This paper presents a comprehensive overview of the preliminary divertor design and plasma exhaust scenario for the reactor-class Spherical Tokamak for Energy Production project. Due to the smaller size of the machine, with a major radius less than half that of most DEMO concepts, the current design features a double-null divertor geometry, comprising tightly baffled extended outer legs and shorter inner legs approaching an X-divertor. Leveraging a significant database of SOLPS-ITER simulations, the exhaust operational space is mapped out, offering valuable insights into the plasma exhaust dynamics. An approach involving the validation of simple, yet robust models capable of accurately predicting key exhaust parameters is detailed, thereby streamlining the design process. The simple models are used to simulate the entire plasma scenario from the plasma current ramp-up, through the burning phase, to the plasma current ramp-down. Notably, the findings suggest that pronounced detachment, with peak heat loads below engineering limits and electron temperatures below 5 eV, is achievable with a divertor neutral pressure between 10 Pa and 15 Pa during the burning phase, and pressures below 5 Pa during the ramp-up to maximise the auxiliary current-drive efficiency. Throughout the scenario, an Ar concentration of approximate to 3% in the scrape-off layer (SOL) is required, in combination with a core radiation fraction of 70% driven by intrinsic emission and extrinsic injection of Xe seeded fuelling pellets. However, significant uncertainties remain regarding key parameters such as the SOL heat flux width, Ar screening, and plasma kinetic effects.

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 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.

A design integration study (KDII4) was conducted in the DEMO Pre-Concept Design Phase with the primary goal to develop a pre-concept feasibility design and concept of operation for the Vertical Blanket Segment Architecture. The primary goal was to develop two workable variants for removal of large in-vessel components. This, prompted by two earlier studies, highlighting the integration challenge finding a self-consistent DEMO design point (KDII) and secondly the complexity of operation required to ensure compatibly with the proposed port-based maintenance schemes. Initially, only a single null (SN) divertor configuration was considered, but due to additional identified issues and technical challenges centred on the extraction of the BB segments, a double null (DN) alternative variant has also been investigated. This was prompted by other KDII's studies, especially KDII1 (Design, performance and feasibility of wall protection limiters during plasma transients) and KDII3 (Advanced Magnetic Configurations). Alternative forms of vertical maintenance architecture have been investigated by breaking down the study into Ports, In-vessel Components, Operations and Safety. SN and DN configurations were studied with split and full blankets numbering seven variants in total. Each variant was studied in some detail, comparatively assessing each on its merits. However, no viable solution has been found for down-selection. Nevertheless, the question of vertical maintenance is now better understood. This paper will describe the findings of KDII4 in conjunction with the Remote Maintenance (RM) technology work package, which suggests that the current ‘reference’ design envelope available for RM is too constrained.

DEMO is a key part of the EU fusion roadmap, and the programme reaches the end of the pre-conceptual phase with a gate review in 2020. As part of the work to complete this phase, eight Key Design Integration Issues (KDII's) have been identified as critical to the programme. Two of these KDII's identified a requirement for a more detailed architectural study. Within KDII#3 (advanced divertor configurations) a double null configuration has been developed; in parallel KDII#4 (vertical segment architecture) identified a need to evaluate a split breeding blanket architecture. These two requirements were combined into a single study that assesses the invessel architecture for a double null DEMO featuring split breeding blankets. This paper presents the configurations developed and an evaluation of their feasibility with respect to integration of key in-vessel components (breeding blankets, divertors), port hardware (service pipes, shielding, vacuum pumping), and maintenance strategy. Furthermore, wider considerations such as the impact of the port orientation on the building architecture are also included. Finally, this paper will identify the risks and further work required to advance the double null configuration.

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.

Substantial progresses have been achieved in the realization of the ITER Neutral Beam Test Facility (NBTF) hosted in Padova, Italy; the buildings, completed by the end of 2015, are being progressively filled with new systems and components. The realization of SPIDER, the ITER full-size negative ion source, is well advanced and important progress is also recorded for MITICA, the full-scale prototype of the ITER HNB injector. The paper gives an overview of the achieved results, highlighting the main challenges faced. (C) 2017 Elsevier B.V. All rights reserved.

Any leak in one of the ITER actively cooled components would cause significant consequences for machine operations; therefore, the risk of leak must be minimized as much as possible. In this paper, the strategy of examination to ensure leak tightness of the ITER internal components (i.e. examination of base materials, vacuum boundary joints and final components) and the hydraulic parameters for ITER internal components are summarized. The experiences of component tests, especially hot helium leak tests in recent fusion devices, were reviewed and the parameters were discussed. Through these experiences, it was confirmed that the hot He leak test was effective to detect small leak paths which were not always possible to detect by volumetric examination due to limited spatial resolution.

The ITER project requires additional heating by two neutral beam injectors, each accelerating to 1 MV a 40 A beam of negative deuterium ions, to deliver to the plasma a power of about 17 MW for one hour. As these requirements have never been experimentally met, it was recognized as necessary to setup a test facility, PRIMA (Padova Research on ITER Megavolt Accelerator), in Italy, including a full-size negative ion source, SPIDER, and a prototype of the whole ITER injector, MITICA, aiming to develop the heating injectors to be installed in ITER. This realization is made with the main contribution of the European Union, through the Joint Undertaking for ITER (F4E), the ITER Organization and Consorzio RFX which hosts the Test Facility. The Japanese and the Indian ITER Domestic Agencies (JADA and INDA) participate in the PRIMA enterprise; European laboratories, such as IPP-Garching, KIT-Karlsruhe, CCFE-Culham, CEA-Cadarache and others are also cooperating. Presently, the assembly of SPIDER is on-going and the MITICA design is being completed. The paper gives a general overview of the test facility and of the status of development of the MITICA and SPIDER main components at this important stage of the overall development; then it focuses on the latest and most critical issues, regarding both physics and technology, describing the identified solutions.

The design of the beamline components of MITICA, the full prototype of the ITER heating neutral beam injectors, is almost finalised and technical specifications for the procurement are under preparation. These components are the gas neutraliser, the electrostatic residual ion dump, and the calorimeter. Election dump panels are foreseen each side of the upstream end of the neutraliser to protect the cryo-panels from electrons, created by stripping and other processes, that exit the 1 MeV accelerator. As the design of the components must fulfil requirements on the beam physics, insight on physical processes is required to identify performance trade-offs and constraints. The spatial gas distribution was simulated to verify the pumping requirements with electron dump panels and local conditions for breakdown voltage. Electrostatic analyses were carried out for the insulating elements of the RID to verify the limits of the electric field intensity. Different criteria were approached to investigate the fracture behaviour of ceramics considering the manufacturing implications and extrapolating the conditions for proof testing. Severe heating conditions will be applied steadily, as the maximum pulse duration is I h, and cyclically so requiring to fulfil fatigue and ratcheting verifications. High heat fluxes, up to 13 MW/m(2) on the calorimeter, with enhanced heat transfer in subcooled boiling conditions will occur in the actively cooled CuCrl Zr panel elements provided with twisted tapes as turbulence promoters. Special R&D activities were undertaken to support the design: manufacturing of thick twisted tapes leading to an increased cooling performance while maintaining flow rate requirements, bending of swirl tubes, verification for permanent deformations due to stress relaxation after heating of swirl tubes, double side deep drilling of 2 m long CuCr1Zr plates. (C) 2015 Elsevier B.V. All rights reserved.

The ITER project requires additional heating by two neutral beam injectors, each accelerating to 1MV a 40A beam of negative deuterons, delivering to the plasma about 17MW up to one hour. As these requirements have never been experimentally met, it was decided to build a test facility, PRIMA (Padova Research on ITER Megavolt Accelerator), in Italy, including a full-size negative ion source, SPIDER, and a prototype of the whole ITER injector, MITICA, aiming to develop the heating injectors to be installed in ITER. The Japan and the India Domestic Agencies participate in the PRIMA enterprise; European laboratories, such as KIT-Karlsruhe, IPP-Garching, CCFE-Culham, CEA-Cadarache and others are also cooperating. In the paper the main requirements are discussed and the design of the main components and systems are described.

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.

Negative Ion Neutral Beam systems on ITER will require actively cooled scrapers and dumps to process and shape the beam before injection into the tokamak. The scale of the systems is much larger than any presently operating, bringing challenges for designers in terms of available sub cooling, total pressure drop, deflection and mandatory remote maintenance. High heat fluxes (∼15–20MW/m2), pulse lengths in excess of 3000s and high number of cycles pose new challenges in terms of stress and fatigue life. The designs outlined in the Design Description Document for the ITER Neutral Beam System [N53 DDD 29 01-07-03 R 0.1. ITER Design Description Document, DDD 5.3, Neutral Beam H&CD system (including Appendices).], based on swirl tubes, have been reviewed as part of the design process and recommendations made. Additionally, alternative designs have been proposed based on the Hypervapotron high heat flux elements with modified geometry and drawing upon a vast background knowledge of large scale equipment procurement and integration. A full thermo-mechanical analysis of all HHF components has also been undertaken based on ITER design criteria and the limited material data available. The advantages and disadvantages of all designs are presented and recommendations for improvements discussed.

The design of the ITER ICRF system has been under revision for several years. The paper presents the status of the design proposal based on a 24 strap antenna plug (6 poloidal by 4 toroidal short radiating conductors) in which the straps are passively combined in 8 poloidal triplets by means of 4‐port junctions. These triplets are connected in parallel pairwise through matching elements to form 4 load‐resilient conjugate‐T circuits. All adjustable matching elements are located outside the plug, i.e. in the ITER port cell and in the generator area.

The European effort for the EC power sources is centred on the development of a 170 GHZ, 2 MW, CW coaxial cavity gyrotron of Collector Potential Depressed (CPD) type. 2.2 MW in single mode were demonstrated in short pulse operation on a pre-prototype gyrotron at FZK. Test on the first industrial gyrotron prototype will start before end 2006 in a new full performance EC test facility in Lausanne (CH). The procurement of the ITER NB system presents several challenges and an extensive R&D programme has been launched. In EU the experimental investigations take mainly place at CEA, Cadarache, where the arc driven source and the concept of an accelerator based on a single gap are pursued and at IPP, Garching, were a RF driven ion source is being developed. For the arc driven source, ITER baseline, the achieved current density is, in long pulses, still lower than the one required for ITER and the Cs consumptions is about three orders of magnitude higher than the one anticipated for ITER. Promising progresses have been made towards a RF driven ion source. Current densities of 330 A/m in H and 250 A/m in D were achieved on a small scale source with good reproducibility and operation in D in ITER relevant parameters space was achieved over many pulses. The RF driven ion source can now be considered a valid alternative to the arc driven one. Finally, the limited performances of the existing test facilities do not allow a reliable extrapolation to the ITER scale. The establishment of a full scale test facility has therefore become a centre piece of the European NB development strategy.