International Tokamak Experimental Reactor Research Articles

Modelling the capacitance of the elongated plasma in tokamak

The capacitance model suitable for the non-circular cross-section plasma is studied based on the capacitance model of the circular cross-section plasma. The coaxial elliptic-torus capacitor property is further derived and used to determine the capacity of non-circular cross-section tokamak plasma, such as EAST (Experimental Advanced Superconducting Tokamak). By testing all the physical terms in this model, we find that the capacitance Cp\\documentclass[12pt]{minimal} \\usepackage{amsmath} \\usepackage{wasysym} \\usepackage{amsfonts} \\usepackage{amssymb} \\usepackage{amsbsy} \\usepackage{mathrsfs} \\usepackage{upgreek} \\setlength{\\oddsidemargin}{-69pt} \\begin{document}$$Cp$$\\end{document}) is increasing exponentially with the increase of elongation ratio (k2/k1), while the minor radius ratio (a2/a1) is just reversed at the flat-top of plasma current, and the capacitance property is implicitly included in the H-mode study during the L–H transition. It is noted that Cp-H mode is the least and Cp-I mode is approximately equal to Cp-L mode under the L-mode, I-mode and H-mode regimes based on this capacitance model in EAST. Consequently, it may be integrated into an equivalent circuit of the tokamak transformer or transport computer code of the edge plasma for use in precise simulations of fusion plasma behavior in the future, such as ITER (International Tokamak Experimental Reactor) or BEST (Burning-plasma Experimental Superconducting Tokamak) in China.

Models of resistive wall tearing mode disruptions

Disruptions are a serious issue in tokamaks. In a disruption, the thermal energy is lost by means of an instability which could be a resistive wall tearing mode (RWTM). During precursors to a disruption, the plasma edge region cools, causing the current to contract. Model sequences of contracted current equilibria are given, and their stability is calculated. A linear stability study shows that there is a maximum value of edge qa≈3 for RWTMs to occur. This also implies a minimum rational surface radius normalized to plasma radius from RWTMs to be unstable. Nonlinear simulations are performed using a similar model sequence derived from an equilibrium reconstruction. There is a striking difference in the results, depending on whether the wall is ideal or resistive. With an ideal wall, the perturbations saturate at moderate amplitude, causing a minor disruption without a thermal quench. With a resistive wall, there is a major disruption with a thermal quench, if the edge qa≤3. There is a sharp transition in nonlinear behavior at qa=3. This is consistent with the linear model and with experiments. If disruptions are caused by RWTMs, then devices with highly conducting walls, such as the International Tokamak Experimental Reactor will experience much milder, tolerable, disruptions than presently predicted.

Investigation of ELM-related Larmor ion flux into toroidal gaps of divertor target plates

A detailed assessment of the thermo-mechanical limits of the International Tokamak Experimental Reactor (ITER) divertor with respect to potential excessive local transient heat loads due to edge localised modes (ELMs) has revealed a particular power loading scenario arising from the fact that ELM ions expelled from the upstream pedestal region will arrive at the divertor target plates without substantial thermalisation. As a consequence of their Larmor gyration around magnetic field lines, they are able to penetrate toroidal gaps between individual monoblocks of the target plate structure and can deliver rather intense heat loads to monoblock side faces near the gap entrance. To verify that this ELM-induced loading, predicted by both ion orbit simulations and particle in cell simulations, really does occur, two dedicated experiments have been performed on the ASDEX Upgrade tokamak. In both experiments a model toroidal gap structure of similar dimensions to those of the ITER divertor target monoblocks was exposed to a series of identical H-mode discharges with strong type-I ELMs. The effects arising from the gyro motion of hot ELM ions were identified by inverting, in the second experiment, the directions of both toroidal field and plasma current, thus reversing the ion gyration direction. The local distribution of incident ion flux on the gap side faces was quantified by pre- and post-exposure analysis of platinum marker layers to determine quantitatively the erosion rate of the platinum marker. The results fully confirm the ion orbit code predictions with respect to the penetration depth of incident ions with gyro orbits of similar or larger radius than the gap width. Moreover, the results confirm that ELM ions do indeed arrive at the divertor with their typical pedestal energies and also allow conclusions to be drawn regarding the corresponding intra-ELM ion particle and power flux, which is not easy to quantify using Langmuir probes.

Novel dual-reflection design applied for ITER core x-ray spectrometer.

A novel dual-reflection configuration is introduced for the International Tokamak Experimental Reactor (ITER) core x-ray spectrometer to fit the allocated space where it will be placed accompanied by moving the detectors backward to reduce the incident radiation dose. The highly oriented pyrolytic graphite, which has a mosaic structure of microscopic crystallites, is chosen for the front reflector motivated by higher x-ray throughput and stronger misalignment tolerance compared to the perfect crystal reflector. In the ITER core x-ray spectrometer, a combination of several reflector-deflected Lines of Sight (LOSs) and a direct LOS is proposed for the first time named X-Ray Crystal Spectroscopy Core (XRCS-Core). The system is optimized to observe lines from externally seeded xenon and the intrinsic tungsten impurity, meeting both port integration needs and measurement requirements. Its spectral performance is simulated using an analytical-raytracing mixed code--XRSA, showing good imaging quality with a spectral resolution higher than 8000. The XRCS-Core system is thought to be applicable in various ITER scenarios through the assessment taking into account the spectrometers' specifications and the chosen lines' emissivity in different plasma parameters.

Progress on European ITER Toroidal Field Coil Procurement: Cold Test and Insertion Work Package

The plasma confinement of the International Tokamak Experimental Reactor (ITER) is provided by the magnetic field generated by 18 toroidal field coils. Fusion for Energy (F4E), the European Domestic Agency for ITER, is responsible for the supply of 10 TFC to ITER project. Their procurement has been divided into three main work packages: I) the production of the radial plates, structural stainless steel components housing the Nb3Sn conductors, II) the manufacture of 10 Winding Packs (WP) and III) the cold test of 10 WP plus their insertion into the Coil Cases. This article gives an update of the status of the production of the third and last work package performed under the framework of an F4E contract assigned to SIMIC SpA, an Italian company. In particular, the details of the WP thermal cycle, the insertion of the firsts TF coil in their coil cases, the closure weld, the gap filling and the machining strategy adopted to optimize the final TF coil shape optimized to minimize the field errors are presented.

Beam energy recovery for fusion and collector design for tests on compact sources

The next fusion project DEMO, which will be the evolution of the experimental fusion reactor [International Tokamak Experimental Reactor (ITER)], would require a high efficient energy production. As in ITER, DEMO will use fast Neutral Beam (NB) injectors to increase the plasma temperature needed for the fusion reaction. A way to recover the electric energy production efficiency in DEMO could be the beam energy recovery in the NB production, which is produced by a D- beam, neutralized by a gas cell with 60% efficiency. A compact energy recovery device with an axisymmetric cylindrical ion collector that uses only decelerating electric fields combined with the beam space charge effect has been recently proposed. It can be used for a test on the beam of the NIO1 (Negative Ion Optimization 1) source, a compact ion source (scaled down from ITER size sources) that has been developed at INFN-LNL and Consorzio RFX (Padua). The detailed collector design to be used on one of the beamlets of the NIO1 source within typical space limitation is presented and discussed here. Furthermore, a preliminary trajectory simulation for a beam with a rectangular geometry similar to the beam used in ITER to verify the beam recovery for a nonaxial symmetric geometry is also shown.

Progress of physics understanding for long pulse high-performance plasmas on EAST towards the steady-state operation of ITER and CFETR

Recently, the first ever 100 s long, steady-state H-mode discharge with good control of impurities, core and edge MHD stabilities, and heat exhaust was demonstrated in the Experimental Advanced Superconducting Tokamak (EAST) using the ITER-like (International Tokamak Experimental Reactor) tungsten upper divertor. Using both radio frequency (RF) power and neutral beam injection (NBI) heating, EAST has demonstrated fully non-inductive scenarios with an extension of fusion performance at high density and low rotation: β P ∼ 2.5, β N ∼ 2.0, H98,y2 ∼ 1.2, bootstrap current fraction fBS ∼50% at q95 ∼ 6.8. With pure RF power heating, plasmas have been maintained for up to 21 s (over 40 times the current relaxation time) with zero loop voltage and small edge localized modes (ELMs) at high density (ne/nGW ∼ 0.6–0.8), β P ∼ 2.0, β N ∼ 1.6, and ƒBS ∼47%. Experimental investigations show how plasma current profiles, turbulent transport and radiation properties self-consistently evolve toward fusion relevant steady state conditions. Modeling and physics experiments have confirmed the synergistic effects between electron cyclotron heating (ECH) and low hybrid wave (LHW), where ECH enhances the heating and current drive from LHW injection, enabling fully non-inductive operation at higher density. Small/no ELMs facilitate the RF power coupling in the H-mode phase and reduce divertor erosion. A low tungsten concentration was observed at high β P with a hollow profile in the core. Reduction of the peak divertor heat flux with f rad of up to 40% was compatible with the high β P scenario by using active radiation feedback control. With features such as dominant electron heating, zero/low NBI torque and an ITER-like tungsten divertor, fully non-inductive high-performance experiments on EAST offer unique contributions towards the succesful operation of ITER and CFETR (the Chinese Fusion Engineering Testing Reactor).

Integrated core-SOL modelling of fuelling, density control and divertor heat loads for the flat-top phase of the ITER H-mode D-T plasma scenarios

The operation of a tokamak designed to test the sustainability of a thermonuclear grade plasma like the international tokamak experimental reactor (ITER) presents several challenges. Among them is the necessity of fuelling the plasma to reach the density required to generate enough fusion power to achieve and, at the same time, to avoid excessive power loads at the divertor targets (beyond 10 MW m). Whether this goal is achievable or not depends on the details of the fuelling scheme, of the atomic species that are injected into the plasma, of the power radiation pattern in the scrape-off layer and in the divertor, of the additional heating schemes and of the transport mechanisms at work in the core and near the edge of the plasma.In this study we present, for different operational scenarios, the results of an integrated modelling approach to the problem taking into account all the different aspects of it, albeit with some limitations and simplifications discussed in the paper. The tool adopted for our simulations is the JINTRAC suite of codes, which can simulate in an integrated fashion the transport of particles and heat in different regions of the plasma.Within the limitation of our modelling our assessment is that, by carefully tuning the gas fuelling and impurity seeding, it is indeed possible for ITER to achieve and at the same time maintain acceptable divertor power loads. We also investigate the sensitivity of this result to some of the uncertainties in the modelling assumptions underlying the simulations presented in the paper.

Suppression of Alfvén Modes on the National Spherical Torus Experiment Upgrade with Outboard Beam Injection.

In this Letter we present data from experiments on the National Spherical Torus Experiment Upgrade, where it is shown for the first time that small amounts of high pitch-angle beam ions can strongly suppress the counterpropagating global Alfvén eigenmodes (GAE). GAE have been implicated in the redistribution of fast ions and modification of the electron power balance in previous experiments on NSTX. The ability to predict the stability of Alfvén modes, and developing methods to control them, is important for fusion reactors like the International Tokamak Experimental Reactor, which are heated by a large population of nonthermal, super-Alfvénic ions consisting of fusion generated α's and beam ions injected for current profile control. We present a qualitative interpretation of these observations using an analytic model of the Doppler-shifted ion-cyclotron resonance drive responsible for GAE instability which has an important dependence on k_{⊥}ρ_{L}. A quantitative analysis of this data with the hym stability code predicts both the frequencies and instability of the GAE prior to, and suppression of the GAE after the injection of high pitch-angle beam ions.

Error Field Impact on Plasma Boundary in ITER Scenarios

Discrepancies in magnetic field maps produced by confinement coils in thermonuclear fusion reactors may drive plasma to lose stability and must be carefully controlled during the whole time evolution of each shot using suitable correction coils. Even when kept below safety thresholds, these error fields (EFs) may alter the geometry of magnetic flux lines, defining the plasma geometry. This paper, using high accuracy 3-D magnetic field computations for confinement coils, addresses the issue of evaluating the effect of EFs on the plasma boundary shape during the shot, modeled as a sequence of equilibrium configurations. In particular, a procedure that can compute the shape perturbations due to given deformations of the coils has been set up and used to carry out an analysis of relationship between the EF and shape perturbations during the time evolution of International Tokamak Experimental Reactor programmed scenario.

Progress on ion cyclotron range of frequencies heating physics and technology in support of the International Tokamak Experimental Reactor

Ion cyclotron range of frequency (ICRF) heating is foreseen as an integral component of the initial ITER operation. The status of ICRF preparations for ITER and supporting research were updated in the 2007 [Gormezano et al., Nucl. Fusion 47, S285 (2007)] report on the ITER physics basis. In this report, we summarize progress made toward the successful application of ICRF power on ITER since that time. Significant advances have been made in support of the technical design by development of new techniques for arc protection, new algorithms for tuning and matching, carrying out experimental tests of more ITER like antennas and demonstration on mockups that the design assumptions are correct. In addition, new applications of the ICRF system, beyond just bulk heating, have been proposed and explored.

Beyond ITER: Neutral beams for a demonstration fusion reactor (DEMO) (invited)

In the development of magnetically confined fusion as an economically sustainable power source, International Tokamak Experimental Reactor (ITER) is currently under construction. Beyond ITER is the demonstration fusion reactor (DEMO) programme in which the physics and engineering aspects of a future fusion power plant will be demonstrated. DEMO will produce net electrical power. The DEMO programme will be outlined and the role of neutral beams for heating and current drive will be described. In particular, the importance of the efficiency of neutral beam systems in terms of injected neutral beam power compared to wallplug power will be discussed. Options for improving this efficiency including advanced neutralisers and energy recovery are discussed.

Equivalent Circuit Analysis for Core Snubber

An equivalent circuit for a core snubber is established in this paper. The circuit can approximately express a parallel resistance and inductance. The resistance can be directly calculated according to the resistivity of the magnetic core and structure dimensions of the core snubber referring to Fink-Baker-Owren (FBO) method, and the eddy current factor is adjusted to one. In this paper, the shunt inductance neglected by FBO is first considered, and it can be obtained from critical characteristics of the magnetic core and structure dimensions of the core snubber. The two electrical parameters are varied with saturated zone of the inner turn of tape. Then, the simulated model of a core snubber is set up. Finally, a miniature core snubber is analyzed, and its parallel resistance and inductance are obtained from the experiment condition. The test data validate the analytic model. The method presented in this paper is proved to be useful in the design of core snubber for the Experimental Advanced Superconducting Tokamak and the International Tokamak Experimental Reactor.

Investigation of tritium analysis methods for ion microbeam application

The trapping and retention of tritium in deposited layers on plasma-facing components is a critical issue for the international tokamak experimental reactor (ITER) and for future power producing tokamak fusion reactors. Cross sections of deposited layers at surfaces in the JET tokamak divertor are being investigated using ion microbeam analysis. To include tritium analysis with high spatial resolution, a number of plausible ion beam techniques have been investigated. Calibration samples with 150 nm tritiated titanium films were used. Absolute concentrations were determined with classical ERD using 2.5–3.5 MeV 12C +. Cross sections for non-Rutherford ERD and for the T( 12C,p) 14C and T( 12C,α) 11B nuclear reactions were measured for different angles in the energy range 2.5–15 MeV. Background spectra were collected from pure carbon, beryllium and deuterium enriched samples and the sensitivity for microbeam NRA measurements of the tritium concentration in thick targets with predominantly Be–C–D matrix was estimated.

Electron dumps for ITER HNB and DNB beamlines

The negative ion accelerators that produce the high-energy particle beams for the neutral injection systems for the International Tokamak Experimental Reactor (ITER) also produce unwanted particles such as electrons. These electrons are emitted in a wide angular spectrum that allows some of them to directly intercept sensitive beamline components such as the cryogenic pumps. As the electrons are also subject to backscattering, indirect interception always occurs. In this article the electron spectra produced by the Heating Neutral Beam (HNB) and Diagnostic Neutral Beam (DNB) accelerators are calculated. It is shown that these are very different. It is proposed to install electron dumps in the beamlines to intercept electron power directed towards inconvenient places in the HNB and DNB beamlines.

Physics design of a 100 keV acceleration grid system for the diagnostic neutral beam for international tokamak experimental reactor

This paper describes the physics design of a 100 keV, 60 A H(-) accelerator for the diagnostic neutral beam (DNB) for international tokamak experimental reactor (ITER). The accelerator is a three grid system comprising of 1280 apertures, grouped in 16 groups with 80 apertures per beam group. Several computer codes have been used to optimize the design which follows the same philosophy as the ITER Design Description Document (DDD) 5.3 and the 1 MeV heating and current drive beam line [R. Hemsworth, H. Decamps, J. Graceffa, B. Schunke, M. Tanaka, M. Dremel, A. Tanga, H. P. L. De Esch, F. Geli, J. Milnes, T. Inoue, D. Marcuzzi, P. Sonato, and P. Zaccaria, Nucl. Fusion 49, 045006 (2009)]. The aperture shapes, intergrid distances, and the extractor voltage have been optimized to minimize the beamlet divergence. To suppress the acceleration of coextracted electrons, permanent magnets have been incorporated in the extraction grid, downstream of the cooling water channels. The electron power loads on the extractor and the grounded grids have been calculated assuming 1 coextracted electron per ion. The beamlet divergence is calculated to be 4 mrad. At present the design for the filter field of the RF based ion sources for ITER is not fixed, therefore a few configurations of the same have been considered. Their effect on the transmission of the electrons and beams through the accelerator has been studied. The OPERA-3D code has been used to estimate the aperture offset steering constant of the grounded grid and the extraction grid, the space charge interaction between the beamlets and the kerb design required to compensate for this interaction. All beamlets in the DNB must be focused to a single point in the duct, 20.665 m from the grounded grid, and the required geometrical aimings and aperture offsets have been calculated.

Application of a SVD-Based Fast Technique for the Analysis of 3D Instabilities of Fusion Plasmas

This paper illustrates the application of a fast SVD-based technique to the analysis of resistive wall modes in fusion devices. The computational cost scales almost linearly with the number of discrete unknowns, allowing the solution of large scale 3D problems of interest for existing or future fusion devices like the International Tokamak Experimental Reactor.

Fusion

Fusion works. It powers the Sun and the other stars, and the Joint European Torus (JET) at Culham in the UK has produced 16 MW of fusion power. Fusion has many potential advantages, including essentially limitless fuel, no carbon dioxide or other emissions, and intrinsic safety. Recent progress has been good and the outlook is promising. Several steps are needed before a prototype (demonstration) power station (‘DEMO’) can be brought into operation. These steps are: (1) build a power-station-size experimental device (an international tokamak experimental reactor (ITER)) and a materials test facility (i.e. an international fusion materials irradiation facility (IFMIF)), which will take 10 years; (2) run these facilities and incorporate the results into the design of DEMO—up to a further 10 years; and (3) build DEMO—up to another 10 years. DEMO could therefore be in operation within 30 years. Fusion power could follow on a significant scale, 10 or more 1.5 GW power stations, before the middle of this century. In the second part of the century, fusion could power large centres of population and perhaps be used to produce hydrogen fuel. Meeting growing energy demand (primarily driven by needs in developing countries) while reducing carbon emissions is a large and growing challenge. A portfolio approach is needed—there is no magic bullet. Given fusion's potential, it is essential that it is developed as rapidly as is reasonably possible (even if success is not 100% certain) as one of very few options available for large-scale production of base-load power.

Removal of carbon deposits in narrow gaps by oxygen plasmas at low pressure

Under a carbon-based divertor operation of the International Tokamak Experimental Reactor, tritium retention by codeposits in hard-to-reach areas represents an important concern, and an urgent need exists for cleaning techniques suitable for the in situ removal of these codeposits. The present design of divertor modules includes castellated structures, with narrow gaps prone to accumulate material. On the other hand, glow-discharge cleaning in He∕O2 mixtures is today one of the main candidates for suppressing tritiated deposits in plasma facing components. In the present work, the effectiveness of this technique for removing a‐C:H films in gaps is addressed. The authors show that a higher-than-expected efficiency, as compared to the penetration ability of plasma ions, can be obtained. Optimization of the plasma parameters for complete oxidation of the films has also been performed. Erosion rates above 3nm∕min at room temperature can easily be achieved in fully exposed films deposited in laboratory plasmas.

The path to fusion power

Fusion is potentially an environmentally responsible and intrinsically safe source of essentially limitless power. It should be possible to build viable fusion power stations, and it looks as if the cost of fusion power will be reasonable. But time is needed to further develop the technology and to test in power station conditions the materials that would be used in their construction. Assuming no major adverse surprises, an orderly fusion development programme could lead to a prototype fusion power station putting electricity into the grid within 30 years, with commercial fusion power following some 10 or more years later. In the second half of the century, fusion could therefore be an important part of the portfolio of measures that are needed to cope with rising demand for energy in an environmentally responsible manner. In this paper, we describe the basics of fusion, its potential attractions, the status of fusion R&D, the remaining challenges and how they will be tackled at the International Tokamak Experimental Reactor and the proposed International Fusion Materials Irradiation Facility, and the timetable for the subsequent commercialization of fusion power.