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We are very pleased to bring you this special collection of recent work published in Nuclear Fusion. The journal aims to facilitate communication between research groups, providing a forum for disseminating results, and the topicality of the journal content complements the on-going activities worldwide. This collection of papers showcases journal content in the areas of fusion technology and fusion materials, and includes articles relating to DEMO concepts, blanket issues, heating systems, ICF targets and the behaviour and testing of tungsten. The articles in this collection will be free to read until 29 February 2016.

Further information on how to access, write for, or subscribe to Nuclear Fusion can be found on the journal’s homepage at iopscience.org/nf.

We hope that you enjoy reading this selection of papers.

The Nuclear Fusion team

Welcome

Cover image: inspired by the neutral beam pattern from the ITER thermonuclear fusion reactor R Hemsworth et al 2009 Nucl. Fusion 49 045006.

4 Nuclear Fusion

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ContentsStatus of the ITER heating neutral beam system 6R. Hemsworth et al

Studies on targets for inertial fusion ignition 6 demonstration at the HiPER facility S. Atzeni et al

R&D of a Li2TiO3 pebble bed for a test blanket module 6 in JAEA H. Tanigawa et al

Compact DEMO, SlimCS: design progress and issues 7K. Tobita et al

Formation process of tungsten nanostructure by the 7 exposure to helium plasma under fusion relevant plasma conditions Shin Kajita et al

The influence of displacement damage on deuterium 7 retention in tungsten exposed to plasma W.R. Wampler and R.P. Doerner

Investigating behaviours of hydrogen in a tungsten 8 grain boundary by first principles: from dissolution and diffusion to a trapping mechanism Hong-Bo Zhou et al

Assessment of compatibility of ICRF antenna operation 8 with full W wall in ASDEX Upgrade Vl.V. Bobkov et al

Taming the plasma–material interface with the 9 ‘snowflake’ divertor in NSTX V.A. Soukhanovskii et al

Neutron-induced transmutation effects in W and 9 W-alloys in a fusion environment M.R. Gilbert and J.-Ch. Sublet

Prospects for pilot plants based on the tokamak, 9 spherical tokamak and stellarator J.E. Menard et al

ITER test blanket module error field simulation 10 experiments at DIII-D M.J. Schaffer et al

First mirrors in ITER: material choice and deposition 10 prevention/cleaning techniques E.E. Mukhin et al

Saturation of deuterium retention in self-damaged 11 tungsten exposed to high-flux plasmas M.H. J. ‘t Hoen et al

Diagnostics for plasma control on DEMO: 11 challenges of implementation A.J.H. Donné et al

Tungsten surface evolution by helium bubble 11 nucleation, growth and rupture Faiza Sefta et al

Progress of the JT-60SA project 11Y. Kamada et al

Scientific and technological advancements in inertial 12 fusion energy D.E. Hinkel

IFMIF: overview of the validation activities 12J. Knaster et al

The ITER magnet systems: progress on construction 12C. Sborchia et al

The ITER blanket system design challenge 12A.R. Raffray et al

Thermal response of nanostructured tungsten 13 Shin Kajita et al

Design concept of K-DEMO for near-term implementation 13 K. Kim et al

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Editors and Editorial Board

Journal scope

Nuclear Fusion archive

Nuclear Fusion publishes articles making significant advances to the field of controlled thermonuclear fusion.

The journal scope includes: • the production, heating and confinement of high-temperature plasmas;•the physical properties of such plasmas;•the experimental or theoretical methods of exploring or explaining them;•fusion reactor physics;•reactor concepts;•fusion technologies.

The journal has a dedicated Associate Editor for inertial confinement fusion.

Nuclear Fusion has aimed to provide a first-class forum for sharing research results ever since its first edition, five decades ago. We are delighted to offer you the complete electronic archive for Nuclear Fusion.

This archive contains more than 50 years of articles, dating back to the journal’s first issue in 1960, and includes key papers from the history of fusion research. You can browse the full Nuclear Fusion collection by visiting the archive website at iopscience.org/nf/archive*.

*access will depend on subscription status.

Editor-in-ChiefA. Fasoli, Switzerland

Associate Editor for Inertial Confinement M. Tabak, USA

Chairman of the Board of EditorsM. Kikuchi, Japan

Editorial Board P. Barabaschi, EU R. Betti, USA L. Chen, China W. Choe, Republic of Korea A.J.H. Donné, EUROfusion X. Duan, China D. Farina, Italy X. Garbet, France R. Hawryluk, USA K. Hesch, Germany K. Ida, Japan

T. Jones, UK Y. Kamada, Japan Y. Kishimoto, Japan T. Kurki-Suonio, Finland B.V. Kuteev, Russian Federation F. Romanelli, Italy J. Sanchez, Spain A. Sen, India H. Shiraga, Japan P. Thomas, ITER V. Tikhonchuk, France

Y. Ueda, Japan M.R. Wade, USA F. Waelbroeck, USA D. Whyte, USA H. R. Wilson, UK R. Wolf, Germany G. Xu, China H. Yamada, Japan M. Zarnstorff, USA S.J. Zinkle, USA H. Zohm, Germany

6 Nuclear Fusion

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Status of the ITER heating neutral beam system 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

2009 Nucl. Fusion 49 045006

Abstract 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: 1) Modifications to allow installation and maintenance of the beamline components with an overhead crane. 2) 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. 3) The RF driven negative ion source has replaced the filamented ion source as the reference design. 4) 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. 5) 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.

Studies on targets for inertial fusion ignition demonstration at the HiPER facility S. Atzeni, J.R. Davies, L. Hallo, J.J. Honrubia, P.H. Maire, M. Olazabal-Loumé, J.L. Feugeas, X. Ribeyre, A. Schiavi, G. Schurtz, J. Breil and Ph. Nicolaï

2009 Nucl. Fusion 49 055008

Abstract Recently, a European collaboration has proposed the High Power Laser Energy Research (HiPER) facility, with the primary goal of demonstrating laser driven inertial fusion fast ignition. HiPER is expected to provide 250 kJ in multiple, 3ω (wavelength λ = 0.35 µm), nanosecond beams for

R&D of a Li2TiO3 pebble bed for a test blanket module in JAEA H. Tanigawa, T. Hoshino, Y. Kawamura, M. Nakamichi, K. Ochiai, M. Akiba, M. Ando, M. Enoeda, K. Ezato, K. Hayashi, T. Hirose, C. Konno, H. Nakamura, T. Nozawa, H. Ogiwara, Y. Seki, H. Tanigawa, K. Tsuchiya, D. Tsuru and T. Yamanishi

2009 Nucl. Fusion 49 055021

Abstract At JAEA, a test blanket module (TBM) with a water-cooled solid breeder is being developed. This paper presents recent achievements of research activities for the TBM, particularly addressing the pebble bed of the tritium breeder materials and tritium behaviour. For the breeder material, the chemical stability of Li

2TiO3 was improved using Li2O additives. To analyse the pebble bed behaviour, thermomechanical properties of the Li

2TiO3 pebble bed were assessed experimentally. To verify the pebble bed’s nuclear properties, the activation foil method was proposed and a preliminary experiment was conducted. To reduce the tritium permeation, the chemical densified coating method was developed and the coating was attached to F82H steel. For tritium behaviour, the tritium recovery system was modified in consideration of the design change of the TBM.

compression and 70 kJ in 10–20 ps, 2ω beams for ignition. The baseline approach is fast ignition by laser-accelerated fast electrons; cones are considered as a means to maximize ignition laser–fuel coupling. Earlier studies led to the identification of an all-DT shell, with a total mass of about 0.6 mg as a reference target concept. The HiPER main pulse can compress the fuel to a peak density above 500 g cm−3 and an areal density ρR of about 1.5 g cm−2. Ignition of the compressed fuel requires that relativistic electrons deposit about 20 kJ in a volume of radius of about 15 µm and a depth of less than 1.2 g cm−2. The ignited target releases about 13 MJ. In this paper, additional analyses of this target are reported. An optimal irradiation pattern has been identified. The effects on fuel compression of the low-mode irradiation non-uniformities have been studied by 2D simulations and an analytical model. The scaling of the electron beam energy required for ignition (versus electron kinetic energy) has been determined by 2D fluid simulations including a 3D Monte Carlo treatment of relativistic electrons, and agrees with a simple model. Integrated simulations show that beam-induced magnetic fields can reduce beam divergence. As an alternative scheme, shock ignition is studied. 2D simulations have addressed optimization of shock timing and absorbed power, means to increase laser absorption efficiency and the interaction of the igniting shocks with a deformed fuel shell.

Nuclear Fusion 7

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Compact DEMO, SlimCS: design progress and issues K. Tobita, S. Nishio, M. Enoeda, H. Kawashima, G. Kurita, H. Tanigawa, H. Nakamura, M. Honda, A. Saito, S. Sato, T. Hayashi, N. Asakura, S. Sakurai, T. Nishitani, T. Ozeki, M. Ando, K. Ezato, K. Hamamatsu, T. Hirose, T. Hoshino, S. Ide, T. Inoue, T. Isono, C. Liu, S. Kakudate, Y. Kawamura, S. Mori, M. Nakamichi, H. Nishi, T. Nozawa, K. Ochiai, H. Ogiwara, N. Oyama, K. Sakamoto, Y. Sakamoto, Y. Seki, Y. Shibama, K. Shimizu, S. Suzuki, K. Takahashi, H. Tanigawa, D. Tsuru, T. Yamanishi and T. Yoshida

2009 Nucl. Fusion 49 075029

Abstract The design progress in a compact low aspect ratio (low A) DEMO reactor, ‘SlimCS’, and its design issues are reported. The design study focused mainly on the torus configuration including the blanket, divertor, materials and maintenance scheme. For continuity with the Japanese ITER-TBM, the blanket is based on a water-cooled solid breeder blanket. For vertical stability of the elongated plasma and high beta access, the blanket is segmented into replaceable and permanent blankets and a sector-wide conducting shell is arranged inbetween these blankets. A numerical calculation indicates that fuel self-sufficiency can be satisfied when the blanket interior is ideally fabricated. An allowable heat load to the divertor plate should be 8 MW m−2 or lower, which can be a critical constraint for determining a handling power of DEMO.

Formation process of tungsten nanostructure by the exposure to helium plasma under fusion relevant plasma conditions Shin Kajita, Wataru Sakaguchi, Noriyasu Ohno, Naoaki Yoshida and Tsubasa Saeki

2009 Nucl. Fusion 49 095005

Abstract Helium irradiation on tungsten changes the surface morphology dramatically by forming a nanometre-sized fibreform structure which could bring about serious problems for fusion reactors. From the experimental results in liner divertor simulators, it is revealed that the incident ion energy and surface temperature are key parameters for the formation of the structure. It is shown that the tungsten nanostructure is easily formed when the temperature is in the range 1000–2000 K, and the incident ion energy is higher than 20 eV. Furthermore, on the basis of the helium irradiation experiments performed in the divertor simulator NAGDIS-I, the initial formation process of the nanostructure is revealed. It is shown that the nanostructure formation is related to pinholes appearing on the bulk part of the material, and then,

The influence of displacement damage on deuterium retention in tungsten exposed to plasma W.R. Wampler and R.P. Doerner

2009 Nucl. Fusion 49 115023

Abstract Trapping of tritium at lattice damage from fusion neutron irradiation is expected to increase the tritium inventory in tungsten components in ITER. The magnitude of this increase depends on the concentration of traps that are produced, and on the depth to which the increased tritium retention extends into the material. Experiments to address these issues are described, in which displacement damage by ion irradiation was used as a surrogate for neutron damage. Irradiated samples were exposed to high flux deuterium plasma to simulate divertor conditions. The resulting deuterium content was measured by nuclear reaction analysis. Measurements were done at various damage levels up to those expected from the end-of-life neutron fluence in ITER. These experiments determine the number of traps produced by displacement damage and the rate at which they are filled during exposure to plasma. The role of defect annealing was explored through plasma exposures at various temperatures. In addition to trapping at damage, near-surface retention from internal precipitation was observed at lower temperatures. Addition of 5% helium to the deuterium plasma greatly reduced D retention by precipitation by localizing it closer to the surface. Results from these experiments indicate that the contribution to tritium inventory in ITER from trapping at neutron damage should be small.

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the rough structure develops to a much finer nanostructure. The nanostructure was also observed on the molybdenum surface that was exposed to the helium plasma. It increases interest in the possibility that nanostructure formation by helium irradiation is a common phenomenon that occurs on various metals.

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Investigating behaviours of hydrogen in a tungsten grain boundary by first principles: from dissolution and diffusion to a trapping mechanism Hong-Bo Zhou, Yue-Lin Liu, Shuo Jin, Ying Zhang, G.-N. Luo and Guang-Hong Lu

2010 Nucl. Fusion 50 025016

Abstract We have investigated the dissolution, segregation and diffusion of hydrogen (H) in a tungsten (W) grain boundary (GB) using a first-principles method in order to understand the GB trapping mechanism of H. Optimal charge density plays an essential role in such a GB trapping mechanism. Dissolution and segregation of H are directly associated with the optimal charge density, which can be reflected by the H solution and segregation energy sequence for the different interstitial sites. To occupy the optimal-charge-density site, H can be easily trapped by the W GB with the solution and segregation energy of −0.23 eV and −1.11 eV, respectively. Kinetically, such a trapping is easier to realize due to the much lower diffusion barrier of 0.13–0.16 eV from the bulk to the GB in comparison with the segregation energy, suggesting that it is quite difficult for the trapped H to escape out of the GB. However, the GB can hold no more than 2 H atoms because the isosurface of optimal charge density almost disappears with the second H atom in, leading to the conclusion that H

2 molecule and thus H bubble cannot form in the W GB. Taking into account the lower vacancy formation energy in the GB as compared with the bulk, we propose that the experimentally observed H bubble formation in the W GB should be via a vacancy trapping mechanism.

Assessment of compatibility of ICRF antenna operation with full W wall in ASDEX Upgrade Vl.V. Bobkov, F. Braun, R. Dux, A. Herrmann, L. Giannone, A. Kallenbach, A. Krivska, H.W. Müller, R. Neu, J.-M. Noterdaeme, T. Pütterich, V. Rohde, J. Schweinzer, A. Sips, I. Zammuto and ASDEX Upgrade Team

2010 Nucl. Fusion 50 035004

Abstract The compatibility of ICRF (ion cyclotron range of frequencies) antenna operation with high-Z plasma facing components is assessed in ASDEX Upgrade (AUG) with its tungsten (W) first wall. The mechanism of ICRF-related W sputtering was studied by various diagnostics including the local spectroscopic measurements of W sputtering yield Y

W on antenna limiters. Modification of one antenna with triangular shields, which cover the locations where long magnetic field lines pass only one out of two (0π)-phased antenna straps, did not influence the locally measured Y

W values markedly. In the experiments with antennas powered individually, poloidal profiles of Y

W on limiters of powered antennas show high Y

W close to the equatorial plane and at the very edge of the antenna top. The Y

W-profile on an unpowered antenna limiter peaks at the location projecting to the top of the powered antenna. An interpretation of the Y

W measurements is presented, assuming a direct link between the W sputtering and the sheath driving RF voltages deduced from parallel electric near-field (E

||) calculations and this suggests a strong E

|| at the antenna limiters. However, uncertainties are too large to describe the Y

W poloidal profiles. In order to reduce ICRF-related rise in W concentration C

W, an operational approach and an approach based on calculations of parallel electric fields with new antenna designs are considered. In the operation, a noticeable reduction in Y

W and CW in the plasma during ICRF operation with W wall can be achieved by (a) increasing plasma–antenna clearance; (b) strong gas puffing; (c) decreasing the intrinsic light impurity content (mainly oxygen and carbon in AUG). In calculations, which take into account a realistic antenna geometry, the high E

|| fields at the antenna limiters are reduced in several ways: (a) by extending the antenna box and the surrounding structures parallel to the magnetic field; (b) by increasing the average strap–box distance, e.g. by increasing the number of toroidally distributed straps; (c) by a better balance of (0π)-phased contributions to RF image currents.

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Taming the plasma–material interface with the ‘snowflake’ divertor in NSTX V.A. Soukhanovskii, J.-W. Ahn, R.E. Bell, D.A. Gates, S. Gerhardt, R. Kaita, E. Kolemen, B.P. LeBlanc, R. Maingi, M. Makowski, R. Maqueda, A.G. McLean, J.E. Menard, D. Mueller, S.F. Paul, R. Raman, A.L. Roquemore, D.D. Ryutov, S.A. Sabbagh and H.A. Scott

2011 Nucl. Fusion 51 012001

Abstract Steady-state handling of divertor heat flux is a critical issue for ITER and future conventional and spherical tokamaks with compact high-power density divertors. A novel ‘snowflake’ divertor (SFD) configuration was theoretically predicted to have significant magnetic geometry benefits for divertor heat flux mitigation, such as an increased plasma-wetted area and a higher divertor volume available for volumetric power and momentum loss processes, as compared with the standard divertor. Both a significant divertor peak heat flux reduction and impurity screening have been achieved simultaneously with core H-mode confinement in discharges with the SFD using only a minimal set of poloidal field coils.

Neutron-induced transmutation effects in W and W-alloys in a fusion environment M.R. Gilbert and J.-Ch. Sublet

2011 Nucl. Fusion 51 043005

Abstract W and W-alloys are among the primary candidate materials for plasma-facing components in the design of fusion reactors, particularly in high-heat-flux regions such as the divertor. Under neutron irradiation W undergoes transmutation to its near-neighbours in the periodic table. Additionally He and H are particles emitted from certain neutron-induced reactions, and this is particularly significant in fusion research since the presence of helium in a material can cause both swelling and a strong increase in brittleness. This paper presents the results of inventory burn-up calculations on pure W and gives quantitative estimates for He production rates in both a fusion-reactor environment and under conditions expected in the ITER experimental device. Transmutation reactions in possible alloying elements (Re, Ta, Ti and V), which could be used to reduce the brittleness of pure W, are also considered. Additionally, for comparison, the transmutation of other fusion-relevant materials, including Fe and SiC, are presented.

Prospects for pilot plants based on the tokamak, spherical tokamak and stellarator J.E. Menard, L. Bromberg, T. Brown, T. Burgess, D. Dix, L. El-Guebaly, T. Gerrity, R.J. Goldston, R.J. Hawryluk, R. Kastner, C. Kessel, S. Malang, J. Minervini, G.H. Neilson, C.L. Neumeyer, S. Prager, M. Sawan, J. Sheffield, A. Sternlieb, L. Waganer, D. Whyte and M. Zarnstorff

2011 Nucl. Fusion 51 103014

Abstract A potentially attractive next-step towards fusion commercialization is a pilot plant, i.e. a device ultimately capable of small net electricity production in as compact a facility as possible and in a configuration scalable to a full-size power plant. A key capability for a pilot-plant programme is the production of high neutron fluence enabling fusion nuclear science and technology (FNST) research. It is found that for physics and technology assumptions between those assumed for ITER and nth-of-a-kind fusion power plant, it is possible to provide FNST-relevant neutron wall loading in pilot devices. Thus, it may be possible to utilize a single facility to perform FNST research utilizing reactor-relevant plasma, blanket, coil and auxiliary systems and maintenance schemes while also targeting net electricity production. In this paper three configurations for a pilot plant are considered: the advanced tokamak, spherical tokamak and compact stellarator. A range of configuration issues is considered including: radial build and blanket design, magnet systems, maintenance schemes, tritium consumption and self-sufficiency, physics scenarios and a brief assessment of research needs for the configurations.

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ITER test blanket module error field simulation experiments at DIII-D M.J. Schaffer, J.A. Snipes, P. Gohil, P. de Vries, T.E. Evans, M.E. Fenstermacher, X. Gao, A.M. Garofalo, D.A. Gates, C.M. Greenfield, W.W. Heidbrink, G.J. Kramer, R.J. La Haye, S. Liu, A. Loarte, M.F.F. Nave, T.H. Osborne, N. Oyama, J.-K. Park, N. Ramasubramanian, H. Reimerdes, G. Saibene, A. Salmi, K. Shinohara, D.A. Spong, W.M. Solomon, T. Tala, Y.B. Zhu, J.A. Boedo, V. Chuyanov, E.J. Doyle, M. Jakubowski, H. Jhang, R.M. Nazikian, V.D. Pustovitov, O. Schmitz, R. Srinivasan, T.S. Taylor, M.R. Wade, K.-I. You, L. Zeng and the DIII-D Team

2011 Nucl. Fusion 51 103028

Abstract Experiments at DIII-D investigated the effects of magnetic error fields similar to those expected from proposed ITER test blanket modules (TBMs) containing ferromagnetic material. Studied were effects on: plasma rotation and locking, confinement, L–H transition, the H-mode pedestal, edge localized modes (ELMs) and ELM suppression by resonant magnetic perturbations, energetic particle losses, and more. The experiments used a purpose-built three-coil mock-up of two magnetized ITER TBMs in one ITER equatorial port. The largest effect was a reduction in plasma toroidal rotation velocity v across the entire radial profile by as much as Δv/v

~60% via non-resonant braking. Changes to global Δn/n, Δβ/β and ΔH

98/H98 were ~3 times smaller. These effects are stronger at higher β. Other effects were smaller. The TBM field increased sensitivity to locking by an applied known n = 1 test field in both L- and H-mode plasmas. Locked mode tolerance was completely restored in L-mode by re-adjusting the DIII-D n = 1 error field compensation system. Numerical modelling by IPEC reproduces the rotation braking and locking semi-quantitatively, and identifies plasma amplification of a few n = 1 Fourier harmonics as the main cause of braking. IPEC predicts that TBM braking in H-mode may be reduced by n = 1 control. Although extrapolation from DIII-D to ITER is still an open issue, these experiments suggest that a TBM-like error field will produce only a few potentially troublesome problems, and that they might be made acceptably small.

First mirrors in ITER: material choice and deposition prevention/cleaning techniques E.E. Mukhin, V.V. Semenov, A.G. Razdobarin, S. Yu. Tolstyakov, M.M. Kochergin, G.S. Kurskiev, K.A. Podushnikova, S.V. Masyukevich, D.A. Kirilenko, A.A. Sitnikova, P.V. Chernakov, A.E. Gorodetsky, V.L. Bukhovets, R. Kh. Zalavutdinov, A.P. Zakharov, I.I. Arkhipov, Yu.P. Khimich, D.B. Nikitin, V.N. Gorshkov, A.S. Smirnov, T.V. Chernoizumskaja, E.M. Khilkevitch, S.V. Bulovich, V.S. Voitsenya, V.N. Bondarenko, V.G. Konovalov, I.V. Ryzhkov, O.M. Nekhaieva, O.A. Skorik, K. Yu. Vukolov, V.I. Khripunov and P. Andrew

2012 Nucl. Fusion 52 013017

Abstract We present here our recent results on the development and testing of the first mirrors for the divertor Thomson scattering diagnostics in ITER. The Thomson scattering system is based on several large-scale (tens of centimetres) mirrors that will be located in an area with extremely high (3–10%) concentration of contaminants (mainly hydrocarbons) and our main concern is to prevent deposition-induced loss of mirror reflectivity in the spectral range 1000–1064 nm. The suggested design of the mirrors—a high-reflective metal layer on a Si substrate with an oxide coating—combines highly stable optical characteristics under deposition-dominated conditions with excellent mechanical properties. For the mirror layer materials we consider Ag and Al allowing the possibility of sharing the Thomson scattering mirror collecting system with a laser-induced fluorescence system operating in the visible range. Neutron tests of the mirrors of this design are presented along with numerical simulation of radiation damage and transmutation of mirror materials. To provide active protection of the large-scale mirrors we use a number of deposition-mitigating techniques simultaneously. Two main techniques among them, plasma treatment and blowing-out, are considered in detail. The plasma conditions appropriate for mirror cleaning are determined from experiments using plasma-induced erosion/deposition in a CH

4/H2 gas mixture. We also report data on the numerical simulation of plasma parameters of a capacitively-coupled discharge calculated using a commercial CFD-ACE code. A comparison of these data with the results for mirror testing under deuterium ion bombardment illustrates the possibility of using the capacitively-coupled discharge for in situ non-destructive deposition mitigation/cleaning.

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Saturation of deuterium retention in self-damaged tungsten exposed to high-flux plasmas M.H.J. ‘t Hoen, B. Tyburska-Püschel, K. Ertl, M. Mayer, J. Rapp, A.W. Kleyn and P.A. Zeijlmans van Emmichoven

2012 Nucl. Fusion 52 023008

Abstract Polycrystalline, annealed tungsten targets were bombarded with 12.3 MeV W4+ ions to various damage levels. Deuterium was implanted by high-flux plasmas in Pilot-PSI (>1024 m−2 s−1) at a surface temperature below 525 K. Deuterium retention has been studied by nuclear reaction analysis and by thermal desorption spectroscopy. We found that deuterium retention is strongly enhanced by the tungsten bombardment and that saturation occurs at a W4+ fluence of about 3 × 1017 m−2. The maximum deuterium concentration in the damaged region was measured to be 1.4 at.%. This is in accordance with other experiments that were carried out at much lower fluxes. We therefore conclude that the saturation behaviour and the maximum retention are not affected by the high fluxes used in our experiments. A simple geometric model is presented that assumes that the saturation solely originates in the tungsten irradiation and that explains it in terms of overlapping saturated volumes. The saturated volume per incident MeV ion amounts to 3 × 104 nm3. From our results, we are able to obtain an approximate value for the average occupation number of the vacancies.

Diagnostics for plasma control on DEMO: challenges of implementation A.J.H. Donné, A.E. Costley and A.W. Morris

2012 Nucl. Fusion 52 074015

Abstract As a test fusion power plant, DEMO will have to demonstrate reliability and very long pulse/steady-state operation, which calls for unprecedented robustness and reliability of all diagnostic systems (also requiring adequate redundancy). But DEMO will have higher levels of neutron and gamma fluxes, and fluences, nuclear heating, and fluxes of particles than ITER, and probably reduced physical access. In particular, the neutron fluence will be about 15–50 times higher than that in ITER. As a consequence, some diagnostics that will work in ITER are likely to be unfeasible in DEMO. It is important, therefore, to develop a new way of thinking with respect to that employed to date in which diagnostics are added after the machine has been basically designed: if certain diagnostics are deemed essential for the control of DEMO, they will have to be taken into account during the entire design phase.

Tungsten surface evolution by helium bubble nucleation, growth and rupture Faiza Sefta, Karl D. Hammond, Niklas Juslin and Brian D. Wirth

2013 Nucl. Fusion 53 073015

Abstract Molecular dynamics simulations reveal sub-surface mechanisms likely involved in the initial formation of nanometre-sized ‘fuzz’ in tungsten exposed to low-energy helium plasmas. Helium clusters grow to over-pressurized bubbles as a result of repeated cycles of helium absorption and Frenkel pair formation. The self-interstitials either reach the surface as isolated adatoms or trap at the bubble periphery before organizing into prismatic 〈1 1 1〉 dislocation loops. Surface roughening occurs as single adatoms migrate to the surface, prismatic loops glide to the surface to form adatom islands, and ultimately as over-pressurized gas bubbles burst.

Progress of the JT-60SA projectY. Kamada, P. Barabaschi, S. Ishida, the JT-60SA Team and JT-60SA Research Plan Contributors

2013 Nucl. Fusion 53 104010

Abstract The JT-60SA project implemented by Japan and Europe is progressing on schedule towards the first plasma in March 2019. After careful R&D, procurements of the major components have entered their manufacturing stages. In parallel, disassembly of JT-60U has been completed on time, and the JT-60SA tokamak assembly is expected to start in January 2013. The JT-60SA device, a highly shaped large superconducting tokamak with a variety of plasma control actuators, has been designed in order to contribute to ITER and to complement ITER in all the major areas of fusion plasma development necessary to decide DEMO reactor construction. Detailed assessments and prediction studies of the JT-60SA plasma regimes have confirmed these capabilities: using ITER- and DEMO-relevant plasma regimes, heating conditions, and its sufficiently long discharge duration, JT-60SA enables studies on magnetohydrodynamic stability at high beta, heat/particle/momentum transport, high-energy ion physics, pedestal physics including edge localized mode control, and divertor physics. By integrating these studies, the project provides ‘simultaneous and steady-state sustainment of the key performance characteristics required for DEMO’ with integrated control scenario development.

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The ITER magnet systems: progress on construction C. Sborchia, A. Bonito Oliva, T. Boutboul, K. Chan, A. Devred, S. Egorov, K. Kim, N. Koizumi, S. Lelekhov, P. Libeyre, B.S. Lim, N. Martovetsky, H. Nakajima, N. Mitchell, K. Okuno, V. Pantsyrny, W. Reiersen, I. Rodin, F. Savary, A. Vostner and Y. Wu

2014 Nucl. Fusion 54 013006

Abstract Construction of the ITER magnet systems has been started at the end of 2007 following the signature of the first procurement arrangements (PA) for the toroidal field (TF) conductors. Six ITER members are involved in the share of the ITER magnet components and, to date, eighteen PA between the ITER Organization and six domestic agencies have been signed. Substantial progress towards full-scale construction has been achieved with the placement of the first large manufacturing contracts, the production of several tens of tons of advanced Nb

3Sn and NbTi strand, and the set-up of large cabling and jacketing facilities. The detailed design of the coils and support structures has also been finalized. The qualification of the fabrication processes for the TF coils and poloidal field (PF) coils has been initiated. The detailed design of the central solenoid (CS) coils is being developed. The design of the correction coils (CCs) with their support structures has been finalized, as well as for the TF gravity supports and clamps of the PF coils. The manufacture of prototypes of the feeder lines and current leads has been started, while ITER is in charge of the procurement of the required magnet instrumentation. This paper provides a progress report on the ITER magnet construction as per December 2010.

Scientific and technological advancements in inertial fusion energy D.E. Hinkel

2013 Nucl. Fusion 53 104027

Abstract Scientific advancements in inertial fusion energy (IFE) were reported on at the IAEA Fusion Energy Conference, October 2012. Results presented transect the different ways to assemble the fuel, different scenarios for igniting the fuel, and progress in IFE technologies. The achievements of the National Ignition Campaign within the USA, using the National Ignition Facility (NIF) to indirectly drive laser fusion, have found beneficial the achievements in other IFE arenas such as directly driven laser fusion and target fabrication. Moreover, the successes at NIF have pay-off to alternative scenarios such as fast ignition, shock ignition, and heavy-ion fusion as well as to directly driven laser fusion. This synergy is summarized here, and future scientific studies are detailed.

IFMIF: overview of the validation activities J. Knaster, F. Arbeiter, P. Cara, P. Favuzza, T. Furukawa, F. Groeschel, R. Heidinger, A. Ibarra, H. Matsumoto, A. Mosnier, H. Serizawa, M. Sugimoto, H. Suzuki and E. Wakai

2013 Nucl. Fusion 53 116001

Abstract The Engineering Validation and Engineering Design Activities (EVEDA) for the International Fusion Materials Irradiation Facility (IFMIF), an international collaboration under the Broader Approach Agreement between Japan Government and EURATOM, aims at allowing a rapid construction phase of IFMIF in due time with an understanding of the cost involved. The three main facilities of IFMIF (1) the Accelerator Facility, (2) the Target Facility and (3) the Test Facility are the subject of validation activities that include the construction of either full scale prototypes or smartly devised scaled down facilities that will allow a straightforward extrapolation to IFMIF needs. By July 2013, the engineering design activities of IFMIF matured with the delivery of an Intermediate IFMIF Engineering Design Report (IIEDR) supported by experimental results. The installation of a Linac of 1.125 MW (125 mA and 9 MeV) of deuterons started in March 2013 in Rokkasho (Japan). The world’s largest liquid Li test loop is running in Oarai (Japan) with an ambitious experimental programme for the years ahead. A full scale high flux test module that will house ~1000 small specimens developed jointly in Europe and Japan for the Fusion programme has been constructed by KIT (Karlsruhe) together with its He gas cooling loop. A full scale medium flux test module to carry out on-line creep measurement has been validated by CRPP (Villigen).

The ITER blanket system design challenge A.R. Raffray, B. Calcagno, P. Chappuis, Zhang Fu, A. Furmanek, Chen Jiming, D-H. Kim, S. Khomiakov, A. Labusov, A. Martin, M. Merola, R. Mitteau, S. Sadakov, M. Ulrickson, F. Zacchia and Contributors from the Blanket Integrated Product Team

2014 Nucl. Fusion 54 033004

Abstract This paper summarizes the latest progress in the ITER blanket system design as it proceeds through its final design phase with the Final Design Review planned for Spring 2013. The blanket design is constrained by demanding and sometime conflicting design and interface requirements from the plasma and systems such as the vacuum vessel, in-vessel coils and blanket manifolds. This represents a major design challenge, which is highlighted in this paper with examples of design solutions to accommodate some of the key interface and integration requirements.

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Thermal response of nanostructured tungsten Shin Kajita, Gregory De Temmerman, Thomas Morgan, Stein van Eden, Thijs de Kruif and Noriyasu Ohno

2014 Nucl. Fusion 54 033005

Abstract The thermal response of nanostructured tungsten, which was fabricated in the linear divertor simulator NAGDIS-II, was investigated using pulsed plasma in the MAGNUM-PSI device and by using high powered laser pulses. The temperature evolution in response to the pulses was measured with an infrared fast framing camera. The temperature increase in response to the pulses on the nanostructured sample was significantly greater than that of the pristine sample both for plasma and laser pulses. After the pulsed plasma/laser irradiation, the nanostructured surface was observed to have melted although the measured surface temperature was much less than the melting temperature. The mechanism to account for the difference in the temporal evolution of temperature between the nanostructured and pristine samples is discussed. It is suggested that there are thermally isolated part on the nanostructures and anomalous temperature increase and melting have occurred locally on them. The amount of vapourized tungsten in response to type-I edge localized mode in ITER is discussed based on experimental observation.

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Design concept of K-DEMO for near-term implementation K. Kim, K. Im, H.C. Kim, S. Oh, J.S. Park, S. Kwon, Y.S. Lee, J.H. Yeom, C. Lee, G-S. Lee, G. Neilson, C. Kessel, T. Brown, P. Titus, D. Mikkelsen and Y. Zhai

2015 Nucl. Fusion 55 053027

Abstract A Korean fusion energy development promotion law (FEDPL) was enacted in 2007. As a following step, a conceptual design study for a steady-state Korean fusion demonstration reactor (K-DEMO) was initiated in 2012. After the thorough 0D system analysis, the parameters of the main machine characterized by the major and minor radii of 6.8 and 2.1 m, respectively, were chosen for further study. The analyses of heating and current drives were performed for the development of the plasma operation scenarios. Preliminary results on lower hybrid and neutral beam current drive are included herein. A high performance Nb

3Sn-based superconducting conductor is adopted, providing a peak magnetic field approaching 16 T with the magnetic field at the plasma centre above 7 T. Pressurized water is the prominent choice for the main coolant of K-DEMO when the balance of plant development details is considered. The blanket system adopts a ceramic pebble type breeder. Considering plasma performance, a double-null divertor is the reference configuration choice of K-DEMO. For a high availability operation, K-DEMO incorporates a design with vertical maintenance. A design concept for K-DEMO is presented together with the preliminary design parameters.

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