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Environmental degradation promoted by the presence of sludge piles in the steam generators of Pressurised Water Reactors (PWR) can pose a threat to their safe and continuous operation. The use of dispersants can reduce the rate at which sludge accumulates. Polyacrylic acid (PAA) is currently the only dispersant used in PWRs. Settling rate tests identified several dispersants with the potential to outperform PAA, notably Hydroxyethylidene-1,1-diphosphonic Acid (HEDP). To estimate the dispersant concentration required during plant operation, optimum concentrations were identified for both PAA and HEDP. Nuclear magnetic resonance spectrometry was used to investigate the thermal stability of HEDP between 230 and 270oC, revealing that HEDP decomposes more rapidly than PAA. The dominant HEDP decomposition product was shown to be orthophosphate but several other long lived intermediate products were detected. The effect of dispersants on the environmental degradation of grade 316 stainless steel was determined by electrochemical methods and by constant extension rate tests. Rates of general corrosion measured by linear polarisation resistance and electrochemical impedance spectroscopy were shown to be very low (on the order 10-5 mm/year) under aerated and deaerated conditions both at room temperature and at 70oC. Corrosion rates were slightly reduced in the presence of PAA and HEDP. Constant extension rate tests demonstrated that neither PAA nor HEDP promote stress corrosion cracking at 250oC. It was concluded HEDP would not be suitable for use in PWRs due to its rapid thermal degradation rate. The decomposition products were shown to rapidly concentrate in steam generators thereby preventing accurate control of water chemistry.

One of the current proposed strategies for the disposal of intermediate level radioactive waste (ILW) is that of an underground facility termed a geological disposal facility (GDF). Under the anoxic alkaline conditions (10.5 < pH > 13) expected to develop within an ILW-GDF cellulose will undergo chemical degradation to a range of cellulose degradation products (CDP). The major components of which will be the isosaccharinic acids (ISA). ISA is of particular interest with regards to the long term performance of a GDF due to its ability to complex radionuclides altering their solubility and thus the long term retention of such radionuclides. The removal of these complexants by microbial action could therefore have an impact upon the long term performance of an ILW-GDF. The biofilm mode of existence is able to increase the resistance of microbes to environmental stresses and could be a mechanism for microbial survival under ILW-GDF conditions. Recent research has shown the ability of microbes from an ILW-GDF anthropogenic analogue site to degrade ISA under hyperalkaline conditions, however, no studies have yet looked at the ability of alkaliphilic biofilms to survive under these conditions whilst using ISA as a carbon source. The aim of this work was to ascertain the ability of alkaliphilic biofilm consortia to survive and degrade ISA under near field conditions and to further investigate the mechanisms of this survival. Using cotton bait, biofilm was able to be grown in-situ at an analogue site located at Harpur Hill, Buxton, UK. Enrichment of this biofilm using CDP driven microcosms under both sulphate reducing and methanogenic conditions at pH 11 showed the degradation of the α-ISA, β-ISA and xylo ISA types through a fermentation pathway to acetate and hydrogen. Investigation into the morphology of these microcosms revealed the microbes to existing within flocs or aggregates of EPS which had a complex EPS composition. Microelectrode profiling of these flocs revealed areas of lower pH than the external pH to be present within the interior of these flocs, a finding which was attributed to microbial survival under the alkaline conditions. Floc cultures were able to form thick dense biofilm within sand columns which enhanced the rate of ISA degradation and facilitated the production of methane and sulphide under relevant conditions. Micro-electrode pH profiling again demonstrated low pH areas within the biofilm which contributed towards biofilm survival and ISA degradation up to pH 13. Biofilms were able to impact ILW-GDF relevant surfaces with evidence of carbonation and EPS substances found upon NRVB, steel and graphite surfaces. The ability of biofilms to form under ILW-GDF conditions could facilitate microbial survival under ILW-GDF conditions and have an impact upon the long term performance of an ILW-GDF. This could be through the carbonation of NRVB surfaces, the blocking of pore throats, gas production, microbial induced corrosion related to sulphide production and through the removal of radionuclide complexants.

Characterization of edge tokamak plasma instabilities by measuring emergent phenomena within a range of frequencies above the ion cyclotron frequency have been explored in two ways: using the inter-event waiting times of Edge Localized Modes (ELMs) occurrences in measured time series of JET plasmas and by performing 2D/3D simulations of filamentary structures dynamics using a hybrid model plasma description, i.e. kinetic ion particles and massless charge neutralizing electron fluid. The analysis of ELMs time series using characteristic emission lines Da of JET tokamak in otherwise similar plasmas was performed with only a minimal number of drivers such as the gas puffing rate. They have shown a key role in changing the underlying system mode cycle where a threshold value revealed its transition from single harmonic behaviour to a transitioning phase into a total lost of the state and born of a higher frequency resonant mode. Hybrid simulations of blobs/filaments are performed in 2D/3D to observe the kinetic evolution of these plasma structures over several ion gyroperiods. Novel 3D simulations represent the first kinetic simulations of these structures along the parallel direction using a kinetic description. We have investigated the evolution and the internal density currents which provide insight of the effects of finite Larmor radius in the blobs dynamics and evolution.

Tritium is a by-product of civil nuclear reactors, military nuclear applications, fusion programmes and radiopharmaceutical production. It commonly occurs, though not exclusively, as tritiated water (HTO) or organically-bound tritium (OBT) in the environment but may exist as other forms in nuclear-related construction and fabrication materials. During the lifetime of nuclear sites (especially those involving heavy water) tritium becomes variably incorporated into the fabric of the buildings. When nuclear decommissioning works and environmental assessments are undertaken it is necessary to accurately evaluate tritium activities in a wide range of materials prior to any waste sentencing. Of the various materials comprising UK radioactive wastes, concrete and metal account for approximately 20% of the total weight of low level waste (LLW) and 12% and 35% of the total weight of intermediate level waste (ILW). Proper sampling and storage of samples are significant factors in achieving accurate tritium activities. The degree of loss of 3H and cross-contamination can be significantly reduced by storing samples in an air/water tight container in a freezer (-18°C). The potential for tritium contamination is dependent on the 3H form. Most 3H loss originates from tritiated water which is easily exchanged with atmospheric hydrogen in the form of water vapour at room temperature. However, the loss of more strongly bound 3H, produced in-situ in materials by neutron activation, is not significant even at room temperature. Such tritium is tightly retained in materials and does not readily exchange with water or diffuse. In nuclear reactor environments tritium may be produced via several neutron-induced reactions, 2H(n,g)3H, 6Li(n,a)3H, 10B(n,2a)3H and ternary fission (fission yield <0.01%). It may also exist as tritiated water (HTO) that is able to migrate readily and can adsorb onto various construction materials such as structural concrete. In such locations it exists as a weakly-bound form that can be lost at ambient temperatures. Bioshield concretes present a special case and systematic analysis of a sequence of sub-samples taken from a bioshield core (from UKAEA Winfrith) has identified a strongly-bound form of 3H in addition to the weakly bound form. The strongly bound 3H in concrete is held more strongly in mineral lattices and requires a temperature of >850°C to achieve quantitative recovery. This more strongly retained tritium originates from neutron capture of trace lithium (6Li and potentially 10B) distributed throughout minerals in the concrete. The highest proportion of strongly bound 3H was observed in the core sections closest to the core. Weakly bound tritium is associated with water loss from hydrated mineral components. Tritium is retained in metals by absorption by free water, hydrated surface oxidation layer, H ingress into bulk metal and also as lattice-bound tritium produced via in-situ neutron activation. Away from the possible influence of neutrons, the main 3H contamination to metals arises from absorption and diffusion via atmospheric exposure to the HTO. Here contamination is mainly confined to the metal surface layer. The tritium penetration rate into metal surfaces is controlled by the metal type and its surface condition. Where metals are exposed to a significant neutron flux and contain 6Li, 7Li and 10B then in situ 3H production will occur which may propagate beyond the surface layer. In such cases tritium may exist in two forms namely a weakly bound HTO form and a non-HTO strongly bound form. The HTO form is readily lost at moderate temperatures (~120°C) whereas the non-HTO requires up to 850°C for complete extraction.

In pressurised water reactors (PWRs), hydrogen overpressure is used to keep the corrosion potential below the threshold for onset of intergranular stress corrosion cracking (IGSCC) in type 304 SS. However, some regions may contain higher oxygen levels resulting in an increase in the potential. These 'dead space' regions are difficult to access and during refuelling; oxygen may become trapped in these locations. The objective of this study was to investigate the influence of PGM additions on IGSCC susceptibility of type 304 stainless steels (SS) in the sensitised state within PWRs. The work presented herein investigates several aspects of the IGSCC problem. Virgin and platinum group metal (PGM)-modified (Ru and Pd) 304 SS have been studied. Material characterisation, including microstructural, tensile properties, hardness and grain size measurements, has been conducted. Crack initiation studies using U-bend samples in autoclaves simulating PWR environments have also been performed. In addition, crack propagation studies using circumferential cracked bar (CCB) specimens under constant extension in potassium tetrathionate solutions, a well-known medium to promote IGSCC on sensitised stainless steels, have been conducted in order to evaluate cracking resistance. Electrochemical studies using model solutions for PWR chemistry (containing boric acid and lithium hydroxide) and also potassium tetrathionate were carried out to look at the influence of the PGM on the kinetics of the main electrochemical reactions. The results revealed that PGM additions appeared to reduce crack initiation on sensitised type 304 SS under oxygenated conditions in high temperature water containing sulphate and chloride. PGM-doped and standard sensitised type 304 stainless steels revealed susceptibility to IGSCC propagation in 0.01 M K2S4O6, at pH=1.5 and 25°C. Electrochemical studies in potassium tetrathionate media showed smaller anodic dissolution peaks with PGM additions and metallography indicated less intergranular attack with PGM additions. In PWR model electrolytes, PGM additions, particularly 1 wt% Ru, were shown to catalyse the oxygen reduction reaction or hydrogen oxidation reaction, depending on the oxygen /hydrogen level. Overall findings showed that Ru additions can improve the IGSCC resistance of sensitised type 304 SS in PWR, while Pd additions are less effective.

In nuclear environments, irradiation hardening and damage have a detrimental effect on materials performance. Among others, fracture toughness of austenitic stainless steels decreases under neutron irradiation. Helium arising from transmutation reactions is one source of embrittlement leading to that decrement and it is here assumed as a case study, austenitic steel 316L being the material under investigation. The experimental reproduction of irradiation hardening effect on yield stress is attempted here by pre-strain under tensile loading at room temperature. The experimental production of porosity is attempted by inducing ductile damage, creep damage or a combination of them. Damage at the microstructural level is analyzed by metallography, fractography, X-ray tomography and quantified by image processing.After calibrating the elastic, the plastic and the porous plastic constitutive equations by the means of tensile tests on smooth and notched specimens, results from damaging experiments are validated by finite element analysis using the Gurson-Tvergaard-Needleman model. The numerical models obtained represent different levels of damage into the material, as induced by the experiments.Material presenting different levels of damage is then machined for fracture toughness evaluation in the shape of sharp-notched round bars. Fracture toughness initiation is inferred from the load vs. displacement plots applying an opportune fracture criterion. In order to test the suitability of the Gurson-Tvergaard-Needleman model, the load vs. displacement results are validated by retrofitting opportune constitutive laws for each “damaged” state. Retrofitting is discussed in relation to the type of damage produced.Results show that the reproduction of the macroscopic effect of irradiation hardening on yield stress may be attempted for 316L by a pre-strain tensile loading at room temperature for levels up to 1.5 dpa or slightly more. These interrupted tensile tests did not give evidence of void volume fraction production. Creep tests at 650 °C showed sensitization at the grain boundaries but not porosity into the matrix. Creep tests at 1000 °C created 1.2% to 1.8% void volume fraction from grain boundary sliding. Finally, one 7% pre-strained specimen was subjected to creep test at 900 °C and stopped at 5% creep strain, without evidence of porosity into the matrix.Fracture toughness tests on the “damaged” states obtained before showed a decrement of fracture toughness initiation when compared with “undamaged” 316L. Specimens with 30% and 40% eng. strain presented a sensible decrement and exhibited a brittle-like behaviour. The differences in porosity size and physical processes involved suggest not stating that a correlation exists with the helium embrittlement effect on the same property. The Gurson-Tvergaard-Needleman model worked for the “undamaged” material. It proved to be not suited for the brittle-like 30% and 40% eng. strain “damaged” materials because it did not capture the experimental progression of damage.In the end, fracture toughness numerical predictions were made using different values of initial void volume fraction. It was argued that, starting from a threshold value, the brittle-like 30% and 40% eng. strain “damaged” materials revert to a ductile behaviour.

Solar energy has been extensively used in the renewable technology field, especially for domestic applications, either for heating, electrical generation or for a combination of heat and power (CHP) in one system. For CHP system solar photoelectric/thermal (PV/T) is the most commonly used technology for roof top applications. However, combination between solar hot water and thermoelectric generators has become an attractive for CHP system, this is due to its simplicity of construction and its high reliability. Moreover, this technology does not rely simply on sunlight and it can work with any other heat source, such as waste heat. However, its main drawback is its low efficiency. Recent publications by Kraemer et al (2011) and Arturo (2013) have shown that the efficiency of solar thermoelectric systems has improved dramatically, especially when combined with a solar concentrator system, as well as within a vacuum environment. The project recorded in this thesis focused on the design, construction and investigation of an experimental solar thermoelectric system based on a flat plate solar absorber. The aim was to study the technical feasibility and economical viability of generating heat and electric power using a solar thermoelectric hot water system. The design procedure involved on determining the heat absorbed and emitted, as well as the electrical power that was generated by the system. It began by obtaining the efficiency of the solar absorber, including selecting its paint, this was done through an experimental technique to determine the heat absorbed by the absorber, and the results obtained were verified by direct measurements of the light intensity. xvi An intensity meter was used, and results from both the experimental and theoretical models showed good agreement. The process also included calculating the heat from the system that was gained, lost and generated, as well as the electrical power provided. This was done to provide the system optimal size optimization to obtain the best and most economical system. Further improvement was made to the system by assembling a vacuum cavity, to improve the system’s efficiency. Although the maximum electrical efficiency obtained was relatively low (0.9%), compared to results recorded in the literature (Kraemer et al ,2011 and Arturo, 2013). However, the results of the electrical power output, under a vacuum level of 5 x 10-2mbar, increased approximately three times compared to the results obtained under normal (atmospheric) conditions. Additionally, the thermal power increased by 37% at this level of vacuum. The process involved determining the best thermoelectric geometries to achieve the optimum power outcome under different environmental conditions. The results showed that the system, which included the Thermoelectric device (TEG) with a larger geometric size, produced the best thermal power among other sizes. It was concluded that the system with the smallest TEG geometric size provided the best electrical power output.

In nuclear reprocessing, PUREX, a solvent extraction process, has long been the separation method employed for the separation of the bulk components of irradiated nuclear fuel (namely uranium and plutonium) from the fission products and other minor actinides produced during the fuel use. The uranium and plutonium constitutes approximately 96 % by mass of the irradiated fuel and for this to be removed, requires large volumes of extractant and equipment with large surface area contactors and therefore floor space requirements. The PUREX process has for nearly 60 years been the largely unchallenged separation technology for the reprocessing of irradiated fuel, for both nuclear weapon production and commercial nuclear power generation. The merits and ability of this process are unquestionable since it achieves the objectives of highly purified plutonium and uranium which both can be eventually recycled. Although well proven and predictable, the PUREX process is not without its challenges: the generation of significant quantities of highly active aqueous liquid containing fission products (FPs) and minor actinides (MAs), and the degradation of the solvent phase reagents and non-specific nature of the extractant TriButylPhosphate (TBP) may have contributed to only a fraction of the total annual output of irradiated fuel being reprocessed. Fission products are elements which are produced in a nuclear reactor and are the atomic fragments left after a large atomic nucleus (typically uranium-235) undergoes nuclear fission, splitting into two smaller nuclei, along with a few neutrons, the release of heat energy (kinetic energy of the nuclei), and gamma rays. Minor actinides such as neptunium, americium, curium, berkelium, californium, einsteinium, and fermium are the actinide elements in irradiated nuclear fuel other than uranium and plutonium; they are minor as they represent a very small proportion of actinides in comparison to U and Pu. This thesis explores the possibility of using a continuous chromatographic method to extract the lesser components of the irradiated fuel. One of the major problems with the use of chromatography as an industrial process is the expansion from the batch separations on the bench top to a continuous efficient process, capable of processing large volumes. This thesis, through existing concepts, will describe a proof of concept chromatographic separation of surrogates and isotopes of the components of irradiated fuels which can be readily scaled up to a continuous chromatographic separation. The project is a radical departure from PUREX and will offer many advantages over PUREX. It is based on the separation of FPs and MAs from uranium and plutonium isotopes using continuous chromatographic separation. This thesis assesses a number of commercial resins for their suitability for the proposed continuous chromatography reprocessing method. The experiments were all undertaken at elevated nitric acid concentrations and as such are describing interactions which are rarely required commercially and therefore seldom reported, with batch studies to assess separation factors between ions, uptake kinetics and isotherms over a range of nitric acid concentrations to more dynamic column breakthrough and eventually separations. The research demonstrates that a separation can be achieved at an elevated HNO3 concentration on a commercially available ion exchange resin.

Nuclear power remains the most dense and reliable primary form of energy supply worldwide. Electricity generation via fission is also inherently carbon free, with environmental footprints rivalling modern renewable options. However issues arise from the production of highly irradiated spent fuels, with current management options limited to geological repository storage or, more desirably, closing of the fuel cycle by partitioning to recover fissionable species. Extensive research has been pursued over the past half century in an effort to address an accumulating oxide spent fuel inventory and to circumvent shortcomings of raw uranium supplies. Pyrochemical reprocessing or ‘pyroprocessing’ using molten salt electrolysis for the recovery of desirable spent fuel components is an increasingly sought after solution to the above issues. First developed in the late 1990’s, the FFC Cambridge Process has shown to be a cornerstone in the electrochemical reduction of metal oxides and potentially offers a new iteration of pyroprocessing by introducing a direct conversion of spent oxide fuels to metals. These metals are easier to reprocess and specifically recovered elements can be used directly in advanced civilian nuclear reactors and metallic fuel cycles. This thesis considers the role that an FFC based process could play in establishing a more sustainable, efficient and safer method for the select recovery of key metals from mixed oxide composites. Using an array of surrogate materials, the appropriation of an effective procedure was investigated in both CaCl2 and LiCl-CaCl2 Eutectic (LCE) molten salts at 810oC and 600oC respectively. At all stages of the process, feeds and electrolysis products have been examined by an assortment of ex-situ analytical tools, primarily SEM, EDS and XRD. A process engineering approach was taken to designing suitable reactors and cells with the aim of improving operational characteristics, greater electrochemical reduction efficiency and high yields of pure products. Preparation of electrolyte and feed oxide electrodes (surrogate or spent fuel) was investigated, including unique electrochemical treatment for the two molten salts and precedent for the creation of the oxide electrode via cold pressing or slip casting to kinetically aid optimal reduction. A series of investigations considering the thermodynamic performance of CaCl2 from a standpoint of electronic conduction were carried out, and considerable improvements found via the implementation of a simple cathodic sheath. Selective partitioning was shown possible by the intended mechanism of partial direct reduction and anodic dissolution in the 2NiO-CeO2 binary. Partitioning of Zr from ZrO2-CeO2, and Ti from TiO2-CeO2 was also achieved, however in both cases it was via the gradual chemical dissolution of partially reduced Ce(III) into the molten salt or phase separation between liquid Ce and solid Zr. Extensive CV experiments were performed to enhance understanding of redox chemistry for each species investigated. CeOCl was found to be the only semi-stable phase of Ce present at potentials between -1.0 V vs. Ag/AgCl and its final reduction potential at approximately -1.95 V in CaCl2 at 810oC. Active CV experiments using PuO2 and a MOX fuel sample containing 5% PuO2 were initiated, revealing remarkably similar electrochemical behaviour of PuO2 and the CeO2 surrogate. Both PuO2 and the bulk UO2 content MOX could be reduced in CaCl2 and in the lower temperature LCE whilst avoiding any decomposition of the electrolyte. Consequently a route for the direct electrochemical reduction of spent oxides fuels was shown plausible and offers a promising alternative to current pyroprocessing technology, with beneficial implications to the wider materials processing field.