Accelerator-based lithium-target neutron sources produce fast neutron spectra ideal for nuclear data studies, but simulations of these neutron spectra are not always reliable. A probabilistic method for characterising neutron spectra with dosimetry foils has been developed, and used to measure a Li(p,xn) neutron spectrum from a new compact fast neutron source. Measurement of 7Be activation in the target was used to estimate total flux, and inventory simulations were used to evaluate the modelled spectrum and unfolded spectrum. Unfolding with Monte-Carlo uncertainty estimation was performed using probability distributions for reaction rate measurements and response functions. Comparison of neutronics model predictions to measurements showed the JENDL-5 proton library inaccurately models some neutron production mechanisms. Foil activities predicted with the unfolded neutron spectrum were consistent with experimental measurements for fast neutrons.

Samples of the AXIOM X2 Zr alloy exposed to neutron irradiation and corrosion in the Vogtle reactor for five years were studied using atom probe tomography. We have shown that the chemistry of the metal substrate below the oxide layer formed during aqueous corrosion has been substantially changed by the incorporation of Li from the coolant water, and, with the support of FISPACT-II inventory calculations, the formation of radioactive 94Nb and Mo isotopes as transmutation products from Zr and Nb. The findings highlight the importance of investigating the degradation mechanisms of Zr cladding alloys under real service conditions, and raise new questions on how these irradiation-driven chemical changes can influence both the in-service corrosion performance and the management of the radioactive waste at end of life.

Predicting with high accuracy (low uncertainty) the performance of tritium breeding systems is critical for the design of future fusion devices. Current modeling relies on limited experimental data and is known to have uncertainties that, if realized, could create a tritium shortfall in the fuel cycle of a power plant. The LIthium BReeding Tritium Innovation (LIBRTI) test facility is a United Kingdom program to develop a combined experimental test platform and model verification architecture to reduce these design uncertainties. A simple, flexible pin cell geometry containing industry-standard ceramic breeder material surrounding a market-leading deuterium-tritium-based neutron source, which will be installed within the LIBRTI facility, is found to produce sufficient levels of tritium to be readily measurable with known techniques for tritium accountancy.The modeling of the potential for interference of the measurements due to the permeation of tritium from the neutron source itself indicates that, depending on scenario, there will be at least hours, and potentially days or more, of viable measurement data in each experiment before noise will become an issue, even before considering any special design to mitigate the permeation. This analysis is used to demonstrate that LIBRTI will provide a critical platform for measurable testing of larger-scale tritium breeding experiments.

The EU-DEMO design is exploring several blanket models, including the Water Cooled Lithium Lead blankets. In this blanket model, liquid LiPb will be used as both the tritium breeder and multiplier. This liquid will experience a spatially varying neutron field as it moves across the tokamak. The production of radioisotopes would depend on the flow parameters of LiPb across this spatially varying neutron field. In this study, we have explored the production of radioisotopes in such a field and compared it with an equivalent, static, non-flowing simulation. While static calculations are simpler and computationally less expensive, we find that the predicted production of a radionuclide deviates from flowing fluid simulations to varying degrees, depending on the radionuclide’s half-life. We also demonstrate how the computational methodology developed can be used to explore online tritium extraction parameters and their impact on the End-of-Life tritium inventory.

Fundamental understanding of tungsten oxidation plays an important role in providing reliable and engineering data relevant to the detritiation, maintenance and waste handing for a future fusion device. The present simulations identify the key mechanisms controlling the capture and release of tritium from bulk tungsten and the oxides that may form on its surfaces. They further predict the rate of tritium release from tungsten and its oxides under variable environmental conditions and material states. These release rates sensitively depend on the interplay between tritium mass transport and tritium binding in WO\textsubscript{x} interstitial sites, lattice vacancies, and grain and interphase boundaries. We report results for the diffusion of tritium from tungsten through bulk phases and interfaces of W|\ce{WO2}|WO$_{2.72}$|\ce{WO3}, and the resulting tritium release from oxide surfaces.

Advancing the theory and simulation of materials for fusion applications remains a key component of global roadmaps aimed at delivering much-needed fusion power. Especially as the drive for commercial application increases, prototypes must be designed against radiation damage before the relevant experimental data can be collected and cost reductions that are possible by testing materials in silico become even more important. Here, we summarise the state of the art as it emerged during the 7th Fusion Materials Theory & Modelling Workshop that took place in 2024, with the aim to highlight present gaps and future directions for the fusion materials modelling community. Of particular interest were the effects of transmutations, chemical complexity with the development of novel alloys and interatomic potentials, advancements in modelling high-dose microstructures, comparison with experimental data and multiscale models for structural assessment relying on high-performance computing and virtual reality.

Accurate prediction of radiation damage in YBa2Cu3O7-5 (YBCO) is essential for assessing the performance of high-temperature superconducting (HTS) tapes in compact fusion reactors. Existing empirical interatomic potentials have been used to model radiation damage in stoichiometric YBCO, but fail to describe oxygen-deficient compositions, which are ubiquitous in industrial Rare-Earth Barium Copper Oxide conductors and strongly influence superconducting properties. In this work, we demonstrate that modern machine-learned interatomic potentials enable predictive modelling of radiation damage in YBCO across a broad range of oxygen stoichiometries, at a greater fidelity than previous empirical models. We employ two recently developed approaches: an Atomic Cluster Expansion (ACE) potential and a tabulated Gaussian Approximation Potential (tabGAP). Both machine-learned models are shown to accurately reproduce Density Functional Theory (DFT) energies, forces, and threshold displacement energy distributions, providing a quantitatively reliable description of atomic-scale collision processes. Molecular dynamics simulations of 5 keV cascades predict significantly enhanced peak defect production and recombination relative to a previous empirical potential, indicating qualitatively different cascade evolution. Moreover, these new machine learning models generate increased proportions of copper and oxygen vacancies compared to previous models, in direct agreement with experimental observations. By explicitly varying oxygen deficiency, we further show that total defect production exhibits only a weak dependence on stoichiometry, providing new insight into the robustness of radiation damage processes in oxygen-deficient YBCO. Finally, fusion-relevant 300 keV cascade simulations reveal amorphous regions with characteristic dimensions comparable to the superconducting coherence length, consistent with electron microscopy observations of neutron-irradiated HTS tapes. These results establish machine-learned interatomic potentials as powerful, computationally efficient tools for uncovering radiation-damage physics in YBCO, enabling predictive simulations across technologically relevant compositions and irradiation conditions.

Fundamental understanding of tungsten oxidation provides reliable engineering data relevant to detritiation, maintenance and waste handling for a future fusion device. Simulations identify the key mechanisms controlling the capture and release of tritium from bulk tungsten and the oxides that may form on its surfaces. They further predict the rate of tritium release from tungsten and its oxides under variable environmental conditions and material states. These release rates sensitively depend on the interplay between tritium mass transport and tritium binding in WOx interstitial sites, lattice vacancies, and grain and interphase boundaries. We report results for the diffusion of tritium from tungsten through bulk phases and interfaces of W|WO2|WO2.72|WO3, and the resulting tritium release from oxide surfaces. Unexpected findings include that, for the oxygen-terminated WO3 surface considered here, tritium is more likely to be released as molecular T2O than as T2; the effects of environmental O/H partial pressures are not resolved. Tritium release rates have been quantified and mapped as a function of temperature and time using the multiscale computational capabilities of the MedeA simulation environment. The results reveal the dominant role of defect trapping in bulk tungsten and its oxide phases, and the relative influence of a range of retarding mechanisms due to phase interfaces and defects in the oxide phases. In tungsten with very high vacancy concentrations, vacancy trapping is expected to be the rate-limiting mechanism for tritium release. For the baseline vacancy concentration, the simulations predict that greater than 90% of the initial tritium inventory escapes within one hour at temperatures of 900 K and above. At 700 K, a significant fraction remains trapped over the same period, indicating that rapid cooling below this temperature could meaningfully extend tritium retention times. This finding could have significant engineering implications for post-shutdown tritium recovery procedures, depending on the extent to which oxidation can be controlled.

Neutron bombardment of high temperature superconducting (HTS) magnets may compromise the integrity of the magnetic confinement in future fusion reactors. The amount of damage produced by a single neutron can be predicted from the threshold displacement energies (TDE) of the constituent ions in the HTS materials, such as the Rare Earth Cuperates. Therefore, in this work a Multiphysics simulation approach is adopted to determine the threshold displacement energies for oxygen in YBa2Cu3O7. Classical molecular dynamics (MD) simulations are employed to determine statistically representative TDEs for all four oxygen sites and these results are validated using Born-Oppenheimer MD employing forces derived from Density Functional Theory (DFT). The simulations were performed at the operational temperature (25 K) and the temperature of existing neutron irradiation studies (360 K) enabling a discussion about the relevance of this data. Overall, these findings enhance our understanding of radiation-induced damage in HTS materials and provide data that can be incorporated into higher order models offering critical insights into shielding design and magnet longevity.

The development of deuterium–tritium (D–T) fusion power plant prototypeswill be constrained by the global availability of tritium. This paper presentsan up-to-date analysis of current and projected tritium supply and demand,spanning near-term fusion devices and the longer-term deployment of fusionpower plants. Tritium production from CANDU and other heavy-water reac-tors is modelled using reactor-specific power output and load factor data fromthe IAEA PRIS (Power Reactor Information System) database, accountingfor operational lifetimes, refurbishments, and tritium extraction capabilities.Three supply scenarios yield peak theoretical global stockpiles of between35.1 and 47.5 kg. These inventories are assessed against low, medium, andhigh demand scenarios that include ITER, DEMO-class reactors, and an ex-panding fusion private sector. ITER’s requirements of 15.9 kg can be satisfiedin all supply cases, but significant deficits arise once multiple DEMO-classand private sector fusion reactors are introduced. Under pessimistic supplyand high-demand assumptions, shortfalls exceed 28 kg by the mid-2050s, ris-ing to 42 kg in 2070, and reach 180 kg in 2070 if fusion power plants requireeven a 10% external top-up due to imperfect tritium breeding. Looking fur-ther ahead, if fusion were to supply 10% of global electricity demand by 2100,annual tritium consumption could exceed 600 tonnes. These findings high-light the urgency of reducing start-up inventory requirements, acceleratingthe development of high-performance breeding blankets and fuel cycles, andexploring supplementary production pathways such as optimised TPBARsystems, if tritium availability is not to become a significant bottleneck inthe commercial rollout of fusion energy.

The addition of Cr and Y into tungsten can dramatically increase the oxidation resistance of the first wall of a future fusion reactor, thereby reducing the risk of formation of volatile WO3 and the release of radioactive material. Experimental observations suggest that in these SMART alloys, yttrium facilitates the formation of a self-passivating layer of Cr2O3 at the metal surface, however, how exactly the Y does this remains unclear. Therefore, this work explores the phase stability of compounds consisting of W–Y–Cr–O and solution energies for the different components in tungsten using density functional theory. The simulations suggest that there is a substantial thermodynamic driving force for the formation of Y2O3, especially from yttrium and oxygen solvated in bulk tungsten. These observations suggest that the role of the yttrium may be to remove the oxygen that may inhibit Cr diffusion to the surface from the tungsten grains. This observation is in accordance with experimental studies showing that the oxidation resistance in the alloy occurs when the oxygen–yttrium ratio in the alloy is close to the stoichiometric ratio for Y2O3.

14 Cis produced from the reactions of O isotopes with high-energy neutrons. Due to its long half-life, high residence time in the environment and ease of assimilation into living matter, the production of 14 C is a major concern when planning for the transportation and disposal of radioactive materials. Observed variation in calculated 14 C production from O isotopes using different data libraries causes uncertainties that are demonstrated to affect predictions of radioactive waste arisings and thus could impact disposal and recycling recommendations. There are also significant inconsistencies in the uncertainty data and co-variance matrices given in the nuclear data files for O isotopes.

The oxidation of pure W and the sublimation of W oxide have been investigated to assess their impact on the lifecycle of a fusion power plant. Pure W has been oxidised at temperatures between 400 and 1050 °C and for durations ranging between 1 and 70 h. The formation of voids and cracks has been observed at temperatures above 600 °C, leading to the formation of dust or oxide spalling, which could be problematic in maintenance and waste-handling scenarios of a fusion power plant. Preferential oxidation taking place at the edge of the specimen was characterised, and its impact is discussed in relation to component design. Characterisation using electron microscopy and Raman spectroscopy revealed that the oxide scale is formed of three main layers: the inner layer is 30–50 nm thick WO2 oxide, the middle layer is a 10–20 μm thick of WO2.72 and the outer layer is formed of WO2.9/WO3 phases — whose thickness varies according to the total thickness of the oxide scale. The observed microstructure is discussed in relation to the parabolic-to-linear kinetics and its potential impact on tritium permeation and detritiation efficiency.

In this paper we present a systematic investigation into the relative stability of hydrogen isotopes within body-centered-cubic (bcc) Fe using the density functional theory (DFT) method with zero-point energy and elastic corrections. We focus on hydrogen interactions in the bulk, trapping in monovacancies and adsorption on clean low-index surfaces. Hydrogen interactions are further considered with Cr in the dilute limit within the Fe matrix. We show that all hydrogen isotopes are highly mobile in the bulk and reside in tetrahedral interstitials, but the diffusivity decreses with mass. Protium migrates via the octahedral site whilst tritium rotates into an adjacent tetrahedral site. The activation barrier for both pathways is degenerate for deuterium which becomes the most mobile isotope at fusion operating temperatures. However, anharmonicity has not been considered. Hydrogen is repelled by Cr which tends to increase migration barriers by up to 20 meV. Trapping energies of deuterium in a mono-vacancy closely match desorption stages from published ion-beam implantation-annealing experiments. The binding of the heavier isotopes is more exothermic than protium. However, dissolution energies between isotopes is typically within 10 meV such that we do not expect practical kinetic isotope effects at operating temperatures. Cr substitutions near a vacancy change the relative incremental binding energies and push hydrogen away from nearoctahedral sites. The binding energy to a Cr-vacancy complex is lower for 1-3 hydrogen atoms relative to a vacancy in the absence of Cr but sincrease as the complex becomes saturated. The binding energy of heavier hydrogen isotopes to both (100) and (110) surfaceis lower than for protium. Subsurface Cr is shown to change the preferred hollow adsorption site to a bridge-site absorption on a (100) surface, increasing the binding energy by as much as 50%. Finally, we present elastic dipole tensors and relaxation volumes of the hydrogen-implanted systems, which can be used to represent source terms in continuum models of irradiated materials.

We develop an atomic cluster expansion (ACE) interatomic potential for lithium that accurately models both the solid and liquid phase and the corresponding melting point. The training data is obtained from 0 K density functional theory (DFT) and finite temperature ab initio molecular dynamics simulations of both solid and liquid Li. The ACE predicted properties for both phases obtained from molecular dynamics simulations are in close agreement with DFT and experimental data from the literature. The potential is able to capture the energy differences of the different competing phases of the solid at 0 K and finite temperature properties of the experimentally observed bcc phase. The potential also accurately predicts the temperature dependence of liquid density, viscosity, and the diffusion coefficient. The melting point is calculated using the two-phase coexistence method and is remarkably close to the experimental value. The potential is used to predict stress-induced phase transformations in solid Li and pressure-volume isotherms in liquid Li. We underline the necessity for a complete training set that includes both solid and liquid configurations in order to obtain a potential that precisely models both phases. By using the ACE formalism, we also systematically investigate the contributions of interactions involving N bodies and the number of radial parameters needed to separately represent both phases since they have a direct consequence on the computational cost of the potential. We shed light on the complexity of the ACE potential needed to model solid and liquid lithium efficiently.

14C is produced from the reactions of O isotopes with high-energy neutrons. Due to its long half-life, high residence time in the environment and ease of assimilation into living matter, the production of 14C is a major concern when planning for the transportation and disposal of radioactive materials. Observed variation in calculated 14C production from O isotopes using different data libraries causes uncertainties that are demonstrated to affect predictions of radioactive waste arisings and thus could impact disposal and recycling recommendations. There are also significant inconsistencies in the uncertainty data and co-variance matrices given in the nuclear data files for O isotopes.

Superconducting material enables fusion reactor magnet concepts to operate with current densities that would melt materials with non-zero resistance. The application of superconducting material is considered essential for net-positive power machines. Catastrophic damage can occur when superconductivity is lost and the current generates heat. This scenario is called a quench. Stabilizer material carries the magnet current (typically copper) during a quench and is the focus of this work. Irradiation-induced defects store energy in the Cu crystalline lattice. The presence of defects reduces thermal conductivity (thermally insulating the superconductor), electrical conductivity (increasing temperature ramp rate during a quench), and specific heat capacity (increasing thermodynamic instability). The release of stored energy in the magnet materials, in combination with the magnet material property changes, has the potential to cause extreme off-normal events in superconducting magnets that worsen with fluence. Stored energy can be released causing local heating and increasing the risk of a quench. For example, following irradiation at 4.6 K and a fluence of 0.45x1018 n cm-2, an energy release of 0.023 J g-1 was measured from Cu when increased in temperature from 10 K to 18 K, which would have been enough energy to create the same temperature increase spontaneously. Extrapolations of experimental data are used to estimate when spontaneous heating can occur due to the release of energy stored in irradiation-induced defects. Critical fluence values are estimated between 1.74x1018 n cm-2 and 2.85x1019 n cm-2 for neutron irradiation of Cu at a temperature of 20 K. In-situ cryogenic calorimetry experiments, operated at high-temperature-ramp rates on irradiated magnet materials, could offer certainty for fusion magnet system designers. Periodic annealing of defects through controlled temperature cycling will be essential in fusion power plants to manage the increasing risk of quench as the superconducting magnets accumulate dose. The ideal frequency and dynamics of these maintenance temperature cycles will be established with further experimental examination.

Tritium production in a fusion reactor is essential for a deuterium-tritium burning plasma source, and designing a lithium-containing breeder blanket is crucial for a closed fuel cycle in an electricity-to-grid power plant. Tritium is a rare resource, currently produced only in a limited number of fission reactors and other low-output facilities. Due to the lack of experimental data in this area, neutronics calculations will guide the design of the first-generation fusion blankets tasked with tritium fuel production as without maximised breeding potential, reactor startup times may need to be postponed due to a low global inventory. This paper describes an optimisation workflow that segments a DEMO-style blanket to allow flexibility in material allocation, along with the necessary validation steps for repeatable results. The developed blanket is tested within the neutronics code OpenMC to evaluate the performance of a hybrid liquid metal–molten salt breeder. FLiBe is found to be an efficient neutron reflector, showing performance increases of 7%, resulting in an overall tritium breeding ratio (TBR) of 1.11, even with a natural abundance of 7.5 % lithium-6. Analysis of the neutron spectral profile indicates a shift towards thermal neutrons when reflected from FLiBe placed at the rear of the design. The use of molten salt as a reflector is novel as a result of its ability to breed tritium and provide a secondary increase in TBR through the reflection of neutrons. A dual system would require more complex engineering, but may offer a solution for compact reactors and/or systems with low lithium enrichment.