Symposium EN17 : Fundamental Materials Science to Enable the Performance and Safety of Nuclear Technologies

At ion energies typically used to emulate fast neutron damage, fission damage or alpha-decay damage in nuclear ceramics, the electronic and nuclear energy losses are often comparable, and local ionization along the ion path can affect damage production and evolution. Experimental and computational approaches are used to investigate the separate and combined effects of nuclear and electronic energy loss on radiation damage in ceramics relevant to nuclear applications. Defect production and damage accumulation have been investigated as functions of electronic energy loss, S_e, nuclear energy loss, S_n, the ratio of electronic-to-nuclear energy loss, S_e/S_n, and the damage energy, E_D. Experimentally, ion mass and energy are controlled to vary electronic and nuclear energy loss, and large-scale atomistic simulations that combine ionization-induced thermal spike and atomic collision processes are used to model these effects. The results demonstrate that electronic energy loss, typical of MeV ions, can lead to competitive damage recovery processes or additive damage production effects in many nuclear-relevant ceramics. An important factor in the effectiveness of the damage recovery or production processes for a single ion event is the spatial coupling of electronic energy dissipation (via electron phonon coupling) and damage energy dissipation (via elastic scattering events) along the ion trajectory. As the radiation damage evolves, electronic energy deposition density can increase, and the dissipation of electronic energy to the lattice can further anneal existing defects along the ion trajectory or interact synergistically with the defects to enhance damage production. These results have significant implications for interpreting and modeling the radiation response of nuclear ceramics in accelerated testing using MeV ion irradiation.

This work was supported by the U.S. DOE, BES, MSED.

Over the past several decades, cluster dynamics methods have become one of the prime modeling techniques used to predict microstructure evolutions during irradiation. This presentaton focuses on the critical question of the predictive capability of these types of methods. In particular two specific points deserve attention pertaining to the level of accuracy and calibration of cluster dynamics methods. First, a novel approach is introduced to couple cluster dynamics methods and discrete dislocation dynamics simulation tools in order to quantify the effects of dislocation content and arrangement on the effective sink strength used in tradiational non-spatially resolved cluster dynamics. In addition to uncertainties associated with sink strength, it has to be acknowledged that other quantities, used to calibrate cluster dynamics models, area slo subject to uncertainty. To address this latent question, a data analytics based approach is used to demonstrate how the use of statistically motivated analysis can precise the effects of the aforementioned uncertainty in the intrinsic defect properties on the prediction of the damaged microstructure state.

Apatite-structured materials have been considered for immobilization of a number of fission products from reprocessing nuclear fuel because of their chemical durability as well as compositional and structural flexibility. It is hypothesized that the effect of beta decay on the stability can be mitigated by introducing appropriate electron acceptor at the neighboring sites in the structure. Decay series ¹³⁷Cs → ¹³⁷Ba and ⁹⁰Sr → ⁹⁰Y → ⁹⁰Zr were investigated using a spin-polarized DFT approach to test the hypothesis. Apatites with compositions of Ca₁₀(PO₄)₆F₂ and Ca₄Y₆(SiO₄)₆F₂ were selected as model systems for radionuclides Cs and Sr incorporation respectively. Ferric iron was introduced in the structure as an electron acceptor. Calculated electron density of states suggests that the extra electron is localized at the ferric iron, which changes its oxidation state and becomes ferrous iron. The calcualtions show that there are minor changes in the crystal and defect structure of CsFeCa₈(PO₄)₆F₂ with Cs⁺ and Fe³⁺ substitutions undergoing Cs → Ba transmutation, and of Ca₃SrY₄Fe₂(SiO₄)₆F₂ with Sr²⁺ and Fe³⁺ substitutions undergoing Sr → Y → Zr transmutations. The results on calculated cohesive energy suggest that transmutations of Cs⁺ → Ba²⁺ and Sr²⁺ → Y³⁺ → Zr⁴⁺ in both apatite compositions are energetically favorable, which are consistent with the minor structure distortions. Stability improvement by incorporating ferric iron is significant with respect without variable valence ions. The results confirm the structural and compositional adaptability of apatites upon beta transmutations. The study suggests that apatite-structured materials could be promising nuclear waste forms to mitigate the beta decay induced instability, by incorporating variable valence cations such as ferric iron in the structure. The study demonstrates a methodology which evaluates the structural stability of waste forms incorporating fission products undergoing beta decay.

Thin film pyrochlores have attracted considerable attention for both their unique electronic and ferroic properties, as well as their radiation tolerance. These materials are able to accommodate radiation damage through the formation of defected fluorite and other structures. Furthermore, by substituting different cation species onto the B-site sublattice, it is possible to engineer various kinds of useful functional properties, but less is known about how such substitution in turn affects irradiation behavior. Here we describe a systematic study of defects generated in model La₂Ti_2-xZr_xO₇ single-crystal thin films that have been irradiated using 1 MeV Zr ions. We employ a combination of atomic-scale scanning transmission electron microscopy (STEM) and electron energy loss spectroscopy (EELS) to examine the local defect microstructure in these materials. Our experiments are informed by density functional theory (DFT) calculations, which offer insight into damage mechanisms and serve as inputs for novel multislice image simulations to interpret our STEM results. We show that this comprehensive characterization and modeling approach can help disentangle the complex interplay of structure, cation chemistry, and defect generation in this promising class of materials.

In mean field rate theory (RT) based models for predicting the microstructural response of materials to radiation exposure, dislocation sinks are typically represented as a homogeneous absorbing medium, and the rate of defect point defect capture at them is based on simplified solutions from assumed defect profiles for ideal dislocation configurations. Recent advances in discrete dislocation dynamics (DDD) have allowed the efficient calculations of the strain field generated by arbitrary dislocation microstructures using the fast Fourier transform. This work presents a coupling of spatially resolved RT to DDD, utilizing the strain fields produced by the latter to inform defect interaction energies for a high fidelity solution of the drift-diffusion problem in the former. In the RT, defect capture rates are computed locally, based on the defect concentrations in the immediate vicinity of dislocations rather than globally or homogeneously. This approach enables explicit calculations of the sink strengths, biases, and spatially dependent point defect super-saturations caused by heterogeneous arrangements of dislocations and the interacting strain fields they produce. Additionally, the rates of point defect capture can be used in DDD simulations to produce climb rates which include with the effect of strain interactions, defect kinetics, and spatial correlations in the microstructure.

The behavior of xenon (Xe) in nuclear fuel is of critical importance to nuclear fuel performance, because the diffusion and precipitation of Xe in fission gas bubbles influences both the amount of fuel swelling and the quantity of fission gas released to the fuel rod plenum. Despite decades of investigation, significant uncertainties exist regarding the underlying mechanisms controlling Xe diffusion, precipitation and release that impact predictions of fission gas swelling and release during both normal operation and transient conditions in accidents. Despite being a key determinant of fission gas effects, which control fuel performance, accurate physically-based models of intra-granular bubble evolution are still lacking in current models. This presentation will first introduce a multiscale modeling approach to simulate the behaviour of fission gas and intra-granular fission gas bubble populations. We will then review the current status of understanding of Xe diffusion mechanisms, and our recent results to investigate the re-solution behaviour of fission gas bubbles as a function of bubble size and xenon concentration (pressure), and finally show initial results of our cluster dynamics model for xenon bubble density. Our approach to multiscale modeling is based on an information-passing paradigm, which allows us to attack the complex fission gas behavior problem from both a “bottom-up” atomistic-based approach, as well as from a “top-down” continuum perspective that focuses on kinetic models of intra-granular gas bubble evolution. Simultaneously attacking such complex and inter-related diffusion and gas bubble evolution processes from both an atomistic and a continuum approach will minimize the risk of using just a single approach and further the prospects for scale bridging, or multiscale integration. The presentation will conclude by benchmarking our continuum clusters dynamics bubble evolution model to available experimental data on intra-granular bubble size and density as a function of temperature and burnup.

The fission gas release (FGR) is a key point that must be accurately assessed for fuel rod design and licensing under both normal and off-normal conditions. FGR is important input data in terms of both the radioactive source term relative to the consequences of a nuclear accident on the environment, and the driving force regarding fuel damage. Correctly predicting FGR under different specific conditions such as accidental scenario remains an important R&D goal. The present work deals with FGR mechanisms during a LOCA type transient. A new model to predict FGR is proposed; it is compared to experimental results and the possible phenomenon responsible of FGR is discussed.
The samples involved (14 different tests) come from a section of PWR UO₂ fuels irradiated up to 72 GWd/t_U. They were subjected to annealing tests on the MERARG facility (CEA Cadarache). They consist of a temperature ramp up to 1200°C at 0.2°C/s under inert atmosphere. The release occurs by bursts according to two main contributions (one between 600°C and 800°C and the second one after 1100°C). The largest FGR occurs at 1100°C and is due to a thermal activated process which is still thoroughly investigated.
The release burst between 600 °C and 800 °C cannot be explained by the same type of mechanism. According to a completely different way, FGR could be induced by the creation of defects themselves created by alpha auto-irradiation (i.e. during the cooling time of the fuel sample). FGR kinetics might be modelled thanks to a two-step process: a phase of gas clusters nucleation and growth followed by their release. This new simplified two-parameter model for describing aggregation kinetics can be borrowed from the Finke-Watzky mechanism. It is a process of slow continuous nucleation linked to an autocatalytic surface growth. Both steps happen simultaneously so that gas clusters are consumed and released.
When experimental data are fitted with the model, a very good agreement is obtained. Even though first results are promising, it is necessary to continue this study in order to confirm the preliminary conclusion.

As part of an effort to develop low enriched U-Mo fuels for application in research and test reactors, the Materials Management and Minimization Reactor Conversion (MMMRC) Program utilizes detailed high-resolution microstructural characterization techniques to characterize irradiated U-Mo nuclear fuel plates. This includes using techniques like transmission electron microscopy, atom probe tomography, electron energy loss spectroscopy, and electron backscattered electron diffraction to uncover the microstructure of U-Mo alloys, and other materials of construction, after irradiation. The data generated using these techniques is imperative for developing a fundamental understanding of the irradiation performance of this fuel under a variety of irradiation conditions. Information like that generated from this work is key for improving computer modeling of the fuel performance under irradiation. This presentation will discuss how recent results can be used to improve understanding of phenomena like recrystallization, grain growth, radiation stability, and swelling of irradiated U-Mo fuel and other fuel plate materials.

The detailed and quantitative understanding of mesoscale in-reactor degradation behavior of nuclear fuels is of critical importance as it is the evolution of the microstructure that has the most profound impact on bulk properties and in-reactor performance of nuclear fuels and thus on the development of next generation reactor systems. In this contribution, we will discuss the recent progress made in microstructural and microchemical characterization of different fuel systems, including mixed oxide (MOX), U-Mo, and U-Pu-Zr. The value of 3D techniques, such as focused ion beam (FIB) nano-tomography and atom probe tomography (APT), in nuclear fuels research will be demonstrated and the challenges in interpreting 3D tomography results on irradiated fuels addressed. In addition, microstructure-property relationship will be discussed and experimental data will be linked to ongoing modeling efforts.

Alloys of uranium (U), plutonium (Pu), and zirconium (Zr) are likely metal fuel candidates for sodium cooled fast reactors. However, our physical understanding of this ternary system is limited. In this work, we use a combination of mesoscale simulation and experimental characterization to study the U-Pu-Zr system. We present our phase field model that captures the behavior of the 13 dominant phases of the ternary system using published free energies for the system. We then demonstrate where comparison with experiments shows weaknesses in the free energies. Finally, we compare the model predictions with experimental data to show where future work is needed.

Uranium-zirconium (U-Zr) alloys are promising candidate fuel materials for the next generation fast reactors. In reactors, the constituent redistribution induced by the thermal gradient and radiation leads to the formation of various intermetallic phases in the fuels. Such phase evolution can significantly impact the thermal transport properties in fuels. Here a mechanistic model is developed to describe the thermal conductivities of U-Zr binary alloys at a wide range of temperatures and Zr concentrations. Different from other models in literature, this thermal conductivity model is based on the intrinsic and residual thermal resistivities. Thermal resistivities of pure uranium and pure zirconium as a function of temperature are estimated from quantum theories and the coefficients are calibrated from the experimental data. The thermal resistivities of U-Zr alloys of different Zr concentrations are determined by the interpolation between the pure U and Zr metals and a compensating term. This model predicts very reasonable thermal conductivities of U-Zr binary system for a wide range of Zr concentrations and temperatures, and its function forms may be applied to other binary or ternary metallic fuel systems.

Metallic nuclear fuels have been studied for decades in various types of reactors, including fast reactors, as well as research and test reactors. Recent improvements in the sectioning and specimen preparation of irradiated fuels and materials have allowed for the small-scale testing of highly irradiated systems. These miniaturized specimens possess a lower dose rate and fewer radionuclides, which enables the possibility of characterization at numerous facilities. Moreover, these improvements are coupled with advanced characterization techniques that allow researchers to gain access to length scales that were previously unattainable. This talk will focus on advanced synchrotron characterization techniques that enable microstructural characterization that spans from two to three dimensions. Radiation effects on atomistic ordering, phases, grains, microstrain, pore formation, and swelling will be discussed. Examples will be provided for both low and high fluence irradiated metallic fuels, including uranium molybdenum and uranium zirconium alloys.

Spent nuclear fuels include a variable amount of minor actinides incorporated within the UO₂ matrix as a result of the ongoing nuclear reactions during operation time. These actinides contribute to the radioactivity of the spent fuel, whose most long-living component is constituted by α-emission, lasting up to millennia and even farther. The accumulation of α-damage and radiogenic He affects the microstructure of the material and might result detrimental for the spent fuel integrity. For this reason, it is fundamental to know exactly how the spent fuel evolves, in order to accurately predict its behaviour over the long term.
To investigate such a slow evolution in a laboratory timeframe, a suitable way is the preparation of UO₂ surrogates doped with highly-emitting actinides that will produce high levels of damage and radiogenic He over a shorter periods. In this work, UO₂ disks containing tailored amounts of highly α-emitting ²³⁸Pu were produced by powder coprecipitation and sintering, and successively characterized. The coprecipitation route was chosen in order to obtain a homogeneous solid solution of the two oxides from the very beginning of the process, while the successive sintering process was tailored to mantain the homogeneous distribution of the α-dopant as well as a nearly complete densification (95%TD, compliant with the real nuclear fuel). The two chosen Pu percentages were 2.5 and 10% and were selected so that all the monitored evolving material properties come to saturation within 2 years.
The samples were then divided into 3 batches and stored at different temperatures: liquid nitrogen (approximately -195°C), room temperature (about 25°C) and wet storage conditions (around 200°C). Particular precaution was taken for the transportation between the storage and the actual characterisation. The choice of these three temperatures should allow differentiating the nature of the defects formed and their temperature-dependant evolution, in particular during spent fuel storage.
The samples evolutions were monitored by means of several characterization techniques: XRD (structure), TEM (microstructure), DSC (defect energy), LAF (thermal properties), Helium thermal desorption spectrometry, mechanical testing, and Raman spectroscopy. The outcome of this work should help predicting the long term behaviour of spent fuel.

This study aims to determine how concrete structures used in waste management scenarios have aged, and any effects on the structural properties. Particular focus is given to the effects of carbonation and gamma radiation, which are investigated through changes in the compressive strength, pore water chemistry, and microstructure. These two degradation mechanisms will be investigated individually as well as simultaneously.

Concrete is used throughout the nuclear industry, in both a radiation shielding role as well as a building material. Previous research conducted into the radiation tolerance of concrete has predominantly focussed on conditions representative of a nuclear reactor, such as neutron or neutron-gamma irradiation. Consequently, only a small proportion of papers have examined the sole effects of gamma radiation on concrete, a situation applicable to waste management structures. Furthermore, the majority of concrete structures present across the U.K. civil nuclear estate are over 20 years old. The vitrified product store is such a structure, housing high level waste incorporated into a glass matrix and stored inside stainless steel canisters. As a result, the store is exposed to elevated temperatures, gamma radiation and potentially carbonation.

Conditions representative of the vitrified product store’s lifetime will be simulated using an accelerated carbonation chamber and an elevated gamma dose rate. Fresh samples are batch produced to replicate the specification of the concrete used for the vitrified product store. A comparison will be made to historic samples recovered from the structure upon construction, some of which have been inside the store accumulating a dose for over 20 years.

Uranium-rich U-Zr alloys such as U-10Zr have been used in previous fast reactor experiments and are materials of interest for future fast reactor designs due to their high fissile density, favorable thermal properties, and ability to operate at higher burnups than traditional oxide fuels. A necessary step in validating U-10Zr and other U-Zr alloys for use in these future reactor designs involves understanding how the properties of the alloys change throughout their operational life-times. This includes the effects of thermal and mechanical stresses, as well as effects produced by the accumulation of radiation damage. Radiation damage events produce displacement cascades that result in the formation of many defects, some of which will anneal over a short time period, while others will remain in the lattice. Vacancy formation energies (VFE) play a significant role in simulations of radiation damage events, due to the impact of monovacancies on the diffusion of defects that were formed during the damage event. The VFE of body-centered-cubic uranium-zirconium alloys was calculated by two methods. In one case, atomistic configurations of a random U-Zr solid solution were generated, an atom was removed, and the vacancy formation energy was calculated. In the other case, the U-Zr solid solution was represented by a series of special quasirandom structures (SQS), before creating the defect and calculating the vacancy formation energy. The impact of nearest neighbor configurations on VFE for U-10Zr alloy was also examined. The molecular dynamics simulation code Large-scale Atomic/Molecular Massively Parallel Simulator (LAMMPS) was used to examine the differences in vacancy formation energies calculated for random structures produced by LAMMPS and SQS generated by the "mcsqs" code of the Alloy Theoretic Automated Toolkit (ATAT). Results indicate that there is not a statistically significant difference in the vacancy formation energies of alloys produced by random atom placement and alloys generated as SQS. Examination of the impact of nearest neighbor configurations on vacancy formation energy in the U-10Zr alloy showed that the vacancy formation energy did not vary significantly for configurations that were statistically likely to form in a random alloy. As the Zr concentration in the bcc solid solutions increases, vacancy formation energy for zirconium removal increases until a maximum value is reached around 50% Zr, then decreases to a value very close to the initial value. For uranium removal, vacancy formation energy initially decreases as the percent of Zr increases from 10%, reaching a minimum around 60 or 70% Zr, then increasing towards the uranium VFE in pure bcc uranium. These results can be used to parameterize higher length scale models of radiation damage and microstructural evolution of metallic U-Zr nuclear fuel.

The phase equilibria of advanced technology nuclear fuel candidates and interactions with alternate options to zirconium-based cladding is being explored. Uranium silicide (U₃Si₂) and silicide nitride (U₃Si₂-UN) composite fuels are the most promising contenders for the future, whereas ferritc alloys such as FeCrAl and SiC/SiC composite materials are under investigation for the cladding. The uranium density of the silicide and U-Si-N composite is advantageous in overcoming the neutron penalty imposed by the FeCrAlY material. This work focuses on thermochemical modeling and experiment to explore current limitations within the literature concerning the U-Si-N and U-Si-FeCrAlY phase space. Experimental techniques to investigate the U₃Si₅-USi₂ region include arc-melting and characterization by SEM-EDS and XRD, which is also extended to ternary nitride compositions. A multiscale modeling approach is used to explore the U-Si phase space including DFT, evolutionary algorithms and cluster expansion to identify stable structure types. Density functional theory is also utilized for formation energies of the U-Si-N ternary as well as the U-Fe-Si phase space to include FeCrAlY cladding compositions. These first principal calculations support thermodynamic CALPHAD assessments these ternary systems, with the cumulative results serving as input for higher order fuel peformance evaluation.

This research is being performed using funding received from the DOE Office of Nuclear Energy's Nuclear Energy University Programs

Oxide dispersion strengthened (ODS) ferritic steels are the most promising candidate structural materials for next-generation nuclear energy systems due to their high temperature stability and strength, and resistance to radiation-induced swelling. The manufacturing technology of ODS ferritic steels is based on powder metallurgy routes commonly involving mechanical alloying (MA) followed by consolidation, such as hot isostatic pressing (HIP) or hot extrusion. In order to enable the safe and efficient operation of current and future nuclear energy systems, improving their powder metallurgy techniques is important to further secure the material soundness at operating temperatures above 923 K. In this work, nano-micro structure control of high-Cr MA/ODS ferritic steels has been investigated to improve their ductility at elevated temperature. Four kinds of Al-Zr added high-Cr ODS ferritic steels have been produced by MA followed by vacuum hot pressing (VHP) or hot extrusion at a temperature of 1423 K. Argon gas is usually used as a milling atmosphere of MA for the manufacturing of ODS ferritic steels. In our group, hydrogen gas is also used as an MA atmosphere. All alloyed powders and solidified steels, processed in argon and hydrogen during MA, are denoted as MA/ODS-Ar and MA/ODS-H. As for alloyed powders, MA/ODS-Ar and MA/ODS-H powders after MA for 48 h exhibit roughly a monomodal particle size distribution (PSD) with size ranging from 1.5 to 55 µm. When MA is performed in hydrogen, the PSD peak width becomes narrower and the peak position shifts toward smaller sizes. As for solidified steels produced by VHP, hydrogen entrapped in the powder particles during MA inhibits phase segregation and grain growth and gives an uniform microstructure in MA/ODS-H steels, which depends on the size and PSD of the alloyed powder. On the one hand, coarse aluminum oxide inclusions (<2 µm in size), medium aluminum nitride (<1 µm in size), and fine zirconium oxide particles (several tens to several hundreds nm in size) exist in/on prior particle boundaries (PPBs) of MA/ODS-Ar steels. As a result, the size of those dispersoids is notably reduced by changing the MA atmosphere from argon to hydrogen, which is attributed to the redox reaction that occurred between dispersoids and hydrogen or other combustible gases during VHP. In addition to this, as for solidified steels produced by hot extrusion, no bubble formation is found at all in MA/ODS-H steels, while argon bubbles (<100 nm in size) can be observed in MA/ODS-Ar steels. Tensile tests conducted at 973 K indicate that an efficient improvement in high temperature ductility of the high-Cr MA/ODS ferritic steels can be achieved by the nano-micro harmonic structure control, suggesting that may be a critical technology for the R&D of ODS ferritic steels.

We developed shielding concrete as a building material that shields the influence from radioactive materials and radiation. This concrete is a material compounded with Anti-Sievert® (high density ceramic material). Conventionally, when Anti-Sievert® is blended, concrete requires high concentration of water. Therefore, sufficient concrete strength could not be realized.
We developed Anti-Sievert® concrete blended with high performance water reducing material and optimized the formulation of cement, water, aggregate, sand, Anti-Sievert® 210, Anti-Sievert® 216 as a material constituting concrete ¹⁾. As a result, we realized shielding concrete with high fluidity to realize construction suitability by concrete pump. The Anti-Sievert® Concrete has a shielding function of about 3 times as compared with ordinary concrete for X-ray (200 kV) and shows strength exceeding 50 N / mm².
In this presentation, we will show the results of the durability by estimating the construction of a nuclear explosion shelter using the Anti-Sievert® concrete. This shelter has the shielding effect with the Anti-Sievert® concrete and utilizes the characteristics of high strength material exceeding 50 N / mm². In this presentation, we report the X - ray shielding function and strength simulation of the Anti-Sievert® concrete shelter. These results of the simulations show that the Anti-Sievert® concrete is a practical material having sufficient strength as a nuclear explosion shelter and has excellent X-ray shielding function.

１）I. Sato ，T. Hatakeyama, Y. Yamashita, Y. Mori, S. Hashimoto, I. Hatsumura, Y. Katakami, T. Nishitsunoi, T. O. Hirano, Development of X-ray Shield Concrete for Secure Safety Against the Impact of the Fukushima Accident, MRS Fall meeting 2017

One of the main challenges remaining on the development of TRISO fuel particles is the understanding of the diffusion of Ag through SiC. Although Ag diffusion does not appear to be controlled purely by changes in chemical composition, Pd is one element that appears to participate in the diffusion of fission products. In this work we show that Pd is capable of transforming the grain boundary character distribution in SiC, leading to a larger concentration of interconnected high-angle grain boundaries, that could enhance the diffusion of other elements. We also show that the globular structures generally seen during the diffusion process are affected by the cooling rate during heat treatment and that Pd can change the composition of the grain boundary as it diffuses through SiC, effectively leading to a grain boundary complexion transition.

High Voltage CMOS (HV-CMOS) Monolithic Active Pixel Sensor (MAPS) is currently one of the most promising candidates for the future upgrade of large detector systems due to many advantages such as being able to withstand higher luminosity, having better spatial and energy resolution, faster response rate and reducing the material budget and manufacturing complexity [1,2]. This paper presents a new Technology Computer Aided Design (TCAD) model of a completed pixel of a HV-CMOS MAPS including the detecting region and readout circuitry. Two types of readout electronics, a simple source follower amplifier and an integrated charge amplifier are implemented, studied and compared. The TCAD model is used to investigate and compare the electrical characteristics and performance of the sensor design with different dimensions, doping profiles, and bias conditions. The responses of the sensor due to incidence of a minimum ionization particle (MIP) at different energy levels are also considered. The result shows that at 120 V bias, the pixel is fully depleted and has a clear response for a MIP which generates approximately 7500 electron-hole pairs (ehps) when penetrating through 100 μm of Silicon. Both readout amplification designs show distinct signals. The integrated charge amplifier is more sensitive and suitable for detecting low signals. A noticeable output is produced for an incident particle that produces only 1000 ehps with the present design.
References
[1] I. Peric, "A novel monolithic pixel detector implemented in high-voltage CMOS technology," in 2007 IEEE Nuclear Science Symposium Conference Record, 2007, pp. 1033-1039.
[2] I. Perić, P. Fischer, C. Kreidl, H. H. Nguyen, H. Augustin, N. Berger, et al., "High-voltage pixel detectors in commercial CMOS technologies for ATLAS, CLIC and Mu3e experiments," Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, vol. 731, pp. 131-136, 2013

Designing radiation tolerant materials is one of the primary challenges associated with advanced nuclear energy systems. As one type of potential materials, pyrochlore materials, encompassing a wide range of chemistry which is coupled to a remarkable variation of properties, are important in numerous technological applications such as catalysis, piezoelectricity, ferro- and ferrimagnetism, luminescence, giant magnetoresistance, and resistance to radiation damage. Some experiments with nanocrystalline scintillating rare earth oxides and rare earth fluorides on the literature have shown that in some cases nanoscopic dimensions provide essential improvement of the most important scintillation parameters: light yield, kinetics of scintillations, radiation hardness, etc. Hence, in this study, based on a combined co-precipitation and molten-salt synthesis method: nanoparticles of yttrium, lanthanum, praseodymium, gadolinium, erbium and lutetium hafnates were synthesized at 650°C. Of the compositions irradiated with different doses of γ-rays, yttrium, praseodymium, gadolinium and erbium hafnates proved to be the most chemically stable samples, maintaing their initial crystal structure throughout the irradiation process even to the highest expsoure, i.e. 12800 Gy. While further investigation is still undergoing, e.g. on the particle size effects, these rare-earth hafnate nanocrystals have demonstrated with desirable performance due to their robust chemical stability and nanoscopic dimension in radioactive environments, their high thermal stability, and their natural structural compatibility with radionuclide species.

Pressure-induced structural modifications in high-entropy alloys with varying Al content, Al_xCoCrFeNi (x=0.1, 0.3, 0.75, 1.5), have been investigated at pressures up to ~50 GPa by synchrotron X-ray diffraction and transmission electron microscopy (TEM). In Al_xCoCrFeNi compounds with x≧0.3, all of which exhibit initial pure fcc structures, the proportionality between the Al content and the transformation pressure is observed. This is attributed to the large size of Al atoms relative to those of the other constituent elements, which leads to more structural distortion in Al_0.3CoCrFeNi and subsequently an increase in the formation energy of the stacking faults. High-resolution TEM results show the variation of the stacking sequence from ABCABC (fcc) to ABABAB (hcp) in Al_0.1CoCrFeNi following exposure to high pressure. In Al_0.75CoCrFeNi, which exhibits an initial dual-phase structure, the result again shows the transformation to an hcp phase despite its higher Al content, which might be due to the presence of the bcc phase that is more amenable to the pressure-induced phase modification. However, the trend of transformation inhibition by increasing Al content is again observed, with Al_1.5CoCrFeNi retaining its initial structures up to the highest pressure achieved.

As interest in the microstructural evolution of nuclear fuels increases, there is motivation to develop computationally efficient mesoscale models consistent with CALPHAD treatments to understand these processes. This work describes development of multiphysics phase-field models which incorporate thermodynamic potentials as driving forces for transport phenomena and phase evolution while controlling implicit interfacial energy contributions. The model is discussed with examples of overpressurised intragranular fission gas bubble migration, and multiphase Al-Mg interdiffusion for advanced research reactor fuel.
Also discussed is the newly developed Included Phase model applied to intergranular fission gas bubble coalescence and percolation on the grain boundary network. This model achieves a substantial reduction in computational expense by solving the 3D phenomenon using a 2D model embedded in 3D space using a projection technique.

Composite materials with outstanding properties such as thermo-mechanical stability and radiation resistance can be engineered tailoring the internal structure and composition of the interfaces between the constituents of the system in addition to the usual compositional design of the bulk phases. These buried interfaces cannot be directly patterned by the usual surface techniques as they are buried inside the system. The design of these interface then hinges on flexible computational approaches that can predict the crystallography and the composition along the direction normal to the interface.
Our focus is specifically on the system (Cu,W) consisting of vertical coherent regions of the two metals separated by a lateral semi-coherent interface where a specific misfit dislocation pattern provides optimum chemical variability and mechanical relaxation. This interface design can be predicted and modelled using atomistic simulations, keeping the computational cost at a reasonable level by using semi-empirical potentials.

We study order transitions and defect formation in a model high-entropy alloy (CuNiCoFe) under ion irradiation by means of molecular dynamics simulations. Using a hybrid Monte-Carlo/molecular dynamics scheme, we generate a model alloy, which is thermodynamically stabilized by configurational entropy at elevated temperatures, but partly decomposes at lower temperatures via copper precipation. Both the high-entropy and the multiphase sample are then subjected to simulated ion irradiation. We analyze the damage accumulation in the studied allow and compared it to an elemental Ni reference system. The results reveal that the high-entropy alloys stabilize at a certain fraction of short-range order even under ion irradiation, independently of its initial configuration. Moreover, the results provide evidence that defect accumulation is reduced in the high-entropy alloy compare to pure Ni. We explain this reduction by the reduced mobility of point defects leading to a steady state of defect creation and annihilation in the course of irradiation. The lattice defects generated by irradiation are shown to act as sinks for Cu segregation.

Concentrated random solid-solution alloys (CSAs) have drawn wide interest as structural materials for next-generation nuclear reactors due to their exemplary mechanical and radiation tolerance properties over conventional dilute alloys. However, recent irradiation experiments show that some of these alloys can undergo disorder-to-order transition, i.e., the atoms that are initially randomly distributed on a crystal lattice undergo ordering (e.g. L1₀ or L1₂) due to irradiation. In nuclear reactors, such transitions can potentially impact the long-term microstructure evolution including grain growth and segregation properties, thereby damaging the chemical and mechanical integrity of materials. In this work, we elucidate the effect of atomic structure (i.e., ordered vs disordered) on the irradiation-induced microstructure evolution of CSAs. While working on Ni and NiFe alloys, from atomistic simulations of over 150 grain boundaries (GBs), we show that there is a direct correlation between Cr segregation and GB energy, i.e., segregation increases with increase in the GB energy. In addition, we show that Cr segregation is higher for disordered NiFe compared to ordered NiFe, albeit both have identical alloy composition. Using molecular dynamics simulation on grain growth, we find that higher grain growth is observed in disordered NiFe compared to ordered NiFe, supporting the previous postulation that disordered structures have higher GB energy compared to their ordered counterparts. Finally, the effect of impurity pinning of GB mobility between ordered and disordered alloys is shown.

Diffusion of point defects during irradiation is simulated via a two-way coupling between mechanical stress and defect diffusion in iron. This diffusion is based on a modified chemical potential that includes not only the local concentration of radiation-induced defects, but also the influence of the residual stress field from both the microstructure (i.e. dislocations, grain boundaries or triple junctions) and the eigenstrain caused by the defects themselves. Defect flux and concentration rates are derived from this chemical potential using Fick's first and second laws. Mean field rate theory is incorporated to model the annihilation of Frenkel pairs, and increased annihilation near grain boundaries is included based on the elastic energy of each grain boundary. Mechanical equilibrium is coupled with diffusion by computing eigenstrain from point defects and adding this to the total strain. Intrinsic stresses associated with the dislocations, grain boundaries and triple junctions are calculated using dislocation and disclination mechanics. Through this two-way-coupled model, defect concentration and sink efficiency is calculated for different types of microstructure. The results show that the two-way mechanical coupling has a strong influence on sink efficiency and provide guidance and metrics to quantify their characteristics.

High-temperature ultrafine precipitation strengthened steel (HT-UPS) is a candidate structural material for advanced nuclear reactors because of its high temperature creep resistance. However, little is currently known about its structure following irradiation. Neutron irradiation induces defect structures over multiple length and time scales. This research examines the three-dimensional (3D) microstructural response of HT-UPS to low neutron doses (0.01, 0.1, and 1 dpa) at ~600^oC. This work will focus on the mesoscale evolution of HT-UPS using high-energy X-rays. High-energy X-rays can provide 3D, non-destructive characterization to study both microstructural and micromechanical aspects. Both far field high-energy diffraction microscopy (FF-HEDM) and near field (NF)-HEDM techniques were used to characterize polycrystalline HT-UPS samples at the Advanced Photon Source. FF-HEDM provides information about center-of-mass, orientation, and strain states of individual grains. NF-HEDM provides inter and intra-granular crystallographic orientation information, akin to non-destructive electron backscatter diffraction, but in 3D. In the present work, FF- and NF-HEDM will be used to quantitatively study these features in irradiated HT-UPS specimens. These results will be compared to the identical, previously unirradiated HT-UPS specimens. This will be first of its kind 3D microstructural comparison study between neutron irradiated and unirradiated materials.

Small-angle neutron scattering is a powerful technique for measuring bulk averaged nanometer length scale structures in a variety of materials nondestructively. It provides complementary information to direct-geometry small-volume or surface techniques like electron microscopy and atom probe tomography. This technique is well-suited for use in studying the properties of alloys due to the widely varying contrasts of different transition metal isotopes when viewed with neutrons, the sensitivity of neutrons to magnetic structure, and the high penetrating power of neutrons in many materials regardless of atomic number. These properties, together with the high flux available at modern neutron scattering user facilities and the existing infrastructure for working with radioactive materials, make SANS a uniquely powerful technique for studying reactor structural alloys.

In this work, we describe how SANS is used to characterize the growth of irradiation induced precipitates in Fe-Cr-Al model alloys ranging in composition from 10-18 wt.% Cr and 3-5 wt.% Al that have been irradiated in the Oak Ridge National Laboratory High Flux Isotope Reactor at nominal damage doses up to 13.8 dpa as a function of dose and to probe the in-situ dissolution of the radiation damage precipitates in a single alloy as a function of temperature and time. We also describe the procedures used to perform measurements using on highly radioactive samples shielded to minimize personnel radiation dose and the risk of contamination.

A systematic study of post-quench ductility of sponge-based E110 samples subjected to steam oxidation at 1000°C, 1100°C, and 1200°C was performed by ring compression testing (RCT). For sponge-based E110 samples hydrogen pickup was not observed during steam oxidation.
Thus, it can be concluded that the ductility of oxidized samples, represented by the offset strain, gradually decreases as the oxygen pickup increases only. The samples oxidized at 1200°C are less ductile than those oxidized at lower temperatures at the same value of equivalent clad reacted. Microstructural examinations were performed on oxidized E110 specimens to correlate material performance with microstructure. Three different layers were observed: oxide layers on sample surfaces, a low oxygen layer in the middle of the cladding wall, and an oxygen rich layer between the oxide and low oxygen layers. Oxide layer thickness is observed to increase as the test time increases. Scanning electron microscopy study shows that the average oxygen content in the center 300 µm increases as the temperature increases for the samples having the same oxygen pickup. Microhardness measurements were performed at room temperature on the oxidized E110 samples. The hardness value of the middle 300 µm increases with both oxygen pickup and steam oxidation temperature. This indicates that more oxygen diffused into the middle of the prior beta layer at higher temperatures thereby reducing the ductility of the prior beta layer. The RCT data show that the strain values decrease as oxygen pickup increases and decreases as temperature increases. This observation is confirmed by the microhardness profile across the metal layer. Low hardness values at 1000°C indicate low oxygen concentration in the metal layer, in good agreement with the observation that the oxygen pickup at 1000°C is lower than that at 1100-1200°C. In addition to the as-received E110 cladding sample, pre-hydrided E110 cladding with hydrogen concentrations in the 100-150 wppm range was also investigated. The material behavior of pre-hydrided specimens is compared to the as-received E110.

The presence of hydrogen (H) can detrimentally affect the mechanical properties of many metals and alloys. To mitigate these detrimental effects requires fundamental understanding of the thermodynamics and kinetics governing H pickup and hydride formation. In this work, we focus on H diffusion in Zr-based alloys by studying the effects of alloying elements, vacancy and stress, factors that have been shown to strongly affect H pickup and hydride formation in nuclear fuel claddings. Parameterized by DFT calculations, a recently developed accelerated kinetic Monte Carlo method is used for the study. It is found that for the alloys considered here, H diffusivity depends weakly on composition, with negligible effect at high temperatures. In contrast, H diffusivity is affected by stress strongly. In addition, H diffusivity is impeded under relative high vacancy concentration.

In this work, we attempt to show that microstructural texture analysis coupled with optical microscopy and morphological analysis can be used to directly elicit and quantify the process-path history of a Zirconium-Niobium alloy. We do this by first determining the processing-path relationships belonging to a specific process by performing the experimental processing steps and measuring the texture evolution for a series of samples. We then use those relationships as input to a reverse process-path model that allows us to predict the initial microstructure and processing history of the sample. Lastly, we show that programs such as the Visco-Plastic Self Consistent (VPSC) model can be used to accurately simulate the texture evolution during processing, thereby reducing the need for experimentation and opening this method up to material systems in which the necessary processing-path relationships may not be readily available.

We do this by incrementally warm-rolling monotectoid Zr-18.8w%Nb, and measuring the texture evolution through both X-ray Diffraction (XRD) and electron back-scatter diffraction (EBSD) methods. As these methods provide slightly different views of the texture, we attempt to quantify this difference in relation to the reverse process-path model. We also perform several VPSC calculations on the BCC Zr-Nb system in which we attempt to match the experimental texture growth by varying the active slip systems and hardening laws. In general, we find that slight variations in the processing paths might significantly affect the final texture but this does not prevent us from acquiring an adequately accurate process-path from our model. For instance, in the case of warm-rolling, the process of repeatedly heating the sample before each roll introduces a recrystallization effect which appears in the formation of a strong γ-fiber in the orientation distribution functions (ODFs), particularly at the {111}<112> and {111}<110> components. However, the typical BCC rolling textures generally remain along the α- and θ-fibers and at the rotated cube components. Similar results were also found from VPSC simulations as the variations in processing made it difficult, but not impossible to reproduce the texture evolution. It was found that if the commonly active BCC slip systems are accounted for the texture can generally be fairly well predicted as long as the hardening law components are chosen carefully to best represent the process and material system.

Hydride precipitates are the major cause for degradation of the mechanical properties (e.g. ductility, fracture toughness) of zirconium (Zr) alloy based clads for nuclear fuel in Light Water Reactors. Experimental studies highlighted the importance of the morphology and orientation of hydrides on hydrogen embrittlement in Zr alloys. In addition to showing a preferred alignment along the circumferential direction, hydrides were commonly observed to self-organize into stacks. The two-level organization of hydrides is connected to the mechanical fields around each hydride and the complex interaction with the surrounding hydrides and defects (e.g. dislocations).

In the present study, a Fast Fourier Transforms based Discrete Dislocation Dynamics (DDD-FFT) technique is used to solve the mechanical fields in and around a δ-hydride precipitate in Zr matrix. The largely anisotropic stress fields developed by an ellipsoidal-shaped δ-hydride precipitate is modified due to the presence of dislocations in the Zr matrix. In addition, the shielding distance, which is introduced as a measure for hydride separation distance in a stack, decreased four-fold in the direction of the largest misfit strain of the δ-hydride. The growth zones, which are the regions for preferred growth direction of a hydride, and the nucleation zones, which are the regions with highest probability for nucleating another hydride, were analyzed by plotting the driving force around the δ-hydride. Nucleation zones at a 45^o angle with the long axis of the δ-hydride were identified as the reason behind the self-organization (stacks) of hydrides. Finally, we derive conclusions on the shielding effect of the δ-hydride due to dislocation nucleation near the hydride-matrix interface.

The phenomena describing the formation of brittle hydrides in nuclear fuel cladding induced by corrosion during reactor operation have been well-established in the literature for years. Further, the degradation of the mechanical properties has been a notorious consequence of these hydrides. As the first barrier against fuel release, it is paramount that nuclear fuel cladding maintains its structural integrity during reactor use and post-use storage. For this reason, developing a method to obtain material properties to evaluate the likelihood of failure has significant merits, especially a method that can test tubes in their customary state without significant modification.

An effective method to determine mechanical properties such as Young’s modulus, yield stress, strain-hardening, and fracture energy of as-received unirradiated and hydrided unirradiated Zircaloy-4 tubes in the hoop direction at room temperature was developed for ring compression tests through experimental and computational work. The hydrogen content was controlled up to 500wppm where failure is isolated at the 3 o’clock position. Since the crack location is isolated, cohesive elements were implemented to model fracture behavior. Length dependence of ring samples was determined experimentally where geometric features were tracked using projection digital image correlation (DIC) along with conventional load-displacement data. Computational work was performed in Abaqus using 2D and 3D models to complement this experimental work. Further, sensitivity analysis was performed on the methodology in this work to determine how strongly the input parameters affect simulation results. As expected, increasing hydrogen content caused embrittlement of the Zircaloy-4 samples tested. The integrated approach will be described and demonstrated to identify the most influential material properties on the damage tolerance of hydrided Zircaloy-4 tubes.

This work focuses on the development of a physics-based thermal creep model aiming to predict the behaviors of Zircaloy-4 under reactor accident condition. The proposed model accounts for the hardening contribution of solutes via their time-dependent pinning effect on dislocations. A recently proposed core-diffusion model is coupled with a transition state theory framework which accounts for the heterogeneous distribution of internal stresses within each grain. The diffusional creep mechanism is also considered. This model, embedded in the effective medium crystallographic Visco-Plastic Self-Consistent (VPSC) framework, is capable of capturing the experimental data for a comprehensive set of testing conditions covering the steady-state strain rate from 1X10^-9s^-1 to 2X10^-3s^-1. The transition between the low (n~4) and high (n~9) power-law creep regimes is reproduced through the dependence of the solute-dislocation binding energy on the dislocation waiting time. The prediction of the anomalous strain rate sensitivity (SRS) through the effect of core-diffusion to junction strength is also discussed. This model is implemented in finite element platform to simulate the burst behavior of Zircaloy-4 cladding tube under Loss-of-Coolant Accident (LOCA) condition.
The proposed model shows two major advantages compare to more empirical ones used as constitutive laws for describing thermal creep: 1) specific dependences on the nature of solutes and their concentrations are explicitly accounted for by the model; 2) this model captures the creep rate evolution in both primary and steady-state creep regimes. The accident conditions in reactors usually take place in short times, and deformation takes place in the primary, not the steady-state creep stage. Therefore, a model that accounts for the evolution of microstructure with time is more reliable for this kind of simulation.

There is an increased interest in integrating thermodynamic calculations in multi-physics codes to predict various aspects of nuclear fuel behaviour in a mechanistic framework. For oxide fuels, progress has been made in providing thermodynamic properties computed by Thermochimica such as chemical potentials, thermochemical activities, and speciation of various phases to the nuclear fuel performance code bison. Furthermore, the output provided by Thermochimica has been used to predict oxygen transport in bison in a more fundamentally correct manner than conventional approaches that do not account for the effects of irradiation. In addition to work in oxide fuels, work is on-going in investigating thermochemical behaviour of molten salt fuels. Here, a great difficulty is that the fuel is liquid and the fission products are not physically separated from the coolant; therefore, the transport of fission product containing phases depends strongly on the state of secondary phases. This presentation will cover a number of recent developments in algorithm and code development in Thermochimica and the coupling to multi-physics codes simulating the behaviour of oxide and molten salt fuels.

Improved thermal conductivity, oxidation resistance, high fissile density and high melting point are all very desirable properties for accident tolerant fuels (ATFs). Recent research has aimed to improve the thermal conductivity of ATFs via the addition of a second phase, with better thermal conductivity, in the existing uranium dioxide fuels. Fabrication of accident tolerant UO₂ based composite fuels is of interest since the primary nuclear fuel fabrication infrastructure is developed around UO₂. Examples of such UO₂ based composites include the addition of nitride, silicide and boride phases. In particular, composites of UO₂ with phases of uranium borides, namely UB₂ and UB₄, can be proven to increase thermal conductivity thus providing better safety margins in an accident scenario. Since ¹⁰B is a neutron absorber, the boron contents of the proposed ATF composites would have to be enriched to ¹¹B which has a lower neutron cross section. An important parameter when assessing the feasibility of UO₂-UB_x composites is the required phase fraction of UB_x to provide a significant increase in thermal conductivity, compared to UO₂. If large phase fractions are required, then the necessary enrichment process may make this fuel too costly to implement. Nonetheless, UB_x can still be included as a third phase in a different ATF composite system where the required volume fraction would be small enough that no enrichment would be necessary. In this case, the volume fraction can be tuned to act as an efficient burnable absorber. In this work, UB₂ and UB₄ fuel pellets have been sintered to high densities (>90% TD) via spark plasma sintering (SPS). A range of SPS conditions was tested to identify which resulted in the highest density pellets and the initial microstructural characterization along with measured thermal conductivity data are presented here.

Over the lifetime of the Transient Reactor Test Facility (TREAT) reactor from 1959 to 1994, a series of historical tests disclosed certain material specific requisites for irradiation testing variables resulted in negligible structural damage after exposure to the reactor and multiple experiments. Comparable fuel design, material testing, and qualifications needs to occur for the high-enriched uranium (HEU) to low enriched uranium (LEU) fuel conversion (<20% U235) of TREAT. This includes, retaining a uniform precipitate dispersion of fueled UO2 micron-size particles throughout a graphite-moderating matrix, where high-density graphite is in direct contact with the fuel and acts as a moderator. On that note, the design of future LEU reactor TREAT fuel core will consider the relative sizing and spacing of fuel as well as the retention of graphite in accordance with the expected thermal, neutron, and energy portfolio for future conversion. In light of the above, LEU conversion fuel blocks currently being manufactured must undergo critical materials testing, irradiation, and examination to determine effects on the manufactured replacement fuel blocks at the millimeter to sub-micron size scale.

Herein, we will report and update the community on value-added studies focusing on the reported particle size distributions of fabricated surrogate and fueled fuel blocks using non-destructive X-ray micron tomography (µCT) and electron-based microscopy. The developed computational image analysis methods to support this effort has included the development of techniques to distill 3-D particle-size distributions, nearest neighbor distances, volume fraction particle curves, and full reconstructions from µCT tomography considered as Big Data spanning thousands of images and several hundreds of terabytes. At this point, the current and pending results provide methodologies to qualify replacement fuel blocks as well as support decisions regarding the final fabrication specifications and tolerances associated with the replacement TREAT fuel assembly and final block configuration.

During service in a nuclear reactor, the chemical composition of nuclear fuel (UO₂) gradually changes with the chain of fission reactions. The high temperatures and continuous irradiation facilitate high diffusion rates and defect production. Oxide solid solutions, perovskites, metallic Mo-Tc-Ru-Rh-Pd-particles and fission gas bubbles have been observed after burnup. The characterization of radioactive materials is time-consuming and expensive. Therefore, in this study similar degradation processes were characterized in a doped simulated fuel (CeO₂). Pulsed laser deposition was used to deposit 1 µm thick polycrystalline CeO₂ films from a CeO₂ target at 550°C. The target was also containing 2 wt% Mo, 1.5 wt% Ru, 0.75 wt% Pd, 0.5 wt% Re and 0.25 wt% Rh. Polycrystalline yttrium stabilized zirconia was used as a substrate. These doped CeO₂ films were irradiated with I⁺ ions at 610°C and 710°C at a flux of 10¹⁶ and 5×10¹⁶ I⁺/cm² each. Sample preparation for transmission electron microscopy (TEM) and atom-probe tomography (APT) was carried out with the focused ion beam. TEM Energy-dispersive X-ray Spectroscopy revealed the formation of small (5-10 nm) Pd precipitates. In many cases bubbles were observed associated with these precipitates. APT was used to quantify the composition of the metallic nanoparticles. Preliminary data from He⁺ implanted material showed Pd precipitation preferentially in the vicinity of grain boundaries. Impurity particles containing SiO₂H_x observed in these specimens highlight the disastrous effect small contaminations could have on fuel degradation. The absence of other precipitates suggests that Pd is the most mobile atom among the five dopants under the irradiation conditions and the precipitation is mainly driven by irradiation-enhanced diffusion processes. Similar precipitation processes are expected in UO₂ fuels in dependence of dose and temperature.

Uranium nitride-silicide composites are being considered as a high-density, high thermal conductivity fuel option for Light Water Reactors (LWRs). During development chemical interactions were observed near the silicide melting point which resulted in formation of an unknown U-Si-N ternary phase. In the present work, U-Si-N samples were produced by arc-melting under a N₂ atmosphere. The resulting samples were characterized by SEM/EDS and XRD, and demonstrated an equilibrium between U₃Si₂, UN, USi and a U-Si-N ternary phase with a distinct crystallographic structure. DFT+U calculations were performed to evaluate the thermodynamic stability of various ternary structures from the analogous system U-Si-C. The combined results provided a good fit to a proposed structure for U₂₀Si₁₆N₃ obtained using U=1.1 eV, where the phase was also found to be energetically stable. The results support the thermal stability of this ternary phase and should lead to further exploration of the U-Si-N system.

The advanced ceramic fuel development program is exploring revolutionary fuels with the potential of “game-changing” impact on reactor operation & response to beyond design scenario. Key properties of advanced fuels include high thermal conductivity, oxidation resistance, high temperature mechanical properties, and thus improved accident tolerance. Composite ceramic fuels possess distinct advantages to fulfill these key requirements. In addition, the US Nuclear Energy Advanced Modeling and Simulation (NEAMS) program is developing science-based next generation fuel performance modeling capability to facilitate the predictive capability of nuclear fuel performance and critical experimental data are needed to validate the multiscale multiphysics MARMOT models. In this talk, recent advancements of using field-assisted sintering technologies, specifically spark plasma sintering (SPS), in fabricating advanced fuels and engineering fuel matrix as the target systems will be reviewed. Different types of concepts are explored for the advanced fuel designs including graphene-based UO₂ composite fuels, large-grained fuel doped by oxide additive and the high uranium density fuel, and the impact on design of accident tolerant fuels is discussed. Recent progresses of using SPS in tailoring and engineering fuel matrix as the target systems for validating MARMOT physics models will also be highlighted. Particularly, monolithic oxide fuels with tailored microstructure including grain size across multiple length scales from nano-metered to micron-sizes, porosity and stoichiometry can be sintered. The impacts of tailored microstructure on thermal-mechanical properties and grain growth kinetics are discussed within the context of the MARMOT modeling.

Uranium-silicide (U-Si) fuels are being pursued as a possible accident tolerant fuel (ATF). This uranium alloy fuel benefits from higher thermal conductivity and higher fissile density compared to UO₂. In order to perform engineering scale nuclear fuel performance simulations, the material properties of the fuel must be known. Currently, the experimental data available for U-Si fuels is rather limited. Thus, multiscale modeling efforts are underway to address this gap in knowledge. Mesoscale modeling methodologies, such as phase-field, require information on the fundamental properties of the material system of interest. This information includes, but is not limited to, point defect energies, diffusivities and interfacial energies. In this study, an interatomic potential for U-Si is implemented for the investigation of interfaces in U₃Si₂. A variety of planar free surfaces, voids, and twist and tilt grain boundaries are investigated. This information is implemented in a phase-field model to predict the swelling behavior of U₃Si₂.

The U₃Si₂ compound is being developed as an advanced accident tolerant nuclear fuel, which benefits from high thermal conductivity compared to the UO₂ fuel currently used in most light water reactors. U₃Si₂ also has higher fissile density than UO₂, giving economic benefits, while an elevated oxidation rate upon cladding breach due to an accident scenario or wear may be a drawback. In order to model the fuel performance of U₃Si₂, material properties such as the diffusion rate of point defects and fission gas atoms must be determined. In this study, the thermodynamic and kinetic properties of U and Si point defects (vacancies and interstitials) as well as Xe atoms interacting with point defects are investigated by means of density functional theory (DFT) calculations. The nudged elastic band (NEB) method is employed to calculate migration barriers. The Hubbard U methodology (DFT+U) is used to describe the properties of U 5f electrons, which has been shown to be necessary in order to recover the experimentally observed U₃Si₂ crystal structure (P4/mbm) as the ground state. The same processes have been investigated using a U-Si empirical potential, which also gives us access to defect entropies and attempt frequencies for migration. A few entropies will be validated against DFT calculations. The DFT and empirical potential results are combined to predict diffusivities. Both U and Si self-diffusion and Xe diffusion are anisotropic as a consequence of the tetragonal crystal structure of U₃Si₂. Self-diffusion of U and Si in U₃Si₂ is faster than U self-diffusion in UO₂. Interstitial diffusion of U is very fast in U₃Si₂ and it is on par with O diffusion in UO₂. Xe diffusion is also faster in U₃Si₂ than in UO₂. Implication of the predicted diffusivities on fuel behavior will be discussed.

Following the 2011 nuclear accident in Fukushima, the United States Department of Energy began funding research to improve efficiency and robustness of nuclear fuel. Among the candidates are uranium-silicides, most notably U₃Si₂. Uranium-silicides have good thermal conductivity and higher uranium density when compared to UO₂. Although the uranium-silicide system has been studied in the past, recent experimental findings indicate gaps in understanding the 60-66 at.% silicon region. In this work we fabricated samples in this compositional range using arc melting and characterized them using X-ray Diffraction and Scanning Electron Microscope-Energy Dispersive Spectroscopy to elucidate phase relations. In addition, phase transitions were determined using Differential Scanning Calorimetry.

This research is being performed using funding received from the DOE Office of Nuclear Energy's Nuclear Energy University Programs.

Uranium-plutonium mixed oxide (MOX) fuel with about 25% of Pu content is the reference nuclear fuel for the future Gen IV fast neutron reactor in France. It will be fabricated from spent fuel coming out of the current pressurized water reactors and will, as a consequence, contain a few percent of Am (less of 5%). The high radiotoxicity of Pu makes the MOX fuel difficult to handle experimentally and, consequently, experimental data on this compound are scarce. The main goal of this study is to determine the effect of americium on the properties of the mixed actinide oxide fuel (U, Pu)O₂ containing americium in low concentration, by comparison of the results obtained for (U, Pu)O₂ to those obtained for (U, Pu, Am)O₂.
Electronic structure calculations based on the DFT+U approach are used. The occupation matrix control (OMC) scheme is also used to avoid convergence of the calculations to metastable states. As a first step, we calculate bulk properties and thermodynamic properties (thermal expansion, variation of enthalpy …) of AmO₂ as the function of U and J parameters of DFT+U. In order to get insight into the influence of americium on irradiation damage in oxide fuels, the formation and migration energies of points defects (vacancies, interstitial, Schottky defects) calculated in (U, Pu)O₂ for various Pu contents are compared to those obtained in (U, Am)O₂.

Nuclear fuels are severely impacted by radiation damage occurring during irradiation in reactor mainly by fission and during storage/disposal by alpha-decay.
One of the most spectacular transformations occurs in the high burnup light water reactor fuels UO2 or MOX and is referred to as the high burnup structure (HBS). To form this structure the original grains with sizes of several micrometers subdivide into thousands of smaller grains of about 200 nanometers, associated with the formation of a high porosity. The underlying mechanisms of this restructuring are associated with the accumulation of defects and their reorganisation to minimize the stresses in the fuel lattice.
It was recently observed by transmission electron microscopy (TEM) that several minor actinide-containing materials also exhibit a structure of nanosized grains formed after the accumulation of high alpha-decay doses.
Experimental observations by TEM of very high burnup UO2 and MOX fuel will be reported. The impact of restructuring on some properties such as thermal conductivity, mechanical properties and fission gas release will be presented. Evidences of polygonisation will be given for minor actinide bearing fuels. Several materials have been studied including pure americium dioxide, UO₂ fuels containing americium or the short-lived ²³⁸Pu.
Correlations will be made with observations of polygonization in ion-implanted materials or with the dose dependence of the defect pattern evolution of alpha-damaged materials.

To date, experimental studies of radiation damage are limited to post-mortem analysis of irradiated samples. Because of this, processes that occur on sub-nanosecond timescales have to be inferred from observations at time-infinity. Ultrafast lasers, which electronically excite materials, can be used to replicate electronic stopping damage in nuclear materials and perform time-resolved measurements. Here, we couple ultrafast laser irradiation with high pressure generated by a diamond anvil cell to study the behavior of compressed, electronically excited UO₂. Excitation was found to produce a photoinduced polaron, as evidenced by a low frequency phonon mode. Frequency and lifetime of the phonon mode were measured as a function of pressure. A change in polaron behavior was observed at ~10 GPa—consistent with a previously observed electronic transition in UO₂—showing a coupling between electronic structure and the polaron. Measurements were made at multiple probe wavelengths to help understand the dispersion relation of the phonon mode, and thus the nature of the polaron. This polaron has been a proposed energy carrier in radiation damage of UO₂, though until now little of its behavior has been understood.

The development of comprehensive, physics-based nuclear fuel performance codes requires integration of the chemical state of the fuel to allow accurate representation of a number of thermophysical properties as well as oxygen and reactive species transport and behavior. A thermodynamic equilibrium solver, Thermochimica, has been successfully coupled with the nuclear fuel performance code BISON to provide that capability. Representing the local fuel chemical state requires an accurate set of thermochemical models and parameters for fuel undergoing burnup, which for LWRs is UO₂ together with fission products and relevant transuranics. Such a set of models and parameters has been under development using elemental subsystems and is being integrated into a comprehensive representation of the fuel thermo-chemical state during burnup. In this paper we will discuss the state of the art of modeling the chemistry of nuclear fuel undergoing burnup and its application in modeling of in-reactor phenomena.

This research was supported by the U.S. Department of Energy, Office of Nuclear Energy, Nuclear Energy Advance Modeling and Simulation Program.

Controlling and predicting the high temperature mechanical behaviour of nuclear oxide fuels is essential to guaranteeing the integrity of fuel rods during normal or incidental operating conditions. Although in the past many studies have been devoted to the creep behaviour of uranium dioxide, few have focussed in detail upon the effect of non stoichiometry. In those which have dealt with this issue, strain rate was usually considered as a function of deviation from stoichiometry although the true intensive thermodynamic variables are the chemical potential, temperature and stress.
In this work we describe very recent developments involving a high temperature compression creep furnace which has been equipped with a system enabling the control and measurement of the oxygen activity in the gas phase. This guarantees the oxygen activity in the solid in equilibrium with it. We report the first creep experiments carried out under these controlled conditions. We also discuss the development of a material model and associated kinematic hardening behaviour law capable of reproducing the near constant strain rate experiments carried out. Although we show that the data may be interpreted in terms of uniaxial loading, due consideration is given to multi-dimensional effects. The physical significance of the material behaviour law parameters is presented, particularly when the parameter is sensitive to oxygen activity. Our ultimate aim is to relate internal variables of the model to the local microstructure. Post test Electron BackScatter Diffraction is shown to be a prospective technique for providing this information and based on it, parallels can be drawn between the mechanisms via which the material accommodates mechanically induced strain and its response to radiation damage.

Uranium dioxide and hyper-stoichiometric oxide have been extensively studied over many years, but still a great deal remains to be understood in relation crystal imperfections and their formation and migration mechanisms. This is because UO_2+x is a complex material in which anion, cation and electronic disorder changes with temperature, oxygen activity and impurity content. There are a number of complicating factors such as clustering and screening of charged defects that make this material difficult to study experimentally or using theoretical methods. The study of uranium vacancies is particularly challenging as these defects are generally present at low concentrations and in different charge states.
Recently, it has been shown by combining first principles approaches and experimentation carried out on charged particle irradiated UO₂, that Positron Annihilation Spectroscopy (PAS) was indeed sensitive to the presence of uranium vacancy containing defects. This, in addition to the fact that the method is a priori sensitive to charge, makes PAS a candidate as a spectroscopic technique that could further shed light on the presence and nature of thermally induced uranium vacancies.
Now, it is reasonably well established that at a given temperature, the uranium vacancy concentration is an increasing function of non-stoichiometry. It is further surmised that the charge of this type of defect, while being negative when the material is close to stoichiometric composition, will increase with increasing deviation from stoichiometry. In order to verify these points, we recently determined the positron annihilation characteristics of dense sintered samples that had first been subjected to controlled oxidation annealing. Doppler broadening and lifetime spectroscopy are shown to be extremely complementary techniques. Both are sensitive to the presence of negative or neutral vacancies and the former is more sensitive to the chemical environment of the positron when it annihilates whereas the latter allows distinguishing trapping rates at different vacancy defects. The results obtained are discussed at length in the light of previous studies and but also with regard to their implications upon material property changes. The method we have used is shown to be extremely promising for understanding and possibly quantifying the complex changes that occur during oxidation of UO₂ which involve the concomitant increase in oxygen interstitials and uranium vacancies.

Additives are widely used to control the microstructure of materials via their effect on defect chemistry during sintering. As the primary nuclear fuel, the properties of UO₂ are crucial for safe and efficient reactor operation. UO₂ has been manipulated by fuel vendors through doping to enhance grain size to provide improved fission gas retention and plasticity. In this work a common phenomenon that governs the effect of Mg, Ti, V, Cr, Mn, and Fe doping of UO₂ for enhanced grain growth is identified, elucidating experimental observations. A combined density functional theory and empirical potential description of defect free energy is used to calculate the doped UO₂ defect concentrations as a function of temperature. At high (sintering) temperatures all dopants studied transition to a 2+ charged interstitial defect. Furthermore, a number of dopants (Ti, V, Cr, and Mn) do so in sufficiently high concentrations to greatly increase the negatively charged uranium vacancy concentration. High uranium vacancy concentrations can enhance grain growth and fission gas diffusion. Mg and Fe also enhance uranium vacancy concentrations slightly, while Al has no impact. The enhanced uranium vacancy concentrations, associated with solution of dopants interstitially, is proposed as the responsible mechanism for enlarged grains in (Ti/V/Cr/Mg)-doped systems as seen experimentally. While Mn-doped UO₂ has been predicted to have higher uranium vacancy concentrations than the more widely used Cr-doped UO₂, Ti- and V-doped UO₂ are predicted to exhibit the highest grain growth and fission gas diffusivity

In recent years, U₃Si₂ is being considered to replace UO₂ as an advanced technology fuel (ATF) candidate, with the Fe-Cr-Al-Y alloys being used as an advanced cladding material. To date there is no assessed CALPHAD (CALculation of PHAse Diagrams) model of the U-Fe-Si ternary space in the literature. While limited experimental data exists, more realistic physical relations for the U-Fe-Si ternary space can be established by complementing existing information with density functional theory (DFT) calculated thermodynamic data, eventually resulting in a self-consistent CALPHAD model, which is the purpose of our study.

First, we expanded the current DFT database by determining the most stable structure of the experimentally observed ternary phases. For the experimentally observed phases with partial occupancies (occupancy less than 1) at a specific Wyckoff position, we used special quasi-random structures to determine the most stable configuration at a specific composition. Subsequently, using DFT, we calculated the formation energies and vibrational entropies of the constituent binary and ternary U-Fe-Si phases to estimate their Gibbs energies. This combined DFT-CALPHAD approach will aid in understanding complex phase formation across the compositional and temperature ranges for prospective ATFs to allow assessment of potential interactions between fuel and cladding systems.

This research is being performed using funding received from the DOE Office of Nuclear Energy's Nuclear Energy University Programs.

With the development of new reactor technologies, such as GenIV or Fusion, there is a concomitant development of new materials tasked with achieving the requried properties, e.g. temperature and radiation resistance. Coupled with this is the continued development of enhanced waste forms, such as glass-ceramic hybrids, which have been designed to accomodate the long terma changes induced by radioactive decay, e.g. recoil damage and transmutation. This presentation will outline the key factors impacting future materials development in these areas, and discuss key impaxcts arising within a new set of carbide based materials, propsoed for use within a reactor core, and the next generation of glass-ceramic hybrids for use as waste host matrices.

Pyrochlores are a class of materials with the general formula A₂B₂O₇, and they have attracted considerable interest both as nuclear waste forms as well as fast ion conductors. The relevant material properties for these applications, namely radiation hardness and ionic conductivity, have been related to disorder on the cation sublattice. Cation disorder may be introduced in a variety of ways: mixing of A and B atoms by heating or ion irradiation, or by the presence of grain boundaries.

In this work we describe the effects of ion irradiation on the cation sublattice of Gd₂Ti₂O₇ (GTO). Single crystals of GTO were grown by the floating zone method, oriented at the <110> pole by Laue diffraction, then sectioned and polished for irradiation. Light ion irradiation (He, 200 keV) was used to gradually introduce disorder into the material as a function of dose. TEM specimens were prepared from the irradiated crystals in the plan-view orientation using traditional dimpling and ion-milling techniques, giving a {110}-type foil plane. The irradiated samples were then examined using high-resolution TEM in a monochromated, image-corrected FEI Titan 80-300 operated at 300 kV. Additionally, the recrystallization and recovery of damage in the irradiated regions of the GTO were examined in these specimens by heating them in-situ using a Gatan double-tilt heating holder. From these experiments, a quantitative understanding of changes in the local cation environment with induced disorder is expected to better inform models of how cation structure imparts functionality to this class of materials.

Radiation damage in materials with fluorite parent structure have always been of interest to the nuclear materials community. This is because they can be used as model materials to study complex material modifications occurring in nuclear fuel or nuclear waste-forms. In the current work, we investigate and compare material modifications in three phases with structurs derived from the parent fluorite lattice existing in the Sc₂O₃:HO₂ phase diagram. These phases are designated as d-(Sc₄Hf₃O₁₂), g-(Sc₂Hf₅O₁₃) and b-(Sc₂Hf₇O₁₇) and are known to have with different degree of rhombohedral distortion. We will discuss the results of grazing incidence X-ray diffraction and transmission electron microscopy analysis performed on samples irradiated ion irradiated studies performed at liquid nitrogen temperatures. Preliminary results at higher fluences, show the usual order-disorder transformation due to formation of anti-sites which is commonly observed in the compounds with a parent fluorite structure. However, in the current, case this transformation is preceded by varying changes in the rhombohedral distortion. These results will be presented by comparing them with order-disorder transformations observed in other fluorite derivative compounds.

Alloys based on the Ni-Cr binary system, such as alloys 625 and 690, are susceptible to mechanical property changes with thermal aging. A disorder-order phase transformation, which is the primary source of embrittlement, has been studied extensively in Ni-Cr model alloys to predict the transformation rate as a function of time, temperature and stoichiometry. Model and commercial alloys have been isothermally aged up to 10,000 hours and characterized via nanoindentation, atom probe tomography, and transmission electron microscopy. Additionally, these alloys have been ion beam irradiated to investigate the role of irradiation in accelerating the ordering kinetics. Preliminary results indicate change in stoichiometry do not change the ordering rate only the amount of ordered phase formed. Also, proton irradiation tends to accelerate the ordering process while Ni+ ion irradiation do not lead to ordering at the dose rates explored.

We examine the thermal unpinning mechanisms of edge dislocations from voids in α-Fe is investigated by means of molecular dynamics simulations [1]. The activation energy as a function of shear stress and temperature is systematically determined. Simulations with a constant applied stress are compared with dynamic simulations with a constant strain rate. We found that a constant applied stress results in a temperature-dependent activation energy. The temperature dependence is attributed to the elastic softening of iron. If the stress is normalized with the softening of the specific shear modulus, the activation energy is shown to be temperature-independent. From the dynamic simulations, the activation energy as a function of critical shear stress was determined using previously developed methods. The results from the dynamic simulations are in good agreement with the constant stress simulations, after the normalization. This indicates that the computationally more efficient dynamic method can be used to obtain the activation energy as a function of stress and temperature. The obtained relation between stress, temperature, and activation energy can be used to introduce a stochastic unpinning event in larger-scale simulation methods, such as discrete dislocation dynamics.

[1] J. Byggmästar, F. Granberg, and K. Nordlund, Phys. Rev. Mater. 1, 053603 (2017)

The buildup of radiation damage in nuclear materials often depends on the spatial distribution of atomic displacements within collision cascades. Although collision cascades have previously been described as fractals, the correlation of their fractal parameters with experimental observations of radiation damage buildup remains elusive. Here, we use a recently developed pulsed-ion-beam method to study defect interaction dynamics in SiC irradiated at 100 C with 500 keV ions of different masses. Experimental data is analyzed with a model of radiation damage formation which accounts for the fractal nature of collision cascades. Our emphasis is on the extraction of the effective defect diffusion length from pulsed beam measurements. Results show that collision cascades are mass fractals with fractal dimensions in the range of ~1-2, depending on ion mass, energy, and the depth from the sample surface. Within our fractal model, the effective defect diffusion length is ~10 nm, and it decreases with increasing cascade density. These results demonstrate a general method by which the fractal nature of collision cascades can be used to explain experimental observations and predict material's response to radiation. This work was performed under the auspices of the U.S. DOE by LLNL under Contract DE-AC52-07NA27344.

The competition between interactions acting over different length scales, keystone of the radiation induced patterning, is clearly analyzed with conserved version of a Swift-Hohenberg (SH) equation. This equation analytically determined clearly shows that produced patterns exhibit universal features independent of the microscopic details of the dynamics. Such an equation allows not only the prediction of all possibles patterns and its generalized phase diagram, but also the solubility limits in emerging patterns. This analysis points out that up to now unexpected morphologies like rolls or hexagonal distributions of spheres with well defined solubility limits can be produced under irradiation varying the average composition of the alloys in the patterning domain.

The recent nuclear evaluations describe more accurately the elastic and inelastic neutron-atoms interactions
and allow calculating more realistically primary damage induced by nuclear reactions. Even if
these calculations do not take into account relaxation processes occurring at the end of the displacement
cascade (calculations are performed within the Binary Collision Approximation), they can accurately
describe primary and recoil spectra in different reactors opening the door for simulating aging of nuclear
materials with Ion Beam facilities. Since neutrons are only sensitive to isotopes, these spectra must be
calculated weighting isotope spectra by the isotopic composition of materials under investigation. To
highlight such a point, primary damage are calculated in pure Ni exhibiting a meta-stable isotope produced
under neutron flux by inelastic neutron-isotope processes. These calculations clearly point out
that the instantaneous primary damage production, the displacement per atom rate (dpa/s), responsible
for the micro-structure evolution, strongly depends on the 59Ni isotopic fractions closely related to the
inelastic neutron isotope processes. Since the isotopic composition of the meta-stable isotope vanishes
for large fluences, the long term impact of this isotope does not largely modify drastically the total dpa
number in Ni based steels materials irradiate in nuclear plants.

The development of corrosion resistant alloys for nuclear applications has been highly empirical in nature. Although in-reactor materials’ performances have constantly been improved, fundamental understanding of corrosion processes are clearly lacking. For instance, the specific effects of very small alloying elements addition in zirconium fuel cladding, either in solid solution or in precipitates, on oxidation kinetics and hydrogen pickup are still not understood. In addition, physically based corrosion models are critically needed to (i) predict materials’ behavior during normal and accidental conditions (ii) understand the role of alloying elements and (iii) characterize the complex coupling effects between corrosion and irradiation.
A holistic approach including precise oxidation kinetics and non-destructive hydrogen pickup measurements, coupled with microbeam synchrotron X-Ray Absorption Near-Edge Spectroscopy (XANES) experiments and in-situ Electrochemical Impedance Spectroscopy (EIS) data, has shed light on the alloying elements effects on corrosion and hydrogen pickup in zirconium fuel cladding. In parallel, a novel physically based corrosion model called Coupled Current Charge Compensation (C4) has been developed, which models the coupled oxygen vacancies, electron and proton currents through the oxide. The C4 model can calculate the oxidation kinetics, hydrogen pickup and space charge across the oxide to predict oxidation and hydrogen pickup kinetics of fuel cladding in normal and accidental (Loss Of Coolant Accident) conditions. The recent C4 implementation and validation into the nuclear fuel performance code BISON will also be presented.
Recently, specific research programs at the University of Wisconsin-Madison have aimed at validating the C4 model further and characterizing the neutron and photon irradiations effects on corrosion. The neutron irradiation effect on corrosion kinetics has been investigated in the framework of the C4 model on ZrNb model alloys using proton irradiation. The proton irradiation induced Nb redistribution has been characterized and its consequence on fuel cladding corrosion mechanism will be discussed. Finally, in-situ photo-corrosion experiments are under development to characterize the effects of photons (UV to hard X-Rays) on corrosion kinetics and hydrogen pickup. The general findings will be summarized and the applicability of this holistic approach to other alloy systems in extreme environments relevant to nuclear energy (molten salts, sCO2 and liquid sodium) will be discussed.

Hydrogen can cause embrittlement which affects safety and performance of Zirconium cladding materials used in nuclear applications. Hydrogen uptake and precipitation in zirconium alloys used in reactors have been potential causes of deterioration of cladding mechanical properties. Liner fuel claddings were developed to protect against Pellet-Cladding interaction (PCI) used as inner liner in BWRs or to improve corrosion resistance in PWRs as an outer liner.
Non-irradiated cladding sections with, without liner, hydrogenated, and with different cooling rates and under stress gradients were analysed. A strong relation between cooling, hydrogen concentration and its diffusion in cladding with the presence of a liner is observed. This paper focuses on the effect of the presence of the liner (inner and outer) and the comparison and quantification of hydrides present and their impact on the mechanical properties of the cladding materials. The result indicates that using very slow cooling rates hydrogen can diffuse to and precipitate in the liner. As in most post-irradiation examinations where the fuel rod typically also undergoes slow cooling, one observes a high number of hydrides in the liner, the hydrides density in the nearby cladding area are depleted. Whether the higher concentration in the outer liner can lead to a higher susceptibility for delayed hydride cracking is unclear.
The investigation of the effect of different liner material on the hydrides under different cooling rates, and with different hydrogen content is important to better assess the mechanical behaviour of the cladding. Understand the hydrogen diffusion, precipitation and re-orientation under certain cooling rates and stress was study by employing the newest method of neutron radiography at PSI. Experimental techniques like OM, SEM and hot gas extraction were also used to compare experimental results as well as validation methods.
Quantitative hydrogen concentration results measured by high-resolution neutron imaging, allowing analysis on stress-induced hydrogen concentration fields in a sub-10 um scale for different stress field and cooling rates will be presented with particular focus on the liner-bulk interface.

In light water reactors (LWR), neutron irradiation is responsible for the embrittlement of the ferritic steels constitutive of the reactor pressure vessel (RPV) stemming from the formation of solute-enriched radiation induced clusters. A deep understanding of the evolution of such cluster population in terms of size and composition is of critical importance in order to assess the macroscopic mechanical properties of RPV steels. To this end, a multi-scale approach has been developed at EDF R&D. Starting from atomistic calculations (usually Density Functional Theory, DFT) unveiling the fundamental interactions between point defects and solute atoms, modeling of the microstructure over time length and dose rate typical of LWR operating conditions is performed via Kinetic Monte Carlo (KMC) or Cluster Dynamics (CD).

In this paper, we will report the main advancements on DFT calculation, kinetic Monte Carlo and Cluster Dynamics for applications to RPV steels carried out in our team at EDF R&D. First, a methodology for thermodynamical calculation allowing for error mitigation in the case of large simulation box calculation has been extended to the case of DFT: it permits to calculate formation entropy of defects inducing large strain fields such as dislocation loops and solute clusters and eventually to foresee the inclusion of temperature effects in KMC calculation. Then, an improved parameterization of the solute-point-defect interaction for complex alloys within KMC has been developed accounting for concentration effects while hybrid Object KMC-Atomistic KMC simulation tool improves the performance of KMC simulation to the point where typical LWR fluences for 40 year operation are accessible. Finally, an explicit model of cluster dynamics for the prototypal dilute Fe-Cu material will be presented, highlighting current challenges in reproducing the dynamics of Cu precipitation with CD and the importance of correct parameterization of the physical model.

Nuclear grade graphite has been widely used as a neutron moderator, reflector and fuel matrix in various types of nuclear reactors since the late 1940s. Its characteristics made it a material particularly suitable for the nuclear application. Consequently, graphite represents the greatest volume of radioactive waste at the end of the reactor's life. To date, about 250,000 tons have been accumulated worldwide. This is typically the case of the French UNGG or the British MAGNOX nuclear reactors developed independently in the same period. The long-term storage or disposal of the nuclear graphite waste requires a special management strategy and the challenges for the fundamental management options are reflecting the chemical, physical and structural properties of the material itself, its retrieval from the core and the associated inventory of long lived radio-isotopes such as chlorine (³⁶Cl) or carbon (¹⁴C) that result from neutron activation processes. Therefore, prior to select any management option for the neutron-irradiated graphite, a comprehensive understanding of the structural properties of raw and structurally modified graphite is needed, so as to provide efficient and effective solutions. Raman spectroscopy appears to be an appropriate technique to probe the structural modifications of nuclear graphite. The present paper will discuss the Raman response through the various types of defects that may appear in the nuclear graphite during its manufacturing and/or surface preparation, after being exposed to neutron bombardments and. upon ion-beam irradiation. For this latter, a special device was developed allowing in situ monitoring of nuclear graphite behavior under light ion beam supplied by a cyclotron, with conditions of temperature, pressure and chemical atmosphere similar to those of an operating UNGG reactor.

An atomistic study is performed on various interfacial defect structures of U-Zr alloy misfit boundaries. A promising next-gen nuclear fuel element, U-Zr alloy commonly operates in harsh thermodynamic conditions that cause phase transformations/separations. These heterogeneities create phase boundaries that function as defect concentration sites and could result in performance degradations. For the high operation temperature, both g-U and Zr exist in body-center cubic(b.c.c.) structures with slight deviations in lattice size. As such, various cube-on-cube stacking misfit interfaces are simulated to study the phase boundary structures. Lattice misfit accommodations in the form of dislocation networks are observed for {100}, {110}, {111} and {112} interface orientations. For {110} and {111} interfaces, structural deviations from classical O-lattice theory predictions are also observed due to dislocation-dislocation interactions. Disregistry for each dislocation set is calculated to compare the importance of the respective set. Strain fields are plotted to identify dislocation emission pathways away from the interface.

Crystalline secondary phases are often present in vitrified nuclear waste.
In this work interfaces are simulated between (Na₂O)_x(SiO₂)_1-x glasses (for x=0.0, 0.1 and 0.2) and TiO₂ crystals using molecular dynamics and empirical potentials.
These calculations were used to investigate features and properties of such interfaces.
They showed that partially ordered layers had been induced in the glass close to the interfaces, with successive O-rich, Si-rich and Na-rich planes being noted.
The first silicate layer in contact with the crystal tended to be highly-structured, with Si ions occupying well-defined positions that depend on interface orientation, and showing 2-dimensional ordering depending on glass composition.
Results are presented suggesting that the structural flexibility of the glass network allows it to conform to the crystal, thereby providing charge compensation and avoiding large relaxation of the crystal structure close to the interfaces.
Such interfacial properties could be crucial to improving phenomenological models of glass-crystal composite wasteform properties.

The goal of this project is to understand the evolution and manifestation of microstructural features in metallic alloy nuclear fuels through a separate effects or parametric test. These are often used to uncover the underlying influence a particular phenomenon or property has on a more complex system into which it is integrated. Such is the situation with the behavior of nuclear fuel under irradiation with respect to phenomena such as grain refinement, fission gas bubble behavior, and element redistribution, all of which are a function of fission rate, fission density, temperature, and material composition. DISECT consists of four uranium-zirconium (U-Zr) low enriched uranium (LEU) fuel assemblies with samples ranging from 6 to 30 weight percent Zr. The U-Zr samples are being fabricated at Idaho National Laboratory (INL) by arc-casting short rods, sectioning the casting into buttons, hot and cold rolling these to thickness, and annealing. These will then be shaped and sealed in a mated zirconium shell that will provide both the vessel for sample retention and the thermal barrier necessary to control fuel temperature. Each assembly is equipped with in-pile instrumentation to monitor sample temperature during irradiation in Belgian Reactor 2 (BR2) operated by Studiecentrum voor Kernenergie (SCK). This novel approach to the experimental design is necessary to target specific temperatures (300-800 C), fission densities (2-6% atomic BU), and fission rates (200-750 W/cc) throughout irradiation without convoluting additional influences such as phase transitions and constituent redistribution. The effects of these well controlled experimental parameters will be used to understand the radiation-induced microstructural changes. As such, the pre-irradiation characterization of the microstructure is a crucial component to this experiment. These activities, discussed herein, include chemical analysis, sample mounting and preparation, optical microscopy, scanning electron microscopy (SEM), wavelength-dispersive X-ray spectroscopy, and transmission electron microscopy.

Polyimides have been widely used in many applications because of their superior properties, such as excellent thermal and chemical stability, excellent dielectric properties, and mechanical strength. Their applications range from microelectronics and aerospace to the medical and nuclear industries. However, even though polyimides show superior durability in harsh conditions compared with other polymers such as polyolefin and polyesters, the polyimides, being organic materials, will decompose under high doses of ionizing radiation. The resulting degradation of physical properties limits the reliable operation lifetime of systems composed of unprotected polyimide materials in this environment.

Recently, NASA has developed new inorganic/polyimide hybrid composites for long duration aerospace structural applications requiring a radiation resistant capability against high energy electrons, protons, neutrons, heavy ions, vacuum ultraviolet (VUV), gamma-rays, and X-rays from solar particle events (SPE) and galactic cosmic rays (GCR). In this study, the effect of inorganic particles on the radiation induced degradation of polyimides, and the mechanism of degradation under high energy radiation, was initiated. LaRC SI^TM polyimide was chosen as a baseline matrix, and several different inorganic nanoparticles were selected as additives. To evaluate the effect of ionizing radiation, the hybrid polyimide composites were exposed to energetic electrons (about 50-60 keV and 9 nA/cm²) or high energy photons (VUV, 184.9 nm and 253.7 nm). The thermal, optical, and mechanical properties were characterized before and after exposure. The pristine polyimide showed a weight loss of about 50% after high energy photon exposure, while the hybrid polyimide composite showed a weight loss of only about 29%. The hybrid polymer composite showed a retarded degradation in mechanical properties (about 20% decrease in elongation at break of the hybrid polyimide composite versus about 85% decrease in elongation at break of the pristine polyimide). The radiation induced molecular degradation mechanism is discussed using various experimental techniques, including Fourier transform-infrared (FT-IR) spectroscopy, dynamic mechanical analysis (DMA), and electron paramagnetic resonance (EPR) spectroscopy. In addition, potential applications in the nuclear and aerospace industries will be discussed.

The Ni alloys generally used for the construction of the molten salt fast reactor can be prone to corrosion attacks and deleterious thermal effects on their mechanical properties at temperatures above 700°. In order to reduce corrosion and thermal damage, YSZ coatings are being deposited by sol-gel/dip-coating. Sol-gel derived coatings are prone to develop cracks due to internal stresses produced during drying. In order to improve the structural integrity of YSZ coatings we have studied the effect of solid concentration and the use of active and inactive additives in the sol-gel solution. Solid concentration in the gel from 30 to 50% has been varied together with the addition of YSZ, NI, C and Al additives with concentrations between 0-20 wt%. We have shown that is possible to produce defect-free YSZ coatings with a high concentration of cubic phase at temperatures below 700 °C.

Composite fuels offer opportunity to tailor fuel behavior during reactor transients. As a result understanding the properties of such composite fuels at microscale is critical for the design of the next generation nuclear reactors. Here we demonstrate the use of laser based experimental techniques to investigate thermal conductivity and microstructure of a coated particle fuel designed to self-regulate itself during a reactivity insertion transient. This fuel concept utilizes fuel negative reactivity feedback. A layer of textured pyrolytic carbon layer acting as a thermal barrier coating is introduced around a fuel particle to enhance the reactivity feedback. We characterized the conductivity of a thin anisotropic pyrolytic carbon layer deposited on the surface of a spherical zirconia particle used as a surrogate fuel. Deposition of textured PyC is important to achieve the lowest thermal conductivity in radial and large conductivity in circumferential directions. This ensures uniform rapid heating of the fuel under transients.
Laser based modulated thermoreflectance technique was used to characterize the anisotropic thermal transport with micrometer scale resolution. The results show that the thermal conductivity in radial direction is 0.28 W/m K, lower than previously reported, while in-plane conductivity is 11.5 W/m K. Microstructure was characterized using Raman spectroscopy. Large intensity of D peak at 1350 cm^-1 and position of G peak at 1580 cm^-1 indicate high level of disorder but confirm textured structure. Intensity ratio of D peak to G peak was used to determine an average grain size of 3.35 nm. Broadening of D peak width indicates some turbulence in the basal plane. A thermal conductivity model was used to analyze low thermal conductivity in circumferential direction as compared to highly ordered pyrolytic graphite and found to be consistent with the G peak broadening. Additional reduction in radial thermal conductivity was attributed to delamination and small pores between graphitic planes. This ultralow thermal conductivity coating offers attractive opportunities in thermal management applications.

Vanadium Carbide has been identified as a potential coating to zirconium-based clad in which it is intended to act as diffusion barrier, thereby mitigating fuel-clad chemical interaction, The self-diffusion behavior of vanadium subcarbide (V2C) is investigated using density functional theory calculations. Three ordered V2C structures, two of which correspond to experimentally observed phases, are characterized in terms of their equilibrium structural, electronic and elastic properties. Our model for self-diffusion in V2C considers diffusion of carbon and vanadium to occur separately on each sublattice. Two sets of self-diffusion coefficients are calculated for each structure: one for vacancy-mediated diffusion of vanadium and the other for interstitial diffusion of carbon. Calculated activation energies and diffusion prefactors are compared to experimental data where available. The diffusion of relevant extrinsic defects such as U, Fe, Ni and Nd is also discussed.

This work was supported by a DOE NEUP Award (DENE0000731).

Internal gelation is one of the most promising production methods when it comes to the fabrication of highly active, minor actinide containing fuel for the integration in a closed fuel cycle or for transmutation purpose. Compared to the classical pellet process, which involves powder handling and multiple mechanical preparation steps, the aqueous gelation can easily be integrated into hot cells, promising much lower contamination risk and maintenance effort.
At PSI numerous programs were conducted, where nuclear particle fuel, consisting of multiple size fractions of spheres, the so called sphere-pac, was produced by the internal gelation method, irradiated, and finally characterized in PIE studies. The fuel material varied from oxide, carbide to nitride ceramics. Over the years the production unit evolved to a continuous process, where the gelation triggering temperature increase was realized with hot silicon oil. In recent programs the immersion into the oil was replace by microwave heating, promising a further simplification of the process, by avoiding the oil contamination and a complex cleaning procedure. The presentation will discuss some of the conducted programs, the fuel performance and some issues.
In the newest approach the internal gelation is suggested to be integrated into an additive manufacturing process, allowing to fabricate arbitrarily shaped fuel bodies with varying internal composition. Furthermore the introduction of local porosity becomes possible, opening the option to introduce some tailored properties. In the presentation the suggested 3D printing is introduced with some potential nuclear fuel concepts.

Microstructural evolution (e.g. grain growth) in uranium dioxide (UO₂) nuclear fuels significantly affects nuclear fuel performance. The anisotropy of both the energy and mobility of grain boundaries determines to a large extent how the microstructure evolves. These boundary properties are also critical input parameters for mesoscale modeling of microstructural evolution. To better understand the anisotropy of these interfacial properties, circular grain boundaries in UO₂ were studied using molecular dynamics simulations. Grain boundary energies and mobilities were extracted for the high-symmetry rotation axes (<100>, <110>, and <111>). The calculated energies for the high-symmetry axes follow the five-dimensional grain boundary energy model well. The mobilities of grain boundaries were calculated across the three axes using the shrinking circular grain method. The <100> mobilities follow typical Arrhenius behavior across a range of high temperatures, followed by an abrupt shift to another Arrhenius-like trend with a much higher activation energy. Mobilities for the <111> axis were found to be the fastest boundaries. Mobilities for the <110> axis were found to be the slowest. The atomic transfer mechanisms across grain boundaries during grain growth are studied to explain the anisotropic mobilities of grain boundaries of different rotation axes.

Despite the high degree of safety of nuclear power plants in normal operating conditions, the risk of a Severe Accident (SA) is still present. As recently demonstrated by the Fukushima-Daiichi events, the three containment barriers of the reactor might fail leading to the release of highly radioactive elements in the environment. Among them, Barium (Ba) is of particular interest. This highly reactive Fission Product (FP) can interact with many other elements present in the fuel matrix such as U, O, Mo, Zr, Cs and Sr modifying its volatility [1], [2]. Furthermore, if Ba remains in the melted fuel, it can be responsible of up to 20% of the residual heat generated after the accident, and if released, it can severely damage human health. Hence, to better predict the progress and consequences of a SA, it is essential to develop realistic models for Ba behavior in the fuel in these conditions.
The development of such models requires experimental data on Ba speciation in the fuel and their comparison to thermodynamic predictions. Due to the limitations in terms of experiments and characterization techniques available up to now to study FPs speciation in irradiated nuclear fuels, model materials, referenced as SIMFuel, can be used [3], [4]. These SIMFuels are manufactured from depleted UO₂ doped with 11 stable oxides as FP surrogates and give access to powerful characterization techniques such as X-ray Absorption Spectroscopy (XAS) [5] [6], [7].
In this work, SIMFuel samples were annealed in different conditions representative of the early stages of a SA (temperatures up to 1700°C in both reducing and oxidizing atmospheres) since the FP behavior during this phase is crucial regarding the final stage. XAS experiments coupled with SEM-EDX analyses were performed after each annealing tests in order to study the evolution of Ba phases. The results tend to show that Ba remains under a zirconate form in reducing conditions whereas its local environment is modified to a molybdate form in oxidizing conditions above 1000°C.

[1] H. Kleykamp et al., J. Nucl. Mater., vol. 131, pp. 221–246, 1985.
[2] Y. Pontillon et al., Nucl. Eng. Des., vol. 240, pp. 1843–1852, 2010.
[3] E. Geiger et al., J. Phys. Conf. Ser., vol. 712, p. 012098, 2016.
[4] E. Geiger, PhD Thesis, Paris-Saclay, CEA Cadarache, 2016.
[5] I. Llorens et al., Rev. Sci. Instrum., vol. 83, p. 063104, 2012.
[6] E. Geiger et al., J. Nucl. Mater., vol. 471, pp. 25–33, 2016.
[7] C. Le Gall et al., Proceedings of the 8^th European Review Meeting on Severe Accident Research, Warsow, Poland, 2017

A previous study of lattice dilatation attributed the structural stability of the α-active ^238,239,240PuO₂, ²⁴¹AmO₂, and ²⁴⁴CmO₂ to radiation annealing [1]. We present a similar study on fuel-like (U,Pu)O₂, (U,²⁴¹Am)O₂, (Pu,²⁴¹Am)O₂ and (Pu,²⁴⁴Cm)O₂. As commonly supposed, an α-decay event creates ~1,500 Ac (actinide) and O Fps (Frenkel pairs), which cause lattice expansion. But despite continuous self-irradiation, accumulation of Fps is limited to a sufficiently low damage level, which prevents metamictization. In our model [1], ingrowth of Fps involves event-by-event production of local defect zones, and prompt annihilation of randomly-occurring closely-spaced interstitial-vacancy pairs. Guided defect mobility is generated by electronic and mechanical distortions, so activation energy is not required. At ambient temperature, mostly O Fps are expected to reside [1], so only similar Fps have been considered in our analyses. Any fresh interstitial or vacancy remains stable if created outside crystal volumes that engulf each a vacancy or interstitial surrounded by empty lattice sites of unstable complementary interstitials or vacancies, respectively. These critical annihilation volumes will expand if a few defect jumps will be generated by local transient thermal spikes. The model has been validated by molecular dynamics simulations on UO₂ at <5K [2]; interstitial and vacancy at closest separation and at some more distant separations combine promptly in both the U and O sublattices.

The current fourteen analyses and the ten previous analyses [1] have resulted in similar parameter values. The mean volume of damage that is created by a single α-decay event is 490±80 nm³, and thus covers ~3,070 unit cells. Damage is saturated even within non-overlapping damage sites. The mean fractional lattice dilatation at damage saturation is 0.0030±0.0003. A steady state of damage and recovery is reached under ~0.2 dpa (displacements per atom). Assuming that the volume increment per Fp is the mean volume of an atom in an undamaged crystal cell, only one surviving Fp is contained in ~9.9 unit cells and ~310 Fps reside in a single damage zone. These features, together with the observation that residual damage is independent of the nature, decay rates, and chemical properties of the Ac cations, manifest the great rigidity of the AcO₂ lattice.

Understanding of mechanisms that prevent radiation-induced crystalline-to-metamict transformations in materials such as nuclear fuels is of practical and academic interest.

1. Yehuda Eyal, A radiation annealing model for maintenance of crystallinity in self-damaged actinide dioxides, Proc. 6th Intern. Conf. on Radioac.Waste Manag. and Environ. Remed., Singapore, Oct 12-16, 1997, pp. 303-307.
2. Yehuda Eyal, Spontaneous annihilation of radiation-induced point defects in uranium dioxide: A molecular dynamics simulation, Proc. 7th Intern. Conf. on Radioac. Waste Manag. and Environ. Remed., Nagoya, Japan, Sept 26-30, 1999, pp. 761-766.

In order to restrict proliferation of nuclear weapons, highly enriched uranium (≥ 20% ²³⁵U) fuel is in the process of being replaced with low enriched uranium (< 20% ²³⁵U) fuel in research and test reactors. As the uranium enrichment is decreased, an increase in the uranium fuel density (> 8gU-235/cm³) is required for high performance research reactors. Accordingly, monolithic uranium-molybdenum (U-Mo) alloys, have been studied and considered as a nuclear fuel for these reactors. U-Mo alloy fuel exhibits excellent irradiation performance because the isotropic, body-centered cubic phase (γ-phase) of uranium is stable at the operating temperatures of research reactors (below 250°C). Understanding the effects of radiation damage on the atomic and microstructure of a nuclear fuel is essential to extend fuel lifetime and to enhance the fuel performance. Extended X-ray absorption fine structure (EXAFS) is a useful technique to study radiation damage of materials since it assists in extracting information on the local environment (~5-6 Å away from an absorbing atom) such as atomic distance and coordination number changes after irradiation. Thus, in the present study, low fluence neutron radiation damage of U-Mo alloys is investigated using EXAFS. Synchrotron X-ray diffraction (XRD) is also utilized to examine the microstructural changes in these alloys, including phase identification, phase fractions, crystallite sizes, and lattice parameters. Additionally, pure depleted U and Mo, are analyzed and compared using both techniques. This work will suggest a new approach for understanding radiation damage of not only U-Mo alloy fuel, but also depleted U and Mo.

In the last years a great effort was devoted to the modeling of uranium dioxide, starting from the atomic scale up to the fuel rod, with the aim of predicting the behavior of nuclear fuel and improving its safety [[i],[ii]]. All the modelling work has been performed with the hypothesis that UO₂ has Fm-3m, fluorite type, crystalline structure, but the existence of a lower symmetry local order in UO₂ was recently demonstrated [[iii]] and the consequences of this finding are still to be evaluated. Here we prove using neutron diffraction that the crystalline structure of UO₂ at 1000°C can be described as nanodomains having Pa-3 symmetry separated by domain walls, having a specific symmetry. This hypothesis of a nanodomain structure dramatically changes the interpretation of UO₂ structural changes as a function of temperature, which was previously based on local distortion due to polarons or point defects. These two interpretations are discussed.


[i] Linking atomic and mesoscopic scales for the modelling of the transport properties of uranium dioxide under irradiation
By: Bertolus, Marjorie; Freyss, Michel; Dorado, Boris; et al.
JOURNAL OF NUCLEAR MATERIALS Volume: 462 Pages: 475-495 Published: JUL 2015


[ii] Predicting material release during a nuclear reactor accident
By: Konings, Rudy J. M.; Wiss, Thierry; Benes, Ondrej
NATURE MATERIALS Volume: 14 Issue: 3 Pages: 247-252 Published: MAR 2015


[iii] What Is the Actual Local Crystalline Structure of Uranium Dioxide, UO2? A New Perspective for the Most Used Nuclear Fuel
By: Desgranges, L.; Ma, Y.; Garcia, Ph; et al.
INORGANIC CHEMISTRY Volume: 56 Issue: 1 Pages: 321-326 Published: JAN 2 2017

Significant quantities of xenon and krypton are produced in nuclear fuels under irradiation and the release of these fission gases induces an increase of the pressure inside the fuel pin and fuel swelling due to gas bubble formation. To prevent cladding failure, a better understanding of the fission gas release process is essential. The purpose of this work is to get further insight into the diffusion mechanisms of these fission gases and their interaction with defects in uranium dioxide. To do this, separated effects studies coupling ion irradiations/implantations and fine characterizations using Thermal Desorption Spectrometry (TDS), Positron Annihilation Spectroscopy (PAS) and Transmission Electronic Microscopy (TEM) have been performed.

TDS characterizations of UO₂ discs implanted with various xenon concentrations were performed on the PIAGARA (Plateforme Interdisciplinaire pour l’Analyse des GAz Rares en Aquitaine) platform at CENBG in Bordeaux to evaluate its thermal diffusion coefficient. The results show that xenon release decreases when the quantity of xenon implanted increases. This proves that fission gases are trapped in radiation induced defects. From the released fraction, we determined a model of gas release taking into account the initial burst and the diffusion in the bulk affected by the trapping effect. The fraction of gas trapped into UO₂ during annealing experiment determined by the model could be related afterwards with microstructure characterizations by PAS (vacancy defect characterization) and MET (cavities observation). Doppler broadening spectroscopy is performed using a slow positron accelerator on UO₂ samples with various xenon concentrations in as implanted state and after annealing at different temperatures. The results suggest that the main defect created after irradiation is the Schottky defect [1-2]. For a higher implantation dose, however, larger defects are detected probably due to vacancies aggregation. After annealing experiments, positrons probe large defects associated to cavities considering annihilation characteristics which is confirmed by TEM observation performed on those same samples.

The fraction of gas trapped determined by TDS measurements coupled to defect characterization by PAS experiments and TEM observations allows us to determine intrinsic diffusion coefficient of xenon and to identify the favorite trap sites related either to the Schottky defect or to the cavities depending on the fluence and annealing temperature. These results will enable us to describe the thermal behaviour of rare gases in UO₂ fuel starting from the atomic scale.

[1] H. Labrim et al. NIMB 261 (2007)
[2] T. Belhabib, PhD Thesis, Université d’Orléans (2012)

Radiation-induced phase transitions in hcp M_n+1AX_n phases (Ti₂AlN, Ti₂AlC, and Ti₄AlN₃) by 1 MeV Au⁺ ion irradiation were investigated, over a wide range of ion fluences from 1×10¹⁴ to 2×10¹⁶ cm^-2., by transmission electron microscopy (TEM) and synchrotron grazing incidence X-ray diffraction (GIXRD). The transformations of the initial hcp phases to the intermediate γ-phases and fcc phases were observed using high-resolution TEM (HRTEM) images and selected area electron diffraction (SAED). Based on phase contrast imaging and electron diffraction pattern (EDP) simulations, the atomic-scale mechanisms for the phase transitions were determined. By comparing the transformation behavior of Ti₂AlN with that of Ti₂AlC and Ti₄AlN₃ under the same irradiation conditions, using both the experimental data and first-principles calculations, the role of the X and n parameters in the radiation responses of different M_n+1AX_n phases are elucidated. The susceptibility of materials in this system to irradiation-induced phase transitions were determined with respect to the bonding characteristics and compositions of these MAX phases.