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Abstarct The negative ion source SPIDER produces a beam power that has to be absorbed by a beam dump component made of stacked hypervapotrons. The manufacturing process produced dimension errors that have given rise to a possible shine-through issue. Finite volume models were developed to carry out transient thermal analyses aiming at validating the design of a system made of copper screens to shield the heat loads passing through the beam dump. The evaluation of the best solution was conducted by looking both at the thermal performances and the technical feasibility. The results show the capability of the system to safely withstand the heat power up to one hour.

Heating neutral beam (HNB) injectors, necessary to achieve burning conditions and to control plasma instabilities in ITER, are characterized by such demanding parameters that a neutral beam test facility (NBTF) dedicated to their development and optimization is being realized in Padua (Italy) with direct contributions from the Italian government (through Consorzio RFX as the host entity) and the ITER international organization (with kind contributions from the ITER domestic agencies of Europe, Japan and India) and technical and scientific support from various European laboratories and universities. The NBTF hosts two experiments: SPIDER, devoted to ion source optimization for the required source performance, and MITICA, the full-size prototype of the ITER HNB, with an ion source identical to SPIDER. This paper gives an overview of the progress towards NBTF realization, with particular emphasis on issues discovered during this phase of activity and on solutions adopted to minimize the impact on the schedule and maintain the goals of the facilities. The realization of MITICA is well advanced; operation is expected to start in 2023 due to the long procurement time of the in-vessel mechanical components. The beam source power supplies, operating at 1 MV, are in an advanced phase of realization; all high-voltage components have been installed and the complex insulation test phase began in 2018. At the same time, construction and installation of SPIDER plant systems was successfully completed with their integration into the facility. The mechanical components of the SPIDER ion source were installed inside the vessel and connected to the plants. Integrated commissioning with the control, protection and safety systems ended positively and the first experimental phase has begun. The first results of the SPIDER experiment, with data from operational diagnostics, and the plans for the 1 MV insulation tests on the MITICA high-voltage components are presented.

The heating neutral beam (HNB) systems at ITER are designed to inject a total of 33 MW of either 1 MeV D0 or 870 keV H0 beams into the ITER plasma using two injectors with a possible addition of a third injector later to increase the injected power to ∼50 MW. The injectors become radioactive due to the neutron flux from ITER and, in order to avoid the resulting complex remote maintenance, the design, choice of materials and the manufacturing process of each component of the injector is, wherever possible, such that they survive the life time of ITER. To ensure a smooth operational phase of neutral beams at ITER a neutral beam test facility (NBTF) is under construction at Consorzio RFX, Padova, (hereinafter referred to as RFX), which consists of 2 test beds, the 100 kV “SPIDER”, and a 1 MV “MITICA” facilities, which will be used to optimize the source operation for H and D beams. MITICA is essentially a full scale ITER prototype injector for the ITER beam parameters. The manufacturing and operation of the facility will allow validation of the operational space of the injectors and provide valuable information about the manufacturing processes applicable to HNB components. Operation of the two facilities is expected to begin in 2016 and 2019 respectively. Currently experiments on the ELISE facility with a half ITER sized RF beam source are underway. ITER relevant parameters for the H beams have almost been achieved. Efforts are underway to optimise the same with D beams. The experimental database from ELISE will be an important input for establishing the ITER relevant parameter space on the SPIDER source. This paper discusses the present status of the design and development of the injectors for ITER and the progress on the test facilities.

The ITER baseline foresees 2 Heating Neutral Beams (HNB’s) operating at 1 MeV 40 A D0, each capable of delivering 16.5 MW of deuterium neutrals to the plasma, with a 3rd HNB injector foreseen as an upgrade option that would bring up the total neutral beam power to 50MW [1]. In addition a dedicated Diagnostic Neutral Beam (DNB) will be injecting 100 keV 60 A of H0 for charge exchange recombination spectroscopy (CXRS) [2]. This paper highlights the impact of the staged approach on the ITER NBI, presents the new installation strategy and reports on the status of the adopted design refinements and the integration of the HNB’s and the DNB into the ITER plant.The ITER baseline foresees 2 Heating Neutral Beams (HNB’s) operating at 1 MeV 40 A D0, each capable of delivering 16.5 MW of deuterium neutrals to the plasma, with a 3rd HNB injector foreseen as an upgrade option that would bring up the total neutral beam power to 50MW [1]. In addition a dedicated Diagnostic Neutral Beam (DNB) will be injecting 100 keV 60 A of H0 for charge exchange recombination spectroscopy (CXRS) [2]. This paper highlights the impact of the staged approach on the ITER NBI, presents the new installation strategy and reports on the status of the adopted design refinements and the integration of the HNB’s and the DNB into the ITER plant.

SPIDER is one of two projects of the ITER Neutral Beam Test Facility under construction in Padova, Italy, at the Consorzio RFX premises. It will have a 100 keV beam source with a full-size prototype of the radiofrequency ion source for the ITER neutral beam injector (NBI) and also, similar to the ITER diagnostic neutral beam, it is designed to operate with a pulse length of up to 3600 s, featuring an ITER-like magnetic filter field configuration (for high extraction of negative ions) and caesium oven (for high production of negative ions) layout as well as a wide set of diagnostics. These features will allow a reproduction of the ion source operation in ITER, which cannot be done in any other existing test facility. SPIDER realization is well advanced and the first operation is expected at the beginning of 2018, with the mission of achieving the ITER heating and diagnostic NBI ion source requirements and of improving its performance in terms of reliability and availability. This paper mainly focuses on the preparation of the first SPIDER operations - integration and testing of SPIDER components, completion and implementation of diagnostics and control and formulation of operation and research plan, based on a staged strategy

The ITER Neutral Beam Test Facility (NBTF), called PRIMA (Padova Research on ITER Megavolt Accelerator), is hosted in Padova, Italy and includes two experiments: MITICA, the full-scale prototype of the ITER heating neutral beam injector, and SPIDER, the full-size radio frequency negative-ions source. The NBTF realization and the exploitation of SPIDER and MITICA have been recognized as necessary to make the future operation of the ITER heating neutral beam injectors efficient and reliable, fundamental to the achievement of thermonuclear-relevant plasma parameters in ITER. This paper reports on design and R&D carried out to construct PRIMA, SPIDER and MITICA, and highlights the huge progress made in just a few years, from the signature of the agreement for the NBTF realization in 2011, up to now - when the buildings and relevant infrastructures have been completed, SPIDER is entering the integrated commissioning phase and the procurements of several MITICA components are at a well advanced stage

The realization of the ITER Neutral Beam Test Facility (NBTF) and the start the experimental phase are important tasks of the fusion roadmap, since the target requirements of injecting to the plasma a beam of Deuterium atoms with a power up to 16.5 MW, at 1MeV of energy and with a pulse length up to 3600s have never been reached together before. The ITER NBTF, called PRIMA (Padova Research on ITER Megavolt Accelerator), is hosted in Padova, Italy; it includes two experiments: MITICA, the full-scale prototype of the ITER injector and SPIDER, the full-size negative ion source. The realization promoted by the ITER organization is carried out with the contribution of the European Union, channeled through the Joint Undertaking for ITER (F4E), of the Consorzio RFX which hosts the Test Facility, the Japanese and the Indian ITER Domestic Agencies (JADA and INDA) and several European laboratories, such as IPP-Garching, KIT-Karlsruhe, CCFECulham, CEA-Cadarache. The early start of operation of PRIMA experiments is urgent because sufficient experimental time is necessary to face and solve the issues related to the achievement of the desired performance in time for the ITER operation, requiring NBI since the beginning. Substantial progresses have been recently achieved: the buildings construction, begun in October 2012, has been completed by the end of 2015 and the installation of some components has been started since the end of 2014. The SPIDER realization is well advanced: the installation phase is proceeding in good agreement with the general plan; it is expected to be almost completed by the end of 2016. In parallel, the commissioning of the SPIDER power supply (PS) and auxiliary plant systems is being proceeding. Tests at full power and remote control are planned, including also those addressed to reproduce the grid breakdowns and to test the relevant protections. The design of the MITICA injector components was completed in 2015their procurement is being made through a number of tenders, some of them already launched. The HV Power Supply system of the MITICA 1MV accelerator, provided by JADA, was delivered on site in December 2015. The challenging installation of these components, including the step-up transformers and the SF6 gas insulated transmission line, started soon after and will go on throughout 2016. The present phase, with the PRIMA buildings continuously filled with new components, with the installation activities progressing and with also the commissioning and testing phase starting represents a substantial step forward toward the main target. The paper will describe the main challenges the Project Team has dealt during this phase and the important feedback derived for the ITER HNB systems both from the technical and the organizational standpoints.

The Indian test facility (INTF) is a R&D facility under development at the Institute for Plasma Research, Gandhinagar, as part of the neutral beam development from negative ion source program. The major advantage of the INTF, besides developing beams from large ion sources, is to characterize and benchmark the ITER diagnostic neutral beam (DNB) to the desired specifications over transport lengths of ~21 m, a unique feature of this test bed. Such a study will establish the expected power to be delivered by DNB into the ITER plasma, an important parameter to estimate the S/N ratio expected from the He ash measurements by charge exchange recombination spectroscopy. It may be noted that DNB is a procurement package for India as an in-kind contribution to ITER. Extensive physics and thermomechanical calculations have been performed to finalize the component design and layout. Adequate choice of materials, and manufacturing and jointing processes compatible to ITER safety standards have been made in order to make the components adhere to the safety and quality classification, thereby ensuring that the components survive the ITER lifetime while operating in harsh nuclear environments. The components are currently in various phases of manufacturing, and the first operations of the INTF are anticipated in Q4 of 2019.

The neutral beam (NB) system for ITER is composed of two heating neutral beam injectors (HNBs) and a diagnostic neutral beam injector (DNB). A third HNB can be installed as a future up-grade. This paper will present the design development of the components between the injectors and the tokamak; the so-called ‘front end components’: the drift duct consists of the NB bellows and the drift duct liner, the vacuum vessel pressure suppression system box (VVPSS box), the absolute valve, and the fast shutter. These components represent the key links between the ITER tokamak and the vessels of the NB injectors. The design of these components is demanding due to the different loads that these components will have to stand. The paper will describe the different design solutions which have to be implemented regarding the primary vacuum confinement, the power handling capability and the remote maintenance operations. The sizes of the components are determined by the large cross section of the neutral beam. The power handling capability is driven by the anticipated re-ionization of the neutral beam and the electromagnetic fields in this region. The drift duct bellows (with an inner diameter of 2.5 m) shall guarantee a leak tight vacuum enclosure during the vertical and radial displacements of the ITER vacuum vessel. The conductance of the VVPSS box must be maximized in the available space. The absolute valve remains a challenging development. The total leak rate through the valve must be ≤1 × 10−8 Pa m3/s when the valve is closed. Due to the radiation environment, the seals of the gate valve will be metallic. An R&D program has been launched to develop a suitable metallic seal solution with the required dimensions. The maximum allowed closing time for the fast shutter shall be less than 1 s. For all these components the leak tightness will be guaranteed by a welded lip seal and the mechanical stability by bolted structures.

CEA has undertaken tests to study the resilience of copper electrodes in vacuum against energetic high-voltage breakdowns using external capacitors to provide the energy. Earlier tests succeeded in dissipating a maximum of 150 J in a 30 mm gap, limited by the equivalent series resistance (ESR) in the external capacitors. Using new ones with an ESR that is a factor of 10 lower it was unsuccessfully tried to produce breakdowns at 200 kV over the 85 mm gap, despite the use of a UV flash lamp and a “field enhancement ring” (FER) that locally increased the electric field on the cathode by 50%. Consequently, the breakdowns had to be produced by raising the voltage to 300–350 kV while maintaining the gap at 85 mm. During these tests, single breakdowns dissipated up to 1140 J in the 85 mm vacuum gap. Inspection of the electrodes revealed that substantial amounts of copper appear have been evaporated from the anode and deposited on to the cathode. Also electrode deconditioning occurred.

The 100 kV negative hydrogen ion source based Diagnostic Neutral Beam (DNB) injector is part of the Indian procurement for ITER. The DNB is designed to deliver an 18–20 A hydrogen neutral beam to the ITER plasma. The exit scraper (ES) defines the profile of the beam after it leaves the calorimeter, the last Beam Line Components (BLCs) during operation. BTR 1 and PDP 1 codes are used to obtain the optimum entry and exit opening dimensions of the ES, so as to deliver the maximum power to the tokamak plasma, and to limit the power to downstream components due to beam interception, to protect front end components from the interception of re-ionized devious particles. Each horizontal heat transfer (HT) panel of ES receives a total power of 85 kW for each symmetric 10 mrad beam (worst case scenario) with peak power density of 1.17 MW/m 2 . The thermo hydraulic design is carried out to withstand the heat flux due to the beam interception on the ES. The mechanical design of the ES is carried out by considering the spatial constraints in the DNB system and the remote handling (RH) system. In the present design, the copper heat transfer panels are directly bolted to the structural member. Protection shall be incorporated to ensure that there is no exposure of the bolt head. The thermo-mechanical analysis has been carried out for the normal, off-normal (accidental/worst-case) events of the ITER DNB. During normal operation, the maximum possible von Mises stress of the order of 62.4 MPa can be expected on the bottom heat transfer panel, and the corresponding deflection of the structure is of the order of 0.62 mm. All the stress checks were carried out with respect to the Structural Design Criteria for In-vessel Components (SDC-IC) criterion. The paper shall present the ES design highlighting its salient features.

SPIDER is one of the two projects of the ITER Neutral Beam Test Facility (NBTF) under construction in Padova, Italy, at Consorzio RFX premises; it is the full-size prototype of the radiofrequency (RF) ion source for ITER Neutral Beam Injector (NBI), with beam energy of 100keV, designed to operate with pulse length up to 3600 s, featuring ITER-like filter field configuration, caesium oven layout and a wide set of diagnostics. These features will allow reproducing the ion source operation like in ITER as cannot be done in any other existing test facility. SPIDER realization is well advanced and the first operation is expected at the beginning of 2018, with the mission to prove the possibility of achieving the ITER NBI ion source target requirements and of improving its performance in terms of reliability and availability. This paper mainly focuses on the preparation of the first SPIDER operations: integration and testing of SPIDER components, completion and implementation of diagnostics and control and workout of operation and research plan, based on a staged strategy.

The NB ITER Remote Maintenance System (NB IRMS) provides the means for the remote maintenance within the NB Cell by removal and replacement of the plant equipment. The NB IRMS will be installed and removed with the assistance of human workers during the preparation, and post-operation phase. During the maintenance operation after opening the Passive Magnetic Shield (PMS) and vessels, the maintenance activity and recovery from failure should be conducted remotely. This paper describes the concept design of the NB IRMS operating inside the NB cell for maintenance of the plant equipment such as NB components, and Upper Port Plugs (UPP). The main tasks of the IRMS, the description of the sub-systems and their specification, and deployment/operation principles are presented. The transportation concept of the NB IRMS to the hot cell facility for storage and maintenance is presented, which is to avoid unnecessary exposure on the equipment inside the NB cell during the machine operation.

The ITER Neutral Beam (NB) System consists of a 100 keV diagnostic injector (DNB) and two 1 MeV heating and current drive injectors (HNB). Injection will be done in the different phases of the experimental programme, ranging from low activation operation with hydrogen and helium plasmas, deuterium plasmas, and finally full performance operation with deuterium–tritium mixtures. The injectors will be operated in a harsh environment with high levels of neutron and gamma radiation from the torus. A high stray magnetic field from the tokamak is reduced to acceptable levels for the region up-stream of the residual ion dump (RID) by a combination of a passive magnetic shield and active correction and compensation coils (ACCC). This broad range of requirements imposes a high flexibility, a high availability and reliability of the instrumentation and on the diagnostics that will be used.

The 100 kV bushing of the Diagnostic Neutral Beam (DNB) injector is a cylindrical feedthrough which forms the interface between the gas insulated transmission line and the torus primary vacuum and provides all necessary services to the beam source. All conventions for safety, voltage holding requirement, vacuum compatibility and the choice of materials have been addressed in the design. Finite Element Analyses (FEAs) for the electrostatic and the structural configuration is carried out to validate the design of the High Voltage Bushing (HVB). Several iterations and optimizations of the stress shields are carried out to meet electrostatic criteria, especially at the triple point (the ceramic, metal and vacuum joint), which is critical for good voltage holding. Structural analyses is carried out to assess the stress distribution in the fiber reinforced plastic insulator, the alumina insulator and the integrated HVB for different load cases like operational orientation (horizontal), normal operation and accidental case. Design is further validated for seismic conditions for Seismic Loading-2 (SL-2).

The 100-kV negative-hydrogen-ion-source-based diagnostic neutral beam (NB) (DNB) injector, which forms a part of the Indian (IN) procurement package for ITER, targets a delivery of ~18-20 A of neutral hydrogen-atom beam current into the ITER torus for charge exchange resonance spectroscopy diagnostics. Considering stripping losses, a ~70-A negative ion current is required to be extracted from the ion source, which leads to a production of 60 A of accelerated ion beam. Subsequent process of neutralization, electrostatic ion separation, and transport to the duct leads to a large separation between the points of generation of the ion beam to the point of delivery of the NB into the torus (~23 m). This forms one of the most important constraints for the transport of NBs to ITER. The requirements are not only for a stringent control over ion optics, the transport to electrostatic separator, minimum loss of beam due to intercepting elements, low reionization loss, and focusing to control interception losses but also for adequate compensation of residual magnetic fields to overcome magnetic field induced deflections also form important design issues for a reasonable transmission efficiency. Due to multiparameter dependence, it becomes necessary to assess the different scenarios using numerical codes. In the present case, the assessment has been carried out for the DNB using the beam-transport codes PDP, BTR, and the MCGF codes which are developed by the Russian Federation. An optimized configuration of the beamline has been arrived at on the basis of these code-enabled studies. These parameters are the following: listing of the vertical and horizontal focal lengths as 20.6 m, a spacing between ground grid and neutralizer of 1 m, and positioning of residual-ion dump at a distance of 0.75 m from the neutralizer exit. Further, optimizing the gas feed to the source and neutralizer leads to a final transmission of ~35% of the extracted beam power to the torus. This paper shall present the methodology, the issues concerned, and the final configuration which forms the basis for the present engineering.

In this paper the results of post-mortem analysis of mirror samples (molybdenum, stainless steel and copper) exposed for the 1 year experimental campaign inside the Tore Supra tokamak are presented. The mirrors were not protected during the experiment and therefore were exposed both to glow discharge conditioning of the vessel walls and to working discharges. After exposure, all samples were found to be eroded and at the same time were coated with a thin contaminating deposit. We pay attention to the fact, that the observed erosion cannot be described using the published sputtering yields. In particular, the difference between the erosion of stainless steel and copper exceeded a factor of 10 and the difference between the erosion of stainless steel and molybdenum was less than a factor of 2 in comparison with the corresponding factors of approximately 2.5 and apprximately 10 known from the literature. An attempt is made to correlate these observations with the different sticking coefficients of carbon on ...

Metal mirrors will be implemented in about half of the ITER diagnostics. Mirrors in ITER will have to withstand radiation loads, erosion by charge-exchange neutrals, deposition of impurities, particle implantation and neutron irradiation. It is believed that the optical properties of diagnostic mirrors will be primarily influenced by erosion and deposition. A solution is needed for optimal performance of mirrors in ITER throughout the entire lifetime of the machine. A multi-machine research on diagnostic mirrors is currently underway in fusion facilities at several institutions and laboratories worldwide. Among others, dedicated investigations of ITER-candidate mirror materials are ongoing in Tore-Supra, TEXTOR, DIII-D, TCV, T-10 and JET. Laboratory studies are underway at IPP Kharkov (Ukraine), Kurchatov Institute (Russia) and the University of Basel (Switzerland). An overview of current research on diagnostic mirrors along with an outlook on future investigations is the subject of this paper.

The early phase of a tokamak plasma discharge can have a dramatic impact on the main heating phase. This has been a persistent problem for the development of the steady state, fully non-inductive scenario using lower hybrid current drive (LHCD) on Tore Supra. The present paper reports on recent experimental and numerical investigations showing that a tearing mode coupled to the internal kink grows on q = 2 in the ohmic phase when the total current is too low, due to the weakening of field line curvature stabilization. Then, the application of LHCD drives the island to a larger size and undermines the development of the non-inductive phase. Decreasing the edge safety factor or increasing the Lundquist number S is found to be beneficial in both the linear and non-linear MHD analyses. The experimental database, which allows covering the edge safety factor dependence, supports this interpretation.

The SPIDER facility is undergoing installation & commissioning in the Neutral Beam Test facility at RFX, Padova. SPIDER Beam Dump (SBD) is mandated to diagnose calorimetrically an ion beam of power up to 6 MW accelerated from the SPIDER Beam Source. SBD utilizes Hypervapotron based Heat Transfer Elements (HTEs) as cooling elements. HTEs are made of Cu-Cr-Zr alloy, designed for removing a steady state heat flux of 10MW/m2. 62 HTEs are arranged in two halves, mounted on a support structure ensuring the required alignment. Diagnostics provisions include independent calorimetric measurement of the coolant temperature at the exit of each of the HTEs and multiple thermocouples embedded in the edges of the HTEs. SBD is designed w. r. t ITER Structural Design Criteria for In Vessel Components (SDC-IC) and is manufactured to meet ITER codes & standards requirements and specific requirements related to vacuum components as specified by ITER Vacuum Handbook. Several challenges experienced in the manufacturing include: i) Production of CuCrZr alloy with specific mechanical properties in aged condition, ii) EB welding of dis-similar material combination like CuCrZr - Ni, Ni -SS 316L and iii) Joining CuCrZr to CuCrZr in lap joint configuration, where an acceptance basis for the allowable extent of ground pores in the lap configuration of EB weld seam has been cast. The complete execution of this activity has conformed to ITER quality requirements. The procurement has been managed by ITER-India and has completed the manufacturing, factory inspection and has been delivered as the first mechanical equipment for the SPIDER facility. This paper will present an overview of the experience gained during SPIDER Beam Dump manufacturing including the identification of technical issues and their resolution and provide important inputs to the technical management of production of similar mechanical components for ITER requirements.

The heating and current drive beam lines (HNB) at ITER are expected to deliver ∼16.7 MW power per beam line for H beams at 870 keV and D beams at 1 MeV during the H-He and the DD/DT phases of ITER operation respectively. On the other hand the diagnostic neutral beam (DNB) line shall deliver ∼2 MW power for H beams at 100 keV during both the phases. The path lengths over which the beams from the HNB and DNB beam lines need to be transported are 25.6 m and 20.7 m respectively. The transport of the beams over these path lengths results in beam losses, mainly by the direct interception of the beam with the beam line components and reionisation. The lost power is deposited on the surfaces of the various components of the beam line. In order to ensure the survival of these components over the operational life time of ITER, it is important to determine to the best possible extent the operational power loads and power densities on the various surfaces which are impacted by the beam in one way or the other during ...

Steady state fully non-inductive current plasmas in Tore Supra offer an opportunity to study the turbulent particle pinch. The local parametric dependence of the particle pinch velocity allows us to discriminate between different theories. The existence of an anomalous pinch is unambiguously demonstrated. The results support the turbulent theories based on ion temperature gradient and trapped electron modes. An anomalous inward pinch related to the safety factor (q) profile is found to be dominant in the gradient region (normalized radius 0.3 ≤ r/a ≤ 0.6). In contrast, the direction of the anomalous pinch in the plasma core (r/a ≤ 0.3) is more sensitive to the electron temperature (Te) profile.

Lower hybrid current drive experiments in the ergodic divertor (ED) configuration have been carried out on Tore Supra for 9 years (1991-1999). This paper gives an overview of the results regarding both the current drive and the divertor efficiencies: waves coupling, current drive efficiency and fast particle confinement, energy confinement, thermal load on the ED plasma facing components, impurity screening. It is shown that the coupling remains good enough for high power transmission up to 26 MW m -2 . The current drive efficiency, measured on not fully non-inductive discharges, is only marginally affected by the loss of fast electrons due to the perturbed magnetic structure which is documented from hard x-ray measurements. Electron confinement follows, as in the limiter case, the Rebut-Lallia-Watkins scaling without any losses related to the width of the ergodic layer extending to 15% of the minor radius. The heat flux on the divertor neutralizers from both thermal particles and fast electrons accelerated near the antennae is detailed. Plasma contamination due to carbon and oxygen is analysed from a large database: it is shown that, for a large enough gap between the ED modules and the antennae, the contamination by these species is reduced by a factor of ∼3 with respect to the limiter case.

A spectroscopic study of VUV emission from injected and intrinsic low-Z impurities has been carried out for Tore Supra ergodic divertor (ED) plasmas. Analysis of plasmas in which a nearby limiter effectively provides a spatially localized source of recycled impurities provides information illuminating the dynamical processes of impurity penetration in ED plasmas. The profile evolution behaviour is found to be consistent with an interpretation which identifies both ED-influenced edge confinement and scrape-off layer and edge phenomena provided by the localized source. Preferentially increased edge impurity transport is a beneficial aspect of the ED.

The formation of a core region with improved electron confinement is reported in the recent full current drive operation of Tore Supra, where the plasma current is sustained with the lower hybrid (LH) wave. Current profile evolution and thermal electron transport coefficients are assessed directly by using the data of the new fast electron bremsstrahlung tomography that provides the most accurate determination of the LH current and power deposition profiles. The spontaneous rise of the core electron temperature observed a few seconds after the application of the LH power is ascribed to a bifurcation towards a state of reduced electron transport. The role of the magnetic shear is invoked to partly stabilize the anomalous electron turbulence. The electron temperature transition occurs when the q-profile evolves towards a non-inductive state with a non-monotonic shape, i.e. when the magnetic shear in the plasma core is reduced to close to zero. The improved core confinement phase is often terminated by a sudden MHD activity when the minimum q approaches 2.

We report experiments aimed at comparing several impurity mixtures (C, O, Cl, N, Ne, Ar) regarding their capability to reduce the power load on the divertor target plates. The divertor conditions required for each mixture to minimise the parallel power flux are determined, along with the resulting core effective charge Z eff and volume averaged density. The radiation efficiency (ratio of edge radiation to plasma core contamination) of intrinsic carbon is found to increase with the total injected power. In the impurity injection experiments, nitrogen is found to be the best choice to reduce the power flux to the target plates: it has the same characteristics as C/O radiation (low core contamination), and it can be controlled. The low Z eff observed in this case is attributed to the large value of the screening of the radiating ionisation stages of the impurity.

Experimental evidence of the location of the ergodic divertor separatrix is shown to agree with the predicted value given by codes. Variation of this position modifies the divertor tightness, defined as the ratio of the divertor to core density. This effect is governed by laminar transport, i.e., transport proportional to the magnitude of the perturbation. Operation with feedback control of the divertor temperature allows one to optimise the choice of injected impurity species. At 10 eV divertor temperature, nitrogen is shown to lead to the largest decrease in energy flux to the divertor at lowest contribution to Zeff. Parallel energy fluxes as low as 2 MW m −2 are thus achieved on the target plates. For this impurity, radiation is localised in the divertor volume thus leading to radiation compression close to 10. The ergodic divertor appears as a powerful tool to control plasma–wall interaction with no loss of core confinement or plasma current.

On Tore Supra, the mean effective charge Z eff of the plasma is deduced routinely from line integrated measurements of visible Bremsstrahlung. In order to get a deeper understanding of this measurement, and to obtain high confidence in the data, we have compared the visible Bremsstrahlung data to line radiation of impurities present in the plasma, removing any error from the processing technique. The paper presents a systematic study of the correlation between the Z eff derived from the two methods, and the dependence on plasma parameters and geometrical factors. A good correlation is found, independent of the position of the outboard pump limiter (OPL). However, the fitting parameter of the empirical method is different for different plasma currents, due to the variation of the emission profiles. The agreement between the two methods is good, confirming the role of carbon and oxygen as the main impurities in the Tore Supra tokamak.

Some models for the suppression of turbulence in the L to H transition, suggest that the width of the H-mode edge barrier is either proportional or is of the order of the thermal or the fast-ion poloidal Larmor radius. This would require that the width of the edge barrier should depend on the plasma current. This dependence has been clearly verified at JET in experiments designed to control the edge MHD stability of ELM-free hot-ion H-mode plasmas. The effects of isotopic mass and the applicability of several edge barrier models to the hot-ion H-mode plasmas were analysed in (Guo H Y et al 2000 Edge transport barrier in JET hot-ion H-modes Nucl. Fusion 40 69) using a large database containing both deuterium-only and deuterium-tritium plasmas. This database has now been enlarged to include discharges from a plasma shape scan, allowing one to study the dependence of the pedestal height on the edge shear. In addition, the range of plasma currents was extended up to 6 MA. It is shown that the edge data are best described by a model where the edge barrier width is determined by the fast ions weighted towards the components with largest poloidal Larmor radii. However, it is not possible to conclusively eliminate the thermal ion model.

During 1997 JET operation with D-T plasmas, 35 g of tritium were introduced into the torus, mainly by gas puffing. It was found that during this period, the torus tritium inventory would accumulate at a rate of about 40% of the input. After tritium operation ceased, the experimental program continued with deuterium- and hydrogen-fuelled experiments, during which time the tritium inventory decreased to about 17% of the total input. Techniques aimed at detritiation of the torus included methods using deuterium gas (such as deuterium pulsing) which were used in the middle of the experimental campaign, and methods which could adversely affect the torus vacuum conditions (such as air purges) which were reserved for the period after the experimental campaign. Whilst it was found that the plasma tritium fraction could be reduced to below the 1% level in a few days, the tritium inventory reached a virtually steady level of about 6 g by the end of the campaign.

The present paper addresses the issue of particle recirculation in discharges where low-energy flux to ergodic divertor target plates is achieved in highly-radiating detached ohmic plasmas. Plasma temperature and particle flux are measured by flush-mounted probes in the divertor plates and by an upstream fast scanning Mach probe. The scalings with core density of the ion flux and electron temperature are well described by the simple two-point model used in axisymmetric poloidal divertors. The detachment signature is a pressure drop that occurs when the edge temperature falls below 10 eV. The parallel ion flux gradient is always positive, indicating that recombination is unlikely to play an important role in detachment. Visible spectroscopy of a neutralizer plate shows that attainment of cold detached plasmas near the density limit coincides with an abrupt increase of fuelling efficiency for both deuterium and impurities. A feedback algorithm based on real-time Langmuir probe measurements has been developed to monitor detachment and avoid disruptions.

Discharge optimization for improving MHD stability of both core and edge was essential for the achievement of record fusion power discharges, in the ELM-free hot ion H mode regime, in the recent JET DT operation. The techniques used to increase edge stability are described. In particular the successful technique of current rampdown used to suppress the outer mode is reported. The increased stability of the outer mode by decreasing the edge current density confirms its identification as an n = 1 external kink. Decreasing the plasma current, however, decreases the ELM-free period, which is consistent with stability calculations that show an earlier onset of the ballooning limit. In order to increase external kink stability without causing a deterioration in the ELM-free period, a compromise was achieved by using plasma current rampdown, while working at the highest plasma current values possible. Results from a plasma current scan show that at the time of occurrence of the first giant ELM, the plasma stored energy, as well as the pressure measured at the top of the edge pedestal increase linearly with plasma current, for a given plasma configuration and power. This is consistent with models of the edge transport barrier, where the transport barrier width is proportional to the ion (or fast ion) poloidal Larmor radius. The MHD observations in DT and deuterium only discharges were found to be similar. Thus the experience gained on the control of MHD modes in deuterium plasmas could be fully exploited in the DT campaign.

In toroidal magnetic confinement configurations all the charged particles align their magnetic drifts with the flux surfaces if, and only if, the lines of force have zero geodesic curvature (omnigeneous configuration). In this condition, the neoclassical transport is zero; unfortunately, standard tokamak devices are far from meeting such a condition. Nevertheless, a minimization of the geodesic curvature of the lines of force in tokamaks minimizes the neoclassical transport, and could as well reduce any other transport terms depending on the noncoincidence between magnetic and drift surfaces. In particular, the geodesic curvature of the lines of force generates both the neoclassical Pfirsch–Schluter current density j∥PS, as well as Hirshman factor q [Nucl. Fusion 18, 917 (1978)], which increases the moment of inertia of the magnetic configuration. The improved confinement observed in the VH-mode (very high confinement) and in reversed magnetic shear discharges of the DIII-D tokamak [Plasma Physics Contro...

The scaling of the energy confinement in H mode plasmas with different hydrogenic isotopes (hydrogen, deuterium, DT and tritium) is investigated in JET. For ELM-free H modes the thermal energy confinement time τth is found to decrease weakly with the isotope mass (τth ~M-0.25±0.22), whilst in ELMy H modes the energy confinement time shows practically no mass dependence (τth ~M0.03±0.1). Detailed local transport analysis of the ELMy H mode plasmas reveals that the confinement in the edge region increases strongly with the isotope mass, whereas the confinement in the core region decreases with mass (τthcore ∝ M-0.16), in approximate agreement with theoretical models of the gyro-Bohm type (τgB ~M-0.2).

A number of experiments with heating of DT plasmas using ICRF waves have been carried out at JET. The results of these experiments have been analysed by comparing experimentally measured quantities with the results of numerical simulations. In particular, four scenarios have been examined: (a) heating of minority (~5-20%) deuterons at the fundamental ion cyclotron frequency, ω = ωcD; (b) second harmonic heating of tritium, ω = 2ωcT; (c) fundamental minority heating of 3He with a few per cent of 3He; (d) second harmonic heating of deuterium, ω = 2ωcD. An important aim of the analysis was to assess whether the present understanding of the ICRF physics is adequate for predicting the performance of ICRF in DT plasmas. In general, good agreement between experimental results and simulations was found which increases the confidence in predictions of the impact of ICRF heating in future reactors. However, when a relatively high deuterium concentration was used in the ω = ωcD scenario, discrepancies were observed. In order to increase confidence in the simulations, the sensitivity of the simulation results to various plasma parameters has been studied.

Results are presented from a series of dedicated experiments carried out on JET in tritium, DT, deuterium and hydrogen plasmas to determine the dependence of the H mode power threshold on the plasma isotopic mass. The Pthr ∝ Aeff-1 scaling is established over the whole isotopic range. This result makes it possible for a fusion reactor with a 50:50 DT mixture to access the H mode regime with about 20% less power than that needed in a DD mixture. Results on the first systematic measurements of the power necessary for the transition of the plasma to the type I ELM regime, which occurs after the transition to H mode, are also in agreement with the Aeff-1 scaling. For a subset of discharges, measurements of Te and Ti at the top of the profile pedestal have been obtained, indicating a weak influence of the isotopic mass on the critical edge temperature thought to be necessary for the H mode transition.

Reactor relevant ICRH scenarios have been assessed during DT experiments on the JET tokamak using H mode divertor discharges with ITER-like shapes and safety factors. Deuterium minority heating in tritium plasmas was demonstrated for the first time. For 9% deuterium, an ICRH power of 6 MW gave 1.66 MW of fusion power from reactions between suprathermal deuterons and thermal tritons. The Q value of the steady state discharge reached 0.22 for the length of the RF flat-top (2.7 s), corresponding to three plasma energy replacement times. The Doppler broadened neutron spectrum showed a deuteron energy of 125 keV, which was optimum for fusion and close to the critical energy. Thus, strong bulk ion heating was obtained at the same time as high fusion efficiency. Deuterium fractions around 20% produced the strongest ion heating together with a strong reduction of the suprathermal deuteron tail. The ELMs had low amplitude and high frequency and each ELM transported less plasma energy content than the 1% required by ITER. The energy confinement time, on the ITERH97-P scale, was 0.90, which is sufficient for ignition in ITER. 3He minority heating, in approximately 50:50 D:T plasmas with up to 10% 3He, also demonstrated strong bulk ion heating. Central ion temperatures up to 13 keV were achieved, together with central electron temperatures up to 12 keV. The normalized H mode confinement time was 0.95. Second harmonic tritium heating produced energetic tritons above the critical energy. This scheme heats the electrons in JET, unlike in ITER where the lower power density will allow mainly ion heating. The inverted scenario of tritium minority ICRH in a deuterium plasma was demonstrated as a successful heating method producing both suprathermal neutrons and bulk ion heating. Theoretical calculations of the DT reactivity mostly give excellent agreement with the measured reaction rates.

JET's 1997 Deuterium Tritium Experiment (DTE1) allows a detailed study of hydrogenic isotope recycling and retention in a pumped divertor configuration relevant to ITER. There appear to be two distinct forms of retained tritium. (1) A dynamic inventory which controls the fueling behaviour of a single discharge, and in particular determines the isotopic composition. This is shown to be consistent with neutral particle implantation over the whole vessel surface area. (2) A continually growing inventory, which plays a small role in the particle balance of a single discharge, but ultimately dominates the hydrogenic inventory for an experimental campaign comprising thousands of pulses. This will be the dominant retention mechanism in long-pulse devices like ITER. The JET retention scaled-up to ITER proportions suggests that ITER may reach its tritium inventory limit in less than 100 pulses.

On the basis of experiments carried out in JET in D:T mixtures varying from 100:0 to 5:95 and those carried out in hydrogen plasmas, the isotopic mass dependence of ELM parameters and the edge pedestal pressure in NBI and ICRF heated H mode plasmas is presented. The ELM frequency is found to decrease with the atomic mass number in both ICRH and NBI discharges. However, the frequency in the case of ICRH is about 8-10 times higher than that in the case of NBI. Assuming that ELMs occur at a critical edge pressure gradient, limited by the ballooning instability, the scaling of the maximum edge pressure is most consistent with the assumption that the width of the transport barrier scales as the ion poloidal Larmor radius governed by the average energy of fast ions at the edge. The critical edge pressure in NBI heated discharges increases with the isotopic mass, which is consistent with the higher deduced width of the edge transport barrier in tritium than those in deuterium and hydrogen. The critical edge pressure in ICRH discharges is smaller, presumably, due to the smaller fast ion contribution to the edge region. As a consequence of the edge pressure scaling with isotopic mass, the edge operational space in an ne-Te diagram increases with operation in tritium. If the evidence that the edge pedestal width is governed by the average energy of fast ions in the edge prevails, the pedestal in ITER would be controlled by the slowing down energy spectrum of alpha particles in the edge.

An overview of JET experimental results in DT plasmas directly relevant to ITER modes of operation is presented. Experiments in D:T mixtures varying from 100:0 to 10:90 and those carried out in hydrogen plasmas show that the H mode threshold power has an approximately inverse isotope mass dependence. Matching some of the key dimensionless parameters to the ITER values, the ITER similarity experiments with ITER shape and safety factor q show that the global energy confinement time is practically independent of isotopic mass (~A0.03±0.08), where A is the atomic mass of the hydrogenic species. Subtracting the edge pedestal energy (which scales as ~A0.57±0.2) from the total stored energy leads to a ~A-0.17±0.1 dependence of confinement in the plasma core, very similar to that expected from the gyro-Bohm transport (~ A-0.2) model. The observed scaling of the edge pedestal energy is consistent with a model in which the edge pressure gradient saturates at the ballooning limit over a region of width that scales as the ion poloidal Larmor radius governed by the average energy of the fast ions in the edge. The steady state total stored energy for a given input power in both ICRH and NBI discharges is the same despite the lower edge pedestal in the ICRH case, which is compensated for by more peaked power deposition profiles in ICRH. The ELM frequency is smaller with NBI; it decreases with isotopic mass in both NBI and ICRH discharges. A steady state, type I ELMy H mode discharge with ITER shape and q at 3.8 T/3.8 MA with an input power of 22 MW produced a Q ≈ 0.18 for 3.5 s and extrapolates well to ignition with ITER parameters. Here, Q is the ratio of fusion output power to input power. The thermal ELMy H mode confinement in both deuterium and tritium gas fuelled plasmas decreases significantly when the plasma density exceeds 0.75 of the Greenwald (nGW) limit, and the maximum density achieved is 0.85nGW. In L mode, the density limit decreases with increasing isotope mass roughly in accordance with code predictions. ITER reference ICRH scenarios have been evaluated. Second harmonic heating of tritium at the densities available in JET produces strong tails and heats electrons predominantly as expected. The 3He minority in 50:50 D:T and tritium dominated plasmas showed strong bulk ion heating leading to ion temperatures up to 13 keV with ICRH alone. Deuterium minority ion cyclotron heating in tritium plasmas at a power level of 6 MW produced steady state record values of Q ≈ 0.22 for more than 2.5 s.

Ion cyclotron resonance heating (ICRH) experiments have been carried out in JET D-T plasmas using scenarios applicable to reactors. Deuterium minority heating in tritium plasmas is used for the first time and produces 1.66 MW of D-T fusion power for an ICRH power of 6 MW. The Q value is 0.22, which is a record for steady state discharges. Fundamental He-3 minority ICRH, in both 50:50 D-T and tritium dominated plasmas, generates strong bulk ion heating and ion temperatures up to 13 keV. Second harmonic tritium ICRH is seen to heat mainly the electrons as expected for JET conditions. All three schemes produce H-mode plasmas.

All major JET systems have been fully commissioned for D-T and the DTE1 series of experiments has started with the D-T fuel mixture and operating conditions foreseen for ITER. In the area of ITER physics, significant results have been produced in both D-D and D-T. In D-D, the LH threshold power database has been extended, the bounds on edge-electron temperature and density in ELMy H-modes have been defined and the advantages of Types I and III ELMy discharges have been compared. In D-T plasmas, the isotope effect on H-mode threshold power and transport has been determined so that a more accurate assessment can be made of the ignition margin and heating requirements for ITER. Trace tritium experiments have provided first particle transport measurements and an assessment of the ITER reference ion-cyclotron resonance-frequency heating scenarios has been started, In the area of fusion performance, record D-D neutron yields have been obtained by controlling the plasma and current profiles in hot ion ELM-free H-modes and optimized shear modes. In D-T, internal transport barriers have been readily established in optimized shear discharges and Alfven eigenmodes have been observed.

In nearly all hot ion H modes in JET, a magnetohydrodynamic (MHD) event is clearly observed just before the time the stored plasma energy saturates and the neutron yield starts to decline. The results of a systematic analysis of MHD observations for a large number of discharges is reported. The relationship between MHD phenomena and the onset of confinement limitation is discussed, as are aspects of the three main types of performance limiting MHD: (a) low-n modes in the outer regions of the plasma, (b) sawteeth and (c) giant edge localized modes (ELMs). Model simulations indicate how the transport is affected and allow an assessment of how much the neutron yield would be improved if the MHD activity were absent

Calorimeter for Diagnostic Neutral Beam (DNB) consists of two movable panels, forming V-shaped configuration with an included angle of 45 Deg. w.r.t. neutral beam axis. Panel's closing and opening motion will be achieved through Calorimeter Motion Mechanism (CMM). The DNB calorimeter design has evolved after analyzing several design proposals serving the basic functionality of the component, but these designs had compatibility issues for the ITER Remote Handling (RH) scenarios, operational maintenance, ITER Vacuum Handbook (IVH) compliance for hydraulic integration, design of flexible elements for large displacements, neutron irradiation susceptibility and joint materials for Vacuum Quality Class 1 (VQC1) application. In the design concept the movement mechanism and hydraulic line flexible element have been deliberately kept outside the DNB vessel, hence there is no sliding and complicated links inside the vessel which cannot be maintained through available RH tools. The paper proposes to present state of the art design of the ITER-DNB Calorimeter.

Positioning of Cesium (Cs) oven modules in the complex interface dominated space envelope of a negative ion source such as Diagnostic Neutral Beam (DNB) source for ITER is a challenge not only for the designer of the ion source, but also that of remote handling. A more user friendly design of the Cs delivery could emerge from the consideration of a possibility of injecting the Cs from an oven located outside the vacuum envelope of the ion source, thereby ensuring an ease of Cs refilling and oven maintenance. The design of such a delivery system involves long transmission path of lengths ∼4 m, from ambient to vacuum. System design involves incorporation of a low loss transmission tube enveloped by highly reflective inner surface pipe to reduce the heat losses and therefore heating of the nearby systems. A combination of all metallic valves operated at high temperatures has been incorporated in such a way that the Cs refilling or oven maintenance can be done without breaking the ion source vacuum. Removable joints in the oven heating elements are provided at specific locations to remove the Cs oven for ion source maintenance. Experimental data on Cs transmission over such a long length, required for an effective design of a co-axial transmission, is not presently available. However, an experiment has been carried out in ITER-India making measurements of Cs distribution in coaxial transmission of a length of more than 5 m. These experiments incorporate an additional feature of multiple nozzle distributor based Cs delivery into the ion source which might help in reducing the need of multiple Cs ovens in large ion sources like ITER. The Cs flux from the oven is measured by surface ionization detector (SID). The angular distribution of the Cs flux is measured by a movable SID in linear direction and has been found in good agreement with the calculations. The Cs inventory in the oven reservoir was measured by electrical resistivity measurements methods. The paper proposes to present the measurement results and also proposes a possible configuration of the Cs oven for ITER DNB ion source.

A steady-state ELMy H-mode profile data set of 68 DIII-D discharges and 74 JET discharges is fitted with an error of 7 - 8%. The advantages of a parametrization of the plasma profiles in terms of a semi-parametric representation, , are described. The shape of the temperature profile depends almost exclusively upon the size, R and , with a secondary dependence on the heating power. The density profile depends primarily upon with a secondary dependence on . The line-average temperature scales as instead of . The predicted ITER temperature is keV.

The motional Stark effect (MSE) diagnostic on the International Thermonuclear Experimental Reactor will need to guide the light through a labyrinth of mirrors to provide neutron shielding. Knowledge of how the mirrors change the polarization is essential for accurate determination of the q profile. The optical properties of the plasma facing mirror are also expected to change with time due to deposition/erosion. For the purpose of examining this experimentally a detector system, identical to the JET MSE system, using twin photoelastic modulators was constructed. Measurements have been performed on freshly prepared mirrors, on mirrors after exposure to plasmas in Tore Supra, and labyrinth designs. The result shows a significant effect on the optical properties and demonstrate the need for in situ monitoring. The measured properties of the labyrinth closely follow the Mueller matrix formalism. With a correct choice of material the angle change introduced by the four mirrors furthest away from the plasma wil...

Diagnostic systems for fusion reactors will require mirrors capable of reflecting electromagnetic radiation in the soft X-ray, near UV, visible, and IR wavelengths. These components will be exposed to significant fluences of fast neutrons during use. Mirrors made from alternating layers of low-Z and high-Z materials were irradiated to a fast neutron fluence of 1.1 × 1023 n/m2 at 270–300°C, and subsequently evaluated for structural changes and in some cases changes in optical properties. Short-wavelength mirrors retained their structural integrity while exhibiting slight changes in reflectance; some long-wavelength mirrors showed structural degradation, while others did not. These results are discussed in terms of materials damage effects and possibilities for improvement of mirror performance under severe operating conditions.

The LIDAR Thomson scattering technique uses the time of flight of a short laser pulse to spatially resolve measurements of electron temperature (Te) and density (ne) in a plasma. The technique was pioneered at JET and the first profiles were obtained in 1986. The initial system used a 3 J, 0.5 Hz, 300 ps ruby laser and microchannel‐plate photomultiplier detectors to make Te and ne profile measurements with ∼10 cm spatial resolution. Since then we have sought to improve both the spatial resolution and frequency of measurement during a JET pulse. It has proved possible to develop a 4 Hz version of the ruby laser and to enhance the rate of cycling of detection and digitizing to match. The main LIDAR system has now been upgraded to 4 Hz and first JET profiles obtained with the new system are presented. For a LIDAR system, the spatial resolution δL along the laser path is given by δL=(c/2)(τ2L+τ2D)1/2, where c is the velocity of light, τL the laser pulse width, and τD the detection system response time. Over a...

The JET neutral beam injection system has proved to be both effective and reliable as a plasma heating device. The ion heating and plasma fuelling characteristics of the system are ideally suited to the production of high fusion performance plasmas while the flexibility in the choice of beam species (H, D, T, 3 He or 4 He) and the ability to inject into almost any JET plasma configuration allows a wide variety of related physics experiments to be carried out. The capability to inject (for the first time) tritium beams was essential to the successful execution of the first tritium experiments in which 1.7 MW of power from DT fusion reactions was generated.