SOFT 2018

Europe/Rome
Giardini Naxos

Giardini Naxos

ATAHOTEL NAXOS BEACH RESORT - Via Recanati, 26 Giadini Naxos, Messina - Sicily (Italy)
Aldo Pizzuto (ENEA)
Description

Dear Colleagues

The 30th edition of the Symposium on Fusion Technology (SOFT 2018) will be held in Giardini Naxos (Messina, Sicily), from 16th to 21st September 2018. The event is organized by the Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA), the agency leading fusion research and technology development in Italy. The biennial Symposium on Fusion Technology is the most important conference in this field in Europe. It regularly attracts more than 800 scientists, engineers, industry representatives and exhibitors from all over the world and focuses on the latest developments on fusion experiments and activities. SOFT includes invited, oral and poster presentations as well as industrial and R&D exhibitions. Thanks to its charming location on the Mediterranean Sea and to its extraordinary history, meeting point of different peoples and cultures, Sicily will provide you with a unique mixture of nature, art, traditions as well as excellent cuisine, all offered with renowned warmth and hospitality.

We look forward to seeing you at SOFT 2018 in Giardini Naxos, Sicily.

 

Paola Batistoni

ENEA
Chairman of the Local Organizing Committee

 

Aldo Pizzuto

ENEA
Chairman of the International Organizing Committee

    • 17:15 17:50
      Opening Ceremony Teatro Antico (Taormina)

      Teatro Antico

      Taormina

      Taormina, Messina - Italy

      SOFT 2018 Opening

    • 17:50 18:30
      SOFT Innovation Prize Teatro Greco (Taormina)

      Teatro Greco

      Taormina

    • 18:30 19:10
      I1.1: B. Bigot Teatro Greco (Taormina)

      Teatro Greco

      Taormina

      Invited 1 (B. Bigot)

      • 18:30
        ITER Construction and Manufacturing Progress Toward First Plasma 40m

        ITER reached in November 2017 completion of 50% of the work required to achieve First Plasma. Progress is most visible in the completion of many key buildings, such as the tokamak assembly building, the cryogenic plant, and the magnet power supply building have been completed. The tokamak building will be ready for equipment in 2020 and the bioshield is already to full height. Key systems begin commissioning in 2018, including the steady-state electric network and the component cooling water, while the cryogenic system and magnet power supply commissioning begins in 2019. Thus, the physical plant is moving rapidly toward completion, and key systems are entering the commissioning phase. Manufacturing progress is equally impressive. The base and lower cylinder of the cryostat have been assembled on the ITER site. Three of the six poloidal field coils and the first of the six modules of the central solenoid are being wound or finished. The winding pack and casing for the first toroidal field magnet are complete and verified to meet the high tolerances required (<0.5 mm). The parts for the first vacuum vessel sector have been fabricated and demonstrated to meet strict tolerances (<1 mm). Therefore, the major components of the tokamak have passed into the fabrication phase. The Heating and Current Drive systems (NB, ECH and ICH) are also in the final design phase. The progression of ITER operation from First Plasma (FP) to the achievement of the Q = 10 and Q = 5 project goals has been formalized in a Staged Approach. This is a stepwise installation of components and auxiliary systems; all systems will be installed before the start of the fusion power operational phase. The ITER Research Plan has been revised in 2017 to be consistent with the systems available in each phase.

        Speaker: Dr Bernard Bigot (The ITER Organization)
    • 19:30 20:00
      Concert Teatro Antico (Taormina)

      Teatro Antico

      Taormina

      Taormina, Messina - Italy

      Concert
      Orchestra a Plettro of Taormina

    • 08:30 09:10
      I2.1: J. Schwemmer Plenary Hall - ATA Hotel Naxos Beach Resort

      Plenary Hall - ATA Hotel Naxos Beach Resort

      Giardini Naxos

      Via Recanati, 26 Giardini Naxos, Messina - Sicily (Italy)

      1st Invited First Conference Day

      • 08:30
        Status European procurement for ITER 40m

        Missing

        Speaker: Dr Johannes Schwemmer (Fusion for Energy)
    • 09:10 09:50
      I2.2: T. Donné Plenary Hall - ATA Hotel Naxos Beach Resort

      Plenary Hall - ATA Hotel Naxos Beach Resort

      Giardini Naxos

      Via Recanati, 26 Giadini Naxos, Messina - Sicily (Italy)

      2nd Oral First Conference Day

      • 09:10
        European roadmap to fusion energy 40m

        The European Roadmap to the realisation of fusion energy breaks the quest for fusion energy into eight missions. For each mission, it reviews the current status of research, identifies open issues, proposes a research and development programme and estimates the required resources. It points out the needs to intensify industrial involvement, to educate the fusion scientists and engineers of the future, and to seek all opportunities for collaboration outside Europe.
        A long-term perspective on fusion is mandatory since Europe has a leading position in this field and major expectations have grown in other ITER parties on fusion as a sustainable and secure energy source. China, for example, is launching an aggressive programme aimed at fusion electricity production well before 2050. Europe can keep the pace only if it focuses its effort and pursues a pragmatic approach to fusion energy. With this objective the present roadmap has been elaborated. The roadmap covers three periods: The short term which roughly covers the period until ITER comes into operation and the DEMO Conceptual Design is completed, the medium term which runs until ITER is in routine operation at high performance and the DEMO Engineering Design is completed and the long term.
        ITER is the key facility of the roadmap as it is expected to achieve most of the important milestones on the path to fusion power. Thus, the vast majority of resources proposed in the short term are dedicated to ITER and its accompanying experiments. The medium term is focussed on taking ITER into operation and bringing it to full power, as well as on preparing the construction of a demonstration power plant DEMO, which will for the first time supply fusion electricity to the grid. Building and operating DEMO is the subject of the last roadmap phase: the long term. It might be clear that the Fusion Roadmap is tightly connected to the ITER schedule. A number of key milestones are the first operation of ITER, the start of the DT operation, and reaching the full performance at which the thermal fusion power is 10 times the power put in to the plasma.
        DEMO will provide first electricity to the grid. The Engineering Design Activity will start a few years after the first ITER plasma, while the start of the construction phase will be a few years after ITER reaches full performance. In this way ITER can give viable input to the design and development of DEMO. Because the neutron fluence in DEMO will be much higher than in ITER (atoms in the plasma facing components of DEMO will undergo 50-100 displacements during the full operation life time, compared to only 1 displacement in ITER), it is important to develop and validate materials that can handle these very high neutron loads. For the testing of the materials a dedicated 14 MeV neutron source is needed. This DEMO Oriented Neutron Source (DONES) is therefore an important facility to support the fusion roadmap.
        The presentation will focus on the strategy behind the fusion roadmap and will describe the major challenges that need to be tackled on the road towards fusion electricity. Encouraging recent results will be given to demonstrate the outcome of the focused approach in European fusion research.
        The author is indebted to the whole European fusion community that is together working to make fusion a reality. This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission.

        Speaker: Dr Tony Donné (EUROfusion)
    • 09:50 10:30
      I2.3: G. Mazzitelli Plenary Hall - ATA Hotel Naxos Beach Resort

      Plenary Hall - ATA Hotel Naxos Beach Resort

      Giardini Naxos

      Via Recanati, 26 Giadini Naxos, Messina - Sicily (Italy)

      3rd Invited First Conference Day

      • 09:50
        Role of the Italian DTT in the Power Exhaust implementation strategy 40m

        In the European road map towards the realisation of fusion energy, one of the challenges is the power exhaust for DEMO. If the ITER baseline strategy can’t be extrapolated to DEMO, tens of years will delay the realization of a fusion plant. So in parallel to ITER exploitation, it is mandatory to test alternative solutions for the heat loads on the divertor as risk mitigation for DEMO.
        In the last years two schemes have been proposed as possible solutions: alternative magnetic configurations and the use of liquid metal divertors. Up to now these solutions have been tested at proof of principle level in devices with a plasma current not exceeding 1 MA and SOL parameters significantly different from reactor conditions. To implement one of these concepts on DEMO it is necessary to make another intermediate step, otherwise the extrapolation to DEMO is too large by upgrading existing facilities or by building a dedicated Divertor Tokamak Test (DTT) facility.
        In this framework, the Italian fusion community with the involvement of European Labs has proposed a new device, the Italian DTT, in which one or more alternative magnetic configurations or/and liquid metals can be tested in DEMO relevant conditions.
        To fulfil the DEMO requirements the device have to operate at relevant heat load on the divertor maintaining high core performance, i.e. operations in an integrate scenario with core and edge parameters as close as possible to ITER and DEMO.
        The Italian DTT project has fully supported by the Italian Government that has also identified and is implementing a funding scheme.
        DTT will be a high field superconducting toroidal device (6 T) carrying plasma current up to 6 MA in pulses with length up to 100s, with an up-down symmetrical D-shape defined by major radius R=2.15 m, minor radius a=0.7 m, and an elongation around 1.7.
        The status of the project will be illustrated, highlighting the reviewed design addressed to increase the flexibility by allowing for fully double null operation.
        The site, the time schedule, and the cost estimation will be presented too.

        Speaker: Dr Mazzitelli Giuseppe (ENEA)
    • 10:30 11:00
      Coffee Break 30m Coffee Break Area - ATA Hotel Naxos Beach Resort

      Coffee Break Area - ATA Hotel Naxos Beach Resort

      Giardini Naxos

    • 11:00 13:00
      P1: Poster Session Pantelleria Hall - Terrace - ATA Hotel Naxos Beach Resort

      Pantelleria Hall - Terrace - ATA Hotel Naxos Beach Resort

      Giardini Naxos

      Via Recanati, 26 Giardini Naxos, Messina - Sicily (Italy)

      Poster Session First Conference Day

      • 11:00
        10 MW FULGOR power supply performance tests and overview of test facility components 2h

        The 10 MW FULGOR test facility (Fusion Long Pulse Gyrotron Laboratory) is built to meet the future needs of the gyrotron development, in particular for future fusion machines like the DEMOnstration power plant (DEMO). In the final stage it will enable KIT to test gyrotrons up to 4 MW RF output power in CW and frequencies up to 240 GHz. One of the main components of FULGOR test facility is the High Voltage DC Power Supply (HVDCPS), which has been developed and provided by Ampegon AG. The 90 kV/120 A CW PS is a so-called Enhanced Pulse Step Modulator which allows tests on gyrotrons with multi-staged depressed collectors. Additionally a Pulsed Power Supply (PPS) extends the capabilities of the HVDCPS to 130 kV/120 A for short pulses up to 5 ms every 2 s.

        First tests with the HVDCPS 90 kV/120 A have been performed successfully on a 750 Ω test load. The pulse length in this phase was limited to 666 ms and maximum duty cycle of 0.22 % due to the limitation in the test load. The paper describes the tests that have been performed to determine the capabilities of the HVDCPS, like short circuit test to determine the power dissipated into an arc (<10 J), maximum rise time (10 % - 90 %) <50 µs, output voltage ripple at 90 kV (<0.33 %), settling time (<100 µs) and modulation up to 5 kHz.

        Additionally an overview of the key components of FULGOR test facility, such as the PPS, Body Power Supply (BPS), 10 MW water cooling system, control and data acquisition system and the superconducting magnet with a magnetic field up to 10.5 T are briefly summarized.

        Speaker: Dr Andy Zein (Institute for Pulsed Power and Microwave Technology, Karlsruhe Institute of Technology)
      • 11:00
        14 MeV neutron streaming calculations for JET-like maze entrance using TRIPOLI-4 Monte Carlo code 2h

        The International Thermonuclear Experimental Reactor (ITER) is currently under construction at Cadarache in southern France. The Joint European Torus (JET) is presently the largest tokamak in the world and the only one capable of using tritium. At JET, D-T fusion experiments will be conducted in 2019 (DTE2) on addressing the future ITER needs and reducing the risks of ITER operations. During the DTE2 operations, 14 MeV fusion neutrons will be generated. One of the aims of the DTE2 neutronics experiments is to investigate the D-T neutron streaming along the penetrations of biological shield of the JET Torus Hall. Continuous-energy Monte Carlo (MC) neutron transport calculations and TLDs measurements were already performed by different teams for the previous D-D neutron streaming benchmark experiments around the JET maze entrance. Using two-step simulation approach, two different calculation ways were performed to decrease the variances of calculated neutron fluence and ambient dose equivalent. Both the MC-MC approach using a midway surface source and the Deterministic-MC one using deterministic improved weight windows produced overestimated C/E results. The TRIPOLI-4 Monte-Carlo transport code has been extensively used on fission neutron radiation shielding analyses. To develop the TRIPOLI-4 application in ITER fusion neutronics, both experimental and computational benchmarks are being performed. In this study, using single-step approach combined with variance reduction (VR) techniques, TRIPOLI-4 shielding calculations were performed for a JET-like maze entrance in order to investigate the D-T neutron streaming from a Torus Hall of 40 m x 40 m x 4.2 m model. The calculation results including dose rate maps, the VR performance, and the user-friendly VR input of TRIPOLI-4 code will be reported in the final paper.

        Speaker: Dr Yi-Kang Lee (CEA-Saclay)
      • 11:00
        3D Tritium Transport Model at Breeder Unit Level for WCLL Breeding Blanket 2h

        The Water-Cooled Lithium Lead (WCLL) Breeding Blankets is one of the European blanket designs proposed for DEMO reactor. A tritium transport model inside the blankets is necessary to assess their preliminary design and safety features. Tritium transport and permeation are complex phenomena to be taken into account in the evaluation of tritium balance in order to guarantee tritium self-sufficiency and to characterise tritium concentrations, inventories and losses. In this context, the study has been performed at breeder unit level in the outboard equatorial breeding blanket module, which is, during the normal operating conditions, one of the most loaded modules and this results in higher permeation phenomena. For these purposes, a 3D transport model has been investigated; the model includes buoyancy effects and a preliminary evaluation of the magneto-hydro-dynamics effect (MHD) is also performed. Moreover, it includes advection-diffusion of tritium into the lead-lithium eutectic alloy, transfer of tritium from the liquid interface towards the steel (adsorption/desorption), diffusion of tritium inside the steel, transfer of tritium from the steel towards the coolant (recombination/desorption), advection-diffusion of diatomic tritium into the coolant and buoyancy effect. The temperature profile, tritium generation rate profile, and Pb-15.7Li flow velocity profile have been also taken into account. Results deriving from the transport equation solution, with the above specified phenomena, input and boundary conditions are illustrated in detail within the paper.

        Speaker: Luigi Candido (Energy Politecnico di Torino)
      • 11:00
        A 15 T large aperture dipole for testing fusion and accelerator superconducting samples 2h

        High field superconducting magnets are an essential technology enabling the development of magnetic confinement fusion and high energy hadron colliders. These two communities have joined efforts to design a facility for testing superconducting cables and small insert coils. We propose a large aperture Nb3Sn dipole to replace the magnet assembly of EDIPO, which was irreversibly damaged in 2016, while the rest of the EDIPO infrastructure, including cryostat, cryoplant, power supplies, and high current transformer, remained intact. The goal of this new magnet is to generate a background field of 15 T over a ±0.5% homogeneous field length of 1000 mm. The aperture will enable the test of samples from the two test facilities existing worldwide for the test of high current superconducting cables in a background field above 10 T: SULTAN (94×144 mm aperture) and FRESCA2 (100-mm-diameter bore). The main features of the magnet design will be discussed along with the status and outlook of this collaborative effort between Fusion and High Energy Physics laboratories.

        Speaker: Dr Xabier Sarasola (Swiss Plasma Center, École Polytechnique Fédérale de Lausanne)
      • 11:00
        A European design proposal on the ITER ELM-Coil Power Supply optimized for ELM mitigation and RWM stabilization 2h

        The paper describes a design proposal for the ITER ELM Coil Power Supply, optimized for the simultaneous ELM mitigation and RWM stabilisation during the ITER non-inductive operation. Slow (ELM) and fast (RWM) rotating magnetic fields are generated by exciting the three sets of nine ELM-Coils at ELM-frequencies up to 5 Hz (N = 4) and RWM-frequencies up to 60 Hz (N = 1). Starting from a basic concept of the power supply scheme proposed by IO, the EU-DA developed, studied and optimized the power supply and control systems, by means of extensive computer models. By combining the nine DC-AC IGBT Inverter Power Units on the same DC-bus, the AC-DC converter is optimized but substantial current harmonics at frwm ± felm are generated in the DC-system. In turn, these excite low frequency LC-resonance arising from the distributed DC-capacitor banks and stray inductance of DC-power connections. Further cancellation of the current harmonics at the DC-system level is primordial (impact on the grid) and requires an accurate phase control of the nine ELM-Coil currents. The optimized power and control scheme, featuring commercial compact water-cooled IGBT Inverter Assemblies with own DC-bus, achieves very good performance in normal operation (amplitude error < 1%, phase error < 1°), is shown to be immune to power system imbalance and is well protected against plasma disruption through the synchronized triggering of bi-polar thyristor crowbars. The equipment layout and component ratings of the power supply scheme have been completed. In view of the limited space at the level-4 of the Tokamak building, a compact layout was developed, in which the AC-DC Converter and nine DC-AC IGBT Inverter Power Units are assembled on a two-level metallic structure, designed to be lifted complete by the ITER station crane.

        Speaker: Dr Michel Huart (Fusion for Energy)
      • 11:00
        A horizontal powder injector for W7-X 2h

        Injection of low-Z powders into fusion plasma has been used to improve wall conditions, similar to the standard boronization process using diborane. Powder injection has the advantage of being much simpler, non-toxic, and efficient. The W7-X stellarator is planning on utilizing powder injection in long pulse discharges; a proof-of-principle test for horizontal injection into the plasma was conceived. A design concept is developed using a polyetheretherketone (PEEK) paddle wheel that is driven by a piezo motor, due to the high magnetic fields, that rotates at 100 deg/s. This small unit (“flinger”) fits into an envelope of 120 mm diameter x 150 mm long, a standard size for Multi-Purpose Manipulator. The device is housed in a carbon cup mounted on a retractable probe that can be placed near the plasma edge, enabling powder injection ~4-8 cm radially into the boundary plasma. The feed for the paddle is via piezo electric actuator that vibrates a funnel filled with powder into a trough for the paddle to push. The 8 paddle arms, 35 mm long and 10 mm wide, are made from 0.38 mm thick PEEK which drag slightly along the powder-filled trough bottom, becoming a spring-loaded paddle which accelerates the powder upon release. Design challenges are the high ambient magnetic field, vacuum compatible materials, high temperature environment, limited rotary-drive options, and compact space. The design and testing of this new device will be presented.

        Speaker: Dr Alexander Nagy (PPPL Collaboration Office PPPL at General Atomics)
      • 11:00
        A novel code for the simulation of plasma equilibrium and evolution 2h

        The simulation of the plasma equilibrium and its evolution is important for the study of plasma physics and for the correct design of fusion devices. For this purpose, a novel code, based on the solution of the Grad-Shafranov equation, has been fully implemented in ANSYS. It exploits the finite element method using the magnetic potential vector formulation. In this approach plasma pressure and current density profiles are described by means of two main parameters: the internal plasma inductance li and the Poloidal Betap. Several ITER equilibria of limiter and diverted plasma configurations have been reproduced and benchmarked with validated codes such as MAXFEA and DINA. Being fully implemented in ANSYS this code will allow the coupling of the 2D axisymmetric plasma equilibrium equations, describing the plasma behavior, with the full 3D eddy currents equations, describing the surrounding three-dimensional structures.

        Speaker: Dr Pietro Testoni (ITER delivery, Fusion for Energy)
      • 11:00
        A preliminary assessment of MCNP unstructured mesh integration in the ITER neutronic model 2h

        The design of nuclear fusion devices like ITER requires the execution of complex multi-physics simulations, involving different analysis disciplines such as mechanical, thermal-hydraulic and neutronics. Nowadays, thanks to the novel implementation of unstructured mesh capability into MCNP6, nuclear responses can be computed over meshes conformal with the components tackling the problematic load transference between different codes and, at the same time, improving the modelling methodology.
        Initially a sensitivity analysis over type, order and number of elements was performed to evaluate the hybrid geometry performances in terms of memory demands, time of loading and the different MCNP implementation methodologies.
        Subsequently, a limited section of the ITER vacuum vessel was modelled with unstructured meshes by means of the HyperMesh code. This region was chosen as one of the examples where the explicit structured mesh tally scoring can produce unphysical results near the cells boundaries where the mesh tally voxels cross two (or more) cells with different material properties and relative cross sections. Results are compared with the standard Constructive Solid Geometry (CSG) ITER model and discussed. Samplings of the unstructured mesh tallies were performed and locally assessed against the structured tallies. The particle fluxes and nuclear heating maps obtained show an overall agreement on the average resulting values, however significant local deviations were observed. In particular, unstructured mesh local results showed higher degree of correlation to the underlying materials and avoided the presence of unphysical peaks. However, the added complexity and the time required for meshing large regions, which include many complicated parts, makes an analysis of the benefits necessary, and it limits the application of hybrid geometries to particular cases.

        Speaker: Marco Fabbri (Fusion For Energy)
      • 11:00
        A two colors interferometer for PROTO-SPHERA experiment 2h

        PROTO-SPHERA (Spherical Plasma for Helicity Relaxation Assessment) is a new concept of torus that aims to produce a Spherical Torus at closed flux surfaces and a force-free screw pinch (SP) at open flux surfaces and fed by electrodes [1]. By replacing the metal centrepost current of the spherical tokamaks with the SP plasma electrode current, the rod at the centre of the plasma, which represents the most critical component of spherical tokamak design configuration, can be eliminated. In this way the aspect ratio can be decreased during the experiment, meanwhile the ratio between the toroidal plasma current and the plasma electrode current, is increased. In order to verify the stability of the configuration and the technical components of the PROTO-SPHERA initial arc, a prototype facility has been constructed; this last experiment has reached its plasma current target in the recent months.
        A diagnostic proposed to equip PROTO-SPHERA for the density measurement, that is of the order of ∼10e20 m−3, is a two color interferometer [2]. It will allow to obtain integrate density line in the equatorial plane, necessary to characterize this uncommon plasma. The design and the interferometer laboratory test are described in this paper.
        [1] F. Alladio et al., Nucl. Fusion 46 (2006) S613–S624
        [2] J. H. Irby et al., Rev. Sci. Instrum. 59 (1988) 1568

        Speaker: Cristina Mazzotta (ENEA)
      • 11:00
        Activation foil measurements at JET in preparation for D-T plasma operation 2h

        In the frame of the JET3 Deuterium-Tritium (D-T) technology project within the EUROfusion Consortium program, several neutronics experiments are in preparation for the future high performance D-T campaign at the Joint European Torus (JET). The experiments will be conducted with the purpose to validate the neutronics codes and tools used in ITER, thus reducing the related uncertainties and the associated risks in the machine operations. One of the requirements for the implementation of successful benchmarks is the accurate measurement of neutron fluence during the irradiation phase. The development of measurement techniques for the accurate monitoring of neutron fluence in the device as well as in the surrounding areas is a crucial aspect with respect to neutronics code validation for shielding design, radiation protection and safety assessment. Among others, the foil activation technique is used at JET to evaluate neutron streaming in regions far from the plasma source and along large shielding penetrations as well as to provide an estimate of the neutron spectra to complement shutdown dose rate measurements.
        In the present study, the neutron fluence measurements performed using activation foils during the JET Deuterium-Deuterium (D-D) campaigns are discussed. The activation foil results are compared against other experimental techniques and Monte Carlo simulations and a satisfactory agreement is observed. Moreover, the pre-analysis of the experiments to be performed in the forthcoming JET Tritium-Tritium (T-T) and Deuterium-Tritium (D-T) campaigns is presented. The results of the study provide important data for the implementation of experimental activities at JET and support the preparation in view of the planned D-T operation, allowing the exploitation of the unique 14 MeV neutron yields anticipated. Furthermore, they contribute to the validation of the tools employed in nuclear analyses, which are fundamental for the design and safety of ITER and future fusion power plants.

        Speaker: Dr THEODORA VASILOPOULOU (Institute of Nuclear and Radiological Sciences Technology Energy and Safety NCSR 'DEMOKRITOS')
      • 11:00
        Alloy element induced vacancy clustering in W-Re/Ta material 2h

        Alloying elements can possibly serve as an important technique in designing W-based plasma facing materials (PFMs) with superior comprehensive performance. To investigate the interaction between the alloying elements and point defect is one of the mainly contents to study the W material service properties under irradiation. The first-principle method based on the density function theory was used to investigate the laws for the movements of the point defects in case of the alloying metal Re/Ta solution atoms. It was found from the formation energy and binding energy that Rhenium would be a potent factor to form the mono-vacancy than Ta in W alloying system. The mono-vacancy diffusion in W-Re/Ta alloy was studied via the pathways models, in which Re atom can steady exist in tungsten lattice while Ta solute atom was beneficial to move to a vacancy. Although the replacement atoms were different, it could find that both solute Re and Ta atoms manifest the positive effects to mono-vacancy migration in irradiation environment. According the average relaxed volume, the attraction interaction of vacancy cluster can be interpreted rationally. The important thing was that Re contributed the vacancy to form clusters configuration, but Ta had an inhibiting effect on diminutive vacancy cluster (n≤3), and it can be furthered concluded that Ta could control the point defects from gathering together, especially when its concentration was enlarged. These results can be used to construct the design basis of the W-based alloying in terms of improving the radiation resistance in the fusion environment.

        Speaker: Min Pan (Superconductivity and New Energy RD Center Southwest Jiaotong University Chengdu 610031 China)
      • 11:00
        An Optimization Study for Shielding Design of D-D and D-T Neutron Generators 2h

        The neutron generator (NG) is being used more increasingly in various industrial and research area such as neutron activation analysis, neutron radiography, neutron capture therapy, and so on. In such an application of neutron generator, compactness is one of the most important issue. Since neutron source is generated by deuterium-deuterium (D-D) or deuterium-tritium (D-T) fusion reaction, relatively thick shield for both of fast neutron and related photon is usually required.
        In this study, optimization of shielding design for portable D-D or D-T neutron generators was investigated by adopting appropriate moderator and shielding materials. The effects of moderator or shield materials to neutron spectra and dose rate were also studied to achieve minimization of shield.
        The polyethylene showed good performance in shielding D-D neutrons source while the tungsten showed considerable performance in shielding D-T neutrons due to relatively high threshold energy of (n,2n) reactions. After neutron shield, appropriate photon shield, such as the lead, was also required because the photon shielding performance of the polyethylene or the tungsten is not sufficient.
        Final required shield thickness was also evaluated for various strength of D-D or D-T neutron generators and maximum neutron source strength for portable neutron generator was also discussed.

        Speaker: Sunghwan Yun (Korea Atomic Energy Research Institute)
      • 11:00
        Analyses of the influence of the recycling coefficient on He confinement in DEMO reactor 2h

        Helium, as the ash of burning D-T plasma, is an unavoidable impurity component in DEMO reactor . Its efficient removal from the burning zone of a D-T fusion reactor is most important in the path towards achievement of economic fusion power production. Edge plasma transport properties: recycling/pumping will play a key role in the problem of helium removal from reactor.
        This work describes integrated numerical modelling applied to DEMO discharges with tungsten wall and divertor, using the COREDIV code, which self-consistently solves 1D radial transport equations of plasma and impurities in the core region and 2D multi-fluid transport in the SOL. The model is self-consistent with respect to both the effects of impurities on the α-power level and the interaction between seeded (Ar) and intrinsic impurities (tungsten, helium). The coupling between the core and the SOL is made by imposing continuity of energy and particle fluxes as well as of particle densities and temperatures at the separatrix. In order to keep the prescribed plasma density at the separatrix, the deuterium recycling coefficient was iterated accordingly. It should be underlined that the recycling coefficient in our approach includes effects related to the pumping efficiency (albedo) as well as the intensity of the puffing.
        The aim of the this work is to analyze the influence of the helium and hydrogen recycling on the He confinement in DEMO reactor. Simulation have been done for two argon seeding puff levels : moderate and strong.
        It is found that recycling coefficient of helium have strong influence on the He confinement, which increases from 8.5s to 16s (going from lowest to highest recycling coefficient), but it has small influence on the effective charge state and radiation in SOL. Alpha power decreases only by about 10%, which is the effect of main plasma dilution.

        Speaker: Irena Ivanova Stanik (Institute of Plasma Physics and Laser Microfusion)
      • 11:00
        Analysis of an actively-cooled coaxial cavity in a 170 GHz, 2 MW gyrotron using the multi-physics tool MUCCA 2h

        Continuous Wave (CW) gyrotrons are the key elements for electron cyclotron resonance heating and current drive in present machines and future fusion reactors. In the frame of the EUROfusion activities, a 170 GHz, 2 MW short-pulse (ms) coaxial gyrotron existing at Karlsruhe Institute of Technology (KIT) is being upgraded for operation at longer pulses (100 ms - 1 s). In the coaxial gyrotron, the resonant cavity is made by a hollow cylindrical resonator and a coaxial inner insert, which permits increasing the gyrotron output power by enhancing the mode selectivity of the cavity. A forced flow of pressurized subcooled water, passing around the resonator as well as through the insert, is used to keep the cavity cooled.
        The MUlti-physiCs tool for the integrated simulation of the CAvity (MUCCA) has been recently developed in collaboration between Politecnico di Torino and KIT. MUCCA is applied here to the analysis of the cavity of the upgraded 170 GHz, 2 MW coaxial gyrotron, to assess the evolution of its working point, starting from cold conditions, when different cooling configurations are considered for the resonator. An iterative procedure is adopted, where thermal-hydraulic and thermo-mechanical simulations are first performed on both resonator and insert with the commercial software STAR-CCM+, evaluating the mechanical deformation. The deformed cavity shape is then used as input in the electro-dynamic module EURIDICE to evaluate the heat load on the cavity wall, which becomes in turn the driver of the thermal-hydraulic analysis in the following time-step.
        The main results of the MUCCA simulations are presented in terms of evolution of temperature, heat load, and deformation of the heated surface of the resonator and insert in the first seconds of operation. It is shown that the cavity behavior evolves towards a stable operation, with a maximum temperature strongly dependent on the cooling configuration.

        Speaker: Dr Andrea Bertinetti (Dipartimento Energia, Politecnico di Torino)
      • 11:00
        Analysis of inner divertor materials of JET C-wall and ILW from viewpoint of spectrometric investigations 2h

        Installation of metallic plasma facing wall (ITER-like wall – ILW) [1] and replacing the previous carbon wall (JET-C) in the Joint European torus (JET) was a unique possibility to collect from the tokamak vacuum vessel the first wall erosion products (EP) – dust and flakes.
        Fundamental investment about the properties of EP to comply with security reasons is given by analysing EP from other tokamak devices. Carbon based materials are considered as plasma facing materials in stellarators. A comparison between tritium release and chemical composition of plasma exposed materials will allow to expand the knowledge about materials behaviour in fusion devices.
        Temperature programmed tritium thermodesorption results show to differences between ILW and C-wall plasma facing surface samples. Tritium release from a sample, cut from ILW inner divertor vertical tile, is in range 470-870 K. From analogous position from JET-C wall tritium releases 450-1180 K, while from EP: 370-1140 [2].
        Selected EP were investigated with means of energy dispersion X-ray (EDX), infrared, electron spin resonance and Raman spectrometry.
        EDX analysis of EP shows presence of metallic impurities (Fe, Ni, W etc.) and carbon as main component. With electron spin resonance spectrometry two types of paramagnetic centres - g=2.002 and g=2.12, are characterized. Raman spectra allowed to estimate that in EP are graphite nano-crystals with size ~15 nm. Infrared spectra show presence of inorganic oxides. The obtained results supplement the information about composition of the EP from fusion devices.

        1.M.Rubel et al./Nuclear Fusion 57 (2017) 066027
        2.L.Avotina et al./Advanced Materials and Technologies, Palanga, Lithuania, 2015, 144, P125

        This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission.

        Speaker: Dr Liga Avotina (Institute of Chemical Physics, University of Latvia)
      • 11:00
        Analysis of technical and economic parameters of fusion power plants in future power systems 2h

        To restrict climate change, it is highly desirable to replace conventional power plants by emission free technologies like solar and wind power. This leads to increased shares of fluctuating sources in the power system challenging the balance between generation and demand. Moreover, as the transportation and heating sectors are expected to shift towards using more electricity the projected demand rises strongly. To ensure both sufficient supply and long-term stability of power systems, fusion power may play an important role in the future due to its advantageous properties. Fusion energy is emission free, controllable and inherently safe.

        In recent years a lot of research has been conducted focusing on the technical principles of commercial fusion power. However, less investigation has been done regarding the interaction of different future power plants with the power system.

        This paper provides a detailed analysis of technical and economic parameters of fusion power plants which are relevant in context of power system such as efficiency, heat losses, investment costs and startup costs. Therefore, two linear optimization models – urbs and evrys – are applied which both represent the European power system by including power plants, storages, transmission lines and loads of 34 countries. Both models optimize the total system costs. urbs is applied for expansion planning whereas evrys present a detailed dispatch in hourly resolution. As a result, clear statements about system costs, system configuration and system operation are made based on a comprehensive scenario framework.

        First results indicate, that investment costs and fixed costs of fusion power plants as well as emission restriction highly affect investment decisions. The more flexible a fusion plant can operate, the more fusion power will be installed. The operation of fusion power plants is more dependent on efficiency rates and startup costs than on thermal properties like cooling and heating.

        Speaker: Inga Maria Müller (Technical University of Munich)
      • 11:00
        ANITA-NC: a Code System for Modelling Material Activation Induced by Neutral or Charged Particles 2h

        The evaluation of the neutron induced material activation plays an important role for the development of future fusion power plants for issues related to safety, engineering design and radioactive waste management. For these devices the activation codes and cross section libraries handling neutron energies up to 20 MeV are quite adequate.
        Besides, in order to study the irradiation effects on fusion materials, some facilities have been proposed to produce accelerator-based neutron sources of sufficient intensity to test samples of candidate materials to be used in future fusion power plants. The International Fusion Materials Irradiation Facility (IFMIF) and, more recently, DONES (DEMO Oriented Neutron Source), more tightly focused on DEMO needs, have been proposed to be such dedicated facilities. In both these plants the neutron source is produced through the reaction of 40 MeV deuterons impinging on a liquid lithium target so that neutrons with energies up to 55 MeV are produced.
        In these facilities the deuterons will themselves cause activation, particularly in the accelerator structure and in the lithium target.
        ANITA-NC (Analysis of Neutron Induced Transmutation and Activation – Neutral and Charged) is an inventory code developed in ENEA-Bologna, capable of modelling material activation induced by several neutral and charged projectiles (neutron, proton, alpha, deuteron and gamma). It is an extension of the previous ANITA versions.
        ANITA-NC uses the EAF-2010 group-wise cross-section libraries in EAF format. The VITAMIN-J+ (211) energy group structure for neutrons, up to 55 MeV, and the CCFE (162) group structure for p, α, d and γ, up to 1 GeV, are used.
        The decay data used in ANITA-NC are based on the JEFF-3.1.1 Radioactive Decay Data Library.
        The present paper summarizes the main characteristics of the ANITA-NC activation system. Some relevant application cases and examples of data and code validation on experimental results are also shown.

        Speaker: Manuela Frisoni (FSN ENEA)
      • 11:00
        Assembly and final dimensional inspection at factory of the JT60-SA Cryostat Vessel Body Cylindrical Section 2h

        The JT-60SA project, a superconducting tokamak developed under the Satellite Tokamak Programme of the Broader Approach Agreement between EU and Japan and of the Japan Fusion National Programme, is progressing on schedule towards the first plasma in 2020. Within the European contribution to JT-60SA, Spain is responsible for providing JT-60SA cryostat.

        The JT-60SA cryostat is a stainless steel vacuum vessel (14m diameter, 16m height) which encloses the tokamak providing the vacuum environment (10-3 Pa). It must withstand the external atmospheric pressure during normal operation, and internal overpressure in case of an accident (0.12 MPa absolute). The cryostat design is subdivided, for functional purposes, in two large assemblies: the Cryostat Vessel Body Cylindrical Section (CVBCS) and the Cryostat Base (CB). For transport and assembly reasons the cryostat is made up of 20 main parts: 7 making up the CB and 13 making up the CVBCS (including the top lid). All of the joints between them rely on bolted flanges together with light seal welds, non-structural fillet welds performed from inside and/or outside of the cryostat. The single wall is externally reinforced with ribs to support the weight of all the ports and port plugs and also to withstand the vacuum pressure. The material SS 304 (Co˂0.05 wt%) with a permeability (µrel) below 1.1. The CVBCS is made of a single wall stainless steel shell with a thickness of 34mm.The CB was manufactured and assembled in-situ in 2013, while the CVBC was manufactured by a Spanish company (ASTURFEITO S.A) and delivered to Japan in November 2017.

        This paper summarizes the assembly and final dimensional inspection at factory of the JT-60SA CVBCS.

        Speaker: Dr José Botija Pérez (Ciemat)
      • 11:00
        Assessment of controllers and scenario control performance for ITER first plasma 2h

        Assessment of Controllers and Scenario Control Performance for ITER First Plasma

        M.L. Walker, D.A. Humphreys
        General Atomics, PO Box 85608, San Diego, California 92186-5608, USA

        G. Ambrosino
        CREATE/Università di Napoli Federico II, Napoli, Italy

        P.C. De Vries, J.A. Snipes
        ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067, St. Paul-lez-Durance, France

        F. Rimini
        CCFE/Fusion Association, Culham Science Centre, Abingdon, OX14 3DB, United Kingdom

        W. Treutterer
        Max Planck Institute for Plasma Physics, Boltzmannstr. 2, 85748 Garching, Germany

        The ITER PCS is a key component of ITER operation, with performance requirements much more stringent than existing devices. We report on assessment of control algorithms and control scenarios comprising the prototype ITER PCS design, which is the starting point for development of the final design for first plasma operation. The scenarios assessed include commissioning of magnetics and gas systems using the PCS and the first plasma scenario, which includes neutral gas prefill, plasma breakdown/burnthrough, and initial evolution of equilibrium and plasma density. Systems involved in first plasma control include ECH, PF and CS coils and power supplies, gas valves, and magnetic, neutral pressure, and electron density diagnostics.

        Assessment involves simulation of an ITER PCS model connected in feedback with an ITER plant model, both executing in the Plasma Control System Simulation Platform (PCSSP). PCSSP is presently undergoing upgrades as part of PCS development to provide support for algorithm development, PCS architecture evaluation, and control performance assessment. In particular, PCSSP provides general methods for extensive testing of performance in the face of multiple adverse events, such as plasma instabilities growth, disruptions, or plant system faults. Definition and application of performance metrics to control simulation results will be discussed.

        This work is supported by ITER Organization under Framework Contract 6000000219.

        Speaker: Dr Michael Walker (General Atomics)
      • 11:00
        Assessment of environmental effects on the ITER FOCS operating in reflective scheme with Faraday mirror 2h

        Plasma current measurements will play an important role in ITER to provide real-time plasma control and machine protection. Fiber Optics Current Sensors (FOCS) with the sensing fiber installed on the external surface of the vacuum vessel is a system intended to perform this task. The FOCS signal is proportional to the current and is more suitable for the steady-state operation as compared to today’s standard electromagnetic sensors. However, the intrinsic and extrinsic linear birefringence significantly degrade the sensor performance. Recent numerical simulations demonstrated that operating FOCS in reflection with an ideal Faraday mirror (FM), i.e. the rotation angle is exactly 90°, can significantly improve the measurement accuracy. Unfortunately, FM detuning above 0.3° gives unacceptable error, while the tolerance of commercial FMs is ±0.5°, at best. The FM calibration offers only a partial solution due to uncontrolled detuning of the FM under external perturbations. This problem is the subject of the presentation.
        An ideal FM can be located far from the vacuum vessel (in the cubicle area) because it fully compensates the linear birefringence in the link between the sensing fiber and the FM. For a non-ideal FM this link gives additional errors and should be avoided. When located near the vacuum vessel the FM will be exposed to thermal and radiation fields which will influence the FM rotation. We have investigated these effects experimentally and found that the thermal detuning sensitivity is ~0.12°/K, i.e. temperature stabilization better than ±2.5K is necessary. Radiation-induced detuning saturates at ~1.2° for gamma-doses of 12.5kGy@520Gy/h. Saturation indicates that pre-irradiation may be a way of radiation hardening.
        In the presentation we will also discuss the physical background for the temperature and radiation sensitivity and show that our experiential results agree well with the theoretical models assuming that the FM uses Bismuth-Iron Garnet as the Faraday rotator.

        Speaker: Dr Andrei Gusarov (SCK-CEN Belgian Nuclear Research Center)
      • 11:00
        Automation of upgraded NBI cooling water system 2h

        NBI Cooling Water System is used during beam operation for the removal of received heat load from vessel sub-components. During the process of shifting the NBI Vacuum Vessel to SST-1, it was needed to shift the cooling water plant. Shifting of Cooling Water Plant leads to the dis-mantling of the actual plant and designing, development and installation of a new plant. The cooling water plant is designed for a capacity of 2895 lpm (max.) with a pressure upto 11 bar (for grids loop).

        Main goal of realizing this system is the full Automation; which is achieved by using Step 7 programming for Programmable Logic Controller (PLC), Supervisory Control and Data Acquisition System (SCADA) GUI for the Control logic and data logging & TIA Portal for the PLC Touch Panel (HMI) along with necessary hardware. The paper describes the hardware tasks and the software tasks accomplished for the automation of the cooling water system.

        Speaker: Dr Karishma Qureshi (Neutral Beam Injector, Institute for Plasma Research)
      • 11:00
        Characteristics and Experiment Measurement of Cascaded Plasma In Linear Plasma Devices 2h

        Finite numerical simulation on plasma generation, confinement and distribution, is lacked in design and research of linear plasma devices. As a high density plasma source, cascaded arc plasma is widely used in plasma and material interaction devices. In this paper, a high-density linear plasma device with cascaded arc source is developed, which plasma parameters and distribution is analyzed by COMSOL Multiphysics. The simulation results show that for argon arc discharge, with magnetic field 2000Gs on axis, argon gas flow 100cm3/s, 80A between cathode and anode, plasma density with distance of 200mm from anode is 1.42×1022m-3, and the election temperature is 1.15eV. To validate the model, results are compared with the experimental findings, in which Langmuir probe is adapted to discover plasma parameters, agreed with the numerical simulation well. Research result can provide references for engineering design.

        Speaker: Dr Bo Li (Manufacture and research center, Institute of plasma physics, Chinese academy of sciences)
      • 11:00
        Characterization of modified Be13Zr beryllide as advanced neutron multiplier 2h

        Hydrogen generation reaction with water vapor of Be at high temperatures and BeO produced by this reaction that is harmful to human bodies are major drawbacks. Advanced neutron multipliers with high stability at high temperatures are desirable for fusion reactors where coolant water is extensively used. Beryllides have strong potential for use in high-temperature environments. In the framework of Broader Approach (BA) activities, beryllide pebbles as the advanced neutron multiplier were successfully fabricated by a combination of a plasma sintering synthesis method and a rotating electrode granulation method (REM).
        Beryllide disturbs the tritium breeding by the metal in Be such as Ti, V. Therefore, Be13Zr was selected, because Be13Zr not only has low neutron absorption property, but also has no peritectic reaction during granulation. Be13Zr pebble has been successfully fabricated directly by REM using the plasma-sintered Be-Zr electrode.
        However, Be13Zr has pest phenomenon. Pest reaction occurs during oxidation at relatively low temperatures, and is disintegration of polycrystalline sample cased by oxidation. The disintegration into powdery products is occurred by the growth of oxide along the grain boundary or crack. As action on this issue, dispersion of insoluble element for reduction of oxidation reaction was tried. Si has been selected as dispersion element, because they have low solubility in Be and low neutron cross-section. From the optimization result, Be13Zr without pest reaction was successful in developing by addition of Si. It is supposed that pest reaction was prevented by reduction of stress of oxidation reaction caused by dispersion of Si in grain boundary of Be13Zr.
        In the present study, the characterization of modified Be13Zr by the addition of Si without pest reaction will be introduced, such as thermal and chemical properties.

        Speaker: MASARU NAKAMICHI (Fusion Energy Research and Development Directorate National Institutes for Quantum and Radiological Science and Technology)
      • 11:00
        Characterization of the SPIDER Cs oven prototype in the CAesium Test Stand for the ITER HNB negative ion sources 2h

        The ITER Heating Neutral Beam (HNB) injector is required to deliver 16.7 MW power into the plasma from a neutralised beam of H-/D- negative ions, produced by an RF source and accelerated up to 1 MeV. To enhance the H-/D- production, the surface of the acceleration system grid facing the source (the plasma grid) will be coated with Cs to reduce its work function. Cs will be routinely evaporated in the source by means of specific ovens embedded in the source. Controlling and monitoring the evaporation rate of Cs inside the source will be fundamental to get the desired performances on the ITER HNB.
        In order to properly design the source of the ITER HNB and to identify the best operation practices for it, the prototype RF negative ion source SPIDER has been developed and built in the Neutral Beam Test Facility at Consorzio RFX. In SPIDER, liquid Cs based ovens will be used to inject Cs vapours inside the source. The CAesium Test Stand (CATS) has been specifically designed and set up for testing, commissioning, and characterizing Cs ovens in vacuum, but also to study the Cs evaporation and deposition onto surfaces. A SPIDER Cs oven prototype has been manufactured and tested in CATS in order to characterize its thermal behavior, by means of thermocouples and thermal camera, and its Cs flux, by means of Surface Ionization Detector and Laser Absorption Spectroscopy.
        The paper will present the CATS set up with a description of the layout and main features. The paper will also show the experimental results on the characterization of the Cs oven prototype for SPIDER.

        Speaker: Dr Andrea Rizzolo (Consorzio RFX)
      • 11:00
        Conceptual design of a Neutral Beam Heating system for DTT 2h

        The main purpose of the Divertor Tokamak Test (DTT) is to study solutions to mitigate the issue of power exhaust in conditions relevant for ITER and DEMO. The key feature of such a study is to equip the machine with a significant amount of auxiliary heating power (45 MW) in order to test different divertor solutions. According to the Italian project, the experiment is foreseen to operate with the following main parameters: BT = 6 T, IP = 5.5 MA, R0 = 2.08 m, a = 0.65 m and a pulse duration of 90-100 s. It shall be able to study different divertor magnetic configurations and reach a reactor relevant power flow to the divertor. The proposed mix of heating power foreseen to achieve the target value of 45 MW delivered to the plasma will be provided by Electron Cyclotron Resonant Heating (ECRH), Ion Cyclotron Resonant Heating (ICRH) and Negative-ion-based Neutral Beam Heating (NNBH).
        In this framework, the conceptual design of a NNBH system for DTT is here presented, with a particular focus on the technical solutions adopted to fulfil the requirements and maximize the performances. The proposed system features two beamlines providing deuterium negative ions (D-) with an energy not smaller than 300 keV and an injected power of 5-8 MW each.
        The design of the main components of the injectors is described in detail, explaining the motivations behind the main design choices. A comprehensive set of simulations was carried out using several physics and engineering codes to drive the development of the design. These simulations mainly regard the efficiency of the main processes, the optics of the beam, the physics reactions along the beamline (stripping, charge-exchange and ionization), the thermo-mechanical behaviour of the acceleration grids and the coupling between the beam and the plasma in the tokamak chamber.

        Speaker: Dr Piero Agostinetti (Consorzio RFX)
      • 11:00
        Conceptual Design of a Toroidal Field Coil using HTS CrossConductor 2h

        The potential of High Temperature Superconductor (HTS) for a Toroidal Field Coil (TFC) of a future fusion power plant has already be demonstrated in a conceptual design within EUROfusion [1]. One of the candidates of a high current HTS conductor for use in a fusion magnet is the so-called HTS CrossConductor (HTS CroCo) where REBCO tapes are arranged in a cross sectional optimized way.
        The basic TFC dimensions have been taken from the PROCESS system code as the starting point for the design of a winding pack, consisting of six CroCos around a copper core, embedded in a stainless steel jacket and cooled by 4.5 K supercritical helium. With this cable geometry, the electromagnetic, structural mechanics, cooling, and thermo-hydraulic performance of an HTS-TFC were investigated. It could be shown that the conductor and winding pack design fulfills the requirements with respect to structure mechanics and hot spot in case of a quench. The current sharing temperature is large enough that it is possible to handle the nuclear heat load on the coil with still sufficient margin.
        [1] R. Heller, P. V. Gade, W. H. Fietz, T. Vogel, K. P. Weiss, "Conceptual Design Improvement of a Toroidal Field Coil for EU DEMO using High Temperature Superconductors", IEEE Trans. Appl. Supercond. 26(4) (2016) 4201105

        Speaker: Dr Reinhard Heller (Institute for Technical Physics, Karlsruhe Institute of Technology)
      • 11:00
        Conceptual studies on optical diagnostic systems for plasma control on DEMO 2h

        The roadmap to the realization of fusion energy describes a path towards the development of a DEMO tokamak reactor, which is supposed to provide electricity into the grid by the mid of the century [1]. The DEMO diagnostic and control (D&C) system must provide measurements with high reliability and accuracy, constrained by space restrictions in the blanket under the adverse effects induced by neutron, gamma radiation and particle fluxes. As a consequence an initial selection of suitable diagnostics has been obtained [2]. This initial group of diagnostic consists in 15 different systems classified in 6 methods, microwave diagnostics, thermo-current measurements, magnetic diagnostics, neutron/gamma diagnostics, IR interferometry/polarimetry, and a variety of spectroscopic and radiation measurement systems.

        The key aspect for the implementation, performance and lifetime assessment of these systems on DEMO is mainly attributable to their suitable location, that must be well protected against neutrons, and meet their own set of specific requirements. Within this paper, we concentrate on spectroscopic and radiation measurement systems that require sightlines over a large range of plasma regions and inner reactor surfaces. In this context, sightline analysis, the space consumption and the evaluation of optical systems are the main assessment tools to obtain a high level of integration, reliability and robustness of all this instrumentation, essential features in future commercial fusion power nuclear plants. This paper summaries the main results and strategies adopted in this early stage of DEMO conceptual design, to assess the feasibility of this initial set of diagnostic methods based on sightlines and the integration of the total number of these needed for DEMO D&C.

        List of references:

        [1] F. Romanelli, “Fusion Electricity – A roadmap to the realization of fusion energy,” 2013
        [2] W. Biel et al., “Diagnostics for plasma control - from ITER to DEMO”, SOFT Conference 2018

        Speaker: Dr Winder Gonzalez (Insttitute of Energy and Climate Research EK-4 01.10, Research Center Jülich)
      • 11:00
        Contribution to safety analyses of DEMO HCPB using AINA code 2h

        The main motivation of the current work, framed under the safety EUROfusion activities to develop DEMO, is to present the conclusions drawn from our contribution to the safety studies of the HCPB DEMO design carried out by the team tasked with AINA code development. During 2016 and 2017 a new AINA version has been built and properly validated in order to evaluate plasma evolution and in-vessel components strains inside the European DEMO designs. As a result, AINA is able to simulate several accident scenarios as plasma disruptions or structural material meltings due to a LOPC (Loss Of Plasma Control) and in-vessel melt either of FW, blanket structure and/or divertor modules because of thermal stresses due to a LOCA (Loss Of Coolant Accident). After due analysis, it has concluded that it would be desirable to carry out a possible design review focused on ensuring a suitable operating temperature range with a bigger safety margin for all the materials which make up the HCPB BB, as well as the need to guarantee a quick detection and actuation by means of a proper system, depending on the affected equipment, when the most demanding transients take place which may drive the reactor to suffer a melting scenario and a very energetic plasma disruption at the same time. These events include an increase of fueling injection above 50%, a permanent improvement in the confinement time and a punctual impurity increase above 300%. Other perturbations has been studied, which some of them only provide information on non-dangerous cases as a decrease of the fueling injection rate, impossible situations from a technical or physical point of view as an unexpected and sudden increase of external power injection above 630% or melting processes as LOCAs (for any severity level) or a decrease of the external power injection.

        Speaker: Dr Eduard Baeza (Polytechnic University of Catalonia (UPC))
      • 11:00
        Convective Baking Test of the ITER Lower Port for Factory Acceptance 2h

        The ITER lower port is designed to support divertor remote handling and vacum pumping. To meet the purpose, it will be assembling to each main vessel on the vacuum vessel manufacturing site. Before delivering to the sector shop, a series of fuctional and mechanical test, which is so-called factory acceptance test (FAT) should be performed by the manufacturer. The ITER FAT should be complying with the RCC-MR 2007 code and French regulations of neclear pressure equipment (ESPN) to assure quality of nuclear pressure equipment. The precedure of lower port FAT has been set to that visual test, pressure test, baking, vacuum leak test, and final dimensional inspection. Especially, the baking is critical cleaning method to satisfy required vacuum condition. The baking condition is challengeable to satify both the given ramp-up/down condition, which is 5 °C/hr, and the temperature difference of the object within 40 °C, simultaneously. The air heating and circulating furnace has been specially designed to apply convective heating and cooling method. This is because the lower port is large complex double wall box structures, convevtive heating and cooling is relatively proper method to satify the given condition with time. In this study, using the mock-up of the lower port, convective baking test is perfomred with maximum holding temperature is set to 220 °C during 12 hours. As a result, the maximmum temperature difference of the mock-up is apperaed 15 °C, which is appeared at the ramp-down condition. In addition, applying the measured result, thransient thermal-structural analysis is performed. The maximum stress of 85.5 MPa is occured, and this is reasonable stress intensity compare to the allowable stress 187.5 MPa. It is a proven that the convective baking method is quite feasible as the FAT baking for the ITER lower port.

        Speaker: Dr Hokyu Moon (Tokamak Engineering Department, ITER Korea, National Fusion Research Institute)
      • 11:00
        Cooling optimization of the electron cyclotron upper launcher blanket shield module 2h

        The four ITER EC (Electron Cyclotron) Upper Launchers inject up to 8 MW microwave power each with the aim to counteract plasma instabilities during plasma operations. The structural system of these launcher antennas will be installed into four upper ports of the ITER vacuum vessel.

        The structural part of the Upper Launchers which forms the plasma facing component is called the Blanket Shield Module (BSM) that during operation will be heated by nuclear heating from neutrons and photons, thermal radiation from the plasma and mm wave stray radiation. The BSM is classified as an ITER VQC1 component since together with its cooling system it is fully immersed into the torus vacuum environment.

        Recently Fusion for Energy has started the manufacturing of a full scale prototype of the BSM with the objective of finding an optimum manufacturing route including both technical and economic aspects.

        This paper describes the design changes that have been implemented for the optimization of the BSM cooling performance by adapting the design and manufacturing route according to the know-how of the prototype supplier (ATMOSTAT) of Hot Isostatic Pressure (HIP) technologies. In particular, a highly efficient flange cooling scheme has been implemented.

        This paper also details the design process that has driven the cooling optimization, including finite element steady state thermal analysis and CFD (Computational Fluid Dynamic) analysis. Specifically CFD was used for balancing the flow of the parallel water channels.

        Finally this paper illustrates experimental tests results of the qualification mockups used to validate the HIP procedures proposed by ATMOSTAT.

        Speaker: Dr Jose Pacheco (Fusion for Energy)
      • 11:00
        Critical Design Issues in DEMO and Solution Strategies 2h

        The EU fusion roadmap defines as a goal the development of a DEMO, which achieves a high plasma operation time and demonstrates Tritium self-sufficiency and net electricity output. A number of design issues have been identified as critical, either because the solution chosen in ITER is not suitable in DEMO or because it is a DEMO-specific issue not present in ITER. All of these will affect in their resolution the design and possibly the technology of several tokamak and plant systems or even the DEMO architecture: (i) Feasibility of wall protection limiters during plasma transients, (ii) integrated design of breeding blanket and ancillary systems, (iii) power exhaust taking advantage of advanced divertor configurations, (iv) tokamak architecture based on vertical blanket segments, (v) direct or indirect power conversion concept, (vi) configuration of plant systems in the tokamak building, (vii) feasibility of hydrogen separation in the torus vacuum pump and direct recirculation, and (viii) plasma scenario.
        For each of these issues potential solutions have been identified and activity plans have been defined for the associated developments and assessments. Four of these particularly affect the integrated design of DEMO, namely (i), (ii), (iv), and (vi). These will be introduced and discussed in this article and for each a summary of the identified risks, the rationale for the chosen solution concepts, and the identified required verifications will be given.

        Speaker: Christian Bachmann (PMU EUROfusion)
      • 11:00
        Decontamination tests of dust under load for the ITER blanket remote handling system 2h

        Radioactive dust will accumulate in the vacuum vessel (VV) of ITER after plasma operations. Thus, the ITER Blanket Remote Handling System (BRHS) will be installed in the VV to handle the blanket modules, which can weigh up to 4.5 ton and be larger than 1.5 m, stably and with a high degree of positioning accuracy. The BRHS itself also needs to undergo regular maintenance in the Hot Cell Facility (HCF). Maintenance workers will be exposed to the radioactive dust that adheres to the surface of the BRHS. Past studies estimated the contaminated surface area of the BRHS, however, in this study, decontamination tests were performed and the dose rate to maintenance workers was calculated using the Monte Carlo N–Particle Transport Code (MCNP5). Decontamination tests were performed by using multiple test pieces of varying surface roughness (Ra 12.5, Ra 6.3, and Ra 1.6) made from SUS329J4L and two different types of brushes to simulate decontamination of the BRHS surface. Tungsten dust was pressed on the test pieces to simulate the loading by the rollers. After the test pieces were brushed the surface of each test piece was observed by using both optical and scanning electron microscopes. Dust was reduced by approximately 99% in all cases where the SUS304 brush was used regardless of surface roughness. Afterwards, the decontamination rate (amount of dust that was cleaned) was used to estimate how much dust will be able to be cleaned from the BRHS surface. This paper describes the optimal decontamination tools with respect to the BRHS surface conditions to calculate and hopefully reduce the dose rate to maintenance workers so as to optimize the BRHS maintenance plan in the HCF.

        Speaker: Makiko Saito (Department of ITER Project National Institutes for Quantum and Radiological Science and Technology Mukoyama)
      • 11:00
        Deformation and fracture behavior of the ODS-Cu/W joint fabricated by the improved brazing technique 2h

        In our previous work, the joint between oxide dispersion strengthened copper alloy (ODS-Cu) and tungsten (W) demonstrated superior fracture strength (~200 MPa). In the present study, deformation and fracture behavior of the bonding layer and its vicinity after the three-point bending test was investigated. Consequently, it was found that the crack initiation site was dominantly in the tungsten bulk side, although it was not clear that the crack initiated from grain boundary or not. In addition, the crack propagation proceeded towards the bonding layer and seemed to end at the interfaces. These results indicate that the strength of the bonding interface is higher than that of the tungsten bulk and the present bonding technique is applicable for severe environments such as a high heat flux component on the fusion reactor.
        The copper alloy has been considered to be useful as a divertor heat sink or cooling tube not only in the helical reactor FFHR-d1 but also in the tokamak DEMO reactor. The special feature of the basic option of the FFHR-d1 divertor is that the ODS-Cu, GlidCop® (Cu-0.3wt%Al2O3) is applied for the heat sink, and the flat plate tungsten armour is supposed to be bonded on the heat sink by using the improved brazing technique with BNi-6 (Ni-11%P) filler material.
        The thickness of the bonding layer between GlidCop® and W was extremely narrow, and the joint strength and toughness were superior. This fact demands attention because it could be generally predicted that the bonding layer itself has a brittle feature. If we would want to obtain much higher fracture strength, tungsten bulk with better mechanical properties should be used. In this paper, the mechanism for producing such a fine and toughened bonding layer, and deformation and fracture behavior of the joint will be discussed.

        Speaker: Dr Masayuki Tokitani (Department of Helical Plasma Research, National Institute for Fusion Science)
      • 11:00
        Delphi Exercise on the possible role of fusion energy in the global energy system 2h

        The Delphi Exercise on the possible role of fusion energy technology in the future global energy system is a research project funded by the EUROfusion Socio-economic Studies (SES) programme. European research on fusion energy has a long tradition of energy scenario modelling as a way to study the economic and social conditions under which fusion would eventually become an energy technology option in the future. Previous research under the SES programme has emphasised for EUROfusion the importance of participatory socio-economic research with the inclusion of stakeholders from informed civil society. Following this, three reflection groups with civil society participants were organised:

        • on the use of the concept of sustainable development in energy governance,
        • on the use of modelling in energy foresight research and
        • on the use of storylines in foresight research in general.

        This Delphi Exercise should be seen as the fourth foresight research activity concerned with researching the possible role of fusion energy technology in the future global energy system. The presentation will discuss the results of the first round in which experts are invited to give their views on four points of attention:

        1. aspects of future energy policy;
        2. the use of reference scenarios such as those of the Intergovernmental Panel on Climate Change or the World Energy Council;
        3. the use of storylines as a mean to construct quantifiable scenario’s;
        4. the use of drivers such as population growth, climate change, direct energy cost or technology availability.

        In conclusion, the presentation will elaborate on how the results of this exercise can inspire future technological and social sciences research on fusion energy.

        Speaker: Gaston Meskens (SCK-CEN)
      • 11:00
        Design and analysis of robot for the maintenance of divertor in DEMO fusion reactor 2h

        DEMO represents DEMOnstration Power station, which is a nuclear fusion power station and it is proposed to be built after ITER experimental nuclear fusion reactor. It is impossible to do any change or repair work during the nuclear operation by human directly due to the high radiation and extreme temperature in the fusion reactor. And the solution to solve these issues is adopt the remote handling robot.
        Divertor is one part of the DEMO fusion reactor. In DEMO, there are 54 cassettes and each cassette is supported by stainless steel structure, the weight of each cassette is around 7 tones. The divertor is devoted to extract the ash and heat generated during fusion reaction, minimize plasma contamination and protect the surrounding walls in the extreme environment where thermal and neurotic loads are existed.
        Concept of remote handling robot is designed, and main components of the robot such as bearings, wheels, rails and cylinders are selected. FEM analysis on critical point especially contact area is carried out. Mechanism of the robot and cassette removal and installation sequences are studied. Challenges here are space limitation and avoiding collision. The rail change system is one of the crucial part in this concept; four V shape wheels integrated with spherical roller bearing can rotate and roll along the toroidal rail. Clearance between blanket and the cassette is eliminated by the hydraulic jack installed on the top of the structure. The driven system of the structure movement along the toroidal rail is two telescopic cylinders, each cylinder controls one direction. Moreover, the merits and demerits are mentioned, in this concept, the structure is simple and have high stiffness, but the modification on the vacuum vessel should be taken into further consideration since the radiation may cause unexpected deformation on the rail.

        Speaker: Changyang Li (Lappeenranta University of Technology)
      • 11:00
        Design and Countermeasures against Cavitation in a Downstream Conduit of the Liquid Lithium Target for International Fusion Materials Irradiation Facility 2h

        A liquid-lithium (Li) free-surface stream flowing under a high vacuum serves as a Li target for the planned International Fusion Materials Irradiation Facility (IFMIF). As the primary Japanese activity for the Li target system of the IFMIF/EVEDA (i.e. Engineering Validation and Engineering Design Activities) project, implemented under the Broader Approach (BA) Agreement, cavitation-like acoustic noise was reported in the downstream conduit of Li target assembly (TA) of the IFMIF/EVEDA lithium test loop (ELTL), which aims to verify the lithium target and purification systems envisioned the IFMIF. To clarify the cause of this acoustic noise, we found that acoustic emissions due to cavitation occurred in a narrow area near the start of the bend pipe where the Li target impinged by using acoustic-emission sensors. And the method to determine this conflict (initial arrival) location was formulated. Intermittent high-frequency acoustic emission can also cause of the cavitation-erosion crack of the structural materials due to cavitation bubbles collapse (cavitation pitting). To examine the detailed location of the cavitation-like acoustic noise occurring in the downstream conduit of TA, the change of initial arrival position in the downstream conduit according to the velocity of Li target was calculated by the flow analysis of ANSYS/Fluent V16.2 and SCRYU/Tetra, and the design and countermeasures against this type of cavitation were discussed.

        Speaker: ChangHo PARK (Department of Fusion Reactor Materials Research National Institutes for Quantum and Radiological Science and Technology)
      • 11:00
        Design and development of the mechanical support structure for ITER in-vessel magnetic sensors 2h

        A mechanical support structure (a.k.a. “platform”) has been designed to provide mechanical support and thermal conductance for the inductive magnetic sensors installed on the inner shell of ITER vacuum vessel (VV) for equilibrium and high-frequency magnetic field measurement. The platform design is modular so as to simplify the on-site installation process. It consists of a permanent and a removable section. The permanent section is welded to the VV surface, forming the base of the whole assembly and allowing electrical connection with the in-vessel electrical service. The removable section houses the sensor head, which is engaged with the permanent base through a knife-switch type connector. Alignment of the two sections is realized via two alignment pins built into the permanent base. The removable section is designed to be compatible with remote handling operation in case of replacement. Real-scale prototypes of the platform components had been successfully manufactured and assembled together according to the design. Good alignment between the two sections has been achieved. Full electrical continuity from the sensor head down to the electrical service cabling has been demonstrated in the platform prototypes. Continued R&D efforts will focus on testing some installation and operation aspects, including mechanical vibration, thermal cycling, cable clamping, etc.

        Speaker: Dr Yunxing Ma (Fircroft Engineering)
      • 11:00
        Design and development of the mechanical support structure for ITER in-vessel magnetic sensors 2h

        A mechanical support structure (a.k.a. “platform”) has been designed to provide mechanical support and thermal conductance for the inductive magnetic sensors installed on the inner shell of ITER vacuum vessel (VV) for equilibrium and high-frequency magnetic field measurement. The platform design is modular so as to simplify the on-site installation process. It consists of a permanent and a removable section. The permanent section is welded to the VV surface, forming the base of the whole assembly and allowing electrical connection with the in-vessel electrical service. The removable section houses the sensor head, which is engaged with the permanent base through a knife-switch type connector. Alignment of the two sections is realized via two alignment pins built into the permanent base. The removable section is designed to be compatible with remote handling operation in case of replacement. Real-scale prototypes of the platform components had been successfully manufactured and assembled together according to the design. Good alignment between the two sections has been achieved. Full electrical continuity from the sensor head down to the electrical service cabling has been demonstrated in the platform prototypes. Continued R&D efforts will focus on testing some installation and operation aspects, including mechanical vibration, thermal cycling, cable clamping, etc.

        Speaker: Dr Yunxing Ma (Fircroft Engineering)
      • 11:00
        Design and measuring performance of the ITER plasma position reflectometer in-port-plug antennas. 2h

        Design parameters of the ITER Plasma Position Reflectometer (PPR) in-port-plug antennas are determined and then their measurement performance is assessed using 2D full wave analysis.
        Two ITER scenarios were selected when considering the optimum antenna position and orientation, namely the baseline scenario (15 MA D-T) and the low density one planned for the initial non-active phase at 7.5 MA. Using them to feed a 3D ray tracing simulation, spatial position and optimum orientation angles of each set of emission and detection antennas were determined. Additionally, a far field analysis of the launching radiation patterns led to the definition of the antenna dimensions in terms of optimal power coupling.
        After this preliminary work, 2D full wave simulations using a finite difference time domain (FDTD) code were performed to assess the measurement performance of the system, in terms of spatial resolution and accuracy. To this end, the detected wave amplitudes and phases were evaluated for each operating scenario and two models of the SOL plasma density. By calculating the spectrogram (STFT technique) of the phase of the detected wave, the reconstruction of the plasma density profile can be carried out and be directly compared with the input profiles. As a result from this synthetic diagnostic analysis, an estimation of the error in the determination of the last closed flux surface position was possible. Both static and turbulent plasmas have been considered, using for the latter a multimodal model to mimic the spectrum of density fluctuations.
        Together with the power gain information previously obtained, the 2D characterization of the measuring performance of the system will be presented.

        Speaker: Dr José Martínez-Fernández (Laboratorio Nacional de Fusión, CIEMAT)
      • 11:00
        Design and preliminar operation of a laser absorption diagnostic for the SPIDER RF source 2h

        The ITER Heating Neutral Beam (HNB) injector is required to deliver 16.7 MW power into the plasma from a neutralised beam of H-/D- negative ions, produced by an ICP RF source and accelerated up to 1 MeV. To enhance the H-/D- production, the surface of the acceleration system grid facing the source (the plasma grid) will be coated with Cs because of its low work function. Cs will be routinely evaporated in the source by means of specific ovens. Monitoring the evaporation rate and the distribution of Cs inside the source is fundamental to get the desired performances on the ITER HNB. In order to proper design the source of the ITER HNB and to identify the best operation practices for it, the prototype RF negative ion source SPIDER has been developed and built in the Neutral Beam Test Facility at Consorzio RFX. In SPIDER, the dynamics of Cs will be monitored by measuring the emission intensity of the Cs 852 nm line along several lines of sight in the source. For a more quantitative estimation of Cs density a Laser Absorption Spectroscopy diagnostic will be installed; by using a wavelength tunable laser, the diagnostic will measure the absorption spectrum of the 852 nm line along 4 lines of sight, parallel to the plasma grid surface and close to it. From the absorption spectra the line-integrated density of Cs at ground state will be calculated. The paper will present the design of the diagnostic for SPIDER, with a description of the layout and of key components. The paper will also show the first experimental results from a preliminary installation of the diagnostic on the test stand for Cs ovens.

        Speaker: Dr Marco Barbisan (INFN-LNL)
      • 11:00
        Design and preliminar operation of a laser absorption diagnostic for the SPIDER RF source. 2h

        The ITER Heating Neutral Beam (HNB) injector is required to deliver 16.7 MW power into the plasma from a neutralised beam of H-/D- negative ions, produced by an ICP RF source and accelerated up to 1 MeV. To enhance the H-/D- production, the surface of the acceleration system grid facing the source (the plasma grid) will be coated with Cs because of its low work function. Cs will be routinely evaporated in the source by means of specific ovens. Monitoring the evaporation rate and the distribution of Cs inside the source is fundamental to get the desired performances on the ITER HNB.
        In order to proper design the source of the ITER HNB and to identify the best operation practices for it, the prototype RF negative ion source SPIDER has been developed and built in the Neutral Beam Test Facility at Consorzio RFX. In SPIDER, the dynamics of Cs will be monitored by measuring the emission intensity of the Cs 852 nm line along several lines of sight in the source. For a more quantitative estimation of Cs density a Laser Absorption Spectroscopy diagnostic will be installed; by using a wavelength tunable laser, the diagnostic will measure the absorption spectrum of the 852 nm line along 4 lines of sight, parallel to the plasma grid surface and close to it. From the absorption spectra the line-integrated density of Cs at ground state will be calculated.
        The paper will present the design of the diagnostic for SPIDER, with a description of the layout and of key components. The paper will also show the first experimental results from a preliminary installation of the diagnostic on the test stand for Cs ovens.

        Speaker: Dr Marco Barbisan (Consorzio RFX)
      • 11:00
        Design and preliminary testing of a sieve tray column for PbLi purification 2h

        Future designs of fusion devices are going to make use of a tritium production systems, several of which are considered using PbLi alloy as a breeder. Apart from tritium, other volatile and non-volatile species are being formed, either as products of the neutron irradiation or as corrosion products. The volatile impurities must be eliminated based on safety concerns (e.g. polonium) or blanket system malfunctions concerns (e.g. helium). For these, a gas-liquid contactor appears to be a suitable unit. A promising arrangement of the gas-liquid contactor is a column using a sieve tray as liquid PbLi distributor allowing certain free falling height for the desired purification to take place. This contribution presents details of the design and initial testing of such unit with the aim to remove highly volatile compounds by desorption into a stream of inert gas at ambient pressure.

        Speaker: Michal Kordac (Research Center Rez)
      • 11:00
        Design and simulation of a cascaded four-quadrant 24- pulse converter based on 6-phase pulsed motor-generator 2h

        This paper designs a cascaded four-quadrant 24-pulse converter fed by a 6-phase pulsed motor-generator (M-G) aiming at the requirements of HL-2M tokamak, which is mainly used for the control of vertical instability of plasma. The optimally designed four-quadrant 24-pulse converter cascaded by four-quadrant converters is able to balance the loads of the double Y of M-G. Owing to the fact that the control of the four-quadrant converters are relatively independent and logical controlled with circulating current, it is available to work under high current in four-quadrant. The output current transits smoothly and continuously over the zero crossing point. In addition, the circular reactors can be saved by taking full advantage of the large inductance of the M-G to restrain circulating current. This paper designs a four-quadrant 24-pulse converter satisfied with the requirements and makes MATLAB simulations to verify the performances of current following, the adjustability of voltage source fluctuating, the smoothness of current passing zero and the presentation of circulating current.

        Speaker: Dr Zheng Xue (Southwestern Institute of Physics)
      • 11:00
        Design concept and thermal-structural analysis of a high power reflective mm-wave optical mirror (M2) for the ITER ECH Upper Launcher 2h

        Each of the 4 ITER Electron Cyclotron Heating Upper Launcher (ECHUL) features 8 transmission lines (TLs) used to inject 170 GHz microwave power into the plasma at a level of up to 1.31 MW (at the TL diamond window) per line. The millimetre waves are guided through a quasi-optical section consisting of three fixed mirror sets (M1, M2 and M3) and one front steering mirror set (M4).
        The M2 mirror set is composed of an upper and lower part, each reflecting 4 nearly-Gaussian beams coming from the M1 to the M3 mirror, which focuses them towards the M4 steering mirror that will aim at the correct location in the plasma for suppression of the q=3/2 and q=2/1 NTMs.

        Mm-wave power is converted into heat by ohmic dissipation, reaching a peak power density of approximately 4 MW/m2 on each of the 4 beam centre spots of the M2 mirror, resulting in a total of 19.4 kW absorbed power.

        EPFL-SPC has developed a novel water cooled mirror design concept which is able to dissipate such high heat loads (with up to 60000 thermal cycles) and also resist the applied external loads and dynamic displacements arising from plasma disruptions and seismic events, while complying with: material, manufacturing and space restrictions.

        This study describes the main design features of the M2 upper mirror, and its design conformity in accordance to the ASME design code and also its conformity to the Essential Safety Requirements (ESR) for Nuclear In-vessel components.

        This work was supported in part by the Swiss National Science Foundation. This work was carried out within the framework of the ECHUL consortium, partially supported by the F4E grant F4E-GRT-615. The views and opinions expressed herein do not necessarily reflect those of the European Commission or the ITER Organization.

        Speaker: Dr Philip Santos Silva (EPFL)
      • 11:00
        Design Criteria of the Electrical Power Supply for Lithium Loop System of DEMO-Oriented NEutron Source (DONES) plant. 2h

        The aim of this paper is to identify the design criteria of the Electrical Power Supply of the Lithium Systems of DEMO-Oriented NEutron Source (DONES) power plant. This facility is planned as a simplified IFMIF-like plant to provide, in a reduced time scale and with a reduced cost, information on materials damage due to neutron irradiation. In particular, a general overview of the current status of the electrical power systems developed for IFMIF-EVEDA has been made and also emphasis has been put on the analysis of the new specific requirements of DONES. This paper describes the Electric Power System (EPS) for Lithium Loop System (LLS). Its scope is to define design criteria, mainly focusing on the design of electric distribution system for electric loads of LLS and its components. A detailed description of design criteria and of key functions are reported, taking into account the specific requirements of the DONES Facility.
        The Lithium Power System, has the following main functions:
         receives the AC power from the commercial power grid and transformed to proper voltage levels and feeds LLS electric loads in normal conditions;
         receives the DC/AC power from the uninterruptible power supply systems and emergency generator in case of the AC power grid lost, for SIC LLS electric loads;
         provides control and protection to the electric equipment, cables and electric loads against faults.
        The Electric Distribution Systems provides the electric power to all electric loads of LLS. These loads are classified based on Safety Important Class (SIC) methodology into three groups based upon the loads served (in according to the IEC Standards):
        • SR/Non-SIC loads, Single Power System (Normal Operation) fed by power grid;
        • SIC-2 loads, Power System plus Emergency Power Supply System;
        • SIC-1a/b loads, Redundancy Power System plus Emergency Power System.

        Speaker: Dr Pietro Zito (ENEA)
      • 11:00
        Design evolution of the diamond window unit for the ITER EC H&CD upper launcher 2h

        The torus window unit is a very particular component of the ITER EC H&CD upper launcher aiming to provide the vacuum and confinement primary boundary between the vacuum vessel and the transmission lines (TLs). The high power 170 GHz millimeter-wave beams generated by the gyrotrons travel along the TLs and pass through the window units, before being quasi-optically guided into the plasma via the upper launchers. The design of the window unit shall thus meet stringent requirements to guarantee the safety function, the millimeter-wave beam transmission and the structural integrity during normal operation and off-normal events. The unit consists of an ultra-low loss CVD diamond disk brazed to two copper cuffs; this structure is then integrated into a metallic housing by welding. The compliance with the requirements shall be assured by applying the ASME Section III – Subsection NC code and a dedicated experimental qualification program.
        This paper reports the way in which the design of the unit, already optimized by FEM analyses against the ITER loading conditions, was further improved by the application of the ASME III-NC code, leading to a more feasible and simpler manufacturing and assembling sequence. In addition, the impact of the ITER project decision to change the inner diameter of the waveguide from 63.5 to 50 mm, to improve the beams’ mode purity, was assessed and it is also discussed. Different materials for the metallic housing and in particular for the millimeter-wave inserts of the unit were compared using appropriate engineering criteria to mitigate the significant increase of the millimeter-wave thermal loads on the waveguides when the diameter is decreased.

        Speaker: Dr Gaetano Aiello (IAM-AWP, KIT)
      • 11:00
        Design of High Current Busbar Contact Connection for ITER Poloidal Field Converter 2h

        The DC equipment of ITER poloidal field converter will be interconnected by DC busbar with water cooled aluminum busbars with cross section 200×60 mm, whose segments will be connected by aluminum flexible links in order to compensate that of thermal expansion. Because the contact surface between DC busbar and flexible links are small and the high current up to 30 kA flowed through, the power dissipated in the contact surface are about 900W, if the contact resistance is assumed only to be 1 uΩ, it is necessary to decrease the contact resistance in order to reduce the temperature rise of contact joint. This paper presents the different design on high current bolted busbar connection to increase the contact area and reduce contact resistance. Firstly, several methods to be discussed to reduce contact resistance based on lots of references. Secondly, these methods are analyzed by using ANSYS software, the pressure and stress between the contact surface is provided, it is obvious that the contact resistance is smaller with the increase of pressure and stress between contact surface. According to comparison among the different design, the best one is decided to be applied, the contact resistance is less than 1 uΩ. Finally, the experiment data are present to prove the validity of the design, the average temperature rise of connect terminal is less than 20℃, it can meet the requirement of ITER operation.

        Speaker: Dr li jiang (Power supply and control system, Institute of Plasma Physics Chinese Academy of Sciences)
      • 11:00
        Design of scalable vacuum pump to validate sintered getter technology for future NBI application 2h

        The vacuum systems of neutral beam injectors have very demanding requirements in terms of gas type, pumping speed and throughput. Due to its high affinity to hydrogenic species, non-evaporable getter (NEG) is in principle a good pumping technology candidate for the deployment in neutral beams, which require the injection of a serious amount of hydrogen in order to operate. Getter materials operate at room temperature, and their use could be particularly welcome in the absence of cryogenic supplies, if high temperature superconducting magnets are successfully deployed in future fusion plants. In the past NEGs have not be used due to their insufficient capacity, but with the new materials developed in recent years, a big step forward has been done.The strategy to validate the use of getter pump technology is based on the realization of a relatively large pump mock-up that has to be tested in fusion-relevant conditions. The objectives of this mock-up are to demonstrate that a pump of large dimensions and capacity is usable. This paper deals with the design of the mock-up, based on conceptual studies which involve at first 3D gas flow simulations considering different modular mock-up pumps based on NEG sintered disks. In addition transient thermal simulations with FE method have been performed with the aim to analyze the thermal response of the mock-up. The conceptual design has been carried out in order to define the best configuration to obtain high pumping speed with low spatial gradient of gas concentration inside the getter material. The suggested solution will exhibit a modular structure of getter disks, which on one hand simplifies the mechanical assembling, and on the other hand allows interpretative modelling at different scales. It is foreseen to test the pump mockup in the TIMO facility at KIT Karlsruhe.

        Speaker: Dr Marco Siragusa (Consorzio RFX)
      • 11:00
        Design of the ITER EC upper launcher nuclear shielding 2h

        ITER will be equipped with four EC (Electron Cyclotron) upper launchers of 8 MW microwave power each with the aim to counteract plasma instabilities during operation. These launcher antennas will be installed into four upper ports of the ITER vacuum vessel.
        Beside their functional purpose the port plugs which are the structural system of the launchers have to provide as much shielding as possible in order to protect adjacent components from neutrons and photons and to minimize the shutdown dose rate in the port interspace and the port cell, being located further back in the ITER Tokamak building.
        In addition to the basic design of the port plug cask several components are particularly shaped in order to achieve maximum shielding performance. Moreover three individual shield blocks will be installed at prominent positions inside the plug.
        This paper presents the general design, the technical integration and mechanical stress analyses for all shielding components. Also the thermal-hydraulic properties and the integration of these water-cooled elements into the port plug´s cooling circuit are outlined.

        Speaker: Dr Peter Spaeh (IAM-AWP, KIT)
      • 11:00
        Deuterium retention behavior in tungsten irradiated with neutron under divertor operation temperature 2h

        In the fusion reactor, tungsten will be exposed to high heat flux, neutrons, ash and fuel plasma of fusion reaction including tritium. The irradiation defects generated by neutrons will dynamically migrate, which results in the accumulation and annealing of irradiation defects. The irradiation defects in tungsten will act as potential trapping sites for hydrogen isotopes and, therefore, increase the hydrogen isotope retention.
        Tungsten samples were irradiated by neutrons in HFIR (High Flux Isotope Reactor) in ORNL (Oak Ridge National Laboratory) up to 0.5dpa at temperatures of 1073 and 1373 K (named as AW-51 and AW-53 according to the sample ID in ORNL, respectively), which are equivalent to the divertor operation temperature in DEMO. Then, the samples were exposed to deuterium plasma at 673 K, and deuterium retention was evaluated by TDS (Thermal Desorption Spectroscopy) conducted in INL (Idaho National Laboratory).
        The deuterium desorption spectrum for AW-51 showed deuterium desorption peak at around 850 K. That of AW-53 was also at the same temperature. This indicates the species of irradiation defects should be the same between these tungsten samples. Besides, deuterium retention in AW-53 was almost half compared to that of AW-51. It was suggested that the irradiation defects induced in tungsten annealed during neutron irradiation under high temperature. Consequently, deuterium retention was reduced for AW-53 due to the lower concentration of irradiation defects.

        This material is based upon work supported by the U.S. Department of Energy Office of Science, Office of Fusion Energy Sciences, under the DOE Idaho Operations Office contract number DE-AC07-05ID14517 and under the UT-Battelle, LLC. contract DE-AC05-00OR22725.
        Sandia National Laboratories is a multimission laboratory managed and operated by National Technology and Engineering Solutions of Sandia, LLC, a wholly owned subsidiary of Honeywell International, Inc., for the U.S. Department of Energy’s National Nuclear Security Administration under contract DE-NA0003525.

        Speaker: Dr Makoto Kobayashi (National Institute for Fusion Science)
      • 11:00
        Development of an electrochemical sensor for hydrogen detection in liquid lithium for IFMIF-DONES 2h

        The structural materials in fusion reactors as DEMO and future power plants are under strong irradiation and will suffer from radiation damages. The knowledge of the radiation induced degradation is planned to be investigated in IFMIF-DONES, a facility in which fast neutrons are produced by a reaction of a D-beam with a liquid lithium target. The operation of such a system requires the control and measurement of impurity concentrations in the melt, thereunder hydrogen. Electrochemistry offers diagnostic tools to measure directly concentrations in such media by online-monitoring systems. Based on this technology, an electrochemical H-sensor for operation in liquid lithium is under development.

        This presentation will line-out the physical background of measuring non-metallic impurity levels in molten metals by measuring electrochemical potentials and their transformation by Nernst correlation into concentrations. Liquid lithium is a very reactive melt, thus material section will be an essential topic for development of a reliably working sensor together with the materials, which can be used as hydrogen-conducting electrolytes in such a sensor. This material behaviour and properties will be discussed. Based on these issues a sensor design for hydrogen in liquid lithium was set up. The successful manufacturing and assembling of the sensor will be shown beyond the synthesis of the electrolytes which are essential for a pre-qualification of the sensor in liquid lithium. These tests are conducted under cleaned Ar atmosphere in a glove-box system as well as the whole sensor assembling. The outlook will deal with measurements in lithium of different hydrogen concentrations.

        Speaker: Nils Holstein (Institute for Applied Materials Karlsruhe Institute of Technology)
      • 11:00
        Development of data acquisition and control system for quasi-2D turbulent electrolyte flow experiment 2h

        A new experimental device has been designed, manufactured and tested for quasi-2 dimensional turbulence studies in magnetized electrolyte system. This experiment can provide significant information about the interaction of large scale shear flows (zonal flows) and smaller scale turbulent vortices. This physics problem has a relevance in different scientific areas such as the turbulent transport reduction via shear flows in magnetically confined fusion plasmas.

        14x14 pieces of N52 10x10x10 mm neodymium magnets have been placed below a plastic container in different geometrical configurations in order to create a large variety of turbulent drives together with a few amps DC electrical current which has been driven through the electrolyte (NaCl). Beside the small vortices induced by the permanent magnets, a large (system-size) seed flow has been also generated by a solenoid placed below the flow. Using this set-up, the interaction of large sheared flows with small scale vortices can be experimentally studied.

        This paper describes the whole system: the hardware and the software developed for the turbulent flow experiment described above. In order to evaluate the velocity field of the flow observed by a high resolution camera, PIV (Particle Image Velocimetry) techniques have been used. Matlab, Simulink, sensors and actuators with microcontroller for data acquisition and process control have been implemented. We also developed our own Matlab routines for automated evaluation of the measurements, enabling remote (overnight) data processing.

        Speaker: Dr Erik Walcz (Department of Plasma Physics, Wigner Research Centre for Physics,)
      • 11:00
        Development of Experimental Helium Cooling Loop (EHCL) for testing nuclear fusion blanket components 2h

        Institute for Plasma Research is developing an Experimental Helium Cooling Loop (EHCL) as a part of R&D activities in fusion blanket technologies. This helium cooling system is designed for testing various nuclear fusion components such as tritium breeding blanket, helium-cooled divertor, and any other components which can be operated within EHCL operating window. The cooling channels of breeding blanket and divertor comprise of complex channel geometry having several parallel channels carrying helium gas for efficient heat extraction. Several mock-ups of these systems need to be tested before finalizing the design and fabrication. In addition to the individual testing of mock-ups of breeding blanket and divertor, integrated operation of the loop as well as understanding the behaviour of high-pressure and high-temperature system components are very essential for the development of Helium Cooling System for a fusion reactor.

        This paper discusses the preliminary process design, process & instrumentation details, and operating & design parameters of the EHCL system. It also describes the characteristics of the EHCL system and the mechanism used to control various loop parameters. It briefly discusses about the architecture of major subsystems and components of the loop. Performance test results of the circulator(s) with its associated loop are also presented in this paper.

        Speaker: BRIJESH KUMAR YADAV (ATOMIC ENERGY INSTITUTE FOR PLASMA RESEARCH)
      • 11:00
        Development of medium size DOME & reflector plate for ITER like tokamak application 2h

        A medium sized water-cooled Divertor DOME has been manufactured at Institute for Plasma Research (IPR), India. Divertor Plasma Facing Components (PFCs) such as DOME and Reflector Plate has multi-layered joints which are made of various materials such as Tungsten (W), OFHC Copper (Cu), Copper alloy (CuCrZr) and SS316L etc. Joining of such multi-layered joints is known to be problematic as being used of several dissimilar materials. Vacuum brazing route was employed to fabricate the medium size DOME as well as the Reflector plate. In order to evaluate the performance of the DOME against ITER-like scenarios, the DOME has been successfully tested for 1000 numbers of steady-state thermal cycles with an incident heat flux of 3.87 MW/m2 in the High Heat Flux Test Facility (HHFTF) at IPR. Subsequent testing with additional 200 thermal cycles was also done with an incident heat flux of approx. 6 MW/m2 at 24.6 kW of Electron beam power. The absorbed heat flux was calculated to be 4 MW/m2 which indicated the absorbed power was nearly 70%. During the HHF tests, the surface temperature of the W tile reached 640oC and the beam power was restricted to 24 kW due to the temperature limit of 450oC at the CuCrZr heat sink. A total 1200 cycles of steady-state thermal cycles have been completed. Engineering analysis on the HHFT of the DOME has been performed using Finite element method (FEM) and Computational Fluid Dynamics (CFD) to simulate and to correlate with the experimental data.
        Ultrasonic immersion technique (NDT) was incorporated to inspect the brazed joint quality of the Reflector plate and the DOME mock-up before and after the HHFT. The results of the experimental details, engineering analysis and methodology adopted to fabricate the DOME and reflector plate will be presented in the paper.

        Speaker: Dr Premjit Singh Kongkham (High Temperature Technology Division, Institute for Plasma Research)
      • 11:00
        Development of plasma control algorithm design via machine learning 2h

        Machine learning has garnered increasing attention within the fusion community in recent years, with much of the focus going toward implementation of disruption predictors. However, disruption detection is but one possible area in which the large body of fusion experimental data, accrued over decades, can be put to use. In particular, this data can be utilized to assist in the implementation of closed loop controllers, either through augmentation of existing model-based approaches, or via purely data-driven methodologies.

        In this work we explore the use of machine learning for vertical stability control of the DIII-D tokamak. We describe the application of a search and machine learning computing toolchain for system identification of highly non-linear coil/vessel/plasma interactions as a function of equilibrium state. Extraction of the training data used in this process is achieved at rates over two orders of magnitude greater than previously attainable. The identified system model is then integrated with a model predictive controller and tested in simulation. Additionally, we investigate creation of a purely data-driven vertical control algorithm (using, for example, reinforcement learning). Development toward integration of these algorithms for real time use in the DIII-D plasma control system is discussed. The use of large-scale machine learning techniques for plasma control is still novel within the fusion community, and this work provides a template for future data-driven approaches.

        Work supported by General Atomics’ Internal Research and Development Funding and in part by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences, using the DIII-D National Fusion Facility, a DOE Office of Science user facility, under Award No. DE-FC02-04ER54698.

        Speaker: Dr Brian Sammuli (General Atomics)
      • 11:00
        Development of power combination system for high-power and long-pulse ICRF heating in LHD 2h

        In the Large Helical Device (LHD), the development of high-power and long-pulse ICRF system is ongoing. Frequency was fixed at 38.47 MHz for the optimization of devices. At this frequency, plasma is heated with the minority ion heating of hydrogen and the second harmonic heating of deuterium. Field-Aligned-Impedance-Transforming (FAIT) antenna has the potential performance of high-power injection of more than 1.8 MW. In order to reduce the voltage in the transmission line, Ex-Vessel Impedance Transformer was also developed. However, the output power of the final power amplifier (FPA) is less than 1.2-1.3 MW for the stable oscillation in the LHD. In order to increase power into the FAIT antenna, a power combination system was developed. The target is injection power of 2 MW for 2-3 s and 1 MW for 10 min. By iterating the simulation of the electromagnetic field, optimized power combiner was designed. Then the combiner was fabricated and installed in the transmission system. The power combiner has two input ports and two output ports. Two FPAs are connected to the input ports. From one output port the combined power will be transmitted to the FAIT antenna. Another output port is connected to a dummy load. Air ducts for the air cooling are attached for the long-pulse operation. It was confirmed by the measurement with the network analyzer that the power combiner has almost perfect isolation between input ports and there are no reflections from these ports when there is no reflection at the output ports. Control of power and phase of foreword waves into the input ports are important for the combination of the waves without power loss. Therefore, a real-time control system was developed and demonstrated. Step time of the control was less than 1 ms and power loss into the dummy load was successfully cancelled.

        Speaker: Dr Kenji Saito (National Institute for Fusion Science)
      • 11:00
        Development of reliable tooling and processes for remote maintenance of ITER cooling water connections 2h

        The high neutron flux inherent in fusion reactors creates high heat loads in the components surrounding the plasma. These heat load needs to be managed through active cooling. These components also become highly activated so require remote maintenance, hence the connection and disconnection of these cooling systems becomes an important functionality of these maintenance activities. The integrity of these connections is also of critical importance both to the operation of the components and to maintain the confinement of the activated effluent they contain; therefore, it is imperative that a reliable system of connection and disconnection is established.
        TIG welding is a process that is recognised within nuclear design standards as a reliable technique that can be used for connections on confinement barriers. RACE has performed extensive work across a range of applications in ITER such as the diverter, diagnostic port plugs and neutral beam vessel, and pipe sizes varying from DIN 25 to DIN 200 to develop tooling and processes to ensure that high quality TIG welding can be conducted remotely.
        This work shows that autogenous TIG welding is an appropriate technology to conduct this maintenance activity. It also notes that for this technology the pipe thickness joint thickness must be restricted to a maximum of 3mm and that the sulphur content of the stainless-steel material must be closely controlled with smaller tolerances that anticipated by the industrial codes. The use of filler inserts to supplement material is also discussed. It concludes that the control of the alignment of the tooling is critical to success and suggests design solutions that can achieve these requirements.

        Speakers: Dr Chris Lamb (UKAEA), Nicholas Sykes (RACE UKAEA)
      • 11:00
        Development of Winding Technology for ITER PF6 Double Pancakes 2h

        The Poloidal Field(PF) coils are one of the main sub-system of ITER magnets. The PF6 coil is being manufactured by the Institute of Plasma Physics, Chinese Academy of Sciences(ASIPP) as per the Poloidal Field coils cooperation agreement between ASIPP and Fusion for Energy(F4E).

        ITER PF6 winding pack is composed by stacking of 9 double pancakes. Series double pancakes are being wound in ASIPP with a “two-in hand” configuration. This paper focus on the main winding process and results of ITER PF6 double pancakes. The winding workshop is composed by two symmetric winding line. During each double pancake winding, two conductors were simultaneously de-spooled, straightened, ultrasonic cleaned, sandblasted and then bent to the correct radius. Followed by manual cleaning, the conductors were wrapped with turn insulation by automatic wrapping head. Finally the conductors were accurately deposited onto the rotary table. 0.05% conductor forwarding length measurement and ±0.5mm radial build-up for each turn were achieved, which indicated well winding controlling. By now, 6 out of 9 ITER PF6 double pancakes winding has been successfully accomplished .

        Speaker: Dr Huan Wu (Institute of Plasma Physics, Chinese Academy of Sciences)
      • 11:00
        Digital valve system for ITER remote handling – design and prototype testing 2h

        ITER-RH system is used to exchange the divertor’s 54 cassette assemblies in the vessel. Water hydraulics and servo valves are currently used in the task requiring high accuracy tracking and the use of de-mineralized water. The main concern has been robustness of the technology. Only few suitable commercial water servo valves exist and problems e.g. with jamming and wear been encountered. A possible mitigation is to use redundant valves but ensuring required level of water cleanliness is still an issue.

        An alternative option is to use digital technology where on/off valves and intelligent control is used to produce proportional output. So far, no proper high-pressure on/off valves existed in the market and therefore a new concept was developed and tested with promising results. The valve has very fast response time, flow capacity that suits well for the required velocities and is compatible with de-mineralized water. In addition, the valve is not sensitive against water cleanliness and can be made rad-hard when necessary.

        A mock-up of the remote handling system was used as test bench. The system was simulated in order to dimension the valve system and to tune the valve controller. A complete valve package with 16 prototype on/off valves and a manifold was manufactured and assembled. The rest of the components were off-the-shelf and the result is fully compatible to be used as direct replacement for the servo valve system.

        Long-term tests of 60 working days and 2000+ hours was made. A new level of performance was demonstrated throughout the tests as tracking accuracies were approximately ten times better than with servo valve. Although some design faults where detected and system had faulty components, tracking accuracy was very good. Tracking and positioning capability of digital valve system exceeds all requirements and seems applicable for ITER-RH application.

        Speaker: Lauri Siivonen (Tamlink Oy)
      • 11:00
        ECART analysis of the STARDUST dust resuspension tests with an obstacle presence 2h

        The activated/toxic dust resuspension inside the vacuum vessel of future fusion devices as ITER or DEMO is a safety issue of main concern. In case of a LOVA or a LOCA, dusts produced during the normal and off-normal conditions can be released inside the tokamak building or towards the external environment. These accidents are not expected during the whole lifetime of the ITER machine, though in the past they were considered in the ITER Generic Site Safety Report with a conservative assumption for the lack of reliable resuspension models in the employed safety codes. To relax this strong assumption and validate resuspension models in fusion like conditions, different experimental campaigns in the STARDUST facility were performed at the ENEA laboratories in Frascati (Rome). In the first experimental campaign (2004), the resuspension of Tungsten (W), Carbon (C) and Stainless Steel (SS) dusts was investigated in an “empty tank” configuration, while the resuspension of the same dust types in presence of an internal obstacle was studied in the second campaign (2005). The obtained experimental results stressed that only a minor fraction of dusts is effectively resuspended. In the present work, focalized on the ECART code validation for the safety analysis of future fusion installations, a further step in the assessment of the semi-empirical “force balance” resuspension model implemented in the ECART code against the data obtained during the second experimental campaign (tank with an inner obstacle) is performed. The code predictions are quite in agreement with these STARDUST experimental data, and three charts (one for each dust type) were elaborated to predict the resuspension magnitude basing on the flow velocity on the structure where the dusts are initially collected.

        Speaker: Dr Sandro Paci (Pisa University)
      • 11:00
        Effect of coil configuration parameters on the mechanical behavior of the superconducting magnet system in the helical fusion reactor FFHR 2h

        FFHR depicts the conceptual design of an LHD-type helical fusion reactor that is being developed by the National Institute for Fusion Science. Several design mechanisms for FFHR have been investigated. For instance, FFHR-d1A is a self-ignition demonstration reactor that operates at a magnetic field intensity of 4.7 T and has a major radius of 15.6 m. FFHR-c1 is a compact-type sub-ignition reactor that intends to achieve steady electrical self-sufficiency and that depicts a high magnetic field intensity of 7.3 T and a small major radius of 10.92 m. For both types of FFHR, the shape and positional relation among the superconducting coils, a pair of helical coils with two sets of vertical field coils, are observed to be similar with each other. Such a relation is based on the coil configuration of LHD. The coil configuration is defined using an aspect ratio, a pitch parameter of the helical coil, the number and geometric position of the vertical field coil, and so on. There is an increasing demand to achieve an optimized coil configuration to anticipate the improvement in plasma-confinement conditions. However, there are a few investigations that study the effect of coil configuration on mechanical behaviors of the coil systems, including the support structure. In this study, the structural design of FFHR-c1 based on the fundamental set of parameters of coil configuration is depicted, which satisfies the soundness of the structure. Further, the effects of the coil configuration parameters on the stress/strain distributions are investigated. A multiscale analysis is performed simultaneously to facilitate a detailed investigation of the elements in the superconducting coil. Furthermore, the mechanical behaviors during the coil excitation process are evaluated based on the contact between the coil winding and the support structure.

        Speaker: Dr Hitoshi Tamura (National Institute for Fusion Science)
      • 11:00
        Effect of plasma screening on pumping efficiency in the DEMO divertor 2h

        The screening of a neutral gas by plasma from the top of the private flux region (PFR) of the latest DEMO divertor configuration without the dome structure is analysed. The effect of the neutral gas compression in the PFR is assessed by using the direct simulation Monte Carlo method (DSMC) and the ambipolar approximation for the simulation of neutral molecule dissociation and ionization as a result of collisions with plasma electrons.
        Preliminary results without the divertor dome have been reported previously [1], in which the atomic and molecular processes as well as the interaction with plasma were neglected. It was shown that a strong neutral reflux from the private flux region towards the x-point occurs. The aim of the present work is to investigate the impact of neutral interaction with electrons on particle reflux. The plasma fan could impede the outflow of neutrals to the bulk plasma due to the sharp electron density and temperature drop towards the PFR, resulting in an effective neutral plugging.
        The numerical analysis includes the calculation of neutral density, temperature and pressure in the divertor plenum and the overall conductance of the sub-divertor structure, which consequently affects the pumping efficiency. It is shown that plasma plugging is more efficient in the case of plasma attachment. Analysis of dome structure requirements for achievement of the effective pumping is extended by including the plasma effect as a plugging of neutrals in the PFR. The DEMO divertor cassette structure can be further optimized to ensure more efficient pumping and achievement of detachment conditions.

        [1] Chr. Day, S. Varoutis, Yu. Igitkhanov, IEEE Trans. Plasma Science 44 (2016) 1636-1641.

        Speaker: Yuri Igitkhanov (ITEP Karlsruhe Institute of Technology (KIT))
      • 11:00
        Effects of process variables on microstructure and tensile properties of friction stir welded ARAA 2h

        Plasma facing component such as breeding blanket and divertor in fusion reactors are supposed to be assembled by welding and joining of parts made of reduced-activation ferritic-martensitic (RAFM) steel, and accordingly the structural integrity is significantly affected by the properties of the joint. Conventional fusion welding results in a wide heat-affected zone (HAZ) where a well-known type IV cracking occurs in RAFM steels under creep condition. In the present work, an attempt was made to butt weld the ARAA (advanced reduced-activation alloy; the Korean RAFM steel) sheets by the friction stir welding (FSW) technique, for which we investigated how the welding condition and post-weld heat treatment (PWHT) affect the microstructure and tensile properties of the joint. Microstructure of the as-welded ARAA was found to consist of a martensitically transformed zone (MTZ), a mixed zone (MZ) of fresh martensite and heat-affected tempered-martensite structures, HAZ, and base material in order outward from the welded interface. The hardness measurements showed a significant hardening in MTZ, an abrupt decrease of hardness towards HAZ in MZ, and the lowest hardness in HAZ. It is noteworthy that increasing the rotational speed and applied load, or decreasing the traveling speed increased the width of MTZ and reduced that of HAZ, which in turn led to increases of the tensile yield strength and ductility of the as-welded ARAA. PWHT conducted at 700 C resulted in transformation of martensite to tempered-martensite in MTZ and MZ while no significant change occurred in other zones. Such changes of microstructure reduced the gradient of hardness between MTZ and HAZ and the degree of strain-localization in HAZ. It is proposed that FSW technique can be utilized for butt welding of thin RAFM steel parts.

        Speaker: Young-Bum Chun (Nuclear Materials Development Team Korea Atomic Energy Research Institute)
      • 11:00
        EHF FW panel for ITER BM with mechanical attachment of the plasma-facing components 2h

        The qualification Full Scale Prototype (FSP) of the Enhanced Heat Flux (EHF) First Wall (FW) for the ITER blanket will be manufactured by JSC NIKIET and JSC NIIEFA as part of the Procurement Arrangement between the ITER Organization and the Russian Federation Domestic Agency. The FSP design is based on the FW panel #14 type A and includes: the supporting structure (FW beam), plasma facing components (FW fingers) and the central slot insert (CSI) to protect the FW beam/fingers welded joint from plasma-induced heat flux.
        In 2017 the EHF FW panel design has been revised in order to improve the repairability and to provide non-destructive inspection of the FW components during the final assembly stage. The modifications include replacing the FW beam/fingers welded joint by a mechanical attachment. The mechanical attachment is located under the FW fingers and is therefore shielded from plasma-induced heat flux. This also allows the possibility of excluding the CSI from the FW panel design which would simplify the cooling channel layout; however the thermal load on the middle part of the FW beam in the absence of the CSI needs to be further assessed before a decision can be taken.
        In this connection JSC NIKIET specialists have performed parametric analyses of this FW panel design, to investigate the effect of the depth of the slot in the middle of the panel.
        This paper describes the revised design of the EHF FW Panel with a mechanical attachment of the plasma facing components to the FW beam, and discusses the results of the corresponding thermal and structural analyses.

        Speaker: Dr Tomilov Sergey (JSC NIKIET)
      • 11:00
        Electromagnetic modelling and design of DEMO and disruption location prediction 2h

        The design study of a DEMOnstration (DEMO) Fusion Plant is one of the main points of the European Roadmap to Fusion Electricity [F. Romanelli, http://www.efda.org/wpcms/wp-content/uploads/2013/01/JG12.356-web.pdf]. The pre-conceptual design phase of DEMO is presently used to explore a flexible range of the main machine geometrical design parameters, including machine magnetic configurations, and optimisation of plasma scenarios and conductive geometries.
        In this paper is presented the electromagnetic modelling of the DEMO baseline scenario, including the analysis and design activities on the plasma surrounding electrically conductive structures, and their influence on the passive vertical stabilisation (VS), which is a machine size driver, due to the large amount of power needed to vertically stabilise the plasma, and the limitations on the poloidal field coils. A careful design of the blanket first wall poloidal geometry was performed, taking into account the plasma heat load on the wall, allowing the minimisation of the distance between the plasma and the vacuum vessel, which is the closest electrically conductive toroidally continues structure . Both the improvements on the passive and active [R.Ambrosino, this conference] VS allowed to increase the maximum controllable plasma elongation at 95% of the separatrix, from 1.59 to 1.65, from the previous to the most recent DEMO baseline, corresponding to an increase on fusion performances.
        Finally a study was carried out on the possibility to predict the plasma final position, following a vertical displacement event, which is essential for a DEMO wall protection strategy from plasma transients. Such point is located where the field after disruption is tangential to the FW and pushing the plasma onto it, under the assumptions that the current quench time is faster than the L/R time constant of the passive structures. The final position was also verified using dynamic simulations and images from JET fast camera during a VDE.

        Speaker: Dr Francesco Maviglia (PPPT, EUROfusion PMU)
      • 11:00
        EM analyses on the 55.NE.V0 loom system and attached components 2h

        The ITER vacuum vessel (VV) contains transducers for vessel and blanket instrumentation and for the measurement of plasma performance. The electrical and optical signals to and from these transducers are managed through the electrical services infrastructure project 55.NE.V0. This system is responsible for transmitting electrical signals from the VV inner skin to the outside of the vacuum where they interface to conventional signal transmission and data acquisition systems. Around 1400 sensors/junction boxes/connectors and more than 1700 MI cables are distributed throughout the vacuum vessel inner surface.

        This paper concerns EM activities performed as part of the full set of engineering analyses necessary to assess the design configuration feasibility.

        Transient electromagnetic analyses were performed for 4 representative looms systems, placed at the inboard and outboard region, and exiting the VV from lower and upper ports. Highly populated looms in the divertor region (named 55.NE.D0) were analysed for the most demanding downward plasma disruptions.

        All the cables were modeled using dedicated EM “wire elements”, an element type developed by LTCalcoli, suitable to model eddy currents in wire-shaped structures.

        The calculated distributed EM loads along the cables and in attachment components allowed the most loaded cable branches, and the worst position and disruption for connecting clamps and junction boxes, to be identified.

        Detailed 3D EM models were developed for the most loaded junction boxes and clamps and integrated in the ITER Global Model.

        The detailed analyses allowed the Lorentz forces and resultant torques on the different sub-structures, and the force distribution element by element, to be evaluated, ready for subsequent structural assessments.

        Since the clamp bosses are welded onto the VV which is a confinement boundary and ESPN equipment, the analyses conducted contribute to demonstrating the absence of impact on the VV. This work is considered to be a Protection Important Activity (PIA).

        Speaker: Dr Claudio Bertolini (L.T.CALCOLI S.R.L.)
      • 11:00
        Endoscopes for observation of plasma-wall interactions in the divertor of Wendelstein 7-X 2h

        The stellarator Wendelstein 7-X has been prepared for long pulse operation in the first operational campaign. Forschungszentrum Jülich has contributed diagnostics for investigation of plasma wall interaction processes in presence of an island divertor and steady state plasma at high density and low temperature.
        A versatile optical observation system has been developed for local characterization of the divertor plasma and the divertor target surface. At two opposite divertor locations linked by magnetic field lines, temperature and density profiles, impurity transport and recycling, hydrogen recycling, and transient heat load deposition are analysed. The analysis is assisted by two gas inlets in the divertor, a fast probe manipulator and a poloidal correlation reflectometer located along the same magnetic structure.
        The optical systems consist of two endoscopes each with perpendicular fields of view with the opportunity of tomographic reconstruction. Mirror based optics has been chosen in order to be independent of wavelength. A narrow field of view allows for high spatial resolution while rotation of the first mirror covers the full poloidal divertor sections. An integrated shutter mechanism and a vacuum window far back minimize coating of optic components. For assessment of change of transmission, a relative calibration function is implemented. The output light is split into wavelength ranges. Both, cameras equipped with narrow band filters as well as spectrometers are connected. Six cameras and four spectrometers can be attached simultaneously to each endoscope.
        In this contribution, the detailed optics, mechanic, and thermal concept, experience from assembly and adjustment as well as first results will be presented.

        This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission.

        Speaker: Dr Olaf Neubauer (IEK-4, Forschungszentrum Jülich GmbH)
      • 11:00
        Enhanced Droplet Control for the Fabrication of Ceramic Breeder Pebbles 2h

        Lithium rich advanced ceramic breeder pebbles composed of lithium orthosilicate with a strengthening phase of lithium metatitanate are intended as tritium breeders for future fusion reactors. The EU breeding blanket being designed for trial in ITER will feature the pebbles in the form of pebble beds in the wall of the reactor. Upon irradiation with neutrons, the lithium will decay into tritium and helium, after which the tritium is processed before being rerouted into the reactor core to react with deuterium and completing the fuel cycle. It is imperative that the pebbles are of sufficient quality to ensure the smooth functioning of the breeder blanket.
        At the Karlsruhe Institute of Technology, a melt based process is used to produce the pebbles. A melt is formed in a platinum alloy crucible from synthesis powders at approximately 1400 °C. Upon ejection through a 300 µm nozzle, a laminar jet is formed which breaks up into small droplets due to Plateau-Rayleigh instabilities. These describe how random instabilities grow on the surface of the jet until the surface tension overcomes the viscous forces, causing droplet break-off. However, as the break-up is considerably irregular, small changes in the droplet sizes will result in droplets with different momentum causing more to coalesce, resulting in oversized pebbles.
        In order to control the break-up of the jet, and indirectly the size of the pebbles, instabilities in the form of audio waves were deliberately applied to the system. A high-speed camera was used to study the effects of varying frequencies on the behaviour of the jet. By applying the correct frequency, it was possible to increase both the yield as well as the monodispersity of the product by minimising the number of droplets coalescing, resulting in less waste and more control of the end product.

        Speaker: Dr Oliver Leys (IAM-KWT Karlsruhe Institute of Technology)
      • 11:00
        Environmenta,l steam-ingress and gamma irradiation tests for optical materials candidates for ITER equatorial-Vis/IR-WAVS diagnostic 2h

        To validate the design of Equatorial Vis/IR WAVS (Wide Angle Viewing System) diagnostic it is necessary to test materials and coatings of optical components and to characterize their behavior under the harsh environment of ITER (in particular in terms of radiation and temperature). The objective of this work is to summarize the results of the different tests: environmental (temperature, air, vacuum, steam-ingress) and gamma irradiation, carried out in different materials (candidates for port plug and optical hinge mirrors, vacuum window, field lens and beam splitters), in order to provide information for choosing relevant optical materials for this diagnostic.

        Speaker: Angel Ibarra (CIEMAT)
      • 11:00
        Equilibrium evaluation for OP1.2a Wendelstein 7-X experiment programs 2h

        Wendelstein 7-X (W7-X) is a modular advanced stellarator, which successfully went into operation in December 2015 at the Max-Planck-Institut für Plasmaphysik in Greifswald, Germany and continued to thrive at the experimental campaign OP1.2a (August-December 2017). The term modular stellarator refers to a generalized stellarator configuration with nested magnetic surfaces created by a system of toroidally discrete coils, providing both toroidal and poloidal field components, and designed with the aim to create optimum equilibrium properties. The optimization criteria included the high quality of vacuum magnetic surfaces, good finite beta equilibrium and MHD-stability properties as well as a substantial reduction of the neoclassical transport and bootstrap current in comparison to classical stellarators. The mission of the W7-X project is to demonstrate the reactor potential of the optimized stellarator line.

        Equilibrium calculations are the basis for the mapping of various diagnostics in different poloidal planes with different shapes onto the flux surfaces and analysis of the experimental results. Simulations were performed with help of the Variational Moments Equilibrium Code (VMEC), available as Wendelstein 7-X web service. Pressure profile modeling was based on OP1.2a experimental results. For this purpose recent data, obtained from the Thomson Scattering system, High Resolution X-ray Imaging Spectrometer and the effective ion charge Zeff measurements, were utilized. The equilibrium evaluation is presented in the context of long pulses (> 10s) dedicated to bootstrap current measurements.
        
        Speaker: Dr Tamara Andreeva (Max-Planck-Institut fuer Plasmaphysik)
      • 11:00
        EUROfusion Plasma EXhaust (PEX) strategy on upgrades to European tokamaks and PFC test facilities 2h

        The European fusion electricity roadmap sets out a strategy for a collaboration to achieve the goal of generating fusion electricity by 2050. It has been developed based on a goal-oriented approach with eight different missions including development of Heat-Exhaust systems which must be capable of withstanding the large heat/particle fluxes of a fusion power plant. This paper summarises the development of the strategy to achieve this mission, focussing on the programme of coordinated upgrades to European tokamaks and PFC facilities.
        The strategy pursues the conventional ITER divertor solution in a combination of radiative cooling and detachment, which is taken as the baseline for DEMO. Nevertheless, a significant risk remains that high-confinement regimes of operation are incompatible with the larger core radiation fraction required in DEMO. Therefore, an investment in assessment of the adequacy for DEMO and proof-of-principle tests of innovative geometries as well as the use of liquid metals (LM) is also required. The broad tasks and milestones for both the conventional and risk mitigation approaches for developing the solutions at the levels of Proof of Principle, Demonstration and Qualification, including implementation on DEMO will be summarised.
        To enable implementation of this strategy a dedicated programme of PEX facilities upgrades and new devices was developed. The selection of projects in the strategy was based on the degree to which the projects enabled the physics and technology gaps to be addressed, prioritising of studies on conventional materials and divertor configurations. Currently the programme includes upgrades on the three MST tokamaks (AUG, TCV and MAST-U), WEST tokamak and a hot cell facility. The assessment PEX activities and the decision to down select alternatives for possible implementation on DTT are scheduled for 2023-2024. The results of this analysis and current status and progress of the PEX upgrades will be presented and discussed.

        Speaker: Dr Mikhail Turnyanskiy (ITER Physics, EUROfusion PMU)
      • 11:00
        European DEMO divertor R&D activities: loads, design concepts and technologies 2h

        As the first funding period of EUROfusion Consortium is nearing its end, the work package “DEMO divertor” (WPDIV) is entering the final project period concluding its preconceptual design activities. The primary mission of WPDIV is to deliver holistic design solutions of the divertor targets as well as the cassette and to assure the availability of required technologies at least with a preconceptual maturity. Numerous tasks have been performed in WPDIV covering the full spectrum of design study and technology development including design rationales, CAD models, multiphysics analysis and load specifications, cooling scheme, design optimization, joining technology, mock-up fabrication, inspection, high-heat-flux (HHF) fatigue tests, post-examination, failure modelling, and guidelines for structural design and analysis. After 5 years of the preconceptual phase, a remarkable progress has been achieved with successful outcomes. In this contribution, an overview and the latest results of WPDIV works are presented. One of the highlights will be the recent HHF qualification testing campaign where several different design concepts of plasma-facing components were evaluated in terms of fabrication quality, HHF fatigue performance and structural reliability using small scale mock-ups. All design concepts shall be subject to down-selection process to be made eligible for the subsequent industrial upscaling stage. Further highlights are specification of loads and optimization of the cooling circuits for the entire divertor cassette system. To this end, an integrated multiphysics approach was employed covering neutronic, thermal, hydraulic, electromagnetic and structural stress analysis. In addition, a brief overview is given on the supporting tasks such as corrosion (coating, testing), inelastic design guidelines, CAD of cassette fixation system and non-destructive inspection techniques. Finally, the future R&D strategy is briefly outlined, including an indicative description of the planning and preparation for the following conceptual design phase.

        Speaker: Dr Jeong-Ha You (Max-Planck-Institut für Plasmaphysik)
      • 11:00
        Examination of ITER Central Solenoid prototype joints 2h

        The ITER Magnet System will be the largest and most challenging integrated superconducting magnet system ever built. For the Central Solenoid (CS), cable – in – conduit - conductors (CICCs) of nearly one kilometre length are produced, but still, it will be necessary to connect several lengths together to wind the gigantic 110 tonnes coils. The creation of these superconducting joints is one of the most delicate parts of the assembly.

        There are three types of ITER CS joints: sintered joints, coaxial joints and twin – box joints. US ITER, the ITER Domestic Agency of the USA produced a prototype containing all three types of joints. The goal is to test the performance of the joints in the SULTAN facility of the Swiss Plasma Center, capable of reproducing close – to – service conditions: high magnetic field (up to 11 T background field), high current (up to 100 kA) and high mass flow rate of supercritical helium for cooling.

        The paper describes the results of a comprehensive examination campaign aimed at understanding the relation between fabrication and performance of the CS prototype joints. The test campaign combines advanced image analysis for the assessment of the void fraction of the conductors and dimensional measurements performed at different cross-sections of the joints, scanning electron microscope (SEM) and energy dispersive x-ray spectroscopy (EDX) to evaluate the quality of the contact between the strands and the sleeve / sole, as well as a full assessment of the welds according to the most stringent acceptance levels of the standards in force.

        Speaker: Dr Stefano Sgobba (European Organization for Nuclear Research (CERN))
      • 11:00
        Examination of ITER Central Solenoid prototype joints 2h

        The ITER Magnet System will be the largest and most challenging integrated superconducting magnet system ever built. For the Central Solenoid (CS), cable – in – conduit - conductors (CICCs) of nearly one kilometre length are produced, but still, it will be necessary to connect several lengths together to wind the gigantic 110 tonnes coils. The creation of these superconducting joints is one of the most delicate parts of the assembly. There are three types of ITER CS joints: sintered joints, coaxial joints and twin – box joints. US ITER, the ITER Domestic Agency of the USA produced a prototype containing all three types of joints. The goal is to test the performance of the joints in the SULTAN facility of the Swiss Plasma Center, capable of reproducing close – to – service conditions: high magnetic field (up to 11 T background field), high current (up to 100 kA) and high mass flow rate of supercritical helium for cooling. The paper describes the results of a comprehensive examination campaign aimed at understanding the relation between fabrication and performance of the CS prototype joints. The test campaign combines advanced image analysis for the assessment of the void fraction of the conductors and dimensional measurements performed at different cross- sections of the joints, scanning electron microscope (SEM) and energy dispersive x-ray spectroscopy (EDX) to evaluate the quality of the contact between the strands and the sleeve / sole, as well as a full assessment of the welds according to the most stringent acceptance levels of the standards in force. The views and opinions expressed herein do not necessarily reflect those of the ITER Organization.

        Speaker: Dr Ignacio Aviles Santillana (European Organization for Nuclear Research (CERN))
      • 11:00
        Experimental analysis of dummy load prototype for ITER coil power supply system 2h

        This paper mainly introduces the experimental analysis of the dummy load prototype, whose functions are to verify the capability of the ITER magnetic power supply systems to operate at their rated power levels without energizing the superconducting coils. The rated inductance of dummy load is 6.73 mH and the pulse test currents are 45 kA, 55 kA and 68 kA. To meet the requirements of the large inductance and different pulse test current classes, a dry-type air-core water-cooling prototype with epoxy resin casting technique has been fabricated, which has 24 layers and 72 turns. The experimental analyses are introduced in detail, focusing on the consistency test, inductance test, and temperature rise test. The consistency of different coil layer is ensured by comparing the voltage wave under the same voltage excitation. The inductance test includes dc inductance test by using the first order circuit method and ac inductance test by LCR instrument and finite-element simulation analysis. The temperature rise test results under different current classes are introduced by comparing with the simulation analyses. The above experimental results are coincided with the requirements and it shows the availability and feasibility of the prototype.

        Speaker: Dr Li Chuan (School of Electrical and Electronic Engineering, Huazhong University of Science and Technology)
      • 11:00
        Experimental and numerical studies on a gas flowing calorimetry for tritium accountability 2h

        A pressure, volume, temperature, and concentration (PVT-c) method, which is widely used to measure the amount of tritium owing to its effectiveness, requires a desorption process from uranium tritides and a transfer process to a measurement tank in the tritium storage and delivery system (SDS) for a tokamak-type nuclear fusion reactor. In addition, the repeated processes for the PVT-c, including the heating of the SDS bed, increase the potential risk of tritium release. For this reason, the application of other tritium measurement methods without the desorption and transfer processes is needed for the SDS bed. In this study, we designed and fabricated a uranium hydride bed with gas flowing calorimetry. Gas flowing calorimetry does not require desorption or transfer processes. Experiments on obtaining the calibration curve for the relationship between the amount of tritium and the difference in gas temperature in the calorimetry have been carried out. In addition, characteristics including the resolution, convergence, and reliability of the gas flowing calorimetry in a uranium hydride bed were obtained. Furthermore, the calorimetry was modeled and simulated under various steady state and transient conditions through a Multi-dimensional Analysis of Reactor Safety (MARS) code. The calorimetry loop consists largely of a uranium hydride bed, a circulation pump, flow controllers, chillers, heaters, buffer tanks, pressure gauges, thermocouples, and resistance temperature detectors. The uranium hydride bed contains 1893.75 grams of depleted uranium as a tritium storage material. A direct current power supply was used to simulate the decay heat of tritium.

        Speaker: Kwangjin Jung (University of Science and Technology)
      • 11:00
        Experimental evaluation of wall shear stress in double contraction nozzle for structural soundness evaluation for liquid Li target of intense fusion neutron source 2h

        This paper reports on an experimental evaluation of wall shear stress in a double contraction nozzle to produce a liquid lithium (Li) target being intended for use as a beam target for the intense fusion neutron sources such as the International Fusion Materials Irradiation Facility (IFMIF), the Advanced Fusion Neutron Source (A-FNS), and the DEMO Oriented Neutron Source (DONES). The current design of the neutron sources requires that the thickness of the liquid Li target be maintained to within ± 1 mm while operating under normal conditions (target speed: 15 m/s, inlet Li temperature: 250 °C, and vacuum pressure: 10−3 Pa). In the IFMIF/EVEDA project, we designed and constructed the IFMIF/EVEDA Li Test Loop (ELTL), and we performed experiments to validate the stability of the Li target. After the end of this project, the ELTL has completed its disassembly work in Feb. 2017 and structural soundness evaluation of the configuration components of the ELTL is being prepared.
        Prior to the structural soundness evaluation, we have investigated the flow characteristics in the nozzle which is a key component producing the stable Li target. The wall shear stress is an essential physical parameter to understand erosion-corrosion by a high speed liquid Li flow, and therefore we evaluated by using an acrylic mock-up of the target assembly experimentally. The working fluid was water whose kinematic viscosity coefficient is similar to that of liquid Li. The velocity distribution in the nozzle was measured by laser-doppler velocimetry and momentum thickness along the nozzle wall was calculated by Buri’s equation. Based on the calculated momentum thickness, we estimated the variation of the wall shear stress along the nozzle wall and showed an importance indicator for erosion-corrosion of the nozzle for the future structural soundness evaluation.

        Speaker: Hiroo Kondo (Advanced Fusion Neutron Source Design Group National Institutes for Quantum and Radiological Science and Technology)
      • 11:00
        Experimental investigation on the interruption performance of a switch based on artificial current zero 2h

        The quench protection switch (QPS) is indispensable to protect the magnet coils from the damage of a quench in a superconducting Tokamak. In this paper, a QPS based on the artificial current zero is involved. The vacuum circuit breaker (VCB), which is driven by a high-speed electromagnetic repulsion mechanism, is used as the main circuit breaker (MCB). Two kinds of commercial vacuum interrupters (VIs), which have electrodes generating axial magnetic field (AMF) and transverse magnetic field (TMF), respectively, are applied. Meanwhile, the breaking current with amplitude of 15-25kA is generated by a LC oscillating circuit. The countercurrent with frequency in range of 500~5000Hz is provided by a commutation branch. The interruption performance of the two VIs under different breaking current and frequencies of countercurrents is investigated. The experiment results indicate that the differences of interruption performance under the two types of magnet fields are enlarging with the increasing of the frequencies of countercurrents.

        Speaker: Dr Sheng Li (Xi'an Jiaotong University)
      • 11:00
        Experimental validation of Enhanced Heat Flux First Wall Panel Mechanical attachment system 2h

        JSC NIKIET is a main supplier of ITER in-vessel components and its responsibility includes the manufacture of the FW beam, finger bodies, and the mechanical attachment system of the Enhanced Heat Flux (EHF) First Wall (FW) Panels in the framework of Procurement Arrangement 1.6.Р1А.RF.01 dated 14.02.2014. The mechanical attachment system comprises the central bolt, threaded barrel and system washers. This FW mechanical attachment system allows the mounting of the FW onto the installed Shield Block (SB) while accommodating all external forces acting during plasma disruptions and normal operation. The design of EHF FW mechanical attachment system has been developed and optimized through a collaboration between the ITER Organization Central Team (IO-CT) and Russian specialists in order to provide the possibility to use it for all the FW panels. Full scale mockups of the FW attachment system components have been manufactured in JSC NIKIET in 2015-2016 and mechanical tests have been performed to demonstrate operation of these mock-ups under a 1 MN cyclic force. This paper summarizes the design description and experimental results of the FW attachment system.

        Speaker: Dr Maksim Sviridenko (JSC NIKIET)
      • 11:00
        Exploration of a fast pathway to nuclear fusion: first thermomechanical considerations for the ARC reactor 2h

        Progress in technological fields such as High Temperature Superconductors, Additive manufacturing, new diagnostics, and innovative materials, has led to new scenarios and to a second generation of Fusion Reactor designs. A new Affordable Robust Compact (ARC) Fusion Reactor, which meets its goal in a cheaper, smaller but even more powerful, faster way to achieve Fusion Energy, has been designed at MIT.

        An investigation of the load-following concept is necessary, in order to prove its feasibility on ARC reactor. We started from ARC’s most close to plasma component, the vessel: finite element analysis models have been designed and thermo-mechanical analysis have been conducted. Thermal fatigue remains the main issue. The study demonstrated that the vessel is able to survive some years in particular conditions such as high temperature and variable thermal loads. This is quite enough, since ARC’s vacuum vessel is thought to be replaced no later than two years, due to neutron embrittlement and neutron induced activation.

        Supports, divertors and connections between the two walls should be designed and investigated, to improve vessel’s resistance to disruptions without causing hotspots and stress concentrations during thermal expansion. Simulations on supports and channels disposition demonstrated that setting channel’s inlet section in the upper side of the vessel, near supports is the best choice.

        Indeed, Inconel 718 is not a good candidate from radioactive waste point of view, due to its neutron-induced activation properties. Therefore, few other materials were investigated, knowing the main material properties needed for vacuum vessel’s structure. Reference materials used in fusion experiments and projects mostly come up to be less attractive than Inconel 718; for example stainless steels 304L and 316L, which show poor thermomechanical properties. However, the study found out that Vanadium alloys such as V-15Cr-5Ti could be a good substitute of Inconel for ARC’s vacuum vessel.

        Speaker: Stefano Segantin (DENERG Politecnico di Torino)
      • 11:00
        Exploratory risk analysis of ITER Cask & Plug Remote Handling System 2h

        An exploratory risk analysis of ITER Cask & Plug Remote Handling System (CPRHS) has been performed under a system engineering approach considering the CPRHS in various operational states with the associated loads.
        A Functional Breakdown Structure was developed from the 4 main functions fulfilled by the CPRHS: to dock, to handle, to transport and to confine. During Tokamak maintenance operations, various operating states and locations were defined for the CPRHS. In regard to the safety function, to confine, specific configurations were considered to capture all relevant loading cases.
        A Failure Mode Effects & Criticality Analysis has been made by identifying potential failures of basic functions to be fulfilled by CPRHS during the maintenance of a Diagnostic Port Plug and quantifying them in terms of Criticality defined as the product of the failure Occurrence and Severity. The Severity rating scale was related to unavailability of the function due to both technical and safety issues.
        Specific analyses of docking operations in the different cells and traveling operations in different rooms while taking into account the safety constraints were made leading to recommend actions for mitigating the failures having the highest criticality levels.
        In order to perform a sensitivity analysis on maintenance duration by means of statistics approach, the failure modes of CPRHS basic functions were introduced in the schedule Primavera Risk Analysis software which uses a probabilistic Monte Carlo approach. The failure modes were considered as task dependent activities with a duration equal to Mean Time To Repair and an existence likelihood equal to the product  x DC where  is the failure rate and DC is the Duty Cycle of the failed basic function.
        CPRHS availability and operations time were then addressed while considering the impact of the failure modes on maintenance operations time and the benefit of mitigation actions.

        Speaker: Didier VAN HOUTTE (IRFM CEA)
      • 11:00
        Exploring the upper measuring limit of pressure gauges for ITER by experimental variation of instrumental parameters 2h

        Neutral gas pressures in the vacuum vessel of ITER will be measured by hot cathode ionization gauges. The design is based on the ASDEX pressure gauge which is operated successfully in many fusion experiments worldwide. Further development is needed to fulfill superior requirements: the upper measuring limit has to be at least 20 Pa in hydrogen at a magnetic flux density of up to 8 T. The required measurement accuracy of 20 % implies sufficient differential sensitivity.
        This work aims at proving compliance of the pressure gauges with requirements of the ITER experiment by means of laboratory tests.
        An experimental campaign was conducted to explore the accessible measuring range by consecutive variation of gauge parameters: electron emission current, electrode potentials and transparency of the acceleration grid. A special prototype was manufactured that features an exchangeable acceleration grid; transparencies from 20 % to 80 % are available. The ion over the electron current as a function of pressure and magnetic flux density was obtained for each parameter set.
        A monotonic behavior was achieved even up to 30 Pa by a reduction of the grid transparency and an increase of the electric field strength at the cathode to accelerate the electrons. While the former leads to a lowering of the ion current, which is unfavorable for the sensitivity limit in the low pressure range, this can be compensated in part by a higher electron current.
        The reproducibility of the gauge response under repeated experimental conditions is to be tested in near future. These results enable the calculation of statistical errors to estimate the measurement accuracy.

        Speaker: Dr Felix Mackel (Max-Planck Institute for Plasma Physics)
      • 11:00
        Fabrication and characterization of Be12V pebbles with different diameters 2h

        Neutron multipliers with lower swelling and higher stability at elevated temperatures are desired for the pebble bed blankets of designed DEMO rector. Among beryllium-based intermetallic alloys, vanadium beryllide Be12V is considered to be an attractive material from the point of view of its potential use as an advanced neutron multiplier of the breeding blanket. Preliminary assessment of its properties showed that, compared to other beryllides, Be12V has a small value of deuterium trapping efficiency and low rate of hydrogen generation by reacting with water vapor.
        Rotating electrode method (REM) was applied in this study for fabrication of sphere-shaped Be12V pebbles using a plasma-sintered Be-V electrode. Be-V electrode as raw material played a role of the target which was melted and centrifugally ejected in the form of spherical droplets in the helium-filled atmosphere of REM apparatus. Since the angular rotation speed of Be-V rod is one of the important technological parameters which significantly influences the sizes of produced pebbles, the wide operating range (2000-6000 rpm) was selected by granulation process and produced batches of pebbles were analyzed. The yield properties, sizes and some characteristics of microstructure of fabricated vanadium beryllide pebbles are summarized and discussed in this work.

        Speaker: Petr Kurinskiy (Department of Blanket Systems Research Fusion Energy Research Development Directorate Rokkasho Fusion Institute National Institutes for Quantum and Radiological Science and Technology QST)
      • 11:00
        Fabrication Status of ITER Centrol Solenoid Modules 2h

        The fabrication of the modules for the ITER Central Solenoid (CS) is in progress at General Atomics (GA) in Poway, California, USA. This purpose built facility has been established with the requisite tools and machines to fabricate the seven 110 tonne CS modules (six required plus one spare). The current schedule has the first module’s fabrication completing in 2018 followed by electrical and full current testing. GA has completed the fabrication of a qualification coil, simulating all of the processes and exercising all of the tooling required to fabricate a production module. After being completed, the qualification coil underwent a cool down cycle to 4.5K which served as the final commissioning step of the helium cooling system on the parallel flow paths of a module. In addition, other final test station systems and procedures were simulated to verify their functionality including simulated Critical Current Sharing Temperature measurements, DC switch operation to dump the energy from the coil, and a helium leak test on the coil at <10K. After warming the coil, Paschen tests were repeated prior to the coil being dissected to evaluate the quality of the vacuum pressure impregnation (VPI). Dissection is scheduled for early spring 2018.
        GA currently has four production modules in fabrication, all in different stages of production. The first module will be vacuum pressure impregnated in the spring of 2018. The second module is currently in the insulation process with the third module being joined together into a 110 tonne module. Module 4 is wound. The description of the VPI process planned for late spring 2018 will be described. To verify the turn insulation of a module, unique impulse testing methods were developed to test the insulation between the 560 turns of a module and these methods are reported in this paper.

        Speaker: Dr John Smith (General Atomics)
      • 11:00
        Factory acceptance test results of ITER EU ECPS 2h

        The power supply set for the EU EC Heating system (ECPS) of ITER provides up to 6 MVA electrical power to two 170GHz/1MW Gyrotrons. The required electrical power for the gyrotrons is both very high and has to comply also with highest quality requirements. These performance indicators were proven with full voltage modulation at rates up to 5kHz.

        Ampegon’s newly developed power supply topology is optimized to cope with these stringent requirements. Two different topologies are combined. On one hand, the PSM topology for the Main High Voltage Power Supply (MHVPS) and on the other hand the enhanced PSM topology for the two Body Power Supplies (BPS). The enhanced PSM topology demonstrates an improved accuracy and very low ripple values. Furthermore, this topology is designed to supply capacitive loads. Besides the demanding dynamic requirements and ripple performance, the power supplies must protect the Gyrotron in case of an arc. Therefore, the energy into the arc is also an important figure for the qualification of the power supply.

        To test the power supply set (one MHVPS and two BPS) in its various operating modes, four different dummy loads have been designed and are part of the Ampegon scope of supply. The complete set, including the relevant control interfaces, has been installed and tested in Ampegon’s test laboratory.
        The test setup and the factory acceptance test results for the ECPS Heating System are presented.

        The work leading to this publication has been funded by Fusion for Energy under the Contract F4E-OPE-454 for ITER. This publication reflects the views only of the author, and Fusion for Energy and ITER Organization cannot be held responsible for any use which may be made of the information contained therein.

        Speaker: Dr André Spichiger (Ampegon AG)
      • 11:00
        Failure and Melting of Intentionally Misaligned Tungsten Castellated Blocks under High Heat Flux 2h

        High heat load test were performed by using 1) E-beam for tungsten blocks and divertor mock-up, and 2) Long pulse H-mode plasmas in KSTAR for tungsten blocks mounted on stainless steel base.
        Tungsten blocks are exposed to a heat flux of 13 MW/m2 from the top with a beam spot size around 11.5 mm in diameter, 100 kV and 12.5 mA, while the divertor mock-up is exposed to much higher heat flux up to 130 MW/m2. In the case of KSTAR experiments, inter-ELM heat flux at the central divertor is in a range between 0.5-3 MW/m2, while that during ELM is up to 50 MW/m2.
        Two tungsten blocks exposed to 13 MW/m2 e-beam show that the Cu interlayer is melted within 4 sec, while the surfaces of the tungsten layer show recrystallization.
        From COMSOL modeling, the temperature of Cu interlayer reaches its melting point (1100 deg C) in 3.7 sec, which is consistent with the experimental observation: The temperature of tungsten layer is around 2100 deg C well above the recrystallization temperature (1400 deg C). With the extreme case of 130 MW/m2 and longer exposure time, Cu interlayer is completely melted and some part of CuCrZr base is melted down. Tungsten blocks exposed to KSTAR H-mode shows melting of Cu interlayers with closed gaps between two tungsten blocks by molten Cu. Simulation with q=3 MW/m2 shows that an exposure time of 15.8 sec is required to elevate the temperature of Cu inter-layer up to the melting point. These results indicate that bonding between tungsten and Cu layers will be failed at a temperature over 1000 deg C resulting in poor thermal contact. Melting of Cu interlayer also causes further misalignment, which can cause further increase of temperature of the tungsten layers up to melting point of tungsten.

        Speaker: Dr Suk-Ho Hong (DEMO Technology Division, National Fusion Research Institute)
      • 11:00
        Fault analysis and improved design of JET In-vessel Mirnov coils 2h

        In vessel Mirnov coils are an essential diagnostic in present day tokamaks. Their use in ITER and future Fusion reactors presents some disadvantages linked to the high radiation environment. Furthermore large Electro Magnetic forces can be experienced by the coil, due to the pulsed operation of the tokamak device [1], and disruptions [2].
        Since the operation with the ITER-like wall, JET has experienced severe faults in the high-bandwidth Ti wire coils. Surface discoloration of the wire pointed to the formation of chemical surface layers that can produce hardening of Ti, leading to increased mechanical failures. Disintegration of 0.5mm thick mica plates covering slots in the coils casing was also found.
        The tensile strength of the failed wire has been tested, comparing it to new samples of the same composition and manufacturer, and hardening has been observed. Scanning Electron Microscope investigations have been carried out on the rupture surface, which is flat but rough, with no wire deformation, compatible with fatigue failure. Modelling of EM loads showed that the forces, combined with fatigue on hardened Ti are enough to cause failure on sharp wire bends.
        During 2016-17 new coils have been designed and installed. These can be replaced using remote handling, and they use Cu alloy wire, in order to reduce chemical interaction with in-vessel gases and increase heath conduction to the ceramic former.
        The presented work includes the failure analysis and modelling, motivating the design differences between old and new coils. The latter will provide valuable information on the long term effects of EM loads during disruptions, as well as chemical degradation processes that will be encountered for ITER HF coils, which are characterized by the same materials.
        1) Vayakis, George, et al. Journal of Nuclear Materials 417.1 (2011): 780-786.
        2) Gerasimov S.N. et al Nucl. Fusion 55 (2015) 113006

        Speaker: Dr Matteo Baruzzo (Consorzio RFX)
      • 11:00
        Fault analysis and overvoltage estimation in the DEMO Toroidal Field coil circuit 2h

        In the European DEMOnstration nuclear fusion power plant (DEMO), the desired toroidal magnetic field is produced by a magnet system composed of 16 Toroidal Field (TF) coils, according to the last 2017 reference baseline. The total stored energy of about 140 GJ, more than three times that of ITER TF coils, has to be quickly dissipated in case of quench by a suitable Quench Protection (QP) system. The energy, the current and the required discharge time constant define the voltage to be applied to the coils; however, the peak value at the coil terminals during the fast transient phase at the beginning of the discharge or in case of faults can be much higher.
        This paper deals with first studies addressed to estimate maximum voltage stresses for TF coils in various operating conditions and for different TF circuit topologies to evaluate their relative merit and to provide inputs for the definition of the number of toroidal sectors as a compromise between requirements for the coil insulation and cost and size of protection system, busbars and current leads.
        The studies were firstly done with 18 TF coils (2015 reference baseline); since three different winding pack options are under study, we selected the one characterized by the highest inductance, thus by the related highest voltage across the coils at the discharge.
        Numerical simulations were carried out to reproduce the voltage waveforms at the coil terminals, across the coils and to calculate the i2t in the coils for the cases analyzed for the different TF QP circuit topologies, operating conditions, and number of sectors (18 / 9). Then, the analyses have been updated for the case of 16 / 8 sectors of the last reference baseline; all the main results are reported and discussed.

        Speaker: Dr Alberto Maistrello (Consorzio RFX)
      • 11:00
        Feasibility of fusion fuel isotope detection below 1% using Penning gauge optical spectroscopy 2h

        The species-selective (or Optical) Penning gauge approach to the measurement H2/D2/T2 fuel isotopic composition [1] and He/D2 concentration [2] in the neutralized particle exhaust of fusion devices is almost universally used nowadays across all fusion facilities. Although recent studies have shown that, through spectroscopic detection optimization, He/D2 detection is feasible down to at least 0.1% [3], an ongoing overview of existing data (JET) points to an apparent isotopic concentration detection limit of ~1% [4]. A laboratory study suggested that surface interactions with the isotopic species may be playing a role in this apparent limit [5].
        In this paper, results from laboratory study, specifically designed to explore the impact of surface interactions, inside the Penning gauge, on this detectability limit will be presented. The study included baking of the gauge and a series of isotopic exchanges.
        Penning bake-out at ~150°C is found to dramatically reduce the H2 background in D2 only fueling cases. However, some self-cleaning by the Penning plasma discharges, after operating at high neutral pressure, can also be achieved.
        A preliminary wall model for the gauge is used to interpret the findings of the laboratory study. The model presently includes hydrogen isotope retention in the surface, isotopic exchange and desorption from the surface.
        Conclusions also include recommended procedures for optimization of isotopic composition analysis. The immediate application will be for the recently upgraded, divertor gas analysis system on JET [3]. However, this will extend to other current devices, as well as to the ITER Diagnostic (DRGA) [4] .

        [1] D.L Hillis et al., Rev. Sci. Instr. 70, 359 (1999)
        [2] T. Denner, K. H. Finken and G. Mank, Rev. Sci. Instr. 67, 3515 (1996)
        [3] C.C. Klepper et al., Rev. Sci. Instrum. 87 (2016) .
        [4] HTPD-2018/San Diego ]
        [5] C.C. Klepper et al., J. Instrum, Proceedings ECPD-2017

        Speaker: Dr Stephane Vartanian (I.R.F.M Cea Cadarache, CEA Cadarache)
      • 11:00
        Feasibility studies of DEMO potential waste recycling by proven existing industrial-scale processes 2h

        This paper is focused on the analysis of existing industrial-scale process for recycling of DEMO steel components (Eurofer, AISI 316L) and Lithium orthosilicates breeder. The aim is the assessment of their practical feasibility and the individuation of preparatory activities to be performed for facilitating and improving the recycling.
        In detail, the thermodynamic analysis of recovering 14C from Eurofer and AISI 316L steels by decarburization processes was performed, based on practices used in steelmaking. The mass of species in the formed phases: gas, oxide, metal, were quantified. The effect of decarburization process on other critical elements contained in the steels like Cr, Mn, W, Ni and Mo was investigated as well.
        The decarburization of steel is a process, currently used in steelmaking for producing Ultra Low Carbon (ULC) Steel and High Chromium Steel, based on the reaction between Carbon and Oxygen, both dissolved in liquid steel, by forming CO gas that is removed by operating under vacuum and/or by inert gas like Argon or Nitrogen. Another process analyzed is the recovery of the expensive 6Li, from used Li4SiO4 solid breeder pebbles, in order to reuse it in the manufacture of fresh solid breeder and to evaluate the way of removing detrimental radioactive impurities. The most promising candidate process is the melting of pebbles with addition of new 6Li enriched material.
        A general trend in the modern steelmaking industry is toward a higher degree of robotization and remote control; this evolution is in line with the need of processing of potential radioactive wastes taking into account the issue of their decay heat and dose rate. The study performed aims to identify and select existing and emerging steel melting technologies that might be suitable for recycling of activated steels and breeder materials of DEMO.

        Speaker: Luigi Di Pace (Fusion Fusion and Technology for Nuclear Safety and Security Department ENEA)
      • 11:00
        First results from a new tritium capable ion implantation materials facility. 2h

        A new tritium facility to study the interaction of tritium with fusion relevant materials, and its retention and release, has been produced. Tritium retention is a major issue for fusion power devices. The new facility allows implanting of a range of gases into samples, including tritium. This facility is currently used for the UKAEA led Tritium retention in Controlled and Evolving Microstructure (TriCEM) project, which includes modelling work investigating the interaction of hydrogen isotopes with different types of microstructural damage, validated by experiments. The experimental section includes sample production, irradiation, implantation with deuterium or tritium, and characterization. Self-ion bombardment with energies of several MeV is used to mimic the defects created by neutrons in fusion power plant and the created traps are then filled with D/T in the new facility. The samples are analysed primarily using Thermal Desorption Spectroscopy (TDS) and Secondary Ion Mass Spectroscopy (SIMS). Part of the characterization and analysis work is carried out at the new Materials Research Facility (MRF). There is a noted lack of tritium retention experiments that this new tritium facility will make up for. Accurate study of isotope effects, such as the isotopic exchange in damaged microstructure, has previously been difficult due to a background signal of light hydrogen. This new capability will allow virtually background free measurements using tritium and deuterium. The design and build of this facility is described. Commissioning results are presented along with the first results from materials with controlled damaged microstructure.

        Speaker: Anthony Hollingsworth (TESG UKAEA)
      • 11:00
        First thermal-hydraulic and thermal-mechanical analysis of a CO2-cooled solid breeding blanket for the EU-DEMO 2h

        Helium Cooled Pebble Bed (HCPB) Breeding Blanket (BB) has been intensively studied for the EU DEMO. However, several feasibility issues remain for a HCPB-class DEMO reactor, namely the large diameter of the Primary Heat Transfer System pipework, the resulting large coolant inventory and large expansion volumes required after an ex-vessel loss of coolant accident, the limited operational experience with relevant size He-turbomachinery and the large circulating power, among other. Due to the larger density of CO2, the use of this gas as primary coolant for DEMO can lead to key advantageous features, mitigating most of the issues posed for He-cooling and resulting in a higher net efficiency than that of HCPB, as reported in a previous study. Therefore, a CO2-cooled Pebble Bed (CCPB) has been proposed as an alternative coolant to He for the EU-DEMO. After identifying that CO2 will have a negligible influence on the neutronic performance, making the CCPB’s TBR almost equal to the HCPB’s one (TBR ≈ 1.15), a full first set of thermo-hydraulic and thermo-mechanical analyses with the commercial code of ANSYS CFX are reported here. The analyses are based on the newly proposed design of breeding zone (BZ) in the enhanced HCPB fuel-pin concept for the EU-DEMO. Such pin-type fuel elements have been already used in liquid metal fast reactors since the 1960s. The paper will show that, despite the lower heat transfer capability of CO2 with respect to He, the fuel-pin design breeding zone improves the thermo-hydraulic performance, meeting the materials’ temperature requirements. For the thermal-mechanical analysis, the structural behavior under normal operation has been assessed according to the available codes and standards (RCC-MRx). The results show that the CCPB can satisfy the basic thermal and mechanical blanket requirements and that CO2 is a realistic option as primary coolant for gas-cooled fusion reactors.

        Speaker: Shuai Wang (School of Physical Sciences University of Science and Technology of China)
      • 11:00
        First version of the W7-X Fast Interlock System 2h

        The central safety system (cSS) of W7-X consists of two parts. The safety related PLC with its corresponding periphery, such as sensors and actors, fulfills the requirements of occupational safety and ensures basic investment protection. The reaction time from the signalization of dangerous faults to the initiation of protection measures like W7-X emergency stop or media shut-off is in the range of some 100 ms. Because this is not sufficient to protect components in the plasma vessel in case of overloading by plasma heating, a fast interlock system was implemented. In the first operation campaigns of W7-X, with limited energy input, only a local fast interlock of the ECRH (using stray radiation probes) has been applied. Now, the central fast interlock system (cFIS) as part of the cSS and local fast interlock systems (lFIS) in several heating systems have been installed. Together with a set of safety relevant diagnostics, this system is able to react to dangerous situations within the few ms-range.
        The paper describes the system design, hardware components as well as the software implementation in detail. First results, which are obtained during the operation campaign 1.2b in summer 2018, are outlined.

        This work has been carried out within the framework of the EUROfusion Consortium and has received co-funding from the Euratom research and training programme 2014-2018 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission.

        Speaker: Dr Reinhard Vilbrandt (Quality Management, Max-Planck-Institut für Plasmaphysik)
      • 11:00
        Heating and in-vessel upgrades of the TCV tokamak 2h

        Two sets of upgrades are being implemented on the TCV tokamak. The first set involves the installation of neutral beam injection (NBI) and new Electron Cyclotron (EC) power sources, to heat the ions and vary the electron to ion temperature ratio, for ITER relevant β values. A tangential 15-30keV, 1MW, 2s NBI is operational on TCV since 2015. A second 1MW, ~50keV beam, also tangential but opposite to the first beam, is foreseen to approach β limits, vary the momentum input and investigate suprathermal ion physics. On the EC side, two 0.75MW gyrotrons at the 2nd harmonic have been installed. The next step, presently ongoing, entails two 1MW dual-frequency gyrotrons (2nd and 3rd harmonics). The total heating power for high-density plasmas will increase from 1.25MW to 5.25MW. The main element of the second set is an in-vessel structure to form a divertor chamber, reach high neutral density and impurity compression and access reactor relevant regimes for a wide range of divertor configurations. Graphite gas baffles will be installed to separate the main chamber from the divertor. The first set of baffles consists of 32 tiles on the high-field side (HFS) and 64 tiles on the low-field side (LFS). The LFS tiles’ dimensions will be varied to modify the divertor closure. Control of the plasma, neutral and impurity densities, and of the He compression in the divertor will be achieved by a combination of toroidally distributed gas injection valves, impurity seeding, and cryo-condensation pumps. Significant developments will be undertaken in plasma diagnostics, to characterize the divertor plasma, measure power and particle deposition at the strike points, and monitor the detachment process. The possibility of installing dedicated high temperature superconducting coils, to expand the range of divertor configurations and improve their control, will be discussed.

        Speaker: Dr Ambrogio Fasoli (Swiss Plasma Center, EPFL)
      • 11:00
        High resolution scanning electron microscope for sequential testing and analyses of full-size PFC components of AUG 2h

        The complex power and particle wall loading conditions in fusion devices lead to various surface modifications of plasma-facing components (PFCs). To assess the consequences of these modifications on power handling capability and lifetime of PFCs, detailed microscopic studies of the surface and internal structure are required. Essential are analyses of the same area before and after plasma exposure. Non-destructive analyses are mandatory for such sequential testing, i.e., the complete tile as installed in the fusion device must fit into the microscope. Therefore, a scanning electron microscope (SEM) with focused ion beam (FIB) and analytics, energy and wavelength dispersive X-ray spectroscopy (EDS/WDS), was procured and commissioned at Max-Planck-Institut für Plasmaphysik.

        This SEM is equipped with a newly developed heavy-duty stage, which allows to analyse samples up to a mass of 10kg, a length of 44cm, and a height of 10cm without and 6cm with additional rotation module. The accessible area on the sample is 23x10cm2. The achieved imaging resolution is better than 5nm. Cross-sections can be prepared by FIB. A multiple gas injection system enables, e.g., to coat markers for erosion measurements. Elemental mapping by X-ray spectroscopy is possible also on FIB cross-sections. Small features (tens of nanometers) can be investigated by using low electron beam energy (3-5keV).

        In this contribution selected SEM analyses using FIB and EDX/WDX capabilities will be presented from material erosion and deposition experiments on divertor and first wall tiles exposed in ASDEX Upgrade (AUG). The examples include analyses of tiles with controlled pre-exposure damage structures and of W monoblock mock-ups pre-damaged by high heat flux testing after their exposure using the AUG divertor manipulator. Also data from heavy-alloy AUG divertor tiles installed for an entire campaign in AUG will be presented. The potential of the analyses capabilities of this SEM device will be elucidated.

        Speaker: Dr Martin Balden (Max-Planck-Institut für Plasmaphysik)
      • 11:00
        High temperature brazing of tungsten with steel by Cu-based ribbon filler alloys 2h

        The work presents the results of high temperature brazing of tungsten with EK-181 steel by rapidly quenched into ribbon filler alloys based on copper. Compositions of the filler alloys were chosen with consideration to the requirement of reduced activation that is necessary for DEMO reactor. All the joints were manufactured at 1100oC in a vacuum furnace. To analyse microstructure and mechanical characteristics before and after thermocycling tests (in the interval of 700 to 25 oC) SEM investigation, microhardness and shear strength tests were used. It is stated, that the use of Cu-27Ti and Cu-20Sn filler alloys together with vanadium interlayer gives an opportunity to make qualitative joint.

        Speaker: Dr Oleg Sevryukov (Brazing Department, MEPHI)
      • 11:00
        Hydraulic characterization of twin box joints for ITER magnets 2h

        The ITER magnet system will be the largest superconducting magnet system ever built. The system, all inside a cryostat, is mainly composed by a central solenoid (CS) split in 6 modules, a set of 18 toroidal field (TF) D-shaped coils and 6 poloidal field (PF) coils. Each of these coils use variable type of cable-in-conduit-conductors (CICC) actively cooled by supercritical helium forced flow. Their electrical supply from the current feedthrough of the cryostat is done with main busbars (MB) using similar CICC. The electrical MB to coils as well as internal PF and TF coils connections rely on the twin box concept developed by CEA in the early R&D phase. After construction and electrical validation of joint prototypes for the PF and the MB conductors through full size samples, the question of their hydraulic behavior in the operating conditions arises. Two specific hydraulic characterization tasks were done through the Magnet Infrastructure Facility for ITER (MIFI) contract between ITER Organization (IO) and CEA devoted to develop, improve and qualify manufactured components and assembly processes. These support tasks, were done on the full size qualification samples by setting the samples on the CEA OTHELLO dedicated facility able to operate with gaseous N2 in a large Reynolds range at room temperature. The paper explains the way followed to get a full hydraulic characterization of the MB and PF5 half joint boxes. The pressure drop for the two flow directions was determined for both joints. The study of the flow distribution between parallel cooling channels inside the PF5 joint revealed a bypass of the active joint region. The paper reports on this hydraulic behavior in the relevant magnets operating conditions and outlines the design changes in the joints provoked by the results of this study.

        Speaker: Dr Sylvain Bremond (IRFM, CEA)
      • 11:00
        Hydrogen isotope permeation experiment - design and first results 2h

        Following the decommissioning of JET, and other future fusion reactors, there will be large amounts of tritiated waste requiring disposal. An appropriate containment strategy is required for storage of this waste. Studies have so far demonstrated that stainless steel appears to be the most promising containment material, but little is known about the permeation of hydrogen isotopes through stainless steel in waste relevant conditions. It is essential that this process is understood prior to the decommissioning of JET and start-up of ITER and DEMO.
        An experiment has been designed to replicate waste storage relevant conditions and allow for the study of hydrogen isotope permeation under these conditions. The effect of temperature, humidity and surface uniformity of the steel are being studied for protium, deuterium and tritium.
        This unique experimental rig has been built at UKAEA, to support JET decommissioning research funded by the UK Nuclear Decommissioning Authority, and is now carrying out testing with deuterium. The first results have been extremely surprising, with deuterium permeation significantly lower than expected for unbaked stainless-steel samples. This phenomenon is being further investigated to determine the cause, but is currently thought to be due to the protium content of the steel preventing diffusion of other hydrogen isotopes. This theory is supported by further experimental results indicating that baking the samples at 400°C to remove protium prior to commencing experimental work increases the permeation dramatically, to be in line with computational models. X-ray photoelectron spectroscopy and thermal desorption spectroscopy are being conducted to further understand the nature of the samples and rule out surface contamination or other unexpected effects.
        The outcomes of this work are entirely unexpected; no published literature appears to have investigated this phenomenon. The results are expected to have wide-ranging consequences for the storage and containment of tritium and tritiated waste worldwide.

        Speaker: Rachel Lawless (Tritium Engineering and Science Group UKAEA)
      • 11:00
        Hydrogen isotopes distribution modeling by "FC-FNS" code in fuel systems of fusion neutron source DEMO-FNS 2h

        Tokamak-based fusion neutron source (FNS) [Kuteev B.V. et al 2010 Plasma Phys. Rep. 36 281, Kuteev B.V.et al Nucl. Fusion 55 (2015) 073035] is the centerpiece of the fusion-fission hybrid reactor (combining nuclear and thermonuclear technologies). In Russia, for the demonstration of stationary and hybrid technologies, the DEMO-FNS project has been developed, which should operate at least 5000 hours per year. The plasma operation in a tokamak requires the continuous injection of the fuel mixture containing hydrogen isotopes (deuterium and tritium) into the vacuum chamber, as well as its subsequent pumping out and processing. Calculation of the distribution of tritium (as well as deuterium and protium) in fuel systems is important for assessing safety features of the facility and for designing these systems. To simulate hydrogen isotope flows and inventories in the fuel systems of FNS, computer code FC-FNS [Ananyev S.S. et al Fusion Eng. Des. (2016), Volumes 109–111, Part A, pp 57–60] has been created that continues to be developed. This report describes capabilities of the code. The results of calculations are presented for the conceptual design of DEMO-FNS. The balance of three hydrogen isotopes is taken into account, the performance of deprotiation systems is calculated to maintain the required level of protium in the plasma tokamak and detritiation (for the variant of the neutral injection system - NBI). Three alternative scenarios for supplying gas to NBI system are simulated. It is shown that the proposed approach to NBI fueling allows reducing the total amount of tritium in FS up to 1.5 times, that leads to the initial loading for DEMO-FNS of 460 g. The time for tritium breeding up to the amount sufficient for starting a new similar device will be 2.5 - 4 years (for different scenarios for FS NBI).

        Speaker: Dr Sergey Ananyev (NRC Kurchatov institute)
      • 11:00
        Implementation of Laser-induced Fluorescence Diagnostics in ITER 2h

        A laser-induced fluorescence (LIF) diagnostic has been designed for measuring helium density (nHe) and ion temperature (Ti) in the outer leg of the ITER divertor. The LIF diagnostic is integrated with the divertor Thomson scattering (DTS) diagnostics via common injection and collection optics. Optimisation of previously proposed spectroscopic schemes, and lasers suitable for nHe and Ti measurements are the focus of this report.
        The LIF method is based on laser pumping of transitions among excited states of atoms / ions and subsequent fluorescent signal processing. Optical parametric oscillators (OPO) or dye lasers with 5-10 ns pulse duration and repetition rate of ~ 1 kHz are suitable for nHe measurements. A set of spectroscopic schemes for He I detection requires the following wavelengths of laser radiation: 388.9, 501.7, 587.6 and / or 667.8 nm. Spectral coverage of OPO and dye lasers combined with currently available power spectral densities allow to achieve a saturation threshold in all the spectroscopic schemes under consideration.
        Using time-modulated lasers (P ~ 20-50 W) for the Ti measurement is discussed in spectroscopic schemes with quenching excited states via pumping to higher excited levels. A tunable ytterbium fiber laser with 1012.3 nm wavelength can be used to quench the 468.6 nm line of He II. Spectral line scanning gives Ti from the absorption line contour Doppler broadening.
        Optimization of the spectroscopic schemes and estimation of the required laser parameters have been performed using dynamic collisional-radiative models (CRMs). The schemes chosen were selected via comparing differences in measurement accuracy. The laser power spectral density has been estimated by calculating saturation threshold of the selected transitions in the ITER divertor plasma.

        Speaker: Dr Alexey Gorbunov (NRC Kurchatov Insitute)
      • 11:00
        Implementing DevOps practices at the control and data acquisition system of an experimental fusion device 2h

        The stellarator Wendelstein 7-X (W7-X) is a fusion device designed for steady state operation. It is a
        complex technical system. To cope with the complexity a modular, component-based control and
        data acquisition system has been developed.
        During operation phases of W7-X components steadily evolve. For instance measurement devices for
        diagnostics get improved, technical processes are optimized, experienced limits of the machine have
        to be taken into account or simple “bug fixing” is done. This requires continuous further
        development of the components while operating them at W7-X – a typical use case for DevOps
        practices.
        DevOps is a software engineering practice. The term is a compound of development and operations.
        It aims at shorter development cycles, while the quality of the changed system must stay at a high
        level. This is achieved by using a highly automated tool chain.
        DevOps at W7-X does not (only) mean deploying new software packages. It also comprises setting up
        new configurations, changing rules and constraints, improving experiment planning views etc. of
        components, based on new scientific and technical experiences. One crucial step of the DevOps tool
        chain at W7-X is the release and reconcile process. It is a well defined and automated process that
        commits a change in the components configuration (Release) to the control and data acquisition
        system with minimal impact. During the process all existing experiment programs are adapted to
        this change (Reconcile). For newly added configuration parameters default values are added. Thus it
        is guaranteed that experiment programs are still executable at W7-X. The quality is ensured by
        preceding unit and integration test, configuration consistency and version checks.

        Speaker: Dr Marc Lewerentz (Operations, Max-Planck-Institute for Plasmaphysics)
      • 11:00
        Improving a Negative Ion Accelerator for next generation of Neutral Beam Injectors: results of QST-Consorzio RFX collaborative experiments 2h

        In large Neutral Beam Injectors for fusion applications, the efficiency of ion beam neutralization and transport to the tokamak plasma strongly depends on the divergence and the deflection angle of each single beamlet with respect to its ideal trajectory. In fact, a very narrow window is available for the particle beam to pass through the neutralizer panels and the duct reaching the tokamak plasma. For this reason, beam optics quality is one of the key requirements in multi-stage multi-beamlet negative ion accelerator, such as the high power Heating Neutral Beam injectors for ITER and JT-60SA.
        In the framework of the collaboration established between Consorzio RFX (Padova, Italy) and QST (Naka, Japan) experimental campaigns have been organized on the Negative Ion Test Stand (NITS) in Naka employing an ITER-like multi-beamlet configuration [1].
        These campaigns were intended to:
        • Test and optimize new solutions for cancelling the undesired ion deflection caused by the transverse magnetic field required for suppressing the co-extracted electrons [2].
        • Better understand the physics underlying the extraction of single beamlets in the initial part of their trajectories, close to the "meniscus" region.
        • Validate and improve the numerical codes used for the accelerator design
        Although it was not possible to explore all the intended parameter space due to power supply limitations, the experiments carried out in 2016 and 2017 allowed to successfully achieve the first most important (engineering) objective and to provide bases for the achievement of the other two (physics) objectives, which will be completed during new already planned activities and collaborations.

        References
        [1] H.P.L. de Esch et al. “Physics design of the HNB accelerator for ITER" Nucl. Fusion 55 (2015).
        [2] D. Aprile, et al. “Realization of a magnetically compensated extraction grid for performance improvement of next generation NBI” , FED 123 (2017) 400–405 .

        Speaker: Sylvestre Denizeau (Consorzio RFX (CNR ENEA INFN University of Padova Acciaierie Venete SpA))
      • 11:00
        In-box LOCA accident analysis for the European DEMO water-cooled reactor 2h

        Transient analysis in a water-cooled fusion DEMO reactor have been performed to support the WCLL (Water-Cooled Lithium Lead) breeding blanket design. In this framework, the Design Basis Accident analysis of an in-box LOCA has been carried out.
        The WCLL breeding blanket concept relies on Lithium Lead (LiPb) as breeder, neutron multiplier and tritium carrier, which is cooled by water at 15.5 MPa with an inlet temperature of 295°C and an outlet temperature of 328°C.
        Water flows in Double-Wall Tubes (DWTs) in order to reduce the probability of water/LiPb chemical interaction. In the case of a LOCA accident, multiple rupture of these tubes is postulated, with consequent leakage of pressurized water in the LiPb side of the module.
        The present safety analysis has been performed with the MELCOR computer code (ver. 1.8.6) modified for the application in the fusion context. Custom models are employed to simulate the chemical water/LiPb interaction in the module. The rupture mass flow rate calculated in water simulation is transformed in its equivalent in terms of hydrogen and unreacted water steam. Both have been treated as non-condensable gas. Two different input decks, one for each fluid considered, have been coupled through an external interface to account for their reciprocal interaction.
        Pressure and temperature transient behavior in the broken module demonstrate that safety margins are respected during the entire accidental sequence, even though no external safety system is foreseen or actuated. Moreover, particular attention has paid to the quantity of hydrogen produced in order to support the development of solutions, suitable for the DEMO, to prevent hydrogen explosion.

        Speaker: Matteo D'Onorio (Dep. of Astronautical Electrical and Energy Engineering (DIAEE) La Sapienza University of Rome)
      • 11:00
        Influences of fabrication conditions on hydrogen isotope retention in W coatings 2h

        Low pressure plasma spraying (LPS) and spark plasma sintering (SPS) are attractive techniques to prepare W armor layers on substrate materials. The properties of LPS-W and SPS-W depend on fabrication conditions. In this study, LPS-W and SPS-W layers were prepared on graphite and carbon fiber reinforced carbon composite (CFC) substrates at different temperatures, and D retention after plasma exposure was examined.

        LPS-W layers were prepared on graphite tiles (IG-430U) at 1073–1353 K or 1233–1453 K. The coatings were subjected to heat treatment in vacuum or mechanical polishing to remove fine W oxide grains formed by fume condensation. SPS-W layers were prepared on CFC tiles (CX-2002U) at 1773, 1873 and 1973 K. The coating thickness was 1 mm. After removing the substrates by mechanical polishing, the coatings were exposed to D plasma at 373 K in a linear plasma device. The flux and fluence were 5×10^21 m^-2 s^-1 and 2×10^25 m^-2, respectively. The incident ion energy was 80 eV. Contents of D and H were measured using thermal desorption spectrometry at 0.5 K/s.

        For all types of coatings, the D retention was in a range of 6–35×10^19 D m^-2. The main desorption peaks appeared at 500–700 K. Fine W oxide grains on LPS-W were successfully removed by heat treatment and polishing. The D retention in the polished samples was slightly higher than that in non-polished ones. The coatings prepared at 1073–1353 K showed smaller D retention than those formed at 1233–1453 K, though columnar grains were more developed in the latter. The D retention in SPS-W was insensitive to the preparation temperature and slightly smaller than that in LPS-W, though the concentration of impurity H was higher. The correlation of D retention with microstructure and surface morphology will be discussed.

        Speaker: Dr Kota Higuchi (University of Toyama)
      • 11:00
        Innovative and emerging melting technologies for fusion power plants wastes recycling 2h

        To fully maintain the promise of energy production in clean, safe and environmentally responsible way, the nuclear fusion technology must include specific recycling and clearance techniques.
        A number of options have been already proposed and investigated to recovery valuable elements, to separate radioactive species, to re-use materials in re-fabrication of components.
        Melting based techniques, where materials after use are treated at high temperature and melted, are considered promising candidates, because the melt can be cast in a more compact volume for easier handling, disposal or for producing new components. The melt can be treated in appropriate conditions in order to remove undesired elements or to add species to recovery some specific characteristics, for further re-fabrication.
        The most investigated melting options have been derived from well consolidated processes and plants widely applied in industrial sectors for mass production, such as steelmaking, glass, ceramic, foundry industries. The idea is to transfer these technologies taking advantage for the treatment and recycling of nuclear wastes. However the specific cases of radio-activated wastes expected from fusion power plant require significant and accurate re-design of technological solutions and operating conditions to make effective these options.
        This paper presents and discusses examples of innovative technologies that could be fruitfully applied in melting processes for fusion power plant wastes treatment. The presented panorama includes technologies already applied at industrial level, in highly specialized, niche sectors, such as skull crucible technique, as well as laboratory and conceptual technology, such as magnetic levitation, still under development, but probably available when fusion power plant will be in service.
        The aim of the paper is twofold: to enlarge the spectrum of melting technologies candidate for the needs of fusion power plant waste treatments: to help the individuation of necessary adaptation and improvement of these technologies for the specific case of radio-activated wastes.

        Speaker: Dr Teresa Beone (RINA)
      • 11:00
        Integrated core-SOL-divertor modelling for DTT tokamak with liquid metal divertor targets 2h

        The behaviour of the SOL plasma of the Italian projected DTT is analysed for the standard divertor configuration by means of the integrated COREDIV code simulations when either Lithium or Tin are used as liquid target materials.
        The DTT tokamak is expected to operate in H-mode, which requires the value of power to scrape-off layer above the L-H threshold. On the other hand it is postulated that the divertor power load should be controlled to not exceed 5MW/m2. Cooling of the plasma can be achieved by seeding or by intrinsic impurities like particles released from the divertor. In the case of liquid metal divertor, vaporization additionally enhances the plate material flux into the bulk.
        This paper analyses possible operational space for DTT device with liquid Sn or Li divertor setup. The impurities originating from the sputtering and vaporization processes are expected to modify plasma characteristics significantly both in the bulk and in the scrape-off layer. Therefore, simulations are performed with COREDIV code which self-consistently solves radial 1D energy and particle transport equations of plasma and impurities in the core region and 2D multifluid transport in the SOL.
        Density and power scans are carried out for different target arrangements, in terms of the coolant temperature and thickness of W substrate. First scenarios with only intrinsic impurity are investigated. Tin appears more promising of lithium in terms of radiative capacity, of wider ranges of applicability both of density and input power and of plasma purity. No clear detachment is observed for either Sn or Li except at very high density. For both solutions regime where evaporation overcomes sputtering is more effective in dissipating the input power, provided that is kept low enough to ensure plasma stability. In this case a sort of vapour shielding seems to develop attached to the impurity source.

        Speaker: Dr Roman Zagorski (Institute of Plasma Physics and Laser Microfusion)
      • 11:00
        Integrated Power Exhaust Modelling for DEMO with Lithium Divertor 2m

        Operation of a future demonstration fusion reactor (DEMO) requires the handling of a significant power flux that crosses the separatrix and enters the scrape-off layer. A considerable amount of energy has to be dissipated before the heat flux reaches divertor plates. The divertor may be exposed to high heat fluxes causing high temperature gradients and material fatigue. Such challenging conditions in the scrape-off layer demands developing solid state plate protection scenarios. In the liquid metal divertor, which is one of the considered solutions to the power exhaust problem, a liquid surface protects the solid plates against high heat loads, allowing for longer operation without the need of the divertor disassembly.
        In this paper a DEMO reactor is considered with a liquid lithium divertor setup. The simulation is performed with the COREDIV code which self-consistently solves radial 1D energy and particle transport equations of plasma and impurities in the core region and 2D multifluid transport in the SOL. Influence of sputtering, prompt redeposition and evaporation of lithium is taken into account. An operational space of parameters is analyzed. Two regimes of operation are identified. The sputtering regime occurs when the divertor is tilted significantly with respect to the magnetic surface. Evaporation is low compared to the sputtering. In such a situation divertor power load is high and additional seeding is required in order to dissipate energy before it reaches the divertor plates. In the evaporation regime, evaporation is higher than sputtering. A high amount of lithium is released into the plasma diluting it and therefore reducing the power to the plates. Although in both regimes lithium dilutes the core plasma reducing fusion power, it is also fully stripped what results in low radiation and high power across the separatrix. Cooling of the plasma can then be achieved by seeding additional impurity.

        Speaker: Dr Michal Poradzinski (IPPLM)
      • 11:00
        Interface and requirements analysis on the DEMO Heating and Current Drive system using systems engineering methodologies 2h

        In this paper we present the analysis of System Requirements and Interfaces of the Heating and Current Drive (HCD) system of the Demonstration Fusion Power Reactor DEMO.
        The work was performed applying Model-Based Systems Engineering (MBSE) refining the HCD System Architecture for assessing the system functions, its interdependencies and its overall integration into DEMO. Two concepts for DEMO have been considered: a conservative design (DEMO1) and a more advanced one (Flexi-DEMO). The effort has been undertaken in the frame of the Work Package Heating and Current Drive, supported by the Work Package Plant Level System Engineering, Design Integration and Physics Integration.
        The scope of the work is, on the one hand, to address the identification and definition of the interfaces occurring, both internally in the HCD system, and between the HCD system and neighboring systems. On the other hand, the impact of requirements coming from the ongoing physics studies has been assessed.
        The rationale is to provide the technical foreground for supporting the decision-making processes related to the HCD system which is planned to be carried out during the Conceptual Design Phase, currently foreseen until 2027. The results we show in this paper are part of the design and integration activities consisting of both systems engineering methodologies and design analysis, all aiming at ensuring consistency in the overall EU DEMO plant design. The set of processes we report here is in line with the common approach established within PPPT and in accordance with the ISO Systems Engineering standard (ISO 15288).

        This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the EURATOM research and training programme 2014-2018 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission.

        Speaker: Dr Giovanni Grossetti (Karlsruhe Institute of Technology (KIT))
      • 11:00
        Interference fit process development for the ITER vacuum vessel gravity support mock-up fabrication 2h

        The ITER Vaccum Vessel (VV) is supported by the nine VV gravity supports (VVGS) located on the cryostat toroidal pedestal. The VVGS is dual hinge type that fastened by dowel on the hinge-block hole. The primary hinge restrains a vertical and toroidal movement of the VV system against fast displacements by the seismic events or fast transients. The secondary hinge restrains steady vertical load. However, the hinges allow radial thermal expansion during temperature increase for operation (100ºC) and baking (200ºC). This paper presents the technical approach and result of interference fitting process of the sleeves and MoS2 coated dowel to the full-scaled VVGS mock-up. Since the sleeve and hinge-block hole have tens of micrometers tolerance and around two meter long length, shrink fit method has been selected for the interference fitting of sleeves. To secure a sufficient time for process, liquid nitrogen was charged to the handling fixture capped sleeve hole’s cavity. As a result, the required contraction time was secured as hundreds of seconds. Since the moisture which could be released from the VVGS in the vacuum evironment might be affect the operation of the VV or cryostat, defrosting treatment of sleeve is required. A local protection system was selected with charged nitrogen injection nozzles. As a result, the defrosted surface of shrinked sleeve was maintained during fitting process. A selected MoS2 coating solution was confirmed to satisfy the lubricating ability and durability of technical specification through the pin-on-disk test. The central heating bar was used to insure uniform thermal expansion of the sleeve hole. As a result, MoS2 coated dowel successfully inserted into the sleeve hole without surface contact. Consequently, the interference fitting was successfully done for VVGS mock-up fabrication, and the technical solutions will be applied to the VVGS manufacturing.

        Speaker: Dr Jason Cheon (ITER Korea(KODA), National Fusion Research Institute)
      • 11:00
        Inverse heat flux evaluation of STRIKE data by neural networks 2h

        The instrumented calorimeter STRIKE (Short-Time Retractable Instrumented Kalorimeter Experiment) has been designed with the main purpose of characterizing the SPIDER negative ion beam in terms of beam uniformity and divergence during short pulse operations. STRIKE is made of 16 1D Carbon Fibre Composite (CFC) tiles, intercepting the whole beam and observed on the rear side by infrared (IR) cameras.
        The front observation presents some drawbacks due to optically emitting layer caused by the excited gas between the beam source and the calorimeter, and the material sublimated from the calorimeter surfaces due to the heating itself. It is then necessary to solve an inverse non-linear problem to determine the energy flux profile impinging on the calorimeter, from the 2D temperature pattern measured on the rear side of the tiles. Most of the conventional methods used to solve this inverse problem are unbearably time consuming, so a ready-to-go instrument to determine the beam condition, while operating STRIKE, is mandatory. In this work, the inverse problem, both in stationary and non-stationary conditions, is faced by using a Neural Network (NN) model, pursuing two different approaches. In the first one, the NN is trained to directly solve the inverse problem, by associating the radiation profile (target) to the measured temperature profile (input). In the second approach, a NN is trained to solve the direct problem, where the input is the radiation profile and the target is the temperature profile. Then, the NN is inverted by determining the input corresponding to a fixed target. Preliminary results show the reliability of the proposed method for STRIKE real time operation.

        Speaker: Dr Rita Sabrina Delogu (Consorzio RFX)
      • 11:00
        Investigation of divertor movement during disruptions 2h

        The divertor, being the main power exhaust of a tokamak, is exposed to high heat
        fluxes and therefore must be precisely aligned to prevent leading edges. Since the
        transition from carbon to tungsten tiles in ASDEX Upgrade it was found that a specific
        assembly in the divertor was misaligned up to 1.5 mm after the experimental
        campaigns. This lead to heat spikes on the edges of several tiles and subsequent
        melting. To understand the origin of this movement, numeric model was created in
        ANSYS Maxwell containing the full 3D coil setup of ASDEX Upgrade, the plasma
        current and a segment of the divertor assembly. The plasma current (1.6 MA) was set
        to decay within 2.5 ms to 10 ms in order to create a poloidal field change of 50 T/s to
        200 T/s while the toroidal field was constant at 2.4 T in the divertor region. The
        simulation revealed a parasitic current flowing from the support structure through the
        outer tiles. The resulting jxB force can reach up to 2500 N and the torque up to 780
        Nm on a single tile depending on the poloidal field change. For carbon tiles, the
        maximal force and torque are five times smaller due to the higher electric resistivity. A
        current flowing through the conducting support of the assembly through the vessel wall
        and back through the assembly causes additional force (max. 5500 N) and torque
        (max. 2300 Nm). The other support is isolated by SiN plates. A friction test showed a
        static friction coefficient of 0.1 under normal forces larger than 5 kN. A FEM model,
        using these results, showed that the friction force at the SiN plates is overcome and
        the assembly is moved.
        The models and detailed results of the calculations will be presented at the
        conference.

        Speaker: Dr Mathias Dibon (Max-Planck-Institute for Plasma Physics)
      • 11:00
        Investigation of the thermal expansion of lithium orthosilicate 2h

        Within the EU, the current grade of advanced ceramic breeder pebbles is composed of a mixture of Li4SiO4 (LOS) and Li2TiO3 (LMT). These pebbles are fabricated at KIT by the melt-based process “KALOS”. The addition of LMT is beneficial for two aspects: the mechanical strength of the pebbles is considerably increased and the long-term stability at high temperatures is improved. Nevertheless, the advanced ceramic breeder pebbles generally show a number of defects that decrease the ideally achievable mechanical strength. In this work, high-temperature X-ray diffraction (HT-XRD) experiments will show that the well-known displacive phase transformation of LOS at temperatures between 665 °C and 723 °C is strongly anisotropic. The formation and growth of the defects of the pebbles may thus not be exclusively attributed to thermal stresses during the fabrication of the pebbles but also to this phase transformation. To address this issue, principally two ways exist. Either the intensity of the phase transformation is attenuated, or the phase transformation is shifted to unobjectionably low temperatures. In this study both ways as well as their combination are demonstrated. It will be shown that the solid solution of germanium in the LOS lattice is an effective way of attenuating the phase transformation while the partial substitution of lithium by magnesium shifts the phase transformation to lower temperatures. Furthermore, it is also demonstrated that by adding magnesium the thermal expansion coefficient of LOS can be aligned with that of LMT, so that intrinsic stresses are reduced as additional benefit. HT-XRD, dilatometry as well as differential scanning calorimetry are used for the analysis of different compositions to evaluate the individual effects of the added elements. Eventually, the impact on the pebble strength is demonstrated for selected promising compositions.

        Speaker: Dr Matthias Kolb (Karlsruhe Institute of Technology)
      • 11:00
        Isotope separation systems for a european DEMO 2h

        Large-scale isotope separation in a DEMO tritium plant poses significant challenges. Alternatives to distillation and palladium-based adsorption (used in the tritium fuel cycle for JET) remain elusive, despite the disadvantages: Cryodistillation is energy intensive and lacks inherent safety due to the high tritium inventories in the liquid phase, that inevitably expand to vapour in the event of a cooling failure. Palladium-based packings are expensive, and the systems are limited in scale by the heat transfer required to liberate hydrogen from the metal hydride. This presentation will outline the approach taken to tackle these challenges by the European DEMO Tritium Matter Injection and Vacuum work package, and highlight the most interesting results.

        Firstly, a candidate list of potential technologies was generated. The isotope separation requirements created by each system block in the fuel cycle was considered, and potential mass balances for different scenarios defined. This work has identified approaches to the plasma fuelling requirements which could reduce isotope separation requirements, and opportunities to integrate isotope separation requirements within the fuel cycle.

        The combination of requirements and mass balances has then been used to assess each technology option qualitatively and, where possible, quantitatively. This has led to surprising results in the size of systems required and the feasibility of both older and novel separation technologies for two distinct sets of isotope separation requirements: isotope rebalancing and protium removal for the inner fuel cycle, and trace tritium recovery for the outer fuel cycle. Studies have established that cryodistillation and temperature swing adsorption type processes (such as TCAP) are feasible within the inner fuel cycle, while gas chromatography is less attractive. Work is on-going to assess thermal diffusion, membrane separations, and pressure swing adsorption, and the suitability of technologies for trace tritium recovery in the outer fuel cycle.

        Speaker: Tamsin Jackson (Tritium Engineering and Science Group UK AEA)
      • 11:00
        ITER magnetic sensor platform engineering analyses 2h

        ITER in-vessel magnetic sensors play a key role for ITER plasma operation. Each of these sensors is accommodated in a platform mounted on the inner surface of ITER vacuum vessel and behind the blanket.
        A full set of engineering analysis has been performed on the platform to assess the feasibility of the design configuration.
        Electromagnetic (EM) Sub-Modelling technique has been used for very detailed evaluation of the EM loads die to plasma disruption events.
        The EM sub-model has been built by taking one sample sensor with a portion of vacuum vessel around it and applying the vector potential obtained from ITER Global Model (IGM) analysis on the boundary nodes.
        With this technique, platforms at 23 locations from a full poloidal array have been analyzed under 8 category III plasma disruption events. The two cases yielding highest EM load were selected to complete the structural assessment.
        Thermal analysis was performed for a platform located on outboard midplane, where maximum heat load is expected.
        The structural assessment has been performed as for the RCC-MR design code and considering a pessimistic input load combination of highest EM loads and highest thermal stresses over ITER operation lifetime.
        An optimization process, dealing with EM and structural analyses, was set up, to reduce the EM loads and increase the structural strength of the platform.
        The engineering analysis showed positive results on the structural integrity of the platform structure with the optimized design.
        Future work is planned to incorporate the thermal deformation of VV surface due to nuclear heating and further refine the calculation.

        Speaker: Dr Anna Marin (L.T.CALCOLI S.R.L.)
      • 11:00
        ITER Upper Visible/Infrared Wide Angle Viewing System: I&C design and prototyping status 2h

        The ITER Upper Visible/Infrared Wide Angle Viewing System diagnostic will provide key measurements for machine protection and plasma control. The system, installed in five upper port plugs, will monitor the ITER divertor but also part of the ITER first wall using both high definition infrared and visible cameras typically running at 100 Hz. The plant system I&C will process about 40 Gb/s of imaging data acquired by the cameras installed in the port-cells up to 200 m away. Then surface temperature of plasma facing components will be computed and abnormal event such as hot spots or ELMs will be detected using dedicated algorithms implemented on fast controllers. Relevant data will be sent to plasma control system and to ITER archiving system for further offline analysis.
        This papers focus on the I&C design for camera control, data acquisition electronics and analysis software. The current progress on I&C design following the ITER methodology for plant system I&C design is presented, including operational procedures, use cases, user and functional requirement tracing, functional analysis and plant I&C architecture. All design documents have been prepared using the Enterprise Architect software with dedicated plug-ins provided by the IO.
        In parallel, prototyping activities related to real-time image processing and high throughput camera data acquisition using a 10 GigE interface have been carried out for the preliminary design review. The goal is to demonstrate some key functions of the system using state-of-the-art techniques adapted to the specific context and needs of the diagnostic. Emphasis is put on current challenges related to real-time computation and archiving of large amount of data.

        Speaker: Dr Sergio Esquembri (Department of Telematic and Electronic Engineering, Universidad Politécnica de Madrid)
      • 11:00
        Joints for cable-in-conduit conductors 2h

        A review of the joints for the ITER CS CICC is given. More detailed discussion of the design and performance of the ITER CS joints is presented including successes and the revealed problems. ITER CS has three types of joints: 1) sintered joints to connect conductor lengths in the CS module; 2) coaxial joints to connect the CS module terminations to the superconducting buses; 3) twin box joints, which are bi-metal copper -stainless steel boxes containing the compacted cables that connect the buses to the ITER feeders. All the ITER CS joints have different requirements for space allocation and ability to be assembled/disassembled. Thus, different designs of the joints are used. The CS joints went through the R&D and qualification effort, but recent verification results showed a necessity of making the design more robust by revisiting some issues that seemed to be established to satisfaction in the qualification phase. We discuss requirements for the strands treatment, quality of interfaces, heat treatment conditions and other technological issues. We also discuss the effect of oxidation of the cable to copper surface interface, effect of the joint compaction, Cr removal and soldering parameters on performance of the joint. Recent results of the ITER CS joints performance and autopsy studies are presented.

        Speaker: Dr Nicolai Martovetsky (US ITER, LLNL/ORNL)
      • 11:00
        Kinetics of double strand breaks of DNA in tritiated water evaluated using single molecule observation method 2h

        Detailed understanding of mechanisms underlying DNA damages by low energy beta-rays from tritium is important for evaluation of impact of tritium release from fusion devices to the environment. In this study, the rate of double strand breaks (DSBs) of DNA in tritiated water was measured using a single molecule observation method.

        Genome size linear double strand DNA molecules of bacteriophage T4 GT7 (57 micrometers, 166 kbp = kilo base pairs) were provided by Nippon Gene Ltd., Japan. The DNA molecules were suspended in sterilized and non-sterilized tritiated water (5.2 MBq/cm3) and irradiated with beta-rays at 10 ℃ for 1, 7 and 14 days. The values of absorption dose were 0.41, 2.9 and 5.7 Gy. After staining with fluorescent dye YoYo-1, DNA molecules were placed on a glass substrate in extended forms. The lengths of DNA molecules were measured using an inverted type fluorescent microscope (Olympus IX73). The number of DSBs were evaluated using a method proposed in [1]. DSBs in non-radioactive sterilized and non-sterilized water were also examined for comparison.

        Under sterilized conditions, significant reduction in DNA lengths due to beta-ray irradiation was observed after irradiation for 7 and 14 days corresponding to 2.9 and 5.7 Gy, respectively. The number of DSBs per 100 kbp was evaluated to be 0.2 at 1 day, 1.0 at 7 days and 1.6 at 14 days. Interestingly, DNA lengths in non-sterilized tritiated and non-radioactive water were clearly shorter than those in sterilized tritiated water. It means that the irradiation effects of beta-rays were far weaker than the influence of microorganisms in water though tritium concentration was as high as 5.2 MBq/cm3.

        [1] S. F. Shimobayashi et al., J. Chem. Phys. 138, 174907 (2013).

        Speaker: Dr Yuji Hatano (Hydrogen Isotope Research Center University of Toyama)
      • 11:00
        KSTAR Tokamak Visible Image sequence classification with Long-term Recurrent Convolutional Networks 2h

        This paper represents the tokamak in-vessel image sequence classification method that used to automatically infer plasma status. Fast framing standard CCD cameras are installed on KSTAR (Korea Superconducting Tokamak Advanced Research) to monitor plasma shape, plasma motion and plasma status. The images generated by the CCD cameras were used for plasma start-up studies and plasma disruption studies. In KSTAR, the images generated by the CCD camera are visually confirmed after the experiment or analyzed manually by the researcher. We introduced long-term recurrent convolutional networks to automatically infer the plasma status from the image. To obtain the description about the image from cameras, we applied Convolutional neural networks(CNNs). To learn a description of image sequences, description from CNN are used as input to Long Short-Term Memory(LSTM) modules. By applying LSTM, we can learn temporal information from variable-length input image sequence. Our results show this model can effectively learn tokamak in-vessel image sequence and infer status of plasma.

        Speaker: Dr Giil Kwon (Control Team, National Fusion Research Institute)
      • 11:00
        Laser Technology for Resident Tritium In-situ Measurement in CEPT Device 2h

        Due to the fact that during Tokamak operation,Plasma Facing Materials(PFM)are able to trap part of the fuel(particularly Tritium), these resident fuel have to be measured and removed. LIBS(Laser induced breakdown spectroscopy) and LIDS (Laser induced desorption spectrometry)are two of the most promising techniques to solve these issues which allow to achieve an on-line and ultra-sensitive measurement of the fuel retention in vessel.
        In this work, LIBS and LIDS were integrated on to a Tritium penetrating and resident assessing instrument named CEPT(Comprehensive ECR Plasma for Tritium) for in-situ tritium retention analysis. An Nd:YAG laser, operating at 1064 nm, with pulse duration of 4-7 ns was used to interact with PFM simulant sample. The spectra emitted from plasma plumes was recorded by the Aryelle 200 spectrometer(/=12500) in the spectral range between 265 nm and 665 nm with an Andor iStar 334 ICCD camera. The mass spectra of particles escaped from the sample due to the laser radiation were obtained using a quadrupole mass analyser (QMA) at range of 1-100 amu.
        A verification experiment was carried out at a pressure range from 1 bar to ultra-high vacuum(10-5 Pa) in He and air for LIBS and range from 10-2-10-5 Pa for LIDS. The LIBS results show that the atomic spectral lines of H, D, W was achieved when a delay time of 0.3 us was installed, in the meanwhile, the released hydrogen isotopes were monitored by the QMA. This technique achieves good spatial resolution without any sample preparation. For comparison, the samples were also analysed by other conventional methods.

        Speaker: Rui-Zhu Yang (Institute of Materials China Academy of Engineering Physics)
      • 11:00
        Layered W-WC composites prepared by FAST 2h

        One of the main difficulties of designing fusion reactor is the development of plasma-facing materials that have to be resilient to the proximity of plasma. Pure tungsten is a primary candidate for this material but has to be strengthened either with particles or fibers to improve its’ brittleness at moderate temperatures and inhibit recrystallization as well as grain growth at higher ones.
        To limit the W grain growth, we proposed the incorporation of tungsten carbide particles in tungsten matrix. Samples were prepared from W and WC powder mixtures and consolidated with field assisted sintering technique (FAST). Layered W-WC composites with the gradually increased content of carbide phase were prepared to tailor the thermal conductivity of the material for monoblock of divertor. The layering will reduce thermal shocks and control heat transfer to the copper-based cooling system.
        We have already confirmed, that sintering of W and WC particles by FAST induces in-situ high-temperature reaction with the final composition of cubic W with hexagonal W2C phase. W-W2C composites exhibit high density and improved mechanical properties at room and elevated temperatures. If the W2C content in the composite is 5 wt % or higher, W-grain growth is inhibited even after aging at 1600°C for 24 h, due to the pinning of W grain boundaries with smaller W2C. The mechanical properties of aged samples did not impair, and chemical composition remained unchanged. Layered composites with the gradually increased content of W2C was successfully prepared. If the difference in the carbon content between two layers is too high, the growth of elongated W2C grains is induced. XRD and EBSD analyses confirmed that these elongated W2C grains had preferred orientation (0,0,2). The microstructures of such phases were thoroughly examined.

        Speaker: Dr Matej Kocen (Department for Nanostructured Materials, Jožef Stefan Institute)
      • 11:00
        Leak detection design for ITER gas injection system 2h

        The main functions of ITER Gas injection system(GIS) are providing gas fueling(H2,D2, T2, 4He/3He, N2/Ne, Ar) for plasma, wall conditioning operation and neutral beam injectors. If there is leak on the gas supply lines during ITER plasma operation state, abnormal gas composition will affect or have potential to affect operation. Furthermore, Out-leak of Hydrogen or Tritium from gas supply lines will lead to safety issues, therefore leak of gas supply lines especially H2(D2)/T2 lines shall be detected as soon as possible during the operation. Leak detection design for GIS H2(D2)/T2 gas supply lines during ITER operation state is introduced in this paper.
        Pressure monitoring of manifold is an effect way to check a problem of a line. Considering the handling of tritium, all the gas supply lines are enclosed in a secondary containment pipe(guard pipe), pressure of which shall be lower than the environment pressure. Since the nominal pressure of the gas supply pipe is also lower than the pressure in the interspace, by monitoring and comparing the pressure in gas supply pipe and guard pipe a leak of gas supply line could be identified with reasonable control logic.
        Besides, hydrogen isotope may leak out to the interspace by diffusion. In this case Hydrogen detector and Tritium detector are employed to the interspace. The manifold interspace is continuously vented to Detritiation System with N2, therefore H2(D2)/T2 detectors are connected to monitor their concentration in the flow. Requirements that the interspace pressure are maintained to be lower than the environment pressure, at the same time the hydrogen isotope gas in the flow could be transferred to the detector as fast as possible are taken into consideration for the design of purge configuration and parameters( pipe diameter, flow rate, etc.).

        Speaker: Xiangmei Huang (Southwestern Institute of Physics)
      • 11:00
        Lithium Loop and Purification System of DONES: Preliminary Design. 2h

        The Demo-Oriented NEutron Source (DONES) is an essential irradiation facility for testing candidate materials for DEMO reactor and future fusion power plants. An intense flux of highly energetic neutrons is generated by the nuclear reactions of a 125mA beam of deuterons at 40MeV striking a liquid lithium target. The neutron flux achieves a damage rate of 8-10 dpa/fpy in a volume of about 0.3 l with a helium production rate of ~10-13 appm He/dpa. The main lithium loop and the related purification system have to generate a stable lithium jet at the target and guarantee: a high speed flow to evacuate the deposited heat (5 MW) and avoid boiling or significant evaporation of the lithium; a constant shape and thickness of the jet to assure a constant neutrons flux and prevent impingement of the beam on the back-plate, being the latter the surface just behind the jet; and an adequate level of chemical impurities solved in lithium. The preliminary design of the two systems is concluded. In this work, the lay-out of the loops, piping dimensioning, pressure drop evaluations, definitions of supports, piping stress analysis, and the design of the traps to remove the impurities are described in detail.

        Acknowledgments
        This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission.

        Speaker: Francesco Saverio Nitti (FSN ENEA)
      • 11:00
        Magnetic interaction between a tokamak reactor and its reinforced-concrete building 2h

        Magnetic interaction between a tokamak reactor and its iron reinforced-concrete basement has been studied using the analytical model and ANSYS electromagnetic code. When the magnetic material is used for tokamak building, the leakage magnetic field from the tokamak is enhanced due to the normal angle incidence of the magnetic field line to the magnetic material wall. As this study is motivated from the ITER building construction, we use the similar dimensional data on ITER building. In this study, the magnetic field is assumed to be stationary.
        We have analyzed the influences of the magnetized material on the position of the magnetic field null in the plasma break down phase, and on the x-point location during the divertor operation. We assume the iron wall plate instead of discrete reinforced-concrete wall, so that this assumption provides the larger magnetic disturbance than that of the actual case with rebars. When we assume floor and bio-shield wall made of 1 m thick iron plate with the relative permeability of 100, the null regime is shifted to the inboard side in the vacuum chamber. This can be understood by image currents induced by the ceiling, floor and outer wall. As the magnetic null regime is in the small magnetic field area, it is disturbed when it is away from the floor and the ceiling. Such shift could be adjusted by the PF coil current. On the other hand, the x-point location during divertor operation is not affected by the magnetic wall because of the dominant contribution of the large poloidal coil current.
        While the effect of the magnetic material on the plasma performance seems to be not crucial, we further investigate the degaussing operation, the mechanical strength of the floor due to the magnetic force and the induction effect by PF coils.

        Speaker: Osamu Mitarai (Institute for Advanced Fusion and Physics Education)
      • 11:00
        Manufacturing and testing of flat type small size tungsten PFC mock-ups by HIP process 2h

        W/CuCrZr PFCs will be used in ITER divertor and are strong candidate for the use in high heat flux regions of the upgraded KSTAR and K-DEMO. Development of hot isostatic pressing (HIP) bonding technology is in progress for the fabrication and qualification of tungsten divertor. We manufactured the first W/CuCrZr flat type small size mock-ups by HIP technology using PVD for interlayer formation. In the developed HIP process, we applied PVD method for interlayer coating of high purity Ni and Cu. The interlayer coating is selected to enhance the diffusion between W and Cu during the HIP bonding process. The Ni and Cu were deposited by two kinds of thicknesses on W surface. The bonded PFC of Ni/Cu coated W, copper interlayer and CuCrZr were manufactured by means of HIP technology employing two kinds of processing conditions. We observed the microstructure of the deposited Ni and Cu layer on the bonding interface by FE-SEM. The HIP processed samples were subjected to shear strength test and high heat flux (HHF) test. There is no significant difference in shear strength due to PVD coating thickness. But the shear strength of samples fabricated by 70 MPa and 700 ℃ HIP condition is higher than those fabricated by 60 MPa and 600 ℃ condition. HHF tests were performed to compare the thermal conductance through the bonded layer and thermal fatigue behavior of the samples in the tested condition as a preliminary test of the HIP process parameters. The layers between W and CuCrZr of samples were observed and compared by FE-SEM before and after HHF test. And bonding with Ti interlayer is compared to those of Ni interlayer in bonding strength and stability. All these results were analyzed synthetically to establish the optimum fabrication conditions and to find a method for performance qualification.

        Speaker: Dr Eunnman Bang (DEMO Technology Division, National Fusion Research Institute)
      • 11:00
        Material Irradiation Tests in the ITER Divertor Relevant Settings 2h

        Various types of multilayer laser mirrors and piezoelements underwent radiation tests to assess the influence of neutron and gamma-ray fluxes similar to those expected in diagnostic ports of ITER divertor. The optical and thermal performance of laser mirrors and the piezoelectric coefficient of the piezo-elements were under investigation. The test was performed in the RIAR irradiation facility RBT-6 using ITER relevant neutron and gamma fields with a neutron flux of ~10^12 n/cm^2/s and a fluence of ~10^19 n/сm^2 (E > 0.1 МeV); heating did not exceed 200 C. Special racks for arranging the test samples of mirrors and piezo-elements in the reactor were designed with neutron shields for adjusting neutron and gamma-ray spectra; also, helium atmosphere for the in-situ cooling was used. The neutron spectra, estimated using MCNP and tested experimentally, are provided and compared with those expected in ITER under baseline inductive burning plasma conditions (Q = 10). The tested laser mirrors (40 samples of 4 types) have a metallic sublayer (Al or Ag) and a multilayer dielectric coating (ZrO2 / SiO2 or Sc2O3 / SiO2). The mirrors are designed for broadband reflection and high laser power breakdown threshold at the wavelengths 956nm, 1047nm and 1064nm (high-power lasers of divertor Thomson scattering), as required for the combined diagnostics 55.EA (laser-induced fluorescence) and 55.C4 (Thomson scattering). The mirror breakdown threshold and heating resistance up to 200-250 C were tested before irradiation; further tests are planned. The piezoelectric coefficient of the two piezo-ceramic types chosen for stick-slip and resonance piezo actuator types was measured before and after irradiation. Experimental approaches and facilities used for irradiation, and pre- and post-irradiation measurements are described in detail. The first results and outlook of further experiments are also presented.

        Speaker: Dr Eugene Mukhin (Ioffe Institute)
      • 11:00
        Material optimization technique to minimize radiological responses in fusion reactors 2h

        High-energy, high-intensity neutrons emitted from the fusion plasma present a stringent environment for the structural materials present in the fusion device. This has significant life-limiting effects on the reactor components. The neutrons interact with the material initiating nuclear reaction leading to the production of radioactive isotopes, gas molecules and material defects. These gases, particularly helium, can cause swelling and embrittlement of the material. Furthermore, the radioactive isotopes produce would cause heating in the material. Some of these isotopes may have long lives which would contribute towards the radwaste produced in the fusion devices. Hence designing of low activation materials for fusion devices is warranted.
        At Iter-India, Institute for Plasma Research a number of computational tools are being developed to estimate the nuclear response of the materials. ACTYS-1-GO, a multipoint neutron activation code which can calculate radiological responses of materials located at various positions in a fusion reactor efficiently, is developed. Along with it, a mathematical framework is developed for accessing the relationship of radiological quantity with the initial elements present in the material. Such framework helps in identifying and thus minimizing the elements/isotopes from the initial material composition.
        In the present study both the methodologies are efficiently coupled for a complete material optimization. Quantities responsible for various radiological effects (like activity, dose, heat, and radwaste) and related defects in the material are at all irradiation times considered and their contributing elements are sorted and minimized accordingly. Also, since a single material faces a gradient of neutron flux over its entire volume, all such optimization is carried out over the entire range of neutron flux faced by that material. This would provide the best material composition in accordance with the neutron environment faced by the material and its functionality, minimizing the radiological effects like activity, dose, gas production and radwaste.

        Speaker: Priti Kanth (Institute For Plasma Research)
      • 11:00
        Maximization of the magnetic flux generated by a DEMO CS coil using HTS conductors 2h

        The present work is performed within the framework of the EUROfusion DEMO project. Previously, it was demonstrated that for a maintained magnetic flux the use of HTS conductors at highest magnetic field in a layer-wound CS coil would allow the reduction of its outer radius by around 0.5 m as compared to the DEMO reference design using only Nb3Sn conductors. Alternatively, the superior high field performance of HTS conductors can be used to maximize the magnetic flux of a CS coil with a given outer diameter, which could significantly prolong the duration of plasma burn phase and thus the overall power plant efficiency. In the present study, the maximization of the magnetic flux, generated by a layer-wound CS coil of fixed outer radius, is considered. HTS conductors are envisaged to be used at highest field, while Nb3Sn and NbTi are foreseen to be used in intermediate and low field layers, respectively. The inner radius of the CS coil is optimized with respect to the generation of a maximum flux taking into consideration the superconductor properties, the hoop stress and the axial stress. In order to provide reasonable starting values for the finite element analysis, first a simplified model with layer-dependent current densities, however, without stainless steel grading will be considered. In the final outline design of the CS coil a superconductor and stainless steel grading will be implemented.

        Speaker: Dr Rainer Wesche (Swiss Plasma Center, EPFL)
      • 11:00
        Measurement of neutron fluence in the High-Flux Test Module of the Early Neutron Source by an activation foils method 2h

        The neutron fluence is an important normalization parameter for the material specimens to be
        irradiated in the Early Neutron Source (ENS). The activation foil method appears suitable for
        this purpose considering cost, low technical requirements and invasivness.
        Small packages of thin activation foils can be placed in several locations: on the outer surface
        of the HFTM, on the outside of specimen capsules inside the HFTM or inside the specimen
        capsules. The latter would provide measurements very close to the specimens while the other
        two options require more corrections to determine the neutron fluence in the place of the
        specimen. Each location has a different access time after completion of the irradiation cycle. If
        the activation foils are mounted on the outer surface of the HFTM the estimated earliest access to
        them for measurement of the induced gamma activity would be approximately one week after
        shut-down. An activation foils package on the surface of a specimen capsule would be accessible
        about three weeks after shut-down while an activation foil package inside the specimen capsule
        becomes available two to three months after shut-down.
        In this work we will present a set of activation foils which is suitable for application in the
        HFTM. The set consists of iron, cobalt, nickel, yttrium, and gold. The selected dosimetry
        reactions lead to radioisotopes with half-lives of several months up to a few years so that they
        preserve neutron flux information over the full irradiation time, and they cover the entire ENS
        neutron energy range. Tests of the measurement method were performed with the cyclotron
        neutron source at NPI Řež which provides a fast neutron spectrum similar to ENS. We will
        present the analysis of these tests together with a review of state-of-the-art evaluated cross
        sections of the dosimetry reactions of interest.

        Speaker: Dr Axel Klix (Karlsruhe Institute of Technology)
      • 11:00
        Mechanical properties of dissimilar TIG welding for ARAA and SS316L 2h

        Korea has designed a helium cooled ceramic reflector (HCCR) breeding blanket for developing the Korean DEMO and fusion reactor, including the development of the reduced activation alloy, ARRA (Advanced Reduced Activation Alloy). From the lesson of the developing procedure of the HCCR test blanket module (TBM) for ITER, it is known that the various fabrication methods, such as electron beam welding, TIG welding, and HIP (Hot Isostatic Pressing) bonding, should be developed for the fabrication of some complex structures. In the current research, the developed dissimilar TIG welding using ARAA and SS316L was introduced. The tensile, impact, bend tests were determined after post-weld heat treatment (PWHT) and hardness, microstructure characteristics were determined before and after to evaluate the welded specimen under the determined welding conditions. The average hardness values before were 338 HV in the HAZ and 199 HV in the WM, and after PWHT, the values decreases to 243 HV in the HAZ and 221 HV in the WM. Additional tests are underway for the ductile brittle transition temperatures (DBTT) on the heat affected zone (HAZ) and on the weld metal (WM). Tensile strength has decreased as the increase of temperature. The yield strength (YS) and ultimate tensile strength (UTS) were 286 MPa and 572 MPa, respectively

        Speaker: Jae Sung Yoon (Korea Atomic Energy Research Institute)
      • 11:00
        Methods and strategies on thermal integrity management of the ITER Thermal Shield 2h

        Due to its main function as provider of a thermal radiation opaque barrier to the superconducting magnets, the ITER Thermal Shield (TS from now) design guarantees an appropriate thermal behaviour during operation. All the methods and strategies implemented with this purpose on the design, manufacturing and assembly of the TS, constitute the so called TS Thermal Integrity Management. The scope of this paper is to provide a comprehensiveness description of the TS Thermal Integrity Management keeping focussed on three main aspects: last methodology for an accurate estimation of the TS heat loads; advanced thermo-hydraulic modelling; combined analytical-experimental method for sizing pipe orifices. The TS heat loads estimation has to be accurate in order to know the margins regarding the agreed limits with the Cryoplant and Magnets. The proposed methodology includes detailed thermal radiation view factors involving all the systems interfacing with the thermal shield. This methodology is applied on a simplified case and compared with previous methods, resulting on a better understanding of the level of conservatism and the accuracy of the estimated heat inputs on the TS. A detailed analysis of the coolant temperature rise on the cooling tubes is also described in combination with experimental data and fluid-dynamic analysis of the required pipe orifices. As a result, a pipe orifices sizing method is described and new thermo-hydraulic parameters are found for feeding the planned final flow balance analysis of the thermal shield. As a conclusion of the paper, it is presented how the described methods are integrated onto the strategy that aims to manage the actual component emissivity data, being a strong tool for assessing the impact of the silver coating quality (during manufacturing and assembly) on the expected TS thermal behaviour. The views and opinions expressed herein do not necessarily reflect those of the ITER Organization.

        Speaker: Dr Germán Pérez-Pichel (ITER)
      • 11:00
        Metrology for integration and installation activities at the PRIMA Test Facility 2h

        The ITER project requires at least two Neutral Beam Injectors, each accelerating up to 1MV a 40A beam of negative deuterium ions, so as to deliver to the plasma a power of about 33 MW for one hour.
        Since these requirements have never been experimentally met, it was recognized necessary to build-up a test facility, named PRIMA, that is in an advanced state of realization and which includes both a full-size negative ion source (SPIDER) and a prototype of the whole ITER injector (MITICA).
        The paper describes the main metrology activities performed in the last three years devoted to the integration and installation of the large number of items and plant units composing the facility. Particular emphasis is given to the propaedeutic activities consisting mainly in the definition of the metrology network (the so called Unified Spatial Metrology Network - USMN) by using technologically advanced laser trackers. The USMN is a feature of the Spatial Analyzer software (SA) compliant with the ISO standard that using the motecarlo method is capable to reduce the global measurement uncertainty. The method is based on the installation and measurement of a large number of fiducial points (approximately two hundreds targets). For PRIMA, some local USMN networks have been built up at different locations resulting eventually in the definition of the PRIMA USMN network. This approach allowed the definition of the global reference frame to be used for the positioning of all items, while respecting the uncertainty requirements of each component. In the paper some instances will be given like the positioning of the transmission line (uncertainty of 0.2mm over more than 100m between first and last tank) and the high voltage bushing support structure for MITICA, the vacuum vessel and the beam source for SPIDER (uncertainty better than 0.01mm of the grid apertures).

        Speaker: Dr Samuele Dal Bello (Consorzio RFX)
      • 11:00
        MHD mixed convection flow in the WCLL: heat transfer analysis and cooling system optimization 2h

        In the Water-Cooled Lithium Lead (WCLL) blanket, a critical problem faced by the design is to ensure that the breeding zone (BZ) is properly cooled by the refrigeration system, thus to keep the structural materials under the maximum allowed temperature. For this purpose, CFD simulations are carried over using ANSYS CFX to investigate how the cooling system performances are affected by the tokamak magnetic field.
        In the configuration studied, the blanket relies on a Single Module Segmentation approach. The LiPb flows inside long poloidal channels, developing along the whole module height; cooling is provided by double-walled tubes which are separated by a vertical pitch and inserted in the BZ from the back-supporting structure.
        The attention is focused on the sub-channel closest to the first wall (FW). The maximum of the neutronic power deposition is foreseen in this region; thus, intense buoyancy forces will arise due to the large temperature gradient in the radial direction (Gr≈10^10). These will interact with the forced convection flow (Re≈10^3), leading to the onset of a mixed convection regime. A constant magnetic field parallel to the toroidal direction is assumed with intensity in the Hartmann number range 0≤M≤10^4. The walls bounding the channel and the water pipes are modeled as perfectly conducting.
        The magnetic field is found to dampen the velocity fluctuations triggered by the buoyancy forces and, for M=10^4, the flow can be reduced to a pure forced convection regime. This effect causes the sharp reduction of the LiPb heat transfer coefficient to one-third of the ordinary hydrodynamic value. Consequently, hot-spots between the nested pipes and close to the FW are observed where the temperature exceeds the maximum allowed value. To address this issue, optimization strategies for the BZ cooling system layout are proposed and implemented in the CFD model, fulfilling the design criterion.

        Speaker: Alessandro Tassone (Department of Astronautical Electrical and Energy Engineering Sapienza University of Rome)
      • 11:00
        MIMO shape control at EAST tokamak: simulations and experiments 2h

        Over the last few years, new magnetic control algorithms have been developed and tested on the EAST tokamak. The aim is to improve the overall plasma performances and to open the way to the control of advanced plasma magnetic configurations [1]. In order to achieve such an objective, an architecture based on a MIMO plasma shape controller was proposed in [2].
        This architecture relies on specific control algorithms for plasma vertical stabilization and for the current control in the Poloidal Field (PF) circuits. These components have been designed exploiting the CREATE magnetic models. In particular, a voltage-driven Vertical Stabilization system, a controller for the plasma centroid position, and a multi-input-multi-output (MIMO) PF currents controller were implemented and experimentally validated in 2016-2017 [3] [4].
        The plan for the 2018 EAST experimental campaign is to test and validate the MIMO plasma shape controller that relies on the architecture tested in 2016-2017, and that adopts an approach similar to the one used by the JET's eXtreme Shape Controller. Such an approach will enable the integrated control of the plasma boundary and of the heat flux on the divertor plates.
        In this paper the simulation results obtained with the CREATE modelling tools are compared with the ones obtained experimentally for different setups of the magnetic control system.

        [1] G. Calabrò et. al., «EAST Alternative Magnetic Configurations: Modelling and First Experiments,» Nuclear Fusion, 2015.
        [2] R. Albanese et. al., «A MIMO architecture for integrated control of plasma shape and flux expansion for the EAST tokamak,» in IEEE Multi-Conference on Systems and Control, 2016.
        [3] R. Albanese et. al., «ITER-like Vertical Stabilization System for the EAST tokamak,» Nuclear Fusion, 2017.
        [4] G. De Tommasi et. al., «Model-based plasma vertical stabilization and position control at EAST,» FED, 2018.

        Speaker: Dr Adriano Mele (CREATE)
      • 11:00
        mitigation of an ingress coolant event in ITER vacuum vessel by means of steam pressure suppression 2h

        An Ingress of Coolant Event (ICE) is postulated to occur in the ITER Vacuum Vessel (VV) due to a breach on the first-wall cooling channels. The pressure raise in the VV is limited by means of a Vacuum Vessel Pressure Suppression System (VVPSS), consisting of relief lines connected to the VV and discharging the steam to four Vapor Suppression Tanks (VST) partially filled with water: one managing small water leaks and three together managing large water leaks).
        The ITER vapor suppression system operates at sub-atmospheric pressure. Since no detailed information has been found in the technical literature for similar operating conditions, a specific study has been launched at the University of Pisa to investigate the performance of the vapor suppression process in the VSTs.
        Both an analytical-numerical analysis and an experimental activity were carried out on a 1:21 scale apparatus, simulating the VVPSS.
        Scaling studies were first performed in order to calculate the transient of steam mass flow rate occurring during the ICE IV event in the scale apparatus, starting from the thermal hydraulic study performed at ITER. Subsequently, a quite extensive experimental campaign was carried out on the scale apparatus in order to study the influence of the main thermal hydraulic parameters that characterize the steam condensation efficiency in sub-atmospheric conditions.
        This paper illustrates the results obtained from the experimental transient simulating the Ingress of Coolant Event in the VST Tanks.
        The experimental results confirm the determined scale predictions and the capability of the VVPSS tank to condense the injected steam at sub-atmospheric pressure, matching the safety goal to limit the VV pressurization.

        Speaker: Dahmane Mazed (DICI Dipartimento di Ingegneria Civile e Industriale universitˆ di Pisa.)
      • 11:00
        Modal and response spectrum analyses of ITER divertor module 2h

        While most of previous numerical analyses have been carried out under thermal and electromagnetic loads due to their significance, severe dynamic loads may also threat its structural integrity. The present study is to investigate resistance of complex ITER divertor module against typical seismic loads. Two kinds of huge finite element models, which consists of cassette body, inner and outer vertical targets, dome and stabilizers, were developed; one is simplified model without coolant tubes and the other is detailed model with three layered coolant tubes. At first, modal analyses to predict dynamic characteristics such as frequencies and mode shapes were conducted by employing either block-Lanczos algorithm or symmetric coupling algorithm considering water in coolant tubes. Subsequently, response spectrum analyses were performed with complete quadratic combination technique by taking into account different seismic magnitudes based on ASME (American Society of Mechanical Engineers) B&PV Sec. III Appendix N. As results, calculated stress intensities at critical locations were compared with corresponding design stress intensities, according to ASME code rule, of which details dependent on sensitivity parameters were discussed.

        Speaker: Dr Sang Yun Je (Nuclear engineering, Kyung Hee university)
      • 11:00
        Modelling of chemical vapor deposition to improve tungsten fiber reinforced tungsten composites (Wf/W) 2h

        Due to the unique combination of excellent thermal properties, low sputter yield, hydrogen retention and activation, tungsten is the main candidate for the first wall material in future fusion devices. However, its intrinsic brittleness and its susceptibility to operational embrittlement is a major concern. To overcome this drawback, tungsten fiber reinforced tungsten composites featuring pseudo ductility have been developed. Bulk material can be successfully produced utilizing chemical vapor deposition of tungsten fabrics. However, a fully dense composite with a high fiber volume fraction is still a huge challenge.

        Therefore, a model is currently developed in COMSOL including the complex coupling of transport phenomena and chemical reaction kinetics. To validate the model with experimental data, fibers were deposited in heated tubes under controlled parameter variation. The temperature and tungsten growth rate were measured along the fibers and inner tube surfaces for different heater temperatures, partial pressures and gas flows. With the experimental results the prediction of the model has been improved. As next step the model will be applied to design infiltration experiments to fabricate fully dense Wf/W composites with a high fiber volume fraction.

        Speaker: Dr Leonard Raumann (Forschungszentrum Jülich)
      • 11:00
        Modelling of MAST-U neutral beam re-ionisation and the impact on the beamline ducts and in-vessel components 2h

        Neutral beam injection is one of the primary auxiliary heating systems for tokamak plasmas. Once the neutral beam leaves the neutraliser collisions with background neutral particles in the beamline and tokamak vessel re-ionises part of the neutral beam. These particles can be deflected by the tokamak magnetic field, potentially damaging unshielded components.
        The first stage of the Mega Amp Spherical Tokamak Upgrade (MAST-U) has two Positive Ion Neutral Injectors (PINIs), one injecting power to be deposited close to the magnetic axis of the plasma (on-axis) and one injecting power that is deposited further out in the plasma (off-axis). Each injector has been designed to run for up to 5 seconds, with 2 seconds of neutral beam required for MAST-U core scope. The increase in neutral beam pulse length and the increase in complexity of the in-vessel hardware compared to MAST increases the risk posed by re-ionisation to the beamline ducts and in-vessel components.
        This paper describes modelling of the neutral beam re-ionisation along the beamlines and into the MAST-U vessel, within a field envelope applicable to the operational space of the machine. Monte-Carlo simulations using the MAGNET code give the re-ionised power loading on both beamline ducts, which was used to position thermocouples on the duct liners. Thermal and structural analysis of the duct liners has been carried out using ANSYS.
        The potential for significant amounts of re-ionised power entering the MAST-U vessel has resulted in the installation of graphite tile plating at the end of the ducts, compatible with co-injection. The power loading on these plates and other in-vessel components as modelled by MAGNET has been verified using the LOCUST code. Analysis has confirmed that the graphite tiles are compatible with core scope operation.

        Speaker: Dr Alastair Shepherd (Culham Centre for Fusion Energy)
      • 11:00
        Multiphysics Analysis of W7-X Control Coils 2h

        The quasi-symmetric fivefold modular Wendelstein 7-X (W7-X) stellarator consists of three groups of coil systems, i.e. superconducting magnet, trim coil and control coil systems. The control coil system contains ten identical 3D shaped control coils (CC) situated behind the baffle plates of corresponding divertor unit, and is designated to rectify the error field and to sweep hot spots on the divertor target plates. The CC is wound from copper conductor with a square cross section of 16 mm x 16 mm and a water cooling hollow of Ø 8 mm. The control coil system was installed in W7-X in 2015, and the integral commissioning has been done in parallel with the completion of W7-X. During the operation phase (OP 1.2a) with limited plasma heating power, a leakage in one of the CC cooling water plug-in was found and dictates a detailed transient thermal analysis of CC to determine the allowable operation time without cooling water flow. The paper presents the transient thermal analysis and is followed by a detailed finite element mechanical analysis with the consideration of temperature gradient loads, dead weight and electromagnetic forces. Moreover, the transient thermal and mechanical performance of actively cooled CC to be intensively operated during state steady operation phase (OP 2) are also analyzed and evaluated with the same FE model.

        Speaker: Dr Jiawu Zhu (Max-planck Institute for plasma physics)
      • 11:00
        Neutronic assessment of HCCR breeding blanket for DEMO 2h

        Main goals of breeding blanket development in Korea are to develop and verify the integrated blanket design tools; to develop the core technologies such as blanket materials, blanket cooling, and tritium fuel cycle technologies; and to develop and evaluate fabrication and joining technologies. Several breeding concepts are considered as candidates for the Korean DEMO blanket concept. As a solid breeder concept, helium-cooled ceramic reflector (HCCR) blanket and water-cooled solid breeder blanket (WCSB) are considered. The HCCR blanket adopts the unique graphite reflector concept to reduce the amount of beryllium multiplier with the Li2TiO3 breeder, reduced activation ferritic-martensitic steel structural material, and helium coolant. This concept of HCCR blanket is adopted to be tested in ITER as a Test Blanket Module (TBM). Currently, the design and R&D activities for HCCR blanket has been performed in Korea.
        In this paper, the design concept of the HCCR breeding blanket system is explored to satisfy the global TBR requirement by a neutronic assessment based on the K-DEMO neutronic analysis model employing vacuum vessel, toroidal field coil, blanket and shield. Firstly, sensitivity studies were performed for a various combination of HCCR blanket breeding layers with fixed blanket thickness. The thicknesses and arrangement of the breeder, multiplier and reflector layers in the HCCR blanket were optimized in the view of tritium breeding ratio. Secondly, the neutron flux distribution inside the blanket was calculated to evaluate the neutron shielding performance of the blanket. Finally, the nuclear heat distribution on the blanket was also estimated to support the cooling system design study.

        Speaker: Seungyon Cho (National Fusion Research Institute)
      • 11:00
        Neutronic assessments towards a comprehensive design of DEMO with DCLL Breeding Blanket 2h

        On the way towards a comprehensive design of DEMO, step by step all the systems and components must be introduced as their definition or refinement progresses, in order to demonstrate the viability of a design on larger scale, i.e. leaving fewer margins to undetermined questions.
        Among the EUROfusion Programme, new aspects have been recently fixed or furtherly developed as the Divertor, the First Wall (FW) and the Flow Channel Inserts (FCI) designs. Furthermore, the integration of Heating and Current Drive (H&CD) systems, as the Neutral Beam Injector (NBI), has started.
        The introduction or modification of these systems and components could seriously jeopardize the nuclear behaviour of an initially validated Breeding Blanket (BB) DEMO concept, since many neutronics criteria - among others - could be no more fulfilled. Since the design of DEMO is a continuous upgrade under iterative process, as the refinement push on, most of the studies have to be repeated to demonstrate that criteria are still respected in a fully integrated design.
        In this work the influence on Tritium Breeding Ratio (TBR) of a new design of detached FW protecting BB from extremely high heat fluxes is investigated. The impact of different typologies of FCIs is assessed also according to the degree of detail in the neutronic description. The divertor composition and its topology also reveal to have strong impact on responses apparently not related with its design, as the tritium production in the BB. Besides, the integration of NBI minimizing its invasiveness in the BB is verified by neutronic analyses concerning the main BB functions: fuel breeding and heat recovery. Accordingly, TBR and Nuclear Heating (NH) are assessed.
        The study is performed for a Dual Coolant Lead Lithium (DCLL) BB DEMO although can be extrapolated to other BB concepts.

        Speaker: Iole Palermo (Fusion Technology Division CIEMAT)
      • 11:00
        Neutronic effects of the ITER Upper Port environment update in C-Model 2h

        The goals of this work are the neutronic modeling of the ITER Upper Port (UP) environment according to the updates of ITER CAD model, the assessment of neutronic effects caused by that update and proposing improvements of the radiation conditions. The update has been applied to the ITER-reference neutronics simulation model called “C-Model” which includes the standard components and generic port plugs. The modifications of UP environment are related to the Vacuum Vessel (VV) UP extension and the port duct. The particular components of the UP environment are the following: VV double and single walls around the Diagnostic Generic Upper Port Plug (DGUPP), UP stub and blanket manifolds, In-Vessel coils (IVC) including Edge Localized Mode (ELM) coils and feeders, and vacuum sealing flange. The neutronics analysis of the newly developed model identified the components which influence predominantly the radiation field at the UP Inter-Space Structure (ISS) where personnel access is anticipated after shutdown. The radiation environment was analyzed in terms of neutron fluxes and Shut-Down Dose Rates (SDDR). High-resolution distributions of the SDDR have been obtained with the R2Smesh method of KIT coupling MCNP radiation transport simulations with isotope inventory calculations using the FISPACT activation code. The material specifications of the model are based on the approved material compositions taking into account the radiological impurity requirements. It improves the reliability of the radiation transport simulations for the assessment of the radiation conditions in the UP area and for the implementation of the ALARA principle for SDDR analysis. In conclusion, several design solutions are proposed to reduce radiation streaming in the direction of the ISS maintenance area.
        Disclaimer: This work has been funded by the ITER Organization (IO) service contract IO/17/CT/4300001478. This paper does not commit the IO as a nuclear operator. The model qualification is provided independently of this paper.

        Speaker: Dr Arkady Serikov (Institute for Neutron Physics and Reactor Technology (INR), Karlsruhe Institute of Technology (KIT),)
      • 11:00
        Neutronics pre-analysis and the status of neutron spectrum unfolding for the development of VERDI 2h

        In future fusion power plants, such as DEMO, D-T neutron emission is predicted to exceed 1e21 neutrons/second. Accurately monitoring neutron energies and intensities will be the primary method for estimating fusion power, and calculating key parameters, including the tritium breeding ratio and nuclear heating. The Novel Neutron Detector for Fusion (VERDI) project, implemented under the EUROFusion Enabling Research 2017 program, aims to develop a novel neutron detector, capable of withstanding the harsh environment of a future fusion power plant. The VERDI detector is based on the foil activation technique, which relies neutron spectrum unfolding methods to process the convolution of gamma-ray measurement and detector response function and infer the neutron energy spectrum. A benchmark experiment has already been performed at the Frascati Neutron Generator, at ENEA, details of this experiment are provided in another submission. In this paper, experimental results from the Frascati Neutron Generator have been applied to neutron spectrum unfolding techniques. Calculations performed using MCNP and FISPACT-II were successful in predicting the gamma-spectroscopy peaks, and the initial use of MAXED has provided a neutron energy spectrum in good agreement with expectations. Results from the benchmark experiment have shown that careful treatment of (n,gamma) reactions are needed for neutron spectrum unfolding to obtain physically realistic solutions at low energy. The application of the VERDI detector is now being extended to JET to explore behaviour in both short-term and long-term experiments. This paper will focus on the pre-analysis calculations that have been performed in preparation for upcoming experiments during the JET D-D campaign. The status of neutron spectrum unfolding will also be discussed in this overview, along with plans for the development of neutron spectrum unfolding.

        Speaker: Dr Chantal Rebecca Nobs (Culham Centre for Fusion Energy)
      • 11:00
        Nuclear analyses of solid breeder blanket options for DEMO: status challenges and outlook 2h

        Within the Power Plant Physics and Technology (PPPT) programme of EUROfusion, an intensive development effort is devoted to the detailed design of a solid breeder blanket for a demonstration fusion reactor (DEMO) with the inherent capability of a highly efficient tritium breeding. A novel design of the Helium Cooled Pebble Bed (HCPB) breeding blanket based on a Single Module Segment (SMS) and an enhanced, near term configuration has been developed for the EU DEMO reactor, having high flexibility with regard to tritium generation and global reactor performances. The key part of the analyses is the development of the geometry model of the breeder blanket with enough detail to perform high fidelity nuclear simulations. To this end a generic geometry model of a DEMO 2017 baseline was adopted to arrange the newly developed SMS HCPB layout using a novel modelling technique. Separate models were developed using detailed CAD designs and the McCad conversion tool: the SMS blanket module, a breeder unit and a highly detailed First Wall (FW) with cooling channels and a roof-top shape. In this way, the model includes a detailed representation of the breeder zone, the back supporting structure and the FW of the blanket.
        This model has been applied for the study of this new generation of HCPB blanket , as well as for the investigation of an alternative concept based on liquid lead neutron multiplier, replacing the Be/beryllide one. The simulations include assessment of the main required nuclear responses: TBR, nuclear power generation and shielding performances. A special technique has been also applied to find the optimal geometry configuration of the different concepts. The study concludes with sophisticated activation analyses of the in-vessel components and with the discussion of different effects coming from the detailed modelling and affecting engineering solutions, current design highlights, challenges and outlook.

        Speaker: Pavel Pereslavtsev (Institute for Neutron Physics and Reactor technique (INR) Karlruhe Institute for Technology)
      • 11:00
        Numerical study of conjugated heat transfer for DONES high flux test module 2h

        Helium flows at low pressure (0.3 MPa) are used to cool the specimen capsules and the structure of the neutron irradiated High Flux Test Module (HFTM) of the DEMO-Oriented Neutron Source (DONES). The flow path includes inlet and outlet ducts with large cross sections, but also mini-channels with 1 mm gap width, where a high velocity low Reynolds number laminar to turbulent transitional heated flow influences the temperature of the irradiated specimens. The large span of Reynolds numbers from laminar to fully turbulent are a significant challenge for the simulation of the complete HFTM and requires validation of models.
        Numerical simulations have been carried out for turbulent (Re = 4500, 6000 and10000) helium flow in a heated mini-channel using the commercial CFD code Star-CCM+ v.12.06. The results were compared with measurements obtained from the IFMIF thermal-hydraulic experiments (ITHEX). Detailed comparisons were made between the k-ϵ, k-ω-SST, and V2F models with different treatment methods of near-wall layer. The appropriated models have been used for the numerical thermo-hydraulic analysis of the conjugated heat transfer the in HFTM. The temperature distribution in the HFTM structure calculated for normal operational conditions serves as a basis for a subsequent thermo-mechanical structure analysis of the HFTM.
        Acknowledgments: This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 under grant agreement No. 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission.

        Speaker: Sergej Gordeev (Institute for Neutron Physics and Reactor Technology Karlsruhe Institute of Technology)
      • 11:00
        On optimization of air cooling system of FDRs dissipating energy from ITER magnet coils 2h

        The Fast Discharge Resistors (FDRs) under development at NIIEFA are intended together with switching equipment to dissipate energy released in case of a quench of the ITER superconducting coils, thereby protecting them against failure. FDRs are made of sections consisting of steel resistive elements enclosed in boxes. Two-four sections stacked vertically form a separate module. During energy release the resistive elements are heated to 250-300⁰С practically adiabatically. The resistors should be cooled to their initial temperature within 3-4 hours. For this purpose, the authors have developed the air cooling system based on the forced air circulation produced by seventeen fans in a complex system of series-parallel channels formed by air supply and return pipes, vertical modules and chimneys. The numerical simulation of the cooling process revealed that distribution of the air flow in the parallel channels formed by the vertical modules is considerably non-uniform, which essentially increases the module cooling time. The study performed by the authors within the last few years has made it possible to propose measures on optimizing the air cooling system mitigating the negative effect of air flow non-uniformity in the FDR modules. The reported analysis continues the previously performed studies of the FDR cooling system. The idea to install diaphragms in each module to equalize the air flow in the cooled-in-parallel modules has made it possible to reduce the time for their cooling to the specified values without a considerable reconfiguration of the air cooling system.

        Speaker: Dr Victor Tanchuk (JSC "NIIEFA")
      • 11:00
        On Structural Analyses of the ITER Vacuum Vessel Bolometer Camera Housing Conceptual Design 2h

        The ITER bolometer provides an absolutely calibrated measurement of the radiation emitted by the plasma which is a part of the total energy balance. The development is especially challenging because of the extreme environmental conditions within the vacuum vessel (VV) during plasma operation. The bolometer has to guarantee reliable measurements within an environment characterized by high neutron flux as well as temperatures exceeding 200 °C. In addition to the thermal loads the bolometer body is exposed to the mechanical loads caused by electromagnetic forces during transient events called disruption.
        This paper describes a possible procedure for a structural analysis of the bolometer camera body. To examine all important structural properties of the bolometer body, a multiple nonlinear finite element model based on a CAD conceptual design, has been generated. Subsequently, a transient mechanical analysis has been performed using the finite element code ANSYS. The input for the analyses was generated by a general electromagnetic model, taking into account the contribution of all structural parts and electromagnetic loading starting with the DINA code. From the wide range of DINA results the worst case load scenario has been chosen.
        This analysis enables the study of the response of the bolometer camera structure to the dynamic excitation caused by electromagnetic forces during the plasma disruption. Moreover, the influence of the different design solutions and different material properties has been investigated. Due to dynamic oscillating excitation by electromagnetic forces the whole bolometer structure is thoroughly shaken, as demonstrated by the results of deformation analyses. The analyses results will be used to validate the design against the required structural integrity during worst-case scenarios and verify the reliability of the bolometer camera design during operation. Finally, this analysis provides the results needed to perform the fatigue analysis of the VV bosses supporting the bolometer camera.

        Speaker: Dr Nikola Jaksic (ITZ, Max-Planck-Institute for Plasmaphysics)
      • 11:00
        On the path towards a Metal Foil Pump – Latest results and new experimental facility 2h

        The current design baseline for the EU DEMO implements the KALPUREX process for the fusion fuel cycle. This process aims to reduce the tritium inventory by separating hydrogen from other gases within the tokamak building and feeding it back to the matter injection system. The best candidate for the hydrogen separation unit close to the torus is a metal foil pump that relies on the effect of superpermeation. For a comprehensive investigation the dedicated HERMES setup at KIT was successfully exploited during the past years. This setup was used recently to demonstrate remarkable improvements in performance. However, due to the limits of operation of this setup a complete redesign was indispensable. Consequently, an improved facility was designed and assembled to overcome the previous limits. This HERMES plus facility allows an extended pressure range for operation due to a different plasma source as well as an advanced metal foil module. This modification aims to fill the missing gap between the operation pressure requirements for use in DEMO and the capability of previous setups at KIT and in literature.
        In this paper two aspects will be presented. On one hand the latest results of superpermeation, achieved with the previous setup, will be shown, highlighting the progress in performance and understanding. On the other hand the new facility will be introduced, demonstrating the motivation of this evolution, the current commissioning status and finally the expected outcome.

        Speaker: Dr Stefan Hanke (Institute for Technical Physics - Vacuum Department KIT - Karlsruhe Institute for Technology)
      • 11:00
        Optimal configuration of a tokamak fusion system with breeding blanket based on ITER TBM 2h

        System parameters and the optimal radial build of a tokamak fusion system with a normal aspect ratio were found through the coupled analysis of a tokamak system and neutron transport. Neutron impact on shielding and tritium breeding capability are self-consistently incorporated together with plasma physics and engineering constraints in determining the radial builds. The plasma physics and engineering constraints moderately extrapolated from the constraints adopted in the design of International Thermonuclear Experimental Reactor (ITER) were used. In a tokamak fusion system with a normal aspect ratio, configuration with only an outboard breeding blanket does not satisfy requirement on tritium self-sufficiency. The optimum system size to produce a given fusion power was determined by the requirements on the shielding, tritium breeding and the magnetic field at the toroidal field (TF) coil. Thickness of the outboard and inboard blanket were determined to allow optimal system size. With a confinement enhancement factor H = 1.3, Q > 30 was possible for fusion power greater than 2,000 MW with an aspect ratio of A = 3.0; however, Q > 30 was impossible with an aspect ratio of A = 4.0. The tritium breeding capability of blanket concepts proposed for testing in the International Thermonuclear Experimental Reactor (ITER) was evaluated by varying the outboard/inboard blanket thickness and the degree of lithium-6 (Li-6) enrichment. Cases with a smaller aspect ratio exhibited better performance since the number of fusion neutrons that contributed to tritium breeding were larger than the case with a larger aspect ratio. Among the blanket concepts, a helium (He)-cooled solid breeder (HCSB) concept showed the best tritium breeding capability and thus allowed for a smaller system size.

        Speaker: Bong Guen Hong (Department of Quantum System Engineering Chonbuk National University)
      • 11:00
        Optimal design of DMA probe for austenitic stainless steel weld of CFETR vacuum vessel 2h

        As to the ultrasonic testing of argon arc seam of 50mm austenitic stainless steel China Fusion Engineering Test Reactor(CFETR) vacuum vessel mock-ups, there are some limitations if we adopt the traditional ultrasonic probe or linear array phased array probe. In this paper, we designed a Dual Matrix Array(DMA) probe based on the CIVA, and then analyze the optimal principle of the probe parameters. The results show that the DMA probe’s signal to noise ratio much higher compared to the traditional probe, and the surface blind area is reduced . The defect detection rate meets the requirements of relevant standards. The research result has much reference value for the application of phased array ultrasonic testing in austenitic stainless steel welding joint.

        Speaker: rui wang (Hefei Juneng Electro Physical High-tech Development Co.Ltd)
      • 11:00
        Optimization and adjustment of impact set-up for testing of insulated pads of ITER blanket module connectors and first wall 2h

        Insulated pads are used on ITER blanket module connectors and the first wall; their main insulating function is to break any current loop between the shield block and vacuum vessel and/or between the first wall and shield block. The design of the pads consists of a cylindrical or prismatic body manufactured from NiAl-bronze, a ceramic insulating coating (Al2O3 or MgAl2O4) which is applied on interfacing surfaces of the pad body and on the shield block or first wall beam. The pads work in ultra-high vacuum conditions and at elevated temperature 100 – 300 оС. Static and dynamic loads up to 2 MN can act on the insulated pads during various plasma events in ITER.
        The qualification program for the insulated pads consist of cyclic and impact tests. The tests are performed for two categories of loading anticipated from plasma events. The number of cycles, force and energy are different for each category and therefore the qualification takes into account the worst case conditions.
        The impact tests are performed on a weight drop test bench, which has been designed and manufactured in JSC “NIKIET”. A measurement system for the test bench consists of specially designed force sensors and accelerometers.
        The description of the pad design and the impact test set-up, the results of analysis for justification of the set-up parameters, and the results of the test bench commissioning are presented in this paper.

        Speaker: Dr Ivan Poddubnyi (Joint-Stock Company “N.A. Dollezhall Research and Development Institute of Power Engineering”)
      • 11:00
        Optimization of GEM based detector structure aimed at plasma soft−semi hard X-ray radiation imaging 2h

        The proposed work refers to the development of gaseous detectors for application at tokamak plasma radiation monitoring. Soft−semi hard X-ray region radiation measurement of magnetic fusion plasmas is a standard way of accessing valuable information on particle transport and magnetic configuration.
        In this work, Gas Electron Multiplier (GEM) based imaging technique is proposed to perform advanced imaging, capable of photon energy discrimination, which can reach a very accurate spatial and temporal resolution and can provide lots of information on radiation, temperature and impurity distribution, MHD, etc., including data processing on the fly (in real-time) usable for plasma control purposes.
        The work will highlight the latest development and optimization on GEM detector structure for utilization in soft−semi hard X-ray radiation detecting system. The photon sensitive configuration is based on triple GEM amplification structure followed by the pixel readout electrode. The efficiency of detecting unit is adjusted for the radiation region of about 2-15 keV. The present work will introduce the preliminary laboratory results on the detector characteristics obtained for the constructed detecting chamber based on newly designed and developed GEM foils. The operational characteristics and capabilities of the detector will be compared with the ones based on standard (commonly used) copper GEM foils. First laboratory tests will be done by means of tuned X-ray laboratory source (2-9 keV), commercial X-ray tube (10-13 keV) and 55Fe iron sources presenting detector’s imaging capabilities. Stream-handling data acquisition mode will be applied for the data acquisition with timing down to the ADC sampling frequency rate (~13 ns) which allows uncovering the invaluable physics information about plasma dynamics due to excellent time resolution. The spatial resolution and imaging properties of this detector will be studied in this work for conditions of laboratory moderate counting rates and high gain.

        Speaker: Dr Maryna Chernyshova (Institute of Plasma Physics and Laser Microfusion)
      • 11:00
        Optimization of high heat flux components for DIII-D neutral beam upgrades 2h

        Upgrade of the DIII-D neutral beams leads to enhanced heat loads on many components, such as pole shields, calorimeter and collimator. Higher power is now desired for the neutral beams, increasing from 2.6 MW to 3.2 MW per source leading to a normal heat flux loads of up to 55 MW/m2 for the calorimeter. Original designs experienced local melting and fatigue cracks during operation at 2.6 MW. The Princeton Plasma Physics Laboratory is responsible for the design and manufacturing of the upgrades of these components.
        Heat flux distribution on neutral beam components is very uneven and leads to significant thermal stresses. High heat flux density impact requires surface optimization to reduce surface heat flux projection, and avoid localized melting. Several new design features were introduced to accommodate increased heat loads, such as molybdenum inserts for the pole shields, two-dimensional shaping for the calorimeter, and three-dimensional shape optimization and replaceable copper inserts for the collimator. Additionally, all three components include an optimized cooling system design featuring peripheral cooling of copper components. This takes advantage of the metal’s high thermal conductivity to achieve cool down between the pulses, and, simultaneously, avoids high thermal stress situations where cooling channels are located close to the high heat affected zones.
        The optimization process included applying analytical relations for the transient temperature distributions on the high heat flux components. These relations were confirmed by previous DIII-D experimental results. To validate the designs, numerical simulations were performed using ANSYS software and consisted of two stages: transient fluid flow simulation in conjunction with heat transfer analysis in the solid parts; and structural analysis using the temperature distribution obtained at the first stage. Special functions were introduced to include complex heat flux effects on the 3D surfaces. Results of the design optimization and numerical simulations will be presented.

        Speaker: Dr Andrei Khodak (Princeton Plasma Physics Laboratory)
      • 11:00
        Optimization of RFX-mod2 gap configuration by estimating the magnetic error fields due to the passive structure currents 2h

        The design of a major refurbishment of the toroidal complex of the RFX-mod experiment is going to be finalized before starting the realization phase. The Inconel vacuum vessel will be removed and the stainless steel supporting structure will be modified so as to become vacuum tight. The plasma facing graphite tiles will be mounted onto the inner surface of the copper shell, allowing an increase of the plasma proximity factor. This could allow different operation regimes expected to provide a significant reduction of the amplitude of RFP tearing modes. This reduction would lead to magnetic chaos mitigation with confinement improvement, better mode control capability and reduced plasma wall interaction. On the other hand, due to the shorter distance from the passive structures, the plasma will be even more sensitive to magnetic field errors produced by the induced currents at the cuts of the same structures. A careful assessment of different gap configuration is mandatory to optimize the final design of the machine modifications. Numerical analyses confirmed by the following experimental results showed that overlapping layers at the shell poloidal gap were an effective solution to minimize error fields in RFX-mod. This suggested analysing the possibility of a similar configuration at the shell inner equatorial gap, too. The direct reproduction is not straight-forward due to strict dimensional constraints and the required compatibility with a much more complex assembly procedure necessary to make the support structure vacuum tight. Parametric analyses of different constructive solutions have been made possible by a computational tool specialized to estimate the induced currents in thin conducting structures with complex geometry and the associated magnetic fields. The magnetic error fields due to other new features of the design such as the conducting cage introduced to assure the electrical equipotentiality of the graphite tiles have also been investigated.

        Speaker: Dr Giuseppe Marchiori (Consorzio RFX)
      • 11:00
        Optimization of single crystal mirrors for ITER diagnostics 2h

        Diagnostic mirrors are planned to be used as plasma-viewing optical elements in all optical and laser-based diagnostics in ITER. Degradation of mirrors due to e.g. deposition of plasma impurities will hamper the entire performance of affected diagnostics. In situ mirror cleaning by plasma sputtering is presently envisaged for the recovery of optical reflectivity of contaminated mirrors.
        Previous studies have demonstrated sound advantages of single crystal mirrors, outlining their ability to withstand plasma sputtering without noticeable degradation of optical reflectivity. At the same time, there are studies made on polycrystalline substrates, showing a signature of sputtering dependence on crystal orientation. Should such a dependence exist, the sputtering of single crystals could be minimized, thus prolonging a mirror lifetime in ITER.
        Four single crystal (SC) molybdenum (Mo) mirrors with different crystal orientation were produced to study the effect of crystal orientation on sputtering. Mirrors were exposed to steady-state argon plasma in linear plasma device PSI 2 under identical plasma conditions. The energy of impinging ions was about 60 eV. Mirror temperature was 250oC. Sputtering conditions in PSI 2 were corresponding to those expected inside the mirror cleaning system of ITER. The average amount of sputtered mirror material was about 1200 nm corresponding to about 120 mirror cleaning cycles.
        Plasma exposures did not affect the optical performance of all mirrors. The maximum measured decrease of specular reflectivity did not exceed 5% in the ultraviolet range at the wavelength of 250 nm. No increase of the diffuse reflectivity was detected. The single crystal mirrors with orientations [110]/[101] demonstrate at least 25% less removal of sputtered material than mirrors with other crystal orientations. Summary of results, their analysis will be presented along with a feasibility assessment of the use of optimized mirrors in ITER.

        Speaker: Dr Andrey Litnovsky (Institute of Energy and Climate Research, Forschungszentrum Juelich GmbH (FZJ))
      • 11:00
        Overview of the JET Operation Reliability 2h

        The JET tokamak has been in operation since 1983, producing ~92500 pulses so far. For the period 2000 to 2016 (not including DTE1 in 1997), information on every shutdown, commissioning phase and experimental campaign has been logged, providing unprecedented operation reliability statistics and a model for studying reliability, availability, maintainability and inspectability (RAMI) in fusion experiments such as ITER.
        The JET Operation Reliability Statistics record the time taken to install major upgrades, commission systems with plasma, the downtime for maintenance and the achieved versus planned experimental campaign days. On average, JET achieves 85% of the planned experimental days and up to 180 campaign days during a calendar year.
        For unplanned interventions the systems that failed are identified, the impact on machine availability is logged and the remedial actions taken are documented. A total of 12 unplanned interventions were recorded over 17 years with durations in the range 2 to 8 weeks.
        Delays during operations are recorded, providing a detailed overview of faults from essential sub-systems such as computer systems, power supplies and neutral beam systems and delays attributed to human operators. On average 21% of the time is lost due to delays during experimental campaigns. This has been consistent over the last 17 years and is compensated by having 20-25% contingency for campaign planning.
        Maintenance schedules and refurbishments are planned at JET, using the operation reliability data. Recently, pulsed power supplies and heating systems were refurbished, enabling continued high operational availability of JET for the upcoming campaigns using deuterium-tritium mixtures in 2019-2020.

        Speaker: Dr Adrianus Sips (JET Exploitation Unit, European Commission)
      • 11:00
        P4.144 Thermal-hydraulic analysis for first wall and vacuum vessel thermal shield of Divertor Tokamak Test facility 2h

        The Divertor Tokamak Test (DTT) machine has been proposed by ENEA, in collaboration with other Italian institutions, to investigate power exhaust solutions with an experiment integrating all DEMO relevant physics and technology issues. The DTT machine will be able to host, in different phases of its life-time, advanced divertor magnetic configurations (snowflake, super-X, double null) and liquid metal solutions, able to withstand the large loads expected in the DEMO fusion power plant. The first wall (FW) has been designed with stainless steel cooling pipes coated with a W layer deposited by plasma spray technique. It shall be compatible with liquid metal divertors and therefore be heated at a temperature of 300 °C to avoid the liquid metal condensation. In this case it is necessary to consider gas rather than water cooling. To minimize the heat transferred to the superconductive coils the vacuum vessel (VV) is kept at 100 °C with water and a Vacuum Vessel Thermal Shield (VVTS) cooled with helium at 70K is interposed between the VV and the coils. In this work the preliminary results of the thermal-hydraulic analysis carried out with ANSYS Workbench software are presented for both the FW, using CO2 as coolant, and the VVTS.

        Speaker: Fabio Maviglia (Department of Economics Engineering Society and Businesses University of Tuscia)
      • 11:00
        Paschen testing of ITER Central Solenoid qualification module 2h

        General Atomics is currently fabricating superconducting magnet modules for ITER Central Solenoid in its Poway, CA facility. A critical step during final testing of the modules is high voltage checks of the insulation in Paschen conditions. A qualification coil was fabricated using the same techniques and equipment as the CS Modules. The qualification coil insulation was tested at voltages up to 30kV at pressures from 10-3 mbar to atmosphere to validate the CS Module insulation design and Paschen testing equipment.
        Using the same vacuum chamber for cooling the coil to 4.5K, Paschen testing was performed utilizing a system to localize discharges without venting the vacuum chamber. Discharge events were detected using a high voltage tester and specially designed in-chamber multi-camera system.
        The initial protection of the ends of the instrumentation wires failed and modifications were made to prevent Paschen discharges at the coil instrumentation wire ends.
        Further testing revealed breakdown in the hand-wrapped area of the piping over the helium pipe weld where the instrumentation wires exited the ground insulation. Two modes of failure were identified: failure of the wire insulation (cracking) and failure of the ground insulation. The insulation scheme was redesigned and requalified. The final liquid helium pipe joint insulation scheme consisted of Glass-Kapton® (GK) tape wrapped with epoxy resin. Methodologies were developed to fill “cusps” between round wires and round pipes and to eliminate a problem of wire insulation cracking after resin curing. This insulation scheme was implemented on the qualification coil, which then passed 30kV in all prescribed Paschen conditions.
        This paper describes the qualification coil Paschen test results, development of Paschenization techniques to be used on ITER CS Modules, development of the insulation joints, and method of extracting instrumentation wires on the helium supply and return pipes while maintaining high voltage standoff in Paschen conditions.

        Speaker: Dr Kenneth Khumthong (ITER Projects / Energy & Advanced Concepts, General Atomics)
      • 11:00
        Path planning and space occupation for remote maintenance operations of transportation in DEMO 2h

        The ex-vessel Remote Maintenance Systems in the DEMOnstration Power Station (DEMO) are responsible for the replacement and transportation of the plasma facing components. The ex-vessel operations of transportation are performed by overhead systems or ground vehicles. The time duration of the transportation operations has to be taken into account for the reactor shutdown. The space required to perform these operations has also an impact in the economics of the power plant.

        A total of 87 trajectories of transportation were evaluated, with a total length of approximately 3 km. The total occupancy volume is, comparatively, between 21 and 45 Olympic swimming pools, depending mainly to the type of transportation adopted in the upper level of the reactor building. Taking into account the recovery and rescue operations in case of failure, the volume may increase up to, between 43 and 64 Olympic swimming pools. The estimation of the total time duration of all expected transportation missions in the reactor building are between 166 hours (7 days) and 388 hours (16 days). This time estimation does not include docking, accelerations or other operations that are not transportation. The travel speed is assumed constant with a maximum value of 20cm/s (the same value assumed for Cask and Plug Remote handling System in the International Thermonuclear Experimental Reactor - ITER).

        The results achieved in this preliminary assessment will help the design process to optimize the time duration of the reactor shutdown and the layout of the DEMO power plant.

        This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission.

        Speaker: Dr Alberto Vale (Instituto de Plasmas e Fusao Nuclear, Instituto Superior Técnico)
      • 11:00
        Plan and progress of the fusion neutron sources at KAERI for fusion and fission applications 2h

        According to the National Fusion Energy Program in Korea, Volumetric Fusion Neutron Source (temporarily called, V-FNS) has been planned and Compact Fusion Neutron Source (temporarily called, C-FNS) development was started at KAERI, which can be used in the fusion and also the fission/industrial applications such as radiotracing isotope production, radiography, and so on, in which the various targets were considered in parallel. For developing the C-FNS, plasma generator based on RF ion source including RF driver/generator was developed and target has been designed, fabricated, and tested with the high heat flux test facility (KoHLT-EB) before assemblying with the ion beam for investigate its integrity under the heating condition. Test conditions were prepared with the preliminary analysis and compared with the test results. From this result, the design was optimized and optimized target design was proposed. So far, the C-FNS is successfully developed, and the preparation of the on-site C-FNS and further V-FNS design in detail is in progress.

        Speaker: Dong Won Lee (Nuclear Fusion Engineering Development Division Korea Atomic Energy Research Institute (KAERI))
      • 11:00
        Plasma light detection in the SPIDER beam source 2h

        The ITER Heating Neutral Beam (HNB) injector RF plasma source is required to generate a 40A D- or 46A H- ion current, with low electron/ion ratio (<1) and high uniformity over the extraction area (800 mm x 1600 mm). The source prototype SPIDER in the Neutral Beam Test Facility at Consorzio RFX has been developed to demonstrate these performances and it is now under final installation and commissioning.
        A set of diagnostics in SPIDER has to characterize the complex behaviour of the plasma in the source, assisting in the optimization of performances and providing the signals for machine protection during operation. To monitor the visible radiation emitted by the plasma in the source a set of photodiodes measures the time evolution of lines of sight integrated intensity; some are equipped with interference filters to select specific spectral lines of particular interest, like of cesium to control its evaporation, which assists the H- production, and of metal impurities lines, which are sign of severe erosion of the source inner wall or even of melting at hot spots. Other lines of sight are fed to grating spectrometers to measure the spectrum at a slower rate, either for low resolution spectral survey of hydrogen Balmer lines and impurities and for high resolution molecular Fulcher bands.
        This contribution focuses on the implementation of the photodiode measurements and on preliminary results during the commissioning phase. The overall diagnostic layout is presented. The custom photodiode electronics is described, which includes a low noise variable gain amplifier, FPGA based remote control and suitably shaped digital output for interlock purposes.

        Speaker: Dr Roberto Pasqualotto (Consorzio RFX)
      • 11:00
        Platinum suported on graphene - PTFE as catalysts for isotopic exchange in a detritiation plant. Manufacturing and physical and microstructural analysis 2h

        The catalytic separation of hydrogen isotopes is of particular interest for nuclear industry from the point of view of tritium recovery and its use in fusion reactors. Isotopic exchange may take place in the homogeneous (gaseous) phase or in the heterogeneous phase (hydrogen or gaseous deuterium and water or liquid heavy water). Catalysts are necessary both for the homogeneous phase reaction and the heterogeneous reaction.
        Recently, graphene and graphene oxides have been investigated as support material due to their excellent physical and chemical properties. The high specific surface, their thermal and chemical stability, high mechanical strength make graphene a potential component in the development of new catalysts that could be used in isotopic exchange. The use of graphene and graphene oxides as support for the active metal has shown an improvement in catalytic activity when compared to the conventional support, degasolination carbon. This behaviour may be explained by the high dispersion of active metal on the surface of graphene.
        The paper presents the preparation of some catalysts that will be used in the future catalytic isotopic exchange experiments (in the beginning VPCS in order to understand its efficiency in isotopic exchange then the catalysts will be integrated into a LPCE-type laboratory installation). The catalysts thus prepared will be characterised from the microstructural point of view for a comparison with existing catalysts. BET, SEM, TGA and XRD will be used for the physical and microstructural analysis of catalysts.

        Speaker: Felicia Vasut (National Research and Development Institute for Cryogenic and isotopic Technologies - ICSI Rm. Valcea)
      • 11:00
        Possibility study of the partial neutron calibration for neutron flux monitors in torus devices 2h

        The absolute calibration of the detection efficiency for the total neutron yield in the whole plasma is one of the most important issues in the neutron diagnostics such a neutron flux monitor (NFM). In many magnetic confinement devises, those neutron detectors are calibrated by moving or rotating a neutron source such as a Cf-252 radioactive source or a compact neutron generator on the magnetic axis in the vacuum vessel. The in-situ calibration work needs a long time and a big effort. Therefore, we consider the possibility of the partial neutron calibration, where the neutron source is located in the vacuum vessel only near the detector or limited number of the toroidal positions, from the in-situ neutron calibration data of the Large Helical Devise (LHD), JT-60U and the other tokamaks. The partial calibration is very useful to estimate the detection efficiency prior to the full calibration, especially on ITER, and might be a final calibration instead of the full calibration. In the case of the NFM using a U-235 fission chamber located outside of the vacuum vessel on the equatorial plane, larger than 90% of the detector counts are contributed by neutrons in the toroidal angle range of -60°~ +60°. The detection efficiency for the neutron source located one toroidal position on the magnetic axis (called point efficiency) decreased exponentially with the absolute toroidal angle of the neutron source. We found that a partial calibration in the toroidal angle range of -60°~ +60° combined with exponential extrapolations to -180° and +180°can estimate total detection efficiency within 5% uncertainty. In the case of the point efficiency measurement at the limited number of the toroidal locations, typically 10-20 locations, the total detection efficiency for the whole plasma can be estimated within 5% uncertainty by the interpolation using the MCNP simulation result.

        Speaker: Prof. Takeo Nishitani (National Institute for Fusion Science)
      • 11:00
        Post-Mortem Analysis of ITER CS Helium Inlets Fatigue Tested at Cryogenic Temperature 2h

        In the ITER Magnet System, ten thousand tonnes of superconducting cable – in – conduit - conductor (CICC) are cooled down by a forced flow of supercritical helium, which is supplied from helium inlets. For the ITER Central Solenoid (CS), consisting of six independent pancake wound modules, the He inlets consist of three overlapping holes covered by an oblong shaped boss, welded to the CS jacket through full penetration, multi-pass Tungsten Inert Gas (TIG) welding. There are 120 CS He inlets and because they are located in a region of high cyclic tensile stresses, i.e. first turn at the inner diameter of the pancake, the CS inlets are one of the most critical structural components.

        Qualification of the design is done by analysis and a comprehensive design optimization has been performed by finite element (FE) analyses. In order to guarantee the required fatigue life at cryogenic temperature of these component, a post – welding process consisting in ultrasonic shot – peening is required.

        Based on a qualified weld procedure, six mock – ups including each two He – inlets on the opposite surfaces have been produced to run a mechanical fatigue testing program at cryogenic temperature with sufficient statistical significance to validate the findings of the FE simulations. Five were peened. One not peened.

        The paper describes the results of a comprehensive post – mortem failure analysis which includes non – destructive (penetrant testing, leak testing, computed tomography) as well as destructive examinations (microoptical and hardness tests, scanning electron microscopy). It also includes a full assessment of the welds according to the most stringent acceptance levels of the standards in force.

        The views and opinions expressed herein do not necessarily reflect those of the ITER Organization.

        Speaker: Dr Ignacio Aviles Santillana (European Organization for Nuclear Research (CERN))
      • 11:00
        Post-test examination of a Li-Ta heat pipe exposed to H plasma in Magnum PSI 2h

        The authors exposed a radiatively cooled, ~195-mm-long, lithium-filled tantalum heat pipe (HP) to a hydrogen plasma in DIFFER’s linear plasma source Magnum PSI continuously for ~2 hours. We kept the overall heat load on the inclined HP constant, varied the tilt and peak heat flux to ~2.5 MWm2. The HP operated at ~1000-1100 C. Diagnostics included near infra-red thermography from two orthogonal ports. [1]

        We stopped after an initial breach occurred near the beam axis intercept. ~0.06 g of lithium formed a 6-mm-diameter nodule and coating covered half of the area wetted by the beam. Several seconds later, a transverse crack opened; lithium flowed out quickly and wetted an area ~30 mm2. Fractography and post-test metallography showed differing breach mechanisms. The first site had prior material damage. The second breach occurred as the HP cooled. With creep and relaxation during the exposure, the site would have tensile loading during cooling. A tantalum disk annealed in a hydrogen furnace was available for comparative evidence of hydrogen embrittlement.

        Sandia had purchased an existing tantalum HP from Aavid-Thermacore, Inc. The test showed prolonged operation, gave useful data and we judged it a success. However, tantalum would not be the right choice for a future PFC. The paper also discusses HP configurations and materials for future PFCs.

        [1] G.F. Matthews, R. Nygren, T. Morgan, S.A. Silburn, “Demonstration of the potential for exchangeable PFCs based on radiatively cooled lithium heat pipes in Magnum PSI,” this conference.

        Sandia National Laboratories is a multi-mission laboratory managed and operated by National Technology and Engineering Solutions of Sandia, LLC., a wholly owned subsidiary of Honeywell International, Inc., for the U.S. Department of Energy’s National Nuclear Security Administration under contract DE-NA0003525. This work was also part-funded by the RCUK Energy Programme under grant EP/I501045.

        Speaker: Dr Richard Nygren (Sandia National Laboratories)
      • 11:00
        Practical Implementation within the Electron Cyclotron Upper Launcher of the French INB Order of 2012 2h

        The ITER project is being undertaken at Cadarache, France, to construct and operate an experimental nuclear fusion facility. The aim of this paper is the description of the implementation of the French Order of February 7, 2012, concerning Basic Nuclear Installation (also called “INB”) within the European Union Domestic Agency (EU-DA), specifically on the Electron Cyclotron Upper Launcher (EC UL).

        The EC UL will be used for the control and heating of the ITER plasma. The launcher includes in-vessel items (nuclear shielding, blanket module, port plug structure and optics) and ex-vessel first confinement items (diamond window, isolation valve, waveguides, miterbends, tapers and port plug back end).

        According to the general rules of the French order, ITER Organization (IO), being the nuclear operator, has to monitor all activities related to the Design, Construction, Operation, Maintenance, Final shutdown and Dismantling of Nuclear facilities during their full life cycle at ITER. The EU-DA, as a tier 1 supplier of IO, applies a Requirements Management and Verification (RMV) process in order to track, control and verify all technical requirements applicable to EC UL components. EU-DA has duties regarding the compliance with the requirements propagated in his full supply chain performing Protection Important Activities (PIA). The methodology of this process will be illustrated for different EC UL Protection Important Components (PIC). The paper will go in more detail on the Nuclear Pressurized Equipment classification (ESPN) of the EC UL and its cooling circuits. The nuclear safety demonstration, traceability, validation of methods, qualifications, prototype tests, calculations and modelling, will be described with specific examples of EC UL.

        ITER Disclaimer: The views and opinions expressed herein do not necessarily reflect those of the ITER Organisation.

        Speaker: Dr Paul Wouters (ITER, Fusion For Energy)
      • 11:00
        Prebaking of T-15MD vacuum vessel 2h

        Presently, the Tokamak T-15MD is being built in the NRC “Kurchatov Institute”. Vacuum vessel was manufactured and passed the preliminary vacuum tests at the plant in St. Petersburg (Efremov Institute) in 2016. Vacuum vessel consists of toroidal shell made of a 321stainless steel of 5 mm and 8 mm thick, horizontal and vertical ports (152 in total), in-vessel elements. The chamber volume is 47 m3 and a surface square faced to plasma is ~ 200 m2. The purpose of vacuum vessel prebaking is the checking of quality of the numerous welds and a workability of the control system. To bake the vacuum vessel up to 220°C at the plant in Bryansk, the ohmic heaters (a single 1.2-mm diam. Ni-Cr alloy wire, housed inside a stainless-steel 6-mm diam. shell, with a magnesium oxide ceramic insulator) have been laid on vessel shell surface both outside and inside. The thermal insulation (cases with mineral wool) closed the vessel surface outside. The surface temperature is controlled by thermocouples. The currents through the heaters are regulated by means of control system. The temperature data processed and stored by means of the data acquisition system. The results of vacuum vessel baking are presented.

        Speaker: Dr Aleksandr KHVOSTENKO (NRC Kurchatov Institute)
      • 11:00
        Preliminary accident analysis of ex-vessel LOCA for the European DEMO HCPB blanket concept 2h

        Based on the reference design HCPB2016 (helium cooled pebble bed) in the pre-conceptual design studies for the European DEMO, the primary heat transfer system (PHTS) for DEMO baseline 2015, and current parameter study for the plasma disruption conditions and the affected FW surface areas, ex-vessel LOCA (loss of coolant) with a double-ended guillotine break of a main pipe in the PHTS has been investigated that helium blows down into the tokamak cooling room (TCR). For the design basis accident (DBA) a fast plasma shutdown (FPSS) followed by a plasma disruption is assumed at 3 s after the detection of the LOCA at 80% of the nominal mass flow rate. Three main cases are identified: case I with the affected FW area of 0.1 m² in one loop and the mitigated plasma disruption, case II with 1.0 m² in two loops and the mitigated plasma disruption, and case III with 5.0 m² in two loops and the unmitigated plasma disruption. If EUROFER reaches 1000 °C an in-vessel LOCA takes place. Since this LOCA starts in case of the beyond design basis accident (BDBA) without the FPSS much earlier than it in the DBA, to save the computation time, the transport of source terms is performed for the BDBA. Also scenarios due to the options of the suppression tank (ST) with or without water as heat sink, and impact of the cooling ability of the vacuum vessel (VV) have been investigated. MELCOR 1.8.6 for fusion is a valid tool for this study. The ex-vessel LOCA is initialized during the normal operation at the steady state. The transient results of different scenarios will be discussed for the time evolution of the accident sequences, pressurization in the TCR, VV and ST, temperature behavior in different volumes and structures, and tritium and dust transport behavior.

        Speaker: Xue Zhou Jin (Institut fŸr Neutronenphysik und Reaktortechnik (INR) Karlsruhe Institut fŸr Technologie (KIT))
      • 11:00
        Preliminary investigation on W foams as protection strategy for advanced FW PFCs 2h

        Among the eight core missions towards the realization of nuclear fusion, a future reactor must ensure efficient and safe power exhaust through the divertor and First Wall (FW). The greatest challenges arise from the occurrence of plasma transients. A simulation of a DEMO-like FW Plasma Facing Component (PFC) was carried out assuming Vertical Displacement Event (VDE) and ramp-up limiter conditions. The results highlighted an extreme heat flux impinging the thin tungsten (W) armour produced by Plasma Vapour Deposition (PVD). As a consequence, localized surface vaporization, melting and re-solidification may occur. Vapour shielding and surface melting play an important role in terms of heat flux reduction, but the failure of the component may occur owing to the high thermal stress and cracking developed, which can compromise safety and prompt return to normal operation. A possible protection strategy is to provide the plasma a sacrificial structure able to promote the flux reduction while preventing the substrate from the failure. W-based refractory foams are a viable solution owing to their peculiar mechanical and thermal management capabilities, that can be tailored by choosing an optimum value of relative density and porosity. However, a proper FE modelling is challenging due to their complexity and anisotropy. Nevertheless, a recent study confirmed that the mechanical behaviour of an existing open cell foam is achievable through the calibration of an equivalent FE model. The present investigation aims to verify the feasibility of W based open cell foams as a possible sacrificial material for FW protection strategy. Thermo-mechanical analyses have been carried on the equivalent FE model of foam exposed to heat loads expected in VDE and limiter conditions. As a consequence, optimal values of relative density and porosity of foam are proposed to achieve the required protection capabilities.

        Speaker: Dr Riccardo De Luca (Department of Economics, Engineering, Society and Business Organization (DEIm), University of Tuscia)
      • 11:00
        Preliminary RAMI assessment for ITER test blanket module ancillary systems 2h

        A RAMI (Reliability, Availability, Maintainability and Inspectability) assessment performed on the ITER Test Blanket Module ancillary systems is presented. The assessment is aimed at evaluating design criticalities possibly jeopardizing the achievement of the overall 75% availability requirements for the considered ITER plant. The Ancillary systems of the European Test Blanket Systems for ITER here analysed are the helium cooling systems (HCSs) of both the Helium Cooled Lithium Lead (HCLL) and Helium Cooled Pebble Bed (HCPB), lithium-lead loop of the HCLL TBM and the Tritium Extraction System (TES) of the HCPB TBM.
        Reliability and availability performance were assessed by means of reliability block diagrams (RBDs) over a foreseen mission of 11 days and 20 years respectively, with operating cycles reflecting ITER schedule. In particular, a set of events leading to unavailability of the systems was initially defined by means of a failure mode and effect analysis. Then RBDs were implemented according to reliability-wise integration of the components included in such systems. Different RBDs were defined depending on the component judged to impact normal operation or start-up operation mode.
        Systems analysis was performed exploiting a modular approach in order to elicit the relative contribution of specific components to the system availability so to support possible design improvement by highlighting critical sub-systems. In particular the cooling, tritium extraction and trapping functions were assessed separately for HCLL and HCPB concepts.
        Both HCLL and HCPB ancillary systems presented design resulted able to achieve the availability and reliability targets with performance ranging from 83% up to 99% for reliability at 11 days and from 89% up to 99% for mean inherent availability. Finally the integrated impact of all ancillary systems on overall tokamak operational availability was estimated and the assessed systems, also considering the current preliminary level of design development, appear able to match expected performance.

        Speaker: Danilo Nicola Dongiovanni (FNS-TEN ENEA)
      • 11:00
        Preliminary structural assessment of the HELIAS 5-B breeding blanket 2h

        The European Roadmap to the realisation of fusion energy, carried out by the EUROfusion
        consortium, considers the stellarator concept as a possible long-term alternative to a tokamak fusion
        power plant. To this purpose a pivotal issue is the design of a helical-axis advanced stellarator
        (HELIAS) machine equipped with a tritium breeding blanket (BB), considering the achievements
        and the design experience acquired in the pre-conceptual design phase of the tokamak DEMO BB.
        Therefore, within the framework of EUROfusion WPS2 R&D activity, a research campaign, aimed
        at the investigation of the structural behaviour of the HELIAS 5-B BB, has been launched at KIT in
        cooperation with University of Palermo. The scope of the research has been the determination of a
        preliminary BB segmentation scheme able to ensure, under the assumed loading conditions, that no
        overlapping may occur among the blanket regions. To this purpose, the Helium-Cooled Pebble Bed
        (HCPB) and the Water-Cooled Lithium Lead (WCLL) BB concepts, presently considered for the
        DEMO tokamak fusion reactor, have been taken into account.
        A 3D CAD model of a HELIAS 5-B torus sector has been adopted, focussing attention on its far end
        regions, namely the triangular and bean shape regions. Due to the early stage of the HELIAS 5-B BB
        R&D activities, the considered CAD model includes homogenized blanket modules without internal
        details. Hence, in order to simulate the features of the HCPB and WCLL BB concepts, equivalent
        material properties have been purposely calculated and assumed. Moreover, proper nominal steady
        state loading scenarios, based on the DEMO HCPB and WCLL thermomechanical analyses, have
        been taken into account.
        A theoretical-numerical approach, based on the Finite Element Method (FEM), has been followed
        and the qualified ANSYS v. 18.0 commercial FEM code has been adopted. The obtained results are
        herewith presented and critically discussed.

        Speaker: Gaetano Bongiovì (Institute for Neutron Physics and Reactor Technology Karlsruhe Institute of Technology)
      • 11:00
        Pressure tests supporting the qualification of the ITER EC H&CD upper launcher diamond window 2h

        The Electron Cyclotron diamond window which is located inside the port cell serves, together with an isolation valve, as primary vacuum boundary between the ITER vacuum vessel, the transmission lines and the atmospheric environment and it functions as confinement barrier. The window consists of an ultra-low loss Chemical Vapor Deposition (CVD) diamond disk brazed into a metallic housing and it has to guarantee the compliance with very stringent nuclear safety requirements and an adequate transmission capability for high power mm-waves (1.31 MW at 170 GHz). The design of the window unit is approaching its final phase including design validation analyses, the development of a dedicated qualification program and prototyping activities.
        In preparation of the testing of complete window prototypes, pressure tests were performed on a mock-up formed by a diamond disk (D = 80 mm, d = 1.11 mm) brazed to two copper cuffs. The scope of the experiments was to show both the capability of the joining between the disk and the cuffs to keep the required vacuum tightness as well as the integrity of the disk when exposed to pressure loads. Based on the requirements defined for the pressure scenarios during normal operation and off-normal events, a test program was developed accounting for cyclic tests at low pressure differentials and overpressure tests up to 2 bar pressure difference over the disk to simulate severe accident conditions.
        This paper gives an outline of the ongoing development of the overall qualification program of the ITER torus window and reports specifically on the experimental set-up and the successful outcome of the pressure tests of the brazed diamond disk mock-up. The test program for the complete window prototype needs discussion and approval by F4E and ITER and will directly determine the final qualification program for the diamond windows during series production.

        Speaker: Dr Sabine Schreck (IAM-AWP, KIT)
      • 11:00
        Progress in Development of ITER Diagnostic Pressure Gauges and Status of Interfaces with ITER Components 2h

        Neutral gas pressure is one of the main parameters for basic control of ITER operation. Diagnostic Pressure Gauges shall provide pressure measurements in the range from 10-4 Pa to 20 Pa with an accuracy of 20 % and a time resolution of 50 ms. In total 52 DPG sensor heads will be installed in 4 lower ports, 4 divertor cassettes and 2 equatorial ports. The overall DPG system has 15 interfaces with various systems and components of ITER.
        Within a Framework Partnership Agreement with F4E, IPP is developing the DPG system including sensor head and front-end electronics. A test campaign aiming at validation of the system baseline design is currently ongoing.
        Performance of the DPG sensor head has been investigated for various parameters of the prototype as electrode potentials, transparency of the acceleration grid and electron emission current. The test results demonstrate the possibility to measure pressures up to 30 Pa.
        Emission properties of several material candidates for the filament (hot cathode) have been studied during approximately three months of continuous operation. Filament samples made of Tungsten with 2 % doping of ThO2 and Tungsten alloy with 26 % of Rhenium coated with Y2O3 demonstrated most promising results. These cathodes required lowest heating currents for fixed emission current. Mechanical tests of these samples showed no considerable deformations for maximum heating currents in transient magnetic fields of up to 8 T.
        The already obtained as well as future test results will be used for further design optimization and integration of the DPG system into the ITER environment.

        Speaker: Dr Alexey Arkhipov (Max-Planck Institute for Plasma Physics)
      • 11:00
        Progress in high heat flux testing of European DEMO 2h

        In the framework of the DEMO divertor project of EUROfusion an extensive R&D program has been carried out to develop advanced design concepts for hot water cooled divertor targets. These plasma-facing components made of W blocks as plasma facing material and CuCrZr tubes as cooling tubes should allow a reliable DEMO operation for 2 h long pulses and maximum heat fluxes up to 20 MW/m². Compared to ITER, the operation at higher coolant temperature of 150 °C, the longer required lifetime, and the significantly higher neutron fluence are the design challenges exceeding the current extent of experience.

        In the pre-conceptual design phase, eight types of target mock-ups were designed and manufactured by the involved research groups. Finally, 25 of these unirradiated mock-ups were assessed by high heat flux (HHF) examination in the test facility GLADIS at IPP Garching. The applied hot water cooling at 130°C inlet and 16 m/s velocity ensures thermal conditions similar to the expected DEMO operation. To reduce the experimental effort of the HHF testing each individual mock-up was subjected to a screening test up to 20 MW/m² at 20°C cooling water inlet for first selection, followed by HHF fatigue tests between 10 and 20 MW/m². After the cold water HHF testing, six concepts were qualified for the subsequent hot water HHF testing with 300 cycles at 20 MW/m². Four of them were successfully loaded up to 500 cycles. Extensive diagnostics such as high-resolution infra-red and visible light cameras were employed for in-situ assessment of surface temperature evolution during the cyclic HHF loading.

        This contribution presents the summary of HHF testing and results from the post exposure investigation. Overall HHF performance of the various design concepts will be discussed together with other non-destructive investigations including ultra-sonic examination and transient IR thermography.

        Speaker: Dr Henri Greuner (Max Planck Institute for Plasma Physics)
      • 11:00
        Progress in the pre-conceptual CAD engineering of European DEMO divertor cassette 2h

        This paper presents the recent progress in the pre-conceptual design activities for the DEMO divertor Cassette Body, performed in the framework of the work package “Divertor” of the EUROfusion Power Plant Physics & Technology (PPPT) program. According to Systems Engineering Principles, the divertor CAD model is reviewed, considering the updates in the DEMO configuration model presented by the Programme Management Unit (PMU) in 2017. The design parameters affected by these changes and their impact on the divertor design and on the interfaced systems are analysed. Then, the paper focuses on the integration on the new cassette geometry of the divertor sub-systems. This includes the design of a “shielding liner” for cassette body and Vacuum Vessel protection, as well as the development of the cassette body-to-Vacuum Vessel fixation system. The design activities related to these main sub-systems are discussed in detail, in terms of CAD model and thermo- mechanical calculations.

        Speaker: Domenico Marzullo (Department of Industrial Engineering CREATE Consortium - University of Naples Federico II)
      • 11:00
        Progress in the production of the W7-X divertor target modules 2h

        The realization of the 19.6 m² highly heat loaded surface of the actively water-cooled divertor of Wendelstein 7-X (W7-X) requires the installation of 100 target modules distributed in ten discrete similar divertor units. A target module is made of target elements mounted onto rails joined by a stiffening plate forming a frame with an attachment system to the plasma vessel. The target modules are water-cooled from manifolds to distribute the water equally between the target elements with flanged connections to the water supply system. A target element is made of a CuCrZr copper alloy heat sink armored with carbon fibre reinforced carbon CFC NB31 tiles and designed to remove a stationary heat flux up to 10 MW/m² on its main area. The main challenge was to fit modules in the limited available space taking into account the 3-D shape of the plasma vessel, the neighboring in-vessel components and diagnostics. The assembly of the modules was carried out in the workshop of IPP-Garching. Special attention was given to the positioning of the individual target elements onto the supporting frame to avoid local heat loads or leading edges, and on the reliability of the orbital welding for the cooling circuits. The quality of the target modules was assessed as follows: visual inspections, measurement of the 3-D CFC surface, dynamic pressure tests, He leak testing under pressure at different temperature (20°C, 160°C) in vacuum oven, high heat flux testing. The paper presents the design, manufacturing process and the results of the quality assessment of the 30 first finished modules to be positioned in the horizontal part of the divertor.

        Speaker: Dr Jean Boscary (Institut Max Planck für Plasmaphysik)
      • 11:00
        Progress of the EU activities for the ITER Divertor Inner Vertical Target procurement 2h

        F4E undertook the qualification of so-called “Additional Suppliers” in order to enhance competition among the potential bidders and secure the procurement of the ITER Divertor Inner Vertical Target.
        In order to assess the performances of W armoured Plasma Facing Components under the conditions expected in the divertor target strike point region, a total of 36 W monoblock mock-ups were manufactured by ATMOSTAT (F) by Diffusion Bonding, and by CNIM (F) and Research Instruments GmbH (D) by brazing. 24 mock-ups were High Heat Flux (HHF) tested in IDTF (Efremov Institute Saint Petersburg, Russian Federation) electron beam test facility.
        The HHF testing program foresaw the performance of 5000 cycles at 10 MW/m2 and 300+700 cycles at 20 MW/m2 with 10 s power on and 10 s dwell time. The coolant conditions were representative of the Inner Vertical Target ones and a swirl tape (twist ratio = 2) turbulence promoter was provided.
        The test results showed some significant improvements, in particular the issues of W monoblocks macro-cracking and heat sink thermo-mechanical fatigue performances, identified during previous HHF tests campaigns. Some critical heat flux experiments were also performed.
        The main results will be presented and discussed in the paper.

        Speaker: Dr Bruno Riccardi (Fusion for Energy)
      • 11:00
        Progress on in-vessel poloidal field coils optimization design for alternative divertor configuration studies on the EAST tokamak 2h

        An upgrade to the lower divertor is currently being planned for EAST superconducting tokamak, aiming at >400s long-pulse H-mode operations with a full metal wall and a divertor heat load of ~10MW/m2. A new divertor concept for EAST, “Tightly Baffled Divertor”, suited to water-cooled W/Cu plasma face components (PFCs) with minimized divertor volume, has been proposed to achieve Te,target<5eV across entire outer target at lower separatrix plasma density and optimized pumping by a simple closed divertor structure combining horizontal target with inclined baffle, dome and duct. This divertor should allow access to high-triangularity small-ELM H-mode regimes and also allow achieving Snowflake or X-Divertor like configurations with the assistance of two water-cooled in-vessel divertor coils (DCs). Preliminary engineering design of in-vessel DCs indicates a maximum current of 8kAt for long-pulse discharges, and 20kAt for the shortest ones. However, flexibility on DCs position optimization is limited to the water cooling system. Initial plasma equilibrium studies by FIXFREE code, used in combination with CREATE-NL and EFIT tools, show that the distance of the two nearby divertor poloidal field nulls, can be decreased up to ~0.8m with a plasma current IP~400kA, leading to a configuration with the secondary x-point located between the primary x-point and the target or close to the target, with a significant increase of magnetic poloidal flux expansion and connection length. This may provide a promising divertor solution compatible with advanced steady-state core scenarios.

        Speaker: Giuseppe Ramogida (ENEA)
      • 11:00
        Progress on thermo-hydraulic and thermo-mechanical performances of Helium-Cooled-Molten-Lead-Ceramic-Breeder as near-term alternative blanket for EU-DEMO 2h

        Within the framework of EUROfusion activities, an alternative Helium-Cooled Molten Lead Ceramic Breeder (HC-MLCB) solid breeding blanket is being also developed at KIT for European DEMO. This concept is proposed as an alternative near-term breeding blanket and it is based on a fission-like “fuel-breeder pin” assembly configuration. Molten lead is used here as the neutron multiplier, Li4SiO4 in form of pebbles inside the fuel-breeder pins as tritium breeder and pressurized helium as coolant. In comparison to typical former cooling-plate configurations, the fuel-breeder pin assemblies greatly reduce the pressure drop, solving the key technology readiness issue of the currently available helium-circulator. Also, the combination of lead and lithium ceramics shows a good tritium breeding performance in a compact configuration (outboard blanket average radial thickness of 1000 mm instead of former 1300 mm).
        After initial design works with this concept for the previous EU DEMO tokamak baseline design 2015 (reported elsewhere), the current status of the design activities on the HC-MLCB integrated for the latest EU DEMO baseline 2017 (EU DEMO 2017) are presented and discussed in this paper.
        Firstly, the structural integrity of the preliminary blanket design under an in-box LOCA event (a key design driver) has been assessed and the design iterated to fulfill the design criteria. Then, and after corresponding neutronics analysis to validate the soundness of the design in terms of nuclear performance, thermo-hydraulic analyses have been conducted to evaluate the blanket temperatures and the coolant pressure drop. After some design iterations to satisfy the temperature design limits and pressure drop requirements, a structural assessment under normal conditions has been conducted with respect to the structural design standard RCC-MRx. The results presented here show that the current design of the HC-MLCB meets the basic nuclear and thermo-hydraulic-mechanical performances, setting the path for a consolidated design of this concept.

        Speaker: Guangming Zhou (Institute for Neutron Physics and Reactor Technology Karlsruhe Institute of Technology)
      • 11:00
        Protection of window assemblies against ECRH and CTS stray radiation in ITER 2h

        Stray radiation at 60GHz and 170GHz is an engineering challenge for the integrity of various window assemblies in ITER. Their protection and long term performance preservation are essential for both the operational safety of the device and its scientific exploitation. This contribution focuses on the assessment of Electron Cyclotron Resonance Heating (ECRH) and Collective Thomson Scattering (CTS) stray loads impacting the windows of one of the most affected diagnostic port plugs: #11. A Neutral Particle Analyzer (NPA), the Low Field Side Reflectometer , Vacuum Ultra Violet (VUV) and Halpha spectrometers are connected to this port plug and may need protection. Indeed, as ECRH/CTS stray radiation may last throughout most of the discharge, an increase in window temperature may occur leading to excessive stresses in the assemblies. Appropriate measures have therefore to be taken to improve the attenuation of their transmission lines. The attenuation of both at 60GHz (CTS) and 170GHz (ECRH) frequencies has been evaluated with both CST simulations and analytical calculations. The use of mitigation measures (such as coatings, grids, etc) has been considered. Solutions with quite good perspectives have already been identified for the NPA, VUV and Halfa lines. Further studies are under way for the design of the Low Field Side Reflectometer, considering active and/or passive components to dampen the stray radiation travelling through the line. Finally, a preliminary set of qualification tests both for the coating materials and for the fused silica window assemblies, with particular attention to the measurement of their dielectric properties, will be discussed.

        Speaker: Dr Michela Gelfusa (Industrial Engineering, University of Rome "Tor Vergata")
      • 11:00
        Purification of Pb-16Li breeder from corrosion products 2h

        This contribution provides summary of two purification experiments of liquid metal breeder Pb-16Li by a cold trap. The behavior of artificially added impurities were studied in non-isothermal ferritic loop Meliloo v1. During these experiments a Mn concentrations followed the solubility curve as published by Barker. More advanced trap design was tested in austenitic loop Meliloo v2. This trap was analyzed for deposits distribution over the internal surfaces using SEM. Large Fe-Cr-Mn-Ni particles and Ni-Mn, Ni-Mn-Sn intermetallic phases were found on the cooled wall, while deposits of Bi oxides were found on the non-cooled internals. The SEM images of typical deposits are presented in the contribution along with the location where they were found.

        Speaker: Lukáš Košek (Centrum vyzkumu Rez)
      • 11:00
        Radiolysis study of EU Li4SiO4 reference breeder material from the HICU experiment 2h

        To proceed the solid breeder concept for ITER and DEMO it is essential to investigate Ceramic Breeder (CB) materials’ properties. To ensure an adequate tritium production of the breeder material several requirements like a high lithium density, good tritium release behaviour, and a high resistance against neutron irradiation as well as thermomechanical stresses have to be fulfilled. Lithium orthosilicate (Li4SiO4), applied as pebbles, has been selected as reference material in the European Helium Cooled Pebble Bed (HCPB).

        Beside standard material characterization, the response of CB materials to neutron irradiation is an important issue. Therefore, CB pebbles of the EU reference material were exposed to neutron irradiation in the HICU experiment (high neutron fluence irradiation of pebble stacks for fusion), that was carried out in the High Flux Reactor (HFR) in Petten (Netherlands) between 2008-2010. Different grades of Li4SiO4 pebbles containing a surplus of 2.5 wt% SiO2 and different 6Li-contents up to 20 % were included in the irradiation under DEMO relevant conditions.

        While the Post-Irradiation Examination (PIE) on tritium release behaviour and material properties were recently presented, new and additional results of a radiolysis study on the Li4SiO4 samples was performed in Latvia will be presented here. Radiation induced Defects (RD) and Products (RP) were investigated using basically Electron Spin Resonance (ESR) and Raman spectroscopy. Further analyses on the tritium release behaviour were performed using Thermally Programmed Desorption (TPD).

        The presented results will reveal new insights of CB pebbles’ behaviour with regard to neutron irradiation and will therefore significantly contribute to the knowledge of CB pebbles' properties in a fusion relevant environment.

        Speaker: Julia Heuser (Institute for Applied Materials Karlsruhe Institute of Technology)
      • 11:00
        Reactivity and thermal stability of ternary Be-Zr-V beryllides 2h

        As a water-cooled solid breeder blanket of a fusion reactor, safety concern has become one of the most critical issues. In specific, Be pebbles as a multiplier have been well-known to generate hydrogen and exothermally react while a loss of coolant accident (LOCA) occurred. In contrary to these Be pebbles, Beryllium intermetallic compounds (beryllides) are one of promising materials because of its much more stable chemical reactivity at high temperatures. Currently, many works on the development of advanced neutron multipliers by Japan and the EU is part of the DEMO R&D activities at the International Fusion Energy Research Center (IFERC) project, which forms a part of the Broader Approach (BA) program. Fabrication methods of beryllides pebbles have been successfully developed by combining a plasma sintering synthesis method and a rotating electrode granulation method.
        By using these methods, preliminary synthesis of the ternary beryllide pebbles with mixtures of Be13Zr and Be12V with ratios of 1:0.1, 1:0.4, 1:0.6, 1:0.8, and 1:1, has been conducted. It was clear that small size of Be13Zr phase as precipitate and Be12V phase formed without Be phase formation since those composition does not contain the peritectic reaction.
        Additionally, the ternary beryllide pebbles found out to have a lower reactivity to water vapor as well as a higher thermal stability. In the present study, not only synthesis process but also characterizations at high temperatures will be introduced.

        Speaker: Jae-Hwan Kim (Department of Blanket Systems Research National Institutes for Quantum and Radiological Science and Technology)
      • 11:00
        Recent experiments with the European 1MW, 170GHz industrial CW and short-pulse gyrotrons for ITER 2h

        The European Gyrotron Consortium (EGYC) is developing the EU 1 MW, 170 GHz CW industrial prototype gyrotron for ITER in cooperation with the industrial partner Thales Electron Devices (TED) and under the coordination of Fusion for Energy (F4E). This hollow cylindrical cavity gyrotron is based on the 1 MW, 170 GHz short-pulse (SP) modular gyrotron that has been designed and manufactured by KIT in collaboration with TED. The experiments with the CW industrial gyrotron are organized in two phases. The first phase was completed successfully at KIT in 2016. In the SP regime (<10 ms pulses), stable excitation of the nominal cavity mode TE32,9 at 170.22 GHz was achieved for a wide range of operating parameters. The maximum RF output power of the tube is higher than 0.9 MW with a total efficiency of 26% in non-depressed collector operation. The Gaussian mode content of the RF output beam is higher than 97%. In long-pulse operation, pulses with duration of 180 s (limited by the high-voltage power supply at KIT) delivered power higher than 0.8 MW with 38% efficiency (in depressed collector operation). The second phase of the experiments is ongoing at SPC, Lausanne, with the goal to further optimize the output power and extend the pulse duration. In parallel, the experiments with the SP prototype are continued at KIT. The SP tube, which in multiple experimental campaigns delivered power higher than 1 MW with 42% efficiency (in depressed collector operation), is further upgraded. Various depressed collector operation schemes are tested with the goal to achieve an efficiency higher than 50%. Moreover, different beam tunnels will be tested in order to have the possibility to go to higher operating beam currents without exciting parasitic oscillations. In this work, the latest results with the CW and SP prototype gyrotrons will be presented.

        Speaker: Dr Gerd Gantenbein (Institute for Pulsed Power and Microwave Technology, Karlsruhe Institute of Technology)
      • 11:00
        Registering micro-indentation of neutron-irradiated low-activation steel at high temperatures 2h

        An ongoing study about the influence of neutron irradiation on the mechanical properties of the first wall’s structure materials is presented in this work. EUROFER97 and an Oxide Dispersion Strengthened EUROFER steel were irradiated in the Petten High Flux Reactor up to a nominal dose of 15 displacements per atom at temperatures between 250 and 450°C and investigated by an advanced method of registering micro-indentation. For the purpose of characterizing irradiated and thus radioactive samples at future fusion reactor conditions i.e. at high experimental temperatures, the Karlsruhe High Temperature Indenter (KAHTI) was developed. In order to safely handle radioactive samples, KAHTI itself is operated by remote control inside a Hot Cell. Due to a new optical depth sensing method, it now is possible to perform registering micro-indentation at temperatures up to 650°C with highly accurate displacement measurement. The results gained by KAHTI show the increase of hardness by neutron irradiation and its dependency on the irradiation temperature. In addition, the sensitivity of the hardness to different testing temperatures is investigated. These results lead to a better understanding and quantification of the different neutron induced damages. The possibility of recombining neutron induced Frenkel-pairs and thus the reduction of irradiation damage without affecting the material’s grain structure is investigated by performing post irradiation annealing inside KAHTI at 550°C. First results indicate an almost complete recovery of mechanical properties. Comparing the results of this study to conventional testing methods approves KAHTI being a complementary testing method with a very high specific amount of information per sample and per sample volume.

        Speaker: Alexander Brabänder (Institut for Applied Materials Karlsruhe Institut of Technology)
      • 11:00
        Reliability Assessment of remote maintenance strategy for CFETR Divertor 2h

        China Fusion Engineering Testing Reactor (CFETR) will be built to test and verify the feasibility of engineering and technology in practice for the future fusion reactor. Long pulse and steady-state operation will be demonstrated with duty cycle time not less than 30~50%.During plasma operation, the in-vessel components of the fusion reactor will be activated and contaminated with tritium. Because of the beta and gamma activation of the component bulk and surface dust (beryllium, carbon, tungsten) special remote handling techniques will be required during machine maintenance periods. The divertor cassettes remote replacement is one of the key maintenance operations for the CFETR to meet the requirements of duty cycle time. RH maintenance strategies will have a significant impact on the layout of the machine and design of components. In this paper we will give an overview of the different CFETR divertor components remote maintenance strategies as well as to describe the maintenance process of the in vessel components. Preliminary assessment of the divertor maintenance scheme is done in order to carry out RH maintenance tasks successfully and efficiently. Considering the number of feasible designs for the divertor maintenance, we concentrate remote handling concept assessments on the follow principles: As simple as possible; high-security; high reliability and availability.

        Speaker: Dr Huapeng Wu (Lappeenranta University of Technology)
      • 11:00
        Repair processes of W7-X target modules 2h

        The highly loaded surface of the actively water-cooled divertor of Wendelstein 7-X (W7-X) is made of 100 individual target modules. In each target module, a set of target elements is water-cooled in parallel and fed by manifolds. A target element is made of a CuCrZr copper alloy heat sink, armored with CFC NB31 tiles. Due to the width of the target elements, CFC tiles had to be successively electron beam welded onto the heat sink from two sides.
        He leak testing under pressure in a vacuum oven was systematically performed for each target element and module. The type 5S target element had a higher percentage of rejection rate during the production, and one series element did not pass the leak test after high heat flux testing. The leakage was the result of the combination of a porosity concentration due to the CFC-CuCrZr - weld in the centre part of the element and the reduced CuCrZr thickness due to machined slots at this location. The selected solution was sealing of the slit between two tiles by electron beam welding.
        The target elements were then connected by orbital welding to the piping system. One of the first produced batch of 30 target modules did not pass the integral He leak testing. Two leaks were detected in the weld seam between two target element connectors and manifold pipes. Their positions did not allow a reliable process for re-welding from outside due to the impossibility to install any jigs. The selected approach was drilling of apertures through the neighbouring manifolds to allow direct access to the leaking seams from inside. The openings allowed installation of an inside orbital welding head. This solution was validated in an intensive prototyping phase. Finally, the repaired target modules passed the He leak test in oven.

        Speaker: Dr Patrick Junghanns (Max Planck Institute for Plasma Physics)
      • 11:00
        Scaling analysis and design for the test model of water-cooled ceramic breeder blanket 2h

        In Chinese Fusion Engineering Test Rector (CFETR), blanket is a key component, responsible for producing and transporting tritium, energy conversion and output, so its safety is of particular concern. The water-cooled ceramic breeder blanket (WCCB) is one of three candidate blankets for CFETR. To confirm safety of WCCB, sufficient data are required to estimate the thermal-hydraulic state and response in the blanket during postulated accidents. Due to strict test conditions and the complexity of WCCB structure, the expense of direct experiments is so great that it is inevitable to design a simplified and scaled-down model instead of the prototype for experiments with some appropriate scaling techniques. In this paper, based on same working fluid—water, full-temperature and full-pressure test conditions, scaling analysis is used. On one hand, the top-down scaling analysis is adapted to preserve the system dynamic responses. On the other hand, the bottom-up scaling analysis is adapted to preserve effective local phenomena. Through that, the characteristic time ratio group and the relation of important parameter ratios are established. The design scheme of reduced height and the number of flow channels is adopted, and the prototype blanket’s structure is reasonably simplified to reduce difficulties of manufacture. Finally, scaling distortions of designed model are quantified, and verify that model can satisfy engineering application requirements. The test section can competently reproduce the thermal-hydraulic behaviors in WCCB.

        Speaker: Mr Zihan Liu (University of Science and Technology of China)
      • 11:00
        Seismic analyses of the Double Closure Plate Sub-Plate for the ITER Electron Cyclotron Upper Launcher during the Vacuum Vessel baking scenario 2h

        Four Electron Cyclotron Upper Launchers (EC UL) will be used at ITER to counteract magneto-hydrodynamic plasma instabilities by aiming up to 20 MW of mm-wave power at 170 GHz. This mm-wave power will be injected through eight ex-vessel waveguide assemblies for each EC UL to the in-vessel waveguides. The power exiting the in-vessel waveguides located inside the Port Plug will be directed by quasi-optical mirrors to specific plasma locations. The Double Closure Plate Sub-Plate (DCPSP), which defines the border between ex-vessel and in-vessel components, was introduced in order to minimize the openings exposing the interior of the plug to avoid the near environment activation in case of maintenance or intervention on the in-vessel components. The seismic event taking place during the Vacuum Vessel (VV) baking scenario was identified as one of the most stringent load combinations for the DCPSP. The modal analysis of the DCPSP shows that the natural frequencies are far from the peaks of the ITER reference spectra at the Port Plug flange. Therefore, the feasibility of analysing the seismic event by using a static method as a replacement of the Response Spectrum method is also investigated. Then, the results due to the seismic event are to be combined with the ones produced due the loads occurring at the DCPSP during the VV baking scenario. The stress distribution produced from this load combination is categorized and compared with the allowable design limits in order to evaluate the mechanical integrity of the DCPSP. This work was supported in part by the Swiss National Science Foundation. This work was carried out within the framework of the ECHUL consortium, partially supported by the F4E grant F4E-GRT-615. The views and opinions expressed herein do not necessarily reflect those of the European Commission or the ITER Organization.

        Speaker: Dr Avelino Mas Sánchez (EPFL)
      • 11:00
        Sensitivity of First Wall thermal-mechanical performance on cooling channel geometry and thermal conductivity 2h

        The First Wall (FW) of DEMO or following fusion power reactors will be exposed to high heat fluxes by thermal radiation and energetic particles from the plasma. During steady state, values of over 1 MW/m² are expected for the EU DEMO concept. The function of the FW therefore relies on (1) good thermal conduction from the plasma facing surface through the channel material, and (2) good heat transfer from the channel wall surface into the coolant medium flow. Those aspects influence the shell-average operation temperature of the structural material - determining the materials mechanical strength - and also the temperature spreads within the component - causing thermally induced secondary stresses.

        For given boundary conditions, FW designs can be thermal-mechanically optimized. In practice, deviations between the optimum design point versus the device under service have to be accepted due to manufacturing tolerances and variable coolant conditions for variably sized channels. This paper assesses sensitivities of the temperature and stress fields by a parameter study performed by finite element analyses, considering also feedback of the channel geometry to the relative flow ratio through individual channels, using validated correlations for helium thermal-hydraulics.

        Further consideration is given to the materials thermal conductivity, which can vary with alloy composition and material history. Thermal conductivities of FW candidate reduced activating ferritic/martensitic steels are reviewed, and own measurements are reported. The possible tolerance in the thermal conductivity due to tolerances in the alloy composition is assessed by application of a neural network trained specifically to the named family of steels.

        Acknowledgments: This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 under grant agreement No. 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission.

        Speaker: Frederik Arbeiter (KIT)
      • 11:00
        Servo valve endurance test for Water-Hydraulic systems in ITER-relevant conditions 2h

        ITER Divertor maintenance equipment work under considerable ambient temperature and radiation load. The heavy components are moved with equipment powered with water hydraulics, with demineralised water as a pressure medium. None of this has yet been tested in ITER-relevant environmental conditions and over projected duty cycles and loading. Hence, a project was undertaken to ascertain the component compatibility with the environment and pressure medium, and their robustness over the required operational period. No irradiation, however, was included. This would be done later on if this test was successful. A heated chamber was constructed to emulate the 50°C ambient temperature of the divertor area. All the tested components were located within this chamber.
        Test trajectories and loads were derived from the Divertor Test Platform 2 (DTP2) at Tampere, Finland, at which the Cassette Multifunctional Mover (CMM) operations and equipment are prototyped. The prototyped operations are along the operations projected for the ITER Divertor area maintenance operations with active-to-idle ratio similar to what was projected for ITER operations. The most important component to be tested was the servovalve. As the hydraulic medium is rather aggressive and the selected servovalve was not originally designed to be used with water, the 2000-hour operational time was considered to be a potential issue. Hence, test routines and measurements were specifically tailored to measure the operational parameters of the servovalve.
        As a result, after the 2000-hour test the servovalve parameters remained within the limits promised by the supplier (Moog hydraulics). Null-point leakage increased by 45% (from 0.45lpm to 0.65lpm which is still quite low) and tare leakage did not change significantly. Pressure gain and hysteresis increased significantly but remained within the allowable limits. Cylinder tracking error remained more or less constant over the test, although it decreased an insignificant amount over the last testing week.

        Speaker: Liisa Aha (Laboratory of Automation and Hydraulic Engineering Tampere University of Technology TUT)
      • 11:00
        SMITER: a field-line tracing environment for ITER 2h

        The ITER plasma-facing components (PFC) are now fully designed and procurement is underway. A key utility in such design is field line tracing for different magnetic equilibria which allows the definition of component front surface shaping. On ITER, this design phase has deployed both analytic theory [1] and the tracing codes CASTEM and PFCFLUX [2]. Attention is now turning towards the critical issue of management and control of PFC heat fluxes, so important in an actively cooled device such as ITER. To facilitate the development of control algorithms, particularly for protection of the beryllium main chamber first wall panels [3], a new field line tracing environment, SMITER, has been developed, featuring a sophisticated Graphical User Interface (GUI) that uses the SMARDDA [4] kernel and has been thoroughly benchmarked against PFCFLUX for specific cases of first wall panel and divertor target loading. SMITER allows power deposition mapping in the full 3-D CAD geometry of the machine, taking as input a user-defined specifications for parallel heat flow in the scrape-off layer. In addition to its use as input for shape control algorithms [3], the tool will also be important for the production of synthetic surface temperatures using built in thermal models for input to diagnostic design.

        The newly developed GUI framework provides CAD and IMAS integration, parametric CAD components catalogue, meshing, visualization, Python scripting, storage in hierarchical data files (HDF) with several simulation cases in one study running in parallel and using MPI for code speedup. Integrated ParaView module can augment CAD geometry, meshes and results.

        [1] P. C. Stangeby, Nucl. Fusion 51 (2011) 103015; [2] M. Firdaouss et al., J. Nucl. Mat. 438 (2013) S536; [3] H. Anand et al., to be published at IAEA FEC 2018; [4] W. Arter et al., IEEE Tr. Plasma Sc. 42(7) 1932, arXiv:1403.7142

        Speaker: Dr Leon Kos (Mech.Eng., LECAD laboratory, University of Ljubljana)
      • 11:00
        Strategies toward the Realization of the Helical Fusion Reactor FFHR-c1 2h

        Design studies on the helical fusion reactor FFHR-c1 has been progressed. The main goal of the FFHR-c1 is to demonstrate one-year steady-state sustainment of the fusion plasma with self-produced electricity and tritium. The major radius of the plasma, R, is ~10 m and the magnetic field strength at the plasma center, B, is ~8 T. High-temperature superconductor (HTS) magnet coils are adopted in the design. During the one-year operation, ~370 MW of fusion power is sustained with ~25 MW of auxiliary heating power supplied by the electron cyclotron heating (ECH) of ~220 GHz. Then, the fusion gain, Q, is ~15. A liquid metal divertor system named the REVOLVER-D is adopted for the reasons of easy maintenance and a high permissible heat load. In this system, ten sets of molten tin showers are discretely inserted to the inboard side of the ergodic layer surrounding the main plasma. A cartridge-type molten salt (or liquid metal) blanket named the CARDISTRY-B is also adopted for easy maintenance. For the realization of the FFHR-c1, it is necessary to accumulate experiences on the new technologies of the HTS magnet coils, the REVOLVER-D, and the CARDISTRY-B. Including these three, we have defined 22 important issues that should be resolved before building the FFHR-c1. The strategy to efficiently address the 22 issues has been also discussed. As a result, we propose a step-by-step approach by developing FFHR-01 (R ~ 0.4 m and B ~ 3 T) for basic studies on HTS coils and the REVOLVER-D, FFHR-a1 (R ~ 2.5 m and B ~ 4 T) for demonstration of one-year operation under a non-nuclear condition, and FFHR-b1 (same R and B as FFHR-a1) for DT operation, before FFHR-c1. The FFHR-b1 plays a role of volumetric neutron source (VNS) and the FFHR-a1 corresponds to the cold test of the VNS.

        Speaker: Dr Junichi Miyazawa (National Institute for Fusion Science)
      • 11:00
        Structural integrity assessment of an ITER ECH&CD Upper Launcher mirror (LM1) 2h

        The Lower Mirror one (LM1) is part of the in-vessel quasi-optical beam propagation system for the ITER Electron Cyclotron (EC) Upper Launcher (UL), in which each of eight beams of mm-waves are reflected from four mirrors during passage to the plasma. 60000 thermal cycles are foreseen at frequencies lower than 3Hz and power levels up to 1.31MW per beam.
        This paper reports the means used to accurately ascertain resistance against the following failure modes: 1) plastic collapse, 2) thermal fatigue and 3) ratcheting. Analyses are carried out in accordance with the design-by-analysis approach. Transient thermo-mechanical effects are investigated via finite elements to support the assessment during the thermal cycle.
        This structural integrity study of the LM1 mirror using the American code and standards for pressure vessels and piping aims to confirm the compliance of the proposed design.
        This work was supported in part by the Swiss National Science Foundation.
        This work was carried out within the framework of the ECHUL consortium, partially supported by the F4E grant F4E-GRT-615. The views and opinions expressed herein do not necessarily reflect those of the European Commission or the ITER Organization.
        Keywords: LM1, Upper Launcher, Transient Thermo-Mechanical Analysis, Structural Integrity, Fatigue

        Speaker: Dr Matteo Vagnoni (Swiss Plasma Center, École polytechnique fédérale de Lausanne)
      • 11:00
        Supplementary neutronics analysis of DEMO WCLL including activity and decay heat 2h

        The WCLL (Water Cooled Lithium Lead) is a European option of the breeder blanket dedicated for DEMO fusion power reactor as being developed in the frame of EUROfusion’s Power Plant and Technology (PPPT) programme. The intense neutron radiation produced results in a strong activation of the breeder blanket structural elements. The activation and decay heat generation of the WCLL components need to be assessed for maintenance, decommissioning and waste management purposes and related safety analyses.
        This paper presents the analyses performed within the SAE (Safety and Environment) project of EUROfusion/PPPT aimed at providing up-to-date estimates of the activity inventories and the decay heat generation in the WCLL. A detailed investigation based on a set of coupled MCNP neutron transport and FISPACT inventory calculations were performed using the 2017 WCLL MMS (Multi-module Segment) model and the FENDL-3.1 nuclear cross-section data library. Activity inventories and decay heat data were assessed for the different breeder blanket segments to take into account heterogeneity.
        The paper discusses the results obtained for the activity and the decay heat as a function of the decay time after radiation and also addresses the issue of the radiation dose loads that have to be expected due to the activated components/systems.

        Speaker: Dr Gediminas Stankunas (Laboratory of Nuclear Installation Safety Lithuanian Energy Institute)
      • 11:00
        Supporting analysis of the ITER TBM Frame and Dummy TBM designs 2h

        The validation and testing of tritium breeding blankets concepts, which are relevant for a future commercial reactor, is one of the goals of the ITER project. To achieve these objectives, mock-ups of breeding blankets, called Test Blanket Modules (TBMs), are tested in three ITER equatorial ports. Each TBM and its associated shield form a TBM-set that is mechanically attached to a steel frame. A frame and two TBM-sets form a TBM port plug (TBM PP). In case a TBM-Set is not available, it can be replaced by a Dummy TBM, steel made only.
        Both the design and manufacture of TBM Frames and Dummy TBMs are fully under the responsibility of the ITER organization.
        This paper describes the summary of recent analysis activities to support the design of the TBM Frame and Dummy TBM, which is presently in the preliminary design phase. In the following, the main items addressed in this work are reported.
        • A refined evaluation through full 3D models of the heat transfer coefficient (HTC) and pressure drop of water cooling system of TBM Frame and Dummy TBM; alternative cooling circuits studied, with the aim of minimizing thermal stress and pressure drop.
        • Electro-magnetic (EM) analyses on the most recent design of TBM-sets and a TBM PP with two Dummy TBMs under cat. II, III and IV disruptions;
        • Evaluation of the impact of the changes from CDR configuration on EM loads on Dummy TBMs
        • Detailed thermo-mechanical analyses and structural integrity assessments of TBM Frame and Dummy TBM carried out according to RCC-MR 2007.
        • Proposal of design improvement from the outcomes of the above analyses, to be possibly included in the FDR configuration.

        The views and opinions expressed herein do not necessarily reflect those of the ITER Organization.

        Speaker: Davide Lumassi (L.T.CALCOLI S.R.L.)
      • 11:00
        Surface oxidation effect on deuterium permeation in reduced activation ferritic steel F82H for DEMO application 2h

        Tritium permeation through structural materials is a significant issue for the Japan’s DEMO reactor blanket concept. Reduced activation ferritic steel F82H is a prime candidate for the blanket structural material. The previous study showed a thin chromium oxide layer formed on a steel substrate worked as tritium permeation barrier; however, heat treatment parameters at atmospheric pressure for the formation of a chromium oxide layer and its permeation behavior in DEMO reactor environments are not clear. In this study, surface oxidation treatments for F82H have been performed to evaluate the effect of the chromium oxide layer on deuterium permeation and its stability under actual DEMO reactor conditions.
        To form a tight chromium oxide layer without formation of iron oxide which causes cracks in the layer, F82H steel plates were heat-treated under low oxygen partial pressure conditions: for 5 min at 700‒720 ºC in gas flow of argon-hydrogen mixture. After surface observation and analysis, gas-driven deuterium permeation measurements were performed at 300‒600 ºC. The selected samples were exposed to purge gas of helium with 1 vol% hydrogen or liquid lithium-lead for 100 h at 500 ºC for the simulation of DEMO blanket conditions.
        The optimized heat-treatment parameter for chromium oxide formation without iron-oxide formation was determined: for 5 min at 710 ºC in half-and-half argon-hydrogen mixture gas. The thickness of the layer was estimated to be less than 100 nm. Deuterium permeation flux of the sample decreased by a factor of 10 in comparison with untreated F82H in the first measurement at 300 ºC, and showed a further decrease by a factor of 150 at 500 ºC due to an increase in the layer thickness by 1.5 times during the permeation measurements. However, the chromium oxide layer was lost with an increase in deuterium permeation after exposure to DEMO blanket environments.

        Speaker: Takumi Chikada (Graduate School of Integrated Science and Technology Shizuoka University)
      • 11:00
        T-15MD tokamak plasma control platform architecture 2h

        A key feature of the developed T-15MD tokamak Plasma Control System (PCS) is its ability to rapidly design, test and deploy real-time shot scenario algorithms. PCS platform consist of two levels:
        1. High application-specific level: model development and linear approximation, calculation of the experiment scenario, controllers design and experiment simulation (Mathlab Simulink RT high-performance computer);
        2. Process control level: real-time control of plasma parameters (National Instruments hardware running LabView RT operating system).
        In the Hardware-in-the-Loop simulation mode communication between the levels (1) and (2) is realized by the reflective memory (RFM) “star” topology network and the middleware S-function package within Simulink RT environment. RFM providing a deterministic real-time method of sharing memory between the systems.
        The proposed architecture of the platform allows to perform testing and configuration of the PCS before plasma shots, which increases the efficiency of the experiments while reducing costs. In future, the plan is to use the high-performance computer Simulink RT to perform real-time calculations of plasma reconstruction and equilibrium in the magnetic control loop, implementing PCS data exchange in the RFM network. It is planned to develop infrastructure for simplified integration and testing of third-party control algorithms and plasma-physical codes.

        Speaker: Dr Mikhail Sokolov (National Research Centre “Kurchatov Institute”)
      • 11:00
        Technical proposals for the IGNITOR control, data access and communication system 2h

        The control, data access and communication system (CODAC) designed to solve the tasks of planning, preparing and conducting the experiment, collecting, processing and complex analysis of the experimental results at the IGNITOR tokamak fusion project.
        It is proposed to build CODAC based on modern failsafe dual-redundant industrial equipment manufactured by National Instruments, Schneider Electric, Siemens, Highland Technology, Hewlett-Packard, GE Fanuc etc. Cross-system interaction provided by high-speed fiber-optic networks: Reflective Memory, MXI-Express, Ethernet network and pulse synchronization with a few microseconds time delay. The high level of Fast Systems (the control loops above the 100 Hz) based on Mathlab Simulink RT (Linux operation system) and National Instruments Labview RT: plasma control system, central safety system, coil power supply system, diagnostics etc. The high level of Slow Systems (the control loops are slower or equal of 100 Hz) based on Wonderware InTouch: auxiliary process systems i.e. vacuum system, cooling water system etc. IGNITOR CODAC should provide synchronized real-time high throughput data processing, high reliability, availability, maintainability and access to experimental data for remote participants through the unified protocol via Internet.

        Speaker: Dr Aleksandr Kachkin (National Research Centre “Kurchatov Institute”)
      • 11:00
        Temporary stagnation in the recrystallization of warm-rolled tungsten in the temperature range from 1150 °C to 1300 °C 2h

        Pure tungsten is a potential candidate for armor material of fusion reactors as it possesses superior thermal properties and radiation resistance. Application at the desired operation temperatures for longer times will result in a loss of strength accompanied by embrittlement due to thermal activated changes in the microstructure, in particular due to recrystallization, undermining tungstens outstanding performance. Investigating the thermal stability of tungsten depending on the manufacturing process is therefore considered crucial. The thermal response of a sintered, hot isostatic pressed tungsten plate warm-rolled to 80% thickness reduction is assessed in the temperature range from 1150 °C to 1300 °C. The restoration processes occurring during annealing are tracked by changes in mechanical properties through hardness testing and identified by supplementary orientation mapping by means of EBSD. Isothermal annealing treatments were performed at six different temperatures. With increasing annealing time, the macro hardness decreased continuously; different stages corresponding to different stages of the microstructural evolution (recovery, recrystallization and grain growth) could be identified and confirmed by the microstructural information gained by EBSD. For the time to half recrystallization, an activation energy comparable to the activation energy of bulk self-diffusion is obtained. The correspondingly extrapolated times to half recrystallization would not allow to operate the material at temperatures above 1100 °C. Nevertheless, for all annealing temperatures a stagnation period in the evolution of the macro hardness was observed where the hardness loss and hence the degradation of mechanical properties halted for a significant amount of time, before it resumed. EBSD investigations revealed that the stagnation occurred when tungsten was still only partially recrystallized, and that recrystallization recommences afterwards. Such a temporary resistance against complete recrystallization and revealing its microstructural origin is of uttermost importance as its understanding could provide interesting insight in opportunities for designing tungsten material with improved thermal stability.

        Speaker: Dr Umberto Maria Ciucani (Department of Mechanical Engineering, Technical University of Denmark)
      • 11:00
        Testing of ceramic membranes for PEG separation and preliminary design of a membrane cascade 2h

        The Plasma Exhausts Gases (PEGs) proposed to reduce the power load over the plasma facing components are separated by the Plasma Exhaust Processing System of DEMO.
        Two kinds of ceramic porous membranes (with top layer of pore size 0.2 m and 3-4 nm, respectively) used commercially for the filtration of liquids have been tested in order to verify their application for the PEG separation. The experiments have been carried out at room temperature with feed pressure in the range 100-180 kPa and permeate pressure of 100 kPa by testing Ar, N2, He and H2. The results of the experiments have been exploited to validate a gas mass transfer model taking into account the permeation mechanisms of Knudsen and Poiseuille and, then, the model has been used to assess the permeance of the molecule DT.
        According to the processing requirements for the PEG separation of DEMO and based on the permeance and selectivity values of the commercial membranes calculated by the model, a membrane system consisting of ceramic membranes (top layer of pore size 0.2 m) followed by a Pd-Ag permeator has been assessed. The calculation of a counter-current recycle cascade of ceramic membranes porous membranes has been conducted via the simplified Underwood-Fenske method under the hypothesis to use Ar as PEG. By vacuum pumping the permeate of the ceramic membrane cascade at 10 mbar the maximum Ar concentration in the retentate is 99.95%, while the higher Ar concentration of 99.995% can be achieved with a vacuum level in the permeate of 1 mbar. In both the cases, the Pd-permeator downstream the ceramic membrane cascade recovers about the 99% of the DT fed in form of ultrapure gas.

        Speaker: Silvano Tosti (CIRDER University of Tuscia)
      • 11:00
        The added value working under ISO 9001 in nuclear fusion technology R&D at ENEA 2h

        The ENEA Fusion Department (FSN) operates in the field of nuclear fusion under a Quality Management System (QMS) according to ISO 9001 since 2011. At the beginning this methodology applied in R&D activities of a Research Institution such as ENEA seemed to be far from the industrial reality according to an internal and external perspective. But now that the construction of ITER reactor became a reality for which the industry and the R&D activities are overlapping, the ISO 9001 approach is becoming a winning solution. Many ITER tenders require that applicants have a certified QMS. In addition to the higher potential to achieve success in these tenders, having a QMS produced in ENEA FSN positive results in terms of added value in planning, risk-based thinking, control services, document management, metrology, performance evaluation and continual improvement. These actions defined by appropriate procedures are useful to satisfy the customer, but also to improve the internal organization of work. The methods applied to obtain the sensitivity to the critical issues and to spread the culture of quality have been adapted to the unique sector of nuclear fusion technology (e.g. design and manufacture of specific components as the divertor HHF components or superconducting magnets, performance of experiments, etc.). The references for these methods are still the performance assessment through the monitoring, measurement, analysis and evaluation of processes and internal audits and the pursuing of continual improvement, but the indicators used for processes measurement are not those typical of a company in which most of the indicators are linked to the economic profit, but some more specific; the impact factor of research activities, the self-assessment efficiency, the continual training of the personnel, the management of resources for measurement, the assessment of the value of ENEA FSN activities by the analysis public press.

        Speaker: Alexander Rydzy (FSN ENEA)
      • 11:00
        The Design of the DONES Lithium Target System 2h

        In the framework of the EU fusion roadmap implementing activities, an accelerator-based Li(d,n) neutron source called DONES (Demo-Oriented early NEutron Source) is being designed within the EUROfusion work-package WPENS as an essential irradiation facility for testing candidate materials for DEMO reactor and future fusion power plants. DONES will employ a high speed liquid lithium jet struck by a 125 mA, 40 MeV deuteron beam to generate the intense neutron flux used to irradiate the materials sample up to the desired level of displacement damage (~10 dpa/fpy for iron in 0.3 l) and He production rates (~10-13 appm He/dpa).
        In order to rapidly achieve a sound and stable design, a new configuration of the DONES target system based on the so-called integral concept has been proposed as reference solution in place of the former baseline design that envisaged a target assembly endowed with a replaceable back-plate, being the latter solution not fully qualified yet and thus not readily implementable.
        Moreover, following the outcomes of a detailed dedicated analysis taking into account several integrated aspects (such as tritium generation, cavitation issues, maintenance strategy, etc…) a decision was taken to move the Quench Tank inside the Test Cell while in the former design it was arranged in the lithium loop area, below the Test Cell floor.
        Further modifications have also been introduced concerning inlet pipe and vacuum chamber configuration as well as support structure layout. A more consolidated design of the interfaces with the lithium loop and the accelerator beam line has been proposed too.
        In this paper, a brief description of the current design of the DONES target system is presented including all the above mentioned aspects. Main results of supporting neutronic and thermomechanical analyses are also reported, showing the capability of the system to fulfil the prescribed requirements.

        Speaker: Pietro Arena (Department of Energy Information Engineering and Mathematical Models (DEIM) University of Palermo)
      • 11:00
        The development of technology of Be/CuCrZr joining using induction brazing 2h

        Beryllium was selected as the plasma facing material for the ITER First Wall. Realization of the advantages of beryllium as a plasma facing material depends on the reliability of the critical beryllium joint with the heat sink made from CuCrZr alloy. This paper considers the method of induction brazing as the technology for this critical joint.
        To prevent the formation of brittle intermetallics and preservation of the material properties of the CuCrZr alloy, it leads to the need to minimize the brazing temperature and time of the brazing cycle. This paper presents the result of selection of the optimal eutectic of STEMET 1101 brazing alloy thereby reducing the temperature of the brazing to 680 оС.
        Using inductor allows to heat locally the beryllium tile/CuCrZr alloy region. The fast local heating of the braze region allows to melt the Stemet whilst ensuring that the major part of the component remains at a relatively ambient temperature. This is an important feature of the induction brazing technique as it will ensure the minimum time above the transition temperature for the CuCrZr and lead to acceptable material properties. The brazing is performed in a vacuum chamber to avoid contamination and oxidization of the component.
        To fix each of the beryllium tiles, a unique clamping tool was developed and applied. This tool allows for an even distribution of the clamping force across the tile and ensures adequate clamping during the complete brazing cycle. The clamping tool does not overheat, can be quickly assembled, and is reusable
        Along with ultrasonic testing of brazed joints, the most important criteria for the reliability of the beryllium armor is its behavior under cyclic heat fluxes. Authors present results of ultrasonic testing of the joints, as well as the first results of the high heat flux testing of the mock-ups.

        Speaker: Dr Aleksandr Gervash (Plasma Facing Components, Efremov Institute)
      • 11:00
        The helium turbo circulator – the heart of cooling systems 2h

        In few last decades, great attention was paid to development in the field of fusion technology. Currently, the International Thermonuclear Experimental Reactor (ITER) is under construction followed by Demonstration Power Station (DEMO) which should be first nuclear fusion power plant in the world. Both of these facilities have one point in common – high power density and thus great demands to supporting cooling systems. Therefore, hand in hand with development of fusion technology, it is also necessary to pay attention to cryogenic, refrigeration and cooling technologies. The paper presents development of helium Turbo Circulator (TC) installed in the form of two pieces in experimental testing facility HElium LOop KArlsruhe (HELOKA) in Germany and as one piece in HElium For FUSion (HEFUS-3) located at ENEA Brasimone, Italy. All installed TCs have output power 232 kW and were/are used in pressured helium circuits for cooling mock-ups of important ITER parts – Test Blanket Module (TBM) and Tested Diverter Module (TDM). Simultaneously, the paper summarizes gained experience and deals with a concept of bigger TCs which could be used in real cooling systems for cooling Inboard Blankets (IB), diverters and Outboard Blankets (OB) of fusion reactor in ITER and DEMO project.

        Speaker: Martin Kroupa (Rotary Machines Workgroup ATEKO a.s.)
      • 11:00
        THE HIGH VOLTAGE DECK 1 AND BUSHING FOR THE ITER NEUTRAL BEAM INJECTOR: INTEGRATED DESIGN AND INSTALLATION IN MITICA EXPERIMENT 2h

        MITICA is the full scale prototype of ITER Heating Neutral Beam (HNB), designed to deliver 16.5MW of heating power to ITER plasma, currently under construction at the Neutral Beam Test Facility in Padova (Italy). In ITER HNB, negative ions (H-/D-) are produced in the Ion Source (IS) polarized to ground at -1012kV, then extracted by 12kV extraction voltage, accelerated to ground at 1MeV energy and finally neutralized. The complex power supply system feeding the IS includes two non-standard equipment, beyond the actual industrial standard for insulation voltage level (-1MVdc) and dimensions: 1) the HVD1 (High Voltage Deck1), a large air insulated Faraday cage (12.5m (L) x 8.4m (W) x 9.6m (H)) hosting the IS power supplies and diagnostics; 2) the HVBA (High Voltage Bushing Assembly), a -1MVdc air to SF6 Bushing which interfaces the HVD1 with the SF6 insulated Transmission Line (TL) connecting the Acceleration Grid Power Supply system (AGPS) with the IS, carrying inside its High Voltage electrode all IS power and diagnostics conductors. The HVD1 and HVBA are installed inside a large HV hall, with controlled ambient conditions. The main design choices leading to the final hall layout, integrating the HV experimental equipment and the conventional building plants by assuring the necessary clearance required by such very high electric insulation level, are presented. Moreover, the paper presents the manufacturing design developed by the supplier to meet the design constraints and requirements of the technical specification for such unconventional devices. The factory type tests to validate the design and release the manufacturing of the HVD1 are described together with the electrical, mechanical and thermal tests carried out on the HVBA. Finally, the paper reports on the on-site installation and commissioning and testing activities carried out in 2017 and on the final acceptance at full voltage foreseen in 2018.

        Speaker: Dr Marco Boldrin (Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete S.p.A.))
      • 11:00
        The ITER in vessel coils – design finalization and challenges 2h

        A design update of the ITER In-Vessel Coils (IVCs) has been launched after the prototype coil manufacture in 2014 revealed some major issues in particular related to brazing and joints inside the coils. In parallel a review and update of the plasma operating scenarios and requirements of the IVCs system has been done and a refined set of plasma pulses and corresponding load scenarios of the IVCs has been elaborated in collaboration with the ITER plasma control team. For the fatigue analysis of the vertical stabilisation (VS) coils, the spectrum type electromagnetic loading has been analysed by applying the Rainflow cycle counting method. Furthermore, the maximum currents during transient plasma events have been assessed considering actual operating currents leading to more representative load cases. The main IVC component modifications are the conductor material and the winding pack support structure. In view of risk reduction and cost efficiency, the conductor design has been unified among the IVC system. Cu brazing has been replaced by welding and the conductors are not brazed to the winding pack bracket but clamped. Huge effort has been made in the structural analysis and design integration of the IVC components. With a strong support from the design integration and teams of the other ITER in vessel components, a fully integrated design baseline of the IVC components has been established.
        Feasibility studies and mock-ups performed in the frame of the design finalization have identified the main critical activities during manufacture such as the joint assembly, the ELM coil bracket welding and the in-situ winding of the VS coils. The status of the IVC conductor procurement involving two suppliers for the first phase covering process qualification is being shown as well. Key challenges experienced during this first phase along with results on mechanical and electrical tests are presented.

        Speaker: Dr Alexander Vostner (ITER Organization)
      • 11:00
        The new attempts for the in-vessel pressure gauge in the KSTAR plasma 2h

        The in-vessel pressure gauge refers to a vacuum gauge installed inside a vacuum vessel of tokamak. The inside of the vacuum vessel in which the fusion reaction occurs have to discharge the impurities and ashes those generated as a byproduct of the fusion reaction to sustain efficient state. Also the impurities and ashes of the plasma impinging on the divertor plate along the magnetic field are transformed into neutral gas and then exhausted. When the divertor contacting with the high-temperature plasma experiences a enormous heat load. And an in-vessel pressure gauge on the back of the diverter is required to analyze and control the amount of the particles impinging on the divertor.
        In addition, the gauge calibration can greatly enhance the convenience of understanding the divertor design and analysis, and the physical phenomena of structures in vacuum vessels when expressed in terms of conventional vacuum pressure.
        We have made new attempts to improve the usability and performance of the in-vessel pressure gauge. There are three new attempts: first, direct filament heating current setting; second, gauge calibration linked to gas input; and third, shot by shot calibration. In this paper we will describe detatils and pros and cons.

        Speaker: Myungkyu Kim (KSTAR Control research team National Fusion Research Institute)
      • 11:00
        The remote handling systems of ifmif-dones 2h

        The DEMO Oriented Neutrons Source (DONES) consists of complex systems and massive components that need to be on site assembled and maintained. For several of them it is required to perform maintenance, inspection and monitoring tasks over many years in a hostile environment and in efficient, safe and reliable manner. The maintenance of DONES’ systems and components, located mainly in the Test Systems (TS), in the Lithium Systems (LS) and in the Accelerator Systems (AS), is classified as a remote handling (RH) Class 1st activity and as such is considered a crucial and essential activity whose success will strictly depend on the DONES RH capability. According to this, a Remote Handling Systems (RHS) for DONES, which comprises the whole set of Remote Handling Equipment (RHE) and tooling for the execution of maintenance tasks, has been designed. A wide range of technologies is involved: special cranes, manipulator arms, lift interface frames, special cameras, control systems and virtual reality.
        In this paper an overview on the design of the main robotic systems and tooling of the RHS of DONES, including design requirements, functions and maintenance tasks to be performed, is given.

        This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission

        Speaker: Gioacchino Miccichè (ENEA)
      • 11:00
        Thermal diffusivity of ceramic breeder beds 2h

        In the solid Breeder Blanket (BB) concepts both tritium release and heat recovery depend on the thermal performances of the breeding zone. Within the R&D activities of the Helium Cooled Pebble Bed (HCPB) breeding blanket, the knowledge of the thermal diffusivity of the breeder beds is of fundamental importance to model the transient heat transfer during the power pulses of the fusion machine. The aim of the present study is to investigate the thermal diffusivity of the breeder beds at BB relevant conditions; to this end the line heat source probe method was employed. The method uses a linear heater (probe) embedded in the material to be investigated. The thermal diffusivity of the material is analytically derived by measuring the temperature rise as a function of time at a point close to the probe. By knowing the thermal conductivity of the material, the specific heat capacity can be derived in addition.
        An experimental facility was conceived for the investigation of the thermal diffusivity of granular beds at breeder blanket relevant temperatures, mechanical state, purge gas type and pressure. Based on the geometrical restriction of the experimental facility, the experimental parameters were tailored with a series of FEM simulations to reduce the error of the measurements.

        Speaker: Simone Pupeschi (IAM-KWT Karlsruhe Institute of Technology)
      • 11:00
        Thermal Induced Electromotive Force Measurements On Twisted Pair Mineral Insulated Cables 2h

        The components of the ITER Diagnostics are located all over on the inner and outer shell of the Vacuum Vessel, in the Ports, on the Divertor Cassettes and in the Cryostat as well. Sensors require electrical transmission lines to transmit both of the diagnostic and control signals across the vacuum boundaries. To transmit the signals, Mineral Insulated cables will be used.
        During the last 2 years under an F4E contract, the Tokamak Services for Diagnostics group at Wigner RCP has performed several experiments on MI cables, in close collaboration with F4E and IO team members. This paper outlines one of these experiments to measure the thermal gradient induced voltage in the cable cores. This was needed to evaluate the electrical noise levels due to thermal gradients in order to know if the noise has significant effect on the signal of some sensitive diagnostics. The subjects of the tests was a couple of Twisted Twin Mineral Insulated Cables, as this type of cables are used by the magnetics in the ITER machine.
        The work leading to this publication has been funded partially by Fusion for Energy under the Specific Grant Agreement F4E-FPA-328-SG07. This publication reflects the views only of the authors, and Fusion for Energy cannot be held responsible for any use, which may be made of the information contained therein.

        Speaker: Dr Miklós Palánkai (Plasmaphysics Department, Wigner RCP)
      • 11:00
        Thermal mixing enhancement of liquid metal film-flow by various obstacles under vertical magnetic field 2h

        Heat removal from liquid metal film flow has been widely studied for liquid divertor concepts of fusion reactor. In this study, thermal mixing characteristics of the liquid metal film-flow with locally heated on the surface under the vertical magnetic field was experimentally investigated by using various types of obstacle as a vortex generator. The temperature distributions on the bottom wall were measured by 40 thermocouples installed on the channel bottom downstream of the vortex generator. In order to evaluate how much heat transports from the locally heated free-surface to the bottom wall, the efficiency of heat transport was investigated for the various vortex shaped generator. The velocity distributions were measured with using a rake consisted of 40 electrical potential probes on the downstream of the channel. A delta-wing, hemisphere or cubic obstacle was used as a vortex generator installed at the center of the bottom. The experiments were conducted for laminar flow region where Re=1000-1700 and in the range of N(=Ha2/Re)=0-36.0 in the presence of the vertical magnetic field in the acrylic rectangular duct. GaInSn alloy was used as a working fluid. According to the comparison of heat flux distributions obtained by the experiments, the entire distributions moved towards the upstream with increasing of the strength of the vertical magnetic field for all the vortex generator types. This tendency meant the heat removal from the free-surface of liquid metal film-flow quickly in case of relatively high vertical magnetic field. The horizontal velocity distribution of the liquid metal film-flow obtained by the probe rake showed the typical M-shape distributions in case of without the vortex generator, some MHD characteristics of the velocity fluctuations were observed in case of various types vortex generators.

        Speaker: Dr Tomoaki kunugi (Department of Nuclear Engineering, Kyoto University)
      • 11:00
        Thermal-mechanical analysis and design optimisations of the IFMIF-DONES HFTM 2h

        The International Fusion Materials Irradiation Facility – DEMO-Oriented Neutron Source (IFMIF-DONES) is planned to generate a high flux of 5E16 neutrons/s with First Wall relevant energy spectrum. The High Flux Test Module (HFTM), is the dedicated assembly to bring the material specimens into the high flux region of the neutron source and maintain the specified irradiation conditions.

        Based on neutronic analysis, the nuclear heating of the HFTM under normal operational conditions was calculated. A subsequent CFD analysis was performed to determine the temperature field of the HFTM structure. This temperature field and the internal operating pressure are the basis for the structural analysis and stress assessment of the HFTM body in regard of the requirements of the RCC-MRx code. Furthermore, design optimisations are presented especially in regard of the installation of the irradiation capsules in the HFTM during assembly. To keep installation time of HFTM in the Test Cell (TC) as short as possible, connectors are foreseen on top of the HFTM, which transmit signals and electrical power from and to the irradiation capsules. A design proposal of a quick and fail-safe HFTM-sided multi-coupling solution for these connectors is presented.

        Acknowledgments: This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training program 2014-2018 under grant agreement No. 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission.

        Speaker: Dr Florian Schwab (Institute for Neutron Physics and Reactor Technology (INR) Karlsruhe Institute of Technology (KIT))
      • 11:00
        Thermo-hydraulic analyses and fatigue verification of the Electrostatic Residual Ion Dump for the ITER HNB 2h

        In the ITER Heating Neutral Beam Injector (HNB) the remaining charged particles after the neutralization process will be removed by an Electrostatic Residual Ion Dump (ERID) where electrostatic fields are used to deflect the ions that are so dumped on to five panels, each one composed of 18 separate CuCrZr Beam Stopping Elements (BSEs).
        The thermal loads applied on panels were calculated for full beam-on power and for the most severe beam conditions (3 mrad beam core divergence and 30 mrad beam halo divergence) taking into account also possible off-normal conditions due to possible variation of the gas throughput in the Neutraliser channels that can affect the heat fluxes on the ERID.
        A customized finite element (FE) code was developed to allow parametric and coupled solid-fluid simulations in the sub-cooled boiling conditions expected in the swirl flow of cooling channels. The model was used to verify the expected temperature at the inner cooling channels and at the external surface, the pressure drop, the Critical Heat Flux (CHF) and bulk boiling.
        Thereafter, thermo-mechanical analyses and fatigue verifications of the ERID panels were undertaken with non-linear material properties by using elasto-plastic properties of CuCrZr and by considering the most demanding condition produced by breakdown and beam-on/off cycles during the operation. The thermal results produced by transient thermo-hydraulic analyses were applied as body load onto the mechanical model to calculate stress and strain fields. The analysis results were post-processed for the fatigue verification by evaluating local effects and creep-fatigue interaction.

        Speaker: Dr Matteo Zaupa (Consorzio RFX,)
      • 11:00
        Thermo-Mechanical behaviour of ITER Blanket Modules interface between First Wall and Shield Block 2h

        The Blanket System provides a physical boundary for the plasma transients and contributes to the thermal and nuclear shielding of the vacuum vessel (VV). It consists of modular shielding elements, the blanket modules (BM), which are attached to the VV. Each BM consists of two major components: a plasma-facing first wall panel (FW) and a shield block (SB). They are connected by means of a preloaded central bolt working in tension, and they further interface each other through a set of eight compression pads installed in each FW. The design of these pads has a major importance because of the very demanding Thermal and Electromagnetic (EM) loads they are subjected to. The thermal gradients taking place during the normal operation thermal cycling tend to bend the SB and the corresponding FW in different ways and they can induce an over loading of the pads or a loss of pretension with the risk of exceeding the damage load of the pads or of the bolt under the EM Major Disruptions and Vertical Displacement events. These phenomena have been numerically investigated on BM#14 (including a representative Enhanced Heat Flux FW) and BM#06 (including a representative Normal Heat Flux FW). A full set of hydraulic (CFD), thermal and structural finite element analyses and related structural (SDC-IC) assessments was performed, focusing on the behavior of the central bolt’s pretension and on the reaction forces at the pads at significant instants of the thermal cycling. The results obtained demonstrated that the pads are within the allowable load for static conditions for all possible combination of thermal and EM loads. However the presence of a physical gap at some pads indicates that a fully conclusive study requires dynamic analysis techniques.

        Speaker: Dr Fabio Vigano (L.T.CALCOLI S.R.L.)
      • 11:00
        Thermometric chains for ITER superconductive magnets 2h

        High environmental constraints are applied on the ITER magnets and therefore on their cryogenics thermometric chains. Accurate and reliable temperature measurements of ITER magnets and their cooling circuits is of fundamental importance to make sure they operate under well controlled and reliable conditions. Therefore, thermometric chains shall reach a high operation reliability. In this paper, we present the full thermometric chain installed on the ITER magnets and their helium piping as well as the associated components production.
        The thermometric chain is described from the sensor and its on-pipe assembly, to the signals conditioning electronics. The thermometric block design is based on the CERN's developed one for the LHC, which has been further optimized thanks to thermal simulations carried out by CEA to reach high quality level of industrial production. The ITER specifications are challenging in terms of accuracy and call for severe environmental constraints, in particular regarding irradiation level, electromagnetic immunity and distance between the sensors and the electronic measuring system. A focus will be made on this system, which has been recently developed by CEA: based on a lock-in measurement and amplification of small signals, and providing web interface and software to monitor and record temperatures. This measuring device provides a reliable and fast system (up to 100 Hz bandwidth) for resistive temperature sensors between a few ohms to 100 kohms.
        During last two years, around 2500 thermo-blocks and nearly 50 crates of 5 boards, with eight channels each, have been produced and tested at CEA low temperature laboratory. All these developments and tests have been carried out thanks to three test benches built up at CEA Grenoble and in one industrial electronics laboratory. Test benches results on the entire production will be presented.
        The views and opinions expressed herein do not necessarily reflect those of the ITER Organization.

        Speaker: Dr Jean Manzagol (INAC/SBT, Univ. Grenoble Alpes, CEA)
      • 11:00
        Three-dimensional model of DEMO-FNS reactor for neutronics calculations and radiation shield problems 2h

        At the present stage of the Demonstration Fusion Neutron Source (DEMO-FNS) design the actual problem is a development and use of the three-dimensional model of this device to the solution of various neutronics problems for the integration of the basic technological systems of tokamak. The radiation safety and the development of the radiation shield are crucial problems which significantly affect the layout and cost of the installation. Besides the basic radiation shield, the FNS blanket obtains shielding properties. The blanket design is currently being developed. The blanket is also a powerful source of neutrons. The recent neutronics studies of the blanket showed a crucial influence of a coolant substance on the rate of transmutations of the transuranium elements and the generation of tritium in the system. This paper presents the three-dimensional DEMO-FNS model developed for the Monte Carlo calculations and the results of its application for the estimation of the characteristics of the DEMO-FNS radiation shield to protect the materials of the superconducting coils of toroidal magnetic field (TMFC) from the neutrons and secondary photons, taking into account the replacement of the water coolant of the blanket to the carbon dioxide one. It was performed the analysis of the neutron balance, the neutron energy spectra and energy release of the neutrons and secondary photons in the shield, case and superconductor on the inner and outer contour of TMFC. It was found that the energy release in the case and superconductor on the outer contour of TMFC located between the injector port and the blanket maintenance port was overestimated in the area facing the injector port by about 10 times in comparison with the monolithic shield option. This indicates that the increase of the local iron-water shield of the injector port is required.

        Speaker: Alexey Zhirkin (Kurchatov Complex of Fusion Power and Plasma Technologies Fusion Reactor Department National Research Centre Kurchatov Institute (NRC KI))
      • 11:00
        Tokamak T-15MD - two years before the physical start--up 2h

        At the present time, in the NRC Kurchatov Institute within the Federal Target Program “Nuclear energy-technologies of new generation for period 2010 - 2015 and to the prospect until 2020” the tokamak T-15MD with supporting facilities is being built. The preassembly of the tokamak T-15MD magnet system together with vacuum chamber was completed at a plant in Bryansk. All elements of the magnet system and vacuum chamber will be delivered to the NRC “Kurchatov Institute” in Moscow for the tokamak T-15MD assembly. It is expected that the T-15MD assembly will be completed by the end of 2018. The reconstruction of sub-station 110/10/0,4 kV for own needs was completed in 2017 and the reconstruction of the main sub-station 110/10/1 kV, 300 MW is completed in 2018 year. Twenty- two of the new transformers 10/1 kV and the same new thyristor convertors will be installed during the 2018-2019. One gyrotron with output power 1 MW for pre-ionization should be installed in 2019. Tokamak T-15MD connection to water and electrical communication and also the adjustment of control system will be completed in the middle of 2020. Physical start-up T-15MD is scheduled for December 2020 year.

        Speaker: Dr Petr Khvostenko (NRC "KURCHATOV INSTITUTE")
      • 11:00
        Towards tritium measurements in W based on ps-LIBS diagnostics 2h

        Determining tritium concentration within plasma facing components (PFC) of a thermonuclear reactor is crucial in terms of safety. As an example, tritium implantation can be high at the material surface (10-2 % at in W) and low in the bulk of the PFC (10s of ppm). In addition, a simultaneous implantation of tritium, deuterium and helium takes place. An in situ technique used to measure concentrations must enable isotopic discrimination and the access to T/D/He concentration profiles. In these conditions, Laser-Induced Breakdown Spectroscopy (LIBS) seems to be an appropriate method.
        LIBS is based on the interaction of a high energy laser pulse and the material. The irradiated material is ablated and transformed into low temperature thermal plasma whose spectroscopic analysis provides its composition and therefore that of the material.
        LIBS using nanosecond pulses has been proved to be irrelevant [Mercadier et al, Journal of Nuclear Materials 414 (2011) 485] in determining the D/T ratio in metallic samples, a specific experimental setup has been elaborated based on the implementation of picosecond pulses. Results obtained on samples implanted with hydrogen isotopes will be presented. The depth of craters following the ablation has been measured. With pulse energy of ~ 20 mJ, low ablation rates (~ 100 nm/pulse) are obtained. The plasma spectroscopic analysis is performed on the [200, 800] nm spectral range for W (for fusion), Al (as Be substitute) and Si (as benchmark) implanted with mainly D+ ions produced by plasma discharges or ion beams. Temporal studies have been performed to identify the appropriate time window compatible with a plasma in thermochemical equilibrium. We will finally present concentration profiles obtained with a double pulse technique, when a second (5 ns) laser pulse is absorbed by the plasma to increase the signal to noise ratio and enables the measurement of He.

        Speaker: Dr Philippe Magaud (CEA Cadarache)
      • 11:00
        Tungsten coatings repair: an approach to increase the lifetime of plasma facing components 2h

        Tungsten coatings have received a great deal of attention as a technical solution for plasma facing components (PFC) in present-day tokamaks owing to their advantages over bulk tungsten, such as lower cost and weight. Nevertheless, tungsten (W) coatings are hard and fragile. Their lifetime is mainly limited by two degradation mechanisms occurring during the operation of the tokamak: erosion resulting from exposure to long plasma discharges and local damages caused by runaway electrons.
        In this paper, an alternative solution to the current replacement of damaged PFCs is proposed. It consists of combining a stripping method and a physical vapor deposition (PVD) method for repairing their surface, with a particular interest for actively water-cooled components made of copper alloy.
        First, the process of removing tungsten was investigated on CuCrZr samples coated with a 15µm layer of W using 3 carefully selected techniques: laser ablation, sandblasting and chemical attack. For each technique, etching properties were assessed by means of microscopic observations and topography analyses. The results presented here also highlight the issues related with controlling the etching depth and the treated surface area. The second part of the paper focuses on the fabrication of a new W coating layer by PVD. To ensure the performance of the coating after repair, the properties of the new W layer were characterized and compared to the one of the old coating.
        In addition to establishing and validating a whole repair procedure for PFCs, this study will provide technical specifications to consider the integration of this technology on a robotic arm. A first application could be the in-situ repair of W coatings in WEST tokamak. From a technical point of view, the integration of different systems on a robotic arm might also be beneficial for ITER where maintenance and inspection will be carried out by robots.

        Speaker: Dr Mathilde Diez (IRFM, CEA)
      • 11:00
        Type tests of the ITER Switching Network Unit components and Protective Make Switches 2h

        In ITER, each circuits of the central solenoid as well as poloidal field coils PF1 and PF6 is provided with a system for plasma initiation, called the Switching Network Unit (SNU), able to provide up to 8.5 kV for the coils. This will be obtained by inserting resistors in series with the pre-energized coils with the help of a DC current commutation unit (CCU) composed of connected-in-parallel mechanical bypass switch and a thyristor circuit breaker.
        The paper describes the results of the work on development and testing of the main mechanical DC switches forming part of the CCU, namely, Bypass Open Switch (BPOS) and Bypass Make Switch (BPMS). These switches provide commutation of currents up to 55 kA at a voltage of 10 kV within 2-4 ms.
        In addition, protective make switches (PMS) similar to the BPMS are considered. These devices are intended to shunt the AC/DC converters supplying power to the coils in case of their failure or in case of a quench, when the energy stored in the magnet system is extracted by a fast discharge to resistors.
        After a short description of the design, the paper focuses on the procedure and results of the type tests on full-scale prototypes. In addition to electrical, hydraulic and pneumatic tests the test program implemented in 2017-2018 included EMC tests, functional tests with control system at rated currents and peak current tests at a current higher than 350 kA.
        The successful results of the type tests confirmed the suitability of the design and compliance with the ITER requirements making it possible to start manufacturing of the switches for delivery to the ITER Site.

        Speaker: Dr Vladimir Bedran (Joint Stock Company "D.V. Efremov Institute of electrophysical apparatus")
      • 11:00
        Type tests of the ITER Switching Network Unit components and Protective Make Switches 2h

        In ITER, each circuits of the central solenoid as well as poloidal field coils PF1 and PF6 is provided with a system for plasma initiation, called the Switching Network Unit (SNU), able to provide up to 8.5 kV for the coils. This will be obtained by inserting resistors in series with the pre-energized coils with the help of a DC current commutation unit (CCU) composed of connected-in-parallel mechanical bypass switch and a thyristor circuit breaker. The paper describes the results of the work on development and testing of the main mechanical DC switches forming part of the CCU, namely, Bypass Open Switch (BPOS) and Bypass Make Switch (BPMS). These switches provide commutation of currents up to 55 kA at a voltage of 10 kV within 2-4 ms. In addition, protective make switches (PMS) similar to the BPMS are considered. These devices are intended to shunt the AC/DC converters supplying power to the coils in case of their failure or in case of a quench, when the energy stored in the magnet system is extracted by a fast discharge to resistors. After a short description of the design, the paper focuses on the procedure and results of the type tests on full-scale prototypes. In addition to electrical, hydraulic and pneumatic tests the test program implemented in 2017-2018 included EMC tests, functional tests with control system at rated currents and peak current tests at a current higher than 350 kA. The successful results of the type tests confirmed the suitability of the design and compliance with the ITER requirements making it possible to start manufacturing of the switches for delivery to the ITER Site.

        Speaker: Dr Vladimir Bedran (Joint Stock Company "D.V. Efremov Institute of electrophysical apparatus")
      • 11:00
        Ultra high vacuum ZnSe window flange design for Phase Contrast Imaging diagnostics for the Wendelstein 7-X stellarator 2h

        The Phase-Contrast Imaging (PCI) system is used to measure plasma density fluctuations in the W7-X stellarator at the Max-Planck-Institut für Plasmaphysik (IPP) in Greifswald, Germany. For this purpose, an expanded CO2 laser beam with a wavelength of 10.6m passes through the plasma and the scattered laser beam components yield information on plasma density fluctuations. The laser beam is expanded on an optical table by telescope optics to the desired beam diameter. A mirror system directs the beam through a zinc selenide (ZnSe) window at the entry port flange directly through the plasma center. Through a second opposite ZnSe window on the exit port flange, the beam is directed via a similar set of mirrors to the receiver optical table where the measurement takes place. A typical diameter of the laser beam is approx. 120mm. Due to the wavelength of the laser beam, ZnSe windows must be used.
        The challenge is to seal the window, which represents the vacuum barrier, into the counter flange ultra high vacuum (UHV) tight. Additionally, the seal needs to fulfill W7-X requirements as compatibility with the bake-out temperature of 150°C and electronen cyclotron resonance heating (ECRH) radiation resistance.
        A standard solution is not provided commercially, since ECRH resistance demands for an all-metal seal and VITON® or another elastomer seal cannot be used. The challenge is the special window material (ZnSe) and the large window diameter. A metal gasket based on the principle of the CF gasket considered as a safe standard connection in UHV technology, but couldn’t be used on the glass side. One solution seemed to be the HELICOFLEX® seal, which combines the elastic component with the temperature resistance. In addition to the special sealing concept for the PCI, the poster shows a market overview of the standard windows for the UHV area.

        Speaker: Dr Christoph von Sehren (E5/Max-Planck-Institut IPP Greifswald, IPP Greifswald)
      • 11:00
        Uncertainty analysis of an SST-2 fusion reactor design 2h

        Systems codes are a powerful tool for designing the next generation of nuclear fusion reactors. By exploring a large design space in a single calculation, they can obtain highly optimised solutions. However, while a single design is informative, it does not give the whole picture. Often new designs will push boundaries, whether that involves scaling to new physical regimes or applying new technologies. All of this will introduce uncertainty which needs to be quantified to give a complete understanding of the performance of a proposed reactor. Uncertainty analysis can then inform about high impact areas, critical design aspects or simply confirm the robustness of the design.

        The SST-2 reactor is a proposed medium sized device with low fusion gain (Q = 5) and capable of producing fusion power from 100 to 300 MW. Tritium breeding will be achieved by having breeding blankets only on the outboard side, while on the inboard side, shielding blankets will be placed due to the limited space available. The magnets will be superconducting in nature to achieve steady state operation. In this work, we are performing an uncertainty analysis of its latest published concept design (Srinivasan et al. 2016; Fusion Eng. Des., 112, 240). We initially use the systems code PROCESS to reproduce the SST-2 design originally obtained using another systems code, SPECTRE, highlighting and explaining the differences between the two. We then apply a Monte-Carlo based uncertainty quantification tool using PROCESS to explore its predicted performance and highlight high impact areas. By building a more comprehensive understanding of the implications of the uncertain parameters we can give better predictions to aid the development of the design.

        Speaker: Stuart Muldrew (Culham Centre for Fusion Energy UK Atomic Energy Authority)
      • 11:00
        Upgrade of Thomson scattering system on VEST 2h

        Upgrade of the Thomson scattering (TS) system in Versatile Experiment Spherical Torus (VEST) is planned for measuring the electron temperature and density with higher reliability and higher time resolution. The existing TS system has difficulties on measuring single plasma discharge, since it uses a laser with energy of 0.65 J and repetition rate of 10 Hz, while the pulse duration of the plasma is ~20 ms. Recently, additional heating sources such as neutral beam injection and electron cyclotron heating are installed, thus, the upgrade of the TS system has been required to provide the time evolutions of the electron temperature and density for heating efficiency analysis. The upgrade is mainly related with two parts. First, a new laser is utilized for increasing both the signal-to-noise ratio (SNR) and the time resolution. The laser generates 10 pulses with 1 ms time interval in every 2 s, and each pulse has the energy of 2 J. And a switched capacitor type digitizer with 32 channels and 5 GS/s is newly employed for storing a number of pulse signals. Second, the optical fibers for transferring the collected TS photons are improved as well as the laser. Although the collection solid angle and the numerical aperture between the collection lens and the optical fiber are well matched each other, it uses only half of the maximum etendue of the polychromator. Therefore, the fiber bundle is designed to optimize the optical properties among the lens, fiber, and the polychromator. Compared to the current system, replacement of the laser, modification of laser injection optics, and fiber bundles are expected to improve the efficiency more than 6 times. Furthermore, in the temporal point of view, it decreases the number of required plasma discharges to be a tenth.

        Speaker: Dr Doyeon Kim (Department of Nuclear Engineering, Seoul National University)
      • 11:00
        Verification of hydraulic performance for the DEMO divertor target cooling 2h

        This paper presents the design activities and test of a vertical target mock-up, developed under the pre-conceptual design phase for DEMO Work Package DIV-1 “Divertor Cassette Design and Integration” under EUROfusion Power Plant Physics & Technology (PPPT) program.
        The activities about the Divertor Outboard Vertical Target cooling mock-up are presented in term of CAD model (CATIA), thermo-hydraulic numerical simulation (ANSYS-CFX), structural analysis (ANSYS Mechanical), structural integrity verification (RCC-MRx) and manufacturing procedure. Moreover, the mechanical dimensions of support systems for plasma facing components (PFCs), manifold and diffuser have been analysed in detail, in order to avoid structural fault during test.
        Test procedures are discussed, taking into account design parameters, design code and facility performances. The actual alloy (CuCrZr) selected for PFCs of EU DEMO divertor has been used also for the mock-up ,while two options are still under evaluation for manifolds/diffuser, CuCrZr and stainless Steel 316L(N)-IG, depending on the welding technology. Since the mock-up is mainly intended to verify hydraulic performances, it has been simplified by removing the W monoblocks from its PFCs.

        Speaker: Giuseppe Mazzone (ENEA)
      • 11:00
        WEST CODAC software quality management 2h

        COntrol, Data Acquisition and Communication (CODAC) real-time software codes are key elements for the operation of a fusion device as they can play a key role both for the machine protection and for the optimization of the experiments. The updating or upgrading of these software codes may be needed quite frequently in order to either correct bugs or include new functionalities, while these code modifications may introduce new errors or bugs that may have an impact on machine operational availability or even on some machine protection items.
        A new software quality management process has been developed and implemented in the WEST (W -for tungsten- Environment Steady-state Tokamak) CODAC in order to tackle these issues. Each source code modification, defined by a unique identifier called “production tag version”, is logged by the corresponding responsible officer, reported to the relevant team and finally submitted to validation by the relevant team leader. The scheduling of the deployment of any new version in the WEST CODAC package is coordinated at the appropriate level in order to minimize the possible impact on operation.
        The “production tag version” of each software code is also archived for each pulse as well as the pulse parameters, which allows coming back to a previous version whenever needed.
        The paper will first describe in details the principles used for the new WEST CODAC software quality management. The workflow process and associated tools will then be discussed. A few examples will eventually be detailed.

        Speaker: Dr Julian Colnel (IRFM, CEA)
      • 11:00
        Work-flow process from simulation to operation for the Plasma Control System for the ITER First Plasma. 2h

        The ITER Plasma Control System (PCS) is an essential component for ITER operations. It will include multiple controls loops as well as a number of support functions dedicated to providing input control parameters and distributing commands to actuators. In addition, a supervisory system within the PCS architecture will manage the orchestration of the PCS control loops during the discharge as well as an Event Handling system to react to real-time changes in the plasma and/or in the plant systems. To investigate and develop controllers and supervisory techniques, a simulation platform is being developed aiming at supporting the staged evolution of the PCS. The PCS Simulation Platform (PCSSP), based on Matlab/Simulink, is used to design and assess PCS control functions and verify the consistency of the PCS architecture. Moreover, the PCSSP is planned to assist in the implementation of PCS algorithms, directly deploying PCS code on the ITER Real Time Framework (RTF).
        This paper will present the overall work-flow processes from control simulations, control assessment, code generation, interfaces and deployment into the RTF and PCS commissioning. This will include interface management and co-ordination with ITER investment protection, the pulse scheduling system, and the plant system configuration during operations. How the entire preparation process will be managed, for ITER first plasma operations, will be reported here.

        Speaker: Dr Luca Zabeo (ITER Organization)
      • 11:00
        X-ray induced defects in advanced lithium orthosilicate pebbles with additions of lithium metatitanate 2h

        Advanced lithium orthosilicate (OSi) pebbles with additions of lithium metatitanate (MTi) as a secondary phase have attracted international attention as an alternative candidate for the tritium breeding in nuclear fusion reactors. In this research, the formation of radiation-induced defects (RD) in the OSi pebbles with various contents of MTi was analysed using X-ray induced luminescence (XRL). After XRL measurements, the accumulated RD were investigate using electron spin resonance (ESR), thermally stimulated luminescence (TSL) and diffuse reflectance spectrometry.
        The advanced OSi pebbles with additions of MTi are biphasic without solid solutions, and thus the formation mechanism and the structure of the formed RD during irradiation with X-rays is similar to the single-phase materials. In the XRL spectra, several bands with maxima at around 380, 420 and 810 nm were detected. The band at around 420 nm can be attributed to the formation of E’ centres (≡Si·) in the OSi phase, while remaining maxima are the result of the additions of MTi. In the ESR spectra, at least three first derivative ESR signals with g-factors 1.93, 2.003 and 2.040 were detected and attributed to Ti³⁺ centres, E’ centres and oxygen related defects. The TSL glow curves are complex and consist of several overlapped peaks between 300 and 450 K, while the TSL spectra consist of one main band with a maximum close to 450 nm. The blue emission was attributed to the radiative recombination of E’ centres or some variants of oxygen deficiency centres with oxygen related defects. It is assumed that the additions of MTi can increase the probability for the recombination processes of primary RD in the advanced OSi pebbles during irradiation and thus reduce the formation of chemically stable radiolysis products, for example colloidal lithium particles, which can interact with the generated tritium and form thermally stable lithium tritide.

        Speaker: Arturs Zarins (Laboratory of Radiation Chemistry of Solids Institute of Chemical Physics University of Latvia)
    • 13:00 14:30
      Lunch Break 1h 30m
    • 14:30 15:30
      O1.A LIPARI Hall - ATA Hotel Naxos Beach Resort

      LIPARI Hall - ATA Hotel Naxos Beach Resort

      Giardini Naxos

      Via Recanati, 26 Giardini Naxos, Messina - Sicily (Italy)
      • 14:30
        Plasma control for EAST long pulse non-inductive H-mode operation in a quasi-snowflake shape 20m

        Advanced magnetic divertor configuration is one of the attractive methods to spread the heat fluxes over divertor targets in tokamak because of enhanced scrape-off layer transport and an increased plasma wetted area on divertor target. Exact snowflake (SF) for EAST is only possible at very low plasma current due to poloidal coil system limitation. However, we found an alternative way to operate EAST in a so called quasi-snowflake (QSF) or X-divertor configuration, characterized by two first-order nulls with primary null inside and secondary null outside the vacuum vessel. Both modeling and experiment showed this QSF can result in significant heat load reduction to divertor target [1]. In order to explore the plasma operation margin and effective heat load reduction under various plasma conditions and QSF shape parameters, we developed ISOFLUX/PEFIT shape feedback control. In experiment, we firstly applied the control of QSF in a similar way to control the single null divertor configuration, with specially designed control gains. Reproducible QSF discharges have been obtained with stable and accurate plasma boundary control. Under Li wall conditioned, we have achieved highly reproducible non-inductive steady-state ELM-free H-mode QSF discharges with the pulse length up to 20s, about 450 times the energy confinement time by using low hybrid wave, ion cyclotron resonance wave (ICRH) and electron cyclotron resonance wave (ECRH) for the plasma current drive and heating. The capability of the QSF to reduce the heat loads on the divertor targets has been confirmed. This new steady-state ELM-free H-mode QSF regime may open a new way for the heat load disposal for fusion development.

        Speaker: Dr Bingjia Xiao (Division of Control and Computer Application, Institute of Plasma Physics, Chinese Academy of Sciences)
      • 14:50
        Implementation and exploitation of jet enhancements in preparation for dt operation and next step devices 20m

        JET presents some unique capabilities: the reactor fuel, ITER wall materials and the capability to confine the alphas. JET next T-T and D-T experimental campaigns can therefore address major physics and technological gaps for the development of fusion energy: the isotopic effects on confinement, the access the H mode and ELM behaviour. The total yield of the final D-T phase is expected to be 1013 n/s·cm2, a factor of six higher than the previous DTE1. In this context, three main aspects of JET capability have been recently improved: 1) scenario development to enhance performance 2) the quality of the measurements to maximize the scientific exploitation 3) specific technologies for ITER and DEMO.
        With regard to the scenarios, the performances of JET with a carbon wall have been reproduced up to a current of 3 MA, which supports the prediction of 15 MW fusion power in full DT. Moreover improved control systems (wall load protection, simultaneous control of ELM frequency and beta, plasma mixture) insure that the plasma configurations are compatible with the wall properties, from melting to retention and dust production.
        In terms of general diagnostic capability, JET can now deploy much better resolution diagnostics, particularly for the edge quantities, and a consistent set of techniques to diagnose the fast particles, from redistribution to losses, using techniques ranging from gamma ray spectroscopy to a scintillator probe and Faraday cups. A full calibration of the neutron diagnostics for the 14 MeV neutrons has just been completed successfully.
        With regard to ITER and DEMO relevant technologies, specific programmes are being pursued to investigate: the radiation field, the induced activity and dose rates and the radiation damage of materials. Dedicated studies for DEMO, including the tests of a new tritium pumping cycle and a tritium breeding blanket mock-up, are also almost completed.

        Speaker: Dr Andrea Murari (Programme Management Unit, EUROfusion Consortium)
      • 15:10
        Electromagnetic FEM studies of disruptions and engineering consequences for the power supply and coils design of planned upper divertor at ASDEX Upgrade 20m

        There is proposed a new upper divertor for the ASDEX Upgrade tokamak experiment [1]. It is planned to be equipped with internal coils for investigation of advanced magnetic configurations like e.g. „snowflake“. Due to the close vicinity of the coils to the plasma, high induced and very stiff voltages are expected during disruption events. Because only very vague analytical estimates of voltages, forces and coupling factors were available, an improvement by the help of finite element method (FEM) was envisaged. Therefore, recorded measurements of currents, plasma position, plasma profile and the geometry were integrated into the electromagnetic simulation as boundary conditions to calculate resulting field distributions during selected AUG disruption events. The time resolution can be better than 100 microseconds and the required computing resources are comparable small due to utilization of 2D axis-symmetry. The results were compared with magnetic probe measurements integrated into the tokamak. They are in good agreement. After this, the simulated geometry was modified to the target geometry including the new divertor to calculate all relevant parameters. The output of these calculations has strong implications for the coil and power supply design: (1) The power supply will be protected with a new kind of crowbar to avoid uncontrolled current and force rise of the coils and power supply damage due to overvoltage. The concept of this so called “ripping crowbar” will be introduced, which is under development, now. (2) The coil cable should be coaxial shaped to monitor isolation faults and to become inherently safe against single-turn shortcuts, identified as a destructive fault scenario.

        [1] A. Herrmann, et al., Fusion Engineering and Design 123 (2017) 508-512.

        Speaker: Dr Markus Teschke (Max Planck Institute for Plasma Physics (IPP))
    • 14:30 15:30
      O1.B NAXOS Hall - ATA Hotel Naxos Beach Resort

      NAXOS Hall - ATA Hotel Naxos Beach Resort

      Giardini Naxos

      Via Recanati, 26 Giardini Naxos, Messina - Sicily (Italy)
      • 14:30
        Myth of Initial Loading Tritium-2 : Practical commissioning strategy of DEMO-Japan without external source 20m

        The authors have pointed out that initial tritium needed for starting operation of fusion reactor can be made by DD and low T discharges with self sufficient blankets. Practical commissioning plan of Japanese DEMO was recently planned as a part of DEMO design activity. The early campaigns will require longer than a year of repeated low power pulses for operational purposes as the “power ascension tests”. Breeding blankets with TBR well above unity is designed based on a water cooled ceramic pebble concept. Complete tritium plants should be continuously operated with torus exhaust and blanket recovery for safety reasons. In the phase 0 of commissioning, DD pulses with small flux neutron followed by relatively long dwell periods are used to confirm all the nuclear functions of DEMO plant including the Balance of Plant and ancillary systems. This study analyzed dynamic tritium behavior in the plant. After the each discharges, produced tritium from DD reaction in the plasma, bred in the blanket, and fed to the vacuum vessel are all collected by the tritium plant and recovered from the isotope separation during the discharge during the dwell periods. After the DD phase, small amount of tritium is added to the fuel, however the required amount if gram level, that is available from the storage in the tritium plant. In the later operation, tritium concentration will gradually increase and the pulse length will longer, however for each shots, sufficient tritium can be prepared prior to the burning. It was found that in this commissioning scenario, no external tritium or additional DD shots for tritium production is required for DEMO program in Japan.

        Speaker: Dr Satoshi Konishi (Institute of Advanced Energy, Kyoto University)
      • 14:50
        Manufacturing of the first ITER Torus Cryopump 20m

        The first of nine ITER Torus and Cryostat Cryopumps has been successfully manufactured and delivered to ITER in summer 2017. This Pre- Production Cryopump is the first of the ITER cryopumps and may be used for the first pump down of the vacuum vessel or the cryostat.
        The pump has a 1.8 m diameter and a length of about 3 m and contains cryogenic pressure equipment with a charcoal coated adsorption stage integrated in a casing with a vacuum vessel plug combined with the largest all-metal vacuum valve ever built resulting in an overall weight of 8 tons. The design of the cryopump is the result of more than ten years of research and development finalized to the ITER built to print design to comply with the demanding requirements to be fulfilled during its operation starting from first plasma.
        The paper will outline the experience gained with the cryopumps manufacture and assembly. We discuss the components with confinement function as ITER style vacuum flanges, double bellows for the valve assembly and electrical feedthroughs. Many different requirements had to be addressed for the manufacture of these components and their integration in the cryopump.
        ITER is a Nuclear Facility, INB-174, and requirements for the operation in the primary vacuum and the nuclear confinement function demand a high level of quality control and inspection needs during all manufacturing stages. The Pre-Production Cryopump has been built in close and successful cooperation with F4E reflected in the adequate surveillance of the 27 suppliers required to fabricate the cryopump. The successfully built and delivered Pre-Production Cryopump will give a reliable basis for the F4E procurement of the eight Torus and Cryostat Cryopumps which are required for first plasma.

        The views and opinions expressed herein do not necessarily reflect those of the ITER Organization

        Speaker: Dr Matthias Dremel (PED, ITER Organization)
      • 15:10
        Neutronics of the IFMIF-DONES irradiation facility 20m

        Within the Early Neutron Source (ENS) project of EUROfusion the design of the accelerator based irradiation facility IFMIF-DONES (International Fusion Material Irradiation Facility- DEMO Oriented NEutron Source) is under development. The main mission of IFMIF- DONES is to provide the irradiation data needed for the construction of DEMO, a fusion power demonstration reactor developed in the frame of the Power Plant Physics and Technology (PPPT) programme of EUROfusion.

        The IFMIF-DONES facility consists of a deuteron accelerator, a liquid lithium target and a Test Cell with irradiation test modules as main systems. Neutronics has to provide the data which are required to design and optimize these systems, evaluate and prove their nuclear performance, and ensure a sufficient radiation protection. In addition, the radioactive inventories, produced during operation, have to be assessed to enable sensible maintenance and waste management strategies.

        A variety of neutronics simulations is needed to compute the nuclear responses for all systems and components and provide the radiation fields during operation, maintenance and shut-down periods. Such simulations require dedicated computational approaches adapted to the needs and peculiarities of the accelerator based IFMF-DONES neutron source. The ENS project thus builds on the development of specific tools and data for simulating the interactions of deuterons with the lithium target and the accelerator structures, the generation and transport of neutrons and photons, and the production of radio-active nuclides with the subsequent emission and transport of decay gamma radiation.

        The paper presents an overview of the IFMIF-DONES neutronics comprising both nuclear analyses and the applied computational approaches. Main issues are the nuclear analyses conducted lately for the Accelerator Facility and the Test Cell utilizing the specific codes and data developed and/or adapted for IFMIF-DONES. Related R&D issues are also addressed.

        Speaker: Dr Ulrich fischer (Karlsruhe Institute of Technology)
    • 14:30 15:30
      O1.C ETNA Hall - ATA Hotel Naxos Beach Resort

      ETNA Hall - ATA Hotel Naxos Beach Resort

      Giardini Naxos

      Via Recanati, 26 Giardini Naxos, Messina - Sicily (Italy)
      • 14:30
        Deep Learning: Towards Autonomous Remote Maintenance 20m

        Traditionally, remote maintenance in fusion and other nuclear plants has made use of man-in-the-loop telemanipulator devices in order to deal with the relatively unpredictable nature of tasks, and complex environments. Future fusion devices will require maintenance orders of magnitude more complex than at present, however it is infeasible to scale remote maintenance operations teams linearly with the increase in device complexity.

        Combined with increasing demand for productivity it will therefore be necessary to automate large numbers of maintenance tasks, many of which have previously been reliant on human-level dexterity and intelligence. This has previously been infeasible due to limitations of automation technology, however recent developments in artificial intelligence are showing a great deal of promise.

        The new and rapidly advancing field of deep learning has developed a number of advanced machine learning techniques which have not only surpassed the performance of previous methods, but also, in some cases, outperform human-level performance in a number of challenging task areas. We present recent developments in deep learning which are relevant to nuclear fusion, as well as a range of research activities which have been taking place at RACE related to deep learning for automation of robotic tasks in fusion environments. We describe how these new techniques are changing what can be considered possible in remote maintenance and how methods for remote maintenance are evolving.

        Speaker: Mr Robert Skilton (UK Atomic Energy Authority)
      • 14:50
        Reconstructing JET using LiDAR-vision fusion 20m

        The containment vessel of the Joint European Torus is a huge, complicated assembly with a myriad of components, all of which are important for plasma operation. As a research device, JET has been operated over many years and has been extensively rebuilt. During each maintenance shutdown, inspections and measurements of the Vacuum Vessel are carried out by means of dual-camera Stereo surveys, High-Resolution single camera surveys and precise Gap Gun measurements. This is a precise but labour-intensive process, taking tens of hours to complete a full survey.

        Due to rapid advancements in the field, combined visual-LIDAR techniques have evolved to the point where it is possible to carry out on-line, high-resolution measurements of the interior of buildings and scientific installations. Since the radioactivity inside the JET vessel is still low enough to allow consumer-grade electronics to survive unprotected, these advancements can be leveraged.

        We present work including the 3D mapping of the inside of the JET Torus using a combined LIDAR-Vision measurement and navigation system. Using one of the remote handling booms, we carry out a scan of the JET vessel. We compare the point cloud model with the CAD models of the JET installation using numerical methods, demonstrating mm and sub-mm accuracy with a dramatically lower survey duration compared to existing techniques. We also compare the estimated path of the scanner through the vessel with the recorded boom joint position data. Conclusions are drawn about the applicability of LIDAR systems to mapping and localisation problems within a Fusion environment as well as assessing the resulting accuracy of the scan.

        Speaker: Dr Emil Jonasson (RACE - Remote Applications in Challenging Environments, UK Atomic Energy Authority)
      • 15:10
        Engineering and integration design risks arising from advanced magnetic divertor configurations 20m

        The divertor configuration defines the power exhaust capabilities of DEMO as one of the major key design parameters and sets a number of requirements on the tokamak layout, including port sizes, PF coil positions, and size of TF coils. It also requires a corresponding configuration of plasma-facing components and a remote handling scheme to be able to handle the cassettes and associated in-vessel components the configuration requires.

        There is a risk that the baseline ITER-like single-null (SN) divertor configuration cannot meet the PFC technology limits regarding power exhaust and FW protection while achieving the target plasma performance requirements of DEMO or a future fusion power plant. Alternative magnetic configurations - double-null, snowflake, X-, and super-X - exist and potentially offer solutions to these risks and a route to achievable power handling in DEMO. But these options impose significant changes on machine architecture, increase the machine complexity and affect remote handling and plasma physics and so an integrated approach must be taken to assessing the feasibility of these options.

        In this paper we describe the work being undertaken, and main results so far, in assessing the impact of incorporating these alternative configurations into DEMO whilst respecting requirements on remote handling access, forces on coils, plasma control and performance, etc.

        Speaker: Dr Richard Kembleton (EUROfusion)
    • 15:30 16:00
      Coffee Break 30m
    • 16:00 17:20
      O1.A LIPARI Hall - ATA Hotel Naxos Beach Resort

      LIPARI Hall - ATA Hotel Naxos Beach Resort

      Giardini Naxos

      Via Recanati, 26 Giardini Naxos, Messina - Sicily (Italy)
      • 16:00
        Integrated current profile, normalized beta and NTM control in DIII-D 20m

        There is an increasing need for integrating individual plasma-control algorithms with the ultimate goal of simultaneously regulating more than one plasma property. Some of these integrated-control solutions should have the capability of arbitrating the authority of the individual plasma-control algorithms over the available actuators within the tokamak. Such decision-making process must run in real time since its outcome depends on the plasma state. Therefore, control architectures including supervisory and/or exception-handling algorithms will play an essential role in future fusion reactors like ITER. However, most plasma-control experiments in present devices have focused so far on demonstrating control solutions for isolated objectives. In this work, initial experimental results are reported for simultaneous current-profile control, normalized-beta control, and NTM suppression in DIII-D. Neutral beam injection (NBI), electron-cyclotron (EC) heating & current drive (H&CD), and plasma current modulation are the actuation methods. The NBI power and plasma current are always modulated by the Profile Control category within the DIII-D Plasma Control System (PCS) in order to control both the current profile and the normalized beta. Electron-cyclotron H&CD is utilized by either the Profile Control or the Gyrotron categories within the DIII-D PCS as dictated by the Off-Normal and Fault Response (ONFR) system, which monitors the occurrence of a Neoclassical Tearing Mode (NTM) and regulates the authority over the gyrotrons. The total EC power and poloidal mirror angles are the gyrotron-related actuation variables. When no NTM suppression is required, the gyrotrons are used by the Profile Control category, but when NTM suppression is required, the ONFR transfers the authority over the gyrotrons to the NTM stabilization algorithm located in the Gyrotron category. Initial experimental results show the potential of the ONFR system to successfully integrate competing control algorithms.

        This work was supported by the US Department of Energy under DE-SC0010661 and DE-FC02-04ER54698.

        Speaker: Dr Andres Pajares (Lehigh University)
      • 16:20
        Conceptual design of the COMPASS-U tokamak 20m

        The Institute of Plasma Physics of the CAS in Prague has recently started construction of new COMPASS-U tokamak. It will be a compact, medium-size (R = 0,85 m, a = 0,3 m), high-magnetic-field (5 T) device. COMPASS-U will be equipped by a flexible set of poloidal field coils and capable to operate with plasma current up to 2 MA and, therefore, high plasma density (~ 10^20 m^-3). The device is designed to generate and test various DEMO relevant magnetic configurations, such as conventional single null, double null, single and double snow-flake. The plasma will be heated using 4 MW Neutral Beam Injection (NBI) heating system with future extension by at least 4 MW Electron Cyclotron Resonant Heating (ECRH) system.
        COMPASS-U will be equipped with lower and upper closed, high neutral density divertors. Due to high PB/R ratio COMPASS-U will represent a device which will be able to perform ITER and DEMO relevant studies in important areas, such as the plasma exhaust or development of new confinement regimes. The divertors will use conventional materials in the first stage, however, in the later stage, the liquid metal technology, which represents a promising solution for the power exhaust in DEMO, will be installed into the lower COMPASS-U divertor. The metallic first wall will be operated at high temperature (approx. 300 °C) during plasma discharge, which will enable to explore the edge plasma regimes relevant to ITER and DEMO operation. The first plasma is scheduled for 2022.
        In this contribution, we will present the conceptual design of the COMPASS-U tokamak as well as the main tokamak components.

        Speaker: Dr Radomir Panek (Institute of Plasma Physics of the Czech Academy of Sciences)
      • 16:40
        Development of HINEG and its experimental campaigns 20m

        Fusion energy becomes essential to solve the problem of increasing energy demands. A high intensity D-T fusion neutron generator is keenly needed for the research and development (R&D) of fusion technology, especially for fusion materials research.
        The Institute of Nuclear Energy Safety Technology (INEST), Chinese Academy of Sciences (CAS) has launched the High Intensity D-T Fusion Neutron Generator (HINEG) project. The R&D of HINEG includes three phases: HINEG-I has been constructed and successfully produced a D-T fusion neutron yield of up to 6.4E12 n/s. The mechanism research of irradiation damage for materials can be carried out. HINEG-II aims at a high neutron yield of 1E15~1E16 n/s neutrons via high speed rotating tritium target system and high intensity ion source, which could be used to conduct material irradiation damage testing. The preliminary design and research on key technologies are on-going. HINEG-III is a volumetric fusion neutron source with yield of more than 1E18 n/s. The integration testing of nuclear system engineering could be performed.
        As an important platform for fusion technology and safety research, HINEG can be used to carry out the neutron activation and irradiation testing not only for structural but also functional materials to assess the performance and reliability, such as structural materials for the blanket, neutron multipliers and ceramic breeders for tritium fuel production, suitable radiation resistant thermosets for the electrical insulation of the superconducting magnets, in-vessel conductor coils, liquid-metal coolants, etc. The fusion neutron irradiation testing is being conducted on China Low Activation Martensitic (CLAM) steel, which has been developed by INEST and selected as the primary candidate structural material for Chinese Helium Cooled Ceramic Breeder ITER Test Blanket Module (CN HCCB TBM). Moreover, the performance of components under neutron irradiation can also be assessed on HINEG platform, such as the tritium breeding blanket and shielding blanket.

        Speaker: Dr Yican Wu (Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences)
      • 17:00
        Neutron spectrum determination at the ITER material irradiation stations at JET 20m

        The experiments that are planned over the next few years at the Joint European Torus (JET), notably including a deuterium-tritium (DT) experimental phase, are expected to produce large neutron yields, up to 1.7E21 neutrons. The scientific objectives of the experiments are linked with a technology programme, WPJET3, to deliver the maximum scientific and technological return from those operations, with particular emphasis on technology exploitation via the high neutron fluxes predicted in and around the JET machine. Importantly, the programme aims to extract experimental data relevant to the international effort to design, construct and operate ITER. The data expected to be retrieved under the JET experimental program will support, develop and improve the radiation transport and activation simulation capabilities via benchmarking and validation in relevant operational conditions. Such capabilities are important and are applied extensively to predict a wide range of nuclear phenomena and impacts associated with components and materials that will be used in ITER operations.

        This paper reports the status of activities conducted as part of the ACT sub-project collaboration under WPJET3. The aim of the subproject is to take advantage of the significant 14 MeV neutron fluence expected during JET operations to irradiate samples of materials that will used in the manufacturing of main ITER tokamak components. The paper will provide analysis of the characterisation work at irradiation stations at JET performed in a previous deuterium campaign using dosimetry foil measurements, and give the status of irradiation experiments at JET that are ongoing in 2018 using real ITER materials. The experimental results are further used, together with calculated dosimetry foil response functions (Ti, Mn, Co, Ni, Y, Fe, Co, Sc, Ta) and spectrometry unfolding methodologies, to derive neutron spectrum information at irradiation positions, which are compared to those derived from neutron transport simulations.

        Speaker: Dr Lee Packer (Nuclear Technology, UKAEA,)
    • 16:00 17:20
      O1.B NAXOS Hall - ATA Hotel Naxos Beach Resort

      NAXOS Hall - ATA Hotel Naxos Beach Resort

      Giardini Naxos

      Via Recanati, 26 Giardini Naxos, Messina - Sicily (Italy)
      • 16:00
        Innovative Technology for 6Li Enrichment using Electrodialysis with Lithium Ionic Conductor 20m

        Tritium needed as a fuel for fusion reactors is produced via neutron capture by lithium-6 (6Li). However, natural Li contains only about 7.8% 6Li, and enrichment of 6Li up to 90% is required for adequate tritium breeding in fusion reactors. In Japan, lithium isotope enrichment methods have been developed to avoid the environmental hazards of using mercury. However, the isotope separation coefficient and efficiency is too low to meet the practical need of large mass production of 6Li.
        Therefore, new Li isotope separation technique using a Li ionic superconductor functioning as a Li isotope separation membrane (LISM) have been developed. First of all, I investigated the ionic mobility of lithium isotopes in ionic superconductor. Combing the first principle and the kinetics Monte Calro simulation, I calculate the diffusion constant of 6Li and 7Li.
        Furthermore, examinations of Li isotope separation using LISM with electrodialysis ware performed. Because the mobility of 6Li ions is higher than that of 7Li ions, 6Li can be enriched on the cathode side of a cell. Using Li0.29La0.57TiO3 (LLTO) as the Li ionic superconductor was prepared. After electrodialysis, I obtained a maximum of 1.05 for the 6Li isotope separation coefficient. This result showed that the 6Li isotope separation coefficient of this method is the same as that of the amalgamation process using mercury (1.06).

        Speaker: Dr Tsuyoshi Hoshino (Fusion Energy Research and Development Directorate, National Institutes for Quantum and Radiological Science and Technology (QST))
      • 16:20
        Status of the EU DEMO breeding blanket manufacturing R&D activities 20m

        The realization of a DEMOnstration Fusion Power Reactor (DEMO) to follow ITER, with the capability of generating several hundred MW of net electricity and operating with a closed fuel-cycle by 2050, is viewed by Europe as the remaining crucial step towards the exploitation of fusion power. The EUROfusion Consortium, in the frame of the European Horizon 2020 Program, is assessing four different breeding blanket concepts in view of selecting the reference one for DEMO. This paper describes technologies and manufacturing scenarios developed and envisaged for the four blanket concepts, including nuclear “conventional” assembly processes as TIG, electron beam and laser welding, Hot Isostatic Pressing (HIP), and also more advanced (from the nuclear standpoint) technologies as additive manufacturing techniques.
        With regard to welding processes, topics as the metallurgical weldability of EUROFER steel and the associated risks or the development of appropriate filler wire are discussed.
        The development of protective W-coating layers on First Wall, with Functionally Graded (FG) interlayer as compliance layer between W and EUROFER substrate, realized by Vacuum Plasma Spraying method, is also propounded. First layer systems showed promising layer adhesion, thermal fatigue and thermal shock properties. He-cooled mock-ups, representative of the First Wall with FG W/EUROFER coating will be fabricated for test campaigns in the HELOKA facility under relevant heat fluxes.
        First elements of Double Walled Tubes (DWT) manufacturing and tube/plate assembly for the water cooled concept are given, comprising test campaign aiming at assessing their behaviour under corrosion.
        Developments described in the paper are performed in conformity with international standards and/or design/manufacturing codes.
        Eventually, further development strategies are suggested.

        Speaker: Dr Laurent Forest (DEN-Service d'études mécaniques et thermiques (SEMT), CEA, Université Paris-Saclay)
      • 16:40
        Experimental refutation of the deuterium permeability in vanadium, niobium and tantalum 20m

        Unique gas retention and transport characteristics of group V elements (V, Nb, Ta) have long attracted a significant interest, in particular among the nuclear fusion community. The nominally high hydrogen isotope permeability and diffusion at the expected operational temperatures, together with the negative activation energy for the solubility present these materials as a promising choice for the fabrication of tritium recycling structures.
        However, before seriously considering these materials, one should question the accuracy of the available data, given the remarkable lack of direct experimental measurements in support of the traditionally accepted transport properties of these materials. Furthermore, it must be considered that data have been mostly obtained by combining results obtained by different authors and methods.
        The extensive literature review presented in this paper shows that existing experimental results not only contradict the semi-empirical values assumed for these materials but also present a broad dispersion.
        In order to clarify this, deuterium permeability data for the three materials was obtained at the THERMOPERM facility at Ciemat (Madrid, Spain) in a relevant range of pressures and temperatures. Experimental difficulties together with the role of surface oxidation which may become a major issue for practical uses are also assessed.

        Speaker: Dr Marta Malo (National Fusion Laboratory, CIEMAT)
      • 17:00
        Multifunctional nanoceramic coatings for future generation nuclear systems 20m

        Several breeding blanket concepts for the DEMO reactor employ the eutectic Pb–16Li as breeder material, namely Helium Cooled Lithium Lead (HCLL), Water Cooled Lithium Lead (WCLL) and Dual Coolant Lithium Lead (DCLL). These three concepts share, with different incidences, three major technological challenges: tritium containment, steel corrosion and magnetohydrodynamic drag. Here, we describe the ongoing work on multifunctional Al2O3 nanoceramic coatings grown by Pulsed Laser Deposition (PLD) and Atomic Layer Deposition (ALD) on T91 steel. In fact, these two techniques are complementary from the manufacturing point of view since the first can produce relatively thick (up to 10s of μms) high performance coatings, while the latter is capable of coating complex 3D objects with thin films (in the order of 100s of nms). Both coatings were tested as tritium permeation barriers with hydrogen at different temperature (from 350 to 650 °C). Results collected in this way indicate an excellent behavior, with a permeation reduction factor (PRF) up to 10^5 for both PLD and ALD coatings. In the case of PLD grown Al2O3 coatings, these results have been shown to be maintained also in the case of deuterium under 2MeV electron irradiation. Moreover, the electrical conductivity of these dielectric coatings is shown to be extremely low even when subjected to irradiation. ALD coatings are being currently tested in these conditions. Finally, to evaluate the chemical compatibility of Al2O3 films in liquid eutectic Pb-16Li, PLD and ALD samples have been exposed to static corrosion tests up to 8000 hours. No corrosive attacks on the steel substrate are detected. In conclusion, alumina coatings deposited by PLD and ALD show great promise to tackle the major technological challenges associated to the BB concepts employing Pb-16Li as breeder materials.

        Speaker: Dr Fabio Di Fonzo (Center for Nano Science and Technology, Istituto Italiano di Tecnologia)
    • 16:00 17:20
      O1.C ETNA Hall - ATA Hotel Naxos Beach Resort

      ETNA Hall - ATA Hotel Naxos Beach Resort

      Giardini Naxos

      Via Recanati, 26 Giardini Naxos, Messina - Sicily (Italy)
      • 16:00
        Methodology for reverse engineering analysis of ITER as-built integrated systems 20m

        The ITER machine consists of a large number of highly integrated and complex systems, with critical functional positional requirements (e.g. accurate positioning of magnets to minimize error fields and location of plasma facing components with respect to magnetic axis) and reduced design clearances to maximize Tokamak performances and limit costs. Deviations from specified part tolerances and product assembly processes/accuracies could have a critical impact on assembly feasibility and achievement of final performances, compromising plasma operation. The assessment of potential deviations on as-manufactured and as-assembled systems, together with early mitigation of non-compliances with Tokamak dimensional requirements, are critical activities during the ongoing manufacturing and Tokamak construction phases.
        This paper describes the detailed methodology implemented within ITER for the assessment of as-built integrated systems, based on Reverse Engineering (RE) techniques. The procedure includes the identification of construction needs (at system and assembly levels), the acquisition of metrology data based on defined needs, the post-processing and alignment of data and the reconstruction and management of 3D as-built configuration models. A detailed description of the assessment of non-conformities and tolerance risks/opportunities based on this methodology is included. Predictive clash detection and assembly process assessment/optimization are also described. RE techniques are a basic tool for the management of dimensional risk and opportunity for mitigation or recovery during ITER Construction phase.
        ITER is a Nuclear Facility INB-174. This paper describes a Protection Important Activity (PIA) for safety

        Speaker: Dr Francisco Javier Fuentes (Central Integration Office, ITER Organization)
      • 16:20
        Digital valve system for ITER remote handling – design and prototype testing 20m

        ITER-RH system is used to exchange the divertor’s 54 cassette assemblies in the vessel. Water hydraulics and servo valves are currently used in the task requiring high accuracy tracking and the use of de-mineralized water. The main concern has been robustness of the technology. Only few suitable commercial water servo valves exist and problems e.g. with jamming and wear been encountered. A possible mitigation is to use redundant valves but ensuring required level of water cleanliness is still an issue.

        An alternative option is to use digital technology where on/off valves and intelligent control is used to produce proportional output. So far, no proper high-pressure on/off valves existed in the market and therefore a new concept was developed and tested with promising results. The valve has very fast response time, flow capacity that suits well for the required velocities and is compatible with de-mineralized water. In addition, the valve is not sensitive against water cleanliness and can be made rad-hard when necessary.

        A mock-up of the remote handling system was used as test bench. The system was simulated in order to dimension the valve system and to tune the valve controller. A complete valve package with 16 prototype on/off valves and a manifold was manufactured and assembled. The rest of the components were off-the-shelf and the result is fully compatible to be used as direct replacement for the servo valve system.

        Long-term tests of 60 working days and 2000+ hours was made. A new level of performance was demonstrated throughout the tests as tracking accuracies were approximately ten times better than with servo valve. Although some design faults where detected and system had faulty components, tracking accuracy was very good. Tracking and positioning capability of digital valve system exceeds all requirements and seems applicable for ITER-RH application.

        Speaker: Lauri Siivonen (Tamlink Oy)
      • 16:40
        Picosecond laser-induced ablation for depth-resolved analysis of first wall components in fusion devices 20m

        Monitoring the fuel content of plasma-facing components is a key challenge for fusion devices like Wendelstein 7-X (W7-X) [1], equipped with graphite PFCs or ITER with beryllium/tungsten components. In the case of ITER, it is essential to limit the tritium content in the first wall to comply with safety regulations and to sustain the tritium cycle. In W7-X the measurement and control of the hydrogen content in the first wall mandatory to achieve stable long pulse operation. Laser-Induced Breakdown Spectroscopy (LIBS) and Laser-Induced-Desorption (LID) [2] are suitable to determine the fuel retention ex-situ in extracted PFCs or in-situ in the vacuum vessel during or between plasma discharges. However, the self-consistent quantification is challenging as the emission is a non-equilibrium process and often a comparative measurement with pre-characterized reference samples is mandatory.

        A combination of LIBS and residual gas analysis is an alternative approach: For volatile sample components, the linearity of the quadrupole signal to gas pressure simplifies the calibration and reduces uncertainties. We use a Nd:YVO4-laser (λ3rd=355 nm) with a pulse duration of τ=35ps and pulse energies up to E=40mJ. With a spot size diameter on the sample of x=0.7mm we achieve a depth resolution of Δh=100nm for graphite samples without significant matrix mixing effects.

        We present a series of composition analysis of poloidal and toroidal locations on graphite limiter tiles of W7-X, exposed to hydrogen plasma in OP1.1. Erosion- and deposition-dominated zones could be identified. In addition, graphite divertor tiles exposed in OP1.2a are analyzed. In contrast to ms-LID, the heat penetration depth of ps-laser-induced-ablation is smaller than the ablation rate. Consequently, all retained hydrogen from the sample is removed and quantitative depth-resolved hydrogen retention information (O(1022/m2)) is gained.

        [1] T.S. Pedersen et al., Nucl. Fusion 55 (2015)126001.
        [2] G. De Temmerman et al., Nuclear Materials and Energy (2016)2352-1791.

        Speaker: Dr Jannis Oelmann (Institut für Energie- und Klimaforschung – Plasmaphysik, Forschungszentrum Jülich GmbH)
      • 17:00
        High temperature microstructural stability of self-passivating W 10Cr-0.5Y alloy for blanket first wall application 20m

        Tungsten is the main candidate material for the first wall (FW) armour of future fusion reactors. However, a loss of coolant accident with simultaneous air ingress into the vacuum vessel would lead to temperatures of the in-vessel components exceeding 1000ºC, resulting in the formation of volatile and radioactive tungsten oxides. A way to prevent this important safety concern is the addition to tungsten of oxide-forming elements, which, in presence of oxygen at high temperatures, promote the formation of a self-passivating layer protecting tungsten from further oxidation.
        A W-10Cr-0.5Y alloy produced by mechanical alloying and hot isostatic pressing (HIP) has been recently developed, exhibiting a strong reduction of oxidation rate compared to pure W and high mechanical strength. After HIP at 1250°C it shows a two phase microstructure, according to the W-Cr phase diagram, with a nanocrystalline structure of the W-rich phase due to the presence of a Y2O3 nanoparticle dispersion inhibiting grain growth. A heat treatment after HIP at 1550°C results in a one-phase material with average grain size of 250 nm and coarsening of the Y2O3 particles to 50 nm. This material exhibits a high thermal shock resistance, as demonstrated by tests at the JUDITH facility consisting of 1000 ELM-like pulses, where the material showed a comparable performance to pure W. Nevertheless, the microstructure is metastable and its thermal stability under operational conditions has to be assessed.
        In this work, results of thermal stability tests on heat treated W-10Cr-0.5Y alloy subjected to temperatures of 650, 700 and 1000°C for times ranging from 50 to 3000 h are presented. After 100 h at 700°C a slight growth of the Cr-rich phase is detected. After 100 h at 1000°C a complete decomposition takes place with the formation of a uniform, fine-scale mixture of W- and Cr-rich phases, typical for spinodal decomposition.

        Speaker: Dr Carmen García-Rosales (CEIT-IK4)
    • 18:30 20:00
      Welcome Reception
    • 08:30 09:10
      I3.1: X. Litaudon Plenary Hall - ATA Hotel Naxos Beach Resort

      Plenary Hall - ATA Hotel Naxos Beach Resort

      Giardini Naxos

      Via Recanati, 26 Giadini Naxos, Messina - Sicily (Italy)

      First Invited 2nd Conference Day

      • 08:30
        European Integrated Programme in support to ITER: Overview of JET and Medium Size Tokamak results 40m

        Europe has elaborated a Roadmap to the realisation of fusion energy in which ‘ITER is the key facility and its success is the most important overarching objective of the programme’. EUROfusion has seized the unique opportunity to develop an integrated programme on devices of different sizes, i.e. on EU Medium-Size Tokamaks (MSTs), and, on JET in order to provide a step-ladder approach for extrapolation to ITER. In addition, the ITER Organization has issued a detailed analysis of the risks to ITER operation and has identified the main R&D needs to mitigate those risks in the revised ITER research plan. In this context, this paper will provide an overview of the recent coordinated contributions of the EU programme to optimise ITER operation.
        Disruptions are considered as the highest operational risk in the ITER Research Plan. The high priority physics studies on JET and MSTs consist of disruption prediction, avoidance, mitigation and associated modelling (including multi-machine run-away electrons model validation). A new shattered pellet injection system is being installed on JET to compare with massive gas injection and elucidate the differences in run-away electrons beam mitigation in view of impacting the design of the ITER disruption mitigation system. The JET and the MSTs programmes have concentrated on the preparation of ITER operating scenarios and on providing a physics basis for optimising fusion performance operation with metallic first wall materials. It is found on JET and ASDEX Upgrade, that plasma performance is significantly affected when plasma boundary conditions are modified which will affect the strategy to achieve the fusion performance in the coming JET deuterium-tritium campaign and ITER QDT=10 main mission. In addition, preparation of the ITER non-active phase has been carried out in hydrogen on JET, and, in hydrogen and helium in the MSTs. The recent progress will be reviewed on plasma surface interaction with ITER first wall materials (e.g. beryllium and tungsten erosion/migration, helium and tungsten interaction), scaling of L to H mode power threshold, Scrape-Off-Layer physics, core and pedestal confinement with different hydrogen isotopes and helium, control of detached divertor scenarios using extrinsic impurity seeding, and options for ELMs control with pellets or with resonant magnetic perturbations, RMPs. ELMs control with RMPs has been established on ASDEX Upgrade in helium using methods developed for deuterium plasmas, addressing a ITER issue of the transferability of ELMs control methods.
        To conclude, the success of ITER operation will also require integrating the experimental progress made in different fusion facilities through theory-based first principle and integrated modelling. The European Transport Simulator, ETS, for integrated modelling has undergone major development and has been benchmarked against TRANSP. The strategic movement towards the adoption of the ITER integrated modelling and analysis suite (IMAS) has been pursued by the continued support and validation of the IMAS infrastructure and extension of the EUROfusion experimental databases in IMAS.
        This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission.
        1 See author list of “X. Litaudon et al., 2017 Nucl. Fusion 57 102001”
        2 See author list of “H. Meyer et al., 2017 Nucl. Fusion 57 102014”

        Speaker: Dr Xavier Litaudon (EUROfusion Consortium JET)
    • 09:10 09:50
      I3.2: Y. Kamada Plenary Hall - ATA Hotel Naxos Beach Resort

      Plenary Hall - ATA Hotel Naxos Beach Resort

      Giardini Naxos

      Via Recanati, 26 Giardini Naxos, Messina - Sicily (Italy)

      2nd Invited 2nd Conference Day

      • 09:10
        JT-60SA Program Contribution to Fusion Energy 40m

        JT-60SA is a highly-shaped large superconducting Tokamak under construction by EU and Japan. The mission of JT-60SA is to support ITER and to complement ITER towards DEMO by resolving key physics and engineering issues. Fabrication and installation of components of JT-60SA by EU-Japan Integrated Project Team are progressing on schedule towards the first plasma in Sep. 2020. On the Cryostat Base made by CIEMAT, the 340-gegree-part of the Vacuum Vessel (VV) has been placed and welded accurately by QST. By Feb. 2018, all 18 TF coils have been manufactured and cold-tested by ENEA and CEA and 14 TF coils have been assembled to the tokamak by QST. Manufacture of all 6 EF coils have been completed by QST. Commissioning of the cryogenic system was completed by CEA in Naka. High Temperature Superconducting current leads have been delivered by KIT. Commissioning of the power supply system (ENEA, RFX, CEA and QST) has also been implemented smoothly. The Cryostat Vessel Body has been delivered by CIEMAT.
        The JT-60SA Research Plan (SARP) ver. 3.3 was issued in March 2016 by 378 co-authors (JA 165 (16 institutes), EU 213 (14 countries, 30 institutes): Using ITER- and DEMO-relevant plasma regimes and its sufficiently long discharge duration, JT-60SA enables studies on all the key physics issues for ITER and DEMO. From ~2030, the first wall will be changed from carbon to full tungsten-coated carbon. By integrating these studies, the project provides ‘simultaneous & steady-state sustainment of the key plasma performances required for DEMO’. Such JT-60SA research activity includes consolidation of a “Plant Simulator”. As for the first plasma and heating experiments, JT-60SA will start earlier than ITER by five years. Therefore, experiences and achievements in JT-60SA are expected to contribute to reliable operation of ITER.

        Speaker: Dr Yutaka Kamada (National Institutes for Quantum and Radiological Science and Technology)
    • 09:50 10:30
      I3.3: J. Fellinger Plenary Hall - ATA Hotel Naxos Beach Resort

      Plenary Hall - ATA Hotel Naxos Beach Resort

      Giardini Naxos

      Via Recanati, 26 Giadini Naxos, Messina - Sicily (Italy)

      3rd Invited 2nd Conference Day

      • 09:50
        W7-X: Technology progress of the experimental campaign with divertor plasmas 40m

        Wendelstein 7-X (W7-X), a fivefold symmetric stellarator located at the Max-Planck-Institute for Plasma Physics in Greifswald, Germany, was successfully taken in operation with short pulse limiter plasmas in 2015. Hereafter, ten symmetrically positioned un-cooled graphite divertors were installed, the plasma facing wall was refurbished with graphite tiles and various auxiliary systems and diagnostics were upgraded. The reinforcements allow for an increase of the energy input from 4 to 80 MJ.
        The experimental campaign with island divertor plasmas was launched in August 2017. The main goal of this campaign is to demonstrate the capability of the un-cooled test divertor in high density and high power plasmas. In the island divertor concept, the divertor is positioned in front of the pumps (for the neutrals exhaust) and it only intersects the relatively cold resonant islands around the core plasma. In this way, convective heat loads are limited to 10 MW/m² and neutral pumping is effective, despite the fact that the divertor area represents only 10 % of the plasma facing surface.
        Second goal of the experimental campaign is to prepare for the installation of the final water cooled divertor. The water cooled divertor is planned for 2020 and has to be installed with great care and high accuracy. It relies on the experience gained with the installation of the un-cooled divertor.
        To improve the density in comparison to the first campaign with limiter plasmas, an injector of frozen hydrogen pellets was installed and a fast hydrogen gas inlet with piezo-valves mounted in cut-outs of the divertor were taken in operation. In addition, various diagnostics were added or enhanced before start of the divertor campaign.
        The tests divertor campaign is split into two parts: During a short break a the middle of the experimental campaign, two so-called scraper elements are installed in front of their corresponding divertor to shield the sensitive edges of the divertor near the pumping gap. Aim of the scraper program is to compare the edge loads on the divertor with and without a scraper and to evaluate the impact of the scraper on the neutrals pumping efficiency. The scraper elements are monitored by additional diagnostics. The break was also used to take the first of two neutral beam injectors (NBI) into operation and to add or harden several diagnostics.

        Speaker: Dr Joris Fellinger (Max Planck Institute for Plasma Physics)
    • 10:30 11:00
      Coffee Break 30m Coffee Break Area - ATA Hotel Naxos Beach Resort

      Coffee Break Area - ATA Hotel Naxos Beach Resort

      Giardini Naxos

      Via Recanati, 26 Giadini Naxos, Messina - Sicily (Italy)
    • 11:00 13:00
      P2: Poster Session Pantelleria Hall - Terrace - ATA Hotel Naxos Beach Resort

      Pantelleria Hall - Terrace - ATA Hotel Naxos Beach Resort

      Giardini Naxos

      Via Recanati, 26 Giadini Naxos, Messina - Sicily (Italy)
      • 11:00
        P.228 Chemistry and Corrosion Research and Development for Water Cooling Circuits of EU DEMO 2h

        The European DEMO design will potentially use single phase water cooling in various components that require protection against corrosion damage. Coolant conditions will be similar to fission PWRs but with additional considerations arising from materials choices (Eurofer-97, CuCrZr), 14 MeV neutron irradiation, the presence of tritium, and strong magnetic fields. Presently, many aspects of the water chemistry and corrosion behaviour are not well defined, and several strands of work, reported here, are ongoing to address these challenges under the EUROfusion framework in collaboration with industrial partners.
        The foundation of this work is a review of relevant operating experience from fission LWR plant to understand the potential for technology transfer, supported by radiolysis modelling to assess options for suppression of oxidising species under high energy neutron irradiation. This has specifically considered the interaction of water with Eurofer-97, utilised in the water-cooled lithium-lead blanket concept. High temperature water corrosion testing facilities have also been employed to expand the corrosion database, reported here an approach using micro-scale samples of structural alloys to study their susceptibility to stress corrosion cracking, and a small-scale flow cell approach for in situ corrosion measurement under changing chemistry conditions.
        Consideration has also been given to the effect of intense magnetic fields on corrosion through exposure of Eurofer-97 coupons to a magnetic field intensity of 0.88 T and temperatures of 80°C. Further work is reported aimed at identifying the nature and extent of any impact on corrosion behaviour in higher temperature water.
        In-vessel cooling of the divertor will use CuCrZr under lower-temperature, high flow conditions, which will lead to different considerations and the potential for flow assisted corrosion. Additionally, high, unidirectional, heat fluxes lead to a radial temperature profile and the possibility of sub-nucleate boiling. A separate test set-up, currently under construction, to expand this corrosion database is described.

        Speaker: Chris Harrington (Culham Centre for Fusion Energy)
      • 11:00
        P2.001 Modelling and experimental validation of RFX-mod tokamak shaped discharges 2h

        Shaped Tokamak discharges with an insertable polarized electrode have been executed in RFX-mod to achieve H-mode regime. This was aimed at reproducing successful experiments of stable operation at q<2 by feedback stabilization of m=2, n=1 mode already performed with low and high-beta circular discharges. Equilibrium magnetic configurations with a wide range of plasma shapes have been experimentally produced and analysed by means of the linearized plasma response model CREATE-L. In order to provide a connection between computational tools and experiments, the purpose of this work was to develop a general procedure for computing linearized plasma response models through the CREATE-L code, in any kind of plasma regime with a high level of accuracy with respect to experimental data. This procedure involves the solution of a constrained non-linear minimization problem to estimate the CREATE-L free parameters by using an iterative scheme trying to minimize the discrepancy between the magnetic field experimental and simulated measurements provided by pick-up coils. Eleven experimental shots have been identified and considered in this study: all of them are Upper Single Null tokamak configurations spanning the whole range of poloidal beta achieved in the RFX-mod tokamak (low-β, intermediate-β, biased induced H-mode regime). A preliminary sensitivity analysis showed a non-negligible dependence of static equilibria on variations of the total plasma current with respect to the measurement provided by Rogowski coils. Thus, the total plasma current has been set as an additional degree of freedom in the minimization problem allowing it to assume values between the Rogowski measurement and the value of the discrete line integral of the poloidal magnetic field measured by the pick-up coils. In all cases under analysis, the iterative procedure showed that the most accurate equilibrium is obtained with a plasma current higher than the Rogowski measurement but lower than the pick-up coils line integral.

        Speaker: Domenico Abate (Consorzio RFX)
      • 11:00
        P2.002 W7-X NBI beam dump thermocouple measurements as safety interlock 2h

        In the upcoming operational phase OP1.2b of the Wendelstein 7-X stellarator in 2018 it is planned to have the Neutral Beam Injection (NBI) Heating System operational. Any un-absorbed heating power is dumped on the NBI beam dump graphite tiles that are cooled using CuCrZr-cooling structures. The Heat Shield Thermography (HST) system is present to prevent damage and overheating of the graphite tiles on these beam dumps. In addition, other interlocks (plasma density and ECRH stray radiation interlocks among others) are present to prevent damage to in-vessel components in case of un-absorbed heating power. Since the HST as well as the other interlocks have a rather low safety integrity level (SIL) it was decided to use the available thermocouple measurements of the beam dumps as an additional interlock. Due to the relatively slow response of this type of measurement, the focus of this safety interlock lies on preventing major damage to the PV wall in case of a chain of malfunctions of HST, other heating interlocks and heating systems control. It is not implemented to fully prevent damage to the beam dump structure itself.

        This paper describes the setup of the beam dump measurements, the upgrade of the electronics cabinets to SIL2-rated thermocouple measurement with alarm trip relais, the sequence for stopping the NBI beam and the analyses performed to determine the interlock alarm trip settings for operation in OP1.2b.

        [1] P. McNeely et al., “Current status of the neutral beam heating system of W7-X”, Fusion Engineering and Design, Volume 88, October 2013
        [2] N. Rust et al., “W7-X neutral-beam-injection: Selection of the NBI source positions for experiment start-up”, Fusion Engineering and Design, Volume 86, October 2011
        [3] B. Mendelevitch et al., “Water-cooling system of the W7-X plasma facing components”, Fusion Engineering and Design, Volume 98-99, October 2015

        Speaker: Paul van Eeten (Operations Department Max-Planck-Institute for Plasma Physics)
      • 11:00
        P2.003 Final design of the JT-60SA pellet launching system for simultaneous density and ELM control 2h

        The key mission of the new tokamak JT-60SA is to conduct exploitations in view of ITER and to resolve key physics and engineering issues for DEMO. Its pellet launching system was designed to cover according requirements by providing a powerful and flexible tool for the control of density profile and ELM frequency. Therefore, the systems lay out had to be adapted for pellet injection via a guiding tube with an already pre-set geometry from the vessel inboard side. Modelling showed inboard launch is a must in order to achieve high fuelling efficiency; by analysing the potential pellet fuelling impact in all relevant plasma scenarios the optimized set of pellet parameters for fulfilling all the tasks requirements was elaborated. The feasibility of a mechanical centrifuge as pellet acceleration unit was studied. This approach would guarantee the precise pellet launch speed as needed to enable the best adaptation to the guiding tube transfer capabilities and accurate control of pellet frequency and particle flux as arriving in the plasma. While the appropriateness of the centrifuge principle has been already proven by several devices, the proposed version possesses a novel design employing several accelerator arms. Such, operation at heighten rates and with a more refined adjustment of pellet flux and frequency could be achieved. Moreover, it allows hosting several sources delivering different pellets and their simultaneous actuation. The appropriate control unit is designed to merge pellets from different sources and form a sequence for the simultaneous control of different basic plasma parameters as e.g. density and ELM frequency. This flexibility can also allow including additional applications like isotope fraction control or radiative power exhaust by the use of compound pellets. At present, the detailed engineering of all major components is in progress, aiming to provide specifications and get ready to prepare the procurement process.

        Speaker: Peter Lang (Tokamak Scenario Development MPI für Plasmaphysik)
      • 11:00
        P2.004 Three-dimensional disruption vertical stability and breakdown analysis of the Italian DTT device 2h

        The reduction of heat loads of divertor target is one of the main challenges addressed by the European roadmap to the realisation of fusion energy. In particular, eight different missions have been identified overall, of which Mission 2 ‘Heat-exhaust systems’ is specifically devoted to this goal. Recently, the Divertor Tokamak Test (DTT) facility [1] has been proposed with the aim of investigating alternative power exhaust solutions in view of DEMO, hence with the capability of including various divertor concepts (e.g. conventional, snowflake, super-X, double null, liquid limiter).
        In this paper, we numerically analyse the recently revised up-down symmetric Italian DTT device, with the aim of evaluating the effects of deviation from axisymmetry on the plasma behaviour. This will be done resorting to CarMa0NL [2] (evolutionary equilibrium in presence of 3D conductors) and CARIDDI [3] (eddy currents equations in volumetric conductors) codes. Vertical stabilization, plasma breakdown and disruptions are analysed in this respect. Three-dimensional effects may be both detrimental (e.g. ports) and beneficial (e.g. first wall) on passive stability, so the overall effect is not obvious and must be carefully evaluated. On the other hand, having significant currents flowing in the first wall during fast transients poses significant challenges on the design, to take into account the aspects related to the field penetration in the breakdown phase and the electromagnetic loads on plasma facing components during disruptions.
        [1] R. Albanese, et al., Fusion Engineering and Design (2017), https://doi.org/10.1016/j.fusengdes.2016.12.025
        [2] F. Villone et al., Plasma Phys. Control. Fusion 55 (2013) 095008
        [3] R. Albanese, G. Rubinacci, 1988 IEE Proc. A 135 457–462

        Speaker: Fabio Villone (DIETI universitˆ di Napoli Federico II)
      • 11:00
        P2.005 Development of reactor relevant pellet launching system technology on ASDEX Upgrade 2h

        Controlling the plasma density in a future fusion reactor will be mainly attributed to pellet injection using a control algorithm based on a rather difficult density measurement. The underlying technology to capacitate the Pellet Launching System (PLS) for the requirements is challenging. The ASDEX Upgrade (AUG) PLS was retrofitted for this task, intensifying the integration into the Discharge Control System (DCS). Lessons learnt and their consequences for the design of a new system “from scratch” will be described.
        The technology of the ice production process is discussed as well as the observed performance in view of isotope ratio accuracy and plasma fuelling performance, mimicking the reactor fuel (mixture D/T) by using a mixture of H/D. Since a fusion power plant will require a steady-state pellet source, the development of a control strategy in view of process control is mandatory. The procurement of a new pellet source was launched in order to enhance the existing centrifuge acceleration system. This contribution will show first considerations.
        An innovative conceptual design will be presented comprising the potential to replace nowadays centrifuge systems. Present-day vacuum rotating feedthrough technology enables the installation of the motor for the accelerating arm on the atmosphere side avoiding a series of technical difficulties.
        The existing PLS on AUG is in operation now for almost 30 years, no fall-back option is available right now. A conceptual design for a new PLS consisting of extruder, launcher and control systems (also for density control) with focus on reactor relevant technologies except tritium compatibility is under preparation.
        Results presented here are complementary domestic activities to the EUROfusion WP TFV.

        Speaker: Bernhard Ploeckl (Tokamak-Scenario-Development Max-Planck-Institute for Plasma Physics)
      • 11:00
        P2.006 Passive control of runaway electron displacement by magnetic energy transfer in J-TEXT 2h

        During disruptions runaway electrons(REs) often drift from high field side to low field side in J-TEXT. It may damage plasma facing components when REs strike the first wall with high energies. In order to mitigate the damage, a novel approach called magnetic energy transfer(MET) based on the principle of electromagnetic coupling is presented in this paper. A set of extra coils with a high coupling coefficient with plasma is installed on the device, and a toroidal current can be induced in the coils during disruptions which can transfer the plasma poloidal magnetic energy out of vacuum vessel. Flowing in the same direction as the runaway current, the induced current can attract the runaway current to high field side, control the displacement of the RE beams and prolong runaway current plateau. Compared with vertical field control, a significant advantage of MET can be seen that the MET is passive control without power supply, while vertical field control needs a power supply and belong to active control. The J-TEXT experiment results show that the increase rate of RE beams’ horizontal displacement can be obviously slowed. The runaway current plateau can be prolonged 4-5 ms and the control effect becomes better as the induced current in the MET coils increases. Thus the MET can control the displacement of RE beams effectively.

        Speaker: Nianheng Cai (International Joint Research Laboratory of Magnetic Confinement Fusion and Plasma Physics State Key Laboratory of Advanced Electromagnetic Engineering and Technology School of Electrical and Electronic Engineering Huazhong University of Science and Technology)
      • 11:00
        P2.007 Status and tasks for modernization of TRINITI site infrastructure for the Ignitor project 2h

        The project of tokamak Ignitor is one of the main themes of long-term scientific cooperation between the Russian Federation and the Italian Republic. Currently, negotiations on the development of technical design tokamak Ignitor with placement on the site of TRINITI (Moscow, Troitsk, Russia). The discussion on preparing of the Russian-Italian Inter-government agreement on realization of Ignirtor Project is in ongoing too.
        Project Ignitor differs significantly from that currently under consideration of the projects of fusion reactors based on the tokamak. Tokamak Ignitor has a super strong magnetic field (13 T), in which the pulse discharge (about 10 sec) flows a powerful discharge current (11MA). The Ohmic heating is the main mechanism of ignition of the thermonuclear fusion reaction
        The main purpose of this stage of research was to determine the current state of the infrastructure of power and engineering-physical complexes of TRINITI and the development of technical proposals for the modernization of these complexes for the task of the Ignitor Project implementing.
        During the research the current state of equipment and communications of practically all the main elements of power and engineering infrastructure of the tokamak with a strong field (TSP) pilot-bench complex TRINITI was inspected, including stationary and pulsed power supply systems, vacuum, cryogenic, fuel systems and diagnostic complex. Based on the comparison of the dates on technical characteristics of the tokamak Ignitor and requirements to the livelihoods systems of the tokamak from the published sources were prepared technical proposals for the modernization of the experimental bench complex of TSF TRINITI under project objectives Ignitor. The work was carried out as a one of the important steps to prepare the Russian side to participate in the joint Russian-Italian development of the tokamak Ignitor technical project.

        Speaker: Dr Mikhail Subbotin (Kurchatov Complex of Fusion Power and Plasma Technologies National Research Centre Kurchatov Institute)
      • 11:00
        P2.008 West Operation management Software Suite (WOSS): software assisting organization and follow-up of West operation 2h

        In order to operate a large research facility, one needs software tools assisting the organization and the operation follow-up. In the past, separated software tools were used on Tore Supra but for West, an integrated approach was chosen. The West Operation management Software Suite (WOSS) allows a streamline management of information from the planned program up to the realized experiments and the follow-up of the West technical status. The WOSS is split up in six modules: Timeline, Roster, Logbook, Physics summary, Systems status, Incidents.
        The Roster module is implemented to schedule experimental campaign with daily experiment information, including links to the experiment description and the list of people who takes part in each experiment, while the Timeline module shows a multi-month calendar of experiments.
        Follow-up of daily experiment is ensured by the Logbook module. The Logbook collects data and comments from responsible officers carrying out experiments on pulse by pulse basis. The Physics summary module, extracting data from logbooks, provides an overview of the current experimental session and is displayed in the control room.
        The Incident and the Systems status modules support the monitoring of West sub-systems. All the operation incidents that happended on any West systems are dully collected in the Incident module as well as current and foreseen status of subsystems are published on the System Status module. Thanks to the integrated database concept, incidents and status data are shown in real time on the Logbooks.
        The WOSS deal with a large quantity of information and is of precious help for the coordination support unit. In the article will be reported, how the WOSS is facilitating the coordination by grouping all information of each sub-system needed for the operation. Moreover, the article will present a report of the last campaign with an analysis of West sub-systems availability.

        Speaker: Elodie CORBEL (IRFM CEA Cadarache)
      • 11:00
        P2.009 Design and implementation of quasi-optical components for the upgrade of the TCV EC-system 2h

        The EC-system of the TCV tokamak is progressively being upgraded with the addition of two MW-class dual-frequency gyrotrons (84 and 126GHz/2s/1MW) being manufactured by Thales Electron Devices with the first gyrotron delivered to SPC at the end of 2017. In order to connect the two gyrotrons to the existing low field side and top launchers, new waveguide routing from gyrotron hall to TCV tokamak was designed and dedicated Matching Optics Units (MOU) have been developed. The internal optics of the system have been determined aiming at optimal coupling to the HE11 waveguides. The laws of quasioptics were used to find quadratic surfaces to shape an incoming Gaussian beam representative of the gyrotron output into a beam matching the proper field distribution at the waveguide entrance and with HE11 content compatible with the system requirements. A solution with one flat (movable) mirror and two shaping mirrors was found and characterized with the physical optics code GRASP. The resulting field distribution is then truncated and projected onto the HE11 component to evaluate the design solution (coupled power at the waveguide entrance >98.4% and HE11 content >96.8% for both frequencies). The model and the results of this analysis will be presented and compared to a model based on the Rayleigh-Sommerfeld scalar diffraction integral. GRASP was also used to evaluate preliminary misalignment effects in terms of coupled power to the waveguide and the first MOU design moved to the manufacturing phase. In parallel to the gyrotron integration and to extend the level of flexibility of the TCV EC-system, a modular closed divertor chamber is developed, requiring the X3 top-launcher to be redesigned. Preliminary antenna conceptual design studies including new curvature to cope with the requirements of modularity and flexibility will be presented.

        This work is supported by the Ecole Polytechnique Fédérale de Lausanne (EPFL).

        Speaker: Alessandro Moro (Istituto di Fisica del Plasma IFP-CNR)
      • 11:00
        P2.010 REMAINING USEFUL LIFE ESTIMATION OF CRITICAL DIII-D SUBSYSTEMS 2h

        DIII-D plays a vital role in the development of the physics basis for fusion energy and the ITER design. Designed in the 1970’s and built in the early 1980’s, the system started operations in 1986 and has provided a reliable platform for fusion experiments for over 30 years. A hallmark of DIII-D operations has been its ability to adapt to the changing needs of the fusion research community and support a wide range of tokamak experiments for various stakeholder’s.
        An important factor in ensuring continued reliable operations is gaining a thorough understanding of the remaining useful life (RUL) of the DIII-D facility. Developing quantitative assessments of RUL requires a complex set of analyses to evaluate failure modes and effects, life models, operating history and present condition. General Atomics has conducted a pilot RUL assessment project to develop and refine a set of RUL estimation tools applicable to various DIII-D systems.
        The pilot assessment categorized DIII-D systems as either repairable/replaceable or facility-life-critical and developed a multi-factor, quantitative assessment of RUL for a facility-life- critical system, the F8-coils. In contrast to repairable/replaceable systems, facility-life-critical systems perform essential DIII-D functions and are expected to last the life of the facility. The F8-coil RUL assessment included life estimates based on conductor and insulation fatigue, insulation ageing, conductor corrosion and material degradation due to radiation exposure.
        The pilot F8-coil RUL assessment project showed that under typical operating conditions seen to date, the F8-coil may be expected to achieve a life of ~50 years or >106 shots. The pilot project further demonstrated that employing reliability engineering techniques coupled with modern analytical tools, quantitative estimates for RUL can be developed with reasonable investment. A description of the methodology employed and results will be presented.

        This work is supported by the US DOE under DE-FC02-04ER54698.

        Speaker: Joel Drake (General Atomics)
      • 11:00
        P2.011 A conceptual system design study for an NBI beamline for the European DEMO 2h

        Neutral Beam Injection (NBI) is a robust, established heating and current drive method in fusion experiments. Among its strengths is high current drive efficiency that may pave the path for steady state operation of a tokamak reactor with an economically viable recirculating power fraction. For large tokamaks like ITER and DEMO the use of negative ions is mandatory due to the vanishing neutralisation efficiency of positive ions at the required approx. 1 MeV beam energy. While the design of ITER’s NBI has long been finalized and the prototype is under construction, NBI for DEMO poses new challenges that go far beyond those of ITER and require new approaches. Besides the demand for very-long-pulse or continuous operation, compatibility with much higher neutron fluences, and higher availability, a significant increase in energy efficiency is required to render NBI a viable option. Currently, the wall plug efficiency is limited to about 27 % on ITER, mostly due to the low neutralisation efficiency in beam–gas collisions. New concepts such as photo-neutralisation promise to overcome this limitation, but their practical feasibility has not yet been demonstrated.

        Within the work package Heating and Current Drive of EUROfusion’s Power Plant Physics and Technology division a systematic and comprehensive system design study of a DEMO NBI beamline was launched this year in order to explore a broad range of options for each beamline component and mutual dependences. The components’ design is mostly constrained by the chosen principle of the neutraliser, e.g. whether there is a need for residual ion energy recovery (ER) that may save the energy efficiency of the beamline even if the particle neutralisation efficiency is far from 100 %. We have chosen this as starting point and present first concepts for ER integration as well as our general approach to a comprehensive system design study.

        Speaker: Christian Hopf (ITER Technology Diagnostics Max Planck Institute for Plasma Physics)
      • 11:00
        P2.012 Long-Pulse High-Power 170 GHz Absorbing Matched Load Tests and Developments 2h

        The future EC systems will consist of several gyrotrons sources providing MW-level millimeter wave power at a frequency around or above 170 GHz. The development of matched loads is necessary to test the new sources, the components for the transmission lines and the launchers, and must ensure high qualification for compatibility with the nuclear environment. The load low reflectivity and high power-handling capability are mandatory for testing as well as for an accurate power measurement capability. The development at IFP and LTC of several compact high-power prototypes during the last decade led to a refinement of the overall design, resulting in the present low-reflectivity vacuum-compatible loads. The activity on loads is supported by F4E in view of the development of the EU gyrotron for ITER and required high power tests at QST (Naka, Japan). Qualification is now supported by new tests at SPC (Lausanne, Switzerland) using the FALCON test-bed designed to test components and the EU gyrotron prototype for the EC system of ITER. These high power tests, performed on the first CW prototype (provided with 16+16 cooling channels in parallel) highlighted the need to improve the mechanical, vacuum and hydraulic design to reach the final goal of 1000s at ~1MW. Two long-pulse loads with new technological solutions have been built for the Japanese ITER gyrotron at QST, while a modified version of the load, designed to test an equivalent input power of 2MW at QST, has been provided for FALCON.

        Acknowledgement

        This work is supported by Fusion for Energy under Grant F4E-GRT-553 to the European Gyrotron Consortium (EGYC). EGYC is a collaboration among SPC, Switzerland; KIT, Germany; HELLAS, Greece; IFP-CNR, Italy. The views expressed in this publication are the sole responsibility of the authors and do not necessarily reflect the views of F4E and the European Commission.

        Speaker: William Bin (Istituto di Fisica del Plasma Consiglio Nazionale delle Ricerche)
      • 11:00
        P2.013 Ion Cyclotron Frequency Range Cold Magnetized Plasma Modelling in ANSYS HFSS 2h

        Achieving the plasma temperature expected for nuclear fusion requires external heating systems, such as dedicated Radio-Frequency antennas. Dimensions, power level and manufacturing cost which are at stake make it impossible to build scale-one mock-up during design and prototyping phases. For that reason, modelling the electromagnetic interactions between magnetized plasmas and Radio-Frequency antennas is mandatory for nuclear fusion research.

        The modelling of the interactions between cold magnetized plasmas and Ion Cyclotron Resonance Heating (ICRH) antennas is generally assessed using specific codes. Antenna coupling codes often approximate the plasma to surface impedances described by 1D half infinite models. Depending of the mathematical approaches used by these codes, the modelled antennas can be described either in simplified 2D dimensions or in 3D complex geometries converted from CAD models. Such approaches do not allow one to work directly on realistic antenna model and to assess rapidly performance changes. Moreover, during the design phase, it is convenient to use the simulation to directly estimate the thermal and mechanical loads in the same software suite.

        Thanks to the collaboration between the fusion community and ANSYS, the RF modelling software ANSYS HFSS supports non-homogeneous gyrotropic medium, which are used to describe magnetized cold plasma away from resonances. In this paper, ANSYS HFSS is used to model ICRH antennas coupling to cold magnetized plasma. A simplified antenna model is compared with the specific coupling code ANTITER for various plasma density profiles. It is found that the coupling performances can generally be reproduced in HFSS. Its finite-element implementation imposes inherent restrictions on the definition and boundaries of the plasma domain due to the gyrotropic media in which two propagation modes can co-exist. These restrictions are discussed and technical recipes are given to conduct satisfying antenna coupling calculations with ANSYS HFSS, which allow faster design phases and experimental comparisons.

        Speaker: Julien Hillairet (IRFM CEA)
      • 11:00
        P2.014 Design and mock-up tests of the RING photoneutralizer concept for an efficient DEMO NBI 2h

        High energy (800 keV) Neutral Beam Injection (NBI) is one of the methods being considered to heat EU DEMO plasma [1]. A major issue of present NBI systems is the limited efficiency of the gas neutralizer (for ITER NBI ~55%), which impacts on the overall system efficiency. An attractive method, but still undemonstrated at full performances, is the photo-neutralization of the negative D-ion beam. In this process the energetic ions pass through an optical cavity where they impact on laser photons with a frequency chosen to maximize the neutralization cross section. The expected neutralization efficiency can be up to 70-90%.
        A possible scheme for photoneutralization is named RING (Recirculation Injection by Nonlinear Gating) [2] where the second harmonic of a Nd:YAG laser is extracted and trapped within a non-resonant optical cavity. A mock-up of the optical cavity has been built in Consorzio RFX to study its performances in order to demonstrate the feasibility of the RING concept and its potentiality for a full-scale NBI photoneutralizer. The mock-up has been operated with a low repetition rate (10 Hz) Nd:YAG laser (λ=1064 nm). The optical alignment of the cavity appears not to be critical and first operations are aimed to achieve the 2nd harmonic generation (SHG) saturation regime. The measurements of the SHG efficiency using a Lithium Triborate (LBO) crystal confirm the non-linearity of the process with the crystal thickness. The optical properties of the recirculating light are described by measuring the beam envelope profile with a laser-triggered CCD camera and the trapped beam pulse decay time with a fast photodiode. Operations with the mock-up are fundamental to assess the optical cavity performance and estimate the loss channels affecting the beam recirculation.

        [1] P. Sonato, Nucl. Fusion 57 (2017) 056026
        [2] A. Fassina, Rev. of Sci. Instrum. 87 (2016) 02B318

        Speaker: Pietro Vincenzi (Consorzio RFX)
      • 11:00
        P2.015 A low power testbed for the WEST ICRF launchers and for the acceleration of their commissioning on plasma 2h

        This paper presents a milliwatt-range testbed that has been recently designed and manufactured for the RF characterization of the WEST ICRF launchers. The low power testbed is integrated into the TITAN test facility. This extends the capabilities of TITAN from testing at high voltage/high current parts of the launchers in vacuum, to the characterization of the launchers coupling capabilities, their impedance matching and their load-resilience at low power. These pre-qualification tests allow checking the launchers performances before their installation in the tokamak which in turn allows accelerating their commissioning on plasma.

        The low power testbed is an RF load based on a mechanically rigid glass aquarium which can host various mixtures such as salty-water or Barium Titanate (BaTiO3) solutions. Indeed, these high relative permittivity isotropic and homogenous dielectrics can mimic qualitatively well the magneto-plasma at ICRF frequencies for the fast wave. Hence, the S-matrix of a launcher radiating into a plasma can be qualitatively well reproduced when the launcher is facing the RF load. The aquarium is furthermore installed on linear guiding system which allows changing the antenna-load distance in a 0-100 millimeter range with an about millimeter-precision. Sweeping the antenna-load distance can be used to assess the launcher’s load-resilience.

        As a first step, the aquarium has been filled up with a salty-water mixture. Hands-on experiments have been conducted with a Tore Supra antenna facing the low power testbed. The optimal salt concentration, maximizing the coupling resistance, has been experimentally assessed and this optimal salt concentration has been confirmed by modeling.

        Finally, the paper discusses the low-power tests of a WEST ICRF launcher and some results are detailed. In particular the launcher is tunable in the specified frequency range (48-60MHz), and even beyond, and its load-resilience has been assessed through the low-power experiments.

        Speaker: Walid Helou (IRFM CEA-Cadarache)
      • 11:00
        P2.016 Physics of the Traveling Wave Array for DEMO with proof of principle on WEST 2h

        To decrease the power density and associated high voltage a distributed antenna system is proposed as ICRH system for the reactor. Among the different solutions a layout made from a set of TWA sections is considered as the most promising [1, 2]. It optimizes coupling to the plasma, is load resilient and avoids large values for the VSWR in the feeding lines. The total radiated power scales as the number of independently fed sections such that high reliability can be expected. A first electromagnetic design for DEMO consisting in 16 sections integrated in the breeding blanket is proposed. This system would have a power capability of 50MW in front of a low coupling reference plasma profile with only 15kV maximum strap voltage. The TWA concept for ICRH is innovative and very different from the traditional IC antennas. A test on WEST should provide a proof of principle of the validity of the TWA approach together with a comparison with the existing WEST IC antennas. The chosen geometry of the TWA section is compatible with one unit of the complete set for a future reactor.
        The paper describes the underlying physics, the antenna design and expected performances from complete modeling of the antenna system including its resonant ring feeding layout. A resonant ring feeding is used to ensure that the total generator power is radiated in the plasma. A comparative modeling with the present WEST antennas will also be discussed.
        The proposed extrapolation to DEMO is discussed together with the problem of its integration in the breeding blanket in an accompanying paper [3].
        [1] R. Ragona and A. Messiaen, EPJ Web Conf. 157(2017)03044.
        [2] A. Messiaen and R. Ragona, EPJ Web Conf. 157(2017)03033.
        [3] J-M. Noterdaeme et al., “Progress with the Ion Cyclotron Range of Frequency System for DEMO”, this conference

        Speaker: Riccardo Ragona (LPP-ERM/KMS)
      • 11:00
        P2.018 Wideband polarizers switches and waveguide for Electron Cyclotron transmission lines 2h

        Overmoded corrugated waveguide is used in high power microwave applications such as Electron Cyclotron Heating systems, where it is necessary to transmit high power at very low loss. In the primary propagating HE11 mode, corrugated waveguide is effective over a large frequency bandwidth. This operational flexibility becomes important in multi-frequency systems. For 50-mm diameter aluminium corrugated waveguide nominally designed for the ITER 170 GHz ECH system, the theoretical ohmic loss of the HE11 mode is around 0.3e-3 dB/m. In a possible dual-frequency system at ITER, the theoretical loss in the same waveguide only increases to a manageable 0.8e-3 dB/m at 104 GHz.

        For ECH, it is also necessary to change the polarization state of the propagating wave. A pair of miter bends with different mirror groove depths is useful for this purpose. By rotating both mirrors, the pair can be used to generate any desired polarization state. The design is frequency sensitive, so it can be difficult to achieve polarizers that function properly over a wide bandwidth. General Atomics has built a pair of 63.5-mm waveguide polarizers for the TCV tokamak’s ECH transmission line that are designed to operate at frequencies ranging from 82.6 to 118 GHz. In addition, polarizers have been designed for ITER’s 50-mm diameter transmission line. A computer code that calculates both the required mirror rotation angles and the ohmic losses in the grooves predicts that these polarizers will function properly at both 170 GHz and 104 GHz.

        General Atomics has recently fabricated a new class of waveguide switches with rotary actuators that have up to four waveguide outputs (or inputs). Such switches with three outputs have been supplied for 82.6-126 GHz transmission lines at TCV. These switches are inherently wideband.

        Speaker: James Anderson (Energy General Atomics)
      • 11:00
        P2.019 Experimental experience and improvement of NIO1 negative ion sources. 2h

        The ion source NIO1 (Negative Ion Optimization 1) is a versatile multiaperture H- source capable of continuous regime operation, with the plasma generated by a 2 MHz/2.5 kW radiofrequency (rf) power supply and nine beamlet extraction. It aims to partly reproduce the conditions of the much larger ion sources built or in construction for neutral beam injectors of fusion devices in a compact and modular ion source, where effects of individual source components can be rapidly verified and compared to simulation code results.
        Several modifications of the magnetic configuration (both inside the ion source and for the embedded magnets inside the accelerator grids) were investigated, with material ranging from hard ferrite to SmCo to NdFeB; evidence of the effect of the accelerator fringe field inside the ion source is discussed. Saturation of filter field beneficial effect at large (>100 G) amplitudes is discussed, as well as the advantages of crossed field operation.
        The radiofrequency system takes full advantage of the frequency tuning amplifier capability (+/-150 kHz installed, 0/+20 kHz used), and it is worth noting that no adjustment of matching box capacitors was necessary in the last 3 years.
        Major diagnostic systems (including CCD cameras and BES measurement) have been integrated with the acquisition system along with the date measured by several power supplies (including the high voltage supplies, the rf generator and bias supplies, the pressure measurements and the beam currents).

        Speaker: Marco Cavenago (INFN-LNL)
      • 11:00
        P2.020 Maximum Likelihood Tomographic Method for the Analysis of Bolometric Measurements on JET 2h

        The accurate determination of the emitted radiation is an important element in the interpretation of Tokamak performance and in the design of experiments. The spatial distribution of the total emitted radiation is typically determined with quite sophisticated tomographic techniques. On JET, a new tomographic inversion method, based on the Maximum Likelihood, has been very recently developed for this purpose. Its main innovative aspect is the analytic determination of the confidence intervals in the emitted radiation levels. The method is computationally quite fast and can therefore be applied on a routine basis. Together with a systematic use of phantoms, it can have several very interesting applications. In addition to allowing a specific optimisation of the tomography for the main plasma scenarios, it permits also a systematic evaluation of various instrumental issues such as the effect of the noise, the impact of missing channels and the influence of the geometry and of systematic errors in the reconstructions. These potentialities are shown with a systematic analysis of bolometric data collected on JET during the experiments with the ITER Like Wall.

        Speaker: Dr Emmanuelle Peluso (Department of Industrial Engineering University of Rome Tor Vergata)
      • 11:00
        P2.021 Safety Systems in the ITER Neutral Beam Test Facility 2h

        The construction of SPIDER, the experimental device of the ITER Neutral Beam Test Facility (NBTF) devoted to the study of the ion source, is completed and SPIDER will soon start operation. The construction of MITICA, the full-size prototype of the ITER HNBI, is in progress. SPIDER and MITICA operation presents many possible hazards including radiation (D.Lgs.230/95–Cat.A), electrical, explosive gas, laser and fire risks. As these safety risks are typically not limited to a single experimental device, the NBTF team decided to use for people and environment safety an integrated approach, referred to as the NBTF Safety System, capable to deal globally with all risks. Safety in the NBTF is implemented first by design in the plant systems where possible; when safety by design is not sufficient, the safety system uses specific devices for active safety. As many different systems collaborate to guarantee the NBTF safety, a coordinating system, called Central Safety System (CSS), has been implemented to supervise all safety systems and provide the safety responsible officer an overall graphics representation of the NBTF safety. The CSS provides the safety coordination and presentation layer, but it is also required to implement directly safety instrumented functions. The paper will present the requirements of the Safety Systems deriving from the safety risk analysis carried out. It will also describe the design and implementation of the safety components with particular focus on the CSS that uses the Siemens SIMATIC S7 400FH PLC technology along with safe input/output remote nodes connected through double ring-topology Profibus redundant network. Complying with the IEC 61508/61511 standards, the system architecture uses the PROFISafe profile and the human machine interface is based on WinCC OA; special provisions are used for communication of safety-related information with the PLC.

        Speaker: Manuela Battistella (Consorzio RFX)
      • 11:00
        P2.022 Initial port integration concept for EC and NB systems in EU DEMO tokamak 2h

        The integration of heating current drive (HCD) systems in the EU DEMO tokamak must address a number of issues, namely space constraints in the tokamak building, remote handling requirements, breeding blanket penetration, neutron and photon radiation shielding, compliance of penetrations of the primary vacuum with safety and vacuum criteria, and a large number of loading conditions, in particular heat, electromagnetic (EM), and pressure loads in normal and off-normal conditions. A number of pre-conceptual design options of the vacuum vessel (VV) port and the port-plug are under assessment and need to be verified against all requirements and related criteria.

        The identification of the functional (or physics) requirements of the HCD systems remains an ongoing process during the pre-conceptual design phase, hence some initial assumptions had to be made as a basis for development of the design of the vacuum vessel ports and the HCD port plugs.

        The paper will provide an overview of present margins in the functional/physics requirements and the rationale behind the assumptions made in order to allow development of the pre-conceptual design options. Furthermore it will introduce the initial design concepts of the Electron Cyclotron (EC) Launchers and the Neutral Beam (NB) Injectors integrated in the equatorial ports. The NB duct design in DEMO and related issues such as transmission and re-ionization losses will be also addressed.

        This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the EURATOM research and training programme 2014-2018 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission.

        Speaker: Dr Thomas Franke (Power Plant Physics and Technology (PPPT) EUROfusion Consortium)
      • 11:00
        P2.023 Conceptual study of an ICRH system for T-15MD using Traveling Wave Antenna (TWA) sections 2h

        The low aspect ratio (a/R=2.2) D-shaped tokamak T-15MD with toroidal field of 2T on axis is currently under construction in the Kurchatov Institute [1, 2]. Ion-cyclotron resonance heating (ICRH) is considered as an important heating method for this device [1]. In addition, ion cyclotron current drive (ICCD) could contribute to sustain the non-inductive plasma current for long-pulse operation. To decrease the power density and associated high voltage a distributed antenna system is proposed for heating of the future fusion reactor with ICRH/ICCD. Among the different possible solutions a set of Travelling Wave Antenna (TWA) sections is considered as most promising [3, 4]. It optimizes coupling to the plasma, is load resilient and avoids a large Voltage Standing Wave Ratio in the feeding lines. A conceptual design of 2 superposed TWA sections with each 8 radiating straps, is made for T-15MD in view of operation e.g. at 60MHz for 2nd harmonic heating of H plasmas. This antenna, loaded by a simulated density profile of T-15MD, is modeled including its resonant ring feeding system. The ring feeding allows the recirculation of the non-radiated power and the termination of the TWA section on its iterative impedance. The paper describes the antenna design, the feeding ring tuning algorithm and expected performances of this antenna concept. The chosen geometry of the TWA sections is compatible with that of a future reactor and therefore this antenna in T-15MD represents also a testbed for DEMO.
        [1] E.A. Azizov, et al., Status of project of engineering-physical tokamak, in: Proc.23rd IAEA Fusion Energy Conf., Daejon, Korea, 2010, Rep. FTP/P6-01.
        [2] P.P. Khvostenko, et al., Fusion Eng. Design 124, 114-118 (2017)
        [3] R. Ragona and A. Messiaen, EPJ Web Conf. 157 (2017) 03044.
        [4] A. Messiaen and R. Ragona, EPJ Web Conf. 157 (2017) 03033.

        Speaker: Jozef Ongena (Plasma Physics Lab LPP/ERM-KMS TEC Partner)
      • 11:00
        P2.024 Experimental Studies on Arc Chamber Failure Mechanisms on DIII-D Neutral Beam System 2h

        Neutral Beam Injection (NBI) is used for non-inductive heating, current drive, fueling and diagnostics in most major magnetic confinement fusion devices. The DIII-D device comprises eight NBI ion sources based on the US Common Long Pulse Source (CLPS), with a total output power of 20 MW.
        Here we report on efforts to improve performance and longevity of the NBI system by initiating a R&D program aimed at studying the most common failure mechanisms for the ion sources. Toward this end, a filament driven plasma chamber has been constructed that attains plasma parameters similar to the arc chamber of a NBI ion source. This Miniature Arc Chamber Experiment (MACE) has a diagnostic suite that includes Langmuir probes, spectroscopy, infrared imaging and mass spectroscopy.
        A report on investigations into two common failure mechanisms is presented here: Firstly, a failure mechanism observed during helium beam operations on DIII-D that results in electrical breakdown of the insulation material that separates the filament plates from the anode. The fault is reproduced in MACE and the proposals for amelioration of the issue will be discussed. Secondly, the failure of the water-cooled Langmuir probes that are necessary for beam current control will be discussed. The original probe design requires a complicated manufacturing technique involving several intricate brazes. When this component fails, it can produce a water leak that impacts operational availability. On MACE, we are testing a redesign of this component to provide a more robust solution.

        *This work supported in part by the US DOE under DE-FC02-04ER54698.

        Speaker: Brendan Crowley (DIII-D National Fusion Facility General Atomics)
      • 11:00
        P2.025 Preliminary conceptual design of the DTT EC heating system 2h

        The Divertor Tokamak Test (DTT) facility has been proposed in the European roadmap to study solutions to mitigate the issue of power exhaust in conditions relevant for DEMO. The Italian DTT tokamak [1] (BT=6T, IP=5.5MA, R0=2.08m, a=0.65m and pulse duration of 90-100s) is being designed to allocate the optimal divertor magnetic configuration under reactor relevant power flow (PSEP/R>15 MW/m) in the scrape off layer. A mix of three heating systems (ECH, ICH and NNBI) will equip the machine to reach the target value of 45 MW at plasma. The present reference design considers a capability of 20-30 MW of EC power at plasma to support and assist different tasks such as bulk electron heating, non-inductive current drive, avoidance of impurities accumulation and MHD control. The gyrotron sources (1MW/170GHz/100s) will be based on the depressed collector technology with 50% efficiency and will exploit the experience gained in developing the solutions for ITER. Two Solid State High Voltage Power Supplies will feed the gyrotrons, the main one for the cathode (−55 kV, 50 A) and a second stage for the anode (35kV, 0.1A). A Transmission Line (TL) with 90% efficiency and 1 MW power handling is being considered and two solutions are studied: evacuated waveguide, as in ITER, and quasi-optical multiple-beam TL, in use at W7-X, are presented and discussed in terms of layout, dimensions and theoretical losses. The conceptual design of equatorial and upper launchers based on the front steering concept is being developed to reach the required deposition location. The EC wave absorption efficiency has been investigated and is presented here, considering a selection of injection angles and launching points with dedicated beam tracing calculations using the GRAY code [2].
        [1] R. Albanese et al, Fusion Eng. Des. 122 (2017) 274
        [2] D. Farina, Fusion Sci. Technol. 52 (2007) 154

        Speaker: Saul Garavaglia (IFP-CNR)
      • 11:00
        P2.026 Advanced NBI beam characterisation capabilities at the recently improved test facility BATMAN Upgrade 2h

        The test facility BATMAN was dedicated since its start in 1996 to the development of radio frequency driven negative hydrogen ion sources for ITER NBI with focus on formation and extraction of negative ions, technological developments and improved concepts. During 2017 the test facility has been upgraded in order to replace the former extraction system (which was derived from a positive ion accelerator from ASDEX Upgrade) with a new ITER-like extraction system comparable in size to an ITER beamlet group. In addition to the standard three grids extraction system a repeller electrode upstream of the grounded grid is installed which can be positively biased by 2 kV with respect to the grounded grid for reducing the amount of back-streaming positive ions and space charge blow up of the beam. For the magnetic filter field a current of up to 3 kA can be driven through the plasma grid, as well as permanent magnets embedded into a diagnostic flange or in an external magnet frame.
        BATMAN Upgrade will focus now on the properties of a beam with an ITER relevant extraction system. One of the main diagnostics is beam emission spectroscopy with line of sight located at two positions from the grounded grid (26 cm and 130 cm) with spatial resolution in vertical direction. A newly developed tungsten wire calorimeter placed just 20 cm downstream of the grounded grid should provide quantitative measurements of individual beamlets, while the previous tungsten wire calorimeter at 2 m distance is still in use for qualitative beam profile diagnostic. Together with a beam dump calorimeter with a crosswise arrangement of thermocouples, beam divergence and uniformity can be studied. This is accompanied by modeling of the meniscus formation, the beam optics, and the beam transport up to the calorimeter with simulated BES diagnostics.

        Speaker: Ursel Fantz (Max-Planck-Institut für Plasmaphysik)
      • 11:00
        P2.027 Equatorial electron cyclotron port plug neutronic analyses for the EU DEMO 2h

        Within the Power Plant Physics and Technology (PPPT) programme in the EUROfusion Consortium design activities are currently in progress for the development of a DEMOnstration Fusion Power Plant (DEMO). In this framework, the design of the machine and the integration of in-vessel components require neutronics analyses fundamental to verify the tritium self-sufficiency, the shielding requirements and the structural integrity of its components. In particular, the penetrations in the blanket and in equatorial port plug introduced by the electron cyclotron (EC) heating system, namely due to openings for the antenna waveguides, can lead to significant leakage of neutrons which may increase the nuclear loads of the superconducting toroidal field (TF) coils and material damages in the vacuum vessel (VV) beyond design limits as well as reduce the tritium breeding capability. In this study a three-dimensional MCNP calculations were conducted for the pre-conceptual designs of the EC port plugs and shielding optimisation were performed in order to ensure that the DEMO design limits are not exceeded.
        Two configurations of the EC heating system both featuring 8 square 63.5 mm × 63.5 mm waveguides (WGs) arranged in two horizontal stacks or a single vertical stack were analysed and shielding solutions for EC port plug proposed. In the first configuration the main focus was on the nuclear heating of the TF coils and in the second case the neutron induced damage of the VV. Analyses were performed using a DEMO Water Cooled Lithium Lead (WCLL) with integrated EC configurations model of a half of the sector (i.e. 10° model) and relevant ones repeated with a full sector model (i.e. 20° model) to test the reliability of results. Additionally, the effect of the radiation on the WGs closest to the plasma was analysed as well as the impact on Tritium Breeding Ratio.

        Speaker: Aljaž Čufar (Reactor Physics Deartment Jozef Stefan Institute)
      • 11:00
        P2.028 Mechanical design of the high powered helicon antenna and strip line feed in the DIII-D tokamak* 2h

        A high-powered “comb-line” helicon antenna for use within the DIII-D Tokamak is currently in design and fabrication at General Atomics. The antenna will drive current in high beta discharges using electromagnetic helicon waves. The high powered helicon antenna (HPHA) is expected to couple up to one MW of power into DIII-D plasmas at a frequency of 476 MHz. The antenna design includes 30 individual, inductively coupled modules fastened to six internally water-cooled Inconel back-plates. The end modules have special design features to provide RF stripline feed attachment points. Each back-plate is mounted to a pedestal, which is secured to the vessel wall via a combination of studs and welded brackets.

        The mechanical design includes features to minimize and survive disruption forces, thermal loading from RF losses, vacuum vessel bake (350C), and installation considerations. Thermal stress design challenges include plasma heat loads, thermal ratcheting, and bake-out cycles that result in thermal growth differences between the strip line and the vessel. Splitting the back-plate into six sections each with a central pedestal support significantly reduces the disruption loads. A single strip line folded in half, fed near the fold provides a 180° phase shift between the strip line module connections for driving the two module straps. This looped mono RF coaxial stripline design was chosen for optimal properties in space limitations, ease of attachment, installation, structural rigidity and RF tuning ability. The strip lines are supported by a quarter-wavelength “stub”, mid span, for support during thermal cycling and plasma disruption events. Multiphysics FEA analyses are performed to optimize the geometric shapes to meet the aforementioned design challenges.

        A description of the mechanical design, fabrication, installation, and analyses of the HPHA and stripline feeds will be presented.

        • Work supported in part by the US Department of Energy under DE-FC02-04ER54698 and DE-AC02-09CH11466.
        Speaker: Benjamin Fishler (Fusion Group General Atomics)
      • 11:00
        P2.029 Review of the JET ILA Scattering-Matrix Arc Detection System 2h

        Arc detection is an essential protection system for high power RF systems. It is commonly realised by monitoring the Voltage Standing Wave Ratio (VSWR) in the transmission lines. The JET ILA is a load tolerant ICRF antenna composed of 8 short straps grouped in 4 Resonant Double Loops (RDLs). In this type of antenna, there is a low impedance section in which the standard VSWR protection is ineffective.

        The Scattering-Matrix Arc Detection (SMAD) was proposed and installed on JET [1] to protect the low impedance section around the T-junction against arcing. It is based on a consistency check of the RF signals around this section using a table of correlation coefficients obtained from RF modelling.
        This contribution reviews the SMAD protection system and its recent improvements, the conditions in which it is essential to protect the antenna during operation, the commissioning of the system and its sensitivity to the input signal levels and accuracy.

        The SMAD error remains small in the full ILA operation frequency range (28-51MHz) during operation on L-mode and H-mode plasmas, showing that both the RF model of the antenna circuit and the measurements are sufficiently accurate for protection purposes. Moreover, due to its insensitivity to the RF coupling properties and the fast (2μs) FPGA error calculation, the SMAD is suitable for detecting arcs during the ELM cycles.

        The time delay to issue a SMAD trip is typically set to 6-8μs (3-4 FPGA cycles), which is confirmed by scope measurements and fast data signals analysis. The time delay between the trip signal and the effective removal of the power in the generator output transmission line is measured to be about 25μs, which is within the general protection specifications of the ICRH system in JET.

        [1] M. Vrancken et al., Fus. Eng. Des. 84 (2009) 1953-1960.

        Speaker: Pierre Dumortier (LPP-ERM/KMS)
      • 11:00
        P2.030 The WEST plasma control system: Integration commissioning and operation on the first experimental campaigns 2h

        The WEST tokamak is aiming at testing ITER like divertor component. This requires to address new control challenges like X-point configuration magnetic control or heat loads control in metallic environment and event handling challenges to sustain long duration H-mode that are in line with ITER needs.
        To address these requirements, a new Plasma Control System (PCS) has been built using a generic version of the DCS (Discharge Control System) Real-Time (RT) framework that is currently used on ASDEX-Upgrade and offers enough flexibility to be adapted to any tokamak. Based on a segmented approach of the plasma discharge, DCS is now the central part of the WEST plasma control system, managing the different actuators and reading the RT data from a large set of diagnostics. In order to reach this goal, several modules have been developed for the synchronization and the communication between DCS and the WEST CODAC infrastructure. The WEST controllers’ layout has been built using Matlab Simulink while the RT source code has been generated with the Simulink coder toolbox. This approach allows asserting the controllers’ performance by coupling them to a tokamak simulator prior to the commissioning phase.
        This contribution will start by summarizing the different concepts used to build the WEST Plasma Control System. The details of the modules developed to integrate DCS into the global WEST CODAC infrastructure will be discussed. An overview of the different plasma controllers will be also presented. The last part of the paper will focus on the results obtained so far highlighting the performance of the system and on the analysis tools used during operation. We will also discuss the ability of the WEST PCS concepts to deal with the machine protection issues.

        Speaker: Rémy Nouailletas (IRFM CEA)
      • 11:00
        P2.031 The dud detector: an empirically-based real-time algorithm to save neutrons and tritium during JET DTE2 2h

        Operations using deuterium-tritium mixtures are envisaged on JET in 2019-2020 (DTE2). Each plasma discharge will be a precious resource during this campaign, being both tritium and neutron budget limited. During DTE2 it will be mandatory to promptly detect and safely terminate those plasma discharges which do not achieve the expected target parameters, due to unsatisfactory plasma performances (i.e. low beta, low fusion power, non-stationarity) or unhealthy conditions that compromise the experiment goals (i.e. inadequate heating power, deleterious MHD, incorrect fuel mixture). A real-time detector of underperforming discharges (dud) algorithm has been developed for this purpose. The algorithm will calculate and monitor the time evolution of plasma performance indicators, which are then used to trigger alarms. Among these, the confinement scaling H98y,2 and the reactivity normalized to the plasma stored energy are the most promising indicators, since they can be easily, yet robustly, estimated based on the available real-time signals on JET. Alarm thresholds for such indicators have been empirically tuned over a wide database of advanced tokamak, baseline and hybrid plasmas. Notably, the robustness of such alarm thresholds will be tested in high heating power regimes and in presence of different isotope mixtures in upcoming JET campaigns. The performance of both detection and alarm generation has been characterized and documented. Coupling the dud detector to other real-time controllers (i.e. radiation peaking detector, isotope mixture and mode locking control) and to proper plasma termination strategies has been investigated in this work. Furthermore, a possible synergy with the real-time state observer RAPTOR code, which will provide a model-based expectation of the plasma state, is also discussed.
        *See the author list of “X. Litaudon et al 2017 Nucl. Fusion 57 102001″

        Speaker: Lidia Piron (CCFE)
      • 11:00
        P2.033 Actuator management development on ASDEX-Upgrade 2h

        The future AUG control research is expected to cope with a large number of control tasks using a limited number of actuators in high performance regime. Essential part of a control system for future tokamaks is an intelligent actuator management that will be responsible for allocating the most convenient actuators to the control tasks of the highest importance.

        Activities in this field have been started at DIII-D, AUG and TCV. The first version of the actuator management at AUG was developed for MHD control and impurity accumulation preemption using ECRH with fast mirrors [Rapson et al, FED 2017]. The next step will be to develop actuator management routines applicable to all heating actuators available at AUG: ECRH, NBI, and ICRH. The first application of this will be in beta control both using NBI and ECRH (so far it has been possible only by NBI) and extension of the applicability of the Te profile controller, which currently uses one