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Nuclear data, describing neutron reaction probabilities (cross sections) and decay behaviour, are critical to the design and operation of fusion experiments and future fusion power plants. Equally vital, are the inventory codes that use the data to predict neutron-induced activation and transmutation of materials, which will define the radiological hazards that must be managed during reactor operation and decommissioning. Transmutation, including gas production, combined with the neutron-induced displacement damage, will also cause the properties of materials to degrade, for example through swelling and embrittlement, eventually limiting the lifetime of components. Thus validated and accurate nuclear data and inventory codes are essential. For data validation there are decay heat measurements performed at FNS in Japan more than 20 years ago. The experiments produced an invaluable database for benchmarking of nuclear data libraries; the latest versions of several international libraries perform well against this data during tests with the FISPACT-II inventory code, although there is still scope for improvement. A recent attempt to provide fusion-relevant validation based on γ-spectroscopy data from neutron-irradiated material samples tests produced predictions for short-lived (several hours or less) radionuclides. The detailed analysis performed for molybdenum demonstrates how these data could eventually provide a new benchmark, and also illustrates the potential benefits of further experiments targeting the longer-lived radionuclides relevant to maintenance and decommissioning timescales. There are also some successful tests of transmutation predictions with FISPACT-II. These direct validations of inventory simulations are critical for lifetime predictions and future experiments should learn lessons from the examples described for tungsten, which demonstrate the importance of an accurate description of the neutron spectrum in experiments. More novel experimental techniques are needed to measure helium production in materials such as Fe and C, but the need to validate the nuclear data evaluations used by simulations should motivate future experimental efforts.

Reduced activation ferritic martensitic (RAFM) and oxide dispersion strengthened (ODS) steels are the most promising candidates for fusion first-wall/blanket (FW/B) structures. The performance of these steels will deteriorate during service due to neutron damage and transmutation-induced gases, such as helium/hydrogen, at elevated operating temperatures. Here, after highlighting the operating conditions of fusion reactor concepts and a brief overview, the main irradiation-induced degradation challenges associated with RAFM/ODS steels are discussed. Their long-term degradation scenarios such as (a) low-temperature hardening embrittlement (LTHE)—including dose-temperature dependent yield stress, tensile elongations, necking ductility, test temperature effect on hardening, Charpy impact ductile-to-brittle transition temperature and fracture toughness, (b) intermediate temperature cavity swelling, (c) the effect of helium on LTHE and cavity swelling, (d) irradiation creep and (e) tritium management issues are reviewed. The potential causes of LTHE are discussed, which highlights the need for advanced characterisation techniques. The mechanical properties, including the tensile/Charpy impact of RAFM and ODS steels, are compared to show that the current generation of ODS steels also suffers from LTHE, and shows irradiation hardening up to high temperatures of ∼400 °C–500 °C. To minimise this, future ODS steel development for FW/B-specific application should target materials with a lower Cr concentration (to minimise α′), and minimise other elements that could form embrittling phases under irradiation. RAFM steel-designing activities targeting improvements in creep and LTHE are reviewed. The need to better understand the synergistic effects of helium on the thermo-mechanical properties in the entire temperature range of FW/B is highlighted. Because fusion operating conditions will be complex, including stresses due to the magnetic field, primary loads like coolant pressure, secondary loads from thermal gradients, and due to spatial variation in damage levels and gas production rates, an experimentally validated multiscale modelling approach is suggested as a pathway to future reactor component designing such as for the fusion neutron science facility.

Using the notion of eigenstrain produced by the defects formed in a material exposed to high energy neutron irradiation, we develop a method for computing macroscopic elastic stress and strain arising in components of a fusion power plant during operation. In a microstructurally isotropic material, the primary cause of macroscopic elastic stress and strain fields is the spatial variation of neutron exposure. We show that under traction-free boundary conditions, the volume-average elastic stress always vanishes, signifying the formation of a spatially heterogeneous stress state, combining compressive and tensile elastic deformations at different locations in the same component, and resulting solely from the spatial variation of radiation exposure. Several case studies pertinent to the design of a fusion power plant are analysed analytically and numerically, showing that a spatially varying distribution of defects produces significant elastic stresses in ion-irradiated thin films, pressurised cylindrical tubes and breeding blanket modules., Comment: To be published in Nuclear Fusion

During operation fusion reactor components will be exposed to long periods of neutron irradiation. As such, a reactor’s structural steels will become activated and need to be disposed of as radioactive waste. Previous studies have shown that such wastes can struggle to meet low level waste (LLW) requirements meaning that costly geological disposal may be required. In order to explore the waste expectations of steels from European DEMO-like fusion reactors, several radioactive waste management systems have been investigated. This includes their LLW criteria, currently available disposal sites and planned future developments. This information was used to analyse the results of DEMO-like inventory simulations of potential reactor steels. The simulations were performed with the inventory code FISPACT-II and the TENDL2017 nuclear data library. The results suggest that when steels are exposed to near plasma neutron fluxes they will struggle to meet the majority of LLW requirements. For lower neutron fluxes, typical of reactor containment vessels, the waste expectations can be more positive, with several steels able to meet some low level criteria. It can be concluded that steels should not be expected to be consistently internationally classified as LLW 100 years after reactor shut down. As all activated fusion waste cannot be disposed of in a single location, it is recommended that waste disposal strategies are included in any fusion reactor proposal before construction begins. These strategies need to align with the radioactive waste regulations the proposed reactor will be subject to.

Molybdenum is a potential material for future nuclear fusion experiments and power plants. It has good thermo-mechanical properties and can be readily fabricated, making it attractive as an alternative material to tungsten (the current leading candidate) for high neutron flux and high thermal load regions of fusion devices. Unfortunately, exposure to fusion neutrons is predicted to cause significant radioactivity in elemental Mo for decades and centuries after exposure, which would be a problem during maintenance and decommissioning operations. Simulation predictions indicate that Mo activation could be reduced by isotopic adjustment (biasing). If these predictions are proven and validated, and if isotopic adjustment is technically and economically feasible, then Mo could be used in future demonstration and commercial reactors without significantly increasing the amount of long-term, higher-level radioactive waste. Transmutation (inventory) simulations used to predict activation rely on nuclear reaction data. The quality of these data impact on the confidence and uncertainty associated with predictions. Recently, UKAEA has developed benchmarks to test and validate the FISPACT-II inventory code and the input nuclear data libraries. Verification of molybdenum inventory simulations is performed against experimental decay-heat measurements from JAEA’s fusion neutron source (FNS) facility and using new data acquired from γ-spectroscopy measurements of Mo irradiated in the ASP 14 MeV facility in the UK. Results demonstrate that FISPACT-II predictions (with TENDL-2019 nuclear data) for Mo are accurate on the short-timescales (minutes, hours of irradiation and minutes, days, weeks of cooling) of these laboratory experiments. However, these kinds of experiments are limited in their coverage of the important radionuclides for decay radiation from Mo on the years, decades and beyond timescales. Further experiments with fusion relevant conditions and timescales, potentially with alternative measurement techniques, are still needed.

Waste-production predictions for the future demonstration fusion power plant (DEMO) are necessary to produce an accurate picture of the likely environmental and economic costs of radioactive waste disposal at end-of-life. An integrated simulation process combining Monte-Carlo neutron transport simulations, inventory calculations, and extensive and reproducible post-processing algorithms has been used for the evolving European DEMO designs to quantify the time-varying mass inventories in different waste classes for individual regions and components of the reactor vessel, as well as for the reactor as a whole. Waste categories based on UK and French regulations reveal that minor impurities contained in structural steels, particularly Eurofer, as well as in functional materials such as tungsten, and beryllium, can have a significant impact on their waste classification prospects. Predictions for current European DEMO concepts suggest that there may be an issue in disposing of fusion structural-steel waste as low-level waste (LLW) in near-surface repositories. Detailed analysis of the subtleties of these predictions, particularly with regard to the production of long-lived radionuclides such as 14C and 94Nb, reveal that the threshold for acceptance as LLW is only just exceeded in some situations. Several mitigation approaches are discussed in this context. The computational framework developed for these assessments can be rapidly and continuously applied to the maturing DEMO design, helping to guide design choices to mitigate long-lived waste production and ensure that most waste becomes LLW or better within a few decades.

To address the needs of activation-transmutation simulation in incident-particle fields with energies above a few hundred MeV, the FISPACT-II code has been extended to splice TENDL standard ENDF-6 nuclear data with extended nuclear data forms. The JENDL-2007/HE and HEAD-2009 libraries were processed for FISPACT-II and used to demonstrate the capabilities of the new code version. Tests of the libraries and comparisons against both experimental yield data and the most recent intra-nuclear cascade model results demonstrate that there is need for improved nuclear data libraries up to and above 1 GeV. Simulations on lead targets show that important radionuclides, such as 148Gd, can vary by more than an order of magnitude where more advanced models find agreement within the experimental uncertainties.

Under the anticipated operating conditions for demonstration magnetic fusion reactors beyond ITER, structural and plasma-facing materials will be exposed to unprecedented conditions of irradiation, heat flux, and temperature. While such extreme environments remain inaccessible experimentally, computational modeling and simulation can provide qualitative and quantitative insights into materials response and complement the available experimental measurements with carefully validated predictions. For plasma-facing components such as the first wall and the divertor, tungsten (W) has been selected as the leading candidate material due to its superior high-temperature and irradiation properties, as well as for its low retention of implanted tritium. In this paper we provide a review of recent efforts in computational modeling of W both as a plasma-facing material exposed to He deposition as well as a bulk material subjected to fast neutron irradiation. We use a multiscale modeling approach-commonly used as the materials modeling paradigm-to define the outline of the paper and highlight recent advances using several classes of techniques and their interconnection. We highlight several of the most salient findings obtained via computational modeling and point out a number of remaining challenges and future research directions.

Inventory calculations have a key role to play in designing future fusion power plants because, for a given irradiation field and material, they can predict the time evolution in chemical composition, activation, decay heat, gamma-dose, gas production, and even damage (dpa) dose. For conceptual designs of the European DEMO fusion reactor such calculations provide information about the neutron shielding requirements, maintenance schedules, and waste disposal prospects; thereby guiding future development. Extensive neutron-transport and inventory calculations have been performed for a reference DEMO reactor model with four different tritium-breeding blanket concepts. The results have been used to chart the post-operation variation in activity and decay heat from different vessel components, demonstrating that the shielding performance of the different blanket concepts—for a given blanket thickness—varies significantly. Detailed analyses of the simulated nuclide inventories for the vacuum vessel (VV) and divertor highlight the most dominant radionuclides, potentially suggesting how changes in material composition could help to reduce activity. Minor impurities in the raw composition of W used in divertor tiles, for example, are shown to produce undesirable long-lived radionuclides. Finally, waste classifications, based on UK regulations, and a recycling potential limit, have been applied to estimate the time-evolution in waste masses for both the entire vessel (including blanket modules, VV, divertor, and some ex-vessel components) and individual components, and also to suggest when a particular component might be suitable for recycling. The results indicate that the large mass of the VV will not be classifiable as low level waste on the 100 year timescale, but the majority of the divertor will be, and that both components will be potentially recyclable within that time.

Accurately quantifying the transmutation rate of tungsten (W) under neutron irradiation is a necessary requirement in the assessment of its performance as an armour material in a fusion power plant. The usual approach of calculating average responses, assuming large, homogenised material volumes, is insufficient to capture the full complexity of the transmutation picture in the context of a realistic fusion power plant design, particularly for rhenium (Re) production from W. Combined neutron transport and inventory simulations for representative spatially heterogeneous high-resolution models of a fusion power plant show that the production rate of Re is strongly influenced by the surrounding local spatial environment. Localised variation in neutron moderation (slowing down) due to structural steel and coolant, particularly water, can dramatically increase Re production because of the huge cross sections of giant resolved resonances in the neutron-capture reaction of 186W at low neutron energies. Calculations using cross section data corrected for temperature (Doppler) effects suggest that temperature may have a relatively lesser influence on transmutation rates.

The high-energy, high-intensity neutron fluxes produced by the fusion plasma will have a significant life-limiting impact on reactor components in both experimental and commercial fusion devices. As well as producing defects, the neutrons bombarding the materials initiate nuclear reactions, leading to transmutation of the elemental atoms. Products of many of these reactions are gases, particularly helium, which can cause swelling and embrittlement of materials. This paper integrates several different computational techniques to produce a comprehensive picture of the response of materials to neutron irradiation, enabling the assessment of structural integrity of components in a fusion power plant. Neutron-transport calculations for a model of the next-step fusion device DEMO reveal the variation in exposure conditions in different components of the vessel, while inventory calculations quantify the associated implications for transmutation and gas production. The helium production rates are then used, in conjunction with a simple model for He-induced grain-boundary embrittlement based on electronic-structure density functional theory calculations, to estimate the timescales for susceptibility to grain-boundary failure in different fusion-relevant materials. There is wide variation in the predicted grain-boundary-failure lifetimes as a function of both microstructure and chemical composition, with some conservative predictions indicating much less than the required lifetime for components in a fusion power plant.

W and W-alloys are among the primary candidate materials for plasma-facing components in the design of fusion reactors, particularly in high-heat-flux regions such as the divertor. Under neutron irradiation W undergoes transmutation to its near-neighbours in the periodic table. Additionally He and H are particles emitted from certain neutron-induced reactions, and this is particularly significant in fusion research since the presence of helium in a material can cause both swelling and a strong increase in brittleness. This paper presents the results of inventory burn-up calculations on pure W and gives quantitative estimates for He production rates in both a fusion-reactor environment and under conditions expected in the ITER experimental device. Transmutation reactions in possible alloying elements (Re, Ta, Ti and V), which could be used to reduce the brittleness of pure W, are also considered. Additionally, for comparison, the transmutation of other fusion-relevant materials, including Fe and SiC, are presented.

The most recent observations of dynamical time-dependent fluctuating behaviour of mesoscopic radiation defects in body-centred cubic metals (Arakawa et al 2006 Phys. Rev. Lett. 96 125506; 2007 Science 318 956–9; Yao et al 2008 Phil. Mag. at press) have highlighted the need to develop adequate quantitative models for the structural stability of defects in the mesoscopic limit where defects are accessible to direct in situ electron microscope imaging. In pursuit of this objective, we investigate and compare several types of mesoscopic vacancy and interstitial defects in iron and tungsten by simulating them using recently developed many-body interatomic potentials. We show that the mesoscopic vacancy dislocation loops observed in ion-irradiated materials are, without exception, metastable with respect to the transformation into spherical voids, but that the rate of this transformation and even the specific type of the transformation mechanism depend on the defect size and the properties of the material.