We have developed a QCM (Quartz Crystal Microbalance) based method for direct gravimetric determination of water adsorption on PuO2 surrogate surfaces, especially CeO2, under conditions representative of those in a typical PuO2 storage can. In this application, the method of transduction of the QCM relies upon the linear relationship between the resonant frequency of piezoelectrically active quartz crystals and the mass adsorbed on the crystal surface. The spurious effect of high temperatures on the resonant frequency of coated QCM crystals has been compensated for by modeling the temperature dependence of the frequency response of the surrogate coated-QCM crystal in the absence of water. Preliminary results indicate that water is readily adsorbed from the vapor phase into porous metal oxide structures by capillary condensation, an observation that may have important ramifications for water uptake within the packed powder beds that may obtain in PuO2 storage cans.

Alloy 617, a high-temperature creep-resistant, nickel-based alloy, is being considered for the primary heat exchanger for the next generation nuclear plant, which is highly influenced by thermal creep. The main objective of this study is to inspect the crept grain boundaries under its imitated working condition, and to determine which boundaries are susceptible to damage and which are more resistant, in order to help improve its creep resistance in future manufacturing. Electron backscatter diffraction was used to measure the proportions of each boundary by observing and analyzing these crept microstructures. The grain-boundary distribution can be expressed in terms of five parameters including three parameters of lattice misorientation and two parameters of the grain-boundary plane normal. Three conditions were analyzed: the original material, metal that was annealed without stress, and ones that were crept at 1000ºC at 19 MPa and 25 MPa for various times. Though observation, it is found that the voids seldom occur at low angle grain boundaries, and coherent twin boundaries are also resist to creep damage.

Graphite is used as a moderator and structural component in the United Kingdom’s fleet of Advanced Gas-Cooled Reactors (AGRs) and features in two Generation IV reactor concepts: the Very High Temperature Reactor (VHTR) and the Molten Salt Reactor (MSR). Under the temperature and neutron irradiation conditions of an AGR, nuclear-grade graphite demonstrates significant changes to it mechanical, thermal and electrical properties. These changes include considerable dimensional change with expansion in the c-direction and contraction in the a/b-directions. As the United Kingdom’s AGRs approach their scheduled decommissioning dates, it is essential that this behaviour be understood in order to determine under what reactor conditions their operating lifetimes can be safely extended.

Two models have been proposed for the dimensional change in graphite due to displacing radiation: the “Standard Model” and “Ruck and Tuck”. The Standard Model draws on a conventional model of Frenkel pair production, point defect migration and agglomeration but fails to explain several key experimental observations. The Ruck and Tuck model has been proposed by M.I. Heggie et al. and is based upon the movement of basal dislocation to create folds in the “graphene” sheets and seeks not only to account for the dimension change but also the other phenomena not explained by the Standard Model.

In order to test the validity of these models, work is underway to gather experimental evidence of the microstructural evolution of graphite under displacing radiation. One of the primary techniques for this is transmission electron microscopy with in situ ion irradiation. This paper presents the results of electron irradiation at a range of energies (performed in order to separate the effects of the electron and ion beams) and of combined electron and ion beam irradiation.

The feasibility of the fabrication of tungsten based nuclear fuel cermets via Spark Plasma Sintering (SPS) is investigated in this work. CeO2 is used to simulate fuel loadings of UO2 or Mixed-Oxide (MOX) fuels within tungsten-based cermets due to the similar properties of these materials. This study shows that after a short time sintering, greater than 90 % density can be achieved, which is suitable to possess good strength as well as the ability to contain fission products. The mechanical properties and the densities of the samples are also investigated as functions of the applied pressures during the sintering.

Zirconium is of interest in the development of high-density diffusion barriers in nuclear reactor applications because of its low thermal neutron absorption cross section and good thermal and mechanical properties. It is used in these applications to protect the low enriched fuel from interacting directly with the cladding material to prevent swelling and cracking of the fuel. In this work, we investigated the use of plasma spraying as a method to produce high quality, dense Zr barrier coatings over large surface areas. A thin sheet of 21Cr-6Ni-9Mn stainless steel (SS) was used as the substrate material. Both sides of the SS sheet were coated, one side at a time. A transfer arc (TA) current between a torch and the substrate was used to vary the substrate temperature to explore the effect of temperature and time on the film grain size, interface quality, and film porosity. The films were characterized using light optical microscopy (LOM), scanning probe microscopy (SPM) and Kelvin probe force microscopy (KPFM) and EDS-SEM Although the film quality did improved with temperature, at the elevated substrate temperatures used in this study, metal atoms from the substrate diffused into and, at the highest temperature and longest time, through the Zr coatings.

From density functional theory (DFT) based ab initio (Car-Parrinello) metadynamics, we compute the activation energies and mechanisms of water exchange between the first and second hydration shells of aqueous Uranyl (UO22+) using the primary hydration number of U as the reaction coordinate. The free energy and activation barrier of the water dissociation reaction [UO2(OH2)5]2+(aq) → [UO2(OH2)4]2+(aq) + H2O are 0.7 kcal and 4.7 kcal/mol respectively. The free energy is in good agreement with previous theoretical (-2.7 to +1.2 kcal/mol) and experimental (0.5 to 2.2 kcal/mol) data. The associative reaction [UO2(OH2)5]2+(aq) + H2O → [UO2(OH2)6]2+(aq) is short-lived with a free energy and activation barrier of +7.9 kcal/mol and +8.9 kca/mol respectively; it is therefore classified as associative-interchange. On the basis of the free energy differences and activation barriers, we predict that the dominant exchange mechanism between [UO2(OH2)5]2+(aq) and bulk water is dissociative.

Currently in the nuclear industry, surface contamination in the form of radioactive metal or metal oxide deposits is most commonly removed by chemical decontamination, electrochemical decontamination or physical attrition. Physical attrition techniques are generally used on structural materials (concrete, plaster), with (electro)chemical methods being used to decontaminate metallic or painted surfaces. The most common types of (electro)chemical decontamination are the use of simple mineral acids such as nitric acid or cerium (IV) oxidation (MEDOC). Use of both of these reagents frequently results in the dissolution of a layer of the substrate surface increasing the amount of secondary waste which leads to greater burden on downstream effluent treatment and waste management plants. In this context, both mineral acids and MEDOC can be indiscriminate in the surfaces attacked during deployment, e.g. attacking in transit through a pipe system to the site of contamination resulting in both diminished effect of the decontaminating reagent upon arrival at its target site and an increased secondary waste management requirement. This provides two main requirements for a more ideal decontamination reagent: Improved area specificity and a dissolution power equal to or greater than the previously mentioned current decontaminants.

Photochemically promoted processes may provide such a decontamination technique. Photochemical reduction of metal ion valence states to aid in heavy metal deposition has already been extensively studied, with reductive manipulation also being achieved with uranium and plutonium simulants (Ce). Importantly photooxidation of a variety of solution phase metals, including neptunium, has also been achieved. Here we briefly review existing decontamination techniques and report on the potential application of photo promoted oxidation technologies to metal dissolution (including process steels) and to the dissolution of adsorbed actinide contaminants.

Oxides with fluorite (or fluorite related) structures form a large class of compounds with a high radiation tolerance, somewhat related to their peculiar ability to accommodate a variety of defects and to form nonstoichiometric compounds with a large homogeneity range. Structural modifications are generally observed when the departure from the ideal composition is large. We discuss these structural features using an approach based on the crystal symmetry analysis based on the phase transition mechanisms in compounds relevant for nuclear applications.

The experimental investigation of actinide materials like nuclear fuels is difficult and usually very costly. Therefore a reliable multi-scale modeling of these often hazardous materials starting at the atomistic level is inevitable to gain further insight into this type of materials. The development of new, more advanced simulation methods accompanied by the rapid growth of the available computational resources provided by high-performance computing facilities, allows the modeling of such materials at a new quality level. Also the recent development of the CP2K program package (http://www.cp2k.org) has been partially focused on enabling state-of-the-art simulations of actinide materials using classical potential as well as electronic structure methods. The long-term goal is to perform reliable molecular dynamics simulations for actinide materials including advanced simulation techniques like nudged elastic band or metadynamics simulations. In this work, the CP2K program package and its application to the simulation of defect migration in uranium dioxide (UO2) using the nudged elastic band method is presented.

During the lifetime of a nuclear facility, radioactive material may become deposited onto process and structural material surfaces. Due to their high corrosion resistance, steels comprise the largest class of metal-based materials encountered on nuclear sites. A greater understanding of the mechanisms of how contaminant radionuclides interact with and attach to process steels in nuclear plant environments is required in order to enable informed decisions to be made about the design and effective application of decontamination techniques, reducing secondary wastes.

There is limited literature relating to radionuclide sorption mechanisms on steels. Key studies have found that sorbed contamination is almost entirely located in the outermost oxide layers formed at steel surfaces. Thus, a molecular level investigation of contaminant uptake during induced oxide formation would be beneficial in developing steel decontamination strategies.

Stainless steel 316L is commonly employed in the nuclear industry in process streams and pipework. Thus, we describe work carried out on electrochemically accelerated oxide growth on 316L and SS2343 (a 316L analog) in nitric acid media and its characterisation using combined voltammetric and microgravimetric measurements. These allow identification of active, passive, high voltage passive, transpassive and secondary passivation regimes in the associated current voltage curves. EQCM on SS2343 coated quartz crystal piezoelectrodes, combined with potentiodynamic polarisation data have allowed us to determine that fastest net growth of surface oxide occurs in the low voltage passive regime. Further, we have directly measured the growth of that layer by using in situ microgravimetry for the first time. We will be shortly using the methods described above and radionuclide surrogates for the study of contaminant uptake during oxide formation and uptake onto preformed oxide layers. XPS will be used to determine layer composition and mode of contaminant uptake.

Simple Hydroxamic acids (XHAs) are salt free, organic compounds with affinities for hard cations such as Fe3+, Np4+, Pu4+ and have been identified as suitable reagents for the control of Pu and Np in advanced nuclear fuel reprocessing. Building upon previous work on the neptunium(IV)-formohydroxamic(FHA) acid system [1], a model that describes the hydrolysis of the acetohydroxamate moiety has been extended to include hydrolysis of bishydroxamatoneptunium(IV) complex. The model has been used to determine the rate constants for hydrolysis of mono- and bis-acetohydroxamatoneptunium(IV) at 25 °C, which were found to be 1.0×10-5 dm3 mol-1 s-1 and 5.0×10-5 dm3 mol-1 s-1, respectively.

Predicting the microstructural evolution of radiation damage in materials requires handling the physics of infrequent-events, in which several time scales are involved. The reactions rates characterizing these events are the main ingredient for simulating the kinetics of materials under irradiation over large time scales and high irradiation doses. We propose here an efficient, finite temperature method to compute reaction rate constants of thermally activated processes. The method consists of two steps. Firstly, rare reactive trajectories in phase-space are sampled using a transition path sampling (TPS) algorithm supplemented with a local Lyapunov bias favoring diverging trajectories. This enables the system to visit transition regions separating stable configurations more often, and thus enhances the probability of observing transitions between stable states during relatively short simulations. Secondly, reaction constants are estimated from the unbiased fraction of reactive trajectories, yielded by an appropriate statistical data analysis tool, the multistate Bennett acceptance ratio (MBAR) package. We apply our method to the calculation of reaction rates for vacancy and di-vacancy migration in α-Iron crystal, using an Embedded Atom Model potential, for temperatures ranging from 300 K to 800 K.

Ferritic/martensitic steels such as HT9 steel, is used for structural components in nuclear power plants because of its high strength and good swelling resistance. Understanding the mechanical behavior of these steels is quite important, since it will affect the strength and the life of the component. In this study, a dislocation density based crystal plasticity finite element model is developed in which different types of dislocation evolves on the activated 12 slip systems in alpha-iron. The dislocation evolves in the form of closed loop and the dislocation density is tracked as internal state variable, the generation and annihilation of dislocations are modeled based on the dislocation interaction laws. The plastic flow is calculated based on the dislocation densities and a generalized Taylor equation is used as the hardening law, and the hardening is assumed to be isotropic in this study. The evolution of polycrystal texture of alpha-iron is presented in the form of pole figures, which indicate the orientation spread and agree with the experimental result. The model also indicates the inhomogeneous dislocation distribution and stress concentration at the grain boundaries.

Uranium dioxide is the most common fuel used in commercial light water nuclear reactors. The fission of the fuel generates fission products (FPs) and minor actinides (MAs), which affects the thermo-physical properties of the fuel. The understanding of the physical and chemical properties of the FPs and MAs is still limited. In this study we have used atomic level simulations to estimate the effect of Ce4+ and Th4+ ions in urania matrix. Our results show that the structural variation depends on the elastic effect, which is guided by the ionic radius of the substituted ion. Ce4+ (ionic radius 0.97 Å) reduces the overall lattice parameter, while Th4+ (ionic radius 1.05 Å) increases the overall lattice parameter of the urania matrix (U4+ ionic radius 1.00 Å). In addition bulk modulus of the U1-xCexO2 system does not change with substitution while the modulus of U1-xThxO2 reduces with an increase in Th4+ ion concentration. This observation is in accordance with Vegard’s law prediction based on the modulus values of bulk UO2, CeO2 and ThO2 systems.

In situ x-ray absorption spectroscopy (XAS) measurements were made on Fe3O4 nanoparticles in supercritical aqueous fluids to 500 °C in order to study their reactivity with Co2+ aqua ions and to investigate the structural properties of the reacted nanoparticles. The analyses of the x-ray absorption near edge structure (XANES) of XAS indicate that reactivity of Fe3O4 nanoparticles with Co2+ ions is minimal to 200 °C but becomes significant in the 250–500 °C temperature range. XANES and angular momentum projected density of states (l-DOS) calculations were carried out using the FEFF8.2 code and analyses were made using multi-peak fitting to determine the origin of the features exhibited in the spectra.

Changes in the material properties of copper (II) phthalocyanine (CuPc) thin-films were studied upon exposure to increasing dose of ionizing radiation using photoluminescence spectrum. We observe generation of new energy states below the band gap upon exposure to ionizing radiation. Organic electronic devices – CuPc based resistor and an organic field effect transistor (OFET) – are proposed in this work as total dose sensors for ionizing radiation. We observe an increase in the conductivity of CuPc thin-films with increasing dose of ionizing radiation. To overcome the possibility of changes/degradation in the electrical properties of CuPc thin-films upon interaction with various gases and moisture in the environment, a passivation layer of silicon nitride, deposited by hot-wire CVD process is proposed. Effect of ionizing radiation on the electrical properties of thin-films of CuPc has been studied. We observe a 170% increase in the resistance of the thin-film for a total of 50 Gy radiation dose using Cobalt-60 (60Co) radiation source. Moreover, significant changes in the electrical characteristics of an OFET, with CuPc as an organic semiconductor, have been observed with increasing doses of ionizing radiation. Experiments with an OFET (W/L = 19350 μm / 100 μm and tox = 150 nm) as a sensor resulted in a ∼100X change in the OFF current for a total of 50 Gy dose of ionizing radiation exhibiting a sensitivity of ∼1 nA/Gy. Moreover, implementing a reader circuit, shift in the threshold voltage of the OFET at 1e-7 A drain current displayed a sensitivity of 80 mV/Gy for a total of 50 Gy dose of ionizing radiation. CuPc based organic electronic devices have advantages as sensors because of their low-cost fabrication, large area coverage on flexible substrates, etc.

Models of radiation growth proposed to date are all based on the assumption that the primary damage is produced by neutron irradiation in the form of single defects. These models do not account for the features of the cascade damage: intra-cascade clustering of self‑interstitial atoms (SIAs) and their one‑dimensional diffusion. During the last twenty years, a ‘Production Bias Model’ has been developed, which shows that the damage accumulation in cubic metals depends crucially on the cascade properties. The cascades in hcp zirconium are similar to those in cubic crystals; hence the model can provide a realistic framework for the hcp metals as well. In this work we present such a model in application to low-temperature (below 300°C) radiation growth.

Ceramics based on the general compositions CaLnXNbO7 (where Ln = La, Nd and Sm, and X=Zr and Sn) have been prepared, and irradiated with 1 MeV Kr ions at the IVEM-TANDEM user facility. The radiation tolerance of these materials has been found to be less than Zr and Hf equivalents. The results also suggest that the amorphisation cross section for these materials is related to the Ln component, and is similar to those observed for Zr and Hf equivalents.

Pentacene and poly 3-hexylthiophene (P3HT) are the most promising p-type organic semiconducting materials for fabrication of organic field effect transistors (OFETs). OFETs with aforesaid organic semiconducting materials have been demonstrated as total dose detectors for ionizing radiation, wherein the changes in the electrical characteristic parameters, such as, increase in the OFF current, increase in the ON current, change in the current ratio, shift in the threshold voltage, change in the subthreshold swing, etc., were used as a measure of ionizing radiation dose. Upon exposure to ionizing radiation P3HT based OFET sensor has shown an OFF current sensitivity of 4.4 nA/Gy while pentacene based OFET sensor has shown an OFF current sensitivity of 26.7 nA/Gy for a total of 50 Gy dose of ionizing radiation. Change in the conductivity of the thin-films of pentacene and P3HT were observed and compared using electrostatic force microscopy (EFM) imaging before and after exposure to ionizing radiation. Effects of ionizing radiation on the energy band structures of the organic semiconducting materials, pentacene and P3HT, have been studied using UV-visible spectroscopy. Moreover, analysis of UV-visible spectra of the thin-films suggested the generation of energy states in larger quantity in case of pentacene thin-film as compared to P3HT thin-film upon exposure to the same dose of ionizing radiation. These results confirm the higher sensitivity observed in pentacene OFET sensor as compared to P3HT OFET sensor in terms of the change in electrical parameters.