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The investigation of time-dependent power requirements for a future nuclear fusion reactor is part of the systems integration task for the European Fusion Programme. A pulsed DEMO will require intermittent and continuous electrical power to operate the various plant systems. For example, the heating and current drive system will require power during I p start-up, heating, plasma burn and I p ramp-down phases. This paper presents the modelling of these power requirements over the whole pulse cycle for all plant systems in the systems code PROCESS. In the EUROfusion baseline case the continuous power requirements are −290 MW and are dominated by the primary coolant pumping requirements. The total intermittent power requirements range from −50 to −375 MW. The time between pulses strongly influences the average plant net electrical power production and for a power plant would have to be minimised to stay commercially viable. The EUROfusion DEMO baseline is estimated to produce enough electrical power to meet the machine's aims given the current burn time, time between pulses and power requirements. [ABSTRACT FROM AUTHOR]

Tritium production is of critical importance to prospective DT fusion power plants. Lithium ceramic and beryllium based solid-type breeder blankets are an option for supplying the tritium required to sustain the DT plasma. This research investigates the time-varying tritium production in solid breeder blankets with different compositions. The breeder fraction was varied in conjunction with the Li 6 enrichment. The parameter study considered 198 different blanket compositions for three blanket thicknesses. The cheapest configuration capable of meeting the tritium requirements were found. The cost of Li 4 SiO 4 (including Li 6 enrichment) and Be 12 Ti were considered. The time-varying tritium production of each blanket configuration was simulated using the interface code, FATI, that couples the radiation transport code MCNP 6 with the inventory code FISPACT-II. Economical blanket configurations capable of self-sufficiency were found. The cost of producing excess tritium for start-up inventories was found to be between $18,000 and $27,000 per g. Fitting functions to predict the time-averaged tritium breeding fraction and the tritium inventory at five years, were obtained for inclusion in the PROCESS systems code. PROCESS is now able to consider different breeding blanket compositions and thicknesses when assessing the engineering, physics and economic feasibility of reactor designs. [ABSTRACT FROM AUTHOR]

PROCESS is a reactor systems code – it assesses the engineering and economic viability of a hypothetical fusion power station using simple models of all parts of a reactor system. PROCESS allows the user to choose which constraints to impose and which to ignore, so when evaluating the results it is vital to study the list of constraints used. New algorithms submitted by collaborators can be incorporated – for example safety, first wall erosion, and fatigue life will be crucial and are not yet taken into account. This paper describes algorithms relating to the engineering aspects of the plant. The toroidal field (TF) coils and the central solenoid are assumed by default to be wound from niobium-tin superconductor with the same properties as the ITER conductors. The winding temperature and induced voltage during a quench provide a limit on the current density in the TF coils. Upper limits are placed on the stresses in the structural materials of the TF coil, using a simple two-layer model of the inboard leg of the coil. The thermal efficiency of the plant can be estimated using the maximum coolant temperature, and the capacity factor is derived from estimates of the planned and unplanned downtime, and the duty cycle if the reactor is pulsed. An example of a pulsed power plant is given. The need for a large central solenoid to induce most of the plasma current, and physics assumptions that are conservative compared to some other studies, result in a large machine, with a cryostat 36 m in diameter. Multiple constraints, working together, restrict the parameter space of the optimised model. For example, even when the ratio of operating current to critical current in the TF coils is increased by a factor of five, the total coil cross-section decreases only a little, because of the need for copper stabiliser, insulation, and structural support. The result is that the plasma major radius hardly changes. It is these surprising results that justify the development of systems codes. [ABSTRACT FROM AUTHOR]

Highlights: [•] A three-stage compound cryopump could be used in fusion reactors. [•] Helium “ash” is adsorbed at 4.5K, and deuterium and tritium are adsorbed at 15–22K. [•] We used a free Direct Simulation Monte Carlo code (DS2 and DS2V). [•] 90% of the deuterium and tritium are successfully trapped at 15K. [•] Fuel is returned directly to the plasma, without isotopic separation or storage. [ABSTRACT FROM AUTHOR]

Abstract: We outline a speculative design for a photodetachment neutraliser for a negative ion neutral beam system, with neutralisation efficiency of 95% or more. The practical difficulties are enormous. The ion beam must pass through an optical cavity capable of reflecting the light many times. For 500 reflections, the laser optical power output ∼800kW, giving circulating power ∼400MW. All sources of light loss combined need to be kept to 0.2% or less per pass. The losses due to photodetachment itself, and due to Thomson scattering in the beam plasma are negligible. A key task is to maintain the reflectance of the mirrors above 99.97% for long periods of operation, protecting all the components from thermal and neutron damage, and from caesium, sputtered atoms and other contamination. A diode-pumped Nd-doped YAG laser can have overall electrical-to-light (“wall-plug”) efficiency up to 25%. A DEMO concept reactor such as the EU Power Plant Conceptual Study (PPCS) Model B requires 270MW heating power. If this is all provided by neutral beams, then a laser neutraliser might reduce the electrical power consumption for this from 900MW to 520MW. [ABSTRACT FROM AUTHOR]

Measurements have been made of the emittance plot of a partly neutralized positive ion beamlet, using a slit, a polymer target, and an infrared camera. This thermographic approach is intrinsically linear and absolute (since the properties of the target are known and approximately independent of temperature). It is sufficiently sensitive that only one short pulse is required to capture the entire range of the angular coordinate y′. Ions and neutral atoms are both detected with equal sensitivity. The measurement is unaffected by secondary electrons emitted by the target and by electrons traveling with the positive ion beam, as both types of electron carry very little energy. No specialized electronics or beam deflection devices are required. Providing the region downstream of the mask is free of fields, the target can be several meters away from the mask, allowing good resolution in the y′ axis. The target can be as large as required. [ABSTRACT FROM AUTHOR]

Power plant studies using systems codes allow the optimisation of designs to maximise or minimise some figure of merit: fusion power gain or cost of electricity, for example. The code should trade off between parameters to find the optimum whilst producing a solution consistent with physics and technology limitations. This paper describes the recently updated superconducting toroidal field coil (TFC) stress model in the systems code PROCESS. The TFC structure is critical in determining the reactor design as it influences key parameters, in particular the radial build and toroidal field. The model was validated with FEA and used to investigate how TF stress influences DEMO concept design in both pulsed (DEMO1) and steady-state (DEMO2) devices. The allowable stress in the TFC structural components was scanned between 440–720 MPa for runs in which PROCESS was minimising the major radius, R 0 , and produced a variation in R 0 of ∼1 m for fixed aspect ratio. The capital cost varied by $2–3 bn over the same range. Understanding how some parameters limit the design is essential for exploring new DEMO concepts and guiding future research. [ABSTRACT FROM AUTHOR]

The emissivity of metallic materials is low and varies with temperature and wavelength inducing errors on surface temperature measurements. High heat flux experiments on beryllium were carried out to investigate the thermal behavior of bulk Be tiles. Thermal modeling aiming at determining the surface and bulk temperatures have been performed using ANSYS®. A Be tile was exposed to heat flux with power density ranging between 1 and 7 MW/m 2 . Surface temperatures were measured using an infrared camera in the 3–5 μm range and two-color pyrometers, one at short wavelengths (1.5–1.7 μm) and one at mid IR range wavelengths (2–4 μm) range. Both the IR camera and two-color pyrometers do not provide accurate temperature measurements on melted Be due to changes in the emissivities and emissivity ratio induced by surface modifications. [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]

As power plant design programmes approach the transition between the pre-conceptual and conceptual design phases, the systems code PROCESS has been improved to incorporate more detailed plasma physics, engineering, and analysis modules. Unlike many systems codes, PROCESS combines the physics modelling with both technology and costs analysis. Some of the key topics in the conceptual design phase are toroidal field (TF) magnet design, divertor power handling, operational sensitivity, and economic uncertainty analysis. Models covering these areas have been integrated or improved in PROCESS. During pre-conceptual design, systems codes are an essential tool for exploring fusion power plant concepts. They allow one to model the interaction of the plant systems and quickly perform reactor optioneering. To be able to carry out these large scoping studies, the fidelity of the models can be restricted to reduce the computational time. For example, the EUROfusion EU-DEMO baseline designs are created using the systems code PROCESS and the ability to measure these trade-offs has led to important design choices being examined during the DEMO pre-conceptual design phase. Ruling out unfeasible designs allows one to efficiently identify where in the design space to carry out detailed design work. The paper describes how PROCESS has begun retooling for use in later stages of power plant conceptual design. Details are given for new additions to the PROCESS uncertainty quantification tools, high-temperature superconducting magnet model, TF coil model, and new models for spherical tokamaks (STEP). Additionally, the paper covers recent developments on the novel systems code BLUEPRINT, which is being used in both the EUROfusion and STEP projects. The paper concludes with an outlook on systems code activities at UKAEA and work with external partners. [ABSTRACT FROM AUTHOR]

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]

Within the DEMO first wall 3D shape design activity, studying the effect of misalignment started in 2017. Such assessments have been conducted in the past for ITER and heat flux penalty factor maps have been created [ 1 ]; this route could be a feasible approach in the case of DEMO as well. This paper details the methodology that allows the effects of misalignments to be assessed for DEMO. The test cases focus on steady-state plasma operation (start of flat top). The aim is to understand the effect of basic misaligned cases, for example radial protrusion/recession or poloidal rotation of a single module. To do so, particle tracing software codes such as SMARDDA and PFCflux have been used to create heat flux maps that reach the first wall surfaces. These heat flux maps, combined with the specified radiative heat load, are used as input for simplified FE models of the blanket modules. As a result, not only the effect on heat flux, but also on the temperature (and later stress) distributions can be estimated. The paper describes how the obtained results can be implemented in ANSYS in the identified critical cases from the test matrix that has been studied. The results obtained from the nominal heat flux map are compared to the misaligned cases. The mitigating effect of the 3D nature of the heat conduction on the peak temperature is discussed. This work paves the way to assessing more realistic combined misaligned cases (such as misalignment from different thermal expansion, or due to electromagnetic loads etc. of neighbouring blankets) in the future. [ABSTRACT FROM AUTHOR]

• CHIMERA will be uniquely capable of 'semi-integral' testing of fusion components. • Key systems have required specific development activities. • The large array first wall heating system is under qualification. • Certain diagnostics are de-risked via testing in a pulsed magnet field. • An integrated systems model of the commissioning mock-up is producing first results. 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 development project is now producing first virtual test results, informing commissioning of the CHIMERA device. [ABSTRACT FROM AUTHOR]

Injection of electromagnetic waves in the electron-cyclotron (EC) frequency range is one of the most promising schemes to drive part of the plasma current in a tokamak fusion reactor. The theoretical calculation of the driven current, as usually performed by ray/beam tracing codes, relies on the knowledge of the magnetic equilibrium, the electron density, and temperature profiles on the one hand and of the wave injection parameters on the other. If the optimum current drive efficiency for a given scenario is sought, extensive parameter scans are usually performed to determine the best injection conditions. This is, however, not a viable approach in typical systems-code applications, where the plasma configuration is not provided in sufficient detail and parameter scans would be anyway too demanding from the computational point of view. In this case, a different approach is required. In this paper, a procedure for the evaluation of the optimum current driven by EC waves for given global parameters is proposed, which relies on a single numerical calculation of the current drive efficiency, based on the adjoint method (including momentum-conserving corrections). The results are shown to be in good agreement with the full numerical optimization of the EC current drive efficiency for a variety of reactor relevant scenarios. This simplified approach also helps clarify the physics underlying the optimum current-drive conditions and the limitations to the achievable current-drive efficiency in reactor-grade plasmas. [ABSTRACT FROM AUTHOR]

Highlights • Preliminary design of DEMO first wall 3D surface is presented. • Steady state and controllable perturbations are preented,evaluating the radiated HF density on the wall and the charged particle power. • Different plasma-wall contact phases have been analysed for a list of foresseable and unforeseeable transients. • 3D field line tracing codes were employed to analyse and design discrete high heat flux components surfaces. Abstract The present DEMO breeding blanked design heat load capability is limited to ≈1 MW/m2 for steady state plasma loading, due to the specific requirements on high neutron irradiation capable materials, and high coolant temperature for efficient energy conversion. While this limit is achievable in nominal conditions in the present DEMO blanket concept designs, the greatest challenges arise from the occurrence of plasma transients. The results of simulations of a number of plasma transients are presented in this paper. 3D field-line tracing codes have been used to analyses the maximum heat flux and energy density for a specific first wall shape design, and optimize it. A scoping study has been performed with the thermal analysis code RACLETTE, using a broad range of transient input heat fluxes, on a series of high heat flux (HF) components concepts with tungsten armor, Eurofer steel or copper alloy as heat sink materials, and helium or water as coolant. The results permit the identification of the operational space of the peak HF density that can be tolerated by different plasma facing components, for the different transient models. [ABSTRACT FROM AUTHOR]

Beryllium (Be) is a precious resource with many high value uses, the low energy threshold (n,2n) reaction makes Be an excellent neutron multiplier for use in fusion breeder blankets. Estimates of Be requirements and available resources suggest that this could represent a major supply difficulty for solid-type blanket concepts. Reducing the quantity of Be required by breeder blankets would help to alleviate the problem to some extent. In addition, it is important that the reduction in the Be quantity does not diminish the blanket's performance in key aspects such as the tritium breeding ratio (TBR), energy multiplication and peak nuclear heating. Mixed pebble bed designs allow for the multiplier fraction to be varied throughout the blanket. This neutronics study used MCNP 6 to investigate linear variations of the multiplier fraction in relation to blanket depth, in order to better utilise the important multiplying Be(n,2n) and breeding reactions. Blankets with a uniform multiplier fraction showed little scope for reduction in Be mass. Blankets with varying multiplier fractions were able to simultaneously use 10% less Be, increase the energy amplification by 1%, reduce the peak heating by 7% and maintaining a sufficient TBR when compared to the performance achievable using a uniform composition. [ABSTRACT FROM AUTHOR]

The European DEMO power reactor is currently under conceptual design within the EUROfusion Consortium. One of the most critical activities is the engineering of the plasma-facing components (PFCs) covering the plasma chamber wall, which must operate reliably in an extreme environment of neutron irradiation and surface heat and particle flux, while also allowing sufficient neutron transmission to the tritium breeding blankets. A systems approach using advanced numerical analysis is vital to realising viable solutions for these first wall and divertor PFCs. Here, we present the system requirements and describe bespoke thermo-mechanical and thermo-hydraulic assessment procedures which have been used as tools for design. The current first wall and divertor designs are overviewed along with supporting analyses. The PFC solutions employed will necessarily vary around the wall, depending on local conditions, and must be designed in an integrated manner by analysis and physical testing. [ABSTRACT FROM AUTHOR]

To prepare the DEMO conceptual design phase a number of physics and engineering assessments were carried out in recent years in the frame of EFDA concluding in an initial design configuration of a DEMO tokamak. This paper gives an insight into the identified engineering requirements and constraints and describes their impact on the selection of the technologies and design principles of the main tokamak components. The EU DEMO program aims at making best use of the technologies developed for ITER (e.g., magnets, vessel, cryostat, and to some degree also the divertor). However, other systems in particular the breeding blanket require design solutions and advanced technologies that will only partially be tested in ITER. The main differences from ITER include the requirement to breed, to extract, to process and to recycle the tritium needed for plasma operation, the two orders of magnitude larger lifetime neutron fluence, the consequent radiation dose levels, which limit remote maintenance options, and the requirement to use low-activation steel for in-vessel components that also must operate at high temperature for efficient energy conversion. [ABSTRACT FROM AUTHOR]

High energy neutrons produced in future fusion reactors will cause significant transmutation reactions in the breeder blanket, including the important tritium breeding reactions. The reaction rate for a given reaction type depends on the neutron spectrum and material properties. The inventory consequences for tritium production of a solid-type breeder blanket are discussed in this paper. A DEMO fusion reactor with 19 homogeneous breeder blanket modules made of Eurofer, helium, Be 12 Ti and Li 4 SiO 4 , with enriched Li 6 content. Each blanket module was segmented radially in order to analyse the time-dependent reaction rate as a function of depth. The resolution of radial segmentation was varied and 5 radial divisions were found to be sufficient to accurately model the tritium inventory. Time-dependent tritium production was simulated with the use of the interface code, FATI, which couples radiation transport code MCNP 6 with the inventory code FISPACT-II. The simulated results show how the tritium production varies over the expected lifetime of the blanket. The overall tritium production of the solid-type blanket decreases as the Li 6 and Li 7 are burnt up. The effect of Li 6 burn-up in the breeder zones nearest to the plasma is identified as the main contributing factor to the decreasing tritium production of the whole breeder blanket. [ABSTRACT FROM AUTHOR]

In normal operation the JET neutral beam injectors have the operating gas supplied to the ion source and the neutraliser. For tritium operation the gas is supplied to both the ion source and neutraliser at a point close to the earth grid (“grid gas”) due to the difficulty in producing a gas line with a secondary containment and a ceramic break for high voltage standoff. In preparation for the next JET tritium campaign the JET EP2 PINIs have been characterised with grid gas flow. This paper reports measurements of arc efficiency, species and divergence in both normal and grid gas operation with hydrogen and deuterium. The data is used to predict the performance in tritium operation. [ABSTRACT FROM AUTHOR]

Abstract: The JET neutral beam injection (NBI) system is undergoing an upgrade of both beam power and pulse duration, which will be completed in 2011. In order to obtain an early assessment of the performance of the upgraded injectors, two positive ion neutral injectors (PINIs) with modified ion source and accelerator configuration were installed on Octant 8 Neutral Injector Box and successfully commissioned in summer 2009. Both PINIs were routinely delivering ∼2MW of deuterium neutral beam power during the JET experimental campaign in autumn 2009. These early tests allowed us to predict with confidence that the JET NBI upgrade objective of injecting 34MW of total deuterium neutral beam power into the JET plasma will be achieved. [Copyright &y& Elsevier]

Abstract: The objectives of the JET neutral beam enhancement (NBE) include raising the delivered power from the present 25MW to more than 34MW and increasing the pulse length from 10 to 20s. The additional power will be obtained partly by increasing the fractional energy components of the beam, resulting from acceleration of molecular ions, hence increasing the total particle flux. These changes place extreme demands on the design of the upgraded protection to the torus entry duct. The present inertial duct protection already reaches its thermomechanical limit in 10s pulses, and active cooling of the upgraded duct protection is therefore essential. Extensive analysis of the pressure and temperature evolution in the present un-cooled duct established the relationship between gas re-emission and surface temperature for copper in this operating environment. This information was used in an integrated physics and engineering approach to the design of the actively cooled duct protection, taking into account the power loads from direct beam interception and re-ionisation. Surface temperature determines power density through the gas re-emission and consequential beam re-ionisation. These considerations define the normal operating point for the chosen enhanced hypervapotron element technology. This approach demonstrated that supplementary in situ duct cryopumping would not be needed, provided that the required heat-transfer performance could be met without any encroachment of the elements beyond the space envelope of the existing inertial duct protection plates. This requirement posed severe constraints on the mechanical design of the hypervapotron element array and its manifolding; the adopted engineering design solutions are presented. [Copyright &y& Elsevier]

Abstract: The JET neutral beam (NB) heating system is being upgraded as a part of the ongoing JET Enhancement Programme. This is one of the largest upgrades of the JET machine carried out within the EFDA-JET framework. The main goals of the project are to increase the NB power delivered to JET plasma, to increase the beam pulse duration and to improve the availability and reliability of the JET NB system. The upgrade of the system is being carried out through the modification of the two existing neutral injector boxes (NIBs), each equipped with up to eight positive ion neutral injectors (PINIs). Significant changes of the JET NB system will be carried out within the next few years and will include modification of all PINIs, modification or replacement of various beamline components and corresponding instrumentation, procurement and installation of new high voltage power supply (HVPS) units and corresponding control systems and refurbishment of the 36kV power distribution. Various physics, engineering and planning issues related to this project, as well as the current status of the project are discussed in detail. Particular attention is given to the results of a PINI prototype test, which are of crucial importance for the successful completion of the entire enhancement programme. Upon the completion of the project in 2009/2010, JET NB system should be capable of delivering more than 34MW of deuterium beam power into the JET plasma for a duration of up to 20s with improved reliability. This will significantly enhance overall capabilities of the JET machine in support of ITER development. [Copyright &y& Elsevier]

• 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]

Abstract: A potentially important feature in a divertor design for a high-power tokamak is an extended and expanded divertor leg. The upgrade to MAST will allow a wide range of such divertor leg geometries to be produced, and hence will allow the roles of greatly increased connection length and flux expansion to be experimentally tested. This will include testing the potential of the Super-X configuration [1]. The design process for the upgrade has required analysis of producing and controlling the magnetic configurations, and has included consideration of the roles that divertor closure and increasing magnetic connection length will play. [Copyright &y& Elsevier]