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In a demonstration (DEMO) reactor, mitigation of the large heat load on the divertor target to the below material and engineering limits is a key requirement for operation. Systems modeling is used to design entire fusion power plants and, therefore, has to be able to appropriately capture the divertor challenge. Therefore, it is important to validate these models against comprehensive SOL-divertor simulation codes and experiments. A 1-D divertor model in PROCESS was investigated, compared to the results of 2-D SONIC simulation under the detachment condition. The comparison shows how the 1-D divertor model handles the power loss mechanisms from the outboard mid-plane to the outer divertor target for a DEMO-like condition. The results show good agreement on the calculated value of the total power crossing the separatrix (<5% difference) and the total impurity radiation power $P_{\mathrm{ imp}}$ (<10% difference). However, the 1-D profiles show differences in density and temperature at the upstream of the target (<10 m of connection length to target, corresponding to 10 cm in the poloidal length). One reason for this difference is that the 2-D model calculates impurity transport, which produces a variable impurity fraction along the connection length in the divertor, while the 1-D model uses a single averaged value. The SONIC code also considers physical processes not covered in the 1-D model, such as radial transport in the SOL and divertor region. A scan of $P_{\mathrm{ sep}}$ values in PROCESS for a DEMO-sized machine show that above $P_{\mathrm{ sep}} = 200$ MW, there is a stronger impact on cost and machine size for higher $P_{\mathrm{ sep}}$. [ABSTRACT FROM AUTHOR]

The official Spherical Tokamak for Energy Production mission aims to demonstrate the ability to generate net electricity from fusion with the STEP Prototype Power plant. One of the key technological and engineering challenges in fusion power plants is managing the loads on the first wall within acceptable limits. Therefore, the conceptual design development of the STEP Prototype Power plant needs to be based on load estimates derived using legitimate plasma physics assumptions through dynamic and flexible tools. The current design foresees the STEP main chamber first wall to withstand steady-state heat loads of up to ∼ 1 MW/m 2 , excluding critical regions expected to receive higher heat loads such as the baffle regions approaching the divertors. These critical areas will require ad hoc assessments and will be designed with the presence of limiters. This article focuses on the models and methodology adopted for designing the 2-D poloidal contour of the STEP first wall, based on the anticipated charged particle and radiation heat loads during normal operation. Firstly, the models adopted for calculating the charged particle and radiation heat loads are introduced. The first model is validated through benchmarking against the particle tracing code SMARDDA, while the second model is verified by comparing it with data from the MAST-U experiment. Secondly, the model used to design the 2-D first wall contour according to the heat loads is explained. We acknowledge that this preliminary design stage assumes certain simplifications, notably an axisymmetric geometry, for computational efficiency and clarity in presentation. It is understood that subsequent design phases will address the complexities of real-world engineering, including non-axisymmetric effects, transient plasma scenarios, and the impact of disruptions on the first wall design. Finally, an automatic procedure based on these models is presented for defining the 2-D poloidal contour of the STEP first wall to minimize heat loads, taking into account the need to radiate most of the alpha-particle and auxiliary heating power. By providing an overview of the models, methodology, and an automatic procedure, this paper contributes to the design process of the STEP first wall, addressing the engineering challenges associated with fusion power plant development. • Proposes a methodological approach for designing the initial first wall contour of the STEP Prototype Power plant. • Addresses significant engineering challenges associated with managing heat loads on the first wall in fusion power plants. • Introduces models and methodology for predicting heat loads resulting from charged particles and radiation. • Develops an automatic procedure for defining the 2D poloidal contour of the STEP first wall to minimize heat loads. • Contributes to the ongoing design process of the STEP first wall and addresses engineering challenges in fusion power plant development. [ABSTRACT FROM AUTHOR]

• Engineering workflow for misalignment studies by using 3D field line tracing codes. • The heat load pattern on geometrical configurations deviating from their reference ones can be retrieved by using penalty factors. • Penalty factors are not self-exhaustive: heat flux magnitude and duration of its applications are parameters to consider in an engineering design. • During Normal Operations, hot spots should be prevented in case of segments' misalignment by defining assembly tolerance levels. • Misalignment tolerance studies help drive the design of FW and limiters. Within the framework of EURO fusion DEMO First Wall and limiter design activities, the protection of the First Wall against power deposition peaks is being considered. During steady-state operation, the radiative power from the plasma could be considered more uniformly spread on plasma-facing components than the charged particle power deposition. The presence of openings (i.e. gaps between segments and ports) and the introduction of limiters breaks the continuity of the wall and opens the possibility of localized high heat flux values on toroidal-facing gaps due to charged particles striking the wall. The heat flux pattern is likely to be altered by misalignments between components due to manufacturing, assembly or non-uniform/non-symmetrical operational conditions in the segments. In this paper, the 3D field line tracing code SMARDDA is used for studying the impact of misalignment on the heat load distribution for a periodically segmented DEMO First Wall, specifically the Single Module Segment Concept. The present work does not address any study on limiter misalignments, which will be treated separately. The work covers normal operation conditions (ramp-up and steady-state), considering both the cases of bare First Wall (without limiters, as reference) and First Wall protected by limiters. The main aim of the work is understanding how the power deposition on the First Wall (due to charged particles) is affected by the presence of a radial or vertical misalignment between segments and starting a workflow to be applied later to all the possible plasma scenarios and First Wall layout. Heat flux penalty factor maps have been created to identify the worst cases among the ones analyzed. The related heat flux maps are relevant for thermal assessments of a simplified DEMO model to be performed later on, particularly in terms of maximum temperature values for the current DEMO design. [ABSTRACT FROM AUTHOR]

• DEMO limiter system overview. • Description of shield block design options, potential candidate for PFC technology. • Introduction of the four different kind of limiters proposed and their roles and main parameters. • DEMO first wall and limiter surface shaping principles. • Status of supporting R&D. The anticipated heat flux limit of the European DEMO first wall is ∼1−2 MW/m2. 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 ∼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 electromagnetic 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. [ABSTRACT FROM AUTHOR]