ITER

Progress in the area of MHD stability and disruptions, since the publication of the 1999 ITER Physics Basis document (1999 Nucl. Fusion 39 2137-2664), is reviewed. Recent theoretical and experimental research has made important advances in both understanding and control of MHD stability in tokamak plasmas. Sawteeth are anticipated in the ITER baseline ELMy H-mode scenario, but the tools exist to avoid or control them through localized current drive or fast ion generation. Active control of other MHD instabilities will most likely be also required in ITER. Extrapolation from existing experiments indicates that stabilization of neoclassical tearing modes by highly localized feedback-controlled current drive should be possible in ITER. Resistive wall modes are a key issue for advanced scenarios, but again, existing experiments indicate that these modes can be stabilized by a combination of plasma rotation and direct feedback control with non-axisymmetric coils. Reduction of error fields is a requirement for avoiding non-rotating magnetic island formation and for maintaining plasma rotation to help stabilize resistive wall modes. Recent experiments have shown the feasibility of reducing error fields to an acceptable level by means of non-axisymmetric coils, possibly controlled by feedback. The MHD stability limits associated with advanced scenarios are becoming well understood theoretically, and can be extended by tailoring of the pressure and current density profiles as well as by other techniques mentioned here. There have been significant advances also in the control of disruptions, most notably by injection of massive quantities of gas, leading to reduced halo current fractions and a larger fraction of the total thermal and magnetic energy dissipated by radiation. These advances in disruption control are supported by the development of means to predict impending disruption, most notably using neural networks. In addition to these advances in means to control or ameliorate the consequences of MHD instabilities, there has been significant progress in improving physics understanding and modelling. This progress has been in areas including the mechanisms governing NTM growth and seeding, in understanding the damping controlling RWM stability and in modelling RWM feedback schemes. For disruptions there has been continued progress on the instability mechanisms that underlie various classes of disruption, on the detailed modelling of halo currents and forces and in refining predictions of quench rates and disruption power loads. Overall the studies reviewed in this chapter demonstrate that MHD instabilities can be controlled, avoided or ameliorated to the extent that they should not compromise ITER operation, though they will necessarily impose a range of constraints.

In response to the 2013 Update of the European Strategy for Particle Physics, the Future Circular Collider (FCC) study was launched, as an international collaboration hosted by CERN. This study covers a highest-luminosity high-energy lepton collider (FCC-ee) and an energy-frontier hadron collider (FCC-hh), which could, successively, be installed in the same 100 km tunnel. The scientific capabilities of the integrated FCC programme would serve the worldwide community throughout the 21st century. The FCC study also investigates an LHC energy upgrade, using FCC-hh technology. This document constitutes the second volume of the FCC Conceptual Design Report, devoted to the electron-positron collider FCC-ee. After summarizing the physics discovery opportunities, it presents the accelerator design, performance reach, a staged operation scenario, the underlying technologies, civil engineering, technical infrastructure, and an implementation plan. FCC-ee can be built with today's technology. Most of the FCC-ee infrastructure could be reused for FCC-hh. Combining concepts from past and present lepton colliders and adding a few novel elements, the FCC-ee design promises outstandingly high luminosity. This will make the FCC-ee a unique precision instrument to study the heaviest known particles (Z, W and H bosons and the top quark), offering great direct and indirect sensitivity to new physics.

The understanding and predictive capability of transport physics and plasma confinement is reviewed from the perspective of achieving reactor-scale burning plasmas in the ITER tokamak, for both core and edge plasma regions. Very considerable progress has been made in understanding, controlling and predicting tokamak transport across a wide variety of plasma conditions and regimes since the publication of the ITER Physics Basis (IPB) document (1999 Nucl. Fusion 39 2137-2664). Major areas of progress considered here follow. (1) Substantial improvement in the physics content, capability and reliability of transport simulation and modelling codes, leading to much increased theory/experiment interaction as these codes are increasingly used to interpret and predict experiment. (2) Remarkable progress has been made in developing and understanding regimes of improved core confinement. Internal transport barriers and other forms of reduced core transport are now routinely obtained in all the leading tokamak devices worldwide. (3) The importance of controlling the H-mode edge pedestal is now generally recognized. Substantial progress has been made in extending high confinement H-mode operation to the Greenwald density, the demonstration of Type I ELM mitigation and control techniques and systematic explanation of Type I ELM stability. Theory-based predictive capability has also shown progress by integrating the plasma and neutral transport with MHD stability. (4) Transport projections to ITER are now made using three complementary approaches: empirical or global scaling, theory-based transport modelling and dimensionless parameter scaling (previously, empirical scaling was the dominant approach). For the ITER base case or the reference scenario of conventional ELMy H-mode operation, all three techniques predict that ITER will have sufficient confinement to meet its design target of Q = 10 operation, within similar uncertainties.

Abstract: We review the physics opportunities of the Future Circular Collider, covering its e+e-, pp, ep and heavy ion programmes. We describe the measurement capabilities of each FCC component, addressing the study of electroweak, Higgs and strong interactions, the top quark and flavour, as well as phenomena beyond the Standard Model. We highlight the synergy and complementarity of the different colliders, which will contribute to a uniquely coherent and ambitious research programme, providing an unmatchable combination of precision and sensitivity to new physics.

• Reviews the fundamental physics aspects of the first ITER W divertor and defines the required operational lifetime within the Staged Approach. • Uses the ITER divertor SOLPS simulation database to establish the target peak heat flux and neutral pressure burning plasma operating domain. • Assesses consequences of narrow SOL heat flux channels, fluid drifts, component shaping and 3D magnetic fields for ELM control. • Uses W recrystallization to define an operational budget and shows that heat fluxes ∼50% higher than previously assumed may be acceptable. • Shows that Ne and N should be equally good as seed impurities and suggests that very strong ELM mitigation will be required at high performance. • Provides a list of key outstanding R&D areas to consolidate the divertor physics basis in the period up to ITER operation. On the eve of component procurement, this paper discusses the present physics basis for the first ITER tungsten (W) divertor, beginning with a reminder of the key elements defining the overall design, and outlining relevant aspects of the Research Plan accompanying the new “staged approach” to ITER nuclear operations which fixes the overall divertor lifetime constraint. The principal focus is on the main design driver, steady state power fluxes in the DT phases, obtained from simulations using the 2-D SOLPS-4.3 and SOLPS-ITER plasma boundary codes, assuming the use of the low Z seeding impurities nitrogen (N) and neon (Ne). A new perspective on the simulation database is adopted, concentrating purely on the divertor physics aspects rather than on the core-edge integration, which has been studied extensively in the course of the divertor design evolution and is published elsewhere. Emphasis is placed on factors which may increase the peak steady state loads: divertor target shaping for component misalignment protection, the influence of fluid drifts, and the consequences of narrow scrape-off layer heat flux channels. All tend to push the divertor into an operating space at higher sub-divertor neutral pressure in order to remain at power flux densities acceptable for the target material. However, a revised criterion for the maximum tolerable loads based on avoidance of W recrystallization, sets an upper limit potentially ∼50% higher than the previously accepted value of ∼10 MW m −2 , a consequence both of the choice of material and the finalized component design. Although the simulation database is currently restricted to the 2-D toroidally symmetric situation, considerable progress is now also being made using the EMC3-Eirene 3-D code suite for the assessment of power loading in the presence of magnetic perturbations for ELM control. Some new results for low input power corresponding to the early H-mode operation phases are reported, showing that even if realistic plasma screening is taken into account, significant asymmetric divertor heat fluxes may arise far from the unperturbed strike point. The issue of tolerable limits for transient heat pulses is an open and key question. A new scaling for ELM power deposition has shown that whilst there may be more latitude for operation at higher current without ELM control, the ultimate limit is likely to be set more by material fatigue under large numbers of sub-threshold melting events.

Abstract: In response to the 2013 Update of the European Strategy for Particle Physics (EPPSU), the Future Circular Collider (FCC) study was launched as a world-wide international collaboration hosted by CERN. The FCC study covered an energy-frontier hadron collider (FCC-hh), a highest-luminosity high-energy lepton collider (FCC-ee), the corresponding 100 km tunnel infrastructure, as well as the physics opportunities of these two colliders, and a high-energy LHC, based on FCC-hh technology. This document constitutes the third volume of the FCC Conceptual Design Report, devoted to the hadron collider FCC-hh. It summarizes the FCC-hh physics discovery opportunities, presents the FCC-hh accelerator design, performance reach, and staged operation plan, discusses the underlying technologies, the civil engineering and technical infrastructure, and also sketches a possible implementation. Combining ingredients from the Large Hadron Collider (LHC), the high-luminosity LHC upgrade and adding novel technologies and approaches, the FCC-hh design aims at significantly extending the energy frontier to 100 TeV. Its unprecedented centre of-mass collision energy will make the FCC-hh a unique instrument to explore physics beyond the Standard Model, offering great direct sensitivity to new physics and discoveries.

A multi-machine database for the H-mode scrape-off layer power fall-off length, λq in JET, DIII-D, ASDEX Upgrade, C-Mod, NSTX and MAST has been assembled under the auspices of the International Tokamak Physics Activity. Regression inside the database finds that the most important scaling parameter is the poloidal magnetic field (or equivalently the plasma current), with λq decreasing linearly with increasing Bpol. For the conventional aspect ratio tokamaks, the regression finds , yielding λq,ITER ≅ 1 mm for the baseline inductive H-mode burning plasma scenario at Ip = 15 MA. The experimental divertor target heat flux profile data, from which λq is derived, also yield a divertor power spreading factor (S) which, together with λq, allows an integral power decay length on the target to be estimated. There are no differences in the λq scaling obtained from all-metal or carbon dominated machines and the inclusion of spherical tokamaks has no significant influence on the regression parameters. Comparison of the measured λq with the values expected from a recently published heuristic drift based model shows satisfactory agreement for all tokamaks.

The ITER neutral beam (NB) injectors are the first injectors that will have to operate under conditions and constraints similar to those that will be encountered in a fusion reactor. These injectors will have to operate in a hostile radiation environment and they will become highly radioactive due to the neutron flux from ITER. The injectors will use a single large ion source and accelerator that will produce 40 A 1 MeV D − beams for pulse lengths of up to 3600 s. Significant design changes have been made to the ITER heating NB (HNB) injector over the past 4 years. The main changes are: Modifications to allow installation and maintenance of the beamline components with an overhead crane. The beam source vessel shape has been changed and the beam source moved to allow more space for the connections between the 1 MV bushing and the beam source. The RF driven negative ion source has replaced the filamented ion source as the reference design. The ion source and extractor power supplies will be located in an air insulated high voltage (−1 MV) deck located outside the tokamak building instead of inside an SF 6 insulated HV deck located above the injector. Introduction of an all metal absolute valve to prevent any tritium in the machine to escape into the NB cell during maintenance. This paper describes the status of the design as of December 2008 including the above mentioned changes. The very important power supply system of the neutral beam injectors is not described in any detail as that merits a paper beyond the competence of the present authors. The R&D required to realize the injectors described in this paper must be carried out on a dedicated neutral beam test facility, which is not described here.

I would firstly like to congratulate all who have contributed to the preparation of the `Progress in the ITER Physics Basis' (PIPB) on its publication and express my deep appreciation of the hard work and commitment of the many scientists involved. With the signing of the ITER Joint Implementing Agreement in November 2006, the ITER Members have now established the framework for construction of the project, and the ITER Organization has begun work at Cadarache. The review of recent progress in the physics basis for burning plasma experiments encompassed by the PIPB will be a valuable resource for the project and, in particular, for the current Design Review.

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 ∼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 1 Hz 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 5–10 MA depending on modelling assumptions.

Since the ITER divertor design includes tungsten monoblocks in the vertical target where heat loads are maximal, the design to protect leading edges as well as technology R&D for high performance armor-heat sink joint were necessary to be implemented. In the R&D, the availability of the technology was demonstrated by high heat flux test of tungsten monoblock components. Not systematically but frequently macro-cracks appeared at the middle of monoblocks after 20 MW/m2 loading. The initiation of such macro-cracks was considered to be due to cyclic exposure to high temperature, ∼2000 °C, where creep, recrystallization and low cycle fatigue were concerned. To understand correlation between the macro-crack appearance and mechanical properties and possible update of acceptance criteria in the material specification, an activity to characterize the tungsten monoblocks was launched.

The heating neutral beam injectors (HNBs) of ITER are designed to deliver 16.7MWof 1 MeVD0 or 0.87 MeVH0 to the ITER plasma for up to 3600 s. They will be the most powerful neutral beam(NB) injectors ever, delivering higher energy NBs to the plasma in a tokamak for longer than any previous systems have done. The design of the HNBs is based on the acceleration and neutralisation of negative ions as the efficiency of conversion of accelerated positive ions is so low at the required energy that a realistic design is not possible, whereas the neutralisation ofH− andD− remains acceptable (≈56%). The design of a long pulse negative ion based injector is inherently more complicated than that of short pulse positive ion based injectors because: • negative ions are harder to create so that they can be extracted and accelerated from the ion source; • electrons can be co-extracted from the ion source along with the negative ions, and their acceleration must be minimised to maintain an acceptable overall accelerator efficiency; • negative ions are easily lost by collisions with the background gas in the accelerator; • electrons created in the extractor and accelerator can impinge on the extraction and acceleration grids, leading to high power loads on the grids; • positive ions are created in the accelerator by ionisation of the background gas by the accelerated negative ions and the positive ions are back-accelerated into the ion source creating a massive power load to the ion source; • electrons that are co-accelerated with the negative ions can exit the accelerator and deposit power on various downstream beamline components. The design of the ITER HNBs is further complicated because ITER is a nuclear installation which will generate very large fluxes of neutrons and gamma rays. Consequently all the injector components have to survive in that harsh environment. Additionally the beamline components and theNBcell, where the beams are housed, will be activated and all maintenance will have to be performed remotely. This paper describes the design of theHNBinjectors, but not the associated power supplies, cooling system, cryogenic system etc, or the high voltage bushingwhich separates the vacuum of the beamline fromthehighpressureSF6 of the high voltage (1MV) transmission line, through which the power, gas and coolingwater are supplied to the beam source. Also themagnetic field reduction system is not described.

A specificity of Brassicaceous plants is the production of sulphur secondary metabolites called glucosinolates that can be hydrolysed into glucose and biocidal products. Among them, isothiocyanates are toxic to a wide range of microorganisms and particularly soil-borne pathogens. The aim of this study was to investigate the role of glucosinolates and their breakdown products as a factor of selection on rhizosphere microbial community associated with living Brassicaceae. We used a DNA-stable isotope probing approach to focus on the active microbial populations involved in root exudates degradation in rhizosphere. A transgenic Arabidopsis thaliana line producing an exogenous glucosinolate and the associated wild-type plant associated were grown under an enriched (13)CO(2) atmosphere in natural soil. DNA from the rhizospheric soil was separated by density gradient centrifugation. Bacterial (Alphaproteobacteria, Betaproteobacteria, Gammaproteobacteria and Acidobacteria), Archaea and fungal community structures were analysed by DGGE fingerprints of amplified 16S and 18S rRNA gene sequences. Specific populations were characterized by sequencing DGGE fragments. Roots of the transgenic plant line presented an altered profile of glucosinolates and other minor additional modifications. These modifications significantly influenced microbial community on roots and active populations in the rhizosphere. Alphaproteobacteria, particularly Rhizobiaceae, and fungal communities were mainly impacted by these Brassicaceous metabolites, in both structure and composition. Our results showed that even a minor modification in plant root could have important repercussions for soil microbial communities.

CRONOS is a suite of numerical codes for the predictive/interpretative simulation of a full tokamak discharge. It integrates, in a modular structure, a 1D transport solver with general 2D magnetic equilibria, several heat, particle and impurities transport models, as well as heat, particle and momentum sources. This paper gives a first comprehensive description of the CRONOS suite: overall structure of the code, main available models, details on the simulation workflow and numerical implementation. Some examples of applications to the analysis of experimental discharges and the predictions of ITER scenarios are also given.

Procurement of the ITER magnets is due to start at the end of 2007/early 2008, with the launch of the longest lead time items, the conductor and the TF coil windings. The base design for procurement was established in 2001, and the build up of the Cadarache ITER team has been accompanied by a review of the most critical, or controversial, features of the 2001 design. At the same time, an urgent R&D program has been launched to complete the necessary verification of the design solutions that are proposed. In this paper an overview will be presented of the main design features and drivers, and some of the recent issues and R&D results will be summarized.

The area of energetic particle (EP) physics in fusion research has been actively and extensively researched in recent decades. The progress achieved in advancing and understanding EP physics has been substantial since the last comprehensive review on this topic by Heidbrink and Sadler (1994 Nucl. Fusion 34 535). That review coincided with the start of deuterium–tritium (DT) experiments on the Tokamak Fusion Test Reactor (TFTR) and full scale fusion alphas physics studies.

An overview of the present status of research toward the final design of the ITER disruption mitigation system (DMS) is given. The ITER DMS is based on massive injection of impurities, in order to radiate the plasma stored energy and mitigate the potentially damaging effects of disruptions. The design of this system will be extremely challenging due to many physics and engineering constraints such as limitations on port access and the amount and species of injected impurities. Additionally, many physics questions relevant to the design of the ITER disruption mitigation system remain unsolved such as the mechanisms for mixing and assimilation of injected impurities during the rapid shutdown and the mechanisms for the subsequent formation and dissipation of runaway electron current.

This review covers basic physics ingredients of the runaway phenomenon and the ongoing efforts (experimental and theoretical) aimed at runaway electron taming in the next generation tokamaks. We emphasize the prevailing physics themes of the last 20 years: the hot-tail mechanism of runaway production, runaway electron interaction with impurity ions, the role of synchrotron radiation in runaway kinetics, runaway electron transport in presence of magnetic fluctuations, micro-instabilities of runaway electrons in magnetized plasmas, and vertical stability of the plasma with runaway electrons. Here, the review also discusses the implications of the runaway phenomenon for ITER and the current strategy of runaway electron mitigation.

Author: Hawryluk, R. J. et al.; Genre: Journal Article; Issued: 2009; Keywords: 22nd IAEA Fusion Energy Conference Govt Switzerland, Geneva, SWITZERLAND, 2008-10-13 to 2008-10-18; Title: Principal physics developments evaluated in the ITER design review

The ITER magnet procurement is now well underway. The magnet systems consist of 4 superconducting coil sets (toroidal field (TF), poloidal field (PF), central solenoid (CS) and correction coils (CC)) which use both NbTi and Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn-based conductors. The magnets sit at the core of the ITER machine and are tightly integrated with each other and the main vacuum vessel. The total weight of the system is about 10000 t, of which about 500 t are strands and 250 t, NbTi. The reaction of the magnetic forces is a delicate balance that requires tight control of tolerances and the use of high-strength, fatigue-resistance steel forgings. Integration and support of the coils and their supplies, while maintaining the necessary tolerances and clearance gaps, have been completed in steps, the last being the inclusion of the feeder systems. Twenty-one procurement agreements have now been signed with 6 of the ITER Domestic Agencies for all of the magnets together with the supporting feeder subsystems. All of them except one (for the CS coils) are so-called Build to Print agreements where the IO provides the detailed design including full three-dimensional CAD models. The production of the first components is underway (about 175 t of strand was finished by July 2011) and manufacturing prototypes of TF coil components are being completed. The paper will present a design overview and the manufacturing status.