Major science and technology issues in the realization of fusion power plants based on magnetic confinement concepts are reviewed. Key socio-economic factors in the areas of safety and environmental impact, public acceptability, operational reliability and availability, and the potential cost of electricity which will influence the eventual adoption of fusion energy for large scale electricity production are discussed. Several representative design concepts based on the tokamak, stellarator and reversed field pinch magnetic confinement configurations are outlined and the key physics and technology elements of each concept are highlighted. A brief introduction is also given to the fusion-fission hybrid concept.

Babinov N. Bader A. Badziak J. Baek S. Bagryansky P. Bakharev N. Ball J. Bandyopadhyay I. Banerjee S. Banerjee S. Banon Navarro A. Bao J. Barbisan M. Barbui T. Bastiani S. Batani D. Batistoni P. Battaglia D. Baylor L. Becoulet M. Belokurov A. Belonohy E. Belova E. Bergmann A. Berk H. Beurskens M. Bhattacharjee A. Bhattacharyay R. Biel W. Bierwage A. Biewer T. Bilato R. Bin W. Bisai N. Bisson R. Bizarro J. Blanchard J. Boeglin W. Bolzonella T. Bombarda F. Bonanomi N. Bonnin X. Bonoli P. Boozer A. Borodin D. Borodkina I. Borthakur S. Bortolon A. Bosch H.-S. Bourdelle C. Breizman B. Bremer P.-T. Brezinsek S. Briguglio S. Brizard A. Brochard G. Bromberg L. Brookman M. Browning P. Brunetti D. Brunsell P. Bruzzone P. Bühler L. Bufferand H. Buller S. Buratti P. Burhenn R. Buzi L. Byggmästar J. Bykov I. Bykov V.

Controlled thermonuclear fusion is the focus of major international research programs that aim to demonstrate the scientific and technological feasibility of large-scale energy production by the fusion of light atomic nuclei. The fundamental elements of the fusion process, the principal approaches to the controlled and sustained production of fusion energy and progress toward the achievement of significant fusion gain are outlined. The advantages and challenges of nuclear fusion as an energy source are summarized and a brief overview of the current status of fusion R&D is presented as an introduction to the more detailed chapters in the Nuclear Fusion R&D Section.

Anti-neuronal antibodies are associated with a range of clinical phenotypes that are often preceded by malignancy. Most of these antibodies can be identified by indirect immunofluorescence (IIF) using commercial primate neural tissue slides. The EUROImmun 12 and upgraded RAVO 14 antigen line immunoblots were directly compared to replace a 9 antigen RAVO immunoblot. Initial assessment was with samples with known positive specificities by IIF. The second phase involved parallel testing of routine pooled negative IIF samples as per institutional standard antineuronal testing algorithm. IIF results and clinical correlation were used to arbitrate on discordant immunoblot results. Correlation in known positives was 80%. The EUROImmun immunoblot did not identify disease causing SOX1 and Recoverin antibodies. Routine pooled analysis did not identify discrepancy between the platforms. Assessment of 318 patient samples did not reveal any cases where a diagnosis was correctly identified by line immunoblot and not IIF. A single sample with positive Yo antibodies was identified by both platforms without clinical or IIF correlation, making this a likely false positive. Subsequent to this audit, Westmead ICPMR has migrated to the 14 antigen RAVO line blot for confirmation of positive IIF or for patients with equivocal results only.

Developments presented at the 27th IAEA Fusion Energy Conference in several areas of experimental magnetic confinement fusion research relating to confinement physics, plasma stability, plasma performance and control are reviewed. First results were reported from significant new stellarator and tokamak facilities and wide-ranging studies of issues such as disruption mitigation, control of edge localized modes (ELMs), the influence of hydrogenic isotope on heat and particle transport, optimization of ELMy H-modes and advanced H-modes and the application of advanced control techniques have been performed in support of the preparations for ITER operation.

ITER is a critical step in the development of fusion energy: its role is to confirm the feasibility of exploiting magnetic confinement fusion for the production of energy for peaceful purposes by providing an integrated demonstration of the physics and technology required for a fusion power plant. Rapid progress is being made in project construction, and the facility is now taking shape at St-Paul-lez-Durance in southern France. In the course of designing and manufacturing of the systems making up the ITER tokamak and the ITER facility, extensive ground-breaking R&D has been implemented by the ITER partners across a wide range of technology and science areas which underpin the achievement of the project’s engineering and fusion plasma performance requirements. Significant developments have been made in the production of high performance Nb 3 Sn superconducting strand and in magnet technologies supporting the construction of the largest superconducting magnets produced to date. High heat flux plasma facing components have been fabricated which are capable of sustaining quasi-stationary heat loads of up to 10 MW m −2 and transient loads of up to 20 MW m −2 . Fusion nuclear technologies such as remote maintenance and tritium breeding have received specific emphasis within the ITER R&D program, since extensive deployment of these technologies is foreseen. Diagnostic systems face particular challenges in the ITER environment, and wide-ranging R&D activities have been implemented to develop novel solutions to ensure an adequate measurement capability in ITER DT operation. Routine and reliable operation in ITER will require a highly effective capability for the detection, avoidance and mitigation of disruptions, and significant science and technology R&D is underway to establish this capability. The overall integration of the control requirements for the ITER plasma and facility, in particular during burning plasma operation, has presented new challenges for fusion control systems, including the need for robust safety and hardware (investment) protection. These challenges are being addressed via the implementation of the most extensive and ambitious control system to date. The paper introduces the ITER project and its major goals in relation to the development of fusion energy and provides an overview of key innovations which have been made in these areas of fusion technology and science in support of ITER construction.

In the four-stage approach of the new ITER Research Plan, the first pre-fusion power operation (PFPO) phase will only have limited power available from external heating and current drive (H&CD) systems: 20-30 MW provided by the electron cyclotron resonance heating (ECRH) system. Accessing the H-mode confinement regime at such low auxiliary power requires operating at low magnetic field, plasma current and density, i.e. 1.8 T and 5 MA for a density between 40% and 50% of the Greenwald density. II-mode plasmas at 5 MA/1.8 T will also be investigated in the second PFPO phase when ITER will have its full complement of H&CD capabilities installed, i.e. 20-30 MW of ECRH, 20 MW of ion cyclotron resonance heating and 33 MW of neutral beam injection. This paper describes the operational constraints and the II&CD capabilities for such scenarios in hydrogen and helium plasmas, to assess their viability and the issues it will be possible to address with them. The modelling results show that 5 MA/1.8 T scenarios are viable and will allow the exploration of the H-mode physics and control issues foreseen in the ITER Research Programme in the PFPO phases.

28 I n practical terms, the project’s goal is to construct and operate a tokamak experiment which can confine a deuterium-tritium plasma in which the α-particle heating dominates all other forms of plasma heating. Formally, the primary mission of the ITER project is to demonstrate sustainment of a DT plasma producing ~500 MW of fusion power for durations of 300 500 s with a ratio of fusion output power to input heating power, Q, of at least 10. ITER is also designed to explore the physics basis for continuous operation of fusion power plants by investigating ‘steady-state’ plasma operation by means of non-inductive current drive for periods of up to several thousand seconds while maintaining a fusion gain, Q, of ~5. If plasma confinement characteristics are favourable, ITER would also be capable of exploring the ‘controlled ignition’ regime of tokamak operation (with Q ~ 30) in which power plant plasmas are expected to operate. The project’s technical goals encompass significant technological demonstrations to prepare the design basis for a fusion power plant. The unique nature of the ITER international collaboration is reflected in the scheme by which the components for the tokamak and auxiliary plant are being constructed. The ITER Organization (IO-CT) in France is responsible for design integration, procurement of components amounting to about 10% of the project’s capital construction cost, management of the on-site installation of the tokamak and plant, and, ultimately, THE FIRST FUSION REACTOR: ITER

The hybrid operating mode observed in several tokamaks is characterized by further enhancement over the high plasma confinement (H-mode) associated with reduced magneto-hydro-dynamic (MHD) instabilities linked to a stationary flat safety factor (q) profile in the core region. The proposed ITER hybrid operation is currently aiming at operating for a long burn duration (>1000s) with a moderate fusion power multiplication factor, Q, of at least 5. This paper presents candidate ITER hybrid operation scenarios developed using a free-boundary transport modelling code, CORSICA, taking all relevant physics and engineering constraints into account. The ITER hybrid operation scenarios have been developed by tailoring the 15 MA baseline ITER inductive H-mode scenario. Accessible operation conditions for ITER hybrid operation and achievable range of plasma parameters have been investigated considering uncertainties on the plasma confinement and transport. ITER operation capability for avoiding the poloidal field coil current, field and force limits has been examined by applying different current ramp rates, flat-top plasma currents and densities, and pre-magnetization of the poloidal field coils. Various combinations of heating and current drive (H&CD) schemes have been applied to study several physics issues, such as the plasma current density profile tailoring, enhancement of the plasma energy confinement and fusion power generation. A parameterized edge pedestal model based on EPED1 added to the CORSICA code has been applied to hybrid operation scenarios. Finally, fully self-consistent free-boundary transport simulations have been performed to provide information on the poloidal field coil voltage demands and to study the controllability with the ITER controllers.

The potential impact of disruptions on the lifetime of ITER components makes the management of plasma disruptions an integral part of the development of the ITER research plan. This paper introduces the concept of a disruption budget consumption for ITER that will aid planning operations and also allows the monitoring of the actual impact of disruptions during operations.

The paper describes the organization of the Test Blanket Module (TBM) program, its overall objective and schedule and the status of the technical activities within the ITER Organization-Central Team (IO-CT). The latter include the design integration of the Test Blanket Systems (TBSs) into the nuclear buildings, ensuring all interfaces with other ITER systems, the design of the common TBS components such as the TBM Frames, the Dummy TBMs, and the TBS maintenance tools and equipment in the TBM Port Cell as well as in the Hot Cell building, the design of the TBS connection pipes and the definition of the required maintenance operations and associated R&D. The paper also discusses the major challenges that the TBM Program will be facing in ITER such as the potential impact of the TBMs ferritic/martensitic structures on plasma operations, the approaches to tritium and contamination confinement, the required mitigation and recovery actions in case of accidents, and the assessment of the reliability aspects that could have an impact on ITER availability.

This paper highlights most urgent issues related to disruption loads, disruption avoidance and prediction as well as load mitigation and runaway control in ITER.

Progress in the definition of the requirements for edge localized mode (ELM) control and the application of ELM control methods both for high fusion performance DT operation and non-active low-current operation in ITER is described. Evaluation of the power fluxes for low plasma current H-modes in ITER shows that uncontrolled ELMs will not lead to damage to the tungsten (W) divertor target, unlike for high-current H-modes in which divertor damage by uncontrolled ELMs is expected. Despite the lack of divertor damage at lower currents, ELM control is found to be required in ITER under these conditions to prevent an excessive contamination of the plasma by W, which could eventually lead to an increased disruptivity. Modelling with the non-linear MHD code JOREK of the physics processes determining the flow of energy from the confined plasma onto the plasma-facing components during ELMs at the ITER scale shows that the relative contribution of conductive and convective losses is intrinsically linked to the magnitude of the ELM energy loss. Modelling of the triggering of ELMs by pellet injection for DIII-D and ITER has identified the minimum pellet size required to trigger ELMs and, from this, the required fuel throughput for the application of this technique to ITER is evaluated and shown to be compatible with the installed fuelling and tritium re-processing capabilities in ITER. The evaluation of the capabilities of the ELM control coil system in ITER for ELM suppression is carried out (in the vacuum approximation) and found to have a factor of similar to 2 margin in terms of coil current to achieve its design criterion, although such a margin could be substantially reduced when plasma shielding effects are taken into account. The consequences for the spatial distribution of the power fluxes at the divertor of ELM control by three-dimensional (3D) fields are evaluated and found to lead to substantial toroidal asymmetries in zones of the divertor target away from the separatrix. Therefore, specifications for the rotation of the 3D perturbation applied for ELM control in order to avoid excessive localized erosion of the ITER divertor target are derived. It is shown that a rotation frequency in excess of 1Hz for the whole toroidally asymmetric divertor power flux pattern is required (corresponding to n Hz frequency in the variation of currents in the coils, where n is the toroidal symmetry of the perturbation applied) in order to avoid unacceptable thermal cycling of the divertor target for the highest power fluxes and worst toroidal power flux asymmetries expected. The possible use of the in-vessel vertical stability coils for ELM control as a back-up to the main ELM control systems in ITER is described and the feasibility of its application to control ELMs in low plasma current H-modes, foreseen for initial ITER operation, is evaluated and found to be viable for plasma currents up to 5-10MA depending on modelling assumptions.

The fast loss of thermal and magnetic energy during an unmitigated disruption in ITER can lead to heat fluxes exceeding melt thresholds for plasma facing components, can generate high electromagnetic loads in some cases close to the design limits, and can potentially cause the generation of high energy runaway electrons. Therefore, the operation strategy in ITER will have a strong focus on limiting the number of disruptions and especially of those that are unmitigated. This will be achieved by a multi-layer strategy including disruption prevention and avoidance, disruption prediction and disruption mitigation. The plasma control system (PCS) will be in charge to react to any deviation from the predicted behavior by applying appropriate correction measures for which an extensive suite of actuators is available including those for active disruption prevention like electron cyclotron heating. It is also the PCS that optimizes triggering of the disruption mitigation system (DMS), should a disruption be imminent despite all control efforts. This paper will give an overview on all aspects of disruption control with a more detailed discussion of the requirements for the DMS.

An assessment of ITER plasma parameters is carried out for the low activation phase that is required for commissioning the basic ITER systems including plasma control, heating and current drive. Such an operation is analysed for hydrogen, helium and deuterium plasmas for full field and current, as well as with magnetic field and plasma current reduced to half of their design values, B0 = 2.65 T, Ip = 7.5 MA. Both hydrogen and deuterium neutral beam injection (NBI) are considered. We assess the possible domain for safe operation, and the possible target plasmas for commissioning the NBI, electron cyclotron heating (ECH) and ion cyclotron heating (ICH) systems, taking into account the constraints imposed by NB shine-through loss, Greenwald limit and access to H-mode operation. Simulations with the Automated System for Transport Analysis (ASTRA) show that for 33 MW of NBI with 20 MW of ECH, H-mode access is marginal for hydrogen plasmas. Good H-mode confinement, expected at PNB + PEC + PIC > 1.5 PL–H, is more likely for the helium and deuterium cases. It is found that plasma parameters, such as normalized beta, plasma density and current flat-top duration, for full power/half field/half current operation can be similar to those required for the DT long pulse operation. Preliminary assessment is also made of the maximum of tritium and neutron yield achievable in a single shot at the deuterium phase of ITER operation.

ITR/P1-35 Effects of ELM control coil on fast ion confinement in ITER H-mode scenarios T. Oikawa, K. Tani, Y. Kusama, T. Sugie, A. Loarte, T. Casper, T. Evans, R. Pitts, K. Shinohara, M.J. Schaffer, P. Cahyna, Y. Gribov, J. Snipes, D.J. Campbell 1 ITER Organization, Route de Vinon sur Verdon, 13115 Saint. Paul Lez Durance, France, 2 Nihon Advanced Technology Inc., Naka, Japan, 3 Japan Atomic Energy Agency, Naka, Japan, 4 General Atomics, San Diego, CA, USA, 5 Institute of Plasma Physics AS CR, Prague, Czech Republic. e-mail contact of main author: toshihiro.oikawa@iter.org Abstract. The fast ion confinement in ergodized edge magnetic fields produced by the ELM control coil (ECC) by linear superposition of the plasma and externally imposed field (vacuum approximation) has been investigated for ITER H-mode scenarios. The dependences of the fast ion loss on the ECC current amplitude and on the toroidal mode number of the ECC applied have been investigated. It has been found that most of the fast particle and energy losses are deposited in the divertor region. The detailed spatial distribution of these loses depends on the toroidal mode of the ECC field. For the largest currents applied to the ECC, the loads can be significant (0.2 − 0.3MW/m) in some areas near the divertor dome which were not originally foreseen to receive plasma loads of this magnitude. Sensitivity studies for possible variations in the density and edge safety factor profiles have also been investigated.