Fusion energy for use in power plants is a continually developing area and many of the related parameters are not yet fixed. The investigation of fusion neutronics and development of computational approaches for assessment is imperative in the road to commercial realisation of fusion power. This research has explored blanket performance, including tritium breeding and the shielding requirements, and assessed radioactive waste, utilising the 3-D Monte Carlo transport code MCNP, and the activation inventory code FISPACT. The performance of some solid and liquid breeder materials has been compared with regards to tritium breeding, energy production and shielding. In the case of novel spherical tokamak concepts, that make use of high temperature superconducting magnets and have no inboard blanket, scoping studies have been performed to investigate the impact of shielding requirements on how small the tokamak can be. Fusion power plants will not produce high level waste, as seen in nuclear fission plants, however the components and structures will become active as a result of interactions with high energy neutrons. A suitable radioactive waste management plan will be required in order to deal with this material appropriately, with an aim to recycle or clear from regulatory control all materials 100 years after shutdown. The study indicates that through suitable material selection and the use of component dismantling the requirement could potentially be satisfied. In terms of computational methods, the neutron flux averaging has been assessed throughout the work and has shown in neutronics estimates to produce some substantial differences. The recently developed unstructured mesh approach to neutronics modelling has been explored and the potential use for more accurate radioactive waste inventory calculations. Although the analysis and comparison shows promising results, it still requires significant development and improvement in the work flow to create a robust neutronic analysis method.

In the Scrape-Off Layer (SOL) of magnetic confinement devices, cross-field transport of particles is dominated by the convection of filamentary plasma structures via self-generated E×B velocity fields. This thesis investigates the dynamics of such filaments using three dimensional simulations to further theoretical understanding of SOL transport. A new 3D SOL simulation code called STORM3D has been developed using the BOUT++ framework to implement an isothermal drift-reduced fluid model in a slab geometry. Verification and validation exercises are documented to demonstrate that the code has been implemented correctly and that the physical model adequately reproduces experimental observations. A comprehensive characterisation of how a filament’s initial geometry affects its subsequent dynamics is provided via a series of 3D simulations of isolated filaments. In particular the size of a filament in the plane perpendicular to the magnetic field, δ⊥, is shown to have a strong influence on its motions, as it determines which currents balance the filament’s pressure-driven diamagnetic currents, which in turn determines its E×B velocity. At small δ⊥, this balance is predominantly provided by polarisation currents and the filament’s radial velocity is observed to increase with δ⊥. In contrast, at large δ⊥, parallel currents closing through the target are found to be dominant, and the radial velocity decreases with δ⊥. Comparisons are made between 3D simulations and 2D simulations using different parallel closures; namely the sheath dissipation closure, which neglects parallel gradients, and the vorticity advection closure, which neglects the influence of parallel currents. The vorticity advection closure is found not to replicate the 3D perpendicular dynamics well and overestimates the initial radial velocity of all filaments studied. A more satisfactory comparison is obtained with the sheath dissipation closure, even in the presence of significant parallel gradients, where the closure is no longer valid. The vorticity advection closure’s poor performance occurs because in the 3D case parallel currents closing through the sheath play an important role in reducing the extent to which polarisation currents are driven. In a conduction-limited or detached SOL regime however, low plasma temperatures and high neutral densities near the divertor will produce significantly higher resistivity values in the region than that used in the aforementioned 3D simulations. Therefore the effect of increasing the normalised plasma resistivity in the last quarter of the domain nearest the targets is examined using 3D simulations. Whilst small δ⊥ filaments are observed to be relatively unaffected by this quantity, large δ⊥ filaments exhibit faster radial velocities at higher resistivity values due to two mechanisms. Firstly, parallel currents are reduced meaning that polarisation currents are necessarily enhanced and secondly, a potential difference forms along the parallel direction so that higher potentials are produced in the region of the filament for the same amount of current to flow into the sheath. This indicates that broader SOL profiles could be produced at higher values of normalised resistivity, and hence at larger reference SOL densities and at colder temperatures.

Cobalt-60 and iridium-192 sources are both generated in nuclear reactors through the irradiation of stable target materials. Variation in: neutron energy, flux and irradiation time; target material characteristics and purity; the activation cross sections of the desired neutron reactions; decay and daughter progeny in-growth; and any post-irradiation processing, can all play a key part in determining the isotopic and chemical composition of the material produced. These isotopic ratios, together with any activated elemental impurities, could potentially be used as signatures in order to provide information relating towards not only the material’s production date, but also the source’s production route, irradiation history and original elemental and isotopic composition. All of which are key indicators for material attribution in the event of a nuclear forensics incident. Computational studies to evaluate the effects of neutron flux spectra on nuclide production within these materials have indicated a number of impurity nuclides that have the potential to be used as nuclear forensics signatures. These signatures though show significant variability depending on the irradiation position and reactor conditions used. When combined with the various issues that investigations into target material purity and manufacturing conditions highlighted regarding signature development, the complexity of the problem starts to become a strong impediment towards the identification of a material’s source of origin. It is therefore unlikely that these impurity signatures could ever be used in nuclear forensics investigations, because investigators would need to know the specific processes and procedures employed by manufacturers in order to attribute materials back to their source of origin. This is an unlikely scenario based on the number of manufacturers and the variety of processes that are employed during source production.

No description available.

Three aspects of the mechanical behaviour of fast reactor fuel have been studied, two being theoretical studies of deformation during operation and the third an experimental investigation of creep in a particular fuel. a. Thermal stress in fuel pellets A computer program was constructed to calculate, by a difference approximation, thermal stress in finite cylinders for a given temperature distribution. This was used to describe the effect of length to diameter ratio, central hole diameter and heat flow conditions, on the stresses and surface displacements of fast reactor fuel pellets. Estimates of the likely crack behaviour of pellets were made and it was concluded that cracking would prevent pellet shape changes from producing clad ridges. However, under certain conditions ridging could be produced by temperature dependent swelling shape change. b. Restraint of the swelling fuel by its clad Mathematical techniques and computer programs were developed to treat this problem in the radial direction for transient and steady state conditions, The programs were then used to derive the basic characteristics of fuel pin behaviour, to determine the effect of changes in variables, and to compare oxide and carbide fuels. c. Compressive creep of zone refined uranium monocarbide Single crystals and polycrystalline specimens of carbon rich and metal rich UC were deformed in a vacuum in a compressive creep furnace, in the 1000-1300°C range. Metallography and Laue back-reflection patterns were used to aid the interpretation of the creep results. A slip system of {111} type was found to be consistent with the results. Comparison with the single crystal results of Bentie et al, (1963) indicated that a single creep mechanism, probably controlled by uranium diffusion, was operating for carbon rich UC in the temperature range 1000-2000°C. Creep in metal rich UC is more complex, being dependent on grain size and metal solubility.

The possibility of gaseous emissions from cyclotron sites during the production of positron emitters for use in nuclear imaging is a public safety issue, and operators are required to measure any such emissions accurately. Despite the fact that stack effluent monitor instrumentation has been available for several years, a suitable calibration facility does not exist. Traceability to a primary standard is essential if such measurements are to be accurate. This project focused on the development of a primary standard based on a system already established at NPL for the standardisation of beta-emitters such as 85Kr and 133xe. Previous standardisations of beta+-emitters have been carried out using liquid or solid sources. Unlike the previously mentioned standardisation methods for positron emitters (which have been carried out using liquid or solid sources), a primary standard for positron emitters in gas has been developed, which can be used to calibrate the instrumentation used at PET/cyclotron sites.11C was chosen as the radionuclide to be standardised, 18F, 15O and 13N could not be used because of experimental factors. To provide an active gas sample suitable for counting and to provide a comparison with an existing standard based on measurement of a liquid, a method has been developed for conversion of sodium bicarbonate to carbon dioxide with subsequent drying and trapping of the gas. A sample of 11C liberated in this way was measured by internal gas proportional counting using a system of three counters. This document describes the standardisation of 11C by gas counting and the comparison of the primary standardisation results with an independent measurement of a sample of the same sodium bicarbonate solution that was used to generate the active gas. The independent measurement was carried out using a secondary standard ion chamber. The experimental determination of counting losses in internal gas proportional counting was difficult. Some of the losses for standardizing 11C,13N, 15O 18F have been calculated using the PENELOPE Monte Carlo Code. The PENELOPE code was selected for the study of the counting losses because it was the most suitable code for positron interaction simulation at the time of this work. The correction factors calculated with PENELOPE were less than 0.6% for the system of the three counters used as the primary standard. A transfer instrument for calibration of local instruments at sites of positron emitting radionuclides can be carried out based on the primary standard.

Ion implantation is increasingly used as a cost effective and safe alternative to neutron irradiation in the study of materials for future nuclear applications. However, the micron-sized layers of radiation damage necessitate the use of small scale testing techniques to obtain mechanical properties data from these materials. This thesis explores such techniques, highlighting the relative merits and limitations of nanoindentation, and the deflection of FIB-machined specimens, in a variety of applications. As a simple starting point, the effect of crystal orientation on nanoindentation hardness and modulus is explored in a pure Fe polycrystal, with selected indents examined further post-test by HR-EBSD and AFM. Hardness in grains with < 111 > and < 110 > surface normal directions is found to be ~ 20 % higher (~ 1.1 GPa) than those with < 100 > surface normals (~ 0.95 GPa), whilst no discernible trend is seen for modulus. A novel L-shaped micro-cantilever technique is developed and used to extract single crystal elastic constants within a single grain of Fe. A computational routine minimising the difference between experimental data and the response of an FE model is used to extract such constants, with a good agreement between experiment and literature found after two runs. Slip transfer across grain boundaries is studied in unirradiated and ion-irradiated samples using nanoindentation. Grain boundaries in the irradiated specimen show less resistance to slip transfer, and the presence of dislocation bursts associated with the operation of dislocation sources in adjacent grains is reduced compared to the unirradiated sample. Finally, the effects of size and orientation on plastic properties are examined in FIB-machined micro-cantilevers. Two different functions are used to fit flow stresses evaluated at different maximum strains to beam width, and compared. Cantilevers orientated for multiple slip are observed to have lower flow stresses than those aligned for single slip with a Schmid factor of 0.5. The work in this thesis highlights the importance of both types of small scale mechanical test in providing information of relevance to candidate materials for nuclear applications. Nanoindentation is able to quickly probe mechanical characteristics, but at the expense of a well-defined stress state under the indenter tip, whereas FIB-machined specimens can provide a simpler stress state to test, for higher costs of time and instrument usage.

No description available.

Currently most commercial nuclear power stations are based on thermal reactor designs called burner reactors which are. net consumers of fissile material. These power stations form one part of a larger system that generates electricity from uraniura. However, in addition to producing energy, such systems also consume energy, in the form of various fuels, during construction and operation. This thesis describes the use of energy analysis to determine the total energy required by these systems. A number of factors are shown to influence energy consumption and, in particular, the effect of extracting uranium from different sources is studied in detail. For ores, an important inverse relationship between energy use and ore grade is investigated and quantified. The physical limit at which the energy input to the system is equal to its output is shown to correspond to an average grade of 15 parts per million of "triuranium octoxide". Analysis of proposals for extracting uranium from seawater indicates that the only schemes giving a positive energy balance are'costly ($500/lb U₃0₈) and limited to low production rates. The effects of feedback within fuel systems are analysed and. the results are used to formulate an economic model in which nuclear electricity prices determine uranium ore costs as well as vice versa. The model demonstrates-that, with present'techniques, the average 6 economic limit to ore grade is 50 ppm U₃0₈ with subsequent resources, on current assessment, of only 107 tonnes U₃0₈. This contradicts most traditional studies which, by assuming fixed, non-dependent fuel costs, suggest an ore grade limit of less than 4 ppm U₃0₈ and economically recoverable resources in excess of 10¹⁰ tonnes U₃0₈.

No description available.