Materials damage by 14.1MeV neutrons from deuterium-tritium (D-T) fusion reactions can only be characterised definitively by subjecting a relevant configuration of test materials to high-intensity 'fusion-neutron spectrum sources', i.e. those simulating closely D-T fusion-neutron spectra. This provides major challenges to programmes to design and construct a demonstration fusion reactor prior to having a large-scale, high-intensity source of such neutrons. In this paper, we discuss the different aspects related to these 'relevant configuration' tests, including:generic issues in materials qualification/ validation,comparing safety requirements against those of investment protection;lessons learned from the fission programme, enabling a reduced fusion materials testing programme;the use and limitations of presently available possible irradiation sources to optimise a fusion neutron testing program including fission-neutron irradiation of isotopically and chemically tailored steels, ion damage by high-energy helium ions and self-ion beams, or irradiation studies with neutron sources of non-fusion spectra; andthe different potential sources of simulated fusion neutron spectra and the choice using stripping reactions from deuterium-beam ions incident on light-element targets.

Materials determine in a fundamental way the performance and environmental attractiveness of a fusion reactor: through the size (power fluxes to the divertor, neutron fluxes to the first wall); economics (replacement lifetime of critical in-vessel components, thermodynamic efficiency through operating temperature etc); plasma performance (erosion by plasma fluxes to the divertor surfaces); robustness against off-normal accidents (safety); and the effects of post-operation radioactivity on waste disposal and maintenance. The major philosophies and methodologies used to formulate programmes for the development of fusion materials are outlined, as the basis for other articles in this special issue, which deal with the fundamental understanding of the issues regarding these materials and their technical status and prospects for development.

The findings of the EU 'Materials Assessment Group' (MAG), within the 2012 EU Fusion Roadmap exercise, are discussed. MAG analysed the technological readiness of structural, plasma facing and high heat flux materials for a DEMO concept to be constructed in the early 2030s, proposing a coherent strategy for R&D up to a DEMO construction decision. Technical consequences for the materials required and the development, testing and modelling programmes, are analysed using: a systems engineering approach, considering reactor operational cycles, efficient maintenance and inspection requirements, and interaction with functional materials/coolants; and a project-based risk analysis, with R&D to mitigate risks from material shortcomings including development of specific risk mitigation materials.

New diagnostic, modelling and plant capability on the Mega Ampère Spherical Tokamak (MAST) have delivered important results in key areas for ITER/DEMO and the upcoming MAST Upgrade, a step towards future ST devices on the path to fusion currently under procurement. Micro-stability analysis of the pedestal highlights the potential roles of micro-tearing modes and kinetic ballooning modes for the pedestal formation. Mitigation of edge localized modes (ELM) using resonant magnetic perturbation has been demonstrated for toroidal mode numbers n = 3, 4, 6 with an ELM frequency increase by up to a factor of 9, compatible with pellet fuelling. The peak heat flux of mitigated and natural ELMs follows the same linear trend with ELM energy loss and the first ELM-resolved Ti measurements in the divertor region are shown. Measurements of flow shear and turbulence dynamics during L–H transitions show filaments erupting from the plasma edge whilst the full flow shear is still present. Off-axis neutral beam injection helps to strongly reduce the redistribution of fast-ions due to fishbone modes when compared to on-axis injection. Low-k ion-scale turbulence has been measured in L-mode and compared to global gyro-kinetic simulations. A statistical analysis of principal turbulence time scales shows them to be of comparable magnitude and reasonably correlated with turbulence decorrelation time. Te inside the island of a neoclassical tearing mode allow the analysis of the island evolution without assuming specific models for the heat flux. Other results include the discrepancy of the current profile evolution during the current ramp-up with solutions of the poloidal field diffusion equation, studies of the anomalous Doppler resonance compressional Alfvén eigenmodes, disruption mitigation studies and modelling of the new divertor design for MAST Upgrade. The novel 3D electron Bernstein synthetic imaging shows promising first data sensitive to the edge current profile and flows.

Operating ITER in the reference inductive scenario at the design values of I-p = 15 MA and Q(DT) = 10 requires the achievement of good H-mode confinement that relies on the presence of an edge transport barrier whose pedestal pressure height is key to plasma performance. Strong gradients occur at the edge in such conditions that can drive magnetohydrodynamic instabilities resulting in edge localized modes (ELMs), which produce a rapid energy loss from the pedestal region to the plasma facing components (PFC). Without appropriate control, the heat loads on PFCs during ELMs in ITER are expected to become significant for operation in H-mode at I-p = 6-9 MA; operation at higher plasma currents would result in a very reduced life time of the PFCs.Currently, several options are being considered for the achievement of the required level of ELM control in ITER; this includes operation in plasma regimes which naturally have no or very small ELMs, decreasing the ELM energy loss by increasing their frequency by a factor of up to 30 and avoidance of ELMs by actively controlling the edge with magnetic perturbations. Small/no ELM regimes obtained by influencing the edge stability (by plasma shaping, rotational shear control, etc) have shown in present experiments a significant reduction of the ELM heat fluxes compared to type-I ELMs. However, so far they have only been observed under a limited range of pedestal conditions depending on each specific device and their extrapolation to ITER remains uncertain. ELM control by increasing their frequency relies on the controlled triggering of the edge instability leading to the ELM. This has been presently demonstrated with the injection of pellets and with plasma vertical movements; pellets having provided the results more promising for application in ITER conditions. ELM avoidance/suppression takes advantage of the fact that relatively small changes in the pedestal plasma and magnetic field parameters seem to have a large stabilizing effect on large ELMs. Application of edge magnetic field perturbation with non-axisymmetric fields is found to affect transport at the plasma edge and thus prevent the uncontrolled rise of the plasma pressure gradients and the occurrence of type-I ELMs. This paper compiles a brief overview of various ELM control approaches, summarizes their present achievements and briefly discusses the open issues regarding their application in ITER.

Operating ITER in the reference inductive scenario at the design values of IP = 15 MA and QDT = 10 requires the achievement of good H-mode confinement that relies on the presence of an edge transport barrier whose pedestal pressure height is key to plasma performance. Strong gradients occur at the edge in such conditions that can drive MHD instabilities resulting in Edge Localized Modes (ELMs), which produce a rapid energy loss from the pedestal region to the plasma facing components. Without appropriate control, the heat loads on plasma facing components during ELMs in ITER are expected to become significant for operation in H-mode at IP = 6 9 MA; operation at higher plasma currents would result in a very reduced life time of the plasma facing components. Currently, several options are being considered for the achievement of the required level of ELM control in ITER; this includes operation in plasma regimes which naturally have no or very small ELMs, decreasing the ELM energy loss by increasing their frequency by a factor of up to 30 and avoidance of ELMs by actively controlling the edge with magnetic perturbations. Small/no ELM regimes obtained by influencing the edge stability (by plasma shaping, rotational shear control, etc.) have shown in present experiments a significant reduction of the ELM heat fluxes compared to type-I ELMs. However, so far they have only been observed under a limited range of pedestal conditions depending on each specific device and their extrapolation to ITER remains uncertain. ELM control by increasing their frequency relies on the controlled triggering of the edge instability leading to the ELM. This has been

The paper investigates the temporal evolution of γ-ray emission induced by nuclear reactions between energetic D–T fusion-born alpha particles with beryllium impurity ions in JET plasmas, 9Be(α, nγ)12C. During and after tritium NBI blips into JET deuterium plasmas there occurs a delay between alpha production and γ-emission. The delay time observed renders possible the assessment of alpha transport, e.g. loss rates of alphas, and hence may be used for fast ion diagnostics if the characteristic confinement time does not exceed ∼20% of the Spitzer slowing-down time. The analysis of the measured relaxation of γ-emission is further demonstrated for examining the energy distributions of ICRH accelerated 4He ions in the MeV energy range.

This paper considers the heating mix of ITER for the two main scenarios. Presently, 73 MW of absorbed power are foreseen in the mix 20/33/20 for ECH, NBI and ICH. Given a sufficient edge stability, Q = 10—the goal of scenario 2—can be reached with 40 MW power irrespective of the heating method but depends sensitively inter alia on the H-mode pedestal temperature, the density profile shape and on the characteristics of impurity transport. ICH preferentially heats the ions and would contribute specifically with ΔQ < 1.5. The success of the Q = 5 steady-state scenario 4 with reduced current requires discharges with improved confinement necessitating weakly or strongly reversed shear, fbs > 0.5, and strong off-axis current drive (CD). The findings presented here are based on revised CD efficiencies γ for ECCD and a detailed benchmark of several CD codes. With ECCD alone, the goals of scenario 4 can hardly be reached. Efficient off-axis CD is only possible with NBI. With beams, inductive discharges with fni > 0.8 can be maintained for 3000 s. The conclusion of this study is that the present heating mix of ITER is appropriate. It provides the necessary actuators to induce in a flexible way the best possible scenarios. The development risks of NBI at 1 MeV can be reduced by operation at 0.85 MeV. NBI is a very important option for steady-state operation of DEMO because of its high CD efficiency. To increase the H-mode accessibility margin, the addition of 20 MW ECH should be considered.

Following an overview of the ITER Ion Cyclotron Resonance Frequency (ICRF) system, the JET ITER-like antenna (ILA) will be described. The ILA was designed to test the following ITER issues: (a) reliable operation at power densities of order 8MW/m(2) at voltages up to 45kV using a close-packed array of straps; (b) powering through ELMs using an internal (in-vacuum) conjugate-T junction; (c) protection from arcing in a conjugate-T configuration, using both existing and novel systems; and (d) resilience to disruption forces. ITER-relevant results have been achieved: operation at high coupled power density; control of the antenna matching elements in the presence of high inter-strap coupling, use of four conjugate-T systems (as would be used in ITER, should a conjugate-T approach be used); operation with RF voltages on the antenna structures up to 42kV; achievement of ELM tolerance with a conjugate-T configuration by operating at 3 Omega real impedance at the conjugate-T point; and validation of arc detection systems on conjugate-T configurations in ELMy H-mode plasmas. The impact of these results on the predicted performance and design of the ITER antenna will be reviewed. In particular, the implications of the RF coupling measured on JET will be discussed.

The contributions to the 22nd IAEA Fusion Energy Conference (FEC) in the categories of Fusion Technology (FT), ITER Activities (IT) and Safety and Economic Studies (SE) are reviewed. In the FT category, 68 papers were submitted, along with 57 papers submitted through the ITER Organisation in the IT category. Finally two papers were submitted in the SE category. The assembled body of work gave a good overview of the worldwide effort in fusion technology and particularly the prolific activity surrounding the ITER Design Review and the major progress with the ITER technology programme.

Several improvements to the MAST plant and diagnostics have facilitated new studies advancing the physics basis for ITER and DEMO, as well as for future spherical tokamaks (STs). Using the increased heating capabilities P NBI ⩽ 3.8 MW H-mode at I p = 1.2 MA was accessed showing that the energy confinement on MAST scales more weakly with I p and more strongly with B t than in the ITER IPB98(y, 2) scaling. Measurements of the fuel retention of shallow pellets extrapolate to an ITER particle throughput of 70% of its original designed total throughput capacity. The anomalous momentum diffusion, χϕ, is linked to the ion diffusion, χi, with a Prandtl number close to P ϕ ≈ χϕ/χi ≈ 1, although χi approaches neoclassical values. New high spatial resolution measurements of the edge radial electric field, E r , show that the position of steepest gradients in electron pressure and E r (i.e. shearing rate) are coincident, but their magnitudes are not linked. The T e pedestal width on MAST scales with rather than ρpol. The edge localized mode (ELM) frequency for type-IV ELMs, new in MAST, was almost doubled using n = 2 resonant magnetic perturbations from a set of four external coils (n = 1, 2). A new internal 12 coil set (n ⩽ 3) has been commissioned. The filaments in the inter-ELM and L-mode phase are different from ELM filaments, and the characteristics in L-mode agree well with turbulence calculations. A variety of fast particle driven instabilities were studied from 10 kHz saturated fishbone like activity up to 3.8 MHz compressional Alfvén eigenmodes. Fast particle instabilities also affect the off-axis NBI current drive, leading to fast ion diffusion of the order of 0.5 m2 s−1 and a reduction in the driven current fraction from 40% to 30%. EBW current drive start-up is demonstrated for the first time in a ST generating plasma currents up to 55 kA. Many of these studies contributed to the physics basis of a planned upgrade to MAST.

The ITER Ion Cyclotron Resonance Frequency antenna must couple 20 MW at an antenna-plasma spacing of ∼ 15 cm for pulse lengths up to 1000 s at frequencies from 40 MHz to 55 MHz, using matching components mounted outside of the torus to allow powering through fast (sub-ms) changes in loading during Edge Localised Modes (ELM's) by the use of either 3dB couplers or a conjugate-T configuration. The chosen design comprises a port plug supporting a close-packed array of 24 straps which are connected in triplets to eight feed transmission lines. Rear sections of the antenna are removable from the rear of the port plug, to allow damaged windows or diagnostics to be replaced, and much of the interior comprises radiation shielding material. The RF specification poses substantial challenges. Computer modelling has been used to maximise the coupled power and/or reduce electric field strength for the straps, feeders and transmission lines, and is now being extended to minimise power loadings caused by sheath effects. The use of closely-spaced straps leads to significant levels of inter-strap mutual coupling that complicates the matching algorithm. Arc detection is also a key issue for this antenna, as recent JET and Tore Supra results have highlighted the need for parallel development of arc detection and ELM-tolerant systems. The mechanical design challenges lie even further beyond the range of present experience. Given the long pulse length, the thermal design dominates much of the detailed mechanical design as peak RF currents of 1 - 2 kA will result in high thermal loads; a situation exacerbated by the power loading from the plasma. Resilience to disruption forces has required the design of RF windows that can transmit the forces on the central RF conductors to the port plug structure. The requirement that the rear transmission line section is removable considerably increases the complexity of the mechanical layout. Achievement of the required level of radiation shielding is challenging, given the need to both maintain the water/steel fraction close to the optimum value and to keep the total antenna weight below 45 tonnes. This paper details the RF and mechanical design features proposed for the antenna and outlines the manner in which the wider EU programme will feed into the design process

Three new improvements have recently been implemented to the existing JET Ion Cyclotron Resonance Frequency antennae to (a) increase power density and (b) develop matching systems that are tolerant to rapid coupling variations during Edge Localised Modes (ELMs); both of which are key developments for the future design of the ITER ICRF antenna. Firstly, 3dB couplers have been fitted to two antennae in 2004/5. Initial results from these antennae have highlighted the need for parallel development of arc detection and ELM-tolerant systems. Secondly, a new ITER-like antenna (ILA) was installed during 2007 to couple an ITER-relevant power density at the required plasma/antenna spacing using a close-packed array of straps. ELM tolerance is incorporated using an internal (in-vacuum) conjugate-T junction with each strap fed through in-vessel matching capacitors from a common vacuum transmission line. The mechanical engineering design challenges posed by the ILA were significant, given the need for the antenna to: withstand high disruption forces; operate for long pulse length; and achieve a demanding positional accuracy on the in-vessel capacitor tuning system. In addition, it has proved necessary to develop both existing and novel arc detection systems. Prior to installation, the antenna has been high power tested on a test-bed. First results, as well as related simulations, have highlighted the challenges inherent in developing a matching system for an ICRF antenna with closely spaced straps. Thirdly, an externally-mounted conjugate-T (ECT) system has been installed on antennae C and D during the 2006/07 shutdown. The detailed engineering design features of all three developments; the results achieved to date; and the implications for the ITER antenna design will all be reported

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.

The ion cyclotron resonance frequency (ICRF) heating system on JET is currently being upgraded to validate new matching concepts in view of coupling ICRF power to ITER-like plasmas and to further increase the total additional heating power on JET. The present paper reports on first testbed results from the new JET ITER-like antenna as well as on the first use of the newly installed hybrid couplers between two of the existing A2 antennas. Several other on-going improvements, such as improved trip management system, external conjugate-T matching circuit and arc detection systems are also discussed.

Recent scientific and technical progress on JET is summarized. The scientific programme on JET over the last 2 years has benefited significantly from various technological developments: advanced real time tools, enhanced heating systems and several upgraded and new diagnostics. JET has also provided more technological input to ITER on questions related to the first wall, divertor and tritium technology. A dedicated Trace Tritium Campaign has delivered a wealth of data on transport and confinement of tritium and alpha particles from the fusion reaction. An enhancement programme is currently underway on JET, consisting of the implementation of an additional set of upgraded and new diagnostics and a further upgrade of the heating systems. This will allow JET to continue to contribute significantly in questions related to the construction and operational phase of ITER. (c) 2005 Elsevier B.V. All rights reserved.

An analysis of helium exhaust experiments on JET in the MkII-GB divertor configuration is presented. Helium is pumped by applying an argon frost layer on the divertor cryo pump. Measurement of the helium retention time, tau(He)(*),, is performed in two ways: by the introduction of helium in gas puffs and measurement of the subsequent decay time constant of the helium content, tau(He)(d*); and by helium beam injection and measurement of the helium replacement time, tau(He)(r*). In ELMy H-mode, with plasma configuration optimized for pumping, tau(He)(d*) approximate to 7.2 x tau(E)(th) is achieved, where tau(E)(th) is the thermal energy replacement time. For quasi-steady internal transport barrier (ITB) discharges, the achieved tau(He)(r*) approximate to 4.1 x tau(E)(th) is significantly lower. The achieved helium recycling coefficient, confirmed by an independent measurement to be R-eff approximate to 0.91, is the same in both scenarios. None of the discharges are dominated by core confinement. The difference in tau(He)(*)/tau(E)(th) is instead due to the confinement properties of the edge plasma, which is characterized by Type I ELMs for the H-mode discharges studied, and Type III ELMs for the quasi-steady ITB discharges. This difference is quantified by an independent measurement of the ratio of the helium replacement time with a helium edge source to the energy confinement time.

One of the scenarios actively studied during the 2003 trace tritium experimental campaign performed on JET includes a gas puff of small amounts of tritium into sawtooth-unstable H-mode plasmas heated by deuterium beams. The sawtooth crashes observed in these discharges, with the electron cyclotron emission diagnostics, are frequently accompanied by the oscillations of the 14 MeV neutron emission, produced mainly due to the reaction between the deuterium beam and the thermal tritium. Such oscillations are clearly seen in low density plasmas while they are weak or absent at high density. The goal of this work is to explain the different behaviour of the oscillations of neutron emission with density. For this purpose, the dynamics of the trace tritium and beam deuterium during the sawtooth crash in discharges with different plasma density is studied numerically using the TRANSP code. In addition, a possibility of using the neutron measurements in sawtoothing low density plasmas for the test of the full and partial reconnection models is discussed and the sensitivity of the simulations to the key parameters of these models is presented.

In preparation for next step burning plasma devices such as ITER, experimental studies of instabilities and confinement of energetic ions were performed on Joint European Torus (JET) and on Mega-Amper Spherical Tokamak (MAST) with innovative diagnostic techniques, in conventional and shear-reversed plasmas, exploring a wide range of effects for energetic ions. A compendium of recent results testing capabilities of the present-day facilities for burning plasma relevant study is presented in this paper. 'Alpha tail' production using 3rd harmonic ion-cyclotron resonance heating (ICRH) of He-4 beam ions has been employed on JET for studying He-4 of the megaelectronvolt energy range in a 'neutron-free' environment. The evolution of ICRH-accelerated ions of He-4 with E >= 1.7 MeV and D with E >= 500 keV was assessed from nuclear gamma-ray emission born by the fast ions colliding with Be and C impurities. A simultaneous measurement of spatial profiles of fast He-4 and fast D ions relevant to ITER was performed for the first time in positive and strongly reversed magnetic shear discharges. Time-resolved gamma-ray diagnostics for ICRH-accelerated He-3 and H minority ions allowed changes in the fast ion distribution function to be assessed in the presence of unstable toroidal Alfven eigenmodes (TAEs) and sawteeth. A significant decrease of gamma-ray intensity from protons with E >= 5 MeV was detected during the 'tornado' modes. This was interpreted as 'tomado'-induced loss of fast ions with the drift orbit width, Delta(f), comparable to the minor radius of tokamak a. Experiments performed in the opposite case, Delta f/a << 1, for ICRH-accelerated He-3 ions with E >= 500 keV, have shown excitation of numerous Alfven eigenmodes without a significant degradation of the fast ion confinement. The stabilizing effect of fast particles on 'monster' sawteeth was experimentally found to fail in low-density plasmas with high power ion cyclotron resonance frequency (ICRF)-heating. The transition from the 'monster' to short-period 'grassy' sawteeth was investigated with different ICRF phasing, which controls the pinch-effect and radial distribution of ICRF-accelerated ions. Instabilities excited by super-Alfvenic beam ions were investigated on the spherical tokamak MAST. Due to higher values of beta and a higher proportion of fast ions on MAST than on JET, a wider variety of modes and nonlinear regimes for the Alfven instabilities were observed, including the explosive TAE-regimes leading to the formation of hole-clump pairs on the fast ion distribution function. The MAST and START data showed that TAE and chirping modes decrease both in their mode amplitudes and in the number of unstable modes with increasing beta.