Future Reactors Research Articles

Are tungsten-based nuclear fusion components truly invisible to x-ray inspection?

The ability to detect undesired volumetric defects in reactor components could affect the safety and reliability of a fusion power plant and change the expected lifetime and performance of the reactor. This is even more true for critical reactor parts like plasma-facing components (PFCs) which have to withstand challenging in-vessel conditions due to a combination of plasma bombardment, radiation, and nuclear heating. The structural integrity of these components prior to their installation in a nuclear fusion reactor needs to be assessed non-destructively. Until now, industrial x-ray radiography and tomography have not been used to non-destructively inspect fusion components due to their lack of penetration power into dense material such as tungsten which is often used to manufacture PFCs. However, aiming to revert this consolidated belief, we have demonstrated for the first time the feasibility of assessing volumetric defects non-destructively on DEMO divertor mock-up by means of MeV energy range x-ray tomography. The authors believe that the application of this technology could be easily extended for inspecting large fusion components and positively impact procedures to be followed in the qualification of fusion components for current and future nuclear reactors.

Considerations on Stellarator’s Optimization from the Perspective of the Energy Confinement Time Scaling Laws

The Stellarator is a magnetic configuration considered a realistic candidate for a future thermonuclear fusion commercial reactor. The most widely accepted scaling law of the energy confinement time for the Stellarator is the ISS04, which employs a renormalisation factor, fren, specific to each device and each level of optimisation for individual machines. The fren coefficient is believed to account for higher order effects not ascribable to variations in the 0D quantities, the only ones included in the database used to derive ISS04, the International Stellarator Confinement database. This hypothesis is put to the test with symbolic regression, which allows relaxing the assumption that the scaling laws must be in power monomial form. Specific and more general scaling laws for the different magnetic configurations have been identified and perform better than ISS04, even without relying on any renormalisation factor. The proposed new scalings typically present a coefficient of determination R2 around 0.9, which indicates that they basically exploit all the information included in the database. More importantly, the different optimisation levels are correctly reproduced and can be traced back to variations in the 0D quantities. These results indicate that fren is not indispensable to interpret the data because the different levels of optimisation leave clear signatures in the 0D quantities. Moreover, the main mechanism dominating transport, in reasonably optimised configurations, is expected to be turbulence, confirmed by a comparative analysis of the Tokamak in L mode, which shows very similar values of the energy confinement time. Not resorting to any renormalisation factor, the new scaling laws can also be extrapolated to the parameter regions of the most important reactor designs available.

Editorial: Reactor Fuels, Materials and Systems Under Extreme Environments

The safe and economical operation of any nuclear power system relies, to a great extent, on the success of the fuel and the materials of construction (Allen et al., 2010). Distinct from conventional power generation methods, a nuclear reactor is characterized by a harsh level of irradiation, very high operating temperature, and extremely corrosive environments. The behavior of fuels and materials as well as systems for fission reactors is thus complex due to the changes that the extreme environments generate on them. These kinds of changes may be seldom seen in other engineering disciplines, but this poses demanding requirements over reactor fuels, materials and systems. The design of reactor fuels, materials and systems subject to extreme environments is a major obstacle to enhancing reactor safety and economic performance of nuclear power systems. It is also essential in increasing innovation for current reactor operation improvement and future reactor design optimization. To address the challenges related to extreme reactor environments, the team of editors organizes the Research Topic "Reactor Fuels, Materials and Systems under Extreme Environments" in the journal Frontiers in Energy Research. Finally, 28 articles are collected, covering the theoretical, numerical and experimental studies on reactor fuels, materials and systems under extreme environments.

A multi-wavelength approach to increase polarimeter diagnostic performance in nuclear fusion reactors

In future tokamaks, the huge variation of the plasma parameters during a discharge (ramp up, flat top, and ramp down) may involve that a diagnostic suitable for the flat top is not suitable for transients, and vice versa. Moreover, future reactors will start the experimental campaigns in safe scenarios, where events like disruptions are not critical, and they will increase their parameters gradually. Also in this case, a diagnostic optimised for the final target scenario may fail at the beginning of the experimental campaign. Laser-based polarimetry, a plasma diagnostic used in magnetized plasma to measure quantities that are related to the electron density, the magnetic field, and the electron temperature (in the case of relativistic effects), is a typical diagnostic that must be optimised for specific scenarios, since it is affected by several issues (refraction, type-I approximation, noise sensitivity) that limit its range of applicability. The aim of this work is to present a method to solve, or at least alleviate, this type of problem by using a multi-wavelength approach. The main idea consists of measuring the polarisation effects (Faraday rotation and Cotton-Mouton phase shift) with more than one wavelength and then calculating the plasma parameters by a weighted average of the measurements, where the weights are derived from the theory of polarimetry. The analysis is performed simulating the process of measurement introducing a casual error. The outcomes demonstrate that the adoption of a multi-wavelength polarimeter system brings a more accurate measurement in a wider range. Considering that next tokamaks will be implemented with a dual-wavelength interferometer, like the dispersion interferometer-polarimeter of ITER, this proposed approach could be taken into consideration to increase the performances of polarimetry.

Benefits of using multiple Raman laser wavelengths for characterizing defects in a UO2 matrix

AbstractRaman spectroscopy is one of the most useful techniques for studying the structure of UO2 and changes due to specific defects caused by doping, changes in stoichiometry, irradiation, or heating under oxidizing conditions. In this paper, we illustrate several aspects of the application of Raman techniques to the study of UO2, including the use of wavelength‐dependent excitation (455, 532, and 785 nm) to assess the effects of doping (Nd, Th, and Zr), ion irradiation, and in situ heating and oxidation (UO2 to U3O8). Additionally, we show examples of how correlative microscopy is possible using electron backscatter diffraction combined with Raman maps of specific vibration bands or of laser‐induced luminescence generated by rare‐earth dopants in the matrix. For each of these applications, we suggest optimal excitation wavelengths that vary depending on the desired data. Blue (455 nm) excitation tends to promote oxidation even at low powers, but because Raman spectra change little with doping, irradiation‐induced changes are easier to observe. Green (532 nm) excitation is optimal for observing electron–phonon resonance effects in UO2 and offers a good compromise for high‐temperature oxidation experiments, delivering high‐quality spectra for both UO2 and U3O8. Infrared (785 nm) excitation is best for observing “defect” bands associated with doping in UO2, as changes with irradiation are small. Raman spectroscopy is particularly suited for studying the stability of UO2 towards oxidation in the presence of dopants simulating fission products, where electron–phonon resonant effects, dopant ion luminescence, and mapping can be used together to investigate structural rearrangement as a function of temperature. These techniques can offer insight into microstructural changes in UO2 fuels at higher burnups envisioned in future reactors.

A safeguards approach for molten salt reactors: Direct gamma-spectrometry of 233U

Multidimensional gamma-spectrometry is a potential solution for safeguarding future thorium-fuelled molten salt reactors (MSRs). A high-sensitivity system developed at Pacific Northwest National Laboratory (PNNL), the Advanced Radionuclide Gamma spectrometer​ (ARGO), has demonstrated direct 233U assay in freshly irradiated fuel salt samples using a dual-detector configuration and the 97.1 keV gamma-emission. These measurements were made using 1.4 × 106 Bq of 233U activity (at 1.4 × 108 fissions), at relatively high dead times (10.4%), and using short count durations of ∼800 s. Other major emissions were also evaluated (42.5 keV, 54.7 keV, 146.4 keV), in additional to limited (1.96%) Compton suppression, but provided less consistent determination of the 233U activity. Accurate 233U measurement was also demonstrated using its four-coincidence gamma-emissions (42.5 keV, 54.7 keV; 42.5 keV, 278.1 keV; 146.4 keV, 174.2 keV; 146.4 keV, 219 keV). Further improvements in coincidence measurements may be achievable with improvements in the detector configuration and nuclear data. Whilst this demonstration utilized the ARGO laboratory system, a similar dual-detector configuration could be utilized for online safeguards measurements.

Edge tokamak transport in regimes with high collisionality

AbstractIn the strongly detached regimes, especially in future tokamak reactors, plasma in the divertor and in some special regions like vicinity of highly radiating X‐point is strongly collisional. Ion gyrofrequency becomes less than Coulomb collision frequency and might be of the order of ion‐neutral collision frequency. For such parameters, parallel and perpendicular ion diffusion and mobility caused by ion‐neutral collision becomes important and can to large extent determine density profiles and currents distribution in the divertor. In the present article, plasma profiles in the vicinity of divertor plates were analysed. It is demonstrated that density profiles are determined by combination of parallel and perpendicular ambipolar diffusion. Results of analytic calculations are compared with Particle‐In‐Cell Monte‐Carlo Collisions (PIC MCC) simulations of magnetized plasma and collisional transport of magnetized ions from quasi‐neutral region to the material surface. Analytic expressions are compared with the results of particle in cell simulations done for different angles between magnetic field and material surface and different plasma parameters. Impact of velocity of neutral particles on ion fluxes and currents is also investigated. Consequences for divertor plasma typical for ITER and other future tokamak‐reactors are discussed.

On the mitigation of cancellation errors in hybrid fluid‐kinetic methods for solving kinetic equations describing neutrals in the plasma edge

AbstractOne of the key aspects when simulating the plasma edge is the treatment of the neutral particles. Neutral particles can be described by a kinetic equation, which is difficult to solve due to the high dimensionality of the phase space and the high collisionality between the neutral particles and the plasma background, especially in the (partially) detached regimes anticipated for future reactors. We consider a recently proposed hybrid fluid‐kinetic method that is based on a micro–macro decomposition of the kinetic equation and mitigates the problems related to both the high dimensionality and the high collisionality. However, this method is prone to cancellation errors due to the introduction of particles with both positive and negative weights. In each iteration of the method, the particles undergo a weight projection step, which we identify as the main source of cancellation errors. We use techniques from numerical linear algebra for solving ill‐conditioned linear systems to improve the numerical stability of the weight projection step and reduce the cancellation errors. The reduction of cancellation errors allows for further speedup in the hybrid fluid‐kinetic method.

Experimental and Numerical Investigation of the Dehydration of Ca(OH)2 at Low Steam Pressures

The CaO/Ca(OH)2 system can be the basis for cost-efficient long-term energy storage, as the chemically stored energy is not affected by heat losses, and the raw material is cheap and abundantly available. While the hydration (thermal discharge) has already been addressed by several studies, for the dehydration (thermal charge) at low partial steam pressures, there is a lack of numerical studies validated at different conditions and operation modes. However, the operation at low steam pressures is important, as it decreases the dehydration temperature, which can enable the use of waste heat. Even if higher charging temperatures are available, for example by incorporating electrical energy, the reaction rate can be increased by lowering the steam pressure. At low pressures and temperatures, the limiting steps in a reactor might change compared to previous studies. In particular, the reaction kinetics might become limiting due to a decreased reaction rate at lower temperatures, or the reduced steam density at low pressures could result in high velocities, causing a gas transport limitation. Therefore, we conducted new measurements with a thermogravimetric analyzer only for the specific steam partial pressure range between 0.8 and 5.5 kPa. Based on these measurements, we derived a new mathematical fit for the reaction rate for the temperature range between 375 and 440 °C. Additionally, we performed experiments in an indirectly heated fixed bed reactor with two different operation modes in a pressure range between 2.8 and 4.8 kPa and set up a numerical model. The numerical results show that the model appropriately describes the reactor behavior and is validated within the measurement uncertainty. Moreover, our study revealed an important impact of the operation condition itself: the permeability of the reactive bulk is significantly increased if the dehydration is initiated by a rapid pressure reduction compared to an isobaric dehydration by a temperature increase. We conclude that the pressure reduction leads to structural changes in the bulk, such as channeling, which enhances the gas transport. This finding could reduce the complexity of future reactor designs. Finally, the presented model can assist the design of thermochemical reactors in the validated pressure and temperature range.

DEMO – The main achievements of the Pre – Concept phase of the safety and environmental work package and the development of the GSSR

The EUROfusion Safety and Environmental Work Package (WPSAE) has the scope to progress the safety studies for the future EU DEMO reactor. A Generic Site Safety Report (GSSR) has been prepared to include all the steps necessary to cover the safety issues of the nuclear fusion plant, from the definition of principles and requirements, through the detection of the source terms at risk, selecting the postulated initiating events, analyzing the accidents, quantifying the doses to the population and investigating waste production and its management. Eleven GSSR volumes collate the studies performed. The final goal is to prepare safety documentation, as complete as possible, to initiate a preliminary safety report when the plant site is selected.In parallel, the safety studies are supporting the maturation of the design of DEMO, providing feedback on the technical choices for the machine, the selection of materials, the use of space, the equipment necessary to correctly manage the safety risks through continuous collaboration with the design teams of other DEMO Work Packages.The GSSR focused on safety and environmental issues, it did not encompass security issues that could impact the design and hence the safety performance of DEMO. Consideration of the interactions between safety and security will be needed in the development of the DEMO detailed design and its associate Preliminary Safety Analysis Report (PSAR).The main achievements of the Pre-Concept Phase are presented in this paper, by summarizing the contents of GSSR volumes. Completion of the work is foreseen during the future Concept Design Phase (2021–2027).

About the enthalpy of formation of the UZr2 compound

The U-Zr phase diagram has been the subject of many recent studies, particularly for the development of metallic fuels for future fast reactors. The so-called δ-UZr2 phase is part of this renewed interest as it is likely to be formed during irradiation of U-Zr10wt% fuel. Although this phase has been the subject of several structural studies, there is a single experimental determination of its enthalpy of formation by dissolution calorimetry. Most of the Calphad and DFT models of this phase rely on this value to support their validation. In this paper, we propose a reinterpretation of this measurement which tends to show that the original interpretation of the authors was probably wrong. Our reinterpretation leads to a new endothermic estimate of the standard enthalpy of formation of δ-UZr2 which is discussed in relation to the other data of the U-Zr system. The need for new heat capacity and formation enthalpy measurements is highlighted. The present work does not invalidate the most recent CALPHAD descriptions.

The effect of radiolytic oxidation on the elastic properties of nuclear graphite measured by electronic speckle pattern interferometry

• Three-point bend measurement data characterised by ESPI and inverse modelling. • Static Young's modulus , Poisson's ratio and strength determined. • Measurements performed on 218 irradiated graphite beams taken from UK AGR stations. • Graphite samples (6 × 6 × 19 mm) cover dose range to 21.1 DPA and weight loss to 46%. • Modulus and strength correlate well with radiolytic weight loss. The Young's modulus, Poisson's ratio and flexural strength of 218 beams of radiolytically oxidised nuclear graphite have been determined by combining electronic speckle pattern interferometry, ESPI, and inverse finite element modelling techniques. The graphite was extracted from three operational advanced gas cooled reactor stations where it had been subjected to fast neutron irradiation and radiolytic oxidation resulting in mass loss. The 6 × 6 × 19 mm beam samples had densities ranging from 1063 – 1840 kg.m −3 , equivalent to weight losses up to 46%, with cumulative neutron doses of up to 161 × 10 20 n.cm −2 equivalent Dido nickel dose, or 21.1 displacements per atom, significantly extending the datasets available in the open literature to higher levels of both dose and oxidation. In situ ESPI measurements were collected during three-point bend tests, creating full field displacement maps of the evolving deformation to which simulations were matched by chi-squared minimisation of the elastic properties. Potential loading abnormalities were accounted for in the fitting process to ensure the simulations were representative, with total uncertainties in the determined Young's modulus and Poisson's ratio values of ± 4.5% and ± 0.08, respectively. Consistent trends were observed in the behaviour of the material from the different stations, with strong positive dependences on density for both the flexural strength and Young's modulus. The strength and Young's modulus remained well correlated even to high levels of oxidation mass loss. Conversely, the Poisson's ratio was practically invariant with oxidation with mean values of 0.21 ± 0.03, 0.24 ± 0.05 and 0.25 ± 0.05 for the three stations. The combined ESPI-finite element measurement approach is demonstrated as robust for irradiated small specimens, making it suitable for monitoring graphite in future reactors. The collected data support the safety assessments of the current reactors, enabling predictions of the responses to progressively higher mean core mass loss.

Transient thermal shock and helium ion irradiation damage behaviors of ODS-W/CuCrZr joints

The transient thermal shock and helium ion irradiation damage behaviors of oxide dispersion-strengthened tungsten (ODS-W)/CuCrZr alloy joints prepared by direct diffusion bonding and by improved diffusion bonding via surface nano-activation and Cu electroplating were studied. The surface morphologies and interface structures of the ODS-W/CuCrZr joints after thermal shock and He+ irradiation were investigated by atomic force microscopy, scanning electron microscopy, focused ion beam/scanning electron microscopy, and high-resolution transmission electron microscopy. The joint of ODS-W and CuCrZr prepared by the improved diffusion bonding method exhibited superior resistance to crack formation. No obvious plastic deformation was found on the ODS-W surfaces of the ODS-W/CuCrZr joints obtained by either bonding method after transient thermal shock (power density = 0.6 GW/m2). Interfacial debonding was observed to occur at the interface of W and Cu in the direct diffusion-bonded ODS-W/CuCrZr joint where Y2O3 particles exist, but did not occur in the joint prepared using the improved method. The nanostructures formed at the interface of the joint via the improved diffusion bonding method therefore elevated the radiation resistance of the joint. These results should be of relevance for the development of plasma-facing components in future nuclear fusion reactors.

Status of the WEST travelling wave array antenna design and results from the high power mock-up

This paper presents the current status of the WEST TWA antenna, its mock-up and a possible extrapolation to DEMO. The updated WEST TWA design has a reduced antenna length and features feeding and mechanical support from a single vessel port. A mock-up of the WEST TWA antenna was designed in 2019, manufactured during 2020 and installed in the TITAN test facility at the beginning of 2021. The results of the mock-up at low and high power, its diagnostic system and the prospects are explained. Extensions towards a TWA antenna for WEST and a possible TWA system for the future DEMO tokamak reactor are briefly discussed.

Non-uniform He bubble formation in W/W2C composite: Experimental and ab-initio study

Tungsten-tungsten carbide (W/W2C) composites are considered as possible structural materials for future nuclear fusion reactors. Here, we report on the effect of helium (He) implantation on microstructure evolution of polycrystalline W/W2C composite consolidated by field-assisted sintering technique (FAST), homogenously implanted at room temperature with 1 MeV 4He+ ions at the fluence of 8 × 1016 ions cm−2 and annealed at 1873 K for 20 minutes. Samples were analysed by scanning and transmission electron microscopy to study the presence and size of He bubbles. Monomodal He bubbles in W (30-80 nm) are limited to point defects and grain boundaries, with a considerable void denuded zone (150 nm). Bubbles do not form in W2C, but at the W|W2C interface and are considerably larger (200-400 nm). The experimental observations on He behaviour and migration in W and W2C were assessed by density functional theory (DFT) calculations, suggesting He migration and accumulation in the composite are determined by the effective He-He binding in clusters, which will give rise to decohesion. In the presence of He clusters, the decohesion of bulk W into free surfaces is energetically highly favourable but not sufficient in the W2C; hence bubbles are only observed in W grains and interfaces and not within bulk W2C.

Hygroscopic additive-modified magnesium sulfate thermochemical material construction and heat transfer numerical simulation for low temperature energy storage.

In this research, the core objective is to explore the effect of super-absorbent polymer material (poly(sodium acrylate)) on the heat storage performance of magnesium sulfate and to investigate the heat transfer behavior of 13X-zeolite, nano-aluminum oxide (nano-Al2O3) and poly(sodium acrylate) modified magnesium sulfate in a reactor. Finally it provides support for future material and reactor design. All characterizations and performance tests were done in the laboratory and a numerical simulation method was used to investigate the heat transfer behavior of the reactor. Through hydrothermal treatment, bulk MgSO4·6H2O was changed into nanoparticles (200–500 nm) when composited with poly(sodium acrylate), 13X-zeolite and nano-Al2O3. Among these materials, MgSO4·6H2O shows the highest activation energy (36.8 kJ mol−1) and the lowest energy density (325 kJ kg−1). The activation energy and heat storage energy density of nano-Al2O3 modified composite material MA-1 are 28.5 kJ mol−1 and 1305 kJ kg−1, respectively. Poly(sodium acrylate) modified composite material, MPSA-3, shows good heat storage energy density (1100 kJ kg−1) and the lowest activation energy (22.3 kJ mol−1) due its high water-absorbing rate and dispersing effect. 13X-zeolite modified composite material MZ-2 shows lower activation energy (32.4 kJ mol−1) and the highest heat storage density (1411 kJ kg−1), which is 4.3 times higher than that of pure magnesium sulfate hexahydrate. According to the heat transfer numerical simulation, hygroscopic additives could prominently change the temperature distribution in the reactor and efficiently release heat to the thermal load side. The experimental and numerical simulation temperatures are similar. This indicates that the result of the numerical simulation is very close to the actual heat transfer behavior. This reactor could output heat at around 50 °C and absorb heat in the range of 100–200 °C. All these results further prove the strategy that thermochemical nanomaterial synthesis technology combined with material-reactor heat transfer numerical simulation is feasible for future material and reactor design.

Void swelling of conventional and composition engineered HT9 alloys after high-dose self-ion irradiation

Ferritic/martensitic (F/M) steels are being considered as potential structural materials for next generation nuclear reactors, and variants of the alloy HT9 are some of the most promising candidates. In this study, two conventional and two composition engineered HT9 alloys were irradiated using 3.5 MeV Fe2+ up to 600 peak displacement-per-atom (dpa) at 450 ℃. Void swelling and microstructure evolution were characterized for each alloy and compared. The two conventional HT9 alloys (INL and ACO3) showed similar void swelling behavior due to their similar elemental composition and processing conditions. The INL HT9 exhibited a maximum of 2.4% swelling and the ACO3 HT9 showed a maximum of 2.8% swelling at 342 and 393 average local dpa, respectively. On the other hand, the two-composition engineered HT9 alloys with varying N contents (10 ppm for low N and 440 ppm for high N) showed disparate swelling behavior. The low N HT9 exhibited a maximum of 4.6% swelling, while the high N HT9 showed a maximum of 0.7% swelling at 342 average local dpa. Changes in the N content also affected Ni/Si rich G-phase formation. The low N HT9 showed a larger size and lower density of G-phase precipitates compared with the high N HT9 after 600 peak dpa irradiation. This study compares the void swelling behavior of the ion irradiated four current HT9 alloys to extremely high doses, with the void swelling data from neutron irradiated HT9 alloys. The comparison lends critical insights into how well these current alloys can withstand high neutron fluxes in future reactors, especially since the low N and high N HT9 alloys have never been exposed to such high doses before.

Neutron irradiation-enhanced grain growth in tungsten and tungsten alloys

To understand the microstructural stability of candidate plasma-facing materials under fusion-relevant environments, neutron irradiation of W and W-3%Re alloys with and without K and La dopants was performed in the mixed-spectrum High Flux Isotope Reactor at nominal temperatures of ~850 °C and ~1100 °C to calculated doses between 0.42 and 0.47 dpa. To the best of our knowledge, this study presents the first experimental evidence of radiation-enhanced recrystallization in W and undoped W–Re alloys at ~850 °C, conditions where thermal annealing does not cause any grain growth in a similar timescale. Potassium- or lanthanum-doped tungsten alloys showed more resistance to radiation-enhanced grain growth. We explain the acceleration of grain growth by analyzing the self-diffusion constant under atomic displacement environments. The microstructural observations of the studied W variants suggest that La doping is more effective than K doping for mitigating recrystallization. This study also found that radiation-enhanced recrystallization is an important consideration when designing and applying W to plasma-facing components in future nuclear fusion reactors.

Spatially-variant isotope production burnup modeling in a CANDU-6 reactor for nuclear treaty monitoring

A CANDU-6 nuclear reactor is modeled as a 3D quarter-core model using Monte-Carlo and burnup techniques to quantify the spatially-variant isotope production for Nuclear Treaty Monitoring applications. Fuel shuffling and refueling techniques were applied in conjunction with the Monte-Carlo and burnup techniques to simulate the realistic operations of a CANDU-6 reactor. The methods developed serve as a prototype for spatially-variant spent fuel analysis for reactors of interest for non-proliferation. A comprehensive, spatially-dependent database of isotopic concentration distributions in spent fuel from a CANDU-6 reactor was developed for the top 237 isotopes expected from CANDU-6 spent fuel. Isotopic ratios of actinides, a common technique used to assess material origin, were determined to have complex distributions dependent on their position within the reactor. Confidence intervals of 67%, 95%, and 99% for common isotopic ratios are presented. For example, the 3-σ confidence interval for 239Pu/240Pu is 1.049-2.599, a significant variance. The database and the methods developed in this work set the framework for future reactor modeling for nuclear treaty monitoring.

Technological Processes for Steel Applications in Nuclear Fusion

Plasma facing components for energy conversion in future nuclear fusion reactors require a broad variety of different fabrication processes. We present, along a series of studies, the general effects and the mutual impact of these processes on the properties of the EUROFER97 steel. We also consider robust fabrication routes, which fit the demands for industrial environments. This includes heat treatment, fusion welding, machining, and solid-state bonding. Introducing and following a new design strategy, we apply the results to the fabrication of a first-wall mock-up, using the same production steps and processes as for real components. Finally, we perform high heat flux tests in the Helium Loop Karlsruhe, applying a few hundred short pulses, in which the maximum operating temperature of 550 °C for EUROFER97 is finally exceeded by 100 K. Microstructure analyses do not reveal critical defects or recognizable damage. A distinct ferrite zone at the EUROFER/ODS steel interface is detected. The main conclusions are that future breeding blankets can be successfully fabricated by available industrial processes. The use of ODS steel could make a decisive difference in the performance of breeding blankets, and the first wall should be completely fabricated from ODS steel or plated by an ODS carbon steel.