The Spherical Tokamak for Energy Production (STEP) programme is an ambitious but challenging endeavour to design and deliver a prototype fusion power plant. It is a rapid, fast-moving programme, designing a first of a kind device in a Volatile, Uncertain, Complex and Ambiguous (VUCA) environment, and digital tools play a pivotal role in managing and navigating this space. Digital helps manage the complexity and sheer volume of information. Advanced modelling and simulation techniques provide a platform for designers to explore various scenarios and iteratively refine designs, providing insights into the intricate interplay of requirements, constraints and design factors across physics, technology and engineering domains and aiding informed decision-making amidst uncertainties. It also provides a means of building confidence in the new scientific, technological and engineering solutions, given that a full-scale-integrated precursor test is not feasible, almost by definition. The digital strategy for STEP is built around a vision of a digital twin of the whole plant. This will evolve from the current digital shadow formed by system architecting codes and complex workflows and will be underpinned by developing capabilities in plasma, materials and engineering simulation, data management, advanced control, industrial cybersecurity, regulation, digital technologies and related digital disciplines. These capabilities will help address the key challenges of managing the complexity and quantity of information, improving the reliability and robustness of the current digital shadow and developing an understanding of its validity and performance.This article is part of the theme issue 'Delivering Fusion Energy - The Spherical Tokamak for Energy Production (STEP)'.

The European DEMOnstration power plant (EU DEMO) project is currently leading the research endeavours within the fusion field with the goal of developing a next generation fusion reactor. Given the challenging nature of the application, it is essential to establish methodologies that can provide convenient and reliable characterisation of the chosen materials within DEMO. In this paper, the recently developed Time-of-Flight Neutron Bragg Edge Imaging (TOF-NBEI) was used on Tungsten (W)/ Copper (Cu) dissimilar joints sample mock-ups of the cooling system design used in the critical divertor component with the goal of mapping the residual stresses across the sample. Residual strain mapping was performed on the W phase with considerable tensile residual strains identified close to the W-Cu interface. The large-grain microstructure of the Cu phase was analysed using the energy-resolved neutron radiographs. The results will be used as a basis for future TOF-NBEI experiments of tungsten monoblocks related to DEMO.

The Spherical Tokamak for Energy Production (STEP) program aims to achieve net electricity production from fusion through the STEP prototype power plant (SPP) by 2040. A major challenge in attaining fusion power is plasma exhaust, which the STEP program addresses through a double-null (DN) divertor configuration, showing promise in reducing radiation requirements. This article presents a case study applying an integrated methodology for the initial design of the STEP first wall in the SPR-45 conceptual design. The methodology involves managing charged particles and radiation heat loads and has been detailed in a separate paper. The models and procedures for predicting wall heat loads, designing the 2-D poloidal contour, and evaluating radiation loading have been seamlessly integrated into BLUEMIRA, a versatile reactor design framework, that supports rapid iterations and sensitivity studies. The integrated methodology enables rapid iterations and sensitivity studies, resulting in an optimized first wall geometry profile. Its flexibility facilitates cross-functional collaboration, allowing the sharing of first wall profiles for additional studies and geometry enhancements by other teams, such as vertical displacement or limiter profiles. These modified profiles can then be reinserted into the methodology for updated assessments of charged particle heat flux and radiation heat loads. This case study demonstrates the integrated methodology’s effectiveness in addressing plasma exhaust complexities and heat management challenges. The seamless collaboration among cross-functional teams exemplifies the methodology’s versatility, making it a valuable tool in the development of fusion power plant designs.

The Combined Heating and Magnetic Research Apparatus (CHIMERA) fusion technology test facility is under construction. The facility will be uniquely capable of semi-integral testing of fusion materials and component modules up to the size of the ITER test blanket module box, under combined conditions of in-vacuum high heat flux, static and pulsed magnetic fields, and high-temperature/high-pressure water cooling. This paper reports the high-level capabilities of the CHIMERA baselined design and the planned program of testing and describes the proposed strategy for use of simulations for virtual testing, qualification, and in-situ monitoring.The first step in testing of a component mock-up is to take data from as-built geometry and other measurements and transmit them to an integrated computational model that can closely mimic the physical asset and form a digital replica. Not only can this digital replica be queried in advance of physical testing in the facility, allowing optimization of the test program, but combined with subsequent test data, it also can deliver much greater insight into experimental results than can be obtained using test data alone. The digital replica is used as the basis for a digital twin, which is live coupled to the running experiment, and is under development as a proposed key facet of fusion reactor surveillance in-service. Physical mock-ups for testing can be subjected to in-vacuum heat flux up to 0.5 MW/m2 over the entire surface while within a strong horizontal magnetic field. The central field can be up to 4 T with a peak in the test region of 5 T. The same component mock-ups can also be subjected to repeated magnetic field pulses with ramp rate 12 T/s, which can simulate loading conditions of a plasma disruption. Facility upgrades are underway to include a liquid metal circulation loop to allow the study of magnetohydrodynamics effects and to add a high-heat-flux system using a very high-power continuous-wave laser to achieve divertor-relevant heat fluxes of 20 MW/m2 over the area of a small-scale mock-up. Four examples are given to illustrate the physical testing program that is currently foreseen.

The current work uses a novel fundamental heat transfer experiment to understand the morphological and thermal performance effects of nanoparticle deposition processes on heating surfaces under high heat fluxes. This is a unique fundamental study of nanosuspension induced nanoparticle coated boiling surfaces under realistic fusion relevant conditions. Al2O3H2O nanosuspensions have been used under forced convection and boiling. The experiments were performed on a test bed able to simulate realistic fusion reactor heat flux. Nanosuspensions are found to deteriorate the cooling performance due to the formation of a complex self-assembled porous nanoparticle layer on the heating surfaces. This negative effect on thermal performance is irrespective of operation in nanoparticulate latent or pure coolants modes. For heat transfer in nanosuspensions, the increase of nanoparticle concentration reduced the observed negative thermal performance effects. Improvement of thermal performance beyond the break-even point, as witnessed for some conditions in the current work, could be potentially achieved by increasing the concentration of nanoparticles in the coolant. When the nanosuspension is removed and the heat transfer surfaces with the nanolayer deposit are washed and operated with pure liquids, it was discovered that the deposited layers survived and still affected (negatively) their heat transfer performance. The deposited layers are porous and are expected to extend the critical heat flux of surfaces in relevant industrial processes. The deposition process and the final thermal properties could be affected by several controlled parameters providing design opportunities for new or retrofitted applications that were otherwise inaccessible or unfeasible.

The CHIMERA fusion technology facility will enable testing of large in-vessel component modules under reactor-like conditions of combined in-vacuum thermal power density and magnetic field. With an integral large superconducting magnet and a PWR-like water loop, CHIMERA is also ideally placed for experiments on liquid metal breeding blanket prototypes. Facility construction is underway at the UKAEA site in South Yorkshire. The superconducting magnet is fully wound and terminated, and the bespoke pulsed magnet power supply has been delivered. Even before construction and commissioning is complete, a parallel and supporting research and development programme is in progress and is reported here. A bespoke infrared heating system has been developed, capable of applying 0.5 MW/m2 to component surfaces up to the size of the ITER test blanket module first wall. The modules of this heater are highly specialised and designed to endure the high magnetic forces from the CHIMERA static and pulsed magnets. CHIMERA will feature a range of diagnostics, including load cells to measure static and pulsed magnetic forces, and induced current sensors, all of which have been tested to confirm acceptable operation in the pulsed magnetic field. Manufacture of the test mock-up to be used for facility commissioning is underway. Testing will be enhanced with simulation 'digital twin' capability, and a develop-ment project is now producing first virtual test results, informing commissioning of the CHIMERA device.

•Systems simulation used for fast-running multiphysics models of fusion components.•CHIMERA provides a platform to validate different simulation approaches.•Thermal, fluid and EM loads of a CHIMERA sample have been coupled together.•In future models’ additional physics will be added, and ran probabilistically.Critical to the successful design of fusion reactor components is the development of coupled simulation models, capable of conducting efficient whole system design optimisation and virtual component qualification, under conditions which cannot be readily tested. Additionally, successful lifetime monitoring and predictive maintenance of fusion components through diagnostic measurements will be limited due to restricted accessibility and operation in a harsh environment. There is therefore a need for a component digital twin, which combines data from the physical instrumentation with simulation to provide abundant virtual diagnostics in real-time.

In a power plant scale fusion reactor, a huge amount of thermal power produced by the fusion reaction and external heating must be exhausted through the narrow area of the divertor targets. The targets must withstand the intense bombardment of the diverted particles where high heat fluxes are generated and erosion takes place on the surface. A considerable amount of volumetric nuclear heating power must also be exhausted. To cope with such an unprecedented power exhaust challenge, a highly efficient cooling capacity is required. Furthermore, the divertor must fulfill other critical functions such as nuclear shielding and channeling (and compression) of exhaust gas for pumping. Assuring the structural integrity of the neutron-irradiated (thus embrittled) components is a crucial prerequisite for a reliable operation over the lifetime. Safety, maintainability, availability, waste and costs are another points of consideration.In late 2020, the Pre-Conceptual Design activities to develop the divertor of the European demonstration fusion reactor were officially concluded. On this occasion, the baseline design and the key technology options were identified and verified by the project team (EUROfusion Work Package Divertor) based on seven years of R&D efforts and endorsed by Gate Review Panel.In this paper, an overview of the load specifications, brief descriptions of the design and the highlights of the technology R&D work are presented together with the further work still needed.

In the conceptual design of EU-DEMO, damage to plasma-facing components under disruption events is planned to be mitigated by specific sacrificial limiter components. A new limiter concept has been proposed using lattice structures fabricated with tungsten powder by additive manufacturing techniques. The major potential benefits of using a lattice structure for limiters are the possibility to customise the thermal conductivity and structural compliance of these components to manage temperatures and stress within material limits and lower the sensitivity to crack propagation. This paper presents the results of the first investigations into the production, characterisation, and high heat flux testing of these lattices to assess their suitability for DEMO limiters. First stage prototypes have been manufactured from tungsten and tungsten tantalum mixed powder with two distinct laser power bed fusion processes, namely pulsed laser and continuous laser with heated bed. The samples are characterised in terms of mass, volume, density, extent of microcracks and voids, level of un-melted or partially melted particulates, texture and grain size, as well as tantalum segregation when applicable. High transient (0.25 ms) heat load testing, with hydrogen plasma of energy density up to ∼3 MJ m−2 was carried out at Kharkov Institute of Physics and Technology on the quasi-stationary plasma accellerator Kh-50. These tests have shown that the energy absorbed by latticed targets preheated at 500 °C is close to that absorbed by solid tungsten, suggesting that they may be used for limiter applications with the added advantage of adjustment of the heat transfer and stiffness performance by geometry design or material properties.

In the conceptual design of EU-DEMO, damage to plasma-facing components under disruption events is planned to be mitigated by specific sacrificial limiter components. A new limiter concept has been proposed using lattice structures fabricated with tungsten powder by additive manufacturing techniques. The major potential benefits of using a lattice structure for limiters are the possibility to customise the thermal conductivity and structural compliance of these components to manage temperatures and stress within material limits and lower the sensitivity to crack propagation. This paper presents the results of the first investigations into the production, characterisation, and high heat flux testing of these lattices to assess their suitability for DEMO limiters. First stage prototypes have been manufactured from tungsten and tungsten tantalum mixed powder with two distinct laser power bed fusion processes, namely pulsed laser and continuous laser with heated bed. The samples are characterised in terms of mass, volume, density, extent of microcracks and voids, level of un-melted or partially melted particulates, texture and grain size, as well as tantalum segregation when applicable. High transient (0.25 ms) heat load testing, with hydrogen plasma of energy density up to ∼ 3 MJ m − 2 was carried out at Kharkov Institute of Physics and Technology on the quasi-stationary plasma accellerator Kh-50. These tests have shown that the energy absorbed by latticed targets preheated at 500 ◦ C is close to that absorbed by solid tungsten, suggesting that they may be used for limiter applications with the added advantage of adjustment of the heat transfer and stiffness performance by geometry design or material properties.

Divertor target is one of the most critical in-vessel components in a fusion power plant being in charge of particle and power exhaust. The targets are exposed to severe thermal loads produced by steady bombardment of impinging plasma flux. Since 2014, integrated R&D efforts have been continued aiming at developing a design concept and high-heat-flux (HHF) technologies for divertor targets of the European DEMO reactor. Recently, the second round (2017–2019) of the R&D program has been concluded. As in the first R&D round, five water-cooled target design concepts were further developed and evaluated. Fabrication technologies were improved reaching a consolidated production quality. Extensive HHF tests were conducted using small-scale mock-ups for extended loading regimes (heat flux: 20–32MW/m²). Comparative studies were performed to investigate effects of copper interlayer thickness (0.1–1 mm) and different tungsten armor materials. In the present paper, the final results of the second round HHF testing campaign are reported. The HHF performance of each design variant is discussed based on in-situ diagnostic data (infrared thermography), ultrasonic inspection images and postmortem metallographic micrographs. All monoblock-type design concepts passed the specified qualification criterion (≥500 pulses at 20 MW/m², coolant: 130 °C) without any failure or armor cracking. Moreover, two of them (ITER-like and composite pipe) remained fully intact even under 25 MW/m² (100 pulses) and 32 MW/m² (5 pulses).

In the framework of the activity of target development of the European DEMO divertor, ENEA carried out an extensive ultrasonic testing campaign on more than 66 tungsten monoblock mock-ups. The EU-DEMO target development activity concerns primarily the comparison between the reference solution for the divertor targets (ITER-like) and two other concepts (Thermal Break Interlayer, Functionally Graded Interlayer), in which the interlayer between monoblocks and tube has been modified with the aim of increasing the component performance. In the same activity, many other aspects were also considered such as analyzing the influence of the tungsten monoblock supplier, interlayer thickness and interlayer manufacturing process (casting or diffusion bonding). Mock-ups were provided by ENEA (ITER-Like), CEA (Thin Graded Interlayer) and CCFE (Thermal Break Interlayer) and tested to thermal fatigue by high heat flux in the GLADIS facility in IPP - Garching center under DEMO relevant conditions. In this work the comparison between UT results obtained before and after the high heat flux test for the more significant mock-ups for phase 2 are carried out. After testing, mock-ups were cut in order to conduct post-mortem analyzes. Substantial agreement is found with UT results.

In the framework of the DEMO divertor project an extensive R&D programme has been carried out to develop advanced design concepts for hot water cooled divertor targets. These plasma-facing components are made of W monoblocks as plasma facing material bonded on CuCrZr cooling tubes. They are designed to allow a reliable DEMO operation of 2 h long pulses up to a 20 MW m(-2) maximum heat flux. Considering the current experiences of the qualification of ITER divertor components, the operation at significantly higher neutron fluence, the longer required lifetime, and the coolant temperature of 150 degrees C are challenges for design and manufacturing. The aim of the performed HHF tests is to support the identification of the most promising design concepts for further development. Eight different types of monoblock mock-ups were designed and manufactured by the involved research groups. These mock-ups were assessed by HHF examination in the test facility GLADIS at IPP Garching. The HHF tests were focused on 20 MW m(-2) loading with 500-1000 cycles using hot water cooling to simulate slow thermal transients in DEMO. In screening tests heat fluxes up to 32 MW m(-2) were applied to study the performance of the components against single thermal events exceeding the expected heat flux limits. This paper gives an overview of the main results obtained from the HHF tests of about 45 components. The microscopic examination after loading analyses recrystallization effects and investigates the W/Cu interfaces in more detail.

Meeting the challenge of realizing fusion power production will require considerable and increasing investment in facilities for testing and development of fusion technology. Particularly important will be testing of components destined for the harsh in-vessel environment of the reactor. To help address this need, the U.K. Government is investing in major new fusion technology facilities, which will offer integrated laboratories covering the complete development life cycle from materials to manufacturing processes and load testing of components. A major part of these facilities shall be a test device named CHIMERA (combined heating and magnetic research apparatus), offering testing under fusion-relevant loads for meter-scale in-vessel component mock-ups. Among the major challenges addressed are electromagnetic loads, high heat flux (HHF), and proving complex and high-risk manufacturing. The ability to test technology in magnetic fields will be unparalleled and could prove vital for breeding blanket designs featuring a ferromagnetic structural material or a liquid metal breeder. The CHIMERA magnet system uses a split-pair NbTi superconducting magnet, combined with a vertical-axis pulsed resistive solenoid to simulate plasma disruptions. Furthermore, in order to provide semi-integrated testing including possible synergistic effects, CHIMERA will enable tests of resilience against magnetic and thermal loads in combination. The heating systems will deliver at least 0.5 MW/m(2) at the module surface, HHF in localized areas, and power for volumetric heating of a module. This article introduces the CHIMERA device, reports the motivation and technical basis, describes the system specification, and outlines the future plan.

A high performing DEMO divertor target mock-up design that uses the thermal break interlayer concept is presented. The design evolved from a previous design of which six mock-ups were designed, fabricated and subjected to high heat flux testing. The new design was generated using optimisation techniques; specifically, software was developed to automatically process the design of experiments data to enable visualisation of the design space. Despite the more challenging geometric constraints of this second phase, this design performs significantly better than that of the previous phase; the strain in the interlayer, which was the dominant damage mode in the phase 1 testing, is reduced by 28%. Four mock-ups of the selected design were manufactured, all of which successfully passed a series of high heat flux testing, including 500 cycles at 20 MW/m2. Design optimisation methods are not widely utilised in fusion engineering, their potential benefits, which are demonstrated here on a plasma facing component, could be applied to many other challenging designs.

The anticipated heat flux limit of the European DEMO first wall is similar to 1 - 2 MW/m(2). During transient and off normal events, the heat load deposited on the wall would be much larger than the steady state heat load and exceed the first wall limit, therefore the breeding blanket first wall needs to be protected. This involves dedicated discrete limiters in certain regions of the machine that would take the brunt of the heat load as well as adequate shaping of the first wall. The current concept envisages limiters at a few (3-4) equatorial ports to cope with the ramp-up of the plasma; upper limiters (in similar to 8 upper ports) are considered for upward vertical displacement events. Two design options have been considered for these limiters: a modular design where the limiter plasma facing components are attached to individual plates that are assembled together so that transient electro-magnetic loads can be reduced, and in case of damage the plates can be replaced/repaired individually; and a divertor-like design where the plasma facing components are attached to a single Eurofer cassette. Other limiters considered include inner wall limiters in case of plasma contraction and lower limiters may be needed for downward vertical displacement events. The thermal hydraulic finite element analysis results show that the integrity of the cooling pipes can be maintained during the anticipated transient events. The limiters are considered to be sacrificial and designed to be replaceable independently from the breeding blanket system. The design has to allow that installation, removal or replacement of the limiters can be performed remotely. Strategy to tackle outstanding issues and required R&D is also discussed.

The EU DEMO plasma is almost completely enveloped by large breeding blanket segments for tritium breeding and power extraction. Shaping of the blanket plasma-facing wall in 3D may prove to be essential, but this strategy alone is not sufficient to protect against anticipated transient plasma events. The high heat flux wall-limiter approach used in ITER is not thought to be viable in a tritium self-sufficient power reactor, and so in EU DEMO wall protection using discrete limiters is pursued. Two types of discrete limiter are described in this paper. One is an equatorial port limiter designed to handle the power during the plasma start-up phase, making use of water-cooled tungsten/CuCrZr monoblock technology. The second is the upper limiter, featuring a plasma-facing component designed specifically for extreme transient loading due to a vertical displacement event. The heat flux channelling and thermal barrier features of this design are shown to considerably reduce CuCrZr pipe temperature, and so reduce the likelihood of catastrophic failure. A preliminary neutronic calculation has shown that the impact of these discrete limiters on overall tritium breeding ratio is relatively low.

The design of plasma facing components (PFCs) requires knowledge of the charged particle heat load in the scrape-off layer (SOL). Ray-tracing codes like PFCFlux can model this heat load assuming that particles follow the magnetic field lines. Calculations on limiter equilibria underestimate the heat load significantly. In fact, not all the power circulating in the SOL is reported on the wall, with 80% of the total power circulating in the SOL missing in the worst cases. This paper explains why some power is missing in this case, and presents different ways to rescale the heat load to recover all the power coming from the SOL. The maximum heat load on the limiter for a given magnetic configuration can change from 1 MW/m(2) without rescaling to values from 3.5 MW/m(2) to 21.7 MW/m(2) depending on the rescaling method.