Symposium EN17 : Structure–Property Processing Performance Relationships in Materials for Nuclear Technologies

While fuels based on thoria have been considered as alternatives for urania for decades, there is considerably less data concerning the accommodation of fission products at the atomic scale. Here atomic scale computer simulation based on density functional theory is used to predict the energies and accommodation sites for a range of fission products and He in stoichiometric ThO₂ and hypo-stoichiometric ThO_2-x.

Nine sites were considered including point defects and vacancy clusters with multiple species. Neutral and all possible defect charge states up to full formal charge were investigated. For Xe and Kr, in ThO_2-x the most favourable solution equilibrium site is a neutral tri-vacancy while in ThO₂ it is the di-vacancy. The most favourable solution site for I and Br is the single positively charged oxygen vacancy in ThO_2-x while in ThO₂, I demonstrates the same solubility in all clusters. Rb, Cs, Sr, Ba, Y and Zr are accommodated at the thorium vacancy but the charge state of the vacancy is important. Finally for He, a relationship is generated that describes the incorporation energy of the x^th He atom, E_x (n, m), into a cluster consisting of n thorium vacancies and m oxygen vacancies.

Oxide nuclear fuels have been widely used in light water reactors. The thermal conductivity of nuclear fuels is closely related to energy conversion efficiency as well as reactor safety margins. Understanding the mechanisms that cause the degradation of thermal conductivity in a high radiation environment is important for the design and development of new high-burnup fuels. For oxide nuclear fuels, phonon scattering by point defects, extended defects such as dislocation loops and bubbles, and grain boundaries plays a significant role in limiting the thermal transport properties. ThO₂ is an actinide-bearing material that does not contain 5f electrons and will serve as a baseline that enables the investigation of phonon transport mechanisms in defective fuel without the effects of 5f electron. In this work, defect evolution in both single crystal and polycrystalline ThO₂ has been studied by a combination of in situ ion irradiation and modeling. In addition, the effects of point defects and extended defects on the phonon transport in ThO₂ have performed on ion irradiated samples. The microstructure of pristine and ion irradiated ThO₂ has been characterized by using electron backscatter diffraction (EBSD), scanning transmission electron microscope (S/TEM), and time-domain Brillouin scattering (TDBS) techniques. The thermal conductivity before and after irradiation has been determined using laser-based modulated thermoreflectance (MTR) technique.

Numerous experimental studies have investigated the melting point of mixed oxides in the ternary UO₂–ThO₂–PuO₂ system, with significant discrepancies. In particular, a few studies have reported anomalous reduction in the melting point of mixed oxide, below that of the end members, which have variously been attributed to uncontrolled deviations of stoichiometry and limitations in the experimental setup. Here we provide a mechanistic understanding of the peculiar melting behaviour using molecular dynamics simulations, which are inherently free of the above limitations, combined with a robust pair-potential and a novel technique for calculating solidus and liquidus lines. We show that a dip in melting temperatures is indeed observed in the solidus and liquidus of the mixed oxides (Th,U)O₂, (Th,Pu)O₂ and (Pu,U)O₂ and that this is an inherent property of the solid solutions. This dip is found at ∼5% additions of the oxide with higher melting point, in agreement with experimental observations. We propose that the root cause for this reduction in melting point is the formation of low-energy Frenkel pairs in the vicinity of small amounts of dissimilar cations of larger ionic radius: the random (and therefore locally heterogeneous) distribution of cations leads to a spread of Frenkel pair formation energies, depending on the local environment around the vacancy and interstitial. Overall this results in an increase in the average Frenkel pair formation energy, but a small portion of Frenkel configurations exhibit a significantly lower formation energy, which may act nucleation sites for early-onset melting.

As a part of EFRC Project titled ‘TETI: Thermal Energy Transport under Irradiation,’ the impact of irradiation induced defects in ThO₂ and U-doped ThO₂ is investigated. Prior to irradiation, the defect content of this material is tied to off-stoichiometry and hence it is dependent upon the U doping level and the temperature and external oxygen pressure or chemical potential. Under irradiation, non-equilibrium defects are produced by energetic particle bombardment, leading to the formation of nanoscale and sub-nanoscale defect clusters that influence thermal transport in ways that are not fully understood to the community. In this presentation we discuss our modeling results for the equilibrium and non-equilibrium defects in ThO₂ and U-doped ThO₂. A thermodynamic defect disorder model was used to investigate the off-stoichiometry behavior in ThO₂ and U-doped ThO₂ in an oxygen partial pressure and temperature environment. The model shows that while ThO₂ remains mostly hypo stoichiometric at relevant thermodynamic conditions, U doping expands the thermodynamics window over which ThO₂ becomes hyper-stoichiometric, thus illustrating the impact of 5f electrons introduced by the U doping on defect disorder. For example, U_0.52Th_0.48O_2+x can sustain disorder up to an oxygen off-stoichiometry content of x = 0.4 at high temperatures, up to 2000 K, and high oxygen pressures, up to 0.316 atm, whereas U_0.2Th_0.8O_2+x can only exist up to an oxygen off-stoichiometry content of x = 0.05. The extent of hyper-stoichiometry in the oxide reduces with increasing temperature and decreasing oxygen partial pressure values. We will also present the results of a recent cluster dynamics model of vacancy and interstitial cluster formation in ThO₂ under irradiation in conjunction with microstructure and chemical analysis of pure and ion-irradiated ThO₂ at different temperatures by EELS imaging technique in TEM. The impact of cluster composition on matrix composition is illustrated both computationally and experimentally, showing that irradiation provides a further mechanism of altering the stoichiometry of the matrix by forming clusters not commiserating with the chemical formula of the crystal.

Since the Fukushima accident in 2011 there has been a large international effort in the development of accident tolerant fuel and cladding materials for water cooled reactors [1]. Uranium silicide intermetallics are being considered as accident tolerant fuels (ATF) to replace UO₂ currently used in light water reactors (LWRs) [2]. Primarily U₃Si₂ is being examined due to its good thermal conductivity (~15 W/m/K at 500°C) [3], higher U metal density, (giving economic advantages by requiring lower ²³⁵U enrichments) and reduced radiation swelling compared to the higher U containing silicides such as U₃Si [4,5]. However, little is known on the oxidation behaviour of U₃Si₂ under high temperature air and steam environments which could be experienced under a loss of coolant accident (LOCA) scenario with water and/or air ingress into the fuel pin.

The oxidation process of Ce₃Si₂ (as a surrogate for U₃Si₂) has been studied to develop an oxidation mechanism. Samples were oxidised under flowing air to 750°C finding discrepancies between the terminal and theoretical mass gains for complete oxidation similar to previous works on the U₃Si₂ system. As oxidised materials were studied using scanning transmission electron microscopy (S/TEM) with energy dispersive x-ray (EDS) and energy filtered (EF)TEM mapping to elucidate the reaction products and develop a mechanism. The fate of the Ce and Si in the system will be presented and mechanistically discussed and discrepancy in the differences in mass gains revealed. The resulting composition and structure of the as-oxidised materials will also be presented, discussing potential consequences on mechanical stability of the fuel material under a LOCA scenario. The oxidation mechanism, comparison with the U-Si system and consequence of the resulting material structure on fuel performance under a LOCA will also be discussed.

[1] B.A. Pint, K.A. Terrani, Y. Yamamoto, L.L. Snead, Metall. Mater. Trans. E 2 (2015) 190–196.
[2] J.M. Harp, P.A. Lessing, R.E. Hoggan, J. Nucl. Mater. 466 (2015) 728–738.
[3] J.T. White, A.T. Nelson, J.T. Dunwoody, D.D. Byler, D.J. Safarik, K.J. McClellan, J. Nucl. Mater. 464 (2015) 275–280.
[4] I.J. Hastings, R.L. Stoute, J. Nucl. Mater. 37 (1970) 295–302.
[5] I.J. Hastings, J.R. MacEwan, L.R. Bourque, J. Am. Ceram. Soc. 55 (1972) 240–242.

The use of uranium-silicide compounds as potential nuclear reactor fuels requires knowledge of thermophysical properties as a function of temperature. At elevated temperatures, a great deal of research has been performed to understand the thermodynamic properties of U-Si compounds; however, in the low-temperature regime, only limited data are available [1-4]. Operating temperatures of nuclear reactors are high, yet important thermodynamic quantities such as Debye temperature and standard entropy (S°) can be derived from low-temperature measurements. S° is an imperative quantity in determination of the Gibbs free energy which depends on S° and enthalpy. Among the U-Si compounds, stoichiometric USi has typically received significantly less attention in the literature with regards to characterization; however, uranium-silicide phases used as nuclear fuels may include appreciable fractions of USi either introduced through fuel synthesis or fabrication. The potential of USi both as a nuclear fuel itself and as a fraction of other U-Si samples, implies a need to further the thermophysical understanding of this material.
Here we present investigations of the low-temperature thermal properties of USi by measuring its specific heat capacities from 2.4 to 398 K using a Quantum Design Physical Properties Measurement System (QD-PPMS). A smooth curve is observed over the entire temperature range measured, and the specific heat capacity at the maximum temperature tested (398 K) was determined to be 67.2 J mol^-1 K^-1. Additionally, the Debye temperature has been determined from fitting the low-temperature (below 30 K) specific heat capacity data as 252.3 K. Furthermore, by performing integration of the specific heat capacity over the temperature range, the standard entropy at room temperature has been determined as 88.5 J mol^-1 K^-1.
[1] J. T. White et al., J. Nucl. Mater., 2015, 456, 442-448.
[2] J.T. White et al., J. Nucl. Mater., 2016, 471, 129-135.
[3] J.T. White et al. J. Nucl. Mater., 2015, 464, 275-280.
[4] D.J. Antonio et al. J. Nucl. Mater. 2018, 508, 154-158.

The U₃Si₂ compound has a high uranium density and thermal conductivity compared to standard UO₂ fuel and, for these reasons, it is being considered as a possible replacement fuel in the current fleet of light water nuclear reactors. It could bring significant performance and economic benefits, while the exothermic reaction with the coolant upon cladding breach is a known drawback that needs to be addressed. Material properties such as the diffusion rate of fission gas atoms and point defects must be determined in order to model the in-reactor behaviour of U₃Si₂. Here we use density functional theory (DFT) calculations and empirical potential simulations to study diffusion of Xe and point defects in bulk U₃Si₂ as well as at grain boundaries. The DFT calculations apply the GGA+U methodology to calculate defect energies, entropies and migration barriers of point defects as well as of Xe atoms interacting with point defects. Xe and point defect diffusion rates are predicted from a point defect model parameterized by results from DFT calculations. The diffusion rate under irradiation due to ballistic damage is derived from molecular dynamics simulations based on a MEAM potential for the U-Si-Xe interactions. Molecular dynamics simulations are also used for investigating grain boundary diffusion and segregation of point defects and Xe atoms. We find that Xe diffusion in U₃Si₂ is faster than in conventional UO₂ nuclear fuel in the intrinsic regime, but lower in the regime where transport is dominated by ballistic mixing. The latter observation is a consequence of the high thermal conductivity of U₃Si₂ preventing high-temperature thermal spikes, which leads to transport by ballistic mechanisms being dominant. The impact of the diffusion properties on fission gas release and swelling in U₃Si₂ is investigated by parametrizing a fuel performance model in the Bison code with results from the atomic scale simulations. The results are compared to recent irradiation experiments available in the literature. Finally, we discuss the importance of bulk and grain boundary diffusion on creep rates in U₃S₂.

Accident Tolerant Fuel (ATF) concepts aim at providing radical performance and safety improvements for the Light Water Reactors (LWR) at economically attractive conditions. Coatings as near-term evolutionary option take advantage of the reliability of traditional Zirconium-based cladding materials for structural purposes, while conferring engineering of the surface properties. In particular, they are expected to reduce corrosion rates, retain structural integrity and adhesion, minimize hydrogen uptake by the cladding matrix and improve high temperature oxidation resistance.
In this work, the process of selection, design, production and testing of ceramic oxides as protective coatings for LWR fuel cladding is presented. It is worth mentioning that all these activities have been performed in the framework of the European project IL TROVATORE. In the first stage, candidate materials are selected by matching neutronic requirements and chemical stability in pressurized-water environment. This phase is followed by the production of ceramic oxides, in the form of sintered pellets or as thin films by the Pulsed Laser Deposition (PLD) technique. Specimens are then exposed to Pressurized Water Reactor (PWR) environment. Specifically, compatibility tests are performed in pure water at 360°C and saturation pressure, in order to simulate the PWR core during normal operating conditions. Mass changes are monitored during the exposure, then correlated to modifications on the surface and in the structure of the tested samples. The specific dissolution and oxidation processes related to the measured mass changes are also investigated. This initial phase allows us to simultaneously evaluate the behaviour of candidate materials as bulk oxides and coatings, hence confirming the feasibility of PLD as a viable technique to produce high-performance ceramic films, with strong adhesion and interfacial bonding. Eventually the screening phase leads us to the selection of best candidate oxide to uptake an additional phase of optimization of the coating fabrication process.
The selected material is grown on relevant substrates such as Zirconium-based alloys and AISI 316L via an optimized PLD process. The exposure of the new coated specimens to the simulated PWR ambient is performed, up to 30 days in saturated pure water at 360°C. The analysis of the mass change confirms that the coatings act efficiently as barriers against oxidation by reducing significantly the oxidation of the metal alloy. In addition, structural integrity and good adhesion to the substrates are confirmed by spectroscopic characterization and electron microscopy.
Furthermore, the improvement of tolerance against severe accidents is considered with respect to the specific case of Loss Of Coolant Accident (LOCA). Preliminary tests at high temperature (up to 1000°C) and in a controlled atmosphere are conducted in order to evaluate the performance of the selected ceramic material during accidental transients. The coatings show structural integrity and good adhesion, suggesting the high capability of protecting the underlying alloy and their potential for a relevant delay of uncontrolled oxidation of the claddings during accidental transients. On the other hand, the importance of correct alignment of the thermal expansion coefficients between coating and substrates becomes evident at higher temperatures (≥1200°C).
In conclusion, the overall results collected show that PLD ceramic coatings are capable of sustaining both normal and off-normal reactor conditions and consequently constitute a promising option for coated-cladding ATF concepts.

Uranium mononitride (UN) is a proposed nuclear fuel for upcoming generation IV nuclear reactors, due to its higher thermal conductivity, greater melting point, and higher concentration of fissile material compared to conventional oxide fuels DFT+U is used to capture the anti-ferromagnetic ground state of UN. However, the +U correction leads to the possibility of convergence to metastable states, thereby disrupting the correct predictions of electronic-level properties, particularly point-defect energies. In this work, we provide a comprehensive analysis of ground and metastable states and the resulting defect properties using three methods, namely occupation matrix control (OMC), quasi-annealing (QA), and U-ramping. We preform each of these calculations using both the Dudarev and Liechtenstein rotationally-invariant forms of DFT+U. We find that QA and OMC reproduce the ground state, whereas U-ramping converges to metastable states as previously observed. All three methods predict the nitrogen interstitial to be the most favorable defect, whereas the uranium interstitial is the least favorable defect among the vacancies and interstitials of both elements. For example, the calculated formation energies of defects calculated using the QA method across the Dudarev form are 1.76 eV and 9.38 eV for the U vacancy and interstitial, respectively, and for the N vacancy and interstitial are 3.86 eV and 0.94 eV , respectively. The migration barriers for both the U and N vacancies are similar, i.e., ~ 3 eV. Finally, our results show there are significant deviations in defect formation energies among various metastable states.

Next generation nuclear systems are meant to outperform current ones, by providing disruptive solutions in terms of non-proliferation, fuel cycle efficiency, radioactive waste management, safety and economics. However, the real occurrence of this scenario is directly linked to the availability of suitable materials for the more demanding conditions of these advanced plant concept, in terms of higher burnup and operating temperature, intense radiation fields, liquid metal or molten salt corrosion and tritium permeation. Interposing an engineered surface layer between qualified structural materials and the reactor environment makes this scenario reachable in the short term. In the last years our group developed multifunctional nanoceramic coatings by Pulsed Laser Deposition and Atomic Layer Deposition for Accident Tolerant Fuel for Light Water Reactors and as a solution to high-temperature Heavy Liquid Metal (HLM) corrosion in Generation-IV (GIV) and fusion systems. Accident Tolerant Fuel (ATF) concepts are developed by depositing on ZIRLO® and Zircaloy-4 suitable nanoceramic coatings, which are found to minimize hydrogen uptake and to improve high temperature oxidation resistance. On the other hand, in respect to GIV and fusion systems, Al₂O₃ and Y₂O₃ on AISI316, 1515-Ti and EUROFER-97, has been tested as anti-corrosion, radiation-resistant tritium permeation, insulating barriers. In particular, the compatibility in Pb and Pb-Li has been proven up to 10,000 hours. The tritium permeation reduction of these films is in the order of 10⁴-10⁵, well above the design requirements. Furthermore, Al₂O₃ coatings have been tested under heavy ions irradiation, at damage levels relevant for fission and fusion applications. The ceramic film under irradiation preserves its integrity and mechanical behaviour, evolving structurally from an amorphous to a crystalline state. To conclude, engineered coatings deposited by PLD/ALD techniques represent promising candidates to face the major issues related to future nuclear technologies and allow the design of innovative and economically attractive power plants.

Experiments show that recrystallization dramatically speeds up the gas bubble swelling kinetics in UMo nuclear fuels. This implies that gas bubbles inside the recrystallization zone, which has hundred nano-meter sized grains in diameter, grow much faster than those inside coarse grains. However, the mechanism of fast gas bubble growth is not well understood. In this work, a gas bubble evolution model integrating cluster dynamics and phase field models is developed to study the effect of vacancy and interstitial clustering on gas bubble evolution and volumetric swelling kinetics in the recrystallization zone. Cluster dynamics is used to describe the evolution of interstitial loops, vacancy clusters, and dislocation density. The phase field model is used to describe the gas bubble nucleation and growth. With this integrated model, the effect of defect generation rate, defect clustering rate, defect sink strength, and interstitial emission rate from interfaces on gas bubble evolution were systematically simulated. The comparison of gas bubble size, density and volumetric swelling from simulations and experimental data demonstrate that the nucleation and growth of interstitial loops in recrystallization zone is one of key mechanisms behind the fast swelling kinetics.

Irradiation induced grain refinement is a high burnup phenomenon associated with the fuel’s response to high strain fields created through in reactor irradiation. The increase in grain boundary area from the refinement increases the nucleation rate of fission gas bubbles which leads to a sharp increase in the swelling rate of the fuel. Understanding this phenomenon will allow for more accurate fuel life predictions and can lead to more efficient fuel design. The fission process creates multi-dimensional defects, including fission products, which strain the lattice. The defects produced by this process propagate through material creating both pinning points for dislocations and grain-boundaries, as well causing dislocation propagation by acting both as a source and a climb mechanism. This increases the internal energy of the material, which provides a driving force for recrystallization. Current models of this phenomenon approach the behavior sequentially; addressing grain nucleation first, then growth of the nuclei. This approach is useful but does not fully capture the temporal inhomogeneity of the system. By utilizing Cahn’s time cone method and adapting it to a temporally inhomogeneous reactor environment the nucleation and growth of the irradiation induced grain refinement can be modeled concurrently and can provide an analytical solution to the phenomenon. To address the temporal inhomogeneity of the system, strain which affects both the nucleation and growth was defined as a function of fission density. The mobility of the system was also adapted for irradiation enhanced diffusion. Heterogenous nucleation was used as the dominant nucleation mechanism. Utilizing this technique to model irradiation induced grain refinement showed comparable results to experimental observations demonstrating that the adapted Cahn time cone method can effectively model the phenomenon, while addressing both nucleation and growth simultaneously.

Monolithic U-Mo fuels have been down-selected for the conversion of several US high performance research and test reactors. In order to qualify this new fuel type, more than a hundred fuel plates have been fabricated and irradiated to date in the Advanced Test Reactor at Idaho National Laboratory. The size of fuel plates fabricated for these irradiation campaigns varies from 4 inches (mini-plates) to more than 40 inches (full-size plates) in length. The fabrication parameters used in fabricating mini-plates and full-size plates are very different, thus resulting in different as-fabricated microstructures. The microstructure of as-fabricated and irradiated mini-plates and full-size plates will be presented. The impact of the fabrication parameters on the microstructure and irradiation performance will be discussed.

The accurate description of the electronic and thermal properties of uranium-based compounds is fundamental in the development of novel nuclear fuels. Density functional theory calculations (DFT), within the plane wave pseudopotential approach using Quantum ESPRESSO (QE), has been applied to the study of uranium-based compounds. The lack of widely accessible pseudopotentials to perform DFT studies of uranium compounds significantly limits the capability of innovation in nuclear materials. As such, we have developed a set of projector augmented-wave (PAW) pseudopotentials for uranium, oxygen, nitrogen, and carbon atoms. The pseudopotentials were optimized to be used with a low kinetic energy cutoff of 37 Ry, which is considerably lower in comparison to PAW pseudopotentials available for QE. The set of pseudopotentials is then used to perform DFT calculations of uranium-based compounds, such as uranium metal, uranium oxides (UO₂, UO₃, and U₃O₈), uranium mononitride, and uranium monocarbide. Systems with strong electron correlations were described using Hubbard corrected DFT+U calculations. As such, monitoring the occupation matrix for 5f electrons of the uranium atoms was required to avoid metastable state solutions, and therefore to obtain reasonable ground properties. The reported results are shown to be accurate in comparison to reported theoretical and experimental results. This new set of PAW pseudopotentials are thus suitable for use in modelling a variety of uranium compounds, but also have the advantage of enhanced computational efficiency, due to the reduced kinetic energy cutoff.

There is limited information in the literature about defect clustering in irradiated alpha uranium in spite of the fact that defect clustering is responsible for irradiation growth of that material. The limited available data shows that two types of dislocation loops form in alpha uranium on the (010) and (100) planes, with Burgers vectors of $\sqrt{a^2+b^2} <110>/2$ and $a [100] $, respectively, under neutron irradiation. Under He ion irradiation, however, small cavities, which are assumed to be He bubbles were observed. We present a cluster dynamics model for loop formation in alpha uranium. The defect clusters considered at prismatic loops of interstitial type, those on (010) planes, and vacancy type, those on (100) planes. The dynamics model parameters were partly taken from open literature and partly fixed by molecular dynamics simulations. Cluster dynamics simulations were carried out and at different temperatures and dose rates, and the comparison with the existing neutron irradiation data for the loop size distribution and average loop size and density shows that the model is able to make reasonable predictions. We also present the results of proton irradiation of alpha uranium, where the proton beam was used to produce a microstructure dominated by dislocation loops. The formation and evolution of dislocation loops were studied by TEM technique, which provided further data for validation of the cluster dynamics model.

The majority of periodic table elements are created within the conventional uranium dioxide fuel matrix during its time in pile. These elements are accommodated in quite different ways within the nuclear ceramic. Ba, Zr, Y and the lanthanide atoms are typically substituted within the uranium dioxide matrix. At high burnup, the number of phases that form and the differences between them can result in the development of highly complex structures and microstructures. Similar effects might be encountered in manufactured advanced actinide fuels for the transmutation of minor actinides or in fuels containing significant amounts of burnable poisons (typically Gd). Those systems constitute then a large class of compounds, generally with fluorite or fluorite-derived structures, exhibiting high radiation tolerance, possibly related to their peculiar ability to accommodate a variety of defects and to exist as nonstoichiometric compounds within a large homogeneity range. Nevertheless, a variety of behaviours and different phases is observed after quenching or annealing in several of the peudo-binary systems of type (Ln,U)O_2-x, suggesting that the thermal history of those systems is of paramount importance. Some of these systems, like (U,Ce)O_2-x are particularly complex, display an intriguing structural variety, and therefore present a fundamental interest. In this talk we would try to address some of those interesting structural features induced by aliovalent substitutions in UO₂.

Advanced fuels are being considered for deployment in light water reactors (LWRs). Advanced fuels are designed to improve performance and fuel economics during normal operation and/or safety margins during accidents. For example, additives such as Cr₂O₃ can be used during fuel fabrication to modify the UO₂ microstructure. Specifically, the resultant enhanced grain size is considered to benefit pellet-clad interactions by modifying the mechanical properties of the fuel and by increasing fission gas retention, which reduces the plenum pressure. The effect of increased grain size on fission gas release can be readily captured within physics-based fission gas models, such as that implemented in the BISON fuel performance code. However, the dopant could modify the defect chemistry of UO₂ during reactor operation, thus altering the fission gas diffusivity and potentially undermining the enhanced fission gas retention obtain through enlarged grains. Fission gas diffusivity exhibits three diffusion regimes for reactor conditions: high T intrinsic diffusion, intermediate T irradiation enhanced diffusion, and low T athermal diffusion. In this work, we present a cluster dynamics model developed and validated for the description of irradiation enhanced diffusion in undoped UO₂. Large defect clusters play a crucial role in the reproduction of irradiation enhanced diffusion. For doped UO₂, it is assumed that the oxygen potential is controlled by the preferential reduction of Cr₂O₃ to Cr (over UO_2+x reduction to stoichiometric UO₂). The resultant doped UO₂ Xe diffusivity is enhanced with respect to undoped UO₂. The enhancement in Xe diffusivity relative to undoped UO₂ has been implemented in BISON. The predicted fission gas release from BISON has been compared to Halden experiments.

During its life in a reactor fission products build up in nuclear fuel. Some of these such as Xe, Kr and He form bubbles in the fuel. Whilst in a reactor, these have a detrimental effect on fuel performance as fission product bubbles may lead to swelling and a degradation of mechanical properties, and when released from the fuel, where the gas degrades the thermal conductance of the fuel-clad gap and leads to an increase in fuel temperature and cladding pressure. Understanding the size, shape and pressure of bubbles also allows the fission product inventory of nuclear fuel to be gauged. This information is important in developing improved fuel performance codes, pushing fuel to higher burn-up and has relevance to back-end processes such as reprocessing and ultimate disposal of spent fuel.

In this work, molecular dynamics simulations have been used to study fission product bubbles in UO₂. The results of this work will show the behaviour of small fission product bubbles in the range of diameters below <100nm. In particular bubble morphology is considered with both spherical and faceted Wulff shapes being compared. The effect of bubble internal pressure on morphology and its dependence on bubble size will be presented. The simulation results will be described in terms of the bubble size and pressure distribution expected in spent fuel.

The formation of high burnup structure (HBS) or rim structure is possibly the most significant restructuring processes at the rim of pellets in-pile with burnup extension in LWR and the effect of HBS on fuel thermo-physical or mechanical properties is a key requirement to ensure the successful implementation of extended burnup. There is independently experimental evidence to support that HBS concomitantly affects thermal and mechanical properties of fuel. But it is still not well understood how such HBS affects fuel thermo-physical or mechanical properties. In this work, we present results how HBS impacts the fission gas behavior of fuel rod under normal operating conditions. The fuel performance code FRAPCON-4.0 was used to simulate fission gas related properties, such as plenum pressure, fission gas release, and gap conductance. The effect of HBS under different boundary conditions, i.e. power history profiles, will also be discussed.

Fracture of UO₂ affects fission gas release in fuels. Predicting, understanding, and controlling within UO₂ fuel pellets is of considerable practical interest. An ongoing fundamental challenge is understanding the constituent contributions form grain boundaries and the bulk to the fracture response of the polycrystal. In this work, micro-scale mechanical testing has been used to induce single grain transgranular fracture and bicrystal intergranular fracture in samples isolated from polycrystalline UO₂. The results are discussed in the context of bulk UO₂ fracture and similar measurements performed on other oxides.

Tungsten has been chosen to cover the divertor and is envisaged for first wall material in fusion power reactors because of its high melting point, good thermal conductivity, low thermal expansion, high strength at high temperatures and high sputtering threshold energy. In such systems, tungsten will be submitted to neutron irradiation, high Helium and Hydrogen fluxes, and high heat fluxes up to 10 MW.m^-2 in stationary and up to 20 MW.m^-2 in transient operation and will have to sustain a temperature up to 1780 K. Such severe operating conditions could have a high impact on the macroscopic properties of the material, such as embrittlement and swelling.
In order to study and understand the evolution of the microstructure of tungsten in future fusion reactors such as ITER and DEMO, well-prepared tungsten samples were irradiated with W ions at 1.2 MeV and 20 MeV. Different damage doses in the range from 0.01 to 0.1 dpa are investigated and irradiation temperatures are fixed in the range from RT to 700°C. Positron annihilation spectroscopy (PAS) and Transmission Electron Microscopy (TEM) are used to characterize defects induced by irradiation. It allows to detect from the small vacancy clusters to large cavities and, for some conditions, the dislocation loops are also studied. The defects size, their number density and for the latter the nature of their Burgers vectors is determined.
The defects distribution is investigated as a function of the irradiation conditions and of the purity of the samples. Their evolution is then followed as a function of the damage dose, the irradiation temperature and also of the annealing temperature after irradiation. The results will be discussed in regard of the role of impurities on the microstructure evolution in the first stages of irradiation in tokamak and of the impact on the swelling of the material.


Acknowledgements:
This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission. Irradiations were performed at JANNuS (Joint Accelerators for Nanoscience and Nuclear Simulation) Orsay at CSNSM (France) and Saclay at CEA (France), which are part of the EMIR French accelerators network.

Void evolution during in-situ irradiation in polycrystalline W and W-5wt.%Ta material has been examined at 800 and 1000°C, via TEM. No dislocation loop formation was observed prior to void nucleation. The irradiated W microstructure was characterised by the presence of a population of voids, randomly distributed in the region. The number density of the voids reduces when the irradiation temperature is raised. In contrast, the excess of free vacancies in the W-5wt.%Ta material irradiated at 800°C only leads to the formation of visible voids after post-irradiation annealing of the sample at 1000°C. Solute Ta atoms also cause a significant decrease in the number density of voids when comparing the microstructure of both materials irradiated at 1000°C, and a gradual progression towards saturation at 0.2 dpa. Moreover, for the first time, we have detected a progressive transition from a spherical to a faceted shape in a number of voids present in both materials at damage levels 0.3 dpa.

A novel W-based refractory high entropy alloy with outstanding radiation resistance has been developed. The alloy was grown as thin films showing a bimodal grain size distribution in the nanocrystalline and ultrafine regimes and a unique 4 nm lamella-like structure revealed by atom probe tomography (APT). Transmission electron microscopy (TEM) and X-ray diffraction show an underlying body-centered cubic crystalline structure with certain black spots appearing after thermal annealing at elevated temperatures. Thorough analysis based on TEM and APT correlated the black spots with second phase particles rich in Cr and V. After both in situ and ex situ irradiation, these precipitates evolve to quasi-spherical particles with no sign of irradiation-created dislocation loops even after 8 dpa at either room temperature or 1073 K. Furthermore, nanomechanical testing shows a large hardness of 14 GPa in the as-deposited samples, with a slight increase after thermal annealing and almost negligible irradiation hardening. Theoretical modeling based on ab initio methodologies combined with Monte Carlo techniques predicts the formation of Cr and V rich second phase particles and points at equal mobilities of point defects as the origin of the exceptional radiation tolerance. The fact that these alloys are suitable for bulk production coupled with the exceptional radiation and mechanical properties makes them ideal structural materials for applications requiring extreme conditions.

Tungsten is the main candidate material for plasma-facingarmor components in future fusion reactors. During operation bombardment with high-energy fusion neutrons will create collision cascades that leave behind lattice-defects. Helium, injected from the plasma and produced bytransmutation, strongly binds to these defects, and modifies their behavior and retention. We investigate the impact ofhelium-implantation-induced and self-ion-implantation-induced damage on the deformation behavior of tungsten, comparing spherical nano-indents in unimplanted, helium-implanted and self-ion-implanted tungsten crystals of 001-orientation. Ion-implantation increases hardness and causeslarge pile-ups that increase with increasing implantation dose. Lattice rotations and indentation-induced residual strains beneath indents, probed using 3D-resolved synchrotron X-ray micro-diffraction, are smaller in the ion-implantedmaterial, suggesting a more confined plastic zone. The increase in pile-up points to a reduction in strain hardening capacity or even a strain softening. Based on these results, as well as TEM observations of defects beneath indents, we hypothesize that while dislocationmotionis initially obstructed by implantation-induced defects, the strength ofthese obstacles is reduced by the passage of dislocations. A 3D crystalplasticity finite element (CPFE) model founded on this hypothesis, with only three fitting parameters, reproduces these effects and is directly compared to the experiments. We also explore the orientation dependence of the indentation-response by considering grains with <001>, <110> and <111> out-of-plane orientations. Indents in unimplanted tungsten show little orientation dependence. In the helium-ion-implanted material, <110> and <111>-oriented grains show a much lower pile-up and hardness than <001>-oriented grains. The CPFE formulation, with all parameters unchanged except the crystal orientation, captures these orientation-dependent changes. The results suggest an orientation-independent mechanism governing the interaction of implantation-induced defects with glide dislocations, with differences in pile-up morphology arising just due to the relative orientations of the crystal slip systems, sample surface and spherical indenter tip.

As a tritium (T) production material, ⁶Li enriched tetragonal lithium aluminate (γ-LiAlO₂) has been used in support of the Tritium Sustainment Program. It also has been investigated as a candidate blanket material for fusion reactor designs. When γ-LiAlO₂ is irradiated with thermal neutrons, 2.75 MeV T and 2.05 MeV He particles from reaction ⁶Li (n, He) T will emit. These energetic particles collide with the lattice atoms and initiate damage cascades, generating point defects in the crystalline structure. The accumulation and interaction of point defects and gas species lead to the formation of extended defects and bubbles. In addition, radiolysis due to electronic energy deposition causes the material to decompose. The modified microstructure will affect T diffusion in the irradiated material.
In order to gain a physical insight into the defect production and gas species diffusion processes, ion irradiation experiments have been designed and performed in reactor-relevant conditions. Both protium and deuterium (D) have been used as surrogates for T in this study. Sequential irradiation with 90 keV He⁺ and 80 keV H₂⁺ ions as well as 120 keV He⁺ and 80 keV D₂⁺ ions with an equal ion fluence was performed over temperatures from 188 to 773 K. The ion energies were so chosen that the H or D profile peak was located at the He⁺ ion damage peak. The ion fluence ranged from 2×10¹⁶ to 4×10¹⁷ He⁺+H⁺ (or He⁺+D⁺)/cm². Isochronal and isothermal annealing experiments were followed up to 873 K and 60 min, respectively. The irradiated samples were characterized using a number of spectroscopy and microscopy techniques, including RBS/C, ToF-SIMS, STEM, STEM-EDS, STEM-EELS, HIM, Nano-SIMS and APT.
This presentation will report our main results from ion irradiation studies of monocrystalline γ-LiAlO₂ and polycrystalline pellets. Major conclusions from these studies include the following. Saturation stages of lattice disorder are found in γ-LiAlO₂ irradiated at 573 K. Complete amorphization with gas bubble formation starts from the surface. Cubic LiAl5O₈ precipitates surrounded by cavities form in the damage peak region. Li out-diffusion and vacuum evaporation from the sample surface occur during ion irradiation at elevated temperatures. STEM-EELS provides evidence for preferred pathways of Li diffusion along the grain boundaries. In situ HIM shows volume swelling, He bubble formation in grains and He release via grain boundaries. In situ TEM reveals a rapid formation of cavities due to radiolysis in γ-LiAlO₂ under low-energy electron irradiation. The thermal annealing data suggest that the D diffusion coefficients in the irradiated material have a value on the order of 2×10^-13 cm²/s at 773 K. There is an exponential decay behavior of D release at room temperature from γ-LiAlO₂ pellets irradiated at 188 K; at the low temperature, D atoms appear to be immobilized in the irradiated structure. The data from this study also show significant H and D diffusion and release from γ-LiAlO₂ during irradiation at 573 K; D retention increases with increasing D₂⁺ ion fluence and tends to saturate at 4×10¹⁷ He⁺+D⁺/cm².

Be12X (X = Ti, V, Mo and W) intermetallics are a promising family of materials for the first wall and neutron multiplying applications in future nuclear fusion reactors. Due to the difficulties of working with Be they are often poorly characterized. Thus, the crystallography, elastic properties, atomic transport processes, thermodynamical stability and deviation from stoichiometry have been investigated using computer simulation based on density functional theory.
Surprisingly there has been considerable controversy regarding the structure of TiBe12, which is variously reported as hexagonal and tetragonal. Lattice dynamics simulations show the I4/mmm tetragonal phase to be more stable over all temperatures. The formation enthalpies of intrinsic point defects and the elastic constants of all the Be12X intermetallics are then predicted. Defect concentrations were shown to be dominated by antisite disorder and Be vacancies, suggesting these materials can accommodate excess X much more easily than excess Be. In Be12Ti it was found that titanium defects have much higher hopping energy than beryllium across the majority of migration pathways studied. Both beryllium vacancy and interstitial diffusion is weakly anisotropic, however, migration of beryllium divacancies is isotropic with energy equal to that of the isolated vacancy.

The use of heavy ions, protons, or electrons to understand the effect of irradiation on microstructures and therefore materials properties presents many experimental benefits over the use neutrons. However, extrapolating the laboratory observations to behaviors found under neutron irradiation in reactor conditions is not a straightforward process. In particular, dose rates that are orders of magnitude apart and can play a significant role on the development of microstructures, through mechanisms that include higher point defect production and cascade mixing effects. Here, we will discuss the behaviors under ion, proton, and neutron irradiation of alloys that would normally phase decompose under thermal conditions. Specifically, using atom probe tomography to characterize microstructures at high spatial and chemical resolution, the evolution of microstructures in model ferritic Fe-Cr alloys and austenitic Ni alloys was quantified as a function of dose and dose rates. In Fe-Cr alloys, precipitation of the α’ phase is expected under long thermal annealing times. In the selected commercial Ni alloys (625, 625 Plus, and 690), the γ” phase, an ordered Pt₂Mo-type phase, and in some cases the γ’ phase are expected to precipitate. Under irradiation conditions, both alloy systems exhibit different precipitation regimes that are dependent on dose rate. Generally, accelerated precipitation by radiation-enhanced diffusion is observed at lower dose rates, establishment of a steady state regime occurs at intermediate dose rates, and precipitation is entirely suppressed at higher dose rate.

Recently there has been a considerable need to extend the lifetime of the world’s pressurised water reactor (PWR) fleet from 40 to 60 or possibly 80 years. The continued safe operation of these reactors is intrinsically linked to the structural integrity of the steel used for the reactor pressure vessel (RPV). Matrix damage and solute clustering induced by neutron irradiation are considered one of the main causes of embrittlement of RPV steel. Intensive research has been ongoing to investigate embrittlement mechanisms of the low alloy ferritic RPV steel, however, testing neutron irradiated materials can take a long time and is very costly. The recent development of micro mechanical testing allows researchers to evaluate mechanical properties of ion irradiated materials which aim to simulate the damaging effects observed on neutron irradiated materials.
This work presents in situ micro tensile testing of a novel manufactured RPV steel: hot isostatic pressed (HIP) SA508 grade 3 steel. Micro tensile specimens with dimensions of 5 x 5 x 15 µm were manufactured using a gallium focused ion beam (FIB) and tensile testing was performed under displacement control at a rate of 20 nm/s in situ in a scanning electron microscope (SEM). Tests were performed under the same conditions on unirradiated and ion irradiated specimens which were irradiated by He²⁺ ions at 5 MeV at room temperature to the damage of 0.6 dpa using a 2 MV tandem accelerator with energy degrader at ANSTO.
The mechanical properties (in terms of ultimate tensile stress and 0.2% proof stress) of the unirradiated micromechanical tests agreed well with data obtained via conventional mechanical testing. The effects of ion irradiation on the mechanical properties of this RPV steel were measured as an increase of ultimate tensile stress and yield strength via micro tensile testing. The hardening of the irradiated specimens was also measured using nanoindentation performed on the top sample surface. The hardening mechanism can be correlated to the implantation induced features observed on the ion irradiated material that were resolved using a transmission electron microscope. Fine bubbles/voids with an average diameter of 1.2 ± 0.5 nm were observed.

There are currently concerns that clusters of Cu, Mn, Ni and Si could limit the operational lifetime of reactor pressure vessels (RPVs) through reductions in fracture toughness, unaccounted for in regulatory models. As there is much debate over the origin of these clusters and whether they are radiation-enhanced or radiation-induced and whether they can be classified as discrete phases, there is an ongoing global research effort to investigate their mechanism of formation.

Using density functional theory (DFT) simulations, we have conducted investigations into the effect of strained environments on the binding energies of solutes, solute pairs, vacancy-solute pairs and dumbbells in bcc Fe to better understand the conditions that favour nucleation and growth of solute clusters as it is expected that these are dependent on both strain fields and the local chemical environment. Atom probe tomography studies have found that clusters are frequently associated with dislocation loops and grain boundaries, suggesting that these could act as nucleation sites for solute clustering. As modelling large dislocation loop sizes is computationally expensive using DFT, we have used interatomic potentials for Fe to model both small (19 atom) and large (65 atom) <111> dislocation loops to identify the range of strain on the lattice. We have then used these results to parameterize DFT simulations by modelling defects in strained lattices that are typical of these loops.

Our findings show that: (1) vacancies will be dragged to these locations as vacancy formation energies are reduced by more than 1 eV when subject to large volumetric strains, (2) solutes will be dragged due to the strong solute-vacancy binding energy for Ni, Mn, Si in both first and second nearest neighbour positions, (3) solutes remain bound to vacancies even in significantly strained environments, and (4) magnetic effects are important to model the energetics of systems, particularly for complex environments such as those including self-interstitial atoms.

UK’s nonprescriptive regulatory system has incentivised the nuclear industry to invest in the research and development of reliable structural integrity codes since 1970’s. To support the UK structural integrity services, the UK project on the ‘Multi-scAle INTegrity assessment for Advanced high-temperature Nuclear system (MAINTAiN)’ offers the ability to further strengthen UK’s position in structural integrity through developing a novel computer model-based assessment code. The aim is to develop the knowledge that is necessary to reliably assess the integrity of critical nuclear components that operate in extremes of high-temperature, high mechanical stress, and irradiation. To this end, a predictive three-dimensional multi-scale creep deformation model, informed by the materials' microstructure, will be established. The MAINTAiN project will provide the scientific foundation for the next generation of high temperature structural integrity assessments.
The University of Liverpool has carried out a part of irradiation damage on the dislocation creep behaviour of three materials that are of interest for the next generation of nuclear power plants: Stainless Steel Type 316LN, ODS Steel, and FeCrAl alloys.
The presentation will illustrate the details of the MAINTAiN and the latest results on the irradiation test of 316LN stainless steel. The irradiation damage was induced by self-ion implantation using the Dalton Cumbrian Facility (DCF). In order to investigate changes, this works has included high resolution transmission electron imaging and spectroscopic analysis of damage in order to investigate the changes in irradiation.

Reactor pressure vessel steel SA508 (a low-alloying steel) and 304 (Fe-Cr-Ni austenitic stainless steel) are widely employed in both PWRs and BWRs. Weldments typically have high residual stress and composition gradients within the fusion zone (FZ) and heat affected zone (HAZ) that lead to greater stress corrosion cracking (SCC) susceptibility. Irradiation assisted SCC can be controlled by factors including water chemistry, localized deformation, tensile stress, radiation induced segregation and radiation damage. In particular, the redistribution of Cr, the element responsible for oxidation/corrosion resistance via chromia formation, can sensitize these alloys to SCC phenomenon. These materials can be preferentially attacked at active paths, such as GBs in the FZ and HAZ, when exposed to chemically reactive environment and to radiation.
Slow strain rate tests (SSRT) with strain rate up to 10^-8/s were applied on proton irradiated EPRI SA508-309 weldment in an autoclave system with simulated the BWR/NWC environment (10.2 Mpa, 288 °C and 2000 ppb oxygen) in order to investigate the radiation assisted SCC. The effect of radiation to induce segregation in the FZ and HAZ of SA508-309 weldment is of typical interest. In addition, SSRTs were employed on as-received SA508-309 weldment to separate the effect of irradiation from environment factors such as LWR water chemistry and applied tensile load. Moreover, SSRTs were conducted on as-received weldment at room temperature and atmosphere pressure. The yield strength and ultimate tensile strength of as-received and proton irradiated SA508-304 weldment under NWC environment is 25% lower compared to that under room temperature and atmosphere pressure. The redistribution of elements such as Cr in the grain boundary of as-received and irradiated weldment is investigated with STEM-EDS. The initiation of cracking was studied by SEM and TEM in the as-received and irradiated weldment. This talk will focus on identification of the primary factors of SCC in Fe-based weldments.

Development of next generation materials for nuclear energy structural applications requires understanding structural changes and failure mechanisms under extreme environments. Reactions at interfaces under corrosive environments play a crucial role in material failures in current and advanced reactors.

Our research focuses on utilizing synchrotron characterization techniques such as spectroscopy, multi-modal imaging and diffraction for characterizing material interfaces and their properties when subjected to extreme environments of high temperature and radiation fluxes present in nuclear reactors. Synchrotron based characterization studies performed for structural materials and molten salt systems, elucidating corrosion at interfaces will be presented.

X-ray absorption spectroscopy has been utilized to investigate local coordination environment, bonding dynamics and electronic properties of metal species, and their changes at elevated temperatures in molten salt systems. Using a custom-designed in situ cell, we used Ni and Co dopants in single component ZnCl₂ and ZnCl₂-KCl eutectic and studied their properties using Ni, Co and Zn K-edge spectra. The local structural changes and speciation of metallic species in molten salt systems will be discussed.

3D multi-modal imaging techniques will be discussed to provide quantitative characterization of Cr segregation at grain boundary inside the sensitized steel sample using X-ray Fluorescence. Tomographic image reconstruction of stainless steel deteriorated by IG corrosion is combined with 3D imaging of phase contrast using Differential Phase Contrast imaging. This combined structural and chemical analysis of the Cr segregation at grain boundaries allows us to understand IG corrosion in sensitized stainless steel at nanoscale.

To track degradation of materials in real time, we built and commissioned an in situ sample environment for studying interfaces under high-temperature and pressure conditions using x-ray diffraction to investigate corrosion mechanisms in cladding materials. Capabilities of the in situ sample environment will be discussed.

The development of new structural materials that can withstand the degradative and corrosive environments exhibited by molten salt systems currently stands among the most important remaining scientific targets that must be reached before viable molten salt systems can be realized at an industrial scale. In corrosion focused studies, the prevailing trend observed in the most commonly investigated alloys (such as Hastelloy N or 316 stainless steel) is the depletion and dissolution of Cr atoms near the alloy surface and along grain boundaries. The atomic level mechanisms and forces that drive the migration of Cr and its eventual removal from the surface are not well understood, and due to the harsh environments intrinsic to molten salt systems experimental efforts may struggle to observe these fundamental processes. Carefully devised computational modelling offers another avenue to study the nature of these interactions and may even help to suggest potential mechanisms and to narrow the field of possible reaction pathways for future experimental studies to investigate.

In this study, the segregation behavior of Cr near an (100) fcc Ni surface is investigated with first-principles density functional theory (DFT), first for a clean surface and then for surfaces with various salt species adsorbed above the segregating Cr atom. Surface segregation energies describe the energetics associated with the location of a solute atom near a surface (i.e. the relative energy of an atom when its located in the bulk versus in a specific surface layer).

Calculations show that under a clean surface (i.e. no adsorbed salt atoms) it is energetically favorable for the Cr atom to inhabit the 3^rd or 2^nd surface layer, but largely unfavorable for it to sit in the topmost 1^st layer. When a Cl adatom is placed above the Cr atom, the segregation energy for the top-most layer decreases, suggesting Cl surface adsorption lowers the barrier for Cr to move to the top layer. When F or O are adsorbed to the surface, the segregation energy not only decreases but becomes significantly negative, suggesting that Cr segregation to the top layer not only becomes more favorable with these species on the surface but also actually preferred relative to a remaining in a bulk position. Further investigation into the effects of H-Cl, H-F, and H-O will also help to understand the role of H and H₂O salt impurities in the corrosion process. The additional presence of an H atom may also allow the highly electronegative surface adsorbates to increase their local electron occupation level, which would more closely resemble the charged ionic nature exhibited by the salt in a molten state. This talk will report new insights into material-salt interfaces through atomistic modeling.

Molten salt eutectics are used as heat transfer fluid in concentration solar power systems. One of the technological challenges with such systems is the control of corrosion of the materials used for transport of the molten salts. For this reason, understanding how the interface between salts and alloys change as a function of temperature is important. In this study, we focused on the interfacial structure at the interface of KCl-MgCl₂ molten salt eutectic and Haynes 230 alloy using neutron reflectometry. This binary molten salt eutectic was deposited on Haynes 230 alloy to characterize the multilayer as a function of temperature. Our measurements were carried out in situ under ultra-high vacuum conditions. First, a thin film of a single Haynes 230 alloy layer was studied to characterize the alloy as a function of temperature before exposing it to molten salts. The structure of the Haynes 230 alloy film was measured between room temperature to 451 °C. Subsequent measurements were made with a thin film of salt sandwiched between two layers of Haynes 230 alloy at temperatures below the salt’s melting point with sample temperatures ranging from ambient to 383 °C. In both cases, a change in the Haynes 230 alloy structure is observed at around 160 °C. We will describe the environment chambers used for these measurements as well as our results of the behavior of the interfacial structure in these thin film systems.

Molten fluorides have been proposed for use as a coolant in various advanced nuclear concepts including molten-salt reactors (MSRs), fluoride high temperature salt-cooled reactors (FHRs) and fusion devices. In these systems, significant quantities of tritium will be produced which must be captured and controlled to prevent a radioactive release. In order to efficiently remove tritium from the system, the transport and chemical properties must be well characterized and understood. Yet, difficulty in the handling of fluoride salts and tritium has resulted in relatively few experiments being conducted and large discrepancies across different studies. In this work, the transport and local chemistry of different tritium species (tritium fluoride and diatomic tritium) has been studied in the prototypical fluoride salts FLiBe (66.6-33.3 mol% LiF-BeF₂) and FLiNaK (46.5-11.5-42 mol% LiF-NaF-KF) using ab-initio molecular dynamics (AIMD). The methods were validated using experimentally well-characterized fluorides. Salts examined in this study include binary fluorides LiF, NaF, and NaF, and ternary fluorides LiF-BeF₂ and LiF-KF at various compositions. Radial distribution functions and coordination numbers from neutron diffraction, and diffusivities and activation energies from nuclear-magnetic resonance are accurately reproduced. Chemical species and structures were found by using graph-theoretic representations of the system at each time step and removing noise with probabilistic filtering methods. Over the temperature range from 873 to 1373K, the T₂ and TF diffusivities in FLiBe were found to be D[m²/s]=4.14E-7*exp(-33.0[kJ/mol]/RT) and D[m²/s]=1.724E-7*exp(-32.0[kJ/mol]/RT) respectively. At the same temperatures in FLiNaK, T₂ and TF diffusivities were D[m²/s]=1.38E-6*exp(-39.0[kJ/mol]/RT) and D[m²/s]=4.94E-7*exp(-41.0[kJ/mol]/RT) respectively. In both systems, TF exhibited 3-4 times lower diffusivity than the reduced form of tritium T₂. In both cases, decrease in diffusivity due to loss in mobility is attributed to TF reactivity and complexation in the salt. In FLiBe, TF was found to be bound to various BeF₂ molecules, while in FLiNaK, TF₂^- complex was formed. In contrast T₂ molecules in both simulations were found to be stable in pure salt with limited chemical interaction in the solvent. The tritium diffusivities fall within range of experimental values and discrepancies between different experimental studies in literature were attributed to a combination of the following factors: differences in measured species (TF and T2), presence of impurities (corrosion products, moisture) and differences in isotopes used (H, D, T). Large differences in tritium transport will have significant impact on assessing tritium control strategies, which typically rely on mass transport (adsorbents, permeation windows) and chemical condition (redox controls). This study provides a deeper understanding of chemical interactions and accurate chemical-transport data that can be used to predict macroscopic behavior and design engineering and control systems.

Next generation concentrated solar power systems and molten salts reactors have created new interest in molten salt chemistry. The corrosion processes of the alloy used to contain the molten salts as a function of temperature remains to be fully understood. In particular, we are interested in the corrosion of Haynes 230 (approximately Ni>57 Cr22 W14) in contact with a KCl/MgCl₂ eutectic mixture for application in solar power systems. Our team has focused on studying this corrosion process in situ using neutron reflectometry. We will present the design of the high temperature cell that will enable us to perform those in situ measurements from room temperature to 850 C.

Zirconium based alloys for nuclear fuel cladding are known to precipitate hydride phases during operation. These precipitates can cause several issues including embrittlement leading to cracking and failure of the cladding. In order to predict the materials properties of the cladding the hydride phases must be accurately simulated. In this work the thermodynamic stability of phases in the zirconium-hydride system was evaluated using first principles-based methods. Cluster expansion techniques were employed to assess the stability of each phase over a range of concentrations at zero temperature. Phonon contributions for each ground state were calculated and combined with results of finite-temperature Monte Carlo simulations to create Gibbs free energies of the phases as a function of temperature and concentration. From these results a new phase diagram was generated, which will clarify discrepancies between different experimental phase diagrams. Moreover, the free energy functionals of all relevant phases were parameterized, allowing for more accuracy calculating microstructure and materials properties of cladding, for instance, in phase field simulations with INL’s MARMOT code.

Amorphous zirconia (a-ZrO₂) has been modelled using a combination of reverse Monte-Carlo, molecular dynamics and density functional theory together. This combination has enabled the complex chemistry of the amorphous system to be efficiently investigated. Interestingly, the a-ZrO₂ system was observed to accommodate excess oxygen readily through the formation of neutral peroxide (O₂²⁻) defects – a result that has implications not only in the a-ZrO₂ system, but also in other systems employing network formers, intermediates and modifiers. Similar conclusions are made for the yttria-stabilized zirconia system and has been corroborated with Raman spectroscopy data. These results have potential implications for materials behaviour in reactor systems: notibly a mechanism for oxygen to be accommodated and transported through thin, amorphous zirconia films such as those expected at highly damaged grain boundaries - providing a mechanism for zirconium cladding corrosion.

Crack initiation in zirconium alloys is an important issue affecting the safety of nuclear reactors. Zirconium hydrides that precipitate in service are potential crack nucleation sites. In this work, the deformation and cracking of zirconium hydrides was studied during the deformation at room temperature of Zircaloy-4 samples up to fracture. Two hydrogen concentrations of 100 and 200 ppm were considered.

Hydrides precipitate both inside and at the boundary of the grains. It is expected that thicker hydrides and hydrides at the grain boundary are more prone to fracture. The main objective of this study was to better understand which of the two is the most important factor, and to identify at which point in the deformation of the alloy the first hydrides break. Microvoids thus nucleated may coalesce and be an important factor in understanding the fracture of these alloys. Scanning microscopy (SEM) images of a number of hydrides, both intergranular and intragranular, were taken at discrete increments of deformation, and the fraction of those that show evidence of fracture was estimated.

The results show that intergranular hydrides possess a lower deformability, which can be rationalised by analysing the conditions for slip transmission into and out of the hydride.

Effect of alloying elements on zirconium alloy degradation properties has been widely studied because of its application in nuclear water reactors. It is observed in experiments that alloying element in zirconium alloy could result in different morphology of the passive film as well as different oxide film growth kinetics. The goal of this work is to study oxygen diffusion and hydrogen pickup processes in the ZrO₂ passive film system under the effect of alloying elements, with varying temperature and chemical environment. In particular, we focus on the chemo-mechanical effect of Fe, Cr, Sn and Nb dopants in ZrO₂ that exist in the oxide matrix as point defects from first-principles calculations. We show that the change in defect chemistry imposed by doping is strongly correlated with the valence states of the dopants. Fe and Cr, which are dominantly in +3 valence states, tend to decrease the concentration of oxygen interstitials and increase the concentration of hydrogen interstitials in ZrO₂, thus suppressing oxide growth while increasing hydrogen pickup. Nb, on the other hand, is dominantly +5 and has the opposite effect. Changes in defect chemistry are accompanied by change in phase stability. We have discovered that Fe and Cr can strongly stabilize tetragonal-phase ZrO₂ while Nb has a strong destabilizing effect. This effect provides insight for the different oxide film morphology of zirconium alloys with varying alloying element.

Uranium diboride is being considered as an advanced fuel, while zirconium diboride remains a leading candidate for use as a burnable absorber, often coated on uranium dioxide fuel pellets. To ensure safe and efficient operation models need to be developed to enable the material to be modelled within a fuel performance code. Thermal properties of these materials are therefore investigated using density functional theory methods. It was found that the boron isotope ratio significantly varied the thermal conductivity for both materials. When compared to experimental data, the thermal conductivity for ZrB2 is well prediced, however, a significant discrepancy is observed for UB₂. Potential reasons for the discrepancy, including the Hubbard factor and phonon relaxation times, are explored and discussed.

Radiation technology is widely used medical radiation therapy, radiation dosimeter, food preservation and other industrial applications. These include scientific research such as high energy physics, radiochemistry, change of material properties and so on. Radiation detection is one of important research area to safely use ionizing radiation. Several devices are investigated for radiation detection such as Silicon diode, metal-oxide semiconductor field effect transistors, thermo-luminescent dosimeters, optically stimulated luminescence dosimeters. These devices use semiconducting properties of organic and inorganic materials with electrical or optical connection for readout.
In this study, we synthesized organic conjugated molecule for application to detection of gamma-ray without any readout devices. This molecule is consisted of two main parts. One is low-band gap conjugated core part, which is dithiophene moiety, for wide absorption in visible range. And the other is dicyanomethylene parts which are possible to induce band-gap change of dithiophene core part by gamma-ray. After irradiation gamma-ray with 1 Gy on film of the molecule, dark blue color is changed to yellow color due to change of band-gap from 1.74 eV to 2.16 eV. We investigate optical and electrical properties of the molecule by gamma-ray and we also discuss detail mechanism of color change and effect on the molecule by gamma-ray.

Pyroprocessing has been investigated as an option to recycle and to manage spent nuclear fuels (SNFs) from pressurized water reactors. The oxide SNFs needs to be transformed to metallic states for electrochemical recovery of transuranic elements. In this respect, oxide reduction (OR) technique has been developed for the metal conversion. LiCl molten salt-based electrochemical reaction is the common OR technique. Conventional OR process is designed based on Li₂O recycle reaction. Li₂O is electrolyzed to form Li metal at cathode and O₂ gas at anode (2Li₂O = 4Li + O₂(g)). The Li metal subsequently reduces the oxide SNFs (ea.g., UO₂ + 4Li = U + 2Li₂O) to release new Li₂O. Hence, Li₂O concentration in the LiCl salt is expected to be constant during the OR operation. Pt has been widely used as the O₂-evolving anode material because of its superior mechanical and chemical stabilities in the high-temperature oxidation environment.
Scale-up is an important issue on feasibility examination of pyroprocessing. The Pt anode, however, is not preferred in the large-scale system because it is gradually damaged upon the operation resulting in high processing cost. The Pt anode is known to have three anodic reactions depending on anode potential: Li₂PtO₃ formation (2Li⁺ + Pt + 3O^2- = Li₂PtO₃ + 4e^- at low potential), O₂ evolution (2O^2- = O₂(g) + 4e^- at medium potential), and anodic dissolution (Pt = Pt²⁺ + 2e^- at high potential). Only the O₂ evolution is required and thus the operation voltage of the Pt anode system should be carefully controlled (normally, ~3 V). Nevertheless, it was revealed that the side reactions cannot be avoided to induce the Pt anode deterioration, even in small-scale experiments (~15 g-UO₂). This would be due to non-uniform potential distribution across the Pt anode (by iR drop and other reasons) and local inhomogeneity of the Li₂O concentration in the salt. This should become more severe with the scale-up and it is considered that the precise control of the reaction pathway of the Pt anode would be very difficult in the large-scale equipment.
KAERI (Korea Atomic Energy Research Institute) has examined several materials to replace the Pt anode. One approach is to develop the O₂-evolving anode with better electrochemical stabilities using conductive ceramics and other noble metals. C anode was also suggested. The C anode system operates at high voltage range (>4 V) compared to the Pt anode system to decompose the LiCl salt to generate Cl₂ gas (2Cl^- = Cl₂(g) + 2e^-) at the C anode. The advantages of the C anode are non-concern on the anodic reaction control and high operation current. In this presentation, the recent studies on the anode material development done at KAERI will be introduced.

A U- 10 wt.% Zr alloy containing lanthanide elements (Nd, Ce, Pr, La) was introduced as a surrogate for transuranium alloys in metallic nuclear fuel. U-Zr alloys containing the lanthanide elements interact significantly with graphite crucibles. Conventional coating material, Y₂O₃ reduced the interaction, but partially interacted and detached from crucible after casting due to the difference in the thermal expansion coefficients of the crucible and the coating layer. The crucible was contaminated and could not thereafter be reused. Thus, it was considered a form of radioactive waste. Moreover, interacted materials could not be removed from the casting part and amounts of the reaction reduced casting yields. Materials such as LaYO₃, SiC, TiC, and TiN along with density-improved Y₂O₃ were considered as candidates for use in the creation of a reusable crucible, which could be used to improve casting yields and to reduce radioactive wastes. The interactive behaviors of the candidate materials were considered by means of a sessile drop test. The annealing temperature was set to 1,450 °C for investigating interactions of the casting method. Microstructures of the candidate materials were observed using a scanning electron microscope (SEM), and their elemental distributions were characterized using energy-dispersive X-ray spectroscopy (EDS). Phase formations of the candidate materials were identified using an X-ray diffractometer (XRD). The material interaction was reduced by the enhanced Y₂O₃ coating and LaYO₃. A shortlist of candidate materials for use in the creation of a reusable crucible was drawn up based on the results of this research.

U-Zr metallic fuels have been developed as a nuclear fuel for Sodium-cooled fast reactor because metallic fuels are related to excellent reactor safety and fuel cycle economy. In the fabrication process of U-Zr metallic fuels, the graphite crucible is commonly used for melting and casting and yttrium oxide is coated on the inner surface of the graphite crucible to prevent the reaction with melted U-Zr alloy. It is crucial to prevent the reaction between the melt and crucible for controlling the loss in the fabrication. In this study, we investigated the reactivity between uranium, zirconium and graphite crucible. The graphite crucible was used to clarify how the graphite react with melt uranium and zirconium. U-10wt.%Zr alloy was melted using injection casting apparatus which is an induction furnace. Raw materials consisting of zirconium sponge and uranium metal were charged in the graphite crucible in the order named. The materials were superheated at 1,600°C for alloying and it was cooled in the furnace after alloying. Microstructures in the surface region of the melt-residue of U-Zr alloy were characterized using Scanning electron microscopy (SEM) equipped with Energy dispersive X-ray spectroscopy (EDS). Although the graphite crucible was used for melting uranium and zirconium, the U-10wt.%Zr melt residue was fabricated without alloying between uranium and the graphite crucible. The melt residue was easily separated from the graphite crucible. However, the reaction layer was formed on the surface of U-10wt.%Zr melt residue. The reaction layer mostly consisted of carbon and zirconium. The thickness of reaction layer was about 5 μm at side region and 10 μm at bottom region. The bottom region had the thickest reaction layer among whole surface because first of all, zirconium was reacted with graphite crucible at the bottom region when the charged raw materials were heated. The thickness of the reaction layer was assumed as the depth penetrated into the U-Zr matrix from the graphite crucible depending on the volume of the zirconium, melting temperature and time. The surface of melt residue was investigated using X-ray diffractometer (XRD). As a result of XRD analysis, ZrC and U3O7 were detected on the reaction region. It is likely that C was contaminated from graphite crucible. The quantitative chemical composition of graphite crucible was investigated using Inductively coupled plasma-atomic emission spectroscopy (ICP-AES). AS a result of ICP-AES analysis, uranium and zirconium were contaminated at contact region. On average, uranium and zirconium contents of graphite crucible were 281 and 31 ppm but it was a considerable difference according to the position. In this study, U-10wt.%Zr alloy was fabricated in the optimizing manufacturing process to examine the reactivity between uranium, zirconium and graphite crucible. The reaction layer consisted of zirconium and carbon is formed on the whole surface of the melt residue because zirconium is commonly reacted with carbon at the contact region with graphite crucible. The graphite crucible is also contaminated with uranium and zirconium when melting.

High-quality thermal conductivity data is critical to the rational design of materials for applications ranging from thermoelectrics to nuclear reactors. However, such data is experimentally and computationally challenging to obtain due to any or all of cost, technical hurdles, and time requirements. The thermal conductivity of these materials can be further influenced by their working conditions, e.g. irradiation, due to the microstructure changes. Therefore, practical models and quantitative understanding of the thermal conductivity, including changes due to microstructure, is essential to support the development of new materials.
In this work, we developed a computational model of thermal conductivity of irradiated metal alloys based on density functional theory (DFT) calculations, semi-classical scattering theory, and experimental data. This approach has been applied to model the electrical and thermal conductivity of alpha uranium previously, and can be extended to model the condictivities of U-alloys like U-Zr and U-Mo. The present model incorporated the contributions of point defects (vacancies and transmutation products), grain boundaries, and noble gas bubbles (inter- and intra-granular) resulting from irradiation. The effects of point defects and grain boundaries are modeled by both DFT calculations and semi-classical scattering theory, and the effect of noble gas bubbles are modeled using empirical models. The utility of our model is demonstrated on U-Mo alloys, which are promising nuclear fuel materials for use in next generation nuclear reactors, and displays semi-quantitative agreement with experimental data with a root-mean-square error of about 0.8 W/m-K (≈7% error versus experiment) over the typical operating temperature and burn-up range for U-Mo alloys. This work provides insight into the semi-quantitative relationships of the thermal conductivity changes resulting from the different irradiation-induced changes in atomic structure and microstructure over a wide range of temperatures, and shows that scattering of electrons from point defects produce the majority of the reduction of the thermal conductivity of U-Mo alloys under irradiation. As an example, at 50% U235 burn-up, the thermal conductivity decreases about 20% due to the effect of point defects, while grain boundary and gas bubble formation contribute only about another 10% decrease. Furthermore, our model can be straightforwardly applied to different irradiation conditions or microstructures by changing the values of of a few corresponding parameters.
The present model both serves as a powerful tool for quantitative semi-empirical modeling of thermal conductivity in irradiated metals, and for materials design insight for the control of thermal conductivity in metal alloys in applications experiencing irradiation, such as the further development of advanced metallic U-alloy nuclear fuels.

The potential strategic application of uranium in nuclear power industry, as well as the environmental and human health problems caused by uranium, promotes the development of reliable methods for in situ extraction of U (VI) from different solutions. Herein, a unique multifunctional three dimensional electrodes have been fabricated by simple hydrothermal method and used as outstanding electrosorbent for uranium extraction from different aqueous solution. The electrochemical sorption method solves the conventional drawbacks of other methods by using fixed potential to guide the movement of uranium ions in the solution and increase the collision rate onto the electrode surface, resulting in faster kinetics and higher adsorption capacity, easier scaling up, and high selectivity due to specific grafted functional groups. The fabricated electrode was characterized by RAMAN, FTIR, XRD, STEM, EDS, XPS before and after uranium loading. The adsorption capacity for U(VI) ions with our multifunctional 3D electrode is estimated to be 4800 mg/g at pH 5.2 without reaching saturation exhibiting the highest reported sorption capacity compared with other electrochemical methods. The fabricated electrode can be efficiently regenerated and recycled for at least 7 times in uranium solution with concentration 40th times higher than initial concentration without any observable change in efficiency. This work represents the first application of multifunctional 3D electrode for uranium extraction from mining solution and contaminated ground water.

Uranium borides are attractive candidates for advanced nuclear fuels. In particular, uranium diboride-based fuels are expected to outperform – in terms of thermal conductivity and spatial uranium density – their currently-used counterparts based on uranium dioxide. Nevertheless, uranium diboride is expected to be less tolerant than uranium dioxide upon accidental contact with water and other oxidizers at high temperatures. The dispersion of uranium diboride in a uranium dioxide matrix is seen as a possible solution that combines the advantages of both compounds, affording a composite material that has a greater thermal conductivity and spatial uranium density than pure uranium dioxide while maintaining its resistance to oxidation.
The interface chemistry of UO₂-UB₂ systems has been investigated via temperature dependent atomic scale simulations.

The realization of the breeding blanket system represents one of the crucial points in the design of future generation fusion reactors. At the time being, issues related to materials compatibility persist. According to the most relevant design, fusion reactors will take the Tritium-Deuterium fusion as the reference reaction for power generation. Thus, the availability of Tritium to fuel the reactor core assumes relevant importance. The breeding process represents one of the main focus points of technological R&D activities and the inhibition of Tritium permeation is mandatory to achieve Tritium balance in the reactor chain: once Tritium is produced, an adequate permeation barrier is required to confine it. At the moment, two are the eligible breeding materials under investigation: a flowing Pb-16Li eutectic alloy and a pebble bed composed by sintered Lithium ceramics and Beryllium. It was demonstrated that both these two materials lead to severe corrosive attack of the steel constituting the breeding blanket moduli (RAFM steel), hence threatening the mechanical integrity of the whole structure. It is thus clear that materials are proving to be one of the major bottlenecks for this technology. In order to undertake these issues, we report on multifunctional nanoceramic oxide coatings grown by Pulsed Laser Deposition (PLD) technique. Oxides were chosen because of their chemical inertia and thermodynamic stability. Among these, Aluminium Oxide and Yttrium Oxide were identified as promising candidates. Deposited films appear dense, compact and well adherent on the steel substrates, showing optimal metal-like mechanical properties combined with high hardness. Morphological and crystalline features were evaluated in the case of as-deposited and annealed samples. A deep characterization was performed by Scanning Electron Microscopy (SEM), Energy Dispersive X-ray (EDX), Fourier Transform Infrared spectroscopy (FTIR) and X-ray Diffraction (XRD). Also, coatings of these materials have been tested as Tritium permeation barriers with Hydrogen at different temperatures (from 350°C to 650°C). They showed a permeation reduction factor (PRF) up to 10⁵ at 650°C. These results have been confirmed also in the case of Deuterium permeation, both under 1.8 MeV electron irradiation and for samples that were previously irradiated with ions. In addition, to evaluate the chemical compatibility of the films with the selected breeding materials, samples have been exposed to corrosion tests (both Pb-16Li and Lithium ceramic pebbles) up to 10000 hours. No corrosive attacks on the steel substrate were detected. Finally, Atomic Layer Deposition technique has been studied and recently employed as a complementary deposition technique to the PLD. The deposition process is still at its early stages of development, but preliminary results show that the coatings thus produced are dense, compact and defect-free. A deeper characterization of these coatings will follow. To conclude, in a DEMO-oriented perspective, oxide coatings present interesting properties as multifunctional protective barriers, proving to be one of the few possible solutions for the material-related issues of all the breeding blanket concepts currently considered.

The typical burnup of currently operating nuclear reactors is around 47 GWd/MTU, which coincides with the onset of permanent bonding between UO₂ fuels and zirconium alloy claddings in water cooled reactors. The bonding between pellet and cladding occurs as a result of the oxidation of Zr-alloy cladding and subsequent bonding with the UO₂ forming a solid solution of (U,Zr)O₂. The impact of this pellet-clad bonded layer increases with increasing burnup. In order to extract more energy from nuclear fuels and increase time between refuelling, operators are increasing burnups beyond 50 GWd/MTU. This research looks at the impact of urania-zirconia mixed oxides [(U,Zr)O₂] on nuclear fuels at high burnups.
The following thermophysical properties of (U_x,Zr_1-x)O₂ were investigated as a function of composition and temperature: lattice parameter, thermal expansion, heat capacity and thermal conductivity. A combination of molecular dynamics and experimental methods were used to provide a means of prediction and validation. Mixtures of (U_x,Zr_1-x)O₂, where x = 0.25, 0.50, 0.75 and 0.9, were produced using sol-gel synthesis to enhance the mixing of cations and achieve a homogeneous solid solution. The results indicate that by increasing the concentration of Zr in (U,Zr)O₂ mixtures the thermal conductivity is degraded slightly at room temperature, becoming negligent at temperatures at and above reactor operating temperatures. However, the formation of t-ZrO₂ on the inner surface of the cladding at moderate burnups (20 – 30 GWd/t) is a barrier to thermal transfer being that the thermal conductivity of t-ZrO₂ is 50% lower than that of UO₂. Additionally, the lower temperature access to superionic transitions due to oxygen disorder could provide some benefit under accident conditions due to the significant increase in heat capacity around 1500-2000 K. This research shows that the impact of (U, Zr)O₂ at high burnups is positive in that it does not affect the thermal conductivity of UO₂ significantly and, due to the permanent bonding of clad and fuel, could enhance thermal throughput given the thermal conductivity of the bonded layer is an order of magnitude greater than that of He.

SiC and W-SiC functionally-graded coatings have been fabricated and exposed to deuterium (D) plasma bombardment at different surface temperatures and ion impact energies in the PISCES-E radio frequency (RF) plasma source [1] to assess their viability as a plasma-facing material (PFM) in a magnetic confinement nuclear fusion device. In situ optical emission spectroscopy (OES) revealed that CD emission, which approximates C chemical erosion, increased with ion impact energy up to 80 eV. Chemical erosion of Si was not observed spectroscopically via the SiD molecular band. Post-mortem Auger electron spectroscopy (AES) measured a <15% increase in relative Si content within the first few hundred nm of the surface. Si enrichment was likely due to preferential erosion of C atoms, as indicated by the asymmetric spectroscopic emission mentioned above. Erosion due to physical sputtering may have also been higher for C than for Si, as shown in [2]. The amount of D retained in samples due to implantation increased by a factor of ~5 between 20 eV and 50 eV, likely due to the onset of ion-induced, extrinsic trap formation. Overall retention was 1.7-2.5× higher in SiC than in W exposed to identical plasma irradiation. The pure W surface of the W-SiC functionally-graded coatings survived plasma exposure at a fluence of 10²⁵ m^-2, as confirmed by AES measurements. Overall erosion of the W-SiC coating was around 2.3× higher than that of pure, crystalline W samples exposed to the same conditions. D retention in the W-SiC coatings exhibited combined trap behavior of both W and SiC samples, and was >10× higher than retention in SiC and W samples. High erosion and retention values may have been an artifact of the coating process, resulting in increased intrinsic trap density and reduced lattice displacement and surface binding energies.

SiC has been considered for use as a PFM in future magnetic fusion devices due to its excellent mechanical strength under neutron irradiation, high thermal strength, and low hydrogenic diffusivity. Tungsten is currently the leading PFM candidate, due to favorable thermal and mechanical properties and low erosion. However, concerns regarding its strength under intense neutron irradiation and its detrimental effect on plasma performance when eroded have led to research on other material candidates, such as SiC. Recent experiments in DIII-D showed that SiC exhibited lower levels of chemical sputtering than that of graphite (a common PFM) [3]. These results motivated further characterization of fundamental plasma-material properties in a dedicated plasma source. SiC samples (coating thickness = 100-200 µm) and W-SiC samples (coating thickness = 10 µm) were fabricated in-house via chemical vapor deposition (CVD) and RF magnetron sputtering, respectively, on ATJ graphite substrates. W-SiC coatings exhibited a stepwise concentration gradient of 100% W at the surface to 100% SiC at the coating-bulk interface. These samples, along with W and ATJ graphite reference samples, were exposed to D RF plasmas in PISCES-E at a flux of 5 × 10²⁰ m^-2 s^-1, T_e of ~ 3 eV, n_e of 3.5 × 10¹⁶ m^-3, and surface temperature of 500 K. In situ OES was used to track erosion products and was quantified via the SXB/DXB method. Overall erosion was measured using a microbalance with a resolution of 0.01 mg. AES was used to measure changes in surface composition due to plasma exposure. Post-mortem thermal desorption spectroscopy resolved the major traps present in the material and estimated overall retention. Qualifying new SiC-based materials in a dedicated plasma source may demonstrate a superior alternative for next-generation fusion reactor materials.

Work supported by General Atomics Internal Research & Development and the U.S. DOE under DE-FC02-04ER54698 and DE-FG02-07ER54912.

[1] G.R. Tynan et al., J. Vac. Sci. Technol. A. 15 (1997) 2885–2892.
[2] S. Bringuier et al., Nucl. Mater. Energy. 19 (2019) 1–6.
[3] T. Abrams et al., MRS Fall Meeting (2019).

Structural components of nuclear reactors must fulfil a series of mechanical and stability properties according to the specific range of experimental conditions they will be submitted during the operation of the reactor. Since the beginning of the construction of generation-II Pressurized Water Reactors (PWR’s) in the 1960’s, austenitic stainless steels from the AISI 300 series have been widely used as structural components for the fuel cladding, due to their high-temperature corrosion-resistance and mechanical strength. The variation of neutron flux along the core of a nuclear reactor presents a challenge to the materials used to build the various components of the core, considering that for a non-steady flux the effects on structural properties also varies. Wiedersich, Okamoto and Lam (WOL) proposed a model to specify irradiation conditions of dose rate and temperature where significant RIS should occurs, being limited by an increase in the thermal concentration of vacancies at high temperatures and recombination of vacancies with interstitials at higher dose rates. In this work, we report the formation of radiation-induced precipitation (RIP) on an AISI 316L steel under in-situ 325 keV Xe ion irradiation/implantation for three different fluxes (1, 2 and 4x10¹² ions.cm^-2.s^-1) up to the same total dose within a transmission electron microscope (TEM) at MIAMI facilities at the University of Huddersfield, UK. All experiments were performed at 550 °C. Statistical analysis from bright-field images under phase contrast and selected area electron diffraction (SAED) were used to characterise the precipitates formed in the samples analysed. The nucleation and growth of second phase precipitation are observed to vary significantly according to the flux the sample is submitted. An overlapping of consecutive cascades at higher fluxes is proposed to explain the different behaviours for the evolution of RIP’s, and a small deviation on the lower limit of the WOL’s model is suggested.

In the continued development of materials for use within a nuclear reactor, whether it be fission or fusion based, how a material behaves under the extremes of damage radiation is a key factor in its applicability. The impacts arising from radiation induced damage are generally damaging, ranging from local damage through to full amorphisation, or the formation of new phases both by equilibrium or non-equilibrium synthesis. Such changes often have macro-scale impacts, for example a reduction in thermal conductivity, or change in a ductility. However, they key length scale at which all changes can be considered is at the nano-scale, where damage can be visualised most easily.

Using electron microscopy as a key aid in determining the impacts from damage, a range of systems, primarily ceramic in nature, will be discussed. Such materials are key components in multiple next generation reactor technologies, whether it be fission or fusion, fuel or structural in application. Regardless of the application, there are common factors that influence a response to radiation induced damage, and the examples presented here will help link the common themes, whether they be compositional or structural in nature.

Zirconium-based alloys have been successfully used for light water reactor fuel claddings since the 1960s. Despite their excellent performance during routine operation, during severe accidents such as the one which occurred at Fukushima Daiichi in 2011, they can undergo rapid exothermic oxidation. This releases hydrogen and supplies a great deal of heat, accelerating the accident. Over the last decade, there has been a concerted effort by the nuclear fission community to develop and implement accident-tolerant fuel designs. One design showing promise is SiC/SiC composite claddings. These consist of a β-phase fibre weave surrounded by β-phase SiC matrix. Despite their strength and improved oxidation performance, the proposed claddings have a number of drawbacks. Namely their irradiation swelling and low thermal conductivity. Their irradiation swelling saturates after a number of months in reactor and is greater on the colder outer surface than the warmer inner surface. The thermal conductivity lower than for zirconium alloys and is worsened by irradiation swelling.

Peridynamics is a developing non-local modelling technique which has the ability to model crack initiation, propagation and branching. Unlike finite element analysis, the technique uses an integral formulation to determine the forces upon material points. This means that crack patterns can be predicted without any a priori knowledge of the loading or crack pattern.

A peridynamic model for composite SiC/SiC cladding has been developed in the Abaqus finite element code. The material model was isotropic and considers matrix cracking and fibre pull-out. The thermal expansion, swelling and the degradation of the thermal conductivity are modelled under typical LWR irradiation conditions. The swelling on the outer surface is greater than the inner surface due to the lower irradiation temperature, causing a tensile stress on the inside of the cladding. This stress increases during the decrease in power at the start of a typical pressurised water reactor refuelling outage and causes microcracking of the matrix on the cladding inner surface. In models without fibres, cracks would propagate through the cladding. If fibres are modelled, matrix cracking will extend to a depth of around 20% through the cladding from the inner surface. If an inner monolith of SiC is additionally modelled, cracking propagates through the monolith and acts as a stress raiser for matrix cracking in the composite. If an outer SiC monolith is modelled, fibre pull-out on the inner surface of the cladding was increased by just under 70%.

Future generations of nuclear reactors are required to operate at higher temperatures and under more extreme radiation conditions than the current generation, creating a need for novel materials for use in these environments. Silicon carbide (SiC) has a number of desirable characteristics for use in reactors but some properties can deteriorate under irradiation. A potential option to improve the radiation resistance is to combine SiC with other carbides such as Titanium carbide (TiC) and Zirconium carbide (ZrC).
To determine the properties of these samples, they can be exposed to high radiation environments and electron microscopy and diffraction techniques can be used for characterisation and analysis of the materials.

Initial samples of SiC/TiC composites have been produced and irradiated in several facilities, with bulk irradiation using 93 MeV Pb ions and in-situ irradiation using 1 MeV Kr ions. The in-situ TEM imaging indicates different responses rates to radiation damage in the TiC phases and the SiC phases. Diffraction of these areas indicates almost immediate amorphisation of the SiC phases, with TiC phases retaining distinct grain boundaries, which is supported by X-Ray Diffraction (XRD) of bulk irradiated samples. The results from these samples are used to guide the fabrication and characterisation of the next set of samples. These samples produced are SiC/ZrC and SiC/TiC composites, each with ratios of 30:70, 50:50, and 70:30. The raw powders are milled at low-speed in inert gas, before being sintered using spark plasma sintering (SPS) at 1600-1900 ^oC. The results of these samples will be discussed in this talk.

Future generation nuclear reactors – either fission or fusion systems – present intrinsic benefits in terms of economics, safety and waste management. However, to allow the deployment of these technologies, innovative material strategies must be designed and implemented. Recently, coatings have earned great attention in the nuclear field as a reliable cost-efficient solution. Indeed, they could tackle major issues such as corrosion, erosion, hydrogen/tritium confinement, etc. In this framework, amorphous/nano-crystalline alumina coatings developed by the Istituto Italiano di Tecnologia (IIT) have been extensively characterized as corrosion-resistant anti-diffusion barriers. Al₂O₃ films have been tested under ion irradiation, proving high stability up to 150 dpa. This radiation tolerance has been related to the amorphousness of the pristine material, which evolves towards a crystalline structure under irradiation. To preserve these enabling mechanisms and improve the range of operation (in terms of life time and operation temperature), the amorphous matrix must be conserved as long as possible. In this work, the stabilization of alumina by doping is evaluated under different conditions. Firstly, pure and doped materials have been compared by thermal annealing, showing a significant increase in the crystallization temperature (from 650 °C to more than 800 °C) for the stabilized alumina. Then, irradiation tests have been performed with in situ TEM at the IVEM-Tandem facility, at ANL. Samples have been irradiated with different ions (namely, 350 KeV Kr and 500 KeV Au) up to 20 dpa. Experiments have been conducted at 600 °C - for the sake of consistency with previous ion irradiation tests - and at 800 °C, to observe the accelerating effect of irradiation on the thermally activated crystallization process. Albeit both the materials withstand the ion irradiation without delamination or structural failure, doped alumina retards the crystallization threshold even under irradiation. Moreover, at significant levels of radiation damage, the presence of doping stabilizes the low-energy metastable phases of alumina, conserving partially the pristine characteristics. This behavior, especially at the highest temperature, improve further the radiation tolerance of the material: while pure alumina suffers from voids formation after crystallization (a traditional phenome for the bulk oxide), no swelling appears in the doped counterpart. To conclude, in situ ion irradiation tests have been performed on stabilized nano-ceramic coatings. The presence of doping slows down the material evolution under irradiation, preserving the well-established properties of the pristine coating. The outcomes of these experiments represent a major step forward in the qualification process of coatings for nuclear reactors.

The accumulation of defects during irradiation leads to material property degradation modes such as embrittlement and swelling, eventually causing material failure. Effective and efficient removal of defects is of crucial importance to design radiation damage-tolerant materials. Here, by biasing defect migration pathways via carbon nanotube (CNT) infiltration, we present a greatly enhanced damage-tolerant Al-CNT composite with defect storage measured to be one order of magnitude lower than that in pure, irradiated Al. Furthermore, extreme-value statistics (largest size) of defect clusters are significantly changed in the presence of CNT. In situ ion irradiation transmission electron microscopy (TEM) experiments and atomistic simulations together reveal the dynamic evolution and convergent diffusion of radiation-induced defects to CNTs, facilitating defect recombination and enhancing radiation tolerance. The occurrence of CNT-biased defect convergent migration is tuned by the thermodynamic driving force of stress gradient in Al matrix due to the CNT phase transformation. This approach to controlling defect migration using 1D interface engineering creates a new opportunity to enhance the properties of nuclear materials.

Metal degradation due to implanted helium (He) is a concern in nuclear energy applications. We investigate the size and distribution of He bubbles in Ta(Ti)/Zr(Ti) composites using transmission electron microscopy. He implanted composites are compared to both the unimplanted material as well as the individual constituents of the composite. Nano-scale He bubbles form in different quantities in the two components of this composite as well as at the curved Ta(Ti)/Zr(Ti) interfaces within the materials. We discuss the consequences of our findings for the development of radiation resistant materials with complex, hierarchical microstructures.

Atomistic modeling of irradiation damages through displacement cascades is deceptively nontrivial. Due to the high energy, high velocity nature of the atom collisions, individual cascade simulations can become very computational expensive and ill-suited for size and dose upscaling. In order to examine microstructural evolutions, and mechanical property changes due to defect accumulation, alternative methods of modeling radiation defect accumulation are needed. Originally developed for application in ceramic materials, the Frenkel Pair Accumulation (FPA) method generates point defect pairs by directly displacing atoms from its initial lattice site. The applicability of method is somewhat limited to metallic/dense materials, as it does not capture the important cascade process known as the thermal spike. The presence of the thermal spikes has shown to be influence both point defect clustering and sequential cascade overlaps. Instead, using FPA as the basis and incorporating a reduced-order approximation for thermal spike, a new method of modeling radiation damage is developed. By adopting the athermal recovery corrected (arc) formalisms, the new arc Damage Insertion (arc-DI) method is able to predict and replicate radiation events across a wide range of recoil energy. Using Cu and Nb as the case studies, arc-DI is verified against standard displacement cascades. Example applications for simulating high energy cascade fragmentation and large dose ion-bombardment are also provided for demonstration.

Sandia National Laboratories is a multi-mission laboratory managed and operated by National Technology & Engineering Solutions of Sandia, LLC., a wholly owned subsidiary of Honeywell International, Inc., for the U.S. Department of Energy National Nuclear Security Administration under contract DENA0003525.

Nanostructured materials with amorphous intergranular films (AIFs) have demonstrated superior strength and ductility. The radiation performance is expected to be as high as grain boundaries, if not better, due to the large fraction of interfacial volumes, which can be efficient sinks for radiation-induced defects. Meanwhile, the concentrated solute atoms within the film can pin down interface migration. However, better understanding beyond qualitative speculation must be pursued towards the radiation behavior of such novel design for the potential deployment in nuclear applications. In this study, we examine the response of an AIF system (nanocrystalline Cu with Zr doped AIFs) under continuous irradiation with molecular dynamics (MD) simulations. The system behavior is evaluated from three perspectives: ion mixing, defect reduction, and interface evolution. We propose a diffusion model that well characterizes the cascade-driven mixing process, and the spread of Zr distribution scales linearly with the damage level. The exceptional radiation resistance of such design can be understood from three aspects: i) Zr mixing into the bulk enhances local defect annihilation, ii) Zr impedes radiation-enhanced grain growth by restraining AIFs from migration, and iii) Zr increases the interface stiffness to maintain AIF integrity. These findings suggest that the AIF-engineered system can be a promising candidate in applications where strong mechanical property, structure stability and self-healing capability of radiation damage are required.

Voids develop in crystalline materials under energetic particle irradiation as in nuclear reactors. Understanding the underlying mechanisms of void nucleation and growth is of utmost importance as it leads to dimensional instability of the metallic materials. In the past decade, researchers have adopted phase field approach to study the phenomena of void evolution under irradiation. In the phase field approach, the boundary between the void and matrix is modeled as a diffuse interface, while it is sharp in reality. This consideration of diffuse interface requires the phase field models to asymptotically match with their sharp interface counterparts in the limit of vanishing interface thickness. The asymptotic analysis of the phase field models has enabled in obtaining a quantitative vacancy-based phase field model, however, it failed to match with the sharp interface model when included interstitials concurrently with vacancies. In this communication, we introduce a thermodynamically consistent, quantitative diffuse interface formulation of type C for void evolution under irradiation, that includes vacancy, interstitial and phase field variable as order parameters. Test cases are presented to validate our new phase field model.

Polymers are widely used in nuclear power plant (NPP) components and in the storage of radioactive materials. Radiation-induced polymer degradation is a known challenge to the lifetime of components. Additive manufacturing (AM) provides a unique capability to cost-effectively replace used parts to key NPP components, thereby extending the life of commercial NPP and reducing maintenance costs and reactor downtime. The radiation damage due to polymers by neutron irradiation is mainly due to secondary protons, recoiled from the polymers themselves. Therefore the study of the effects of proton irradiation of polymers informs the damage study of neutron radiation and is a viable simulation of neutron damage. The radiation damage of 3D printed polymers with gamma beams has already been studied in terms of changes to mechanical properties such as tensile strength and flexural strength. The study aims to obtain an understanding of the proton radiation impact on the mechanical performance of fused filament fabrication (FF) acrylonitrile butadiene styrene (FF-ABS). The study addresses the effect of proton radiation damage at 40 MeV and high-level proton doses through experimental evaluation of both mechanical performance and material characteristic changes. Furthermore, it considers the comparison of mechanical damage caused by protons versus that caused by Co-60 gamma-ray radiation. The key components of this research are (1) material characterization to investigate irradiation damage to FD-ABS parts caused by exposure to protons with energies up to 40 MeV and a maximum dose of 1 MGy; and (2) tensile experiments to investigate the ionizing radiation effects on the mechanical properties of the FD-ABS. Preliminary results from irradiation performed at Texas A&M University Cyclotron Insitute show a significant drop in tensile strength as the radiation dose increased. The molecular weight and decomposition temperatures decreased as the radiation dose increased.

Tungsten has a high melting point and low sputtering rate, this makes it ideal for use in fusion reactor walls and divertors [1]. In this environment the material is exposed to temperatures up to 1000°C and 14.1MeV neutrons leading to 150dpa of damage in a lifetime [2]. For structural components (e.g. divertors) it is important to understand the effect that this extreme environment has on the mechanical properties throughout working lifetime. Previous work suggests that high irradiation doses can lead to embrittlement in tungsten [3].
To investigate this, Helium implantation of Tungsten with 2% Tantalum (WTa2%) (including the role of transmutants) was irradiated with He+ up to 3000appm at a temperature of 300°C [4]. The implantation was conducted using a range of energies (0.05-1.8MeV) and fluences (1-5x10¹⁵ ions/cm2) to produce a relatively flat implantation profile ~1.56μm deep.
This scale of damage therefore requires small scale characterisation to understand the mechanical properties that change during radiation damage. In this work we illustrate the use of nanoindentation and microscale cantilevers to find the changes in/to the elastic and plastic response of the material. We have developed a new geometry of cantilever in order to test material volumes that are removed from the edge of the sample. This methodology allows for observation of the elastic and plastic properties of irradiated material and opens up a wide range of possibilities for future work in terms of location specific characterisation.

[1]:- P. Norajitra et al Journal of Nuclear Materials 2007 367-370 pp 1416-1421
[2]:- N. Baluc et al Nuclear Fusion 2007 47 pp S696-S717
[3]:- N. Yoshida Journal of Nuclear Materials 1999 266-269 pp 197-206
[4]:- C. E. Beck et al Symposium HH – Advances in Materials for Nuclear Energy 2013, Vol 1514 pp. 99-104

Additive manufacturing (AM) refers to the technique of producing complex three-dimensional objects directly from a digital model through layer-by-layer computer-controlled deposition and consolidation of feedstock powder materials. The AM technique is emerging into a mainstream manufacturing process and has the potential to find widespread application in nuclear technologies. Inconel 625 is a Ni-Cr-based alloy widely used in nuclear reactors. In this study we investigated the relations between the microstructure and associated high-temperature oxidation properties of additively manufactured IN625. Using Laser Engineered Net Shaping (LENS) technology, we fabricated 1x1x1 cm cubes from argon-atomized IN625 powders. The microstructure of these LENS-processed IN625 consisted of primary dendrites and fine cellular grains. Characterization revealed a significant microsegregation of Nb and Mo at the boundaries of the fine grains due to local rapid solidification processes. Thermogravimetric analysis was then conducted to determine the intrinsic oxidation resistance of the LENS-processed IN625 samples exposed to dry air at 800 C and 1000 C for up to 100 h. The kinetics results indicated that the LENS-processed IN625 exhibited less resistance to oxidation as compared to wrought IN625. The results will be discussed in the context of Nb and Mo microsegregation to the grain boundaries detrimentally affecting the oxidation resistance of the LENS-processed IN625.

Understanding the behaviour of complex materials in complex environments over millennia is a challenge for both modellers and experimentalists but must happen to have confidence in the safety cases for permanent radioactive waste disposal. Current durability testing of ceramic materials (including glass, glass composite materials and cements) as hosts for radioactive wastes is based on overly simple materials and procedures including using powders or samples of defined size and shape, polished samples, distilled water, controlled flow rates, closely defined and constant temperatures, simple and inactive compositions (International Simple Glass, ISG) etc. While steps are being taken towards more realistic testing procedures to account for package corrosion products e.g. iron oxide from container, type of water in repository environments (salts), Engineered Barrier Systems (e.g. clay) and rock dissolution products etc. more is needed.
Recent work [1-2] has highlighted the need for improved standard testing procedures and modelling developments. In [1] the impact of partial water immersion on the International Simple Glass (ISG), a condition that is inevitable as a repository gradually becomes saturated with penetrating groundwater was examined. This revealed quite different degradation behaviour compared to standard tests in which the material is fully immersed. In particular, whilst in standard tests the system reaches a steady state with a very low alteration rate thanks to the formation of a protective gel layer, in partially-immersed tests this steady state could not be reached because of the continuous alteration from the condensate water film. The constant input of ions from the emerged part of the sample caused a supersaturation of the solution, which resulted in early precipitation of secondary crystalline phases. In [2] using highly radioactive ²³⁸Pu-doped and non-radioactive samples of borosilicate glass with chemical compositions and synthesis procedure similar to French standard SON68 glass showed dramatic differences in behaviour. Over 4 months the radioactive glass is fully covered by mechanically unstable alteration layer whereas the model glass remains virtually pristine. Such studies highlight the importance of understanding the evolution of (T, environment, activity) conditions with time and the inherent inhomogeneity of the wasteforms and how they change with time, and the need to work with active material.

This talk uses these, and other, data to highlight the need for improved durability testing of wasteforms.

[1] RK Chinnam, PCM Fossati and WE Lee, “Degradation of Partially Immersed Glass: A New Perspective,” J. Nuclear Mats. 503 56-65 (2018).
[2] BY Zubekhina, AA Shiryaev, BE Burakov, IE Vlasova, AA Averin, VO Yapaskurt and VG Petrov, “Chemical alteration of ²³⁸Pu-loaded borosilicate glass under saturated leaching conditions,” Radiochim Acta doi.org/10.1515/ract-2018-3097.

Nuclear energy offers many benefits including stable electricity generation and low carbon output; however, one of the major challenges of nuclear energy is the management of the used fuel. In Canada, the Nuclear Waste Management Organization (NWMO) has proposed a deep geological repository (DGR) solution where used Canada Deuterium Uranium (CANDU) fuel bundles are packaged into a used fuel container (UFC), which is then placed at a reference depth of 500 m underground in a geologically stable rock formation and surrounded by bentonite clay. The UFC consists of a structural steel containment vessel which is copper coated on the exterior surface for corrosion protection. Two methods are used to copper coat the UFC. The first method is electrodeposition (ED) process which is used to coat the shell and hemi-spherical factory components in a conventional manufacturing environment. After UFC is loaded with used fuel, the vessel is welded shut and the UFC corrosion barrier is completed by applying cold spray (CS) copper coating to the weld closure zone under radiological conditions. The as-deposited CS copper coating exhibits high hardness and low ductility. In order to restore ductility, local annealing (1 hour at 350°C) with a heating band is performed.

Since the Cu on the UFC are applied using two different methods, it is important to develop a better understanding of the structure-property-processing-performance relationships of these coatings. In this study, we apply various electron microscopy techniques including electron backscattered diffraction (EBSD), focused ion beam (FIB) and transmission electron microscopy (TEM) to establish the relationships between the coating processing conditions (ED, CS, annealing of CS-Cu) and microstructure of the coatings at different regions including the ED-Cu and CS-Cu interfaces, as well as the Cu particle-particle interfaces in the CS-Cu coatings. These structural characterization results will provide a basis for the ongoing mechanical and corrosion testing programs seeking to advance knowledge on strength and corrosion resistance of these Cu materials for long term storage in the repository.

In a deep geological repository scenario, oxidized uranium in aqueous systems will be stabilized as UO₂²⁺ (hexavalent uranium), as a consequence of tetravalent uranium oxidation by radiolytic byproducts. Uranyl cationic species (UO₂²⁺) in different complexation forms, such as uranyl oxyhydroxide hydrates are expected to be found at the whole pH range conditions. The importance of uranyl lies in its potential incorporation of trace radioelements onto secondary uranyl phases. In view of the difficulty of uranium chemistry in natural groundwater, it is worthy to improve speciation assessment techniques so as to characterize precipitated solid phases and understand chemical processes.
Recently Raman spectroscopy has been shown as powerful tool to analysis the speciation of various actinyl (UO₂²⁺_,NpO²⁺ and PuO₂²⁺) and to determine distribution of those elements which are more likely to be stable in a near-field groundwater environment. Therefore, the aim of this work is to develop a quantitative procedure to follow uranyl speciation in aqueous solution under repository conditions, and understand the bonding, coordination and local surroundings of uranyl species, which will brings light to understand and predict the precipitated solid phases formed (solubility-limiting phases). Hydroxide and hydroxo-carbonate complexes of uranyl will be spectroscopically identified in aqueous solution in the absence and presence of carbonate. The proposed identification procedure of uranyl complexation will give a precise basis to assign better estimates of the predicted solid phase stoichiometry combining solubility analysis with speciation and precipitation of U(VI).
In this work, it is shown the use of Raman spectroscopy adapted to the empirical analysis of different nuclear applications for uranium concentrations higher than 0.05M at ambient atmosphere, i.e. as monitoring tool for UO₂²⁺ precipitation as a function of pH studying UO₂(NO₃)₂x6H₂O stability in aqueous solutions representative of groundwater, in particular at ionic strength I = 0.02 – 0.4 M and pH from 7 to 13.2 and to evaluate the effect of gamma radiation fields. Therefore, the obtained results will provided a complete picture of secondary phase formations, as a result of corrosion of spent nuclear fuel in a deep geological repository.

The radiation damage tolerance of nuclear waste forms is dependent on the materials resistance to defect formation and its ability to accommodate the structural distortions that arise from defect formation. This study illustrates how the radiation tolerance of hollandite can be improved thorough compositional control of cesium stoichiometry. The barium (Ba) to cesium (Cs) ratio was varied in the tunnel sites referred to as the A site of the hollandite structure. Zinc (Zn) was substituted for titanium (Ti) on the B site to achieve the targeted stoichiometry with a general formula of Ba_xCs_yZn_x+y/2Ti_8-x-y/2O₁₆(0<x<1.33; 0<y<1.33). The tunnel cross-section depended on the average A site cation radius while the tunnel length depended on the average B site cation radius. The enthalpies of formation from binary oxides of Zn doped hollandite measured using high temperature oxide melt solution calorimetry were strongly negative, indicating thermodynamic stability with respect to their parent oxides. The formation enthalpies became more negative, indicating hollandite formation is more energetically favorable, when Cs was substituted for Ba across the range of Zn doped compositions investigated in this study. This hollandite series was exposed to heavy ion (Kr²⁺) irradiation at 27 °C, 100 °C, 200 °C and 300 °C followed by characterization with grazing incidence X-ray diffraction , transmission electron microscopy , and aqueous leaching tests.. The radiation tolerance increased at elevated temperatures with a critical amorphization temperature between 200 °C and 300 °C. Elemental leaching decreased with increasing cesium content. Irradiated samples exhibited twice the fraction cesium release of pristine samples.

The performance of advanced nuclear wasteform materials containing radionuclide absorbing nanoparticles such as zeolite, hexacyanoferrate (HCF), and salt-inclusion materials (SIMs) depends on hierarchical microstructures as well as thermodynamic and kinetics properties of ion exchange processes. For example, the wasteform morphology including particle size and particle aspect ratio, anisotropic mobility, and electrochemical potential affect the uptake kinetics of radionuclides. A physics-based mesoscale model of ion exchange in hierarchical materials is presented that takes into account the effects on ion exchange processes of inhomogeneous microstructures, anisotropic and inhomogeneous diffusivity of charged particles, electrochemical potentials, and operational conditions. Wasteforms consisting of zeolite or SIMs particles are taken as model systems. The thermodynamic and kinetic properties of diffusive ions are assessed using atomistic simulations and thermodynamic calculations. Simulation results showed that anisotropic mobility and larger aspect ratios of particles dramatically decrease ion exchange kinetics if the migration mobility is faster along the largest dimension anisotropic particles, such as needle-shaped particles. Electric fields associated with inhomogeneous ion distributions is an additional driving force that typically increases ion exchange kinetics depending on particle aggregation and ion mobility. The capability of the model for investigating the physics and mechanisms behind ion exchange in complicated microstructures and for predicting the effects of microstructure on ion exchange kinetics will be presented.

Early studies on the onset of flash sintering (FS) for UO₂ have identified that flash is induced by a critical field, at a critical onset time (incubation time) and they are both shown to be dependent on temperature and O/M. UO_2+x has been shown to flash at lower critical fields, compared to UO_2.00, and a bifurcation stability model has been used to model the Joule heating effects [1] taking place during flash onset with reasonable agreement with experimental data [2]. Densification of stoichiometric and hyper-stoichiometric UO₂ has been achieved via FS and has resulted in 81% and 92% theoretical density (TD) pellets, respectively [3]. Still, the resulting microstructures were significantly cracked and featured porosity. The latest studies on FS of UO₂ have concentrated on producing high density pellets with structural integrity and uniform microstructure. FS using the current control method has been implemented to minimize thermal shock and achieve control in the resultant microstructures. The current control method has been shown to produce more structurally sound pellets, although additional difficulties with FS of UO₂ result from atmosphere and subsequently stoichiometry control of the pellet. More specifically, during FS of UO_2+x using Direct Current (DC), an O/M gradient can develop due to the concentration of O₂ close to the anode, resulting in hyper-stoichiometric material in the vicinity which can sinter at a faster rate. This produces pellets with differential densification along their length. To overcome this issue, a pseudo-Alternating Current (pseudo-AC) has been used and it has been shown to produce pellets with slightly higher densities and more homogeneous macroscopic characteristics. Finally, the pseudo-AC and current controlled methods have been applied to FS of doped UO₂ and the results are compared to doped UO₂ that has been sintered conventionally.

[1] J. G. P. da Silva, et al., “A dynamic bifurcation criterion for thermal runaway during the flash sintering of ceramics,” J. Eur. Ceram. Soc., vol. 36, no. 5, pp. 1261–1267, 2016.
[2] A. M. Raftery, J. G. P. da Silva, et al., “Onset conditions for flash sintering of UO2,” J. Nucl. Mater., vol. 493, pp. 264–270, 2017.
[3] J. A. Valdez, D. D. Byler, E. Kardoulaki, et al., “Flash sintering of stoichiometric and hyper-stoichiometric urania,” J. Nucl. Mater., vol. 505, pp. 85–93, 2018.
[4] D. Sprouster, E. Kardoulaki, et al., “In situ X-ray Characterization of Uranium Dioxide During Flash Sintering,” Materialia, vol. 2, pp. 176-182, 2018.

In this report, we will review the crystal growth and scintillation properties of simple and mixed cation and mixed anion ternary halides that belong to the ABX3, AB2X5, A4BX6 families (A – one or more alkaline metals, B – one or more alkaline earth metals, X – one or a mixture of chlorine, bromine or iodine). They are of interest for scintillation applications due to their relatively high density, congruent melting, large band gaps, and availability of a divalent lattice site to accommodate a Eu2+ luminescence activator. They have attractive spectroscopic performance that is competitive with the industry standard, NaI:Tl, or state-of-the-art SrI2:Eu and LaBr3:Ce scintillators, and therefore are suitable for use in portable radiation detectors for radio-isotope identification. Details of compositional uniformity, growth defects, and gamma-ray spectroscopic performance as a function of Eu concentration, crystal volume and growth parameters are presented to provide understanding of the best achievable performance. Particular attention is paid to thermal and structural properties to design the crystal growth parameters via the vertical Bridgman method. Additionally, the potential for scale-up is evaluated by simultaneous growth of multiple large size crystals (1” – 1.5” in diameter) using the Multi-Ampoule Growth Station.

Gamma-ray detection and spectroscopy is the quantitative analysis of gamma energy spectra, and is of critical importance in many applications, such as nuclear safeguards, nuclear forensics, and many more. Recently, single-crystalline, hybrid (and inorganic) perovskites have been proposed as new semiconducting materials for gamma-ray spectroscopy. In just a few short years, early reports have shown promising results for perovskite-based semiconductor gamma detectors. Some of the most impactful results in this new application for perovskite materials include: CsPbBr₃ showing gamma-ray spectra with a resolution of 3.8% at room temperature for a 662 keV ¹³⁷Cs source at room temperature,¹ MAPbI₃ showing gamma-ray spectra with a resolution of 6.8% at a lower temperature of 2 °C for a 122 keV ⁵⁷Co source,² and MAPbBr_2.94Cl_0.06 showing gamma-ray spectra with a typical resolution of 12% for a 662 keV ¹³⁷Cs at room temperature.³ Although first reports for this class of materials have shown significant results with high-resolution gamma detectors, many new issues have arisen for this new application of radiation sensing.

High electric field device stability and ion migration are deeply explored to understand the effects of high bias applications in hybrid perovskites and ways to suppress/eliminate the negative effects towards high-resolution, room temperature gamma-spectroscopy. In a Cr/MAPbBr_2.85Cl_0.15/Cr device, low electric fields (≤25 V/mm) pulses with extremely low rise times (average of 30 to 65 µs). In order to increase the rise time, the device needs to be operated at higher electric fields, (≥ 50 V/mm) which increases the rise time to an average of 15 µs. However, at high electric fields, the signals become noisy in roughly 10 seconds to 1 minute. We directly probed the cause of the noise using long-term high voltage biasing, while studying the hysteretic behavior via IV characteristics. A new Cr/MAPbBr_2.85Cl_0.15/Cr exhibits pure Ohmic behavior with negligible hysteresis between -200 to +200 V (E = 100 V/mm). However, after biasing the device for only 10 minutes at +200 V, a large hysteresis loop is observed. After letting the device rest in the dark for two days, the hysteresis loop becomes even larger, rather than restoring to its original properties, showing that the interfacial damage caused is non-reversible. Using the same electrodes in the device architecture allows us to use a technique termed voltage cycling to avoid permanent interfacial damage while operating the detectors. The signals from positive and negative polarity are roughly the same in the metal/semiconductor/metal device. Therefore, when noise appears, we are able to switch polarity to eliminate the noise and continue counting for gamma spectra. Cooling the detector to temperatures of -30 °C further increases the speed of the detector pulses by about 30%. However, the standard deviation in the pulse heights and rise times are still quite large, resulting in low resolution for the detectors with about 30-35% resolution for 59.6 keV γ-rays and 20-25% resolution for 662 keV γ-rays. Further development in hybrid (or non-hybrid) perovskite materials, such as CsPbBr₃ and MAPbI₃, should prove beneficial, as they are more stable under high electric fields required to achieve high resolution γ-ray spectroscopy. Also, further development in areas such as interfacial engineering and pixelated detector designs will be the next required steps in investigating the improvement for single crystalline, perovskite radiation detectors.

References
1. He, Y., et al., High spectral resolution of gamma-rays at room temperature by perovskite CsPbBr3 single crystals. Nature Communications, 2018. 9(1), 1609.
2. He, Y., et al., Resolving the Energy of γ-Ray Photons with MAPbI3 Single Crystals. ACS Photonics, 2018. 5(10), 4132-4138.
3. Wei, H., et al., Dopant compensation in alloyed CH3NH3PbBr3−xClx perovskite single crystals for gamma-ray spectroscopy. Nature Materials, 2017. 16, 826.

Distributed ionizing radiation dosimetry is crucial in diverse security areas with significant environmental and human impacts such as nuclear waste management, radiotherapy, or radioprotection devices. We present a fast, real-time dosimetry detection system based on flexible oxide thin-film transistors that show a quantitative shift in threshold voltage of up to 3.4 V/gray upon exposure to ionizing radiation. The transistors use indium-gallium-zinc-oxide as a semiconductor and amultilayer dielectric based on silicon oxide and tantalum oxide. [1] Our measurements demonstrate that the threshold voltage shift is caused by the accumulation of positive ionization charge in the dielectric layer due to high-energy photon absorption in the high-Z dielectric. The high mobility combined with a steep subthreshold slope of the transistor allows for fast, reliable, and ultralow-power readout of the deposited radiation dose. The order-of-magnitude variation in transistor channel impedance upon exposure to radiation makes it possible to use a low-cost, passive radiofrequency identification sensor tag for its readout. In this way, we demonstrate a passive, programmable, wireless sensor that reports in real time the excess of critical radiation doses. [2]

[1] T. Cramer, A. Sacchetti, M. T. Lobato, P. Barquinha, V. Fischer, E. Fortunato, R. Martins and B. Fraboni,“Radiation-Tolerant Flexible Large-Area Electronics Based on Oxide Semiconductors,” Adv.Electron.Mater., vol. 2, no. 7, pp. 1–8, 2016.
[2] T. Cramer, I. Fratelli, P. Barquinha, A. Santa, C. Fernandes, F. D’Annunzio, C. Loussert, R. Martins, E. Fortunato, B. Fraboni, “Passive radiofrequency X-ray dosimeter tag based on flexible radiation sensitive oxide field effect transistor”, Sci.Adv., vol. 4, eaat1825, 2018.

We study the degradation of thermal diffusivity in self-ion implanted tungsten, and utilise this degradation to predict point defect densities. The defect density predictions are consistent with previous transmission electron microscopy (TEM) and molecular dynamics (MD) studies of self-ion implanted tungsten.
Tungsten, due to its high melting point, high thermal diffusivity and low sputtering yield, is the prime candidate for plasma facing armour components in future fusion reactors. Armour components in the so-called divertor, the exhaust of the fusion reactor, will receive the highest heat and radiation loads in the reactor. The damage created in the tungsten matrix as a result of intense neutron irradiation and plasma exposure decreases its thermal diffusivity, reducing its heat removal capability. Self-ion implantation is a convenient, low cost proxy to neutron irradiation to study the displacement damage created by neutrons. It efficiently singles out the displacement damage from other effects such as transmutation. However self-ion implantation even at energies as high as 20 MeV create damage layers that are only a few microns thick. The transient grating spectroscopy method used in this study enables us to measure the thermal diffusivity of these thin layers with high accuracy. In this study, a simple kinetic theory model is used to predict the point defect densities from the measured changes in thermal diffusivity as a function of ion dose. These predictions are compared to dislocation loop and point defect densities observed in TEM and predicted by MD.
20 MeV self-ion implanted tungsten, over a dose range of 0.0001 – 10 dpa, shows a significant degradation in thermal diffusivity above doses of 0.001 dpa. This degradation progresses up to 0.01 dpa, beyond which the decrease in thermal diffusivity saturates. Full saturation is achieved by 0.1 dpa, and the thermal diffusivity ceases to deteriorate with additional implantation. The reduction in room temperature thermal diffusivity at this damage level is 55% of the value of pristine tungsten.
The extracted point defect densities from the TGS data are about 2 orders higher than anticipated when considering solely defects visible in TEM. By combining TEM observations with predictions from MD, the density of point defects associated with defect clusters too small to be seen by TEM can be estimated. Once these “invisible” defects are accounted for, the total point defect concentration agrees to within an order of magnitude with that inferred from the TGS measurements of thermal diffusivity.
The reduction in thermal diffusivity observed in this study provides a much more complete picture of the degradation of thermal diffusivity with increasing damage dose. It also suggests that the diffusivity degradation due to the displacement damage from neutron irradiation will saturate beyond 0.1 dpa, which is an important result for predicting the in-service performance of armour components in future fusion reactors. The defect prediction results demonstrate the novel capability of TGS in quantitatively determining defect densities in irradiation-damaged materials. It also shows that the finely distributed smaller defects affect material properties such as thermal diffusivity to a larger extent than the bigger dislocation loops that are visible to TEM.

Ferroelectric (FE) materials show a spontaneous polarization, reversibly switchable with an electric field, in addition to large dielectric, pyroelectric and piezoelectric response. Such properties make them attractive for fulfilling multiple functionalities in MEMS devices as sensors, actuators, and energy harvesting units. Traditionally, lead zirconate titanate (Pb[Zr_1-xTi_x]O_3, PZT) has been the most used ferroelectric in piezoelectric MEMS. However, relaxor-FE bulk single crystals, such as (1-x)Pb(Mg_1/3Nb_2/3)O₃-xPbTiO₃, (1-x)PMN-xPT, exhibit even larger electromechanical response than ceramic PZT, when cut and poled along the (001) direction with compositions on the rhombohedral side of the morphotropic phase boundary (x~0.32).
Lead-based materials are often radiation-hard, and numerous studies have shown PZT’s exceptional capabilities of retaining functional response after radiation exposure. The radiation hardness enables applications in high radiation exposure environments (e.g., aerospace, medical physics, x-ray/high energy source measurement tools, continuous monitoring of nuclear power plant applications, radiation disaster relief, etc.). However, PZT films still suffer from response degradation, especially for radiation doses at, or superior to 5 Mrad(Si). The high B-site chemical heterogeneity and disorder of PMN-PT is expected to result in a more radiation-hard material than PZT. Such perceived advantages are offset by challenges in processing pore- and crack-free PMN-PT thin films. Here we report on the radiation hardness of 0.7PMN-0.3PT thin films with varying microstructures, as modified through processing parameters.
All films were deposited on platinized Si substrates through chemical solution deposition (CSD). A 0.3M precursor solution of PMN-PT was prepared through a 2-methoxyethanol route and deposited onto the substrate via spin coating. A PbTiO₃ seed layer was used to induce (100) crystallographic orientation in PMN-PT films. Heat treatment conditions were modified in order to control porosity and grain morphology. Specifically, the samples were pyrolyzed at 250 to 430 °C for 1 to 5 minutes, and annealed at temperatures of 700 to 800 °C for 1 or 2 minutes. The deposition and heat treatment process were repeated in order to obtain final film thicknesses up to 500 to 600 nm. All samples showed strong (100) perovskite orientation with Lotgering factors > 80%, as evidenced by X-ray diffraction patterns. The samples were irradiated using a ⁶⁰Co source with ionization doses ranging from 0.2 to 10 Mrad(Si). A zero-dose sample was used for aging reference and background radiation control. Dielectric, polarization, and piezoelectric responses were conducted using sputtered Pt top electrodes, before and after irradiation.
In general, higher crystallization temperatures led to larger grains and columnar grain growth. All PMN-PT samples showed degradation in dielectric and saturated polarization within 5%, and in piezoelectric coefficients within 15%, within the radiation dose range used. Some films showed improvement in response up to 5% at lower doses, possibly due to annihilation of defects through low radiation exposure. Greater radiation hardness and lower functional response was observed in films with smaller, stacked grains, compared to films with larger, columnar grains. The influence of grain microstructure and porosity on both functional response and radiation-hardness, and a phenomenological model to quantify defect-defect interactions in irradiated functional materials will be discussed.

In the past five years, it was only hypothesized that morphological features could be used as signatures for nuclear forensics. The physicochemical connection between the processing parameters such as precipitation conditions, calcination temperature, and the composition of the starting material was generally thought to affect the morphology of the resulting material. However, quantifying these features was the limitation that prevented this signature from being utilized. In the past two years, we have published multiple papers demonstrating quantitative morphological analysis of UO₃, U₃O₈, and UO₂. To move from qualitative morphological analysis to quantitative analysis, the SEM images were processed using the Morphological Analysis of MAterials (MAMA) particle segmentation software. Following particle segmentation, the MAMA software was used to calculate morphological particle attributes for each of the segments. We generally segment enough images to acquire several hundred particles for statistical analysis. Our initial studies focused on synthetic parameters from the uranyl peroxide route in which we showed quantitative differences in the particle morphology of Am-UO₃ and U₃O₈ based on the calcination temperature.

We have also shown that incorporation of impurities with the U-oxide will alter the final product morphology. For example, four impurities: calcium, magnesium, vanadium, and zirconium were separately introduced to pure uranyl nitrate hexahydrate from stock solutions. Following the addition of the impurities, the solutions were precipitated using hydrogen peroxide. Incorporation of the impurity was quantified using inductively coupled plasma mass spectrometry (ICP-MS). The uranyl peroxide precipitate was then calcined to U₃O₈ at 800°C. ICP-MS was used again, to evaluate the incorporation of the impurity in the calcined uranium oxide. SEM was used to investigate morphological changes, and SEM-Energy dispersive spectroscopy (EDS) was used to identify any concentrated regions of the impurity in the SEM images. Results will be discussed highlighting the potential of using particle morphology for nuclear forensics, but also highlighting the challenges that need to be overcome for this to be a viable signature of processing history.

Nuclear materials age in service, with their properties changing as a combined result of how they were produced and to what they were exposed. This aging, which can be due to a combination of thermal and irradiation origins, sometimes results in degraded performance which can lead to failure. It also often leaves fingerprints, measurable either directly or indirectly. In this talk we will explore the use of transient grating spectroscopy (TGS) as a method to both assess the health of key light water reactor (LWR) components due to spinodal decomposition, short range ordering, and phase precipitation, and to reconstruct their irradiation history. Changes in thermal, elastic, and acoustic material properties are indirectly linked to the quantities of interest, such as Charpy impact energy via surface acoustic wave peak splitting for structural materials, and irradiation dose history to reconstruct reactor usage, particularly for additively manufactured materials. This wide variety of examples highlights the vast, unexplored space where non-destructive, indirect techinques can far more rapidly uncover new science and assist industry in nuclear materials forensic and health assessment issues.

Metal-Organic Frameworks (MOFs) are emerging as a novel class of hybrid materials for the purpose of radionuclide sequestration. Significant research has been reported on MOFs for applications such as gas storage, sensing and heterogeneous catalysis; however, radionuclide incorporated MOFs are relatively unexplored. The unique and desirable properties of MOF structures like high porosity, modularity, and synthetic diversity make them attractive candidate materials to contain the radionuclides present in the nuclear wastes. Density Functional Theory (DFT) calculations were used to determine the favorability of ion-exchange (at the metal node) of relevant radionuclides in various parent MOF clusters. A range of DFT methods, including various flavors of exchange-correlation functionals, DFT+U, relativistic effects, and magnetic effects were employed to establish robustness in results. The electronic structures of MOFs are also determined and are correlated with experimental data.

Thermal transport in insulators is of great interest in high temperature nuclear reactors application. External irradiation by swift heavy ions (SHIs) such as Bi at hundreds of MeV energy is used to simulate the fission product impact in nuclear and model materials like LiF with well-known structural response to dense excitation effects [1, 2].
We report the results of color centers formation analysis in bulk lithium fluoride crystals irradiated at room temperature by 98 MeV, 244 MeV, 465 MeV and 710 MeV Bi ions at fluence 10¹¹ cm^-2 and 710 MeV ions at fluencies ranging from 10¹⁰ to 10¹³ cm^-2 using optical absorption (OA) and photoluminescence (PL) spectroscopy.
The irradiation of LiF samples was carried out at the U-400 FLNR JINR cyclotron (Dubna, Russia). The proliferation of ion-induced F (245 nm) and F₂ (445 nm) color centers were studied by OA spectroscopy. As the fluence of irradiation ions grows, the integrated absorption intensity of F-peak gradually decays, whereas that of F₂-peak broadens and increases, which indicates that color centers tend to aggregate. The corresponding defects densities were estimated using Smakula-Dexter formula [2].
Since F₂ and F₃ color center related bands in absorption spectra overlap, the complementary technique, PL spectroscopy, was performed. Specifically, the F₂ and F₃ color centers in irradiated LiF samples were distinguished by their characteristic peaks observed at around 530 and 670 nm, respectively. For the fluencies 10¹⁰ - 10¹³ cm^-2, the integrated fluorescent intensity of F₂ appears to dominate in the spectrum for the lower ion doses and drops to the level comparable to that of F₃ as the ion dose increases.
Cross-plane near-surface thermal conductivity of SHI irradiated LiF crystals was characterized by picosecond time-domain thermo-reflectance (TDTR) in the thermally probed region extending down to 290 nm, where the electronic stopping / ionization processes dominate over nuclear stopping. TDTR results show that thermal transport in this sub-surface domain irradiated by 710 MeV ions with the dose of 10¹² cm^-2 has degraded by the factor of ~ 2 compared to non-irradiated sample.
Theoretical analysis was based on Klemens analytical thermal transport model [3] by fitting into TDTR measured thermal conductivity values. Given TDTR probed sub-surface region overlapped with ionization depth determined by SRIM calculation, we conclude that defects are formed primarily by electronic energy loss / ionization. Measurement and simulation of nano-scale thermal transport degradation in irradiated LiF will be discussed with results of structural defects proliferation as determined by OA and PL measurements. Funding by MES RK grant AP05130446, state-targeted program BR05236454 and NU FDCRG grant 110119FD4501 is acknowledged.
References
[1] Manika I. et al. Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms 196, no. 3–4 (November 1, 2002): 299–307.
[2] Schwartz K. et al. Physical Review B 58, no. 17 (November 1, 1998): 11232–40.
[3] Klemens P.G. Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms 1, no. 2–3 (February 1, 1984): 204–8.