Future Reactors Research Articles

Helium's impact: Unraveling bubble formation in Fe2AlB2 under extreme conditions of temperature and fluence

Layered carbides and borides, blending ceramic and metallic characteristics, present compelling prospects for future nuclear reactor applications due to their complex nanolaminate crystal structure. To explore the behavior of Fe2AlB2 under different conditions, helium irradiation is conducted across saturation fluences and under different annealing and irradiation temperatures. Room temperature (RT) irradiation causes some changes: a-lattice expands 0.3 %, c-lattice contracts 0.07 %, and an embedded amorphous layer into the matrix is formed. 400 °C annealing keeps an amorphous structure whereas 800 °C annealing fully recrystallizes with bubble growth. Increasing the irradiation temperature to 400 °C with the same fluence, smaller bubbles form without an amorphous layer, indicating defect recombination depends on helium attachment to vacancies. Under irradiation at 700 °C, faceted bubbles develop exclusively within Fe2AlB2, with no such bubbles in Al2O3. These faceted bubbles align themselves along (110) planes. Simultaneously, irradiation prompts tangled dislocations within the damaged layer, fostering the congregation of polygonal bubbles along grain boundaries (GBs) and creating denuded areas at GBs. Fe2AlB2 demonstrates remarkable oxidation resistance under high-temperature irradiation, maintaining surface stability without the cracks seen in ion-irradiated Ti3SiC2 and Ti2AlC at RT.Fe2AlB2 shows promise for use in reactors operating at temperatures above 500 °C due to its resilience to irradiation. Further investigation is warranted for its application in reactors operating in environments exceeding this threshold temperature.

Radiation shielding properties of composites of TiZrNbHfTa refractory high entropy alloy reinforced with TiZrNbHfTaOx high-entropy oxide

Radiation shielding materials play a critical role in advancing the next generation of nuclear reactors. Among these, high-entropy materials with refractory properties emerge as a new category for radiation shielding applications. This study introduces TiZrNbHfTa-TiZrNbHfTaOx composites tailored for such purposes. Initially, a TiZrNbHfTa refractory high-entropy alloy (RHEA) is synthesized via mechanical alloying, followed by the oxidation of the alloy to obtain its refractory high-entropy oxide (RHEO), TiZrNbHfTaOx. Composites are then prepared by blending 5%, 10%, and 20% weight fractions of RHEO with RHEA using mechanical alloying. Comparative analysis of shielding properties reveals a decrease in mass attenuation coefficient and effective atomic number with increasing RHEO content. While RHEA exhibits superior shielding performance against gamma photons and neutrons, the composite containing 20% RHEO emerges as the optimal shield against alpha and proton irradiations. The equivalent neutron dose rate and removal cross-section values are obtained 49.8% and 0.142 for the composite with 20% RHEO, respectively. These findings, supported by computational simulations in this study, contribute valuable insights for the advancement of novel shielding materials crucial for future nuclear reactor technologies.

Stochastic processes shape microbial community assembly in biofilters: Hidden role of rare taxa

Stochastic and deterministic processes are the major themes governing microbial community assembly; however, their roles in bioreactors are poorly understood. Herein, the mechanisms underlying microbial assembly and the effect of rare taxa were studied in biofilters. Phylogenetic tree analysis revealed differences in microbial communities at various stages. Null model analysis showed that stochastic processes shaped the community assembly, and deterministic processes emerged only in the inoculated activated sludge after domestication. This finding indicates the dominant role of stochastic factors (biofilm formation, accumulation, and aging). The Sloan neutral model corroborated the advantages of stochastic processes and mainly attributed these advantages to rare taxa. Cooccurrence networks revealed the importance of rare taxa, which accounted for more than 85% of the keystones. Overall, these results provide good foundations for understanding community assembly, especially the role of rare taxa, and offer theoretical support for future community design and reactor regulation.

Study of helium diffusion in yttria: A multiscale approach based on the density functional theory and kinetic Monte Carlo, with transmission electron microscopy and thermo-desorption spectroscopy

A known issue for future nuclear reactors is helium accumulation inside the steel structure materials, responsible for structural issues such as embrittlement and cracking. One possible solution is using new types of reinforced steel, such as oxide dispersion strengthened (ODS) steel. It consists of adding oxide nanoparticles to the Fe-based material, especially yttrium oxide (yttria, Y2O3), improving its properties. Therefore, one first step is understanding the helium diffusion inside this system. Very little is known about helium inside yttria, with most studies being theoretical ones. Based on this context, this work proposes a combined theoretical and experimental multiscale approach to investigate helium diffusion inside yttria.The theoretical approach starts with the density functional theory, used to model the atomic yttria cell and determine helium insertion sites. The transitions between the sites were described using the Nudged Elastic Bond (NEB) method. Kinetic Monte Carlo was then employed to obtain the interstitial diffusion coefficient expression for the first time for this material. It showed a limited diffusion at temperatures below 600 K, which may indicate a tendency for He to be blocked in the oxide. Then, the charged vacancies were explored. It showed that the vacancy further reduces helium diffusion. Finally, it was demonstrated that helium spreads across different vacancies and interstitial sites.The experimental part involved implanting helium ions at 50 keV in samples with nanometric or micrometric grains. Then, the specimens were characterised with transmission electron microscopy (TEM) and thermo-desorption spectroscopy (TDS) techniques. TEM did not evidence detectable bubbles even at the highest studied fluence (1 × 1016 cm−2). The TDS highlighted different mechanisms for helium diffusion and the grain size's role, providing a novel model for diffusion coefficient calculation based on interstitial diffusion.

Status of the MINERVA cryomodules and associated cryogenic system (MYRRHA phase 1)

MYRRHA at SCK CEN in Mol/Belgium will be a pre-industrial large-scale Accelerator Driven System for unparalleled research opportunities in spent nuclear fuel, nuclear medicine, and fundamental and applied physics. In 2018 the Belgian government released funding for MYRRHA’s first phase, MINERVA, for a staged implementation and operation. It covers the design, construction, and commissioning of the continuous-wave superconducting RF proton linac up to 100 MeV, as well as dedicated user target stations. The MINERVA proton linac will accommodate 30 identical cryomodules to boost the beam energy delivered by the normal-conducting frontend from 16.6 MeV to 100 MeV. Each cryomodule will contain two superconducting RF single-spoke cavities immersed in a superfluid Helium bath at 2 K. The design and architecture of the associated cryogenic system is derived from the stringent linac reliability requirements imposed by the future subcritical nuclear reactor. We present the architecture, design, and development status of the MINERVA cryomodules and associated cryogenic system towards the implementation phase of the project as part of a collaboration between ACS, CEA/DSBT, IJCLab, and SCK CEN. We also provide an overview of the initial outcomes of cryogenic and RF tests for the prototype MINERVA cryomodule, which are still ongoing at IJCLab.

VINUS: A neutron transport solver based on the variational nodal method for reactor core analysis

Compared to traditional transverse integration methods, the variational nodal method, with its unique advantages, is more suitable for high-fidelity calculations of reactor physics in reactors with complex geometries and finer detail descriptions. In this study, the basic theory of the variational nodal method was derived and the VINUS code is developed. The neutron solver based on this method is adaptable to various geometric models, and showcased the code's fundamental framework. On this basis, a set of self-designed macroscopic cross-section benchmarks, actual macroscopic cross-section benchmark VVER-440, and few-group microscopic cross-section benchmark RBEC-M for fast spectrum reactors were used to verify different functionalities of VINUS. The results were shown that VINUS code maintains good computational accuracy and convergence trends. For the VVER-440 benchmarks, the deviation of keff of VINUS from reference is less than 100 pcm, and the maximum power deviation is less than 4 %. For the RBEC-M, the deviation of keff is 125 pcm, and the maximum power deviation is less than 5 %. These outcomes collectively demonstrate the solver's potential for engineering applications in future advanced reactor designs.

Combining Chemical Vapor Deposition and Spark Plasma Sintering for the Production of Tungsten Fiber‐Reinforced Tungsten (Hybrid – Wf/W)

Successful upscaling of tungsten fiber‐reinforced tungsten composites (Wf/W) on industrial level could represent an important milestone for future nuclear fusion reactors. The primary objective of these materials is to enhance the durability and operational lifespans of critical components. Developing mature manufacturing approaches remains a challenge, highlighting the need for innovative solutions. This study evaluates the feasibility of merging chemical vapor deposition (CVD) with spark plasma sintering (SPS) for producing such composites. This analysis indicates that combining CVD‐W sealed tungsten fabrics with SPS requires additional manufacturing steps or the utilization of tungsten powders for effective sintering. The process is currently only suitable for simple textile structures utilizing single filaments, mitigating one of the main advantages of CVD. Configurations such as radially braided yarns are currently less compatible to the high stress levels during SPS. A key outcome of this work is the introduction of a thin secondary CVD‐W interface into the composite design, substantially improving the stability of the yttria‐interface and effectively shielding the W‐fibers from potential matrix interactions. This innovation reduces issues such as carbon embrittlement and allows the potential integration of tungsten fibers into different matrix materials such as ceramics, broadening the potential application range of tungsten fiber‐reinforcements.

Comparison of reduced model predictions for divertor detachment onset and reattachment timescales in ASDEX Upgrade and JET experiments

Building on prior analysis of ASDEX Upgrade (AUG) experiments (Henderson et al 2023 Nucl. Fusion 63 086024), this study compares simple analytical formula predictions for divertor detachment onset and reattachment timescales in JET experiments. Detachment onset primarily scales with divertor neutral pressure, impurity concentration, power directed to the targets, machine size, and integral perpendicular power decay length. JET experiments, focusing on seeding mixtures of Ne and Ar, align with the detachment onset predictions. Radiation efficiencies among the impurities show good agreement with the model predictions, contrasting with AUG observations which suggested higher efficiency for Ar and lower efficiency for Ne. The time taken to re-ionise the neutral volume in front of the outer target in fully detached divertor conditions was measured following both abrupt increases in injected neutral beam power and, separately, cutting of the impurity gas flow. Re-ionisation of the neutrals occurs within approximately 1 s on JET, which aligns with the simple model prediction derived from AUG data. While the AUG results are not new, their comparison with the JET results enhances understanding, reinforcing confidence in using simple models to predict future reactor scenarios.

Plasma surrogate modelling using Fourier neural operators

Predicting plasma evolution within a Tokamak reactor is crucial to realizing the goal of sustainable fusion. Capabilities in forecasting the spatio-temporal evolution of plasma rapidly and accurately allow us to quickly iterate over design and control strategies on current Tokamak devices and future reactors. Modelling plasma evolution using numerical solvers is often expensive, consuming many hours on supercomputers, and hence, we need alternative inexpensive surrogate models. We demonstrate accurate predictions of plasma evolution both in simulation and experimental domains using deep learning-based surrogate modelling tools, viz., Fourier neural operators (FNO). We show that FNO has a speedup of six orders of magnitude over traditional solvers in predicting the plasma dynamics simulated from magnetohydrodynamic models, while maintaining a high accuracy (Mean Squared Error in the normalised domain ≈10−5 ). Our modified version of the FNO is capable of solving multi-variable Partial Differential Equations, and can capture the dependence among the different variables in a single model. FNOs can also predict plasma evolution on real-world experimental data observed by the cameras positioned within the MAST Tokamak, i.e. cameras looking across the central solenoid and the divertor in the Tokamak. We show that FNOs are able to accurately forecast the evolution of plasma and have the potential to be deployed for real-time monitoring. We also illustrate their capability in forecasting the plasma shape, the locations of interactions of the plasma with the central solenoid and the divertor for the full (available) duration of the plasma shot within MAST. The FNO offers a viable alternative for surrogate modelling as it is quick to train and infer, and requires fewer data points, while being able to do zero-shot super-resolution and getting high-fidelity solutions.

In-Situ synchrotron investigation of elastic and tensile properties of oxide dispersion strengthened EUROFER97 steel for advanced fusion reactors

The augmentation of mechanical properties of reduced activation ferritic martensitic steels through the introduction of creep resistant nano-oxide particles produces a class of oxide dispersion strengthened steels, which have attracted significant interest as candidates for first wall supporting structural materials in future nuclear fusion reactors. In the present work, the effect of temperature on the elastic properties and micro-mechanics of 0.3 wt% Y2O3 oxide dispersion strengthened steel EUROFER97 is investigated using synchrotron high energy X-ray diffraction in-situ tensile testing at elevated temperatures, alongside the non-oxide strengthened base steel as a point of comparison. The single crystal elastic constants of both steels are experimentally determined through analysis of the diffraction peaks corresponding to specific grain families in the polycrystalline samples investigated. The effect of temperature on the evolving dislocation density and character in both materials is interrogated, providing insight into deformation mechanisms. Finally, a constitutive flow stress model is used to evaluate the factors affecting yield strength, allowing the strengthening contribution of the oxide particles to be assessed, and correlation between the thermally driven microstructural behaviour and macroscopic mechanical response to be determined.

Laser and microwave instrumentation for ITER and future reactors

ITER diagnostics include an extensive set oflaser and microwave diagnostics to give access to a wealth of information on thecore and edge plasma and to support high performance operation of ITER. Forexample, Core and Edge Thomson scattering systems build detailed density andtemperature profiles on time scales much faster than τ E to follow transientevents; ECE and reflectometry add time resolution to follow MHDevents. Implementing these diagnostics is challenging, needing a panoply oftechnologies to keep them functioning reliably for thousands of hours despiteextreme events such as disruptions and wall conditioning cycles. Shielding,shutters and cleaning systems protect the forward elements of most opticalsystems from the build-up of deposits and damage. Still, plasma-facing mirrorsmust survive laser loads and endure erosion, deposition and in-situ RFcleaning. Calibration and monitoring systems ensure accurate and drift-freeoperation. These support systems are also not straightforward and requiredspecific R&D. Access also drives the design: to deal with the neutronand gamma sources yet allow maintenance of activated components, ITER useslarge, multi-purpose ports that couple otherwise distinct systems into modulesfor maintenance. Machine movement requires provisions to maintain alignment andcalibration, from these port plugs, shown in figure 1, to theaccessible areas 10–50 m away. A final complication comes from the difficulty ofemploying electronics near the plugs. Extensive qualification for radiationresistance is needed. This paper examines design adaptations that ITER adoptedfor its near-reactor environment, consider the lessons learnt from the ITERdesign activity specifically for laser and microwave systems and lays out somepossible evolution paths for the reactor diagnostician that must follow a moreindustrial approach.

Development of radiation sources based on CAD models for the nuclear analysis of IFMIF-DONES lithium loop

IFMIF-DONES is a neutron source designed to irradiate materials to be used in future fusion power reactors such as DEMO. The facility is based on a deuteron beam impinging onto a liquid lithium jet to generate the neutron flux. Lithium and Corrosion Products will get activated and produce Be-7 and Activated Corrosion Products (ACP). These products will be distributed along the lithium loop, both dissolved in lithium and deposited locally. This complex gamma source should be properly represented to perform radiological safety studies. A new tool, called CAD2CDGS, was created to represent these sources. This tool creates decay gamma sources based on CAD geometries and codifies them into CDGS format. Sources are specified in the CAD model, allowing a user-friendly approach. The tool is based on Open-Source tools (FreeCAD) and EUROfusion codes (cR2S European SDDR tool). In this article the CAD2CDGS tool methodology, inputs and workflow are explained. Two verification tests have been done to check the correctness. Finally, it is demonstrated the applicability of the tool to IFMIF-DONES, with the radiological zoning analysis of the rooms surrounding the Lithium Loop Cell impact by the Be-7 and ACP.

Decoupled magnetic control of spherical tokamak divertors via vacuum harmonic constraints

Power exhaust is a critical challenge for spherical tokamak reactors, making the design, optimisation and control of advanced divertor configurations crucial. These tasks are greatly simplified if the poloidal magnetic fields in the core and divertor regions can be varied independently. We present a novel method which facilitates decoupling of the core plasma equilibrium from the divertor geometry optimisation and control, using vacuum spherical harmonic (SH) constraints. This has the advantage that it avoids iterative solution of the Grad–Shafranov equation, making it easy to use, rapid and reliable. By comparing a large number of MAST-U equilibrium reconstructions against their approximations using SHs, a small number ( ∼4 ) of harmonics is found to be sufficient to closely reproduce the plasma boundary shape. We show experimentally that poloidal field changes designed to leave harmonics unaffected indeed have no effect on the core plasma shape. When augmented with divertor geometry constraints, this approach gives a powerful tool for creating advanced magnetic configurations, and its simplicity brings improvements in speed and robustness when solving coil position optimisation problems. We discuss the clear benefits to real-time feedback control, feed-forward scenario design and coilset optimisation with a view to future reactors.

Editorial: Nuclear material for current and future reactor design

Editorial: Nuclear material for current and future reactor design

How turbulence spreading improves power handling in quiescent high confinement fusion plasmas

Viable magnetic fusion devices necessitate combining good confinement with effective power flux handling. A major concern for ITER, and devices beyond, is the divertor heat load width, which sets peak boundary heat loads on the plasma-facing materials. Current estimates of the heat flux width are narrow for future reactors. Here, we demonstrate how pedestal turbulence can expand into, or entrain, the stable scrape-off-layer and so broaden the heat flux width beyond these neoclassical predictions. Employing combined theoretical, computational, and experimental approaches, we focus on quiescent high confinement discharges on the DIII-D tokamak, but the results are of broader significance. Our findings uncover common trends in the edge turbulence intensity flux, the pressure perturbation skewness, and the turbulence mixing length, which together determine the heat flux width. This research demonstrates the physics of scrape-off-layer broadening by turbulence and highlights the promise of a turbulent pedestal for successful core-edge integration in ITER and future fusion devices.

DONES EVO: Risk mitigation for the IFMIF-DONES facility

The International Fusion Materials Irradiation Facility- DEMO Oriented Neutron Source (IFMIF-DONES) is a scientific infrastructure aimed to provide an intense neutron source for the qualification of materials to be used in future fusion power reactors. Its implementation is critical for the construction of the fusion DEMOnstration Power Plant (DEMO).IFMIF-DONES is a unique facility requiring a broad set of technologies. Although most of the necessary technologies have already been validated, there are still some aspects that introduce risks in the evolution of the project.In order to mitigate these risks, a consortium of companies, with the support of research centres and the funding of the CDTI (Centre for the Development of Industrial Technology and Innovation), has launched the DONES EVO Programme, which comprises six lines of research:•Improvement of signal transmission and integrity (planning and integration risks)•Optimisation of RF conditioning processes (planning and reliability risks)•Development of a reliable beam extraction device (reliability risks)•Development of technologies for the production of medical isotopes (reliability risks)•Improvement of critical parts of the lithium purification system (safety and reliability risks)•Validation of the manufacture of critical components with special materials (reliability risk).DONES EVO will focus on developing the appropriate response to the risks identified in the IFMIF-DONES project through research and prototyping around the associated technologies.This contribution will present a discussion of the risks, the proposed response to them and the evolution of the technologies involved, following selected experiments carried out throughout the project.

Application of two-dimensional temperature response functions for reconstruction of divertor heat flux profile in commercial fusion reactors.

To keep the tritium breeding rate TBR > 1 and to meet the high heat load and neutron shielding requirements for the first wall and divertor in fusion demonstration (DEMO) reactors, the number of port plugs and other openings must be limited. To accomplish this, it is necessary to develop alternatives to the use of infrared (IR) thermography to determine the peak heat flux and the heat flux profile onto divertor targets. A divertor tile equipped with multiple temperature monitoring channels can be used to reproduce the temperature profile. To avoid the high temperatures and high neutron flux environment in a DEMO, the monitoring positions can be set well away from the irradiated surface. However, the spatial resolution of this method is lower than that provided by IR thermography. In the present work, we apply two-dimensional temperature response functions and the corresponding heat conduction model to temperature data obtained from a divertor tile surface in the large helical device to study the effects of the spatial resolution of the monitored temperature profile on the reconstructed heat flux profile. The findings provide information that will be useful in defining a method for embedding thermocouples into the divertor tiles of future DEMO reactors.

Investigation of Li–Be and B halides as blanket in future fusion molten salt reactor

Abstract FLIBE (LiF–BeF2) is both a nuclear reactor coolant and a solvent for fertile or fissile materials. FLIBE can also dissolve a variety of fissile and fertile materials such as uranium, thorium and plutonium and has the ability to serve as liquid fuel for molten salt reactors (MSRs). It’s very high thermal capacity and chemical stability are among its other valuable properties. In addition, the low atomic weights of lithium, beryllium and to a lesser extent fluorine make FLIBE an effective neutron moderator. In this study, cross-section values were determined by using various level density models (constant temperature + Fermi gas, back-shifted Fermi gas, generalized superfluid, microscopic level density models) for reactions of 19F(n, α)16N, 19F(n, p)19O, 35Cl(n, α)32P, 35Cl(n, p)35S and 79Br(n, α)76As, 79Br(n, p)79Se, 127I(n, α)124Sb, 127I(n, p)127Te using TALYS 1.95 code and these datas are compared with the values in the EXFOR database.

Recent developments in coupled experiments and simulation to understand the fluid–structure dynamics of a Pressurized Water Reactor fuel assembly

The proposed paper is dedicated to the description of an extensive research program devoted to understanding fluid–structure interaction mechanisms at all scales of a fuel assembly for a Pressurized Water Reactor (PWR). This work carried out over the last decade, closely combines modeling and experiments at the highest level, leading to a production of knowledge in the fields of instrumentation, fluid and structure measurements, numerical methods, and simulation, which goes beyond the strict framework of the PWR assembly to apply to a large number of dynamic systems of interest for current and future fission reactors. The article provides an overall synthesis of the past work, as well as a focus on some key outcomes and a description of the numerous remaining perspectives and challenges in an active ongoing research effort.

Corrosion behavior of Cr coating on ferritic/martensitic steels in liquid lead-bismuth eutectic at 600 °C and 700 °C

The surface coating technology, encompassing ceramics, refractory materials, metallic alloys containing Al or Si, and multicomponent composites, presents a viable approach to improve the corrosion resistance of ferritic/martensitic (F/M) steels with (9–12) wt.% Cr in liquid lead-bismuth eutectic (LBE) environment. Among these coating materials, chromium (Cr) coating emerges as a particularly noteworthy option. This study specifically focused on depositing a 3 μm thick Cr coating on on T91 and SIMP steels using magnetron sputtering. Subsequently, the corrosion behavior of the Cr coating was investigated in LBE at temperatures of 600 °C and 700 °C. The results revealed that, after 300 h at 600 °C, T91 and SIMP steels formed oxide scales with approximately 32.6 μm and 19.3 μm thicknesses, respectively. At 700 °C for 140 h, these oxide scales increased to about 82.4 μm and 73.1 μm for T91 and SIMP steels, respectively. However, the application of a Cr coating resulted in the formation of a dense layer of chromium oxide with a thickness of 4–5 μm. This layer effectively impeded oxygen diffusion and Fe migration leading to a significant reduction in the corrosion rate of the steel. Notably, the Cr coating maintained secure attachment to the steel even after exposure to high-temperature LBE corrosion. These findings underscore the capacity of coating to markedly enhance the corrosion resistance of T91 and SIMP steels in high-temperature LBE environments, providing robust protection against the detrimental effects of challenging conditions. Consequently, Cr coating emerges as a promising solution for future fission nuclear reactors.