Recent improvements to the ITER neutral beam system design

Fusion Engineering and Design 87 (2012) 1805–1815

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Recent improvements to the ITER neutral beam system design L.R. Grisham a,∗ , P. Agostinetti b , G. Barrera c , P. Blatchford d , D. Boilson e , J. Chareyre e , G. Chitarin b , H.P.L. de Esch f , A. De Lorenzi b , P. Franzen g , U. Fantz g , M. Gagliardi d , R.S. Hemsworth e , M. Kashiwagi h , D. King d , A. Krylov i , M. Kuriyama e , N. Marconato b , D. Marcuzzi b , M. Roccella j , L. Rios c , A. Panasenkov i , N. Pilan b , M. Pavei b , A. Rizzolo b , E. Sartori b , G. Serianni b , P. Sonato b , V. Pilard e , M. Tanaka e , H. Tobari i , P. Veltri b , P. Zaccaria b a

Princeton University, Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543, USA Consorzio RFX, Euratom-ENEA Association, C.so Stati Uniti 4, I-35127 Padova, Italy c EURATOM-CIEMAT Association, Avda. Complutense 40, 28040 Madrid, Spain d Culham Center for Fusion Energy, Abingdon, Oxon. OX14 3DB, UK e ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance, France f CEA-Cadarache, IRFM, F-13108 Saint-Paul-lez-Durance, France g Max-Planck-Institut für Plasmaphysik, EURATOM Association, D-85748 Garching, Germany h Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193, Japan i Russian Research Centre, Kurchatov Institute, Moscow, Russia j L.T. Calcoli SaS, Via C. Baslini 13, 23807 Merate (LC), Italy b

h i g h l i g h t s     

Improvements to ITER accelerator voltage holding. Improvements to ITER negative ion source design. Improvements to ITER megavolt bushing. Improvements to beamline components. Accelerator design improvements.

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Article history: Received 27 June 2012 Accepted 13 August 2012 Available online 7 September 2012 Keywords: Neutral beam injector Negative ions ITER

a b s t r a c t The ITER [1] fusion device is expected to demonstrate the feasibility of magnetically confined deuterium–tritium plasma as an energy source which might one day lead to practical power plants. Injection of energetic beams of neutral atoms (up to 1 MeV D0 or up to 870 keV H0 ) will be one of the primary methods used for heating the plasma, and for driving toroidal electrical current within it, the latter being essential in producing the required magnetic confinement field configuration. The design calls for each beamline to inject up to 16.5 MW of power through the duct into the tokamak, with an initial complement of two beamlines injecting parallel to the direction of the current arising from the tokamak transformer effect, and with the possibility of eventually adding a third beamline, also in the co-current direction. The general design of the beamlines has taken shape over the past 17 years [2], and is now predicated upon an RF-driven negative ion source based upon the line of sources developed by the Institute for Plasma Physics (IPP) at Garching during recent decades [3–5], and a multiple-aperture multiple-grid electrostatic accelerator derived from negative ion accelerators developed by the Japan Atomic Energy Agency (JAEA) across a similar span of time [6–8]. During the past years, the basic concept of the beam system has been further refined and developed, and assessment of suitable fabrication techniques has begun. While many design details which will be important to the installation and implementation of the ITER beams have been worked out during this time, this paper focuses upon those changes to the overall design concept which might be of general interest within the technical community. Published by Elsevier B.V.

1. Introduction ∗ Corresponding author. E-mail address: [email protected] (L.R. Grisham). 0920-3796/$ – see front matter. Published by Elsevier B.V. http://dx.doi.org/10.1016/j.fusengdes.2012.08.001

ITER [1] is envisioned to be the first magnetically confined fusion device to release more power through fusion reactions than the

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Fig. 1. Principal components along the beam flight path of an ITER beamline. The ion source and multi-aperture multi-grid accelerator produce the energetic ion negative ion beam. The neutralizer is a gas cell vertically partitioned to reduce the conductance. The RID is the residual ion deflection unit, which employs electrostatic fields to remove positive and negative ions remaining after the neutralizer. The calorimeter can be closed to measure the beam power. A fast shutter valve to nearly eliminate gas interchange between the beamline and the tokamak except during beam injection will be located just upstream of the absolute valve. The VVPSS box provides access to a conductance path other than the beam box for pressure excursions in the tokamak vessel due to unplanned events. The bellows allows for thermal expansion and contraction at the interface between the tokamak assembly and the beamline.

power being delivered to the plasma by its heating systems, with the intended goal of Q = 10 (Q is approximately the ratio of fusion energy released to the heating power applied by the inductive current and auxiliary heating systems). It will also be capable of very long pulses of up to 3600 s. Since ITER is a tokamak [9] with a toroidal plasma configuration, it requires both a toroidal magnetic field produced by coils and a poloidal magnetic field arising from current flowing around the toroid. Because even the initial operating mode will involve plasma discharges many times longer than those which could be sustained by inductively driving the plasma current, it must have other means of driving this current. While ITER will have a number of auxiliary current drive technologies available, the circulating ion beams following magnetic field lines around the torus which are produced when energetic neutral beams are impact-ionized within the plasma, are the most efficient technique available in terms of the amount of current which can be driven with a given amount of driver power applied to the plasma, and also in terms of the current driven per unit of electrical power fed into the driver [10]. The neutral beams [2] will represent a majority of the heating power delivered to the ITER plasma, which is needed both to produce the high temperatures needed for substantial fusion reaction rates, and to induce sufficient pressure gradients to drive large amounts of bootstrap current [11]. Fig. 1 illustrates the principal components of an ITER beamline, with the exception of the bushing, which is positioned above the ion source and connects it to the transmission line coming from the power supplies, water cooling, and gas supplies. The early generations of magnetically confined fusion devices had sufficiently low plasma line densities along the injection path to permit the use of hydrogen isotope beams with energies low enough that positive ions could be converted to neutral atoms in a gas cell with acceptable efficiency. However, in order to heat and drive a current in the plasma of larger devices, neutral hydrogen isotopes having higher specific energy are required, which cannot be efficiently created by neutralization of positive ions. For this reason, more recent devices have begun to use neutral beams formed from negative hydrogen isotope ion beams, which can be converted to neutrals in a hydrogen isotope gas cell of appropriate line density with an efficiency of about 60% over a wide range

of energies up to many MeV/amu [12] as will be required by all large future fusion devices. Because hydrogen has only a modest electron affinity, 0.75 eV, producing useful current densities of negative hydrogen ions has proven far more challenging than for their positive counterparts. Additionally, the low binding energy of their outermost electron makes them much more vulnerable than H+ to premature conversion to neutrals by collisions with gas in the extractor and accelerator. This in turn has proven more constraining in terms of acceptable source operating pressures, since they determine the gas flow into the ion source and thus the gas densities in the extractor and accelerator. Additionally, the higher voltages which are generally associated with the use of negative hydrogen isotopes have meant that not only were more stages needed in the accelerators, but the acceleration gaps between successive accelerator stages have also grown much longer. The longer gaps mean that the distance over which the direction of each beamlet can be influenced by perturbations to the electric field, whether from the space charge of other beamlets, or the shapes of conductors such as the grid holders or support structures, is greater than in previous generations of magnetic fusion ion sources. Thus, the direction in which each beamlet is steered is more sensitive to the distribution and space charge of the beamlets on each side of it, as well as edge structures, than in the multiaperture positive ion sources of the first three decades of beam injectors. A multi-stage accelerator has been chosen for the ITER injectors because it appears to have better voltage holding than a single stage accelerator and fewer electrons are accelerated to high energies, leading to a better electrical efficiency and lower electron power exiting the accelerator. The downside is that steering the beamlets to a focus becomes more complex, as the direction in which a beamlet exits each aperture can depend upon its direction and alignment upon entering the aperture as usually each aperture forms an electrostatic lens. Moreover, unanticipated beamlet deflection due to imbalance in the space charge forces from the surrounding beamlets, or edge conductors, which were left out of the early accelerator modeling, can seriously degrade beam transmission through the multistage accelerator assembly. Another complication of hydrogen isotope negative ion sources relative to their positive ion counterparts is that, due to the low

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electron affinity of hydrogen, it is necessary to add cesium vapor in order to produce adequately high negative ion current densities of hydrogen isotopes at sufficiently low ion source pressures to allow a substantial majority of the beam to exit the accelerator without being stripped to neutrals before it has gone through all the lenses and reached full energy. While the mechanism through which this increases the extractable negative hydrogen current density, whilst reducing the unwanted co-extracted electrons, remains somewhat unclear, it is generally observed that the most important factor is the presence of an appropriate amount of cesium on the plasma grid, through which negative ions are extracted to form beamlets. The cesium reduces the electron work function of the plasma grid, making it easier for neutral hydrogen isotopes and positive ions impinging on the grid to be converted to negative ions, which, depending upon where they are formed, may be directly extracted or accelerated across the sheath into the source plasma, where they can charge exchange with other neutrals heading towards the extraction plane. Theoretically, the electron work function should be a minimum with a coverage of about half a monolayer of cesium, although in practice this is not clear, particularly since the cesium is not all atomic, but also in the form of oxides, nitrides, hydrides, or hydroxides. Moreover, the surface layer will contain impurities, and the density of those on the grid may vary also as the grid temperature. The cesium may also be ‘polluted’ by tungsten (in a filamented arc discharge ion source), or metal sputtered from the ion source walls. In any event, although the amount of cesium on the plasma grid is a function of its temperature, the dependence of the extracted H− current is rather weak around the optimum temperature of about 150 ◦ C for a well-conditioned RF driven ion source of the type being developed for the ITER injectors. This means that the “cooling” water for the plasma grid must itself be warm enough to sustain this temperature, yet able to carry away excess heat from the plasma discharge when the source is operating. Earlier generations of neutral beam systems for magnetic fusion devices were designed for beam pulse lengths ranging from tens of milliseconds to tens of seconds. The ITER beams, on the other hand, are specified to produce deuterium beam pulses of up to 3600 s. This is long compared to the thermal time constants of any of the system components, so everything has to be actively cooled, such that the rate of heat removal balances the rate of heat deposition into each component. Within the accelerator, there is the additional complication that significant power (a few MW) is likely to be intercepted by the extraction grid and the grids forming the accelerator, leading to thermal expansion, which must be accommodated by the mounting assembly. If successive grids expand by different amounts, then the relative aperture alignment will change, perhaps causing increased beamlet interception and altering the directions in which the beamlets are steered if electrostatic lenses are present at the shifted grids. In the presently contemplated design, the electrical gradient throughout the accelerator is constant, resulting in negligible lens strength, so the only significant lenses in this configuration are at the extraction grid and the grounded grid. Another significant difficulty of negative hydrogen isotope beams is that, due to the low electron affinity of hydrogen, it does not naturally form ion–ion plasmas with nearly equal amounts of positive and negative ions, and few free electrons, but instead ordinary electron–ion plasmas in which the greater mobility of the lighter negative charge carriers relative to the much more massive positive ions leads to the formation of a positive potential well in the plasma in order to balance the ambipolar diffusion of positive and negative charges to the walls bounding the plasma. This potential well tends to trap the negative ions, rendering their extraction from electron-ion plasmas difficult [13]. A magnetic field transverse to the extraction/acceleration direction is imposed across the portion of the ion source plasma in front of the plasma grid. Historically this has been called a filter field

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[14] because its original purpose was to shield the plasma in the extraction region from high energy primary electrons in arc discharges which would have large cross-sections for the destruction of negative hydrogen. This field had the additional consequence of impeding the diffusion of electrons to the plasma grid, thereby reducing the imbalance in the ambipolar diffusion of the positive and negative species, and reducing the depth of the resulting ambipolar potential well, thereby increasing the probability that negative ions reach the extraction plane, and decreasing the accompanying electrons [13]. The filter field has largely been formed by flowing a current of several kA through the plasma grid in some high current negative hydrogen sources. Such a field may extend downstream of the plasma grid and deflect the beamlets, producing additional steering which must be accounted for in the relative alignments of the grids. This downstream field can also be employed to control electrons in the accelerator arising from stripping of negative ions, ionization of neutrals, and secondary production at surfaces. In the ITER accelerator this field may be augmented in the accelerator for electron control by adding permanent magnets or other techniques. Despite the presence of the magnetic filter field, a significant fraction (33–50%) of the beam extracted through the plasma grid consists of electrons, which are mostly removed by dumping them onto the extraction grid by a combination of the downstream field from the magnetic filter and the dipole magnetic fields formed by implanting powerful rare earth magnets on each side of each beamlet aperture in the extraction grid, resulting in additional beamlet deflections, all of which must be accounted for in the design of negative hydrogen isotope accelerators. In ITER this can be achieved by offset aperture steering, specifically by offsetting the exit of each aperture in the extraction grid. An alternative solution using correction magnets and a ferromagnetic plate in the grounded grid is being considered, and the basic concept will be tested on the SPIDER ion source test facility to be built at RFX Consorzio. Due to these and other differences between the ITER beams and earlier generations of positive and negative ion neutral beam systems, the design and development challenges have been significant. This paper describes some of the efforts to improve and further define the design over the past 1–2 years by the many parties participating in the design and development of the ITER neutral beams. While many design details which will be important to the installation and implementation of the ITER beams have been worked out during this time, this paper focuses upon those changes to the overall design concept which might be of general interest within the technical community.

2. Ion source progress 2.1. Ion source research and development IPP Garching has operated three test stands [3,15,16] to understand and improve the operating modes of cesiated RF-driven negative ion sources [3–5,17]. Among the recent results are that good negative ion operation was achieved with the body walls of the source on the BATMAN facility at 35 ◦ C [18]. It appeared that earlier results suggesting the need for higher temperatures may have been attributable to poor vacuum conditions rather than intrinsic source physics. This is a positive development, as it means the ion source body water at ITER will not need an additional heating circuit. The negative ion sources on both the MANITU and RADI facilities were able to operate at a pressure of 0.2 Pa, which is a third lower than their normal pressure of 0.3 Pa, which is itself the ITER requirement. MANITU, however, experienced lower extractable negative ion currents, and higher co-extracted electron fractions during the

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low pressure operation, which may be linked to inadequate control of the cesium during the long pulses. RADI was not designed for beam extraction, so similar measurements were not available across its larger area, but it did obtain good plasma uniformity and operating conditions using a magnetic filter field that was topologically similar to that planned for the SPIDER facility (originally suggested by IPP and elaborated upon by Consorzio RFX), which will test beam extraction from a full size ion source for full pulse length. While the ITER specifications can be met at an ion source operating pressure of 0.3 Pa, if it should prove practical to obtain sufficient ion current and tolerably low co-extracted electrons through design and operations improvements, this would reduce stripping losses of the beam in the accelerator, and therefore also power loading on the accelerator grids and downstream components. One of the RADI test stand’s principal missions was to understand and optimize the RF circuits of the ion source drivers when more than one driver is used to illuminate a common expansion region. Previously problems had been encountered with crosstalk between driver units, which impeded the flexibility to externally control plasma conditions. The RADI team has now shown, however, that the addition of cylindrical electrostatic screens around each RF driver eliminated cross-coupling. This should enhance the ability to produce uniform conditions across the extraction plane along the vertical direction in large multi-driver RF sources, such as the eight-driver ITER sources, and it should also enhance the capability to obtain uniform extractable D− current density across the faces of these large sources. The RF drivers for the ITER source consist of four pairs of drivers, with each pair in a horizontal configuration driven by its own RF source, and with the four pairs stacked vertically. This is a significant result, since the achievement of uniform extractable negative ion current density from large area sources was one of the principal remaining issues, and while the recent RADI work does not settle the issue, it goes a very significant way towards providing reassurance that it can be straightforwardly handled. 2.2. Engineering design of the ion source The most significant materials issue for the ITER ion source which had earlier been identified was how to manufacture large ion source components, such as the four plates which comprise the backplate of the ion sources for SPIDER, the full size ion source test facility, MITICA, the full size ITER beamline test facility, and ITER, which have thick (about a mm) molybdenum coatings bonded to copper plates. The molybdenum is needed in order to suppress the copious sputtering that is engendered when copper is in places where it can be struck by significant energetic ion fluences, and a thick molybdenum layer is needed on the back plate to cope with the very high (≈40 MW/m2 ), but very localized (<1 mm diameter), power density due to back streaming positive ions from the accelerator. Enquiries several years ago by Consorzio RFX engineers revealed that no company was confident that the backplate of an ITER ion source could be produced as a single piece if it required a thick layer of molybdenum bonded to copper. In response, the design was changed such that the back plate consisted of four separate plates. Even this less demanding design required demonstration of an acceptable fabrication technique. To this end, RFX Consorzio has carried out tests of bonding techniques. The original candidate was plasma spraying of molybdenum onto copper, a long-established industrial technique that is suitable for coating many types of dissimilar materials. While this was readily able to produce uniformly thick (1 mm) layers of molybdenum firmly bonded to the copper substrate, it was observed that the coating was highly porous, leading to excessive out-gassing under vacuum conditions, which was not in conformity with ITER guidelines. Subsequently, an explosive bonding fabrication technique

was tried, resulting in fully acceptable components consisting of a millimeter of solid molybdenum joined to copper. This is now the planned manufacturing technique for molybdenum-coated components, provided the tests of the bonding to be carried out on the GLADIS high heat flux test facility at IPP, Garching, are positive. Because hydrogen has an electron affinity of only 0.75 eV, and thus does not naturally lend itself to the formation of ion–ion plasmas in the way that halogens do, hydrogen plasmas require a magnetic field transverse to the direction of extraction to impede electron flow to the extraction plane, especially the flux of energetic electrons which could destroy the fragile D− . After calculating the resulting magnetic fields from a wide assortment of permanent magnet configurations, configurations with the field produced by current flowing along the plasma grid, and combinations of the two, along with various spatial arrangements of the current return paths for the plasma grid current, and then comparing the results with some measurements, Consorzio RFX have concluded that the most uniform magnetic filter field can be obtained with the filter field arising primarily from current flowing along the grid, but with carefully located current return paths. After further optimization, a design has been found in which the current returns located at the back of the source are compatible with the layout of the RF plasma driver array [19–26].

3. Accelerator progress 3.1. Development of physics basis for accelerator design JAEA has been carrying out a number of tasks for ITER in an effort to ensure that all relevant physical processes are accounted for in the design of the ITER accelerators. In particular, the space charge repulsion between beamlets and other beamlets and groups of beamlets is being modeled using 2D and 3D codes. The interaction of the beamlets with the electric field distortions produced by edge structures is also being modeled, as is the interaction of the beamlets with the magnetic field from the plasma grid magnetic filter. These interactions all produce steering of the beamlets, and are important to account for accurately, as they determine the overall structure, uniformity, and defocusing of the beam envelope. None of these effects were included in the original design of the JT-60U negative ion neutral beam system, leading to many problems in the accelerators of those beams. Their inclusion in the ITER design should result in a first version of the accelerator which behaves much closer to expectations than previous designs did [6–8]. The horizontal focusing of the beam envelope is to be controlled by offsetting grid apertures horizontally by small amounts so as to shift the electric field structure to cause net deflections of beamlets. In the vertical dimension, the individual beamlets are not steered, but the beam profile is concentrated by tilting the four grid sectors in each vertical column so as to bring about an overlap at 25.4 m from the exit of the accelerator, which is the position of the opening into ITER. Originally it was intended to do the horizontal steering of the beamlets by offsetting the apertures in the grounded grid. This type of arrangement had encountered some problems in the original JT-60U half-megavolt accelerator because the space charge forces between beamlets, which act in opposition to beam envelope compression, were not taken into account when the accelerator was designed. With the advantages of modern beam simulation programs, all relevant forces can be included in the calculations. However, the practical implementation of horizontal steering through aperture displacement is itself dependent upon constraints imposed by the engineering design. Following experiments at JAEA which showed improved voltage holding with larger gaps between grids, and between their support structures, it is expected, as discussed below, that the acceleration gaps will

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be somewhat longer than originally planned. This is likely to have the unintended consequence that all the acceleration gaps are of equal length. Obviously the smallest gap, the last one, has to be sufficiently long to hold the voltage, and if, as in ITER, equal voltage differentials are applied across each accelerator gap, then all the upstream gaps would have to be successively larger in order to produce the differences in field strength which produce lenses which can be exploited for steering and beamlet control. This results in an unacceptably long accelerator with increased beam loss due to stripping, and beamlet expansion in the longer gaps. As the acceleration gaps are to be of equal length, to a first approximation the acceleration grids do not form electrostatic lenses. However, recent design studies have found that because the grids are not electrostatically thin, the entrance and exit of each aperture form electrostatic lenses, of opposing polarity, and offsetting those with respect to the beamlet passing through the aperture gives steering at the entrance and exit of each aperture. The present status is that offsetting of the exit of the extraction grid apertures (or possibly compensation magnets plus ferromagnetic material in the grounded grid) will be used to compensate for the deflections caused by the electron suppression magnets, which are embedded in the extraction grids, and that inclined apertures in all the acceleration grids except the last, grounded grid, plus offsets of the apertures in the grounded grid, which does form a conventional electrostatic lens, will be used to steer the beamlets in the horizontal plane. This requires drilling the apertures in the upstream acceleration grids such that the beamlet trajectory passes through the center of each aperture at its midpoint in order to keep the beamlet interception to an acceptable level. Studies of the beamlet optics concluded that the minimum divergence at optimum perveance was reduced slightly (a change of less than 0.5 mrad), by decreasing the thickness of the grids from the previous 20 mm to about 10 mm. This was attributed to better field penetration into the apertures so that there was not a dead region with loss of control over the beam. It was calculated, and shown experimentally, that reducing the grid thickness reduced the secondary electrons produced by stray particles hitting the aperture walls. Since the power loads to the grids from the beamlets arrive on the upstream side only, they produce a temperature gradient and differential expansion across the thickness of each grid. Calculations implied that with a 10 mm thick grid this would drive a significant out-of-plane distortion of up to 9 mm, and that the thickness needs to be about 17 mm to reduce this distortion to acceptable levels (<2.4 mm). The diameter of the entrance to the apertures in the extraction grid was also increased from 11 mm to 13 mm and the exit of those apertures was increased from 15 mm to 17 mm so that the beamlets did not encounter the field aberrations near the aperture edges, thus reducing the single-beamlet divergence. A variety of different accelerator gaps were studied, resulting in the conclusion that gaps between acceleration stages greater than 90 mm resulted in increased beamlet divergence. The present proposal is to use acceleration gaps of 85 mm, which is larger than the earlier set of gaps, in order to ensure robust voltage holding. This increase in the gap length will slightly increase the gas line density in the accelerator, but as the lateral pumping is also improved, the stripping which has been calculated (about 30%) does not change significantly. Additionally, the larger the gaps, the more sensitive each beamlet is to the space charge of surrounding beamlets and to structural perturbations to the field at greater distances, such as the support structure. Recently the megavolt test facility at JAEA, Naka, Japan demonstrated beamlet steering which adequately compensated for space charge repulsion between beamlets and for magnetic deflection, and incorporated some of the techniques discussed in Section 3.3 for improving voltage holding, The most significant of these was increasing the acceleration gaps to 120 mm, which also allowed

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more distance between the beamlets and the electric field from the supporting structure, and between parts of the structure at different potentials. With these enhancements, it accelerated H− beams of 0.94 MeV at 190 A/m2 and 0.98 MeV at 185 A/m2 [27–31]. 3.2. Beamlet halo research and mitigation An experimental program [32] at the Institute for Magnetic Fusion Research (IRFM) at Cadarache, France determined the origin of much of the halo of high divergence ion trajectories which often surround the core beam emerging from H− and D− extractors. They found that there was a negligible accompanying halo when the ion source was run without cesium. Under these conditions hydrogen negative ions are produced by dissociative attachment of low energy electrons to highly excited rovibrational levels of hydrogen molecules. Because this volume production process cannot yield sufficient extractable current densities of negative hydrogen ions at pressures which are tolerably low to allow most of the beam to traverse a fusion beam accelerator unstripped, all sources presently employed for magnetic confinement fusion research add cesium, primarily to coat the plasma grid, where the ions are extracted, with (very approximately) a half monolayer of cesium to reduce the electron work function of the surface. It is thought that the dominant negative ion production process in cesiated sources occurs at cesiated surfaces, where incident H0 or D0 pick up an electron from the low work function surface. Because the survival distance is usually short compared to source dimensions, it is mostly negative ions produced at the plasma grid which contribute, either directly or indirectly, to the extractable negative ion flux. One possible explanation for the halo was that some cesium migrates onto downstream parts of the plasma grid apertures such that when neutral atomic hydrogen strikes those surfaces, the resulting negative ions emerge across the field at the edge of the apertures, acquiring relatively large velocity components transverse to the beam propagation axis, and thus engendering a high divergence halo. When the single aperture ion source used in the halo experiment was run with cesium, and with a typical plasma grid aperture, which has a sloped region on the side facing the extraction grid so as to shape the electric field, the experiment found a halo carrying 8% of the beam particles, whereas in the absence of cesium the same configuration had produced no detectable halo. This appeared to confirm the hypothesis that the halo formed due to surface production in the extraction apertures. A new plasma grid aperture was tested which had no slope on the side facing the extraction grid. When operated with cesium, this aperture produced a beam with no measurable halo. While this result is of great value in understanding and ameliorating the halo, there may still be some halo in practical designs. The halo-free aperture did not have any shaping facing the extraction grid to shape the extraction field, and in some form this shaping is probably necessary in an accelerator that is designed to have the optimum beamlet divergence. Moreover, it was noted by the experimenters that the electron suppression magnetic field in single aperture experiments produced an asymmetry in the beamlet profile which, in a multibeamlet beam, such as those for ITER, would give the appearance of a halo consisting of 2% of the beam. 3.3. High voltage holding Holding high voltage across gaps, whether spanned by vacuum, gases, or insulators is a fundamental challenge in any accelerator design. Despite the fact that electrical breakdowns are important in most aspects of electrical engineering, and are a major cost driver in many designs, the physics governing voltage holding is, at best,

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poorly understood. Based upon a compilation made by ITER neutral beam staff of voltage holding experience within the magnetic fusion energy neutral beam community, a preferred gap length in vacuum of at least 915 mm was adopted for holding a potential difference of a megavolt, and a maximum electric field gradient of about 3 kV/mm near surfaces in megavolt gaps. Due to the long-known but little-understood fact that, for lower voltages and shorter gap lengths, higher electric field gradients are sustainable without breakdown, somewhat higher gradients may be tolerated in lower voltage gaps [33]. These criteria are intended to ensure that the megavolt beam system will be robustly safe against vacuum breakdown both in the support structure and surroundings of the accelerator. The gap length criterion led to a redesign of the beam source vessel so as to allow a minimum distance of 915 mm between the beam source and the connections between the high voltage bushing and the beam source and the vessel when the source is aimed such that the end of the source assembly is closest to the vessel. This comprised a substantial increase in the minimum distance beyond the previous 700 mm, and necessitated an increase of 100 mm in the height of the beam source vessel. In general, the design of the accelerator and its support structure is being changed to reflect the fact that the magnitude of the peak electric field at surfaces is much more important in determining the vulnerability to discharges than is the average electric field across the gaps between the surfaces. The design is also being changed to take into account that the electric fields in the accelerator support structure are just as important in determining robustness against breakdowns as are those in the accelerator. The maximum electric field at surfaces can be reduced both by increasing the distance between surfaces at different potentials and by increasing the minimum radius of curvature of component structures of the ion source, accelerator, and their environs. To this end, the curvature of many structures within the accelerator and the exterior of the ion source are being increased, either by reshaping component details or by enshrouding them in smooth covers with large radii of curvature. As part of this effort to adopt more conservative standards for voltage holding, several desirable practices are being incorporated into the accelerator and source design. The heads of bolts, which by nature have small-radius corners, are being moved to the anode sides of accelerator grid support structures, and wherever possible, they are being covered by gently-curved surfaces. Where practical, structures such as corona rings are being reshaped so that their cathodic surfaces (those on the more negative side of a gap), have higher radii of curvature than do their anodic surfaces (those on the positive, or less negative, side of a gap), thereby making the most efficient use of the available space envelope. Among the practices adopted to ensure robust voltage holding by the ITER neutral beam systems, priority is given to minimizing electric field gradients on the cathode, or more negative, side of gaps, since it is from these surfaces that one would expect field emission to be initiated. Spontaneous field emission from microprojections on surfaces, or from macroscopic projections if they are present, is thought to constitute the most ubiquitous instigator of high voltage arcs. Nonetheless, some models of high voltage breakdown postulate the emission of charged clumps or ions from surfaces as the precursors, and thus introduce the possibility that surfaces at anodic, or more positive, potential, might also initiate an arc. To mitigate the risk of such phenomena, the anode sides of gaps in the accelerator are also being carefully shaped to minimize the surface electric field gradients. The mechanical and electrical feasibility of adding an electrostatic shield around the ITER and MITICA ion source so as to break up the now-large (900–915 mm) gap between the source and the beam vessel into two gaps of the order of half the distance and

half the electric potential change is being explored. If practical and compatible with the tilting range of the source, this could smooth the equipotential lines and, more importantly, take advantage of the fact that, generally speaking, smaller vacuum gaps can sustain higher electric field gradients without breakdown. There may be a low magnetic field in the gap between the beam source and the vacuum vessel wall, produced, for example, by the currents in the plasma grid current busbar system, which are necessary for the production of the transverse magnetic field in the negative ion source and the accelerator. Such a field might cause a degradation in the voltage holding across the gap, so an experimental campaign is to be carried out at the University of Padua to assess the effect of such fields upon voltage holding. 3.4. Engineering design of the accelerator Numerous design studies [34,35] have been underway by Consorzio RFX and their partners to explore the many tradeoffs that must be made to optimize the accelerator design for MITICA and the ITER heating beams, which will be identical in design. A portion of these studies is evaluating different magnetic filter field configurations within the accelerator to dump electrons arising from co-extraction from the source, beam stripping, ionization of the gas in the accelerator, and secondary electrons. Generally speaking, it is desirable to deposit the electrons onto the nearest available grid as soon after they are born as possible, so as to minimize the amount of energy they draw from the acceleration field, so long as this does not produce an excessive concentration beyond the local cooling capability of the impacted surface. The original plan was to use just the downstream magnetic field produced by the plasma grid filter field current to deflect electrons, but this appears to face difficulties because the magnetic field from the return current busbars weaken the field in the accelerator. de Esch of the IRFM, Cadarache and the design team at Consorzio RFX have proposed a class of novel alternative designs which incorporate permanent magnets into all the accelerator grids except the ground grid, oriented so as to produce deflection of electrons in several alternating sequences, and which in some versions also incorporate ferromagnetic inserts in the accelerator grids to act as field clamps [36]. There are also engineering tradeoff studies underway on the absolute length and relative lengths of accelerator gaps, as well as grid thicknesses, taking into account the beamlet steering requirements discussed in Section 3.1. These studies are in turn coupled with design goals of increasing the vacuum conductance of the grids in the direction transverse to the acceleration direction so as to reduce the gas line density along the beam flight path within the accelerator, and thus reduce beam stripping. All of these tradeoffs are being evaluated, so that a balanced engineering design can emerge. Since there are several competing design considerations which determine the spacing between accelerator grids, the RFX engineering team is trying to arrive at a design for the MITICA accelerator which will allow each acceleration gap to be adjustable over a considerable range, such as 80–100 mm, by adding or subtracting shims in the frames of the support structures. This general technique has been found to be tractable on a smaller scale in previous generations of positive ion sources at several laboratories, although the larger gap range planned for MITICA might require different engineering solutions than were employed in those cases. RFX have also conducted tests of friction welding between stainless steel tube samples and copper similar to that which will be used for the extraction grid. While friction welding is a common technique in industry for mass-producing welds between dissimilar materials, it requires extreme uniformity of materials, and requires significant time to find the appropriate tooling conditions for a particular joint, rendering it less adaptable than might first

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appear to applications using special materials produced in relatively small batches, and incorporated into joints which are only produced in quantities of dozens rather than millions. In tests with several batches of stainless steel, Consorzio RFX found that each nominally identical batch behaved differently when friction welded to copper was attempted, resulting in the destruction of an unacceptable fraction of the test pieces. As a result, they have decided to investigate different joining techniques, including non-welding processes, which are much more forgiving of minute differences in compositions from batch to batch, and do not require the exquisite adjustments of the tooling (such as setting the resistance and time profile of the clutch release which holds the initially stationary side of a friction weld) necessary in friction welding. Consorzio RFX is also considering an electron beam welding process, with the option to use a nickel insert between the copper and the stainless steel, which is welded to both, a technique which has been extensively used on Tore Supra at the IRFM for hundreds of welds on the floor limiter with 100% success.

4. High voltage bushing design and development The bushing transmits across the boundary between SF6 in the high voltage transmission line and vacuum everything needed for the operation of the ion source and accelerator [37] except the cooling water for the grounded grid. It carries all of the electrical power, at various voltages needed for the ion source and accelerator, along with cooling water, diagnostic lines, and H2 or D2 to fuel the plasma discharge. One of the major changes is the pressure of the dry air between the alumina ceramic, which faces the primary vacuum of the injector, and thus of ITER, and the fiber reinforced epoxy insulator, which carries the main longitudinal mechanical load of the structure, and which faces SF6 insulation gas. Dry air is used as a guard gas to prevent SF6 infiltration into the vacuum in order to protect the tritium plant from being damaged. In the earlier design, the pressure of the dry air was 1 MPa, while the SF6 was at a pressure of 0.6 MPa, resulting in a net radial load of 0.4 MPa. Subsequently, it has been found that optimizing the shape of the electric field relaxation rings near the flanges joining the five stages of the bushing results in sufficient reduction in the electric field gradients to allow the dry air pressure to be reduced to 0.6 MPa, the same as the SF6 . This in turn gives rise to less radial mechanical stress across the ceramic inner insulator of the bushing, and none across the epoxy outer insulator during operation, and also reduces the radial loads during initial pressure testing at elevated pressures to demonstrate the robustness of the design. Experimental tests are planned by JAEA to ascertain if the pressure of the dry air can be reduced to <0.6 MPa, which is desirable as it would allow the monitoring of the interspace pressure to be used to detect leaks into the interspace of SF6 , or into the vacuum of the dry air. Calculations of the heating within the dry air due to the dark current induced by neutrons and gammas from the accelerator, beamline components, and tokamak, have led to the conclusion that it will be low enough (<1 W) for the heat to be effectively removed through convection to the water cooled metal flanges, without the need for active circulation of the air. This significantly simplifies the design of the bushing flanges, since they now do not need passages for the air. The additional heat load due to the dark current produced by X-rays from the accelerator is still being assessed, but is not expected to change the total heat load enough to require circulation of the air. To ensure the integrity of the bushing seals over long term exposure to SF6 and high ionizing radiation, it has been concluded that the ITER bushing should have permanent epoxy bonding between the fiber-reinforced epoxy and the stainless steel flanges which

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separate each of the bushing stages, instead of the Viton O-ring previously envisioned. At least one likely candidate for the epoxy bond has been identified: the epoxy which was used to bond epoxy insulators to stainless steel flanges in the accelerators of the TFTR and DIII-D positive ion systems. This bonding material has a history of robust trouble-free reliability across 25 years of operations, during which time it was exposed to SF6 and its breakdown products for cumulative totals of years, and to much more mechanical stress than will be required of the bonding in the ITER bushing application. On TFTR, the accelerators were mounted parallel to the ground, so that the force of gravity was across the accelerators, putting the top of each segment in tension, and the bottom in compression, and the sides were also in various degrees of tension. The orientation with respect to gravity was somewhat different on DIII-D, but still resulted in a substantial amount of each bond being in tension. These configurations on TFTR and DIII-D thus were more mechanically stressful than the ITER bushing configuration, where the bonds will all be in compression due to the vertical stacking of the bushing segments, but no mechanical problems were encountered with the TFTR or DIII-D configurations. Some of the TFTR accelerators have continued to be used on NSTX. Particularly on TFTR, the epoxy was also exposed to appreciable neutron radiation, although much less than will be the case for the ITER bushing, and its tolerance of the expected ITER fluence requires further validation. The internal structure of the high voltage bushing consists of nested layers of concentric screen shields at the potentials of each of the five stages of the accelerator. They act as conducting leads to carry the electrical current associated with each accelerator stage and they also smooth the electrical field arising from pipes carrying water or gas to the different stages. Fig. 2 shows the general structure of the high voltage bushing. The detailed design of the screen shield is being conducted, including the evaluation of mechanical resonances to ensure that it does not vibrate due to the excitation of normal modes by the flow of water through the pipes it contains. Tests are planned for 2012 to test the proposed screen shield structure for mechanical vibrations under realistic water flow conditions, allowing the modification of the structure or its supports, if needed, to damp resonances. The voltage holding characteristics of the nested screen shields will also be tested. Voltage holding tests across two sample stages of the megavolt bushing reached stable operation at 370 kV before further tests were terminated by the 2011 East Japan earthquake. A single stage of the bushing was tested under conditions similar to those which will be required of the ITER bushing for 5 h at 220 kV and for 1 h at 240 kV, corresponding to the same gradient which would be across the whole five stage bushing if it were holding a total voltage of 1.1 MV and 1.2 MV respectively, offering further reason for confidence. Since the radiation and tritium requirements will be much less severe for the tests of the full beamline at the MITICA facility at Consorzio RFX, the MITICA bushing employs O-ring seals in order not to delay the schedule. There is also a good chance that the voltage holding tests of the bushing sample piece will have resumed and been completed by the time it is of interest to operate MITICA at a megavolt.

5. Beamline progress 5.1. Engineering development of beamline components Flow tests have been conducted on an experimental rig to tune and validate the design for the flow of 80 K helium through pathways similar to those that are expected to be used in the infrared radiation and gas thermal accommodation shield between the main cryopumps and the beamline environment. These pumps have both

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Fig. 2. Diagram of the one megavolt bushing which connects the transmission line from the power supplies, diagnostic readouts, water, and gas to the ITER ion source.

cryocondensation and cryosorption modes. Only one side of the 6.5 K panels will be coated with carbon, so during hydrogen operation the carbon pumps the H2 by sorption, but in deuterium both surfaces condense the D2 , and during deuterium–tritium operations, the carbon will also adsorb helium ash entering the injectors. The Culham Center for Fusion Energy (CCFE) has continued the design of the beamline components, resulting in more robust arrangements of the swirl tubes in the calorimeter which will be more stable in operation. They are currently exploring several options for permanently fixing in place the steel ribbon spirals which impart a radial force component to the flowing water to break up sheet boiling so as to improve heat transfer. Among the methods under consideration and testing are welding the ends of the ribbons to their channels, along with the possibility of also chilling the spiral ribbons to cryogenic temperatures for insertion into channels, and then letting their thermal expansion when they warm up lock them in place. The latter of these methods is the inverse of, but otherwise functionally similar to, the swaging technique which was used to fabricate early swirl tubes 40 years ago. In those early designs, inconel ribbons at ambient temperature were inserted into straight copper tubes, which were subsequently stretched to draw them down into a tight fit over the ribbons. Recent engineering innovations by CCFE include designing high heat flux elements for the residual ion dump which incorporate four parallel swirl tubes into each element and investigating gundrilling as a fabrication technique for drilling long straight cooling channels through components such as the high heat flux elements for the residual ion dump and the duct liners. Gun-drilling, which allows the lead bit of a flexible drill to find a perfectly straight path down a long stretch of metal, sounds like it would be difficult to control, but in fact has a centuries-long history in the fabrication of firearms.

5.2. Gas density profile in the beamline Monte Carlo calculations have shown that one baffle after the downstream end of the neutralizer is adequate for controlling the line density along the beam flight path from the ion source. In particular, using two baffles, with the first one near the upstream end of the neutralizer, does not appreciably reduce the line density in the accelerator or in the region between the accelerator and the neutralizer. This is because most of the gas in the accelerator, which causes stripping of some of the negative ions to neutrals and concomitant grid loading, comes either directly from the source or directly from the neutralizer. Similarly, the extra baffle adds little value to that of the downstream baffle in reducing the line density in the beam path downstream of the latter baffle, which causes reionization and eventual loss to the duct walls, of a portion of the neutral beam. 5.3. Gas density profile in the duct Present estimates for the gas efflux from some classes of ITERs high fusion performance plasmas are a factor of 20 higher than those previously envisioned. While these high edge recycling discharges will not necessarily constitute a large fraction of ITER plasma shots, it is presently thought that they may be well-suited to attaining large fusion energy gains for the Q = 10 operating phase of ITER. Thus, it is important that the heating beams be capable of operating under these conditions. Detailed calculations have now been performed of the resulting gas density profile along the long ITER beam duct, and the resulting loss of beam due to re-ionization, including the extra line density arising from gas desorbed by the re-ionized beam hitting the duct liner. There are two consequences of the tokamak operating conditions with high gas efflux; about 1 MW of the 16.5 MW from each neutral beam that would be injected into ITER when operating at

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Fig. 3. The simplified coil configuration presently envisioned for the active magnetic field compensation system, consisting of three coils above and below the beamline.

full deuterium voltage and power is lost, thus reducing the power delivered to the tokamak plasma, and this 1 MW of beam power is deposited on the duct liner. In any event, the additional re-ionized beam power is spread out over a substantial area, and does not concentrate to the degree that occurred in previous systems, such as the JET beamlines. The inductive portion of the current drive in tokamaks comes from a transformer effect, with the primary loop incorporating either an iron or air core. JET was an iron core tokamak, leading to sharper stray field gradients than occur around air-cored tokamaks such as ITER. Since water molecules, oxygen, and hydroxyl ions all have ionization cross sections of the beams substantially larger than those of hydrogen molecules, it has recently been concluded that the duct region should be monitored for water, with this added to the interlock tree. This should be possible to implement by spectroscopically monitoring a strong oxygen line. While access for the monitor could in principle be done with an optical fiber during the hydrogen/helium initial operation phase of ITER, some other transmission system, such as mirrors, will be required in order to survive the neutron fluence during the deuterium and deuterium–tritium phases of the operation. Because of the cross sections for neutralization of H− on H2 and of re-ionization of H0 on H2 change with beam velocity, the gas target needed for neutralization of 870 keV H− is ≈40% higher than that required to neutralize 1000 keV D− , with a consequent increase in the gas flow into the neutralizer. However, the gas target for re-ionization is dominated by the high edge efflux from the tokamak, so because beam re-ionization is a declining function of beam velocity, plasma discharges with high edge efflux will have somewhat less effect, by ≈40%, upon the heating beams during the hydrogen phase than during the deuterium phase, since 870 keV H has a higher velocity than 1000 keV D. Because the diagnostic neutral beam operates at 100 keV with H, it will have a lower velocity than the heating beams, and thus a larger re-ionization loss cross-section. The implications of high edge efflux plasma conditions upon the diagnostic beam are thus a bit more severe in terms of the percentage of the beam which will be lost, and the implications for the diagnostics which use the

beam have not been fully assessed, although it seems like a signal reduction on the order of 20% should probably be tolerable. The DNB duct is also being modified to withstand the extra re-ionized power. 5.4. Tokamak stray magnetic field exclusion from the beamline The magnetic field from the tokamak extends far beyond the coils, pervading the region occupied by the beamlines. It is essential to reduce the level of this field within the ion source and the beam flight path because it produces three undesirable effects. Within the ion source, it alters the distribution of the plasma, and therefore the uniformity of negative ion production. Within the accelerator and the beam flight path to the residual ion deflection dump, it deflects the ionized portion of the beam, resulting in a deflection of the eventual neutral beam. It also induces growth in the beam divergence because each ion is neutralized after traversing a different path-integrated-flux through the 3 m long neutralizer and the gas preceding and following it. 5.4.1. Active compensation correction coils The earlier design of the active compensation correction coils, which produce magnetic fields to cancel a large portion of the stray magnetic field from the tokamak, entailed nested coils, one of which was very large and of a complex shape, above the beam and ion source boxes, and a very large coil underneath. This has been simplified. The large coil of complicated shape was difficult both to manufacture and to install, and the large coil underneath the beamline could not be removed for repair. The new design uses three smaller coils adjacent to each other above and below the beamline. They will be easier to manufacture and install, and the coils under the beamline can be removed for repair if that should ever prove necessary. Because the coil shapes should be much easier to manufacture, it is also expected that they will cost less than the previous design. Fig. 3 illustrates the configuration of the revised active control coils.

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5.4.2. Passive magnetic shielding The passive magnetic shielding surrounds the beamline and beam source box. The active control coils, which lie above and below the shielding, only cancel part of the total stray flux from the tokamak. The passive magnetic shielding shunts most of the remaining stray flux around the beamline and beam source box. The earlier design used a single thick layer of soft iron. The new design for the heating beams consists of two concentric layers, each of which is one half of the total thickness of the former single layer, with the two layers separated by a 10 cm concentric air break, while the diagnostic neutral beam will have 3 layers separated by air breaks. So long as the magnetic reluctance (analogous to resistance in electrical circuits) of such an air break is large compared to the magnetic reluctance of a layer of the shielding for a flux line running from the centerline of the beam box to the edge, and so long as the layers do not saturate, then the magnetic field strength in each layer will decrease in going from the outer layer to the inner layer, resulting in a lower residual magnetic field inside the beam box than could have been obtained with a single layer of the same total thickness. Compared to a single layer of shielding, multiple layers can also offer much better cancellation of magnetic fields arising from electrical eddy currents if they are electrically insulated from each other, although in the case of ITER, this is probably not necessary, since the beams will be operating when the tokamak field is more or less constant, and thus not inducing eddy currents, except during disruptions. The steel specified for the shielding was also changed slightly, to one with less cobalt, so as to reduce activation. The passive shielding also acts as a neutron shield to reduce the activation of the beam hall. 6. Performance validation The designs of all the neutral beam subsystems will be validated in terms of performance at RFX Consorzio using the ITER ion source test facility (SPIDER) and the ITER neutral beam system test facility (MITICA), which will also identify necessary optimization and potential improvement. Therefore, they will be equipped with a set of diagnostics which is much more extensive than that foreseen for the ITER beams. On SPIDER an instrumented calorimeter observed by infrared thermography [38], complimented by tomography [23], beam emission spectroscopy [39] and a neutron imaging detector [40] will allow the beam to be fully characterized in terms of uniformity and divergence. The source will be monitored by emission spectroscopy [41,42], thermocouples, and electrostatic probes [20], with particular attention to the H− production and cesium dynamics using respectively cavity ring down spectroscopy and laser absorption spectroscopy [43,44]. On MITICA, beam emission spectroscopy, tomography and neutron imaging may be used to deduce the divergence and beam profile in three positions between the beam line components [45]. 7. Conclusion While the basic outline of the beam system has changed little from the conceptual design study some seventeen years ago, except for switching from an arc ion source to an RF-driven ion source, the details continue to evolve, leading to what is expected to be a robust and versatile heating and current drive system for ITER. References [1] B.J. Green, ITER International Team and Participant Teams, ITER: burning plasma physics experiment, Plasma Physics and Controlled Fusion 45 (2003) 687. [2] R. Hemsworth, et al., Status of the ITER heating neutral beam system, Nuclear Fusion 49 (2009) 045006–045021.

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