J. Nucl. Eng., Volume 7, Issue 1 (March 2026) – 23 articles

Qualification is the evidence-based process through which confidence is established that a component will perform its intended function, in its intended environment, for its intended lifetime, with the required reliability. It is an owner-led activity that defines the type, quantity and quality of data required for codification and for the industrial deployment of components and their structural materials. This paper presents a structured qualification framework and applies it to a fusion machine breeder blanket structure as a representative component. It demonstrates that qualification, rather than material properties alone, dictates the use of fusion structural materials and the deployment of such materials under ASME BPV and AFCEN RCC codes. Current limitations in addressing irradiation synergy, liquid metal corrosion, and joint integrity expose gaps that these codes cannot yet prescribe. Two contrasting structural blanket material case studies: metallic-based ferritic-martensitic steel Eurofer97 and non-metallic-based silicon carbide fibre-reinforced composites (SiC_f/SiC) are used to illustrate the differing evidence requirements for each system type. Industrial scale-up considerations, including alloy specifications, manufacturing readiness, inspection reliability, and supply-chain maturity, are evaluated alongside the need for internationally harmonised datasets and design methodologies. Fusion programmes can use a phased qualification strategy in which early, time-limited operation under controlled conditions builds the evidence needed for codification and scale-up, with the required pre-operation qualification level depending on risk, component criticality and failure consequences, and with the pace of qualification ultimately setting how quickly industry can supply components for commercial fusion. Codification remains essential for commercial deployment because construction codes express codified material behaviour through allowable stresses and permitted fabrication routes, enabling designers to use advanced materials without disclosing proprietary data. In jurisdictions where ASME BPV compliance is mandatory, codification determines whether a material may enter pressure boundary service and must therefore form part of the fusion machine owner’s long-term strategy for deployment. Full article

Silicon carbide (SiC) and SiC fiber-reinforced SiC matrix composites (SiC/SiC) are receiving renewed attention for use in next-generation fusion reactors due to their ability to withstand extreme conditions, including high temperatures, neutron irradiation, and plasma interactions. General Atomics Electromagnetic Systems (GA-EMS) has demonstrated significant progress in scaling up the fabrication of SiC/SiC, achieving high mechanical uniformity and meeting dimensional requirements in components up to 12 feet in length. Key developments are discussed including scale-up of the chemical vapor infiltration (CVI) process from lab-scale to full sized parts, high-dose (100 dpa) irradiation testing, nuclear-grade ceramic joining technologies, and production-focused quality control with the collective aim to establish SiC/SiC as a reliable solution for structural and functional components in fusion systems. Beyond manufacturing, the paper addresses supply chain barriers, particularly the limited availability and high cost of nuclear-grade SiC fiber. GA-EMS is developing a novel SiC fiber production method based on a thermochemical cure step that is anticipated to reduce costs compared to traditional approaches. Additionally, advancements in engineered SiC materials, such as SiC foams and tungsten-graded SiC composites, are discussed as promising solutions for specific fusion reactor components. Full article

During the operation of nuclear power plants, nuclear fuel undergoes significant compositional changes. After several cycles of use, the fuel must be removed and stored. Currently, spent fuel is stored mainly in pools or casks, and it is necessary to demonstrate the subcriticality of these systems. Spent nuclear fuel has a complex composition, and because computational codes are typically validated using fresh-fuel experiments, subcriticality assessments are usually performed conservatively with fresh-fuel compositions. These approaches demonstrate subcriticality but are very conservative and can lead to storage system designs that are more expensive or have reduced capacity. This paper focuses on the validation of computational codes using nuclear power plant critical start-up tests (referred to as reactor criticals). These tests include spent fuel and are well documented, allowing them to serve as validation experiments. Codes validated using reactor criticals can be applied to systems containing spent fuel calculation if sufficient similarity is demonstrated. Similarity is evaluated using the SCALE TSUNAMI-IP module, which is widely used for this purpose. Based on a database containing dozens of reactor criticals and similarity analyses, we developed a methodology for demonstrating the subcriticality of spent-fuel storage systems. Full article

High-temperature gas-cooled reactors (HTGRs) employ spherical fuel elements containing thousands of tristructural-isotropic (TRISO) particles, each centered on a UO₂ fuel kernel. The internal gelation process is a key technology for preparing these UO₂ fuel kernels. However, its application is limited by the poor room-temperature stability of conventional broths and the inherent trade-off between broth stability and mechanical strength. In this work, a novel five-component broth system composed of ZrO(NO₃)₂, hexamethylenetetramine (HMTA), urea, acetylacetone (ACAC), and glucose was developed. The synergistic effects of ACAC and glucose on broth stability and gelation kinetics were systematically investigated. An optimal ACAC/glucose molar ratio of 1:1 and an ACAC/ZrO₂⁺ ratio of 1.5 yielded a zirconium broth stable for over 5 h at 25 °C. Yttrium-stabilized zirconia (YSZ) microspheres prepared under optimized conditions exhibited excellent sphericity (1.04 ± 0.01), high density (5.84 g/cm³), and a crushing strength of 8.0 kg sphere⁻¹. Importantly, this stabilization strategy was successfully extended to the uranium broth, increasing its room-temperature stability from minutes to 6 h. The results demonstrate that the synergistic stabilization strategy effectively decouples the trade-off between broth stability and mechanical strength during the internal gelation process, providing an energy-efficient, scalable route for the preparation of nuclear fuel microspheres. Full article

Using factual information on background knowledge, costs, personnel numbers, resources, and facilities from the Manhattan Project, we examine the feasibility of the development of nuclear weapons in Germany in World War II. We conclude that, while for various reasons, a uranium bomb would have been technically and economically out of reach in Germany, a few plutonium bombs might have been possible had a coordinated aggressive project been initiated no later than about mid-1940. However, the German scientists involved never established an understanding of the functioning of an atomic bomb as contained in the Frisch–Peierls memorandum and were never asked to provide such a basis on which a decision on an atomic bomb program could be based. This means that a German atomic bomb program did not fail as is often assumed; rather, it was never started. The German uranium project was never more than a scientific mission to study the possibilities offered by the newly discovered source of nuclear power. Full article

The paper is focused on broader considerations regarding the commissioning process of the 3rd Unit of nuclear power plant VVER-440 type in Mochovce (Slovakia). The new nuclear plant built in Europe is getting much more slowly than expected, declared or scheduled. Besides the nuclear power plant in Olkiluoto (Finland) and also Flamanville (France), the 3rd Mochovce Unit has finally been in full operation since 6 November 2024. Nevertheless, the more than 30 years of construction process, which was intermittently stopped and frozen, make this success story exceptional. Lessons learned from commissioning are every time specific for different countries but commissioning of nuclear power plant without presence of general designer, respecting all safety requirements and taking full responsibility for this process is unique. Still, in general, the actual Slovak experiences and knowledge could help optimise new buildings in Europe, including dreams about small modular reactor deployment or the building of other clean and sustainable use of advanced nuclear facilities in the future. Full article

Molten Salt Reactors (MSRs) offer significant advantages over conventional reactors but introduce unique modeling challenges due to their circulating liquid fuel and strong coupling among nuclear, chemical, and fluid transport processes. These challenges are amplified in depletion calculations, where MSR specific phenomena such as online refueling, off-gas removal, material redistribution, and other flow driven processes must be accurately represented. This work presents a novel multi-point depletion model that efficiently and accurately predicts isotopic evolution in MSRs by explicitly accounting for these characteristics. The mathematical formulation is derived from first principles and is computationally implemented in the open-source depletion code ONIX using neutronics solutions from open-source transport code OpenMC. The new model represents the entire primary loop by dividing it into interconnected depletion zones and tracks nuclide transport, irradiation, and removal mechanisms through a system of coupled ordinary differential equations. This approach enables parallel computation and improves performance over traditional sequential depletion methods. Validation of the developed model against Molten Salt Reactor Experiment data shows good agreement for salt-seeking isotopes and those without noble gas precursors, while discrepancies for other nuclides suggest underestimation of the corresponding removal rates. The depletion model was further applied to a reference Molten Salt Fast Reactor design to assess a new reprocessing scheme intended to expedite the achievement of equilibrium operation. Full article

The global demand for medical radionuclides is rapidly increasing, driven by the expansion of diagnostic and therapeutic radiopharmaceuticals, and by recurrent vulnerabilities in international supply chains. While high-flux reactors remain the backbone of large-scale isotope production, low- and medium-power research reactors—such as TRIGA facilities—offer valuable opportunities for decentralised, flexible, and alternative radionuclide generation. Several studies have demonstrated their capability to produce emerging therapeutic or diagnostic isotopes, including ¹¹¹Ag, ⁹⁹Mo/^99mTc, ⁶⁴Cu, ¹⁷⁷Lu, and ¹⁶¹Tb, although with yield limitations inherent to moderate neutron flux levels. In Europe, recent initiatives such as PRISMAP, SIMPLERAD, and SECURE aim to strengthen production capacity and diversify radionuclide availability. Within this framework, Italy—lacking operational power reactors—seeks alternative routes to ensure a resilient national supply. This work presents the investigation carried out within the SECURE project to assess the feasibility of an Italian production cycle for medical-grade ¹⁶¹Tb at the ENEA TRIGA RC-1 Research Reactor (Rome). The study integrates reactor-specific irradiation analyses, the development of chemical separation and target recovery processes, and a comprehensive economic evaluation within a full lifecycle perspective. The results highlight the potential and constraints of a TRIGA-based production for supporting future Italian theranostic needs. Full article

Objective neural network-based two-phase flow regime classifiers are developed for vertical, narrow, rectangular channels and a 1 inch round pipe using Kohonen Self-Organizing Maps. In the rectangular channel, the classifier uses five geometric inputs obtained from a two-sensor droplet-capable conductivity probe (DCCP-2): the bulk gas void fraction ${\alpha }_g$, ligament void fraction ${\alpha }_{\mathrm{lig}}$, normalized ligament chord length $y_{\mathrm{lig}}$, normalized large bubble chord length $y_{\ell ,\mathrm{bb}}$, and a droplet indicator. These parameters allow for the objective identification of bubbly/distorted bubbly, cap-turbulent, churn-turbulent, annular, rolling wispy, and wispy flow regimes, and yield quantitative transition boundaries in the $(j_f,j_g)$ plane for a densely populated test matrix. In the round pipe, a four-sensor droplet-capable conductivity probe (DCCP-4) provides the mean and standard deviation of droplet, bubble, and ligament chord length distributions, which are used as inputs to a Self-Organizing Map (SOM) classifier that separates rolling annular and wispy annular regimes at high void fractions. The resulting regime maps are discussed in terms of the associated phase geometries and their impact on interfacial area, drag, and entrainment, providing regime-dependent geometric inputs that can be used to improve Two-Fluid Model closures for reactor downcomers, core channels, and other nuclear thermal–hydraulic applications. Full article

The neutron survival probability (and related quantities including probabilities of extinction and initiation) is a central element of the broader stochastic theory of neutron populations and finds application in fields including reactor start-up, analysis of reactor power bursts and criticality accidents, and safeguards. In a full neutron transport formulation, the equation governing the single-neutron survival probability is a backward or adjoint-like integro-partial differential equation with the added complexity of being highly nonlinear. Analogous formulations of this equation exist in the context of many approximate theories of neutron transport, with the point kinetics formulation having received significant theoretical attention since the 1940s. This work continues this tradition by providing a novel analysis of the single-neutron survival probability equation using the tools of boundary layer theory. The analysis reveals that the “fully dynamic” solution of the single-neutron survival probability equation—and some key probability distributions derived from it—may be cast as a singular perturbation around the underlying quasi-static single-neutron probability of initiation. In this perturbation solution, the expansion parameter is the ratio of the neutron generation time to a macroscopic time scale characterizing the overall system evolution; this interpretation illuminates some of the fundamental structural aspects of neutron survival phenomena. Full article

The SCALE transport lattice code, Polaris, has been previously developed to generate few-group homogenized cross sections for whole-core nodal diffusion simulators in which the embedded self-shielding method (ESSM) is used for resonance self-shielding calculations to process cross sections. Although the ESSM capability has been very successful in light-water reactor analysis, it may require enhancements in computational efficiency; treatment of spatially dependent resonance self-shielding effects; and handling of interrelated resonance effects among fuel, cladding, and control rod materials. Therefore, this study focuses on improving computational efficiency by using a Dancoff-based Wigner–Seitz approximation combined with a material-based resonance categorization, through which a spatially dependent ESSM capability is developed to accurately estimate self-shielded cross sections inside the fuel. Benchmark results show that the new capability significantly enhances computational efficiency and accuracy for spatially dependent local zones within the fuel and through depletion. Full article

KANECS is a 3D multigroup neutronics code based on the Simplified Spherical Harmonics (SP_N) approximation and the Continuous Galerkin Finite Element Method (CGFEM). In this work, the code is extended to solve the time-dependent neutron kinetics by implementing a fully implicit backward Euler scheme for the neutron transport equation and an implicit exponential integration for delayed neutron precursors. These schemes ensure unconditional stability and minimize temporal discretization errors, making the method suitable for fast transients. The new formulation transforms each time step into a transient fixed-source problem, which is solved efficiently using the GMRES solver with ILU preconditioning. The kinetics module is validated against established benchmark problems, including TWIGL, the C5G2 MOX benchmark, and both 2D and 3D mini-core rod-ejection transients. KANECS shows close agreement with the reference solutions from well-known neutron transport codes, with consistent accuracy in normalized power evolution, spatial power distributions, and steady-state eigenvalues. The results confirm that KANECS provides a reliable and accurate framework for solving neutron kinetics problems. Full article

This study evaluates the performance and scalability of fluoride-salt-lubricated hydrodynamic journal bearings used in primary pumps for Fluoride-salt-cooled High-temperature Reactors (FHRs). Because full-scale pump prototypes have not been tested, a scaling analysis is used to relate laboratory results to commercial conditions. Bearings with different length-to-diameter (L/D) ratios were assessed over a range of shaft speeds to quantify geometric and hydrodynamic effects. High-temperature bushing test data in FLiBe at 650 °C were used as inputs to three-dimensional computational fluid dynamics (CFD) simulations in STAR-CCM+. Applied load, friction force, and power loss were computed across operating speeds. Applied load increases linearly with shaft speed due to hydrodynamic pressure buildup, while power loss increases approximately quadratically, indicating greater energy dissipation at higher speeds. The resulting correlations clarify scaling effects beyond small-scale testing and provide a basis for bearing design optimization, prototype development, and the deployment of FHR technology. This work benchmarks speed-scaling relations for fluoride-salt-lubricated hydrodynamic journal bearings within the investigated regime. Full article

The performance of conventional precipitation-strengthened copper alloys drastically degrades at temperatures exceeding 500 °C, hindering their application under extreme conditions like those in nuclear fusion reactors. Oxide dispersion–strengthened copper (ODS–Cu) alloy surmounts these constraints by incorporating thermally stable, nanoscale oxide dispersoids that simultaneously confer strengthening, microstructural stabilization, and enhanced irradiation tolerance, while preserving high thermal conductivity. This review comprehensively examines the state of the art in ODS–Cu alloy from a “processing–microstructure–property” perspective. We critically assess established and emerging fabrication routes, including internal oxidation, mechanical alloying, wet chemical synthesis, reactive spray deposition, and additive manufacturing, to evaluate their efficacy in achieving uniform dispersions of coherent/semi-coherent nano-oxides at engineering-relevant scales. The underlying strengthening mechanisms and performance trade-offs are quantitatively analyzed. The review also outlines strategies for joining and manufacturing complex components, highlights key gaps in metrology and reproducibility, and proposes a roadmap for research and standardization to accelerate industrial deployment in plasma-facing components. Full article

Targeted radionuclide therapy (TRT) is an innovative and flexible approach for treating various forms of cancer, enabling selective delivery of cytotoxic radiation to cancerous cells while minimizing damage to healthy tissue. Although TRT has proven to be highly promising for treating even advanced-stage cancers, ensuring a stable supply of the radionuclides essential for its use remains a significant challenge today. This is also true for radionuclides utilized in nuclear imaging procedures, such as Positron Emission Tomography (PET) and Single Photon Emission Computed Tomography (SPECT). Liquid-fueled molten salt reactors (MSRs) are promising for producing large quantities of highly desirable radionuclides for imaging and therapy, offering the ability to recover these radionuclides online without the need for interruptions to power production. In this study, the production of numerous beta- and alpha-emitting radionuclides for use in TRT and diagnostic procedures was studied in two small, geometrically identical, thermal spectrum MSR models—one operating with LEU fuel, and the other with a mixture of HALEU and thorium—using a novel MSR refueling and waste management concept. For therapeutic alpha emitters such as ²²⁵Ac and ²¹³Bi, the impact of thorium utilization on production yields was significant, facilitating greatly increased production. Full article

Accurate prediction of the effective thermal conductivity (ETC) of nuclear fuels is essential for optimizing fuel performance and ensuring reactor safety. However, the experimental determination of ETC is often limited by cost and complexity, while high-fidelity simulations are computationally intensive. This study presents a novel hybrid framework that integrates experimental data, validated mesoscale finite element simulations, and machine-learning (ML) models to efficiently and accurately estimate ETC for advanced uranium-based nuclear fuels. The framework was demonstrated on three fuel systems: UO₂-BeO composites, UO₂-Mo composites, and U-10Zr metallic alloys. Mesoscale simulations incorporating microstructural features and interfacial thermal resistance were validated against experimental data, producing synthetic datasets for training and testing ML algorithms. Among the three regression methods evaluated, namely Bayesian Ridge, Random Forest, and Multi-Polynomial Regression, the latter showed the highest accuracy, with prediction errors below 10% across all fuel types. The selected multi-polynomial model was subsequently used to predict ETC over extended temperature and composition ranges, offering high computational efficiency and analytical convenience. The results closely matched those from the validated simulations, confirming the robustness of the model. This integrated approach not only reduces reliance on costly experiments and long simulation times but also provides an analytical form suitable for embedding in engineering-scale fuel performance codes. The framework represents a scalable and generalizable tool for thermal property prediction in nuclear materials. Full article

Based on lessons learned from the Fukushima Daiichi Nuclear Power Plant accident caused by the 2011 off the Pacific coast Tohoku Earthquake, and the subsequent tsunamis, Japanese utilities have been upgrading their tsunami countermeasures. To understand the residual risk from beyond-design-basis events, it is important to assess seismic and tsunami risks independently while also recognizing how a plant’s risk profile changes when these events occur concurrently. This study developed a framework for a multi-hazard probabilistic risk assessment (PRA) to evaluate risks to nuclear power plants (NPPs) resulting from the superposition of earthquake and tsunami events. The framework is proposed on the assumption that the targeted plant has previously conducted single-hazard PRAs for both earthquakes and tsunamis. This study presents an approach to define the scope of risk assessment for the superposition of earthquake and tsunami events, based on the results from single-hazard PRAs for each hazard. It provides an analytical framework for superposition scenario analysis and a simplified method for multi-hazard assessment using data from single-hazard assessments. Moreover, a series of steps for the multi-hazard fragility assessment of superposed earthquake and tsunami events are proposed, clarifying the relationship between superposed impacts and the damaged parts and damage modes, accompanied by illustrative examples. Full article

Boron Neutron Capture Therapy (BNCT) is a radiotherapeutic modality which couples selective pharmacological delivery of ¹⁰B with irradiation by low-energy neutrons to achieve highly localized tumor cell killing. The BNCT therapeutic approach is undergoing rapid evolution driven primarily by advances in compact accelerator-driven neutron-source and associated facility-level nuclear infrastructure. This review examines the key physical and radiobiological principles of BNCT, with emphasis on the current engineering and operational aspects, such as neutron production and moderation, spectral shaping, beam optimization and dosimetric quantification, that critically influence clinical translation. Recent progress in ¹⁰B production and enrichment, as well as in strategies for efficient ¹⁰B delivery, is also briefly addressed. By tracing the pathway from neutron source to clinical target, this review defines the state of the art in BNCT technology, identifies the main physical and infrastructural challenges, and delineates the multidisciplinary advances needed to support widespread clinical implementation of next-generation BNCT systems. Full article

Lead–bismuth eutectic (LBE), while advantageous for advanced nuclear reactors due to its thermophysical properties, faces oxidation and corrosion challenges during operation. This study aims to optimize gas-phase oxygen control by computationally analyzing oxygen transport dynamics in an LBE loop. High-fidelity simulations were performed using ANSYS Fluent and STAR-CCM+ based on the CORRIDA loop geometry, employing detailed meshing for convergence. Steady-state analyses revealed localized oxygen enrichment near the gas–liquid interface (peaking at ∼$3\times {10}^{-6}$ wt%), decreasing to ∼$5.0-6.8\times {10}^{-8}$ wt% at the outlet. Transient simulations from an oxygen-deficient state ($1\times {10}^{-8}$ wt%) demonstrated distribution stabilization within 150 s, driven by convection-enhanced diffusion. Parametric studies identified a non-monotonic relationship between inlet velocity and oxygen uptake, with optimal performance at 0.7–0.9 m/s, while increasing temperature from 573 K to 823 K monotonically enhanced the outlet concentration by >$200\%$ due to improved diffusivity/solubility. The average mass transfer coefficient (0.6–0.7) aligned with literature values ($\pm 20\%$ deviation), validating the model’s treatment of interface thermodynamics and turbulence. These findings the advance mechanistic understanding of oxygen transport in LBE and directly inform the design of oxygenation systems and corrosion mitigation strategies for liquid metal-cooled reactors. Full article

The structural integrity of nuclear graphite is paramount for the lifespan of High-Temperature Gas-Cooled Reactors. The nuclear graphite components operate under extreme conditions involving high temperature, pressure, and intense neutron irradiation, leading to complex service behavior that is difficult to characterize only by experimental methods. This study employs the finite element method (FEM) to assess component stress and failure risk. The ManUMAT simulation method was first validated against irradiation data for Gilsocarbon graphite from an Advanced Gas-Cooled Reactor and was subsequently applied to stress–strain analysis of the nuclear graphite bricks in the HTR-PM side reflector layer. The 3D micropore structure of nuclear graphite was obtained via X-μCT and reconstructed in Avizo to establish an FEM model based on the actual pore geometry. Simulations of nuclear graphite over a 30 full-power-year service period predicted a significant contraction on the core-side and minimal thermal expansion on the out-side driven by the neutron doses. This research establishes a finite element framework that extends the ManUMAT approach by integrating a realistic pore structure model, thereby providing a foundation for quantifying the microstructural effects on macroscopic performance. Full article

Fretting wear due to flow-induced vibration (FIV) remains a primary cause of fuel failure in light water nuclear reactors. In the study of axial FIV, i.e., FIV caused by axial flows, three vibration characteristics, namely natural frequency, damping ratio, and root-mean-square (RMS) amplitude, are critical for mitigating fretting wear by avoiding resonance, maximising overdamping, and preventing large-amplitude instability motion, respectively. This paper presents a set of best practices for simulating axial FIV with a focus on predicting these parameters based on a URANS-FSI numerical framework, utilising high-Reynolds-number Unsteady Reynolds-Averaged Navier–Stokes (URANS) turbulence modelling and two-way fluid–structure interaction (FSI) coupling. This strategy enables accurate and efficient prediction of vibration parameters and offers promising scalability for full-scale nuclear fuel assembly applications. Validation is performed against a semi-empirical model to predict RMS amplitude and experimental benchmarking. The validation experiments involve two setups: vibration of a square beam with fixed and roller-supported ends in annular flow tested at Vattenfall AB, and self-excited vibration of a cantilever beam in annular flow tested at the University of Manchester. The study recommends best practices for numerical schemes, mesh strategies, and convergence criteria, tailored to improve the accuracy and efficiency for each validated parameter. Full article

The temperature fluctuations due to the mixing of two streams in a T-junction induce thermal stresses in the piping material, resulting in a pipe failure in Nuclear Power Plants. The numerical modeling of the thermal mixing in T-junctions is a challenging task in computational fluid dynamics (CFD) as it requires advanced turbulence modeling with scale-resolving capabilities for accurate prediction of the temperature fluctuations near the wall. One approach to address this challenge is using Partially Averaged Navier–Stokes modeling (PANS), which can capture the unresolved turbulent scales more accurately than traditional Reynolds-Averaged Navier–Stokes models. PANS modeling with k-ε closure gives encouraging results in the case of the Vattenfall T-junction benchmark case. In this study, PANS k-ω closure modeling is implemented for the WATLON T-junction Benchmark case. The momentum ratio (M_R) for two inlet streams is 8.14, which is a wall jet case. The time-averaged and root mean square velocity and temperature profiles are compared with the PANS k-ε and LES results and with experimental data. The velocity and temperature field results for PANS k-ω are close to the experimental data as compared to the PANS k-ε for a given filter control parameter f_k. Full article

Future fusion power plants require structural materials that can withstand extreme operating conditions, including high coolant outlet temperatures, mechanical loading, and radiation damage. Reduced-activation ferritic martensitic (RAFM) steels are a primary candidate as a structural material for such applications. This study demonstrates the successful production of a 5.5-tonne RAFM billet via electric arc furnace (EAF) technology, enabling scalable, cost-effective manufacturing. The resulting UK-RAFM alloy offers superior tensile strength and creep lifetime performance compared to Eurofer97. This is attributed to alterations in the initial forging process during manufacture. Modified thermomechanical treatments (TMTs) were subsequently applied to the UK-RAFM, which are shown to enhance the tensile strength further, particularly at 650 °C. Building on this, an Advanced RAFM (ARAFM) steel was designed to exploit the benefits of optimised chemistry to encourage metal carbonitride (MX) precipitate evolution alongside bespoke TMTs. Challenges around ensuring suitable processing windows in these steels, to avoid the over-coarsening of MX precipitates or the formation of deleterious delta-ferrite, are discussed. A subsequent 5.5-tonne ARAFM billet has since been produced using EAF facilities, with performance to be reported separately. This work highlights the synergy between alloy design, process optimisation, and industrial scalability, paving the way for a new generation of low-cost, high-volume, fusion-grade steels. Full article