Atomic scale simulations have been performed investigating various phenomena governing nuclear fuel performance during reactor operation and during post irradiation storage or disposal. Following a review of some of the key features of irradiated nuclear fuel, such as fission product distribution, two key factors were identified as the focus for this investigation: i) the role of uranium dioxide non-stoichiometry and ii) the effect of temperature. The former has been carried out using a previous pair potential model, whilst a new many-body potential was developed to enable temperature effects to be studied over the full range of temperatures of interest. Secondary oxide precipitates are known to exist in irradiated nuclear fuel with Ba, Sr and Zr precipitating to form the perovskite (Ba,Sr)ZrO3 grey phase. The binary BaO, SrO and ZrO2 may also be formed. The precipitation enthalpies of these oxides were predicted as a function of hyper-stoichiometry. Additionally, CrUO4 can also precipitate from Cr-doped UO2+x. The possibility of fission product segregation to the phases from UO2 or UO2+x was also investigated with a broad range of species preferring segregation from stoichiometric UO2. The role of defect cluster configuration on vacancy mediated uranium migration was investigated for UO2 and UO2+x. In both cases the lowest enthalpy migration pathway involved reconfiguration of the cluster to a metastable configuration. Furthermore, there were a very large number of alternative pathways that had similar migration enthalpies, especially for UO2+x. A new potential model was developed that uses a novel approach to include many-body interactions in the description of the actinide oxide series. This represents a significant improvement on previous models in the ability to describe the thermal expansion, specific heat capacity and elastic properties of CeO2, ThO2, UO2, NpO2, PuO2, AmO2 and CmO2 from 300 to 3000 K. Using the new model the thermal expansion, specific heat capacity, oxygen diffusivity and thermal conductivity of the mixed oxides (UxTh1-x)O2 and (UxPu1-x)O2 were predicted. Enhanced oxygen diffusion and a degradation in thermal conductivity were predicted in terms of the non-uniform cation sublattice.

The presence of uranium in grouted Magnox intermediate level waste (ILW) containers is a concern for the nuclear waste industry. Recent inspection of 16 randomly selected ILW containers revealed bulging around the circumference of 3 containers, the cause of which is currently unknown. The bulges are suspected to be related to the intermittent existence of large uranium masses, the corrosion of which produces products almost twice the volume of uranium metal. Formation of hydrogen and uranium hydride through the oxidation of uranium and Magnox with water is of particular concern owing to their pyrophoric nature on exposure to oxidising conditions should the container fail. The understanding of uranium corrosion mechanisms and rates in ILW containment is therefore vital in assessing the risk posed by the bulging waste containers during container storage, transportation and disposal. The following study aimed to contribute to this understanding by performing experiments that transition from laboratory based investigations to conditions more relevant to real ILW scenarios.

The decommissioning of the UK fleet of Magnox nuclear power stations will require the disposal of some 57,000 tonnes of irradiated graphite from the reactor cores. Currently, this graphite is classified as intermediate level waste due to the incorporation of long half-life radioisotopes such as carbon-14 and chlorine-36. The plan for disposal of such waste is to place it in an engineered geological disposal facility, and currently the graphite accounts for approximately 15% of the total volume of the UK intermediate level waste inventory. An understanding of the concentration, distribution and potential release of carbon-14 from the graphite is critical for ensuring the safe disposal of this ,vaste. In this study 49 irradiated PGA graphite samples trepanned from the fuel and interstitial channels of Oldbury Reactor One were examined. They were compared with virgin PGA graphite used to fabricate the core. It was observed that the graphite exposed to the service environment had a significantly different structure to the virgin PGA. A distinct surface deposit of up to a few tens of micrometre thickness was observed on both the fuel and interstitial channel wall faces. In subsequent experiments the carbon- 14 concentration and distribution were determined using two separate techniques, one based on mass spectrometry and the other using thermal oxidation and subsequent . scintillation counting. The main observations from this work were that the surface deposit was relatively enriched in carbon-14 compared to the underlying graphite, with a higher concentration measured on samples originating lower in the fuel channels. The carbon-14 content measured was not correlated with the lifetime neutron dose and was therefore likely to have arisen from transmutation of the precursor species present in the coolant gas. The remaining carbon-14 was determined to be split between two fractions, firstly, a fraction originating from nitrogen surface complexes formed on the surface and near sub-surface, that may be released slowly in a geological disposal facility, and a second fraction from nitrogen and carbon-13 precursors located in the graphite lattice, that may never be released. These results correlate well with previous reports in the literature on the leaching of irradiated graphite and results from this work may provide useful inputs into radioactive release models as well as developing analytical techniques that can be used in the future examination of graphite arising from different Magnox reactors to determine the total fraction of carbon-14 that may be released from a geological disposal facility.

No description available.

This thesis outlines the development of a nodal method with the purpose of addressing difficulties encountered in the modelling of advanced fuels. The standard calculational route used when modelling a Uranium (U) fuelled Pressurised Water Reactor (PWR) is not accurate enough to analyse a PWR containing U and Plutonium (Pu). This is because the assumptions made when developing the standard route are not necessarily representative of situations involving advanced fuels. To address some of these poor assumptions a nodal method has been developed which can solve the SPN equations in multiple energy groups. The SPN equations are an asymptotic approximation of the full neutron transport equation, and as such will include more physical effects than the neutron diffusion equation. The theory behind the development of this nodal method is outlined in this thesis along with an extensive set of benchmark tests for verification of the method. It is found that through a similarity transformation of the determining equations, existing nodal diffusion solvers can obtain solutions to the SPN equations without any approximations. Previously EDF Energy have developed an embedded methodology to address the shortcomings of the standard calculational route. This procedure solves the diffusion equation in greater detail on local sub-meshes in order to correct the standard 2 group nuclear data, and reduces the pin power errors by ≈ 50% by capturing spectral effects on the interface between two significantly different fuel types. In this thesis the incorporation of the SPN nodal method into the embedded methodology is described. A small light water reactor benchmark is solved to test the accuracy of the embedded methodology combined with the SPN nodal method. It is concluded that similar accuracy to diffusion is attained with the SPN equations. This is because the homogenisation procedure produces an error larger than the improvements due to the use of the SPN equations. To address the limitations discovered in this thesis future work is proposed based on the author’s experience of research in the area.

The Canadian nuclear waste management concept envisages using carbon steel as a primary engineered barrier for isolating nuclear waste in a deep geological repository (DGR) located in sedimentary rock. Steel corrosion in anticipated repository environments has been studied, but was mostly focused in two main areas: (i) aerobic or oxygen containing environments (both in vapour and liquid phases); and (ii) anaerobic, solution environments. The atmospheric corrosion behaviour of steel in a humid, anaerobic or anoxic environment, is a new topic, with virtually no published data to on which to rely. This study was undertaken to improve the existing knowledge of anaerobic, atmospheric corrosion of carbon steel. Atmospheric corrosion testing was conducted on carbon steel wires in anoxic atmospheres at various temperature and relative humidity (30-100% RH), with and without sodium chloride (NaCl) contamination of the wire surfaces. Hydrogen evolved from corrosion was monitored and converted to an estimated corrosion rate. With salt on wire surfaces, sustained final corrosion rates in the range of 0.01 to 0.8 μm·y-1 were observed over test durations of 935 to 1725 hours. Without salt contamination, the corrosion rates are very low, and can only be detected using a solid-state electrochemical hydrogen sensor. The hydrogen sensor can detect the pressure increase down to a limit of ca. 0.1 Pa, corresponding (depending on the exact procedure) to a corrosion rate as low as ca. 0.0001 μm·y-1. The estimated corrosion rates for the degreased and pickled wires were found to be < 0.01 μm·y-1. In parallel with the corrosion experiments, corrosion product surface analyses were performed using different techniques. Oxides formed on steel surfaces consist mostly of the mixed Fe2+/Fe3+ spinel oxide Fe3O4. The experimental results of this study will be applied to assess the corrosion behaviour of carbon steel containers during the anoxic, unsaturated phase of a deep geological repository in Canadian sedimentary rock.

As nuclear waste is being considered for geological deposition, a safe and durable method for sequestering it from the environment is needed. The radiation tolerance of various ceramic compounds was analysed using alpha-decay event molecular dynamics simulations. As the methods traditionally used for damage identification tend to over-predict the amount of damage under nonequilibrium conditions, a method novel for this type of simulation was developed. Using spherical harmonics, the Steinhardt method uses the angles between atoms to build up a fingerprint of the local structure which is translationally and rotationally invariant. This can then be compared to reference structures in order to quantify the amount of each structure type within the simulation cell. Both single and multiple decay event simulations were used to probe the pyrochlore series Gd₂(TixZr₁-x)₂0₇,where 05 x 51, in order to investigate the underlying mechanisms of damage and healing as well as how these mechanisms change with the varying of x. The single decay event simulations outlined the importance of cascade overlap in the accumulation of damage and so high pressure multiple decay event simulations were run. Changes in volume and total energy were compared to structural data from the Steinhardt order parameters in order to gain insight into the atomistic mechanisms responsible for damage and healing. Atmospheric pressure simulations were also run and structural changes consistent with ion bombardment experiments were observed. The results showed that the radiation tolerance of the solid solution increases with increasing Zr content due to the enhanced healing that high ion mobility brings with it. Proposed models for amorphisation kinetics were tested against the structural data from these simulations and a new model was also suggested. Multiple alpha decay simulations were also run on PU02 in order to investigate its tolerance of alpha radiation. The local structure surrounding each atom was analysed using Steinhardt order parameters in order to quantify the damage caused by the alpha decay events. The results of these simulations show that like Gd₂Zr₂0₇, PU0₂ is extremely tolerant of alpha radiation due to its high ion mobility.

The production of nuclear waste is undoubtedly a major downside to nuclear energy. In the UK, much of our intermediate and high level waste is currently stored in temporary facilities with the aim of burying it in deep underground facilities by 2040. In order to achieve this, the radionuclides require immobilising to prevent them from leaching into the environment. This can be achieved by incorporating the radionuclides into minerals, ceramics or glasses before encapsulation in disposal containers. In order for a material to be successful in this role it must be chemically durable, thermally stable and radiation resistant. The current plan for the disposal of high level waste is to combine it with molten borosilicate glass before encapsulation in stainless steel containers (vitrification). This is a far from perfect solution however as, amongst other faults, these glasses have been found to undergo amorphisation of the newly-formed crystalline phase which over time can lead to microcracking and swelling; thus reducing the integrity of the wasteform. This study is therefore focused on ceramic wasteforms as an alternative to vitrification, specifically the orthosilicate, ASi04, and pyrochlore, A2B207, structure types. This research will describe how low temperature hydrothermal synthesis (at temperatures of just 150-240 QC) offers significantly greater control over the product structure and morphology than traditional solid state methods. The wasteform materials produced were also found to exhibit outstanding thcll11al stability and excellent chemical durability.

No description available.

No description available.