular method to complete initial predictions of the properties and characteristics of the fuel. A new potential model was developed for both the Th-O and the Gd-0 interactions using two different derivation methods. In both cases the potential model included the shell model rather than the previously used rigid ion model; the shell model has been proven to be superior in modelling defects and defect interactions. Potential validation using bulk properties confirmed the robustness of the potentials and allowed confidence in taking them forward to investigate defects such as the introduction of fission products, surface simulations and molecular dynamic simulations. Within this work the pure and mixed oxide fuels have been examined using various atomistic modelling codes including GULP, METADISE and DL_POLY to allow a robust understanding of the properties and features of the fuel.

Nuclear power is currently receiving a great deal of attention in the international arena as nations seek to tackle the joint energy challenges of supply security and sustainability. Discussions have focused on questions of efficiency, cost, waste disposal and the safety profile of the technology. Within the social sciences attention has been given to the decision making processes surrounding nuclear technologies, and to themes of dialogue and participatory decision making. This body of research conceptualises nuclear industries as products of social and political processes. In this thesis, I have conceptualised the Polish nuclear industry as the product of social, political, but also material forces which together co-shape dialogue surrounding the pursuit of nuclear development. The thesis presents a case study of nuclear development in Pomerania, northern Poland. During the 1980s, the Communist government of the People’s Republic of Poland began to construct the first of a series of nuclear power plants beside Lake Żarnowiec in the north of the Province. However, following protests and political upheaval in the region, construction work was abandoned and the site left empty for nearly twenty years. Since 2009, Lake Żarnowiec has once again taken centre stage as Poland seeks to construct a nuclear facility in Pomerania. This thesis examines the agency of the infrastructural assets of the original development, landscape changes, notions of identity, and the social and political processes which underpin decision making surrounding the pursuit of nuclear development within the region. It demonstrates the presence of a complex entanglement of social, material and conceptual elements present in dialogue surrounding contemporary nuclear development. The research calls attention to the forces and pressures, both seen and unseen which together shape political and social attitudes towards nuclear development.

Approximately 96,000 tonnes of the UK Higher Activity Waste (HAW) inventory consists of irradiated nuclear graphite. The current Nuclear Decommissioning Authority (NDA) baseline strategy for irradiated graphite in England and Wales is isolation in a future Geological Disposal Facility, with Scottish policy endorsing an alternative decision of near surface long-term storage. Irradiated graphite disposal routes in the UK remain under review, however, as there are concerns surrounding timing and whether deep geological disposal is the most appropriate course of action for graphite. An alternative waste management solution is treatment prior to disposal to separate mobile radioactive isotopes such as 3H and 14C from the bulk material, allowing for HAW volume reduction and concentration. Optimisation of an existing thermal treatment process at the Nuclear Graphite Research Group (NGRG) of the University of Manchester has been effected and a detailed review of the uncertainties associated with quantitative determination of radioisotope releases during thermal treatment of irradiated graphite samples has been conducted. Thermal treatment experiments in both an inert atmosphere and 1% oxygen in argon atmosphere have been conducted for temperatures ranging from 600°C to 800°C, and durations from 4 to 120 hours, to determine the effects of oxidation time and temperature, and the consequent oxidation characteristics on the release rate of prominent radioisotopes, with a focus on the release of 14C. Lower temperature treatments in an oxidising atmosphere have shown that a preferential release of 14C-enriched graphite can be achieved from the bulk material of Oldbury Magnox reactor irradiated graphite, with evidence demonstrating that this liberated 14C-enriched region is located at the graphite surfaces throughout the porous structure. A large proportion of radiocarbon found in this irradiated graphite, however, is uniformly distributed throughout the bulk material and cannot be selectively oxidised. It is found that prominent metallic radioisotopes such as 60Co are not mobile at these temperatures and remain in the bulk graphite material, inclusive of radioactive caesium which the literature suggests will volatilise. The preliminary results were undertaken as part of the EU FP7 EURATOM Project: CARBOWASTE.

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.