No description available.

Phosphate minerals such as β-tricalcium phosphate and fluorapatite are abundant, exhibit a large chemical variability and samples of both phases are often used for geochronometry; so their stability over thousands of years is known. This makes them attractive as nuclear waste forms as they are stable over extended periods of time and are able to accommodate a range of waste species. In particular, due to the difficulty of incorporating halides in conventional waste forms (in particular glass), fluorapatites are considered as potential hosts for radioactive waste streams containing both actinides and halides. β -tricalcium phosphate, a structure related to fluorapatite, is not only used as a precursor for preparing apatites, but could itself be used as a host for radioactive waste. To better understand the structure and stability of these phases, atomic scale computer simulation has been employed (both dynamic and static), using classical pair potentials. One site in the β -tricalcium phosphate structure has previously been assigned a half occupancy from fitting to Rietveld power diffraction experimental data. The half occupancy gives rise to an ordering effect where the Ca2+ ions are arranged over the half occupied Ca(4) sites, in many different configurations. A comparison of different cell sizes (and increasing configuration number) is given and reported in terms of lattice energy and structural parameters. The largest cell size considered here generates the most stable structures, which have a symmetry group related to the experimentally derived average primitive unit cell. Simulated X-ray diffraction patterns indicate a difference in spectra for these low energy configurations. However, experimental X-ray diffraction patterns fail to differentiate between low energy structures. The results of the experiment and small difference in lattice energy for the most stable structures, indicate that domains containing different low energy structures are likely to exist. The configurations described here are also discussed in terms of statistical analysis. Substitutions of a range of isovalent and trivalent cations have been carried out (and compared) at Ca2+ sites in both fluorapatite and β-tricalcium phosphate structures. The defect and solution energies were calculated for both structures and compared, in order to predict the partitioning across the two phases. For the isovalent defects investigated, the defects segregate to the β-tricalcium phosphate lattice. The trivalent defects present a different trend such that defects with a smaller ionic radii than Ca2+ have an energy preference for β-tricalcium phosphate, but for those with atomic radii closer to that of the host cation the preference for either structure is indistinguishable. The ramifications of cation partitioning are discussed. Cation and anion migration in fluorapatite is considered, where the overriding feature of migration in this structure is that ionic transport (of either cations or anions) occurs preferably along the c-axis. Consideration of fluorine transport yields a more sophisticated migration mechanism than that reported previously. A “concerted mechanism” for fluorine ion transport in the lattice is discussed fully and described using fluorine density plots, over a range of temperatures. Radiation damage effects in these minerals and their ability to recover and resist damage is a crucial consideration when designing a nuclear waste host. This thesis compares the Kinchin- Pease model of threshold displacement energies to full radiation cascades. Furthermore, the effect of radiation damage on the lattice is considered by virtue of incident damage, defect types, phosphate group response and recovery of the lattice. These results indicate that the fluorapatite lattice is “relatively tolerant” to radiation damage, especially with regard to the phosphate tetrahedra, but more cascades should be considered to obtain a more statistically reliable prediction. Finally, loss of material from the waste form is most likely to occur at a surface and understanding the processes by which this occurs is important. Therefore, it is necessary to first model the surfaces, which involves static simulations of surfaces to predict surface formation energies. These energies are used as the basis for predictions of particle morphology and the effect of isovalent defect incorporation near the surface of fluorapatite.

The research described in this thesis is directed at advancing the state of the practice of the non-destructive gamma-ray assay of nuclear waste containers. A number of potentially accuracy-limiting issues were identified and addressed, resulting in new developments which were implemented on an instrument prior to entering it into service. A set of Pu reference sources used for experimental data have been studied to determine the internal composition (density and fill height) of the sources to assist with validation of a point kernel model. This model has been used to observe the behaviour of gamma-rays in lumps of fissile material from plutonium over the mass range 0.001g to 350g, for a number of densities corresponding to Pu, PuO\(_2\) and PuF\(_3\). Established lump corrections have been analysed and have been found to produce large over- and under-corrected results for the range of masses. Due to the inadequacies of current techniques, a new Pu self-absorption correction method has been developed using the data from numerical simulations, allowing nature to reveal the correlations rather than traditional approaches based upon approximate models. For a 25g 1cm-high Pu flat-plate of density 15g.cm\(^{-3}\), the developed Pu correction produces a result of (24.9 ± 8.8)g compared to (19.5 ± 0.9)g for the Fleissner 2-line method, and (14.7 ± 0.4)g for the Infinite Energy Extrapolation method. The developed Pu correction method has been extended to the application of uranium lumps in waste matrices, provided the enrichment of the sample is known or may be determined via sophisticated isotopic analysis methods such as MGAU or FRAM. The U self-absorption correction method has been found to produce results within 30% of the true mass of the sample for the lumps studied. An analysis of ‘real drum’ effects has been performed, including the revisiting of the Total Measurement Uncertainty (incorporating the uncertainty components of the new Pu and U self-absorption corrections) and results from known sources placed in artificial inhomogeneous waste matrices assayed inside a Canberra Auto Q2 system.

The nuclear industry in line with many other industries has a problem with contaminated organic waste in aqueous effluent. During the PUIREX process operated by British nuclear Fuels Ltd, the recovery of radioisotopes using 30% tributyl phosphate in odourless kerosene, the organic phase is degraded due to the effects of radiolysis. This results in a very complex mixture with dibutyl phosphate, monobutyl phosphate and phosphoric acid amongst its components; the odourless kerosene is also degraded forming long chain carboxylic acids. Another consequence of this process is the production of finely divided solids such as zirconium phosphate, these solids together with the complex organic mixture form stable emulsions, known as cruds. These tend to form at the interface between the organic and aqueous layers. These cruds cause problems with further recovery of metal and also physical problems such as mass transfer. This project was initially aimed to ascertain whether microwaves could be used to assist the total oxidation of these cruds. However, due to the complexity of these cruds it was decided that a model system should be studied. Tributyl phosphate was chosen as this model, this was deemed to be appropriate as it is the single most stable component of the "cruds". The theory behind the technique is if a suitable sensitiser was used tributyl phosphate would absorb onto its surface. When exposed to microwave energy a localised plasma would be generated around the surface of the particles. It is in this very highenergy field that oxidation of the tributyl phosphate would occur. Work performed at the University by John Rawcliffe for his MSc thesis, had demonstrated the concept that tributyl phosphate could be converted to its degradation products by the use of microwave radiation. The aim of this project was initially to establish sound analytical methods for measuring the degradation products of tributyl phosphate and other chemicals that were to be investigated during the work, and to optimise the reaction conditions to greatly improve the conversion of tributyl phosphate to its degradation products. It was hoped this could be achieved by altering a number of parameters including air and liquid flow and the way in which the microwaves are applied to the type of sensitiser used. When the optimum conditions had been established then further work was undertaken to scale-up the size of the reaction vessels to ascertain the difficulties that would be faced in scaling up to pilot plant as this was the ultimate aim of the project sponsor. In order for scale-up to go ahead new reaction vessels and microwave cavities had to be designed and commissioned, and the initial microwave equipment adapted to accommodate the larger reaction vessels. It was envisaged that this technique could have much wider application. To demonstrate this hypothesis the system was tested on other chemicals that were potentially problematic for the nuclear industry, in terms of their disposal.

This thesis describes the application of a passive sampler, Diffusive Gradient in Thin Films (DGT), to the measurement of uranium in natural waters. Four resins (Chelex-100, manganese dioxide [MnO2], Diphonix® and MetsorbTM) were trialled with the DGT device. In freshwater environments, the MetsorbTM accumulated uranium in line with the DGT equation for 7 d with an acuracy of 75%; Chelex-100 did not accumulate uranium past 2 d; MnO2 accumulated up to 75% of that predicted by the DGT equation for 4 d; and the Diphonix® accumulated uranium for 7 d with an accuracy of ~100%. None of the resins tested in this study accumulated uranium in a marine setting in line with DGT predicted values past 2 d. The application of DGT to regulatory environmental monitoring schemes was investigated with MetsorbTM. The MetsorbTM DGT devices were deployed for 7 days at a time over a 6 month period at two freshwater field sites. Fluctuations in water chemistry were monitored and the size of the diffusive boundary layer (DBL) was measured. The uranium accumulated by the MetsorbTM DGT showed close agreement with the grab samples. The size of the DBL was found to be significant, particularly in low flow conditions. This study showed that DGT could be used as a tool to both monitor radioncludes in the environment, and to obtain information on the speciation and organic interactions. The lability of uranium-humic acid complexes was also examined in this study. Initial data shows that the uranyl-humic complex is labile in low pH environments, but becomes increasingly kinetically limited the higher the pH and the higher the humic acid:uranium ratio. Data is also presented on the penetration parameter of the uranyl ion into the resin gel layer, and how this can be used to indicate lability. Lability is important in determining bioavailability and potential toxicity of uranium.

Understanding the behaviour of actinide species is of importance when removing and processing all nuclear waste. Examples include the safe clean-up of contaminated waste ponds and aspects of the TALSPEAK (Trivalent Actinide-Lanthanide Separation by Phosphorus Reagent Extraction from Aqueous Komplexes) process. The chemistry of the ponds and the TALSPEAK process has been studied by probing the aqueous solution behaviour of Ln(III), Am(III), Cm(III) and Th(IV) ions in the presence of organic (EDTA4- (ethylenediamine tetraacetate), DTPA5- (diethylenetriamine pentaacetate) and lactate) and inorganic (CO32- (carbonate) and OH- (hydroxide)) ligands by a variety of techniques including Nuclear Magnetic Resonance (NMR), Ultra Violet-Visible (UV-Vis) and luminescence spectroscopies, as well as potentiometry. Various ternary complexes have been shown to exist, including [M(EDTA)(CO3)]3-(aq), (where M = LnIII, AmIII or CmIII) and [Th(EDTA)(CO3)2]4-(aq), which form approximately over the pH range 8 to 11, and also [M(EDTA)(lactate)]2-(aq) (where M = Ln(III) or Am(III)) and [Th(EDTA)(lactate)]-(aq), which predominantly occur over the pH range 4 to 6. The nature of lactate interaction with [M(DTPA)]2-(aq) complexes (where M = Ln(III) or Am(III)) is unclear, as it may be possible that lactate can coordinate directly to the metal ion or to the acetate groups of DTPA5- (via a H-bonding interaction). The knowledge gained in this research has given a deeper insight into the nature of lanthanide and actinide coordination chemistry in mixed-ligand environments. For example, the increasing solubility of actinide metal ions in the contaminated waste ponds is probably due to the ability of organic ligands present in the ponds to solubilise metal ions at high pH, and also under TALSPEAK conditions of pH 3.5, there is likely to be minimal interaction of lactate with the [Ln(DTPA)]2-(aq) complexes. The determination of metal ion speciation using a combination of NMR, UV-Vis and luminescence spectroscopies, coupled with potentiometry, could be applied to new characterisation challenges faced in the future of the nuclear industry.

No description available.

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.

Geological disposal is the preferred option for the long term management of British intermediate level radioactive waste. The disposal site is currently being identified, with possible geological environments including fractured crystalline rocks and low permeability rocks such as clay. The selection of the host rock will have an impact on the design of the waste repository. This thesis investigates the ways the behaviour of repository borne gas can be affected by the repository design and the selection of the host rock. Commercially available TOUGH2 package is used to model the resaturation of the disposal facility, along with gas migration out of the repository and towards the ground surface in a generic geology. A facility located in fractured rock is estimated to resaturate within 6.5 years of its closure. The resaturation time is found to be strongly dependent on the presence and properties of a low permeability liner around the disposal vaults. The inflowing water starts gas generation processes within the repository; gas initially accumulates within the facility, but it is estimated to find its way into the host rock approximately 450 years after the facility has been closed. A maximum outflow rate is reached after approximately 1,000 years. The flow of gas migrating through the host rock is strongly affected by site-specific features. In the case of a uniform crystalline rock, gas is found to break through at the surface after 29,000 years. For a disposal site with a very slow groundwater flow rate, the resaturation phase may take several decades and gas outflow will occur much later. It is estimated that, in very low permeability environments, gas breakthrough may not occur before 100,000 years.

The cores of early UK graphite moderated research and production nuclear fission reactors operated at temperatures below 150°C. Due to this low temperature their core graphite contains significant amounts of stored (Wigner) energy that may be released by heating the graphite above the irradiation temperature. This exothermic behavior has lead to a number of decommissioning issues which are related to long term "safe-storage", reactor core dismantling, graphite waste packaging and the final disposal of this irradiated graphite waste. The release of stored energy can be modeled using kinetic models. These models rely on empirical data obtained either from graphite samples irradiated in Material Test Reactors (MTR) or data obtained from small samples obtained from the reactors themselves. Data from these experiments is used to derive activation energies and characteristic functions used in kinetic models. This present research involved the development of an understanding of the different grades of graphite, relating the accumulation of stored energy to reactor irradiation history and an investigation of historic stored energy data. The release of stored energy under various conditions applicable to decommissioning has been conducted using thermal analysis techniques such as Differential Scanning Calorimetry (DSC). Kinetic models were developed, validated and applied, suitable for the study of stored energy release in irradiated graphite components. A potentially valid method was developed, for determining the stored energy content of graphite components and the kinetics of energy release. Another parameter investigated in this study was dedicated in the simulation of irradiation damage using ion irradiation. Ion bombardment of small graphite samples is a convenient method of simulating fast neutron irradiation damage. In order to gain confidence that irradiation damage due to ion irradiation is a good model for neutron irradiation damage the properties and microstructure of various grades of ion irradiated nuclear graphite were also investigated. Raman Spectroscopy was employed to compare the effects of ion bombardment with the reported effects of neutron irradiation on the content of the defects. The changes of the of defect content with thermal annealing of the ion irradiated graphite have been compared with the annealing of neutron irradiated nuclear graphite.