27th IEEE Symposium on Fusion Engineering

Marriott Shanghai City Center

Marriott Shanghai City Center

555 Xi Zang Road (Middle), Huangpu District Shanghai 200003 China
George Neilson (Princeton Plasma Physics Laboratory) , Paul Humrickhouse (Idaho National Laboratory)
    • 8:30 AM 5:30 PM
      Mini-course 1 - Plasma Diagnostics Meeting Room 1

      Meeting Room 1

    • 8:30 AM 5:30 PM
      Mini-course 2 - Plasma-Material Interactions: Fundamentals and Applications Meeting Room 2

      Meeting Room 2

    • 6:00 PM 8:30 PM
      Welcome Reception 2h 30m Grand Ballroom

      Grand Ballroom

    • 8:00 AM 10:10 AM
      M.PLN: Plenary M Grand Ballroom

      Grand Ballroom

      • 8:00 AM
        Opening Remarks 10m
        Speaker: George Neilson (Princeton Plasma Physics Laboratory)
      • 8:10 AM
        Opening Remarks 10m
        Speaker: Prof. Jiangang Li (Institute of Plasma Physics, Chinese Academy of Sciences)
      • 8:20 AM
        Welcome Message 15m
        Speaker: Delong Luo (Director, ITER-China)
      • 8:35 AM
        Conference Logistics 5m
        Speaker: George Neilson (Princeton Plasma Physics Laboratory)
      • 8:40 AM
        Technical Program Overview 10m
        Speaker: Paul Humrickhouse (Idaho National Laboratory)
      • 8:50 AM
        Overall Status of the ITER Project 40m

        The ITER project, established by an international agreement among seven Members (China, the European Union, India, Japan, Korea, the Russian Federation and the United States of America), is a critical step in the development of fusion energy: its role is to confirm the feasibility of exploiting magnetic confinement fusion for the production of energy for peaceful purposes by providing an integrated demonstration of the physics and technology required for a fusion power plant. Rapid progress has been made over the past two years in the design, manufacturing, construction and R&D activities, and the facility is now taking shape at St Paul-lez-Durance in southern France.

        Supported by impressive achievements in fusion technology R&D, manufacturing of ITER components is advancing in factories and laboratories around the world. The international collaboration formed around the production of superconducting magnets for the ITER tokamak has produced over 600 t of Nb3Sn and 300 t of NbTi superconducting strand. 90% of the superconductors required for the ITER magnets are now complete, contributed by 6 out of the 7 ITER partners. Winding packs for the first 4 toroidal field coils have been produced and stacked in the EU and Japan, and central solenoid and poloidal field coil fabrication activities on the first-of-kind coils are underway in partners’ factories in China, France, Russia and USA. Successful tests of prototype high temperature superconducting leads for ITER magnet systems using Bi-Sr-Ca-Cu-O (2223) tapes have also been completed and series production of the current leads has been launched. Fabrication of the vacuum vessel is moving forward, with structures being manufactured under the responsibility of four contributing Domestic Agencies. Manufacturing of the thermal shield is also in progress, and the cryostat elements delivered to the ITER site by India are currently being assembled into large-scale sections of the cryostat (~29 m diameter  ~29 m height). Substantial elements of the power supply and cryogenic systems have also been delivered and several captive (water) drain tanks have been installed, the first equipment incorporated in the Tokamak Complex and the first steps in a multi-year on-site installation programme of tokamak and plant systems which is about to be launched.

        ITER Management is continuing its efforts to strengthen project integration, streamline decision making and ensure the efficient use of project resources while accelerating construction activities. During 2015 and 2016 the ITER Organization and Domestic Agencies worked closely to redevelop the project baseline schedule, providing a realistic framework for the completion of construction while meeting the Members’ budget constraints. The ‘staged approach’ strategy endorsed by the ITER Council in November 2016 has established a target for First Plasma of December 2025 as the earliest technically achievable date, with the transition to DT operation scheduled for December 2035.

        The presentation will review the progress made in developing the advanced technologies required for ITER and in the manufacturing of major components, describe the status of construction of the ITER facility, discuss measures taken to establish a more effective project organization and summarize the revised baseline schedule.

        Speaker: Dr Bernard Bigot (ITER Organization)
      • 9:30 AM
        CFETR- New Design and R&D Activities 40m

        Roadmap of MFE research in China has been discussed by MF research community after CN joint ITER. The most consensus conclusion is showed in figure 1. The key step after ITER for FE development in China is CFETR. It will be the next key device for CN MFE program and will aims to bridge the gaps between ITER and the demonstration reactor DEMO.

        Mission and objectives of CFETR: 1) a good complementarities with ITER; 2) demonstration of full cycle of fusion energy; 3) demonstration of full cycle of T with TBR over 1.0; 4) long pulse or steady-state operation with duty cycle 0.3 ~ 0.5; 5) develop the new advanced technologies such as diagnostics/control for burning plasma, CW H&CD, materials, RH etc for DEMO.

        Status of CFETR project: 1) the first concept design has been completed; 2) a new design is under developing; 3) some important R&D supported by different channels has made important progress; 4) further budget support for engineering design and R&D activities will be provided by government soon possibly.

        • Both first design and the new design will aim to operate in two phases. Steady-state operation and tritium self-sustainment will be two key issues for the first phase with a modest fusion power up to200 MW. The second phase aims validation for DEMO with a FP over 1 GW.

        • Advanced H-mode physics, higher TF magnetic fields up to 7T, CS coil up to 14 T, larger size (R, a/b ), less number of TF magnets for higher accessibility, high frequency ECRH & LHCD together with off-axis negative-ion neutral beam injection will be used for achieving steady-state advanced 1 GW fusion power operation for the new design. The more detailed design information and new R&D activities will be introduced in the paper.

        Further efforts and challenges: CFETR should get financial support for two phases respectively:

        • The budget for engineering design and some R&D have been approved;

        • To be approved for CFETR construction by government will be a great challenge and further significant efforts will be needed.

        • Wide international exchanges and collaborations will be promoted and
          welcome !

        Roadmap of CN MFE research

        First design version of CFETR

        New design version of CFETR


        [1] Wan Y.X, Li J. Liu Y. Wang X.L and CFETR team “Overview of the Present Progresses and Activities on the Chinese Fusion Engineering Test Reactor” 26th IAEA 2016 FEC OV-3

        [2] Wan B.N, et al “ Physics Design of CFETR: Determination of the Device Engineering Parameters” IEEE Transactions on plasma science, vol. 42, No. 3, March 2014

        Speaker: Yuanxi WAN (USTC/ASIPP)
    • 10:10 AM 10:40 AM
      Break 30m
    • 10:40 AM 12:40 PM
      M.OA1: Experimental Devices I Salon 1

      Salon 1

      • 10:40 AM

        The NSTX Upgrade (NSTX-U) team recently completed a scientifically productive research campaign with 10 run weeks of operation. NSTX-U achieved H-mode on the 8th day of operation, surpassed the maximum magnetic field (achieved Bt = 0.65 T) and pulse-duration (achieved 2 sec long pulse) of NSTX, matched the best NSTX H-mode performance at ~1MA, identified and corrected dominant error fields, and commissioned all magnetic and kinetic profile diagnostics. In addition to the new centerstack of NSTX-U, which will ultimately double the maximum field and current capability relative to NSTX, NSTX-U also has a more tangential second neutral beam injector (NBI). NSTX-U researchers discovered that this second NBI can suppress Global Alfven Eigenmodes, which have previously been observed to influence core thermal electron transport. Thus the second NBI may provide means of modifying fast-ion and thermal transport in additional to controlling rotation and current profiles. Finally, NSTX-U researchers implemented new techniques for controlled plasma shut down and disruption detection and commissioned new tools for disruption mitigation. The 2016 run campaign was interrupted by an internal short in a divertor poloidal field coil, and the NSTX-U team is actively developing a recovery strategy. NSTX-U results and future plans will be described

        Speaker: Dr Rajesh Maingi (PPPL)
      • 11:00 AM
        Technical issues toward the steady state operation at KSTAR 20m

        Fusion reactor needs the steady state operation and sustaining the long pulse operation beyond transient period in term of physics and engineering parameters is essentially one of key requirements in present non DT operation tokamaks.
        Recently, KSTAR reported the long pulse operation beyond 1 miniute at the injected power of about 5 MW and the plasma current of 0.5 MA. The normalized beta is about 1.5 and the total injected energy to the plasma reaches to about 300 MJ. It is shown that the bootstrap current fraction is below 40% and the discharge is interrupted by the surface temperature rise at in-vessel coil current other than physics issues.
        In this talk, technical issues for extending to 100s operation with injected power of 12 MW and the plasma current of 1 MA in KSTAR are discussed in heating, plasma facing components and diagnostic system conjecting from present data of 60s operation and preparation efforts are shown.

        Speaker: Dr Jong-Gu Kwak (NFRI)
      • 11:20 AM
        Status of the ITER Vacuum Vessel Manufacturing 20m

        The ITER Vacuum Vessel (VV) has major functions of being the first confinement barrier and removing nuclear heating during fusion reaction of plasma. Also the VV provide mechanical support for all in vessel components such as Blankets, Divertors, In-vessel Coils, Diagnostics, etc. The VV has been designed as a fully welded torus-shaped, double wall structure with in-wall shielding (IWS) and cooling water between the shells in order to satisfy the main functions. Therefore in accordance with French regulation the VV and ports are classified as Nuclear Pressure Equipment due to presence of radioactive products in the plasma chamber and in water cooled structure. The VV procurements consist of five Procurement Arrangements (PAs) and four direct investments. The PAs have been signed for the fabrication of nine sectors (five sectors by the EU Domestic Agency (DA) and four sectors by the KO DA), IWS (IN DA), upper ports (RF DA), and equatorial & lower ports (KO DA) in 2008 to 2009. These direct investments are Field joint welding, Instrumentations, In service inspections, and Bellows.
        Manufacturing design of VV regular sectors and upper/lower ports have been completed by industries with accommodation of requirements of the RCC-MR 2007 edition and approved by the VV project team and the Agreed Notified Body (ANB). The EU VV sectors are being manufactured by the EU DA with the consortium of Ansaldo, Mangiarotti, and Walter Tosto (AMW). Progress of poloidal segments of the first sector, Sector #5, is about 20 % and other sectors are progressing for manufacturing. The KO VV sectors are also being manufactured by the KO DA with the Hyundai Heavy Industry and progress is about 55 % for the first sector, Sector #6, and about 23 % for the second sector, Sector#1. IN DA with Avasarala Technologyes Limited has completed manufacturing of In Wall Shield (IWS) of amount for about 2.5 sectors out of 9 sectors. Remaining of IWS is being manufacturing in order to complete it in end of 2018. Manufacturing of the first upper port stub extension of RF DA with MAN Diesel & Turbo has been completed and full factory acceptance tests have been completed under inspection of ITER organization (IO), related DAs and the ANB. All related manufacturing dossiers have been reviewed by the IO, related DAs/industries, and the ANB under established procedures. Other components for direct investment are under manufacturing design or procurements according to their planed schedule that will be introduced during presentation.
        In this report, current progress of manufacturing, intermediate manufacturing results, major difficulties/issues with solutions, and future plan will be presented.

        Speaker: Dr C.H. Choi (ITER Organization)
      • 11:40 AM
        MAST Upgrade Divertor Facility: A test bed for novel divertor solutions 20m

        The challenge of integrated exhaust consistent with the other requirements in DEMO-class tokamaks (ITER-like and alternative DEMOs, FNSF approaches) is well-known. The exhaust solution is likely to be fundamental to the design and operating scenarios chosen. While no facility can address all of the challenges, the new MAST Upgrade tokamak can explore a wide range of the aspects related to the divertor plasma. MAST Upgrade has unique capabilities to produce conventional and novel divertor configurations for detailed studies and comparison in a single device. The two closed divertor chambers are each surrounded by 8 poloidal field coils for detailed control of the magnetic geometry, including strike point location, field line length within the divertor, poloidal flux expansion and their variation across the scrape-off layer, whilst keeping the shape of the core plasma unchanged. It will be equipped with neutral beam heating, and a wide range of high resolution diagnostics with a strong emphasis on the scrape-off layer and divertor plasma, allowing new levels of detail in testing of models.

        To extrapolate to future devices where full tests in advance are not feasible, theory-based and semi-empirical models can be used. These models, and their necessary compromises and simplifications, need to be validated and improved using the plasma physics mechanisms expected to be important at DEMO-scale, and this is at the heart of the MAST Upgrade programme. Possible paths to confident performance predictions will be outlined, with the role of MAST Upgrade indicated.

        Specific physics areas to be explored include:

        i) Plasma detachment, especially how novel magnetic configurations can make detachment easier and more controllable, e.g. the role of variation in mod(B) along the divertor leg.

        ii) How divertor configuration and detachment state affect the plasma pedestal and access to H-mode.

        iii) Controllability of double null, with potentially different detachment behaviour in upper and lower divertors.

        iv) Behaviour of the inner leg in double null for different configurations (SX, SF, conventional).

        v) Cross-field transport which determines the power footprint on the divertor and the ease of detachment. Longer divertor legs allow cross-field transport to be more effective.

        In most cases the studies will focus on the underlying mechanisms, e.g. plasma filaments / blobs are often actors, and are affected by the scenarios and divertor configurations and state of detachment.

        While MAST Upgrade is not a prototype, in this presentation we will address how it can be used to inform design questions for alternative and novel divertors in DEMO-class devices.

        This work has been funded by the RCUK Energy Programme [grant number EP/I501045].

        Speaker: Dr William Morris (CCFE, UKAEA)
      • 12:00 PM
        Progress of Interface Design between Test Cell and Lithium Systems in IFMIF-DONES 20m

        The Test Cell (TC) in the IFMIF-DONES (International Fusion Material Irradiation Facility- Demo Oriented Neutron Source) facility is the central confinement to envelop the end section of the accelerator, the lithium Target Assembly (TA), and the test modules. The major functions of the TC include: hosting fusion material irradiation experiments in a leak-tight controlled environment, providing sufficient biological shielding to surrounding rooms against the in-TC neutron and gamma irradiation, and allowing media (mainly lithium and helium) and signal/power penetrations between inside and outside of the TC.

        The Lithium System (LS) is connected to its in-TC components through an inlet pipe and an outlet pipe, which penetrate the TC confinement. Two main problems are related to these penetrations: compensation of thermal stresses and minimization of neutron streaming.

        The latest TC design optimization has suggested including the lithium collecting tank, the so called Quench Tank (QT), inside the TC, to find a trade-off solution among the simultaneous and conflicting issues of lithium flow stability, irradiation shielding, penetrations into TC confinement, tritium production, and remote handling access.

        In this paper, the IFMIF-DONES TC design is updated by introducing a TC-Lithium systems Interface Cell (TLIC) below the TC floor to accommodate thermal stress compensation sections of lithium pipes and irradiation shielding materials. In this configuration, the leak tight boundary of the TC is extended to the inner surface of TLIC through the gaps between the lithium inlet/outlet pipes and main body of the TC floor, and the fixing points of the lithium pipes on the TC boundary is arranged on the wall of the TLIC. Inside the TLIC, lithium pipes will be bended in such a way that the thermal stresses are compensated and direct neutron streaming to the LS area is minimized in combination with removable neutron shielding materials. Preliminary thermal mechanical analysis and neutronic simulations are applied to assist the design of the TLIC and the lithium pipe bends.

        The TLIC will be equipped with an air-lock door which can be used as remote handling access to compensation pipe sections and shielding materials during the maintenance periods. Corresponding maintenance scenarios of these components are briefly discussed.


        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 Kuo Tian (Institute for Neutron Physics and Reactor Technology, Karlsruhe Institute of Technology)
      • 12:20 PM
        Development and Application of High Intensity D-T Fusion Neutron Generator HINEG 20m

        Fusion energy becomes essential to solve the energy problem with the increase of energy demands. Although the recent studies of fusion energy have demonstrated the feasibility of fusion power, it commonly realizes that more hard work is needed on neutronics and safety before real application of fusion energy. A high intensity D-T fusion neutron generator is keenly needed for the research and development of fusion technology. However the intensity of D-T neutron generators currently on operation around the world is lower than 1013n/s, which is severely restricting the research capability.

        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 to develop an accelerator-based D-T fusion neutron generator with the neutron yield higher than 10^(15)-10^(16) n/s. HINEG consists of two phases: The first phase, named HINEG-I, aims to have the intensity of 10^(12)-10^(13)n/s in order of magnitude, and the second phase, named HINEG-II, is designed to reach a neutron yield of 10^(15)-10^(16) n/s via high-power tritium target system and high-intensity ion source. HINEG-I has been completed and commissioning with the neutron yield of up to 10^(12)n/s, while the related research on the key technologies of HINEG-II are on-going. HINEG can be used for research and development of nuclear technology and safety, including the validation of neutronics method and software, radiation protection, materials activation and irradiation damage as well as neutronics performance of components. Its application can also be extended to nuclear medicine, radiotherapy, neutron radiography, and other nuclear technology applications. This contribution will summarize all the latest progresses and future plans for the research and development of HINEG.

        Speaker: Chao Liu (Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences)
    • 10:40 AM 12:40 PM
      M.OA2: Divertors and High Heat Flux Components Salon 2

      Salon 2

      • 10:40 AM
        Design and test of W7-X water-cooled Divertor Scraper 20m

        For steady state operation up to 30 minutes pulse duration of the stellarator Wendelstein 7-X, an actively water-cooled divertor will replace the currently installed adiabatically loaded divertor designed for maximal 10 second plasma pulses. Heat load calculations taking into account the effect of bootstrap current have indicated the possible overloading of the ends of the divertor beyond their technological limit. The intention of the additional scraper is the interception of some of the plasma fluxes both upstream and downstream before they reach the divertor surface. To check the effect of the scraper on the divertor for long pulse operation, an adiabatically loaded scraper element will be installed during the phase of the short pulse operation.

        Design activities including the manufacturing and testing of prototypes have been carried out to prepare a possible fabrication of the water-cooled scraper. One scraper is made of 24 identical plasma facing components (PFCs). A PFC is 247 mm long and 28 mm wide. It has 13 monoblocks made of CFC NB31 bonded by hot isostatic pressing onto a CuCrZr cooling tube equipped with a copper twisted tape. Due to pressure drop limitation the scraper is divided into 6 parts of 4 PFCs; each part has 4 PFCs hydraulically connected in series by 2 water boxes (inlet and outlet). Individual full-scale prototypes of PFCs have been successfully tested in the GLADIS facility up to 20 MW/m².

        This paper discusses the challenges of the design and manufacture of the water box prototypes. The scraper and water boxes have to be integrated in a very limited available space and require a very compact design. Prototypes have been manufactured to select the best technology for the water boxes. The results of the successful HHF testing of a component made of 4 PFCs will be presented. The results of these activities have defined the technological basis for a possible fabrication of the water cooled scraper.

        Speaker: Dr Jean Boscary (Max Planck Institute for Plasma Physics)
      • 11:00 AM
        Modeling and Experimental Validation of Physics Enabled by W7-X Scraper Element Divertor Components 20m

        A set of add-on components referred to as Scraper Elements (SE) were designed as a passive solution to a predicted heat flux overload of certain areas of the main Wendelstein 7-X (W7-X) stellarator divertor during long-pulse operation. W7-X will soon begin its first phase of operation using the first set of plasma facing components (PFCs) and a magnetic topology to realize an island divertor configuration (OP1.2). In an island divertor the core plasma is surrounded by an island chain with a helicity determined by the edge value of the rotational transform. The island chain is intersected by the PFCs, leading to heat and particle fluxes that typically manifest as a set of stripes with neutral baffling to guide recycled particles into pumping volumes. One challenge associated with stellarator island divertor configurations is to keep the edge rotational transform constant to maintain the desired topology. A net toroidal current evolving during a discharge will modify this transform unless it is opposed using applied driven current (e.g., ECCD) or by changing the toroidal field to maintain a fixed value. This issue is mitigated in W7-X as one of the optimization goals targeted during the design process was a low bootstrap current. However, this property is persistent only in certain configurations. In some long-pulse configurations of interest the steady-state toroidal current is predicted to be sufficient to modify the edge transform by ~10%. Such a current and thus boundary topology evolution, without mitigation, would sweep heat flux across regions of the divertor with a reduced rating, resulting in transient overload before the island divertor configuration is restored in steady-state.
        We designed the SE as a passive solution to this problem. In the long-pulse, high-power operational phase (OP2) ten SE, one for each divertor unit, would be installed to intercept heat flux to the overloaded areas during the transient phase while receiving a load less than its 20 MW/m2 rating. Due to geometric limitations on the design, the SE continue to receive loading during the steady-state configuration, possibly leading to deleterious effects on pumping of neutral particles and impurity influx into the core plasma. To test both the positive and negative impacts, two inertially cooled SE will be installed in the middle of the OP1.2 campaign. Experiments both before and after the installation of SE will be used to determine if the predicted overload will exist, whether the SE protects the affected areas, and to assess any side effects associated with the SE. Special configurations will be used to mimic the effect of the OP2 configuration evolution, which is not directly accessible in OP1.2. The modeling associated with the design will be presented, along with the plans for validation using the W7-X diagnostic set. Details of the inertially cooled SE design and the finite element modeling performed to determine power and pulse length limitations will also be shown. Support from D.O.E. contracts DE-AC05-00OR22725, DE-AC52–06NA25396, DE-AC02-09CH11466.

        Speaker: Dr Jeremy Lore (Oak Ridge National Laboratory)
      • 11:20 AM
        Status of the ITER Cooling Water System Design 20m

        ITER Cooling Water System (CWS) is designed to reject all the heat generated in the plasma and transmitted to the In-Vessel components through the Tokamak Cooling Water System (TCWS) to the intermediate closed loop Component Cooling Water System (CCWS) and then to the environment via the open Heat Rejection System (HRS).
        The TCWS is designed to remove the total peak heat load of about 1100 MW and is divided into three Primary Heat Transfer System (PHTS) loops, two Chemical and Volume Control System (CVCS) units, a Draining and Refilling system (DR) and a Drying System (DY). The TCWS has a safety role for the primary confinement of radioactive inventory due to Activated Corrosion Product (ACP) and Tritium content in the water. The three PHTS are: Vacuum Vessel (VV PHTS), Integrated Blanket ELMs and Divertor (IBED PHTS) and the Neutral Beam Injectors (NBI PHTS). The VV PHTS has also the safety function to provide the decay heat removal functions even when the other PHTSs are not available during off-normal accidental events like LOCA, LOSP etc.
        The paper describes the main design challenges faced and the changes that have been carried out to prepare the CWS final design phase.
        The paper also reports the main functional requirements for the CWS considering the phased installation, commissioning and operation of the CWS from the preoperational activity to the First Plasma and eventually to the nuclear DT phase.
        Detailed information will be also provided about the physical and functional interfaces between CWS and the main clients (e.g. Vacuum Vessel, In-Vessel Components, Diagnostics, Power Supply, Cryoplant etc.) with the progress on the integration the CWS to the other systems in the Tokamak Complex as well as in the other non-nuclear buildings.

        Speaker: Giovanni Dell'Orco (ITER IO)
      • 11:40 AM
        Virtual Engineering of a fusion reactor: application to divertor design, manufacture and testing 20m

        Owing to ever-growing computational power, Virtual Engineering is transitioning from a possibility to an absolute necessity, and has exciting potential to accelerate the realisation of a fusion power reactor. Virtual Engineering is the use of sophisticated computational modelling to enhance the traditional route of component qualification by providing deeper insight, shortening the design cycle, reducing the burden of costly experimental testing or by answering questions that are simply not possible to answer by testing or real-world measurement. Finite element analysis (FEA), has been used for decades to perform engineering calculations and support design substantiation. In Virtual Engineering, the execution of FEA is highly parallelised, and mathematical optimisation is used to efficiently explore the design space, dramatically reducing design time. Further, typical FEA uses a much simplified and idealised representation of a real component, often taking input from computer aided design models. The advent of high performance computing and 3-D volumetric scanning offers the capability to create a highly accurate virtual “twin” of an as-manufactured component, including any unintended features or imperfections. These virtual components can be virtually tested under realistic conditions to simulate manufacturing, assembly, commissioning or operating phases.

        In this work, the potential of Virtual Engineering in the design and validation cycle is demonstrated for the first time in an application to DEMO water cooled divertor target design, manufacture and testing. First, FEA-based design search and optimisation is used to improve the established tungsten monoblock divertor concept [1]. Fabricated mock-ups are imaged using X-ray tomography and analysed using image-based finite element modelling (IBFEM) to simulate the in-service high heat flux conditions. The IBFEM captures imperfections in the manufacturing process; in the example here an incomplete braze between components of the monoblock mock-up. The response of the FEA model to in-service simulations of heat flux indicated that the stress within the monoblock exceeded design limits leading to mechanical failure. A revised manufacturing procedure was introduced to improve the joining procedure and eliminate voids. The in-service simulations can be used to investigate the impact of an imperfection not only to identify early failure but also to identify acceptable imperfections that do not compromise lifetime performance, avoiding unnecessary component rejection and thus reducing cost and time to realisation. Mock-ups fabricated to the improved joining procedure have been successfully tested by infrared thermography, and under high heat flux at up to 25 MW/m2 with no sign of damage.

        [1] T. Hirai et al., Fusion Engineering and Design, 88 (2013), p1798-1801.

        Speaker: Dr Thomas R. Barrett (CCFE, Culham Science Centre, Abingdon OX14 3DB, United Kingdom)
      • 12:00 PM
        Expermental and numerical investigation on anti-fatigue and anti-thermal shock performance of the divertor first wall 20m

        After years of exploration and development, research of magnetic confinement nuclear fusion is progressed into stage of experimental fusion reactor construction and test. As a key plasma-facing component, the anti-fatigue performance of first wall of fusion reactor receives widely concerns. Due to the fact of enduring both periodic loads of pulse operating mode and shock loads of transient events such as disruption, ELMs etc, the coupled fatigue responses of material and structure are in the state of very complex. It is significant and necessary to research the coupled mechanism of fatigue by both transient and periodic heat loads, which will be beneficial to develop the key and new technology of promoting anti-fatigue performance for the first wall of fusion reactors. With such motivations, a multi-purpose experimental platform integrated both high heat flux loading and heat shock loading as well as mechanical force loading is established. And meanwhile, a relative complete finite element analysis method based on a full coupled thermal/structural heat transfer equation with consideration of elastic/plastic constitutive relation as well as multiple kinds of thermal physical effects such as melting, solidification, evaporation etc. is established. Based both experimental and numerical works, the thermal/mechanical response of first wall and its fatigue performance are investigated. It is concluded that the fatigue life time of first wall is decreasing nonlinearly with increase of heat loads magnitude,and the coupled periodic normal loads and shock loads induced by transient events will greatly reduce the fatigue life time of first wall. And serverial techniques to improve anti-fatigue and anti-thermal shock performance are explored with both experiments and numerical tests.

        Speaker: shenghong Huang (university of science and technology of china)
      • 12:20 PM
        Adjourn 20m
    • 10:40 AM 12:40 PM
      M.OA3: Inertial Fusion Engineering and Alternate Concepts Salon 3

      Salon 3

      • 10:40 AM
        Status of the ICF program in China 20m

        The inertial confinement fusion (ICF) program in China is to perform thermonuclear ignition and burning, and has made great progress so far. For ICF drivers, the laser facilities of SG-IIU (upgrading) and SG-IIIP (prototype) with both 8 beams and total laser energy output of tens kJ for 0.35μm wavelength (same below) and of SG-III with 48 beams and total laser energy of ~ 200kJ are operating and serving for target physics experiments. The ignition facility with laser energy output of more than 2.0MJ has been considered. The SG-III are very important to provide a scale-up from physical experiments in laser energy of hundreds kJ to MJ ignition. In addition, we are deeply investigating the target physics toward ignition. The innovative ignition schemes different from that performed on NIF, where the fusion ignition was unsuccessful, have been proposed as well.

        Speaker: Mr Wanguo Zheng (Research Center of Laser Fusion, CAEP. )
      • 11:00 AM
        Magnetized Target Fusion at General Fusion 20m

        Magnetized Target Fusion (MTF) involves rapidly compressing an initial magnetically confined plasma by >300X volume compression. If near adiabatic compression is achieved, the final plasma the plasma can be heated to > 10 keV, and confined inertially to produce interesting fusion energy gain. General Fusion is developing a compression system using pneumatic pistons to collapse a cavity in formed in liquid lead-lithium, heating a plasma target such as a spheromak or spherical toroid trapped in the cavity. With a low-cost driver, straightforward heat extraction, good tritium breeding ratio and excellent neutron protection, the concept is promising as a practical power plant. We will review the plasma formation and compression results achieved so far and our plans moving forwards. Work on the compression system will also be described.

        Speaker: Dr Michel Laberge (General Fusion)
      • 11:20 AM
        Properties of a Clean and Economic Boron Laser Fusion Reactor 20m

        Fusion reactions of protons with the boron isotope 11 (HB11) were considered as extremely difficult and impossible for a power reactor. This changed by several orders of magnitudes using picosecond (ps) lasers with powers >petawatt (PW) igniting fusion in a non-thermal way by direct conversion of laser energy into ultrahigh acceleration of plasma blocks [1]. The HB11 reaction produces primarily only clean helium without nuclear radiation problem. The design of a new kind [2] of a fusion power reactor, (see Fig. 19 of Ref. [3]) contains a reaction unit in the center of the reactor sphere with a cylindrical solid stoichimetric hydrogen-11born fuel (see Fig. 10 of Ref. [3]). The unit is charged at about -1.4 Megavolt within the reactor sphere and the fusion reaction is generated end on at the fuel cylinder by a 30 kJ laser pulse of ps duration. The 2.9 MeV helium (alpha particles) convert their energy into electricity when moving against the wall of the reactor. At a one Hertz operation rate, the current of 780 Amps is converted into ac three-phase electricity resulting in power generation on a profitable level [3].

        The fusion reaction in the cylindrical fuel is trapped by a magnetic field by a 4.5 kilotesla magnetic field for one nanosecond. The trapping field is generated for one nanosecond by the capacitor laser driving device following Fujioka et al. [4]. The conditions for sufficient magnetic trapping of the HB11 reaction for binary reactions in a fuel cylinder of 1 and of 0.2 mm diameter are confirmed by hydrodynamics [2][3] and extended to experimentally confirmed avalanche reactions [5]. Results on physical solutions are reported focusing on direct drive ignition conditions and the theory of avalanche reactions by elastic nuclear collisions [6]. This was elaborated on block ignition by laser pulses of >30kJ-ps producing >GJ energy in the 2.9 MeV alphas. It is estimated that the available 10PW-ps pulses per minute [7] are developed within reasonable time to the 30PW-ps laser pulses for one Hz operation for the new reactor type.

        [1] H. Hora, G.H. Miley, et al. Energy & Environmental Science 3, 479-486 (2010); H. Hora, et al. IEEE
        Transact. Plasma Science 42, 640 (2014).

        [2] P. Lalousis, et al. Laser & Particle Beams 32, 409 (2014); H. Hora & G.J. Kirchhoff, International Patent

        [3] Heinrich Hora, et al. SPIE Proceedings 9515, 951518 (2015).

        [4] S. Fujioka et al. Scientific Reports 3,1170 (2013).

        [5] A. Picciotto et al. Phys. Rev. X4, 031030 (2014); H. Hora, G. Korn, etal. Laser and Particle Beams. 33, 607

        [6] Shalom Eliezer, H. Hora, G. Korn et al. Physics Plasmas 23, 050704 (2016).

        [7] T.R. Ditmire, 2 nd Conf.HPLSE Suzhou/CN 15 March 2016 paper C-12.

        Speaker: Prof. Heinrich Hora (University of NSW Sydney)
      • 11:40 AM
        Fusion chamber dynamics and first wall response in a Z-pinch driven fusion-fission hybrid power reactor (Z-FFR) 20m

        The Z-Pinch driven fusion-fission hybrid reactor (Z-FFR) concept utilizes energetic neutrons produced by D-T fusion to drive a sub-critical fission blanket for energy production. Benefiting from an innovative local-holistic-ignition Z-pinch fusion target and advanced sub-critical fission blanket made of natural or depleted uranium, the Z-FFR has significant advantages in safety, economy and environment, and has great potential to be a millennial energy that could address the main issues of long-term sustainability related to nuclear power: fuel supply, energy production, and waste management. In Z-FFR, the fusion target will produce enormous energy of ~1.5 GJ per pulse at a frequency of 0.1 Hz. Almost 20% of the fusion energy yield, approximately 300 MJ, is released in forms of pulsed X-rays. Radiation hydrodynamics in the fusion chamber were investigated by MULTI-1D simulations. To evaluate the influence on thermal and mechanical loads on the first wall brought by the uncertainties of calculated radiation opacities, as well as limitations from employing a single-group treatment in chamber radiation transport, artificial adjustments of opacities by multiplying a coefficient were adopted in the simulations to increase the design reliability. The thermo-mechanical response in a tungsten-coated Zr-alloy first wall was performed by FWDR1D calculations using the derived thermal and mechanical loads as inputs. The temperature and stress fields were analyzed, and the corresponding elastic strains were conducted for primary lifetime estimations. Both pure tungsten and W-Re alloy were tested on an intense pulsed Z-pinch X-ray source to find the safety thresholds of certain materials for designing requirements, as well as to verify and validate the FWDR1D code.

        Speaker: Dr Jianmin Qi (Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics)
      • 12:00 PM
        Experimental results from the SPECTOR device at General Fusion 20m

        General Fusion (GF) is operating a new sequence of plasma devices called SPECTOR (Spherical Compact Toroid) capable of generating and compressing plasmas with a more spherical form factor, avoiding the concave liner geometry used on previous compression tests at GF. SPECTOR forms spherical tokamak plasmas by coaxial helicity injection into a flux conserver (R= 19 cm, $\lambda$$_{Taylor}$ = 23.9 m$^{-1}$, minor radius of 8.3 cm) with a pre-existing toroidal field created by ≤ 500 kA of current in an axial shaft. The initial poloidal flux of up to 30 mWb and toroidal plasma current of 100 - 300 kA is formed rapidly in the spherical flux conserver during a Marshall gun discharge (850 kA peak, 90 $\mu$s duration), and then resistively decays over a time period of ~2 ms. SPECTOR 1 has an extensive set of plasma diagnostics including a surface magnetic probe array, 3 interferometer chords, visible and VUV spectroscopy, multi-point Thomson scattering as well as a 4-chord FIR polarimeter system in development. SPECTOR 2, 3 are mobile test platforms that can be transported out of the lab for compression tests. Plasma facing surfaces include plasma-sprayed tungsten and bare aluminum, and can be coated with ~5 $\mu$m of vacuum deposited lithium for the purpose of gettering impurities out of the base vacuum and to reduce the gas recycling coefficient of the wall. Working gas has included helium and deuterium. Experimental characterizations have been made of formation dynamics, MHD mode activity, evolution of plasma profiles during its lifetime, and trends in FWHM magnetic lifetime with respect to system control parameters. Control of safety factor profile q($\Psi$) can be achieved through a choice of the amount and axial distribution of poloidal gun flux and the amount of shaft current. Grad-Shafranov equilibria are reconstructed from the surface magnetic data using Caltrans/Corsica. Ideal and resistive MHD stability can be tested with DCON and NIMROD over a range of pressure and current profile parameters. Realistic compression scenarios have been simulated using the 3D MHD code VAC. The SPECTOR geometry is stable for a wider range of plasma parameters than previous experiments at GF. Relatively hot (T$_e$ ≥ 400 eV) and dense (~10$^{20}$ m$^{-3}$) plasmas have achieved energy confinement times $\tau$$_E$≥ 100 $\mu$s and are being used in field compression tests.

        Speaker: Dr Michel Laberge (General Fusion)
      • 12:20 PM
        Radiation Safety Design for the North Pole Neutron Time-of-Flight System at the NIF 20m

        The National Ignition Facility (NIF) at Lawrence Livermore National Laboratory is the world’s largest and most powerful laser system for inertial confinement fusion. During the ignition campaign, the NIF is expected to generate shots with varying fusion yield (up to 20 MJ or 7.1e18 neutrons per shot). Neutron time of flight (nTOF) detectors are fielded in the NIF to measure neutron yield, ion temperature, and downscattering in the cold fuel for D-T implosions. A collimated nTOF line-of-sight (LOS) has been fielded near the Target Chamber (TC) North Pole to examine any possible anisotropy in the cold fuel. A fast scintillator is placed inside a diagnostics hut located above the roof of the Target Bay (TB). The scintillator is located at 21.6 m from the Target Chamber Center (TCC). The line-of-sight passes through the TC, the 60.96-cm-thick concrete 69’9” TB floor and the 76.2-cm-thick concrete TB roof. Radiation streaming through the LOS represented a potential radiation hazard if personnel were accidentally present on the 69’9” floor or on the top of the TB roof. The potential hazard at these two locations is caused by radiation streaming through a 30.48-cm-diameter hole in the 69’9” concrete floor and the TB roof. Additional potential hazard to personnel present on the roof during a shot is caused by radiation scattering off the scintillator. The un-scattered radiation is eventually intercepted by a beam dump made of 45.72-cm-thick iron followed by 30.48-cm-thick concrete. The front surface of the beam dump is located at 23.85 m from TCC. The beam dump is designed to fully intercept the radiation and eliminate skyshine hazard due to neutrons passing through the LOS and interacting with the surrounding air.
        A detailed MCNP model of the TB is used to estimate dose values at the previously identified locations of concern during shots. Before adding the new LOS, the area above the 69’9” floor wasn’t normally accessed or swept before low yield shots (< 1e16 neutrons) due to expected low dose values. The MCNP simulation indicated that adding the LOS will result in a maximum effective dose of 2 mSv on the 69’9” floor during 1e16 shot. Following construction of the LOS, the 69’9” TB floor is routinely swept and no access is allowed during all shots. Similar analysis showed that radiation scattering off the scintillator resulted in higher dose values inside the diagnostics hut. A maximum effective dose value of 3 mSv is expected outside the hut during a 20 MJ shot. Currently, no access is allowed inside the hut during all shots, and access control of the TB roof is required for shots with yield above 1e16 neutrons. Finally, in addition to reducing the skyshine dose, the beam dump effectively eliminated any potential hazard to planes flying over the facility. In conclusion, contribution of the new North Pole nTOF system to dose outside the facility and near the site boundary is negligible.

        *This work performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.

        Speaker: Dr Hesham Khater (Lawrence Livermore National Laboratory)
    • 12:40 PM 1:40 PM
      Lunch 1h Shanghai City Bistro/EZO Restaurant

      Shanghai City Bistro/EZO Restaurant

    • 1:40 PM 3:40 PM
      M.POS: Poster Session M Junior Ballroom

      Junior Ballroom

      • 1:40 PM
        A construction design of helium recovery and purification system on HL-2M 2h

        According to the needs of the development of China nuclear fusion research, HL-2M is being built in Southwestern Institute of Physics (SWIP) as a transformation and upgrade device of HL-2A ,which is a plasma physics and controlled nuclear fusion tokamak experimental platform. The liquid helium cryogenic system (LHCS) with a 500W@4.5K power is under construction to provide a cryogenic and vacuum environment for HL-2M and its related equipments by total capacity of 7000L liquid helium. So it’s necessary to construct a helium recovery and purification system (HRPS), which contains a recovery part, a storage part and a purification part. The recovery part mainly includes 2 piston compressors with capacity 50m3/h and 100m3/h respectively, the storage part mainly includes 4800m3 higher purity helium tanks, 1200m3 dirty helium tanks and 50m3 gasbags, the purification system mainly includes a high pressure helium purification equipment, which consists of 2 purification storages, dryers and a helium purity analyzer. The maximum recovery rate of the system is designed to 150m3/ h, and the helium purification equipments can improve the purity from 95% to 99.999% or even more.
        Key words: HL-2M, LHCS, HRPS

        Speaker: Mr Xin Chen
      • 1:40 PM
        A Method for Diagnosis of Current in PF Magnet based on Inversion of Measured Magnetic Field 2h

        More and more Tokamak devices have been designed and constructed to challenge the controlled fusion problem, such as ITER, J-TEXT, EAST and HL-2M. The coil systems of the Tokamak devices are usually composed of Poloidal Field (PF) coils, Toroidal Field (TF) coils and Central Solenoid (CS) coils. Among them, the PF coils play a role for adjustment of the confining magnetic field to control the fusion plasma. In order to avoid the plasma disruption, it is necessary to know the correlation of the magnetic field in vacuum vessel (VV) with the current in coils and the plasma current and to predict the current distribution of PF coils from a distribution of perturbed magnetic field due to unstable plasma current. In this paper, taking the HL-2M device as an example, a method to predict the current distribution in the PF coils from the perturbed magnetic field in VV is proposed based on inverse analysis of the magnetic field information. Firstly, based on the Biot-Savart’s law, a forward code to calculate magnetic field due to currents in PF coils, TF coils and the plasma current was developed. Secondly, an inversion code to predict current distribution of PF coils from the measured magnetic field was developed based on the conjugate gradient optimization method. The validity of the proposed inversion method and the corresponding numerical codes was investigated through reconstructing PF current distributions from several groups of perturbed magnetic fields with artificial noises.

        Speaker: Mr Zichu Huang
      • 1:40 PM
        A Quasi-Periodic Linear Feeder for the Impurity Granular Injection on DIII-D 2h

        Injection of solid non-fuel pellets has been actively used as a tool for pacing and mitigation of edge localized modes (ELMs). In DIII-D, effective ELM pacing has been demonstrated by high frequency injection of Li and C sub-millimeter spheres, using the Impurity Granule Injector (IGI) [1], which injects granules into the plasma at speeds up to 150 m/s, through impact with a rotating impeller. In the IGI, high frequency granule delivery was accomplished through a vibrational granule dropper, in which high time-average rates are obtained at the cost of lack of period control [2].
        We present a new in-line granule feeder, capable of delivering granules of size 0.2-2.0 mm with no restriction of material properties, at quasi-periodic rates up to 150 Hz, for 0.7 mm diameter Li granules (600 Hz using 0.3 mm granules). The new dropper mechanism combines two piezo in-line units; one to feed, and one to circulate granules that are filtered out of the feeder path. A remotely adjustable filter eliminates granules that are stacked, oversized, or side-by-side to form a single moving granule injection line. The granules fall off the in-line feeder exit one at a time, hence achieving a quasi-periodic delivery at a rate proportional to the exit speed. At drop rates <60Hz, the granule delivery period has a variation of +/- 25%. At higher rates, the periodicity deteriorates. This behavior was studied using high-speed cameras and electrostatic measurements, and the variation appears caused by to gaps that develop in the last centimeter of the injection line, as granules exit off the moving track.
        The linear feeder concept is robust against bridge instabilities and clogging issues, thanks to the simple diverter filter and constant recirculation of granules. Furthermore, the open-top design of the device allows easy access for refilling the device from separate reservoirs, and has easy access for directly monitoring operation and adjustment.
        This paper describes the in-line feeder design details, along with several design iterations. The goal is a robust in-vacuum mechanism that can deliver flow ranging from a single particle to a line of particles at 150 per second, using different sizes and materials from the same apparatus.

        [1] Bortolon et al. Nucl.Fusion 2016
        [2] Nagy Proceedings of SOFE 2015

        This work was supported by the U.S. Department of Energy under DE-AC02-09C11466 and DE‑FC02-04ER54698.

        Speaker: Alexander Nagy (PPPL)
      • 1:40 PM
        A rapid non-destructive inspection method applied to EAST lower divertor by IR thermography technique 2h

        Graphite is used as plasma facing material in EAST lower divertor consisting of hundreds of graphite tiles which are connected to heat sink using screw bolt currently. A soft graphite sheet is inserted between graphite tile and heat sink to improve the ability of thermal conductivity. To evaluate the quality of the thermal contact between graphite tile and heat sink , efficient non-destructive inspection is essential before assmbling divertor to EAST device. This paper introduce a rapid non-destructive inspection method for EAST lower divertor by infrared(IR) thermography which records the surface temperature of each graphite tile . The poor quality of thermal contact can be examined by comparison of the transient thermal response of surface of graphite tiles in symmetric region of the same divertor module at a rapid switch from hot to cold water flowing in the tube welded to heat sink. Three-dimensional (3D) thermal finite element (FE) analyses have been performed to simulate the excellent quality of thermal contact and as a reference for the experimental observations obtained by IR thermography .

        Speaker: Mr Yanwei Liu (Institute of plasma physics, Chinese Academy of Sciences)
      • 1:40 PM
        Advances in Technology, Performance, and Power and Polarization Measurements for the ECH System on DIII-D 2h

        The DIII-D electron cyclotron heating system (ECH) has six gyrotrons installed at this time and is operated for injection into the plasma of rf power up to 3.6 MW at 110GHz frequency. The rf power injected at the tokamak is measured on a shot to shot basis with a calibration based on the heat deposition in the gyrotron water cooling circuits.
        A technique for calibrated ECH power measurement using both orthogonal polarizations of the transmitted rf wave at the last miter bend in the line was tested. Polarization scans for each system show H-plane and E-plane rf waveforms can be combined using square law detectors to provide a reliable calibrated power signal at the closest access point near the tokamak. Previous attempts to calibrate the power at this location were limited by the detection of only one polarization component at the last miter bend.
        The elliptical polarization of the injected rf wave was measured for all the transmission lines. Final wave polarization is controlled using a pair of corrugated mirrors installed in miter bends. This allows for the launching the extraordinary mode that is absorbed at the second resonance for different plasma configurations and injection angles for heating or current drive. Two types of corrugated mirrors with different power handling capability were investigated. The smaller size mirrors performed better in terms of ellipticity control than the larger size mirrors with tapered mode convertor.
        System upgrades include a new operating frequency, a new upgraded collector map measurement system better adapted to different configurations of collector RTDs, 4-port power monitors for reflected power and polarization measurements, and robust beam refraction protection using a density interlock, visible cameras, visible light monitors, and reflected power monitors used as sniffers. Increased levels of reflected power can indicate low absorption in the plasma or arcing in the launcher. Updates of the protection circuits to allow recovery after faults such as rf dropouts are included in the future plans for the system.
        A newly designed depressed collector gyrotron in the 1.5 MW class, operating at 117.5 GHz, will be added to the ECH system. This new gyrotron has achieved 1.8 MW for short pulses during factory testing and is expected to be installed and operated in the spring of 2017. The system expansion is expected to reach a total installed power of over 11 MW and a total injected power of 8 MW with ten gyrotrons, some operating at 117.5 GHz and some at 110 GHz.
        This material is based upon work supported 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 DE FC02-04ER546981.

        Speaker: Dr Mirela Cengher (General Atomics)
      • 1:40 PM
        An approach to the study of crack initiation at the divertor tungsten target plates of ASDEX Upgrade 2h

        The solid tungsten divertor tiles at ASDEX Upgrade experimental device have been gradually installed as substitution for W coated graphite tiles in 2012.
        ASDEX Upgrade is equipped with an adiabatically loaded divertor as a compromise between available heating power, plasma discharge length and heat removal capability of divertor tiles. Accordingly, the design of the new solid tungsten plates has been conditioned by in-vessel surrounding and supporting structure. The full-scale prototypes, dimension of 250x80x15 mm³, have been intensively tested in the high heat flux test facility GLADIS (Garching Large Divertor Sample Test Facility) [1]. The GLADIS heat loading profiles are Gaussian with central heat flux of 10 - 30 MW/m², resulting in an integrated absorbed power of the W tile between 100 and 280 kW. Thus simulating the expected highest power and energy loads in ASDEX Upgrade. The corresponding measured surface temperatures reached values between 1500 °C and 3300 °C. In addition, cyclic loading tests have been performed with 200 cycles at 10.5 MW/m², 3.5 s duration. These applied loads correspond to the expected thermal loading of about 4 years of ASDEX Upgrade operation with approximately 50 high power discharges per campaign.
        During the cyclic loading in the GLADIS facility, no crack initiation at the tungsten tiles has been detected. However, after one campaign of AUG operation (about 1200 plasma shots) almost all tungsten divertor tiles exhibit cracks. The inspection of the plasma exposed tiles has revealed 126 tiles with deep cracks. Nearly all of 128 tiles have shown shallow cracks in the high heat load region. Nevertheless, none of these divertor tile damages have caused an operational interruption of ASDEX Upgrade. A comprehensive investigation of the damages has been performed to find out the origin of the crack initiations [2].
        This paper is presenting a bundle of numerical simulations of the ASDEX Upgrade solid tungsten divertor tiles on the basis of the theoretical hypothesis for failure of brittle materials. Accordingly, thermomechanical analyses with cyclic loading, simulating both the GLADIS and ASDEX Upgrade load profile have been performed. Additionally an assessment of the crack initiation induced by material fatigue under thermal cyclic load has been studied. Finally, the design optimisation considerations of divertor tiles are discussed.

        [1] JAKSIC, N., et al., "FEM investigation and thermo-mechanic tests of the new solid tungsten divertor tile for ASDEX Upgrade", Fusion Engineering and Design 88 (2013) 1789–1792, http://dx.doi.org/10.1016/j.fusengdes.2013.04.048.

        [2] HERRMANN, A., et al., "Experiences with a solid tungsten divertor in ASDEX Upgrade", Journal of Nuclear Materials and Energy (2016)

        Speaker: Mr Nikola Jaksic (Max Planck Institute for Plasma Physics, EURATOM Association)
      • 1:40 PM
        Analysis of Short Circuit Fault for 4.6GHz/6MW LHCD High Voltage Power Supply 2h

        4.6GHz/6MW Lower Hybrid Current Drive (LHCD) is one of plasma current heating methods for Experimental Advanced Superconducting Tokamak (EAST). High Voltage Power Supply (HVPS) is the power supply subsystem of 4.6GHz/6MW LHCD system, which was designed, built and accepted successfully by Chinese National Development and Reform Commission in 2015. Then the new system has been in use for the 2015 EAST campaign. This paper presents the structure of 4.6GHz/6MW LHCD-HVPS and its transient operation state when its klystron load has short circuit fault. In order to protect the klystron and HVPS itself, the short-circuit fault and its transient process are analyzed and calculated in detail. And a three-electrode gas switch has been built to eliminate the short-circuit fault in microseconds. In addition, the effectiveness of the three-electrode gas switch has been verified by simulation and experiment result. The HVPS has been used in 4.6GHz/6MW LHCD system and it has good performance for the entire 2015 EAST campaign.

        Speaker: Mr Yang Zhigang (Institute of Plasma Physics, Chinese Academy of Sciences)
      • 1:40 PM
        Application of Contour Fitting Method in CFETR VV Assembly 2h

        China Fusion Engineering Test Reactor (CFETR) vacuum vessel (VV) is a forming and welding product, which is composed of formed shells and ribs, and welded to a toroidal double shell structure device. However, there are manufacturing errors in the manufacturing process, which will lead to accumulative deviation of the VV dimensions and affect the VV contour and the welds quality. To reduce and decentralize errors, the contours of welded parts are measured and fitted with least-squares method. According to the fitting result, new benchmarks are reset in form of controlling points for the next assembly. This paper is about the application of contour fitting method in manufacturing of the first 1/32 VV sector of the CFETR 1/8 VV mock-up.

        Speaker: Mr Xiaosong Fan
      • 1:40 PM
        Application of EPICS in HL-2A Host Centralized Control System 2h

        In order to improve the capability of Tokamak measurement and control system, centralized measurement and control system with compatibility and continuous upgrade ability is needed to build. EPICS has advantages of safety, easy expansion etc. in this paper, HL-2A host centralized control system based on EPICS is developed and designed. Communication between SoftIOC and PLC was realized using the s7nodave device driver module. Through the CSS application development, real-time communication between OPI layer and subsystem has been realized, and the PLC of the subsystems can be integrated to the EPICS control system. In order to manage subsystem of HL-2A host device more intuitive and convenient, the state mode of HL-2A centralized control system is deeply researched based on the theory of CODAC state machine. Standard system states are defined, subsystems are converted to the state in a given order. The state of other system synchronize as the state of a system changes. HL-2A host control system based on the state machine mode is not only easy to centralize management of each subsystem, but also can realize the interlock and protection among subsystems.HL-2A host centralized control system prepare for design on the next generation device host centralized control system.

        Speaker: Jie Xu
      • 1:40 PM
        Assembly methodology and tools developed for Tore Supra transformation into WEST platform 2h

        Since 2013 Tore Supra has turned to WEST (Tungsten (W) Environment in Steady state Tokamak) Platform targeted at supporting the ITER divertor detailed design, manufacturing and future operation. The major changes included the modification of the magnetic configuration (from limiter to divertor), the replacement of carbon plasma facing components by new tungsten plasma facing components (PFCs), the upgrade of the high frequency heating systems and diagnostics. This resulted in replacing 100% of the inner components and about 80% of port-plugs components.
        The main technical challenges consisted in:
        - Assembling interlinked new elements in the existing device;
        - Designing the in-vessel interfaces and accurately positioning the components in order to maximize the plasma volume;
        - Designing and in-situ manufacturing of the divertor coils.

        The key phase of the assembly sequence was the in-situ Divertor coil construction, which required developing specific techniques such as in-situ brazing, wrapping, etc.… and controlling perfectly operations as no repair is possible after the divertor structure closing. The assembly work has been organized in several steps before and after the in-situ Divertor coil construction.

        Sub-millimeter metrology was key to provide input data on the existing environment, to transfer the magnetic references from Tore Supra to WEST and also accurately position the elements. In addition, the CEA virtual reality room was widely used for kinematics definition and, in a second step, for generic tooling validation.

        The paper will describe the main sequences defined for WEST assembly and associated qualification processes applied before and after component installation. Lessons learned in terms of design to assembly and associated tooling cycle will be detailed.

        Speaker: Cyril BRUN (CEA)
      • 1:40 PM
        Automatic Deployment of a Nuclear Fusion Experimental Data Storage Cluster 2h

        With the rapid growing of experiment data, widely using of distributed database or file system will be the trend in future fusion storage. For such storage systems, a cluster is required, but its setting up and service’s deploying are challenging. Besides, plenty of configurations should be set carefully, including environment variable and software runtime, which is a complicated task. Apart from that, due to the variety of hardwares and difference between test and production environment, particular setup procedures are required.J-TEXT Cloud Database (JCDB) is a nuclear fusion experimental data storage and management system, aiming to satisfy the requirements of future long pulse experiment. Based on MongoDB server, Cassandra cluster servers and web applications servers, it will cost much time for JCDB to setup these servers from the beginning. To scale out and configure more conveniently and automatically, JCDB used Docker, a lightweight software containerization platform, to build cloud database and web applications, guaranteeing consistency of the all servers and reducing plenty of time. JCDB pushed two images to Docker repository and scaled out by pulled specified images. Even without any servers on hand, you can also deploy it on cloud services such as Amazon Web Services, Microsoft Azure.

        Speaker: Dr Wei Zheng
      • 1:40 PM

        Previous research has demonstrated that linear plasma accelerators can achieve the energy, temperature and densities necessary to reproduce the conditions of plasma impact in a divertor plate [1]. Some initial experiments have been performed with plasma foci at University of Mexico [2]. At the Instituto Politecnico Nacional a new device, a flexible Plasma Gun that can be operated in snowplow and in deflagration mode is under development. The first step is to tailor the design in such a way that the parameters of the ejected plasma fall within the desired range. It is therefore of great importance to develop a computational model that can predict plasma parameters as a function of input parameters such as the electrical input circuit characteristics, the electrodes geometry, the operating gas pressure, the gas feed, the distance to the exposed sample, among others. Using a snowplow model such as the one developed originally by Hart as a starting point [3-4], a 2D-axisymmetric model for a coaxial plasma gun is constructed. This model includes circuit effects and calculates expansion of the plasma plume and its density and temperature, so reasonable mechanisms of the interaction of the sample with this plasma can be predicted and later be contrasted against experimental measurements once the physical device is constructed.

        [1] J. Rapp, T.M. Biewer, J. Canik, J.B.O. Caughman, R.H. Goulding, D.L. Hillis, J.D. Lore, L.W. Owen, “The development of plasma-material interaction facilities for the future of fusion technology,” Fusion Science and Technology, vol. 64, Issue 2, p. 237-244, 2013.

        [2] G. Ramos, M. Martinez, J.J.E. Herrera, F. Castillo, “The plasma focus as a tool for plasma-wall-interaction studies”, Journal of Physics: Conference Series, vol. 591, art. 012031, 2015.

        [3] P. J. Hart, “Plasma Acceleration with Coaxial Electrodes,” Phys. Fluids, vol. 5, Issue 1, p. 38, 1962.

        [4] P. J. Hart, “Modified Snowplow Model for Coaxial Plasma Accelerators,” J. Appl. Phys., vol. 35, no. 12, p. 3425, 1964.

        Speaker: Mr Christian Gómez-Samaniego (CICATA Queretaro-IPN)
      • 1:40 PM
        Concept Design of GDT-Based Fusion Neutron Source for Improving the Q with High Field Neutral Beam Injection 2h

        Gas Dynamic Trap (GDT) is very attractive as a kind of fusion neutron source for testing fusion material and component as well as driving transmutation reactor due to its linear and compact structure, easiness of construction and maintenance, relatively low cost and tritium consumption. These years, the conceptual designs of GDT-based neuron source for above two purposes, named FDS-GDT, have been proposed as candidate of fusion neutron source by Institute of Nuclear Energy Safety Technology (CAS) • FDS Team in China, which focus on fusion safety and fusion nuclear science and technology research. However, the fusion energy gains (Q) in current international designs are still far lower than one, even about 0.05.
        In order to improve the Q and reduce the technologies requirement of magnet and neutral beam injection (NBI) for GDT-based fusion neutron source, a new method was proposed with high field neutral beam injection (HFNBI) for substituting the conventional method that the neutral beams are obliquely injected at middle plane of GDT where the field is minimal. This method will benefit for confining higher density of fast ions at turning point in the zone with higher magnetic field, as well as getting higher mirror ratio by reducing mid-plane field rather than increasing the mirror field. In this situation, the collision scattering loss of fast ions with higher density will be critical and change its confinement performance, power balance and particles balance.
        Two optimal designs of GDT-based fusion neutron source was proposed with HFNBI by using updated calculation model and based on SYSCODE. One is for improving Q to 0.5, about 10 folds comparing to conventional design scheme, and the fusion power is 18MW. The other is for reducing the NBI power and mirror field to enabling level, such as 10MW and 10T respectively, and the fusion power is 2MW and Q is 0.2.

        Speaker: Dr Dehong Chen (Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences)
      • 1:40 PM
        Conceptual Design of a Bidirectional Hybrid DC Circuit Breaker for Quench Protection of CFETR 2h

        The conceptual design of quench protection circuit for CFETR (China Fusion Engineering Testing Reactor) rated for current up to 70 kA and voltage of 15 kV is presented. The proposed scheme is based on a mechanical switch paralleled to the controlled static breaker. Static breaker is composed of a IGBTs unit and four diode units in a rectifier bridge allows it to be used in both current directions. The feasibility of bidirectional static breaker such as the reliable turn-on IGBTs unit, voltage and current sharing of each IGBTs and effect of rectifier diode recovery characteristic are also investigated for conceptual design of quench protection. The voltage and current unbalance are discussed in detail by simulation analysis which includes the influence of the gate signal delay and stray inductance in each IGBTs branch. Finally, a discussion of the conceptual design of quench protection circuit is given.

        Speaker: Mr Shusheng Wang
      • 1:40 PM
        Conceptual design of the cryogenic system for CFETR 2h

        China is planning to construct the China Fusion Engineering Test Reactor (CFETR), in order to bridge the gap between ITER and the Prototype Fusion Power Plant. The cryogenic system of CFETR is indispensable to provide cooling for the superconducting magnet system and cold structures at 4.2 K, the cryopumps at 3.7 K, the HTS current lead at 50 K, and the thermal shields at 80 K. In this paper, the magnet system of CFETR is introduced and the cold mass is estimated. The cryogenic system heat load is extrapolated from that of ITER. The required refrigeration capacity of the cryogenic system is evaluated, taking into account the dimension of CFETR and especially the large burn duty cycle of 50%. The conceptual design of the CFETR cryogenic system is introduced.

        Speaker: Mr Xiaogang Liu (Institute of Plasma Physics Chinese Academy of Sciences)
      • 1:40 PM
        Conceptual design of the torus cryopump for CFETR 2h

        A torus exhaust pumping system comprising 6 identical cryosorption pumps for China Fusion Engineering Test Reactor (CFETR) is designed to provide specified pressure levels and throughputs for various plasma operation modes. A conceptual structural design is performed and recommendations are presented in accordance with the design requirements. Based on the analysis of conductance, the effective pumping speed is calculated and the influencing factors are discussed. The thermal loads of cryopanels and shields are then calculated to verify the feasibility of the concept and determine the consumption of coolants at steady-state operation. The temperature distribution of the cryopanels is displayed by a 3D thermal-hydraulic calculation and regarded as one of the boundary conditions to obtain the stress distribution of the cryopanels. Simulation results indicate that the temperature distribution and the maximum stress meet the design requirements proposed by the cryogenic properties of the gases pumped. Finally, the configuration and operation scheme of the torus cryopumps are given in accordance with the overall pumping characteristics. The conceptual design of the torus cryopump provides methods and experience for the overall design of CFETR in the future.

        Speaker: Dr Chen Chen
      • 1:40 PM
        Control and protection system for the W7-X ECRH plant – experience from the first and plans for the next campaign 2h

        W7-X is a steady state capable optimized stellarator. The main heating system is electron cyclotron resonance heating (ECRH) operating at 140GHz providing up to 9MW microwave power.
        A set of diagnostics has been developed to protect the machine from non absorbed ECRH power which can easily damage in vessel components.
        The power is launched into the machine by front steerable quasi-optical launchers in X- or O-mode. While in X-mode the first pass absorption is ~99%, it is only 40... 70% in O-mode. The non absorbed power hitting the inner wall is measured by waveguides embedded in the first wall (ECA diagnostic).
        In order to prevent the inner wall from overheating or arcing, a near-infra red sensitive video diagnostic with a dynamic range of 450...1200°C was integrated in the ECRH launchers. Thermal calculations for the carbon tiles predict a temperature increase above the detection threshold for scenarios of plasma start-up failure or poor absorption on a time scale of ~100ms and the risk of overheating after ~300ms. However, no temperature rise above the detection threshold could be observed in experiments with failed break down, i.e. poor ECRH absorption for up to 100ms.
        The stray radiation level inside the machine is measured by so called sniffer probes which were designed to collect all radiation approaching the probing surface independent of incident angle and polarization. Five sniffer probes are installed at different toroidal positions. They were absolutely calibrated.
        The sniffer probes are integrated in the ECRH interlock system. During the first operational phase of W7-X this was the only available plasma interlock system. The signal quality proofed to be high enough for a reliable termination in case of poor absorption. After a breakdown phase of ~10ms, the sniffer probe signals dropped by more than an order of magnitude. However, especially in the very first days of operation, most discharges died by a radiative collapse due to impurity influx. In this case the heating power was reliably switched off due to the increased level of stray radiation.
        During OP1.1 the gyrotrons which are mostly capable of delivering >900kW power were operated at reduced power to increase the reliability. At maximum power there is an increased risk of losing the main mode in the gyrotron and thus a stop of RF emission. This causes an interlock for the gyrotron. An intelligent control system being able to operate stabile at maximum output power is currently being developed. One approach is to bring the gyrotron back in operation after a mode loss by switching off the HV supply for a short time (<1ms) and then back on at a slightly lower value. The second approach is a feedback control system to stabilize the output power by identifying a pre cursor signal for a mode loss and minimizing it. The mode activity at parasitic frequencies has been identified a possible pre cursor for a feedback control system.

        Speaker: Stefan Marsen (Max-Planck-Institut für Plasmaphysik)
      • 1:40 PM
        Current profile measured by the motional Stark effect polarimeter in the HL-2A tokamak 2h

        The safety factor and current density profiles play a very important role in understanding magnetohydrodynamics and micro-instability. Motional Stark effect (MSE) is one of the most powerful tools to measure the current density. A 4-channel MSE polarimeter based on dual photo-elastic modulators (PEMs) has been developed in the HL-2A tokamak. For each channel, 6 1-millimeter silicon fibers are applied. And off-the-shelf avalanche photodiode detectors with frequency band of 250 kHz are adapted due to its quantum efficiency up to ~83% at 660 nm. The beam emission spectra are filtered by a monochrometer; and the filter is controlled by an absolutely calibrated rotator, which can change the tilting angle of the filter with velocity of 720 degree/s, corresponding to the wavelength change of 288 nm/s with the filter. The rapid angle change of the monochrometer enables the wavelength to be swept during the discharge. The accuracy of the MSE can be up to ±0.15° in the calibration experiments.

        On HL-2A, the motional Stark effect is rather weak [1]. During the pilot experiment, the pitch angles of magnetic field are obtained for 3 spatial channels covering 10 cm along the major radius with time resolution of 5 ms. The profiles of current density and safety factor are obtained by the Current Profile Fitting (CPF) code, as shown in Figure 1. The q profile is monotonic, and the minimum q value is around 0.7. And the position of the q=1 surface consists with the sawtooth inversion radius measured by ECE.

        [1] D. L. Yu et al., Rev. Sci. Instrum., 85, 053508 (2014).

        Speaker: Mr Wenjin Chen
      • 1:40 PM
        Current Status and Progress on the Shield Blanket Design of CFETR Reactor 2h

        The main function of the CFETR shield blanket (SB) system is to provide the neutron shielding capability for the in-vessel components and the external environment. The SB concept design has been carried out during the 2009-2011 campaign. The current design on the SB system is concentrated on the neutronics analysis, the SB modules design, and the mechanics analysis using finite element method (FEM). The neutronics analysis on the SB thickness estimation in radial direction and the neutron shielding performance on the shield materials is carried out. The two kinds of SB module structure are taken into account respectively, which are mainly for exploring the heat removal capability comparing for the SB cooling channel system. In addition, the mechanics analysis is made on the structure static stress and the electronmagnetic (EM) force considering the core plasma disruption. This paper summarizes the SB components design activities and the progress at present status.

        Speaker: Dr Liu Changle (Institute of Plasma Physics, CAS)
      • 1:40 PM
        Design a Suitable Test Scheme for Triggering Bypass Protection Test of ITER PF Converter Unit 2h

        The external bypass, as an important components of the international thermonuclear experimental reactor (ITER) poloidal field converter unit (PFCU), will provide a freewheeling loop for the load current to protect the magnets and PF converter modules from being damaged by over-current and over-voltage under fault conditions. The triggering bypass protection test is used to verify that the designed bypass can be triggered normally and endure the rated load current. In this paper, a suitable test scheme is designed for triggering bypass protection test of ITER PFCU based on the PF converter integrated test platform in ASIPP. This test scheme includes the triggering bypass method, calculating the trigger angle of the converter in inverter and a method of reducing the bypass current to zero in the absence of ITER mechanical switch. The feasibility of the test scheme is successfully verified by the simulation results and test results, both of which prove that the external bypass of ITER PFCU can be triggered normally and withstands the load current for 100ms. Due to the integrated inductance of the actual DC output circuit, the actual transfer time of bypass current in the test is more than the simulation results.

        Speaker: Dr Xiuqing Zhang (Institute of Plasma Physics Chinese Academy of Sciences)
      • 1:40 PM
        Design and Analysis of the CFETR TF Coils with REBCO tapes 2h

        Yong Ren*, Xiaogang Liu, Zhaoliang Wang, Junjun Li, Shijun Du, Guoqiang Li, Xiang Gao

        Institute of Plasma Physics, Hefei Institutes of Physical Science, Chinese Academy of Sciences, PO Box 1126, Hefei, Anhui, 230031, People's Republic of China
        *E-mail: renyong@mail.ustc.edu.cn; renyong@ipp.ac.cn.

        Abstract—The China Fusion Engineering Test Reactor (CFETR), which has the potential to produce a fusion power above 2 GW, is being designed to bridge the gap between the ITER and Demo in China. The fusion power is strongly dependent on the plasma pressure to magnetic pressure ratio and toroidal magnetic field at the plasma major radius. The maximum magnetic field above 20 T is expected in the CFETR TF coil. The state-of-the-art low temperature superconducting (LTS) magnet has a relatively low coil current density in magnetic field above 15 T. Therefore, a high performance superconducting magnet with high temperature superconductor (HTS) is required to provide the superior superconducting property. The RE–Ba–Cu–O-coated (REBCO) tapes, which has the higher critical current density than LTS in magnetic fields above 15 T and endure a high tensile stress, will be used for the CFETR TF coil. The CFETR TF coil is composed of the REBCO HTS coils and Nb3Sn LTS coil.
        This paper will describe the design of the CFETR TF coil. The electromagnetic and thermal-hydraulic analysis of the CFETR TF coil is performed.

        Keywords—CFETR, Double-pancake (DP), REBCO coil, Superconducting magnet, Thermal-hydraulic behavior.

        Speaker: Dr Yong Ren
      • 1:40 PM
        Design and Analysis of the High Power DC Water-Cooled Busbar Connecting Type 2h

        High power aluminum water-cooling DC busbar is used to connect the converter and the reactor for the ITER PF converter. The water temperature difference between import and export is less than 20℃, and the temperature rise of the busbar does not exceed 70℃ according to the ITER standards. Three DC busbar connection types is designed to resolve soft connection overheat under the rated current. This paper mainly presents the temperature rise of DC busbar connection in different water flow and different steady state current. A temperature simulation of directly weld connection, lamination weld connection and soft connection are investigated by finite element method (FEM). The test is carried out to verify the correctness of simulation. The comparison results of three type’s shows that directly weld connection method has a better performance.

        Speaker: Mr Zhongma Wang (Institute of Plasma Physics Chinese Academy of Sciences)
      • 1:40 PM
        Design of a local oscillator for the 2.45GHz/4MW LHCD system on EAST 2h

        The paper describes the design process and experimental validation of a local oscillator for the lower hybrid current system(LHCD) on EAST. The local oscillator is designed to provide original RF energy for the whole LHCD system, which plays an important role. The local oscillator must be of high spectral purity and stability. Only phase noise is better than -90dBc/10KHz can satisfy the requirements of the LHCD system. The local oscillator consists of the phase locking loop(PLL), the PIN switch, the regulable attenuation, the amplifier, the coupler and the power divider. Among these components, PLL determines the phase noise to a large extent. Usually the design of PLL is used with PLL chips integrated VCO. That makes the design simpler, but the phase noise is not better than that with PLL chips disintegrated VCO. Measures such as using fourth-order passive filter circuit and reducing the power supply ripple are adopted to optimize phase noise in addition. The test shows that the phase noise of the local oscillator is -94dBc/10KHz. The local oscillator must have the function that rapidly shut off in case an accident. The PIN switch plays the rule and it can shut off the oscillator in 1us. Due to requirements of the LHCD system, the local oscillator has three output ports, two ports must more than 30dBm, and another one must more than 20dBm. Before the production of the amplifier, coupler and the power divider circuit, the design must begin with theoretical calculation and regulated models built in the software ADS. In order to improve the stimulation accuracy, joint stimulation combined with the actual circuit is adopted in ADS. Test shows that the output power meet the requirements. Stimulations and experimental results fit well and the local oscillator has worked in LHCD system for months. That successfully indicate the reliability of the local oscillator.

        Speaker: Dr ZHU Liang (ASIPP)
      • 1:40 PM
        Design of Ground Plane of NSTX-U Ohmic Heating Coil 2h

        A ground plane is a conductive, grounded, electrostatic shield surrounding high voltage insulated conductors. The purpose is to contain the electric field developed by the high voltage conductors within the insulation and to provide a return path for capacitive currents during transients.
        A resistive paint was applied to the outer surfaces of the NSTX-U OH coil to form an outer ground plane. It serves to shield the surrounding instrumentation (flux loops, thermocouples, etc.) from electrical noise generated by the switching of the thyristor converters that supply power to the OH coil.
        This paper provides a detailed description of the equivalent circuit model of the multi-layered OH coil with the ground plane. An analysis of surface voltage of the OH coil is performed using the PSCAD, a profession transient simulation tool, to confirm that the ground plane performs as required under normal operation and fault conditions.This work is supported by US DOE Contract No. DE-AC02-09CH11466.

        Speaker: Weiguo Que (Princeton Plasma Physics Laboratory)
      • 1:40 PM
        Design of Inverter Module on RMP coil Power Supply in EAST 2h

        In Experimental Advanced Superconducting Tokamak (EAST), Resonant Magnetic Perturbation (RMP) coils, which are powered by RMP coil Power Supply (PS), are set to research Edge Localized Mode (ELM), Resistive Wall Mode (RWM) and Error Field Correction (EFC). The RMP coils are grouped in 8 sets with 2 coils in series as 1 set. And the 8 sets of coils are respectively powered by 8 sets of PSs.
        To generate a wide range of perturbation, the RMP coils will be supplied independently by DC or AC with the amplitude up to 2.5 kA in the frequency range of 50 Hz to 1 kHz. The maximum output voltage is 408 V. For further research of physicists, the output voltage and current should be up to 450 V and 4 kA. The current latency time should be less than 0.35ms (0-2.5 kA50%) while the voltage latency time should be less than 0.25ms (0-450 V50%). And the rise rate for voltage and current should be larger than 300 V/ms and 5 kA/ms. The voltage ripple should be less than 2% of output voltage, when the output voltage is less than 5% of the rated value, the voltage ripple should be less than 10%.
        To meet the design requirements of output current, output voltage, current and voltage latency time and output voltage ripple, the switch frequency of IGBT, equivalent output frequency, RC compensation branch and topology structure should be designed carefully. The specific design progress including theoretical analysis and simulation results are given. The test results about different design requirements validate the correctness and reasonability of the scheme.
        By applying carrier phase shifting PWM between branches, the ripple of output voltage can be reduced, and the equivalent switch freqyency has increased. And the whole 8 sets of RMP coil PSs have been put into use in the latest EAST experiment. Until now, they can have an effect on ELM control according to the physical experimenters.

        Speaker: Dr Deyong Song (Institute of Plasma Physics, Chinese Academy of Sciences)
      • 1:40 PM
        Designing a Power Module for Compressed Plasma 2h

        Abstract—Magnetic compression (MC) technology was suggested for tokamak to study compressed plasma in [Li, G., Scientific reports, 2015, 5] and a power module is designed here for developing its server power supplier. The minor-radius compression is one of the most effective method to improve the performance parameters of existing tokamaks, enabling the plasmas operated at high density, high temperature and high beta. In this paper, a high frequency and high power AC/Pulse converter is proposed, used for powering coils of minor-radius magnetic compression within vacuum chamber of the experimental advanced superconducting tokamak (EAST). The basic of the power module is a AC/Pulse converter of buck type, implemented by full-bridge phase-shift circuit and controlled by pulse-width-modulated (PWM). Also, control method adopts current closed loop Proportional-Integral (PI) control, has less than 1 ms current response time in real time. The converter is analyzed and the design procedure is discussed. Experimental results obtained from a 3kA converter prototype are presented to validate the converter’s performance with the re-designed control board.

        Keywords—magnetic compression; phase-shift PWM; AC/Pulse converters; tokamaks; fusion plasma; Lawson trinity parameter.

        1.Li, G. High-Gain High-Field Fusion Plasma. Sci. Rep. 5, 15790; doi: 10.1038/srep15790 (2015).

        Speaker: Dr Zhiyuan Weng (Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP))
      • 1:40 PM
        Development of primary vacuum windows for ITER diagnostics 2h

        Most of ITER’s diagnostics will be provided with viewing lines (optical, microwave, spectroscopic) for the monitoring of key characteristics of the plasma or for the achievement of physical measurements inside the vacuum vessel. The nature of the physical signal transmitted through the viewing lines requires the implementation of window assemblies incorporating non-metallic window. Placed at the vacuum boundary, the window assembly shall also ensure the vacuum integrity required for the plasma. Moreover, the diagnostic window assemblies form part of the ITER primary confinement boundary. Their integrity is consequently of prime importance in containing the reactant materials such as tritium in the inside of the vacuum vessel and, thus, directly related to the Nuclear Safety. The window assemblies are Protection Important Components (PICs) and their design, procurement and operation are considered as Protection Important Activities (PIAs). The primary confinement boundary shall be fully ensured during all the normal and accidental conditions. Window assemblies are part of the unpressurized area of the radioactivity confinement barrier, with safety related functions. These assemblies are made from components with non-metallic materials, particularly high grade ceramics, which are generally not covered by pressure vessel codes and for which there is no existing industrial standard that specifies the criteria for the design, manufacturing, and testing directly applicable to ITER. To incorporate non-metallic replaceable window assemblies in the confinement barrier of ITER, it shall be a requirement that the Operator (ITER Organization) has these window assemblies designed, procured, installed and operated based on procedures and records the French Nuclear Order known as “the Order 7th February 2012”, concerning basic nuclear installation design construction and operation quality.

        The paper will discuss the progress in the ongoing design and development of the primary vacuum window assemblies, covering several aspects, such as design, interfacing loads, integration in the port infrastructure and impact on the diagnostic performance. Also, maintenance and associated tools for window assemblies will be discussed.

        Speaker: Dr Victor Udintsev (ITER Organization)
      • 1:40 PM
        Divertor heat flux study of H-mode with NBI in EAST 2h

        H-mode with ELMs (Edge Localized modes) plasma regime is considered to be a preferable scenario in the future fusion devices as ITER. Heat loads on divertor during ELMs especially the type I ELMs is an important issue [1-2]. An infrared (IR)/visible endoscope system was built on the Experimental Advanced Superconducting Tokamak (EAST) in 2014. Based on the IR data in the experiment of 2014, the heat fluxes on the lower outer divertor were calculated with a code named DFLUX developed by ASIPP, aimed to provide reference for the H-mode operation of EAST [3]. Heat fluxes on lower outer divertor during ELMs without neutral beam injection (NBI) and with different NBI power were calculated and compared. The analyzed discharges were lower single null (LSN) divertor configuration discharges. In the case with lower-hybrid wave current drive (LHCD) only (Ip ~ 400kA, PLHCD ~2MW), ELM-free occurred after L-H transition accompanied by the increasing electron density (ne). The peak heat fluxes on lower outer divertor during ELM-free were not more than 1MW/m2. Then ELMs occurred and ne began to reduce and eventually lead to H-L transition. The peak heat flux on lower outer divertor during ELMs was not more than 2MW/m2 mostly. In the case LHCD combined with NBI (Ip ~ 300kA, PLHCD+PNBI ~2MW), type I ELMs occurred after L-H transition and ne was still increasing. The peak heat flux on lower outer divertor during ELMs was more than 3MW/m2 or even 5MW/m2 and ELMs disappeared when NBI was turned off mostly. By comparing the heat flux profile of divertor target versus radius in ELM, it may be because that the heat flux profile in ELM with NBI was narrow. The peak and averaged heat flux on lower outer divertor during ELMs did not increase with the increase of PNBI. The work of this paper will provide reference for H-mode discharge with NBI of EAST.

        Speaker: Mrs Bo Shi (Institute of Applied Physics of AOA)
      • 1:40 PM
        Effect of Rapid-forging and annealing on the properties of W-TaC alloys 2h

        Tungsten materials have been attracting growing interest as a promising candidate for plasma facing materials (PFMs) based on its high melting point, high temperature strength, and good thermal conductivity as well as its low erosion in fusion radiation environment and low tritium retention. But the inherent disadvantages of tungsten materials, including poor low temperature machinability, low ductility, high ductile-brittle transition temperature (DBTT) and irradiation-induced embrittlement, cannot be ignored for fusion reactor application. In recent studies, some process strategies focus on manufacturing nanostructured tungsten materials through thermo-plastic deformation treatment to resist the defect of tungsten.
        In this work, the mechanical properties and transient thermal shock performance of W-TaC alloys prepared by hot pressing(HP) followed by rapid-forging and annealing treatment were investigated. Tungsten powder and TaC powder were mixed through High-energy ball milling and then sintered by hot pressing with temperature of 1800℃ and pressure of 30MPa. Density、hardness 、Tensile strength and total elongation at different temperature of W-TaC alloys were tested respectively. The polished tungsten surfaces were exposed to repetitive ELM-like thermal shock loads at different base temperature and various absorbed energy densities. The thermal shock-induced damages and the microstructure were analysed by scanning electron microscope. The results indicate that strength and ductility can be improved by rapid-forging and subsequent annealing.

        Speaker: Dr Fan Feng
      • 1:40 PM
        Electronic Transport Properties of NbTi in Cooper Matrix Superconducting Wires for ITER Applications 2h

        The International Thermonuclear Experimental Reactor(ITER) device should demonstrate the scientific and technological possibility of commercial fusion energy production in large scale in order to solve the worldwide energy problem in the future. The superconducting magnet system is the key part of the ITER device to supply high magnetic fields for confining the deuterium–tritium plasma. The multifilament NbTi wires extruded in a Cu matrix with high quality have been studied to meet the specifications of superconducting strands. This work is presenting the study of signal-to-noise assessment, the electronic transport properties of NbTi wires extruded in a Cu matrix with 0.4mm in diameter and volume ratios of NbTi:Cu = 1.35:1. Normal-state magnetoresistance, I-V characteristics and superconducting state critical currents are thoroughly investigated.
        Additionally, the critical current density has been investigated as a function of temperature and field using the expressions for the critical temperature, critical magnetic field and pinning force in NbTi. The measurements undertaken in this research cover a range of the magnetic field between 0T to 7T at temperatures ranging from 1.9K to 10K.
        In order to measure the electrical resistance down to cryogenic temperature (2K) a Physical Property Measurement System (PPMS) is being used. The measurements have been done in various magnetic fields, up to 7T. The values of the measured resistance are the bases of the calculating the electrical resistivity, critical current density (Jc) and pinning force (Fp).
        Key words: superconducting wires, vortex matter, critical current density, pinning force

        Speaker: Mrs Alina Elena Niculescu (ICSI Rm. Valcea, Romania)
      • 1:40 PM
        EM Analysis of ITER Diagnostics Upper Port Plugs 14 (US port) and its in-Port components during Plasma Disruptions 2h

        ITER diagnostic port plugs perform many functions including nuclear shielding, structural support of diagnostic system, while allowing for diagnostic access to the plasma. With design advancing, the in-port diagnostic components are integrated into the port plug structure, and the diagnostic shield modules (DSM) are customized to accommodate various in-port diagnostic components. This paper summarizes results of transient electro-magnetic analysis using Opera 3d in support of recent design activities for ITER diagnostic upper port plug 14 (UPP14). A complete distribution of disruption loads on each component in UPP14 is presented. Impacts of different design features, such as the locations of the electrical contact, to the EM loads are discussed, and the solutions for improving the port structure are proposed.

        Speaker: Mr russell feder (pppl)
      • 1:40 PM

        Forschungszentrum Jülich and partners have been developing the ITER upper port plug diagnostic system (cCXRS) that is to transmit the visible light from the plasma to the end diagnostic via optical mirrors.
        Each port plug (PP) and its on-board components should withstand severe loads due to the plasma transients when the eddy currents and electromagnetic (EM) forces occur in the PP massive structures and its on-board components. The worst loading case can be found, and the forces be applied as a time-history loading to the PP mechanical model.
        In addition, huge transient eddy current and shock dynamic EM forces occur in the vacuum vessel (VV). In response to this loading the VV vibrates, thus exciting the PP and, consequently, its on-board components. ITER Organization has studied a wide range of plasma transients to provide the enveloped Floor Response Spectra (FRS) for different VV locations.
        The main computational problem is a reasonable superposition of a deterministically calculated time-history PP response to the applied EM forces with the PP response to the VV excitations that are specified as a FRS at the port stub (port attachment to VV) when only the maximum values of the structure response are calculated over a range of frequencies. On top of this, the EM loads should be combined with the seismic ones which are also specified as a FRS.
        This paper considers a potential methodology for combining plasma transients with seismic events. It is not yet intended for design purposes. The paper presents a step-by-step numerical modeling of the upper PP hosting some representative cCXRS component. Approaches to calculate the EM forces in the PP and its on-board components with the use of the dedicated global EM ITER model and to perform a subsequent structural dynamic analysis using a dedicated PP model are presented. The response spectrum analysis (RSA) of the PP and its on-board component that are excited via the VV due to the plasma transients and seismic events are then discussed. The challenge of combining closely spaced modes is highlighted. The approach to superimpose the time-history and RSA results is represented. A conservatism of the proposed approach, its requirements and merits are discussed. The technique proposed herewith is especially demanded when the dynamic behavior of the on-board component is a key feature of its design. This methodology gives a direct and transparent engineering way to design and estimate mechanical strength of the PP on-board components. The analysis uses reliable port stab FRS input and does not depend on spectra-to-spectra recalculation procedure (from port stab to component attachment) that is well established for the seismic-type response spectra but needs to be validated for the FRS due to plasma transients.
        This work was supported by Fusion for Energy (F4E) under the Framework Partnership Agreement F4E-FPA-408 (DG). The views and opinions expressed herein do not necessarily reflect those of F4E.

        Speaker: Mr Anatoly Panin (Forschungszentrum Juelich GmbH)
      • 1:40 PM
        Experimental Study on Natural Circulation Heat Transfer of Square Channel in Water Cooled Blanket 2h

        Square channel is widely used in the conceptual design of water cooled blanket of fusion reactor for cooling and providing appropriate inner temperature field for tritium breeding. Blanket is one of the most important safety components and thermal-hydraulic characteristics of blanket directly determine the heat transfer efficiency and safe operation of fusion reactor. Under accident conditions, the natural circulation phenomenon occurs without any mechanical devices intervention when the field forces acting on the fluid produce density gradients able to induce natural convection, which is the main heat transfer mechanism and a important measure to mitigate consequence of the reactor accident. For square channel(8mm*8mm) in blanket, the experimental study of natural circulation heat transfer has been conducted. Experimental results showed that natural circulation flow was not a independent parameter, which increased with the increase of the heat flux and the decrease of system pressure within the experiment scope. Simultaneously, natural circulation heat transfer was strongly affected by system pressure and heat flux. A new correlation was developed on the foundation of experimental data, which could predict the heat transfer coefficient of natural circulation with the maximum relative error of 30%. Comparing the experimental results with the results of forced circulation, it could be found that the heat transfer coefficient of natural circulation was lower than the heat transfer coefficient of forced circulation.

        Speaker: Mr hui bao (Institute of Plasma Physics, Chinese Academy of Sciences)
      • 1:40 PM
        Final Design and Fabrication of the TDU Scrapper for Wendelstein 7-X 2h

        A divertor “scraper” has been designed to protect weakly-cooled regions of the Wendelstein 7 X (W7-X) divertor targets from overheating under certain steady-state conditions [1]. The scraper is a limiter-like component with a plasma-facing profile geometry that was numerically designed to shadow those vulnerable target regions under such conditions. The “TDU” scraper is an inertially cooled component designed to test the protective function of the scraper, and its impact on particle pumping, under pulsed operating conditions during the OP1.2 campaign in 2017-18. The TDU scraper design matches the steady-state profile geometry within 0.20 mm, but was simplified to reduce cost while satisfying requirements for pulsed W7-X operation with 80 MJ total plasma energy input and pulse intervals up to 20 minutes, constrained by the very limited space available near the divertor. The scraper units are instrumented with in-situ Langmuir probes, thermocouples, and a pressure manometer to support the test program. These test divertor unit scraper elements "TDU-SE" components will be located in two of the ten half modules during W7-X OP1.2 operation. This paper describes the design considerations and fabrication of these scraper elements for Wendelstein 7-X.

        Speaker: G. Douglas Loesser (Princeton Plasma Physics Laboratory)
      • 1:40 PM
        Implementation of an Excitation Controller for an Impulse Motor-Generator 2h

        An energy-stored impulse motor-generator (MG) is used in power supply system of HL-2A tokamak to produce short high-voltage or high-current surges of desired parameters that are usually used for magnetic field coils and auxiliary heating equipment loads. The operation changing of these loads will cause disturbances in generator’s terminal voltage and the remark drop in MG’s rotating speed. This paper describes the implementation of an excitation controller using LabVIEW and CompactRIO for the 125MVA impulse MG in power supply system of HL-2A/2M. Both a staged control strategy and digital PID algorithm built in LabVIEW are applied to the excitation controller that runs preciously in a one millisecond cycle to achieve voltage feedback control. The proposed excitation controller, composed mainly of host computer and CompactRIO embedded reconfigurable system, is available to restore and stabilize terminal voltage in accordance with the desired voltage waveform set by operators when pulsed loads and motor speed change quickly, also implements real-time monitoring of the working condition and some electrical parameters of excitation system and communicates with central control system via Ethernet to either download discharge control files at time interval between two impulse discharges or upload waveform data generated in control process after a pulsed discharge. Engineering experiment results show that the use of excitation controller improves the voltage stiffness of the power supply system and provides effective control of generator’s terminal voltage under the management of central control system.

        Speaker: Mr Chi Wang (Southwestern Institute of Physics)
      • 1:40 PM
        Improvement of the plasma current density profile by the polarimeter/interferometer system on the EAST tokamak 2h

        A novel method has been developed to improve the accuracy of plasma current density profile by combining the polarimeter/interferometer (POINT) measurements with the external magnetic measurements on the EAST tokamak. The POINT system, measuring the accurate line-integrated electron density and Faraday rotation angle, provides the magnetic field information inside the plasma. By adding these data to the equilibrium confinement, the results from POINT measurements show a difference with the original equilibrium and the difference becomes larger from boundary to core of the plasma. This correction process makes up for the deficiency of magnetic probe measurements, the details of the correction process are specified, which bypass the equilibrium fit (EFIT) code. Results with and without these corrections are presented, comparisons of the corrected results and experimental results are also shown and they are found agree well with each other. The feasibility and reliability of the correction process are also discussed in this paper.

        Speaker: Xiang Zhu
      • 1:40 PM
        Improving accuracy of interceptive current measurement for use in IFMIF/EVEDA accelerator 2h

        ACCT/DCCTs are used as a non-interceptive means of beam current measurement in the IFMIF/EVEDA accelerator. Current measurement using ACCTs for long pulses with 100 ms or longer suffers such problems as drooping due to the fact that this measurement is based on the current induction through transformers. The electrical circuits connected to ACCT have been improved in order to reduce the drooping and to obtain a waveform as close as to the waveform of the original beam current.
        In this presentation we propose a method for measuring the beam currents derived from the waveform of ACCT output signals. Since the ACCT and associated electrical circuits can be considered as a linear system, there must be a unique transfer function connecting the input and the output of the ACCT, etc. This transfer function and the backward transfer function can be obtained numerically from simple experiments. The conversion from the output waveform to the input waveform is “ideal” since they are free from restrictions of real circuits.
        This method has several advantages: (1) no detail information about the ACCT and the electronics is necessary; (2) the transfer function is easy to obtain from simple experiments with a function generator and an oscilloscope other than the ACCT system; (3) effects of stray capacitance and inductance are inherently reflected in the transfer function; (4) the use of FFT speeds up the calculation for obtaining the transfer function. On the other hand, it does not allow a real-time beam monitoring since retrieving the accurate input waveform requires the whole waveform of the ACCT output.
        For verification of this method, we conducted a set of simple measurements using a function generator, an ACCT and an amplifier to determine the transfer function numerically. Applying this transfer function to the ACCT output, the waveform of the original beam current was retrieved. In order to reduce the noise in obtaining the transfer function, a set of waveforms were carefully chosen so that the FFT window is not an integer multiple of the input pulse length.
        For an FFT window of 3 seconds, five waveforms with pulse lengths between 0.7 – 1.3 seconds were used to determine the transfer function. The backward transfer function so obtained was applied to ACCT outputs with pulse lengths of 0.1 - 1.3 seconds. The retrieved waveforms show a very good agreement with the original square inputs without any drooping observed, showing the validity of this retrieval method.
        In the presentation, we will show the theory and the procedure of retrieval as well as several results for waveforms with different kinds of shapes and lengths. We will also discuss potential problems and limitations of this method. Approaches to making this close to real-time current monitoring and implementation for use in the accelerator will be also discussed.

        Speaker: Dr Yosuke HIRATA (National Institute for Quantum and Radiological Science Technology, IFMIF Accelerator Group)
      • 1:40 PM
        Insulation Systems for the ITER Central Solenoid Modules 2h

        General Atomics is currently fabricating seven (six plus one spare) modules for the ITER Central Solenoid in its Poway, CA facility. One of the critical steps in developing the fabrication process for the modules was to develop the insulation system providing the electrical isolation between turns and to grounded components. The insulation system for a module was required to withstand a maximum test voltage of 30kV, with a design goal of 150kV to ground for all penetrations and leads. The complex geometries of the coil, the helium penetrations, leads, and helium piping required the use of novel materials and approaches for the insulation.
        Materials were developed to improve the handling of large sheets of fiberglass and Kapton® polyimide of the ground insulation. On the vertical surfaces and around the corners of the CSM, sheets of lightly-bonded Glass-Kapton®-Glass (GKG) and Glass-Kapton® (GK) were used. These sheets were developed and tested for electrical strength and resin permeability in comparison to glass and Kapton® sheets.
        Around the leads and helium penetrations, electrical strength and tracking distance were obtained using built-up layers of a thermoplastic polyimide (TPI). Sheets of DuPont Mitsui AURUM® TPI were thermoformed and interlaced with other sheet materials to obtain the required tracking distance.
        The liquid helium supply and return pipes were insulated with 20 layers of Kapton® coated on one side with a B-stage epoxy resin that was dry to the touch. These Kapton® tapes were wrapped around the pipes (without interstitial layers of glass), with a ground mesh and a final outer layer of prepreg glass for durability, and then cured. The insulation was tested for electrical strength in air and in Paschen conditions.
        Instrumentation wires attached to the conductor exited the ground insulation along the helium pipes. The sealing of the “cusps” between the round wires and the round pipes was accomplished with high resin content prepreg glass. A methodology was developed to eliminate a problem of wire insulation cracking after curing the epoxy. These wire exits were tested in Paschen conditions.
        A ground screen was installed around the entire coil, its penetrations, and the helium pipes. The ground screen consisted of stainless steel mesh sheets spot-welded together with a single-point ground. The ground screen for the leads and helium penetrations was made by forming the screen pieces and connected to the pipe ground mesh with spot welds.
        A qualification coil was insulated and tested to 10kV prior to resin injection. A vacuum pressure impregnation of the resin was completed after which the coil passed a 30kV hipot. This qualification coil serves as a final test article to validate the design and fabrication of the insulation system.
        This paper describes the insulation system and the development and testing of the novel insulation materials used in the ITER CSM.
        Acknowledgement: This work was supported by UT-Battelle/Oak Ridge National Laboratory under sponsorship of the US Department of Energy Office of Science under Awards 4000103039 and DE-AC05-00OR22725.

        Speaker: Nikolai Norausky (General Atomics)
      • 1:40 PM
        Integral Benchmark Experiments on a Large Copper Block using GELINA accelerator to validate natural Cu neutron inelastic scattering cross sections from different neutron cross section databases. 2h

        A neutronics integral benchmark experiment on a pure Copper block (dimensions 60x60x60 cm3), aimed at testing and validating recent nuclear data libraries has been performed at GELINA. GELINA is a powerful photo-neutron source using 75 A, 110 MeV electron current impinging on a depleted 238U rotating target, producing a white spectrum with neutron energies ranging from epithermal region up to about 20 MeV with a mean energy of about 1.4 MeV and intensity up to 3.2E13 n/s. A large natCu block has been positioned at 100 cm from the target. The block had seven positions at different depths respect to the main neutron propagation direction where thin activation foils were used as neutron flux probes. Materials which are activated by different neutron energies were used and the measured fluxes were compared with calculations performed with MCNP5 neutron transport code employing different neutron cross sections database for comparison. With the MCNP5 it was modelled the neutron spectrum produced by GELINA accelerator and the neutron transport inside the block describing all the most relevant components of the experiment. This is the first time that a neutronics integral experiment on Copper is performed using such a white neutron spectrum and the results of the our comparison are used to validate the neutron inelastic scattering cross sections.

        Speaker: Dr Mario Pillon (ENEA)
      • 1:40 PM
        ITER PF6 Dummy Double Pancake Winding 2h

        The Poloidal Field (PF) coils are one of the main sub-systems of the 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 signed between ASIPP and Fusion for Energy (F4E). The ITER PF6 winding pack is composed by stacking of 9 double pancakes. Dummy double pancake fabrication will be implemented to qualify the critical fabrication processes, before series production.
        This paper describes the ITER PF6 dummy double pancake winding with a “two-in hand” configuration. Conductor de-spooling, straightening, ultrasonic cleaning, sandblasting, bending, turn insulation wrapping and deposition were carried out through the whole winding process. Joggles forming with high accuracy, helium inlets manufacturing with automatic welding and x-ray test, tail manufacturing with on-line assembly and PAUT test were performed. High synchronization of each unit in one line and between two lines were achieved. Tight tolerances of turn positioning and deviation between turns were obtained. The dummy double pancake winding has been completed with radial build-up deviation of 1mm, which met the requirement of less than 3mm.

        Speaker: Dr Wei Wen (ASIPP)
      • 1:40 PM
        Latest results from the Hybrid Illinois Device for Research and Applications (HIDRA) 2h

        The Hybrid Illinois Device for Research and Applications (HIDRA) is a toroidal fusion device at the University of Illinois. HIDRA is the former WEGA stellarator that was operated in Greifswald at the Max Planck Institut für Plasmaphysik. The machine is a five period, l = 2, m = 5 stellarator, with major radius $R_0$ = 0.72 m and minor radius r = 0.19 m. Initial heating is achieved with 2.45 GHz ECR heating at $B_0$ = 0.087 T magnetic field, which can go as high as $B_0$ = 0.5 T. HIDRA has the ability to operate as both a stellarator and a tokamak, initially operating in the stellarator mode. The focus of research on HIDRA will be to do dedicated PMI studies using the wealth of knowledge and experience at the Center for Plasma Material Interactions. In early 2016 the first experiments were started to be performed in HIDRA. This paper presents some of the initial results obtained from the machine.

        Speaker: Mr Rabel Rizkallah (University of Illinois)
      • 1:40 PM
        Lessons learned on design, manufacturing and commissioning of IRVIS endoscopes prototypes for W7-X divertor temperature monitoring 2h

        The Wendelstein 7-X fusion experiment at Max-Planck-Institut für Plasma Physik (IPP) in Greifswald produced its first hydrogen plasma on 3rd February 2016. This marks the start of scientific operation. Wendelstein 7-X is to investigate this configuration’s suitability for use in a power plant. In order to allow for an early integral test of the main systems needed for plasma operation (magnet system, vacuum, plasma heating, control and data acquisition, etc), one of the five divertor unit (module 5) and most of the carbon tiles covering the wall protection elements are being installed before the next experimental campaign (OP1.2). For the later operation phases, the heat fluxes coming from the plasma will be distributed over an area provided by the plasma facing components (i.e. divertor target plates, baffles). An important diagnostic for W7-X will be thermography systems monitoring the surface temperature of the divertor target plates by collecting and processing infrared (IR) and visible (VIS) light from the divertor region of the plasma. For this purpose the company Thales SESO has been assigned to design, build, test, deliver and install two prototypes of IRVIS (InfraRed-VISible) endoscope systems for the divertor of the W7-X Stellarator. Thermography is part of the operational and protective divertor diagnostics and has to detect signals indicating anomalous operation scenarios. The design of the horizontal and vertical target plates and the baffles in the divertor should keep the local power load below 10 MW/m2. The IRVIS endoscope systems are designed to operate under heavy-duty conditions.

        Speaker: Didier Chauvin (CEA)
      • 1:40 PM
        Li2Be2O3 pebbles prepared via sol-gel method: multifunctional blanket material designed to both tritium breeder and neutron multiplier 2h

        As a notable multifunctional material in breeding blankets, Li2Be2O3 could play a role on both tritium breeder (lithium) and neutron multiplier (beryllium), which is being considered for use in increasing the tritium breeding ratio. However, synthesis of Li2Be2O3 ceramic powder still needs to be further studied, and the preparation of Li2Be2O3 pebble-type breeder has not been reported in recent reports. In this paper, two-step method is carried out: firstly, Li2Be2O3 ceramic powder is synthesized by sol-gel method. Stoichiometric CH3COOLi and Be (NO3)2 are dissolved in distilled water and transfer to a gel by aqueous solution polymerization of acrylamide. Then the nano-sized ceramic powder is observed by sintering process at high temperature in the air. The sintering temperature is established by thermogravimetric/differential scanning calorimetry (TG–DSC). The Li/Be molar ratio of products are detected by inductively coupled plasma atomic emission spectroscopy (ICP-AES). Crystal structure of this powder is characterized by the X-ray diffraction spectroscopy (XRD). Secondly, ceramic injection molding methods in several conditions are used to prepare the Li2Be2O3 pebbles. The ceramic slurry is prepared by milling 85 wt% synthesized Li2Be2O3 powder and 15% polymerizable binder solution (acrylamide and N,N'-Methylenebisacrylamide). Then the slurry is doped to the liquid paraffin with the temperature at 80 ℃, and each drop is about 10μL. The liquid drops could be solidified during the sedimentation in the hot oil. The obtained gel balls could transfer into ceramic pebbles after the second sintering and the strength of the pebbles is also measured by mechanical tester. The results show that when the concentration of polymerizable binder solution is selected about 10 wt%, the resulting pebbles had a better spherical shape and a higher strength.

        Speaker: Dr Wei Lu (Institute of Applied Physics, Army Officer Academy)
      • 1:40 PM
        Lithium Evaporation System Design for the NSTX-Upgrade Fusion Device 2h

        The National Spherical Torus eXperiment (NSTX) has recently undergone a major upgrade to NSTX-U at Princeton Plasma Physics Laboratory (PPPL). To improve plasma performance for NSTX-U, control of the influx of impurity gases and fuel recycling from plasma facing components (PFCs) are critical. On the NSTX fusion device, two lithium oven evaporators (LITERs) were mounted on the upper dome of the vacuum vessel to apply a thin layer of lithium coating downward onto lower divertor, which has resulted in effective deuterium retention and improved energy confinement time. For NSTX-U, it is desirable to have lithium directly coat the upper divertor for double-null plasma operation, and increase the coverage of PFCs in general with lithium. In this paper, the design of a new lithium evaporation system capable of coating NSTX-U PFCs in all directions will be reported. Porous stainless steel (SS) tubes will be used to hold the lithium. Lithium will be loaded into the pores of the SS tube as a liquid at elevated temperature within an argon glovebox. A vacuum heater will be inserted into the lithium-loaded SS tube to heat lithium to more than 600 C for lithium evaporation in the NSTX-U vacuum vessel. With significantly less thermal mass compared to the LITERs, the new lithium evaporator can be heated up and cooled down much faster. The time needed to reach operating temperatures and unwanted lithium evaporation while at temperature will be drastically reduced. The new evaporation system setup and thermal modeling results of the temperature distribution achieved in the porous evaporator will be covered in detail. Safety design considerations and analysis regarding lithium handling during operation will also be included.

        Speaker: Dr Dang Cai (Princeton Plasma Physics Lab)
      • 1:40 PM
        Mechanical Designs for High Magnetic Field Tests for ITER Applications 2h

        In the ITER (International Thermonuclear Experimental Reactor) program, core imaging X-ray spectrometer (CIXS) and electron cyclotron emission (ECE) are two diagnostic systems in the US ITER project, where X-ray Dectric PILATUS single-photon-counting pixel detectors for X-ray energy and piezo actuators are required to operate under the conditions of high magnetic fields and the accompanying rapid field transient rates. The two devices were tested under the conditions in Princeton Plasma Physics Laboratory. For the tests, a transrex power supply was employed to provide high intensity of magnetic fields (up to 3T), and a series of mechanical devices were designed and made to carry and secure the testing devices in the magnetic field. In addition to structural integrity, material magnetism was a major concern and an analysis of the magnetic properties was carried out. In this paper the testing fixtures are described, mechanical setups, instrumentation, generation of the high magnetic field, safety aspects and procedure are also introduced. With the testing setups, the tests for CIXS and ECE components were successfully completed, and the results are provided to ITER applications.

        Speaker: Dr Luis Delgado-Aparicio (Princeton Plasma Physics Laboratory)
      • 1:40 PM
        Mechanism for Plasma Fusion with Major Ionic Species at Only Ten Million Kelvins 2h

        Ten-thousand-fold enhancements of $^3$He abundance in impulsive solar energetic particle (SEP) events indicate that $^3$He ions, before accelerated to high energies (~ 1 MeV/nucleon) in solar flares, should have been preferentially heated significantly ($T_{^3He}/T_{^4He} ~ 10 - 100$). The best mechanism developed so far for preferential heating of $^3$He is the first or second harmonic resonances of $^3$He with current-driven electrostatic $^4$He or H-cyclotron waves. This mechanism for minor ionic species to be strongly heated by major ionic species cyclotron waves may also play an important role in the heating of laboratory plasmas for nuclear fusion. It is well known that nuclear fusion is a high-energy reaction in which two extremely energized lighter atomic nuclei, after overcoming the sturdy Coulomb barrier, fuse tightly into a heavier one via the nuclear force. For the fusion to occur among nuclei in labs such as in a tokamak, the plasma must be heated to or above 100 million Kelvins (MK) with sufficient confinement time and sufficient plasma density. Efforts on experiments of plasma fusion conducted in the past decades indicate numerous mysteries and difficulties surrounding how to control heat bursts and confine plasmas of extreme temperatures. Based on our previous work for $^3$He-rich SEP events, we propose a new general mechanism for plasma fusion of minor ionic species extremely heated with major ionic species at only about 10 MK in order to reduce difficulties of fusion technology and engineering in the plasma confinement and control. We consider multi-ion plasmas composing of various major and minor ionic species with a current drive. As an electric current is driven through, a plasma can be ohmically heated by the current up to 10 MK, at which the resistivity in the plasma is too low for the current to be significantly dissipated further and the entire plasma saturates its temperature at this level in this first-stage of the heating process. When the current is continuously driven up to a critical point, e.g. thirty percent of the electron thermal current, electrostatic ion-cyclotron waves of the major ionic species are destabilized, which can have frequencies at around a multiple of the ion-cyclotron frequencies of the minor ionic species and thus can further heat the minor ionic species via particular harmonic resonances to 100 MK and higher, at which the nuclear fusion between the extremely heated minor ionic species and the relatively cold major ionic species can occur. In this second-stage of the heating process, only the minor ionic species are preferentially heated. This mechanism of plasma fusion, because temperatures of the major ionic species and electrons are only around 10 MK, can greatly reduce difficulties of technology and engineering in confinement and control of the fusing plasma.

        Speaker: Prof. Tianxi Zhang (Alabama A & M University)
      • 1:40 PM
        Modeling of advanced nuclear fuel cycles incorporating hybrid fission/fusion devices. 2h

        Two of the main cited flaws of nuclear fission power to label it as a non-sustainable energy source are linked to the nuclear fuel cycle: one is the fuel availability, and the other is the radioactive spent fuel legacy. The incorporation of fast neutron systems to the fuel cycle can help reduce this two problems, by breeding additional fissile material (thus extending the nuclear resource's lifetime) and by eliminating minor actinides present in the spent fuel that contribute to reducing its radiotoxicity by many orders of magnitude (thus greatly reducing the spent fuel legacy issue). Traditionally, the fission community has explored this alternative via fast breeder reactor designs and their incorporation into the fuel cycle, but the use of fusion-based fast neutron sources needs consideration as well. Jointly, UT Austin and IPN have developed a nuclear fuel cycle modeling platform, which can perform detailed neutronic calculations of both thermal fission and fast fission/fusion systems, and allows material exchange between them at the end of each burn cycle. A Compact Fusion Neutron Source (CFNS), a simplified spherical tokamak design developed at the University of Texas at Austin, generates the neutrons in the fission/fusion device. The CFNS has around it an annular space where zones that contains fresh fertile material can breed fissile material, while other zones may contain spent fuel material that can be “rejuvenated” (i.e. breed additional fissile material) and reduce its radiotoxicity by destroying the minor actinides with fast neutrons. Results from the use of this platform to analyze the self-sufficiency of Th/U and U/Pu fuel cycles with and without reprocessing stages, in particular with regard to neutron economy in the hybrid system, will be presented in this paper.

        Speaker: Prof. Martin Nieto-Perez (CICATA Queretaro - IPN)
      • 1:40 PM
        Modeling of Ohmic Disruptive Discharge in J-TEXT Using the Tokamak Simulation Code 2h

        A simulation of J-TEXT Ohmic disruptive discharge has been achieved by using of the Tokamak Simulation Code (TSC) model. The anomalous transport of disruption was adjusted according to heat transport across large magnetic island. The simulation result of the plasma current, electron temperature, loop voltage and the disruption are compared with experimental disruptive discharge data. According to simulation result the dynamic response of one test plasma-facing module were calculated, the numerical results show that the maximum stress of the test module is in safety range.

        Speaker: Dr Jinhong Yang (New Star Research Institute of Applied Technology)
      • 1:40 PM
        Multiple laser system for high resolution Thomson scattering diagnostics on the EAST tokamak 2h

        The high temporal and spatial resolution of laser Thomson scattering (TS) diagnostics is an important research subject of fusion plasma diagnostics. Currently, the temporal resolution of TS based on single laser in EAST is limited to 20 ms, which is too low to resolve the evolution of pedestal structures directly. A critical part of this diagnostic is the high-frequency laser source. A multiple laser system for high resolution TS diagnostics has been designed and installed on the EAST tokamak in ASIPP (Institute of Plasma Physics, Chinese Academy of Sciences). To achieve the specified parameters, a multilaser solution including four 10-50 Hz 5 J Nd:YAG laser systems with the fundamental wavelength of 1064 nm, at a distance of ~40 m from the tokamak, is utilized. The design of the laser beam transport path is presented, using multi-beam combiner technology to improve the time resolution(up to 8 microseconds) and real-time monitoring of laser power to improve density measurement accuracy of the system. The requirements for this system are very stringent with approximately ~7mm spatial resolution at the edge region. After several weeks trial running on the superconducting EAST tokamak, the system was proved to be capable of measuring plasma electron temperature and density with high resolution. The setup of multiple laser system is described in detail in this paper, as well as the analysis of the measurement capability. Finally, the experimental results are presented. The completion of this project will provide the basic tools for other fast physical processes study, such as L-H transition.

        Speaker: Dr Xiaofeng Han
      • 1:40 PM

        In four of the upper ports of ITER, Electron Cyclotron launchers will be installed for heating and plasma stabilization. The launchers are designed as stainless steel casks (so-called port plugs), accommodating microwave mirrors and waveguides with the capability to inject up to 24 MW total microwave power into the plasma.

        The inner volume of the port plugs represents a relatively open structure which is unavoidable since the propagation of the microwave beams shall not interfere with any structural components. Thus it is essential to fill all remaining volumes with shielding components to guarantee the compliance of the launchers with the neutronic design requirements.

        That is why the EC upper launchers will be armed with three particular shielding elements, of which the first one is installed into the upper area of the plasma-facing Blanket Shield Module, the second one in the front area of the launcher main structure and the third one in its rear part. All shielding components must be equipped with suitable internal cooling structures regarding volumetric heat dissipation, acceptable pressure drop and proper steel/water ratio for optimum shielding performance.

        This paper outlines the general design of the shielding components, including mechanical structure, cooling layout and integration, interfaces with the MW-components, manufacturing, installation and maintenance aspects. Also analyses to prove mechanical integrity, thermo-hydraulic behavior and shielding capability will be presented.

        This work was supported by the European Joint Undertaking for ITER and the Development of Fusion Energy (Fusion for Energy) under contract No. F4E-2010-GRT-161.

        Speaker: Peter Spaeh (KIT)
      • 1:40 PM
        Neutronics and thermomechanical analysis of a conceptual shielding blanket for CFETR 2h

        The conceptual design of Chinese Fusion Engineering Reactor (CFETR), which will be operated in two phases has been proposed to fill the gap between ITER and DEMO plant [1]. In order to maintain the performance of Vacuum vessel (VV) and external machine components especially the superconductors, a reasonable shielding blanket is needed between tritium breeding blanket (BB) and those superconductors. On account of different BBs, the shielding blanket may have different structures. So far, mainly three conceptual BBs have been designed for CFETR: Helium gas cooled T breeding blanket, water-cooled and LiPb liquid metal coolant blankets. In this paper, based on the water-cooled BB, a conceptual shielding blanket (SB) structure has been proposed. The SB module structural materials are investigated, and the detailed shielding scheme is researched with the simplified 3 D model by using the MCNP program. The cooling structure of the SB module is provided, and the thermomechanical behaviors are analyzed by finite element method. These neutronics and thermomechanical analysis results indicate that the conceptual SB module meets the requirement for shielding and in compliance with thermomechanical standards of structure materials.

        Speaker: Dr Jie Zhang
      • 1:40 PM
        New features of the W7-X Safety Control system for OP 1.2 2h

        After the successful first operation campaign of Wendelstein 7-X in 2016 the experiment is being upgraded for the next stage called OP 1.2.
        Due to some new or extended components like heating systems and diagnostics, the
        increased energy level in the machine, and some new safety requirements, the central safety system (cSS) has to be extended. In addition, a safety level based switchover is being implemented to support the engineer in charge and the cSS-operator in their assessment of enabling a predefined set of components in different operation modes of W7-X. Furthermore, the safety levels define the entrance conditions to the experiment areas. This safety level concept comprises only 5 different states, “off”, “stand by”, “experiment pause”, “experiment”,
        and “W7-X emergency stop”. For test and calibration purposes a certain set of components can be enabled in some special operational modes, like Laser or ICRH calibrations, boronization etc.

        Speaker: Dr Reinhard Vilbrandt (Max-Planck-Institute for Plasma Physics, D-17491 Greifswald, Germany)
      • 1:40 PM

        A successful conceptual design was completed to develop in-vessel control coils (a.k.a. Non-Axisymmetric Control Coils or NCC). The NCC coils are a series of saddle coils that are intended to satisfy a number of physics criterion including magnetic breaking, error field control, fast Resistive Wall Mode (RWM) control and ELM stabilization. Customized Mineral Insulated Cable (MIC) was selected for the conductor material. The MIC was made from oxygen-free copper conductor, high purity Magnesium Oxide powder insulation and 304 stainless steel sleeve cover. Sample MIC was purchased from an outside supplier and various tests conducted for design and performance verification including high voltage testing, which necessitated the development of special test terminations. A concept design was also developed for terminating the MIC ends using non-conductive and vacuum sealed technique. Two alternative designs were proposed for joints inside the vacuum vessel. The NCC Coils are designed to be mounted in-front of the primary passive plates and underneath the PFC tiles. The passive plates will be modified to accommodate the coils. A new PFC tiles design concept was developed using High-Z materials. New penetrations design on the vacuum vessel wall was developed to prepare one port per coil. A new power patch panel will be required to provide the ability for various combinations of connections between the NCC, the existing RWM Coils and the existing SPA power suppliers.

        Speaker: Mr Neway Atnafu (Princeton University)
      • 1:40 PM
        Numerical simulation of particle dynamics in the magnetic mirror 2h

        The motion of electron in the magnetic field from a pair of current-carrying circular coil was studied using the standard numerical method for solving the differential equations. The situations of two coils with the same and reverse directions current were studied respectively, and the results were compared.

        In both of the two magnetic fields, the trajectory of an electron (or other charged particle) is sensitive to the initial state. Both types of magnetic mirrors are not capable of effective magnetic confinement for various initial states of electrons, even if confined at the beginning, but in subsequent movements, the electrons may escape at any time (when the time is long enough after).

        【Key Words】magnetic mirror; MATLAB ; magnetic confinement

        Speaker: Mr Hui Wang (Army Officer Academy)
      • 1:40 PM
        Numerical Simulation Research on the New Design Scheme of the EAST Divertor using Multi-physics Coupling Method 2h

        The surface thermal flux density on the EAST divertor would reach 10MW/m2 due to improvement of heating capacity. The lower divertor as one of the main in-vessel components in EAST must be updated to fit the future experiments for achieving high performance and long pulse plasma, which aims to bridge the gaps between the high performance experimental device EAST and the Chinese Fusion Engineering Test Reactor (CFETR).
        Now, some new design schemes of the lower divertors have been proposed and studied. All the possible loads including thermal flux, electromagnetic force, eddy electromagnetic moment and so on will be considered by applying the multi-physics coupling simulation method, so the simulation results would be closer to actual condition. The design schemes which couldn’t satisfy the requirements will be optimized until success.
        The research efficiency could be improved and the research costs could be decreased greatly by the multi-physics coupling simulation method. All the research results will be provide valuable references to the design of the divertor or other in-vessel components on EAST and CFETR.

        Speaker: Dr Lei LI (Institute of Plasma Physics, Chinese Academy of Sciences)
      • 1:40 PM
        Numerical–experimental benchmarking of a probabilistic code for prediction of Voltage Holding in High Vacuum, for ITER N-NBI. 2h

        In the framework of the program for the construction of 1 MeV–16 MW negative neutral beam injector (NNBI) for ITER, an R&D activity on voltage holding in vacuum has been initiated since 2009, aimed at supporting the design, construction, and development of the NNBI accelerator. For this purpose the voltage holding prediction model (VHPM) previously developed [1] has been updated. In the VHPM the effect of the electric field anode and the electric field cathode on the probability of breakdown is evaluated, by two exponents: alpha and gamma. On the basis of the experimental results from different test stands and of detailed 3D numerical simulation of the corresponding electric field configurations, the predictions of the VHPM numerical code have been benchmarked. New exponents, alpha and gamma have been proposed to obtain a more precise location of the weak point of the system and a better prediction of the maximum withstanding dc voltage in high vacuum.

        [1] N. Pilan, P. Veltri and A. De Lorenzi, IEEE Transactions on Dielectrics and Electrical Insulation Vol. 18, No. 2; April 2011

        Speaker: Nicola Pilan (Consorzio RFX)
      • 1:40 PM
        Operation Analysis of Impulse Current Mode on ITER High Power DC Test Platform with SVC System 2h

        The ITER Poloidal Field (PF) converter is comprised of four converter bridges, DC reactors, DC switches, etc. and most are non-standard. To evaluate the performance of these equipments, an ITER high power DC test platform has been built to carry out the rated current test and impulse current test. The latter is significant to verify the fault suppression capability of product. The DC test platform can output the rated 400 kA impulse current. The principle design and structure of the DC test platform is introduced in this paper. In addition, the impulse current test procedure is also discussed. The transient large current in impulse current test can produce huge impact reactive power, which impacts the fundamental reactive power and power grid voltage drop. The effect is analyzed by theoretical calculation and simulation. In order to suppress the transient effects of reactive power, a Static Var Compensator (SVC) system is added, which consists of Thyristor Controlled Reactors (TCR) and Fixed Capacitors (FC) with rated compensation capacity 83.2 Mvar. An impulse experiment is implemented on test platform with SVC system. The results of theoretical calculation, simulation and experiment are compared, which demonstrate that the SVC system is effective in compensating impact reactive power and it also performs well on inhibiting the power grid voltage drop.

        Speaker: Mr Xudong Wang (Institute of Plasma Physics, Chinese Academy of Science)
      • 1:40 PM
        Optimized Shape of TF Coil 2h

        The effect of gravity is Insignificant on the Conductor tension of TF coil. For The Chinese Fusion Engineering Test Reactor (CFETR), the influence is about one ten thousandth. Therefore, the design of TF coil shape generally does not consider its gravity.
        The shape of the TF coil is determined by the constant-tension and without bending moment equation regardless of its gravity. By calculating the equation, the Princeton-D curve can be worked out in the case of giving the radial position of the TF centerline. Because the radius of curvature of Princeton-D curve varies continuously in space, it is difficult to manufacture. Therefore, the Princeton-D curve is generally fitted with three symmetrical arcs up and down.
        There are four variables in the fitting process, and the fitting result may be not optimal because of different fitting methods. In this paper, a criterion is given to determine whether the fitting result is optimal or not. Based on the criterion, it is possible to determine whether the shape of TF coil is optimal or not for each fusion device.

        Speaker: Wang Zhaoliang
      • 1:40 PM
        Particle model of the driver of ITER NBI system 2h

        The Neutral Beam Injection (NBI) heating system of ITER tokamak is a key stage for the yield of the full fusion machine. The NBI yield in turn strongly depends on the performance of the first component of the system, the negative ion source of D- ions.
        This negative source starts with a “driver region” where a RF discharge is induced in the deuterium gas and a plasma is created. The sources of this kind are therefore referred as Inductively Coupled Plasma – Radio Frequency (ICP-RF) sources. The formed plasma expands in a larger chamber and is then extracted.
        While several simulation tools have been developed for the expansion and extraction regions, a full simulation of the driver region is still lacking, mainly because of the difficulties created by the self-consistent inclusion in the codes of the inductive coupling of the RF frequency with the plasma.
        For this reason we developed a 2.5D Particle-In-Cell Monte-Carlo-Collision (PIC-MCC) model of a cylindrically symmetrical ICP-RF source, keeping the grid spacing and time step of the simulation small enough to respectively resolve the Debye length and the plasma frequency scales. We report the results of these simulations, which require a massive parallelization, and give some details about the computational side. This is a first step for the modelling of the full negative ion source.

        Speaker: Nicola Ippolito (INFN - Italy)
      • 1:40 PM
        Performance test of CICC joint for ITER correction coil 2h

        Abstract: In the frame of CICC testing for correction coils (CC) of ITER, the soldered joint design was developed and tested up to 12kA in a loop comprising the secondary winding of a superconducting transformer. The transformer which consists of two concentric layer-wound superconducting solenoids with the primary inside secondary coil was designed and manufactured. The primary coil was wound by 0.87mm diameter multifilamentary NbTi wires and secondary coil was wound by ITER CC conductor. The quench protection system was also introduced. The joint was test in liquid helium (LHe) temperature. The hall sensor was installed on the CC conductor to measure the current of secondary loop. Test results are present and showed that the joint resistance remained about 2nΩ in the current range from 8kA to 12kA, which was satisfied the requirement of ITER CC’s design.
        Index Terms: superconducting transformer, CICC, NbTi, joint resistance

        Speaker: Dr Yuanyuan Ma
      • 1:40 PM
        Physics and engineering progress of CFETR integration design 2h

        Chinese Fusion Engineering Test Reactor (CFETR) aims to bridge the gaps between the fusion experimental reactor (ITER) and the demonstration reactor (DEMO). CFETR will be designed and operated in two phases. Phase I focuses on a modest fusion power of up to 200 MW, where steady-state operation and self-sufficiency will be the two key issues. Phase II aims for DEMO validation with a fusion power over 1 GW. A new design has been made by choosing a larger machine with R = 6.6m,/a=1.8m, BT= 6-7T recently. Over 1GW fusion power can be achieved and technically it is easier to transfer from Phase I to Phase II with the new design. Physics and engineering progress of CFETR integration design are introduced in this paper.

        Speaker: Xiang Gao (Institute of Plasma Physics, Chinese Academy of Sciences)
      • 1:40 PM
        Polarizer designed for the electron cyclotron resonance heating system on J-TEXT 2h

        A polarization-controlled launcher has been designed for the 60 GHz electron cyclotron resonance heating system on J-TEXT. The polarizer is an essential component of the polarization-controlled launcher which is used to change the polarization of the electron cyclotron wave. The coordinate transformation method is applied for design of the polarizer with sinusoidal grooves. The polarizer was tested with the lower power test platform. The results agree well with the numerical results of the coordinate transformation method, indicating that the designed polarizer can meet the requirements of the electron cyclotron resonance heating system of the J-TEXT.
        Keywords: J-TEXT, ECRH, polarizer, C-method.

        Speaker: Dr C.H. Liu (First author)
      • 1:40 PM
        PPPL's Project Management Office: Work Planning System 2h

        Princeton Plasma Physics Laboratory (PPPL) is a U.S. national laboratory under contract by the Department of Energy to conduct plasma physics and nuclear fusion research. With an annual budget of approximately 100 million dollars, the laboratory supports projects in various divisions such as: experimental physics, engineering, operations, IT, and facilities and infrastructure. PPPL’s Project Management Office (PMO) is responsible for the oversight and implementation of all projects at the laboratory by using an online Work Planning System database.

        The Work Planning System implements project management and integrated safety management tools into a comprehensive checklist for projects. It captures the scope of the project, determines project risks, interfaces, support groups, and reviews all engineering work to combine these categories into a holistic approach for the life cycle of a project. Moreover, it guidelines risk management processes and focuses on project roles and approvals.

        The Work Planning System is important because it communicates project requirements and progress across a large board of people, consisting of engineers, supervisors, and directors. Monthly reviews are held for all active projects listed in the system to ensure accuracy and compliance, as well as provide oversight and feedback. This type of Work Breakdown Structure (WBS) is used by everybody at the laboratory; however, it is most important to the engineering staff working on large jobs that require multiple personnel and have a significant cost and time involvement.

        In this paper, we will further explore the key functions of the PPPL's Work Planning System, highlight its importance in the project’s life cycle, and discuss the development of a planned upgrade.

        Speaker: Ms Soha Aslam (PPPL)
      • 1:40 PM
        Preliminary Assessment of Tungsten as an Optional Plasma Facing Material in CFETR 2h

        Since tungsten is considered as the optional divertor target material for the future fusion device, e.g. CFETR, it is crucial to keep this high-Z impurity concentration under an acceptable level to avoid significant degradation of core performance. In this work, a parameter scan study is performed to preliminarily assess the tungsten impurity. The OEDGE (OSM-EIRENE-DIVIMP) code package is employed, where OSM-EIRENE provides 2D scrape-off layer (SOL) plasma background, and DIVIMP code then simulates the impurity distribution. Instead of specifying the upstream condition, the target plasma parameters are scanned by assuming the heat load of the tungsten divertor lower than the engineering heat flux limit (10MW/m2). A large range of plasma profiles are sampled by the scan of the edge plasma temperature and density decay lengths, which are assigned based on empirical equations. The results reveal both the temperature and density decay lengths have a noteworthy effect on tungsten sputtering flux, divertor tungsten retention and core concentration. The impact of the poloidal drift velocity, radial pinch velocity and cross-field diffusion coefficient on the tungsten transport is also studied.

        Speaker: Mr Guoliang Xu (School of Nuclear Science and Technology, University of Science and Technology of China)
      • 1:40 PM
        Preliminary Design for Diagnostic Port Integration at ITER Upper Port #18 2h

        ITER has many ports to install various diagnostics which view and measure various plasma parameters. One of the ports, the upper port #18 (UP18) is designed to integrate three tenant diagnostic systems: VUV (Vacuum Ultra Violet) spectrometer, NAS (Neutron Activation System), UVNC (Upper Vertical Neutron Camera). The key design drivers for the port integration are requirements on neutron shielding and maintenance. In this paper, we discuss the neutron shielding design made following the ALARA (as low as reasonable achievable) principle in order to reduce the shut-down dose rate in the interspace and port cell which are human-accessible areas. The design choice for radiation shielding of electronics in the port cell is also discussed. The port maintenance in ITER consists of RH (remote handling) operation for the port plug and manual (or assisted-manual) operation for the interspace and port cell areas. The compatibility with the ITER maintenance strategy is investigated for UP18 and the associated issues are addressed.

        Speaker: Sunil PAK (National Fusion Research Institute)
      • 1:40 PM
        Preliminary Probabilistic Safety Assessment of Tokamak-type Fusion Power Plants In Conceptual Design Stage 2h

        Owing to their inherent safety features and lacking of high level long-lived radioactive waste, fusion power plants have long been expected to be a safe, clean and ultimate solution for human energy crisis. Among all the fusion reactors, the tokamak-type fusion power plant (FPP) with deuterium and tritium fuels is considered to be one of the most promising fusion energy systems. However, there has been no probabilistic safety assessment for severe accidents of this type of FPP. Then how is the radiation risk of a tokamak-type fusion power plant in the viewpoint of severe accidents? This paper is such an effort to assess its radiation risk under the conditions of severe accidents in a risk way.

        Since there is no reactor core in fusion power plants, core damage frequency concept from fission nuclear power plants cannot be adopted as the risk metric for fusion power plants. But the frequency of large off-site release of radioactive material could be a possible effective risk metric, as there will also be specific radioactive material release in the accidents of a fusion power plant. According to the recommendations of international atomic energy agency, a large release of radioactive material can be specified in a way as a specified dose to the most exposed person off the site. Therefore, a large release concept could be defined for fusion power plants.

        On the basis of this large release concept, preliminary probabilistic safety assessment was applied to the safety design concept of a typical FPP based on the European fusion power plant conceptual study. This complex assessment work is finished with the assistance of reliability and probabilistic safety assessment program RiskA developed by FDS Team. Not only the total large release frequency of a fusion power plant was calculated, but also representative large release accident sequences initiated from specific accident types of a fusion power plant were identified. And their characteristics in happening frequencies, radioactive material release fractions, releasing time were analyzed and compared. Results showed that fusion power plants were not so safe as public’s imagination. There are still accident sequences which would arise significant radiation risk, although inherent safety features exist in the tokamak-type FPP.

        Speaker: Dr Shanqi Chen (Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences)
      • 1:40 PM
        Progress in the development of CFC/CuCrZr components for HL-2M divertor 2h

        HL-2M is a new medium-sized copper-conductor tokamak device under construction at Southwestern Institute of Physics and can perform advanced divertor configurations, such as snowflake and tripod. An open cassette divertor structure with active water cooling has been designed to meet the operation requirements of HL-2M tokamak. The divertor consists of a flat-tile CFC/CuCrZr component and a cassette structure. The CFC/CuCrZr component is made of a water-cooled CuCrZr copper alloy heat sink armored with CFC tiles CX-2002U. The CFC surface was modified by using slurry technique to improve its wettability to copper. An oxygen-free copper (OFC) buffer layer was cast on the modified CFC surface in order to mitigate the internal stresses caused by mismatch in the coefficient of thermal expansion of CFC and CuCrZr. Vacuum brazing of OFC/CFC tiles to CuCrZr heat sink was performed by using a silver free brazing alloy. Non-destructive examination followed by high-heat-flux testing was performed to access the manufacturing quality of the joint interfaces between the CFC tiles, OFC and the heat sink. The CFC/CuCrZr components experienced cyclic tests of 7-10 MW/m2 for 1000 cycles without visible damages. High-quality bonding between CFC and the heat sink was achieved to ensure the heat removal capability of the components.

        Speaker: Dr Youyun Lian (Southwestern Institute of Physics)
      • 1:40 PM
        Progress of Auxiliary Systems for Linear IFMIF Prototype Accelerator (IFMIF) 2h

        Progress of Auxiliary Systems for Linear IFMIF Prototype Accelerator (LIPAc)
        G.Pruneri1, P.Cara2, R.Heidinger2, P.Guy2, A. Kasugai3, J. Knaster1, K. Kondo3, M.Sugimoto1, K.Sakamoto3, and the LIPAc Integrated Project Team
        1IFMIF/EVEDA Project Team, Rokkasho, Japan, 2F4E, Garching, Germany, 3QST- National Institute for Quantum and Radiological Science and Technology, Rokkasho, Japan, 1LIPAc Integrated Project Team, Rokkasho, Japan

        The International Fusion Material Irradiation Facility (IFMIF) aims at qualifying and characterising materials capable to withstand the intense neutron flux originated in the D-T reactions of future fusion reactors thanks to a neutron flux with a broad peak at 14 MeV capable to provide >20 dpa/fpy on small specimens also qualified in this Engineering Validation Engineering Design Activity (EVEDA) phase. All its broad mandate has been successfully achieved, the only pending, is the validation of its Linea IFMIF Prototype Accelerator (LIPAc) with its Auxiliary Systems.
        The validation of LIPAc will be achieved in this on-going phase until December 2019 with the operation of a deuteron accelerator at 125 mA CW mode and 9 MeV, which is presently under installation and commissioning in Rokkasho (Japan). The successful operation of such a challenging plant, demands careful assessment of its auxiliary systems, that holding adequate redundancies will allow the target plant availability. The target availability of LIPAc was considered top priority even due to the inherent administrative difficulties of an “in-kind” project.
        LIPAc, the Linear IFMIF Prototype Accelerator presents a broad spectrum of ancillary equipment to optimize its operational beam time.
        A description of the Nuclear HVAC of IFMIF has already been reported [1].
        The present paper describes the auxiliary systems of LIPAc, (and their construction status) among which we address the Cryoplant System (Cryo), the Heating Ventilation & Air Conditioning (HVAC), Electrical Power Supply (EPS), the Service Water System (SWS), the Service Gas System (SGS), the Heat Rejection System (HRS) and the Fire Protection System (FPS).
        [1] G. Pruneri et al., Design principles of a nuclear and industrial HVAC of IFMIF, Fusion Engineering and Design 103 (2016) 81–84

        Corresponding author: giuseppe.pruneri@ifmif.org
        Topic: Project management, systems engineering
        Oral or poster preference: Poster

        Speaker: Mr Giuseppe Pruneri (IFMIF/EVEDA Project Team (RFX))
      • 1:40 PM
        RAMI analysis for PFCs of EAST divertor 2h

        Experimental advanced superconducting tokamak (EAST) is a D-shaped full superconducting tokamak device. Plasma facing components (PFCs) of divertor are most likely to be damaged during the operation, which become failure easily and have to be replaced and repaired frequently. To be able to reach the design objective, an assessment of technical risks by means of RAMI (Reliability, Availability, Maintainability and Inspectability) of the PFCs has to be performed on the graphite tiles as examples.
        A functional breakdown of the graphite tiles was performed in a bottom-up approach, which are described using the IDEFØ method. Reliability block diagrams (RBDs) were prepared to calculate the reliability and availability of each function under stipulated operating conditions. Failure Mode, Effects and Criticality Analysis (FMECA) of the graphite tiles was performed to evaluate potential causes of failures and their consequences. Criticality charts highlight the risks of the different failure modes with regard to the probability of their occurrence and impact on operations. It was assessed that the RAMI analysis results meet the EAST project requirement during this design phase and the result will be qualified further when the device is updated.

        Speaker: Yang Zhang
      • 1:40 PM
        Real-time Two-dimensional Optical Polarization Properties of the Fusion Reactor First Mirror Based on Active Polarized Beams 2h

        With the construction of the next generation of large size Tokamak marked by ITER and the realization of steady state plasma discharge, the first mirror(FM) will become the core of optical elements which will be used in the optical diagnosis system of future fusion reactor. Under the action of plasma,the FM surface morphology, structure and composition, and optical properties will be changed, which seriously affect the optical performance and service life of the FM. FM problem, directly related to the effectiveness of the optical diagnosis system, is an extremely important research topic of the future plasma physics diagnosis. At present, the FM research on optical properties mostly stays on the reflectivity, the polarization properties in nature are little studied. Besides, most researches on FM are performed on a point and off-line detection, and the use of polarized beams to carry out real-time on-line two-dimensional detection of the FM are less. In this manuscript, based on the polarization characteristics of a metal object, the experiment system of active polarized beam detecting FM optical polarization properties was established. The two-dimensional polarization information varieties of the same incident beam reflecting from different FMs, including the degree of polarization (DOP), polarization azimuth (PA) and circular polarization angle (CPA), were in real-time studied. It is found that the two-dimensional DOP and CPA can be used to detect different roughness degree Ra and regular degree of the measured surface, and provide significant basis for further quantitative detection of FM damage degree.

        Speaker: Mr Junli Qi (University of Science and Technology of China)
      • 1:40 PM
        Recent Development in Structural Design and Optimization of ITER Neutral Beam Manifold 2h

        The Neutral Beam (NB) manifold is a major sub-system of ITER fueling system with complex combination pipes with the aim at distributing the working gases for NB and Diagnostic NB injector. During the final design (FD) phase, NB manifold design has been completed based on configuration management model defined to use in FD. After the FD review, the design of NB manifold suffered several design changes so as to meet the different manifold routing requirements. Additional, structural integrity assessment during FD revealed that the NB manifold design has potential for more robust structure performance, as well as potential for a significant simplification of the support layout by redefining the constraint form of the support and the whole structural architecture. This paper describes the new design of NB manifold based on a more optimized support system. The former complex manifold supports and internal pipe supports have been compacted and replaced with an alternative scheme in order to more effective, decreasing about 90mm of structural deformation. Detailed analyses on internal pipe support layout are dedicated to confirm both the structural reliability and feasibility. Comparative analyses between two typical types of manifold support scheme, with emphasis on space feasibility, embedded plate location and etc., have been performed. All relevant results of thermo-mechanical analyses for different operation scenarios and fault conditions are presented as well as the mechanical behaviors and manufacturing aspects. Future optimization activities are described, which shall give useful information for a refined setting of components in the next phase.

        Speaker: Dr Chengzhi Cao (Southwestern Institute of Physics)
      • 1:40 PM
        Rectangular Magnetic Sensor Array for Current Measurement Based on Numerical Quadrature Method 2h

        Due to the feasibility of analytic solution, abundant researches have been conducted to study the measurement of line current by magnetic sensor array with circular arrangement. Since the large size bus is hard to be simplified as line current, transducer with circular arrangement is no longer suitable for large current measurement, especially in power system for fusion magnet coils, which has a rated current of tens of kiloampere and an impulse current up to hundreds of kiloampere when short circuit appears. Based on numerical quadrature algorithms, a magnetic sensor array based current transducer with rectangular arrangement is proposed in this paper. For the sensor array, the dimension of the rectangle depends on the magnetic field distribution, while the quantity and arrangement method is deduced by numerical quadrature theory which can improve the measurement accuracy. In addition, the effect of installation error is also discussed in this paper. A transducer prototype is developed to measure the current, which is 420 kA, through bundle busbars with a total section size of 2990 mm×650 mm. Compared with high accuracy fiber transducer, the consistency of experimental results demonstrates the high accuracy and reliability of the proposed transducer.
        Key words: Current measurement, magnetic sensor array, numerical quadrature theory, large size bus

        Speaker: Mr Qi Guo
      • 1:40 PM
        Research and Analysis on the Compatible Structures of the CFETR Divertor Based on the Remote Handling Requirements 2h

        Abstract: The Chinese Fusion Engineering Test Reactor (CFETR) is the next device in the roadmap for the realization of fusion energy in China, which aims to bridge the gaps between the fusion experimental reactor ITER and the demonstration reactor (DEMO). As a key in-vessel component in tokamak fusion reactor, the divertor which eliminates impurities, Helium ashes and heat flux often exposes to tritium environment and neutron radiation. Therefore, the divertor is very easily damaged, and it needs to be maintained regularly. Its maintenance should be handled by remote handling (RH) ways rather than by personnel directly, so the structures for remote handling ways should be suitable and feasible. It not only can meet the requirements for the RH but also should have adequate strength to resist high electromagnetic forces generated by Eddy current and Halo current due to the plasma disruptions. In this paper, the CFETR RH compatible structures are analyzed and two schemes are put forward based on the RH requirements. Then their working conditions, especially electromagnetic analysis are discussed to validate the design quality for the newly designed structures. After that, the maintenance processes of two schemes are simulated in the virtual environment by the software Delmia. Through simulation, the installation and dismantlement processes of the divertor can be vividly seen in Delmia where can check whether there will be interferences between the divertor and other components. What’s more, the distribution of other components and the location of RH ports can be determined by simulation. And the maintenance procedures of the divertor are planned rationally according to the simulation PERT chart. Finally, the optimal path for the divertor RH process is chosen through simulation.

        Speaker: Dr Huaichu Dai (1.Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, China; 2. University of Science and Technology of China, Hefei 230026, China)
      • 1:40 PM
        Research on Brazing technique of tungsten materials and reduced active ferritic–martensitic steels 2h

        An active helium cooled structure was adopted in the high performance divertors, which requires metallurgical bond of plasma facing materials (PFMs) and structural materials. The tungsten materials were one important part in the PFMs, and the structural materials were made of reduced active ferritic–martensitic steels (RAFMS). The tungsten materials were joined to RAFMS using Fe-based amorphous reduced active fillers by vacuum brazing. Joint interface was analyzed by several methods , such as optical microscope, SEM , EDS.The results showed that The weld zone forms a good metallurgical bond with no voids or cracks. And then the mechanical properties testing were performed preliminarily. The shear strength is 250MPa,which is much higher than other results.

        Speaker: Jianbao WANG
      • 1:40 PM
        Reverse Processing of CFETR Vacuum Vessel Mock-up 2h

        The 1/32 sector of CFETR vacuum vessel mock-up is welded together with four poloidal segments (PS), which are made by forming and welding. So each PS has certain machining allowance and contour error, lacks the processing datum in the process of machining welding groove. Considering each PS has complex surface profile, the method of reverse engineering is adopted to get the 3D model of each PS which has been made. On the basis of this 3D model, the model for CNC machining can be designed. 3D datum transformation is used for solving the problem of lacking the processing datum. An additional part with regular shape is used for converting the model for CNC machining into the machine tool coordinate system. Now, four poloidal segments have completed the welding groove processing. The machining deviation of welding groove is less than 0.5 mm, which has meet the requirements of assembly and welding.

        Speaker: Haibiao Ji (Institute of Plasma Physics, Chinese Academy of Sciences)
      • 1:40 PM

        Institute of Nuclear Energy Safety Technology (INEST), Chinese Academy of Sciences (CAS) is one of the leading teams undertaking its corresponding research and development and mainly responsible for structure material development and safety analysis. After the kick-off meeting for PD phase of Helium Cooled Ceramic Breeder (HCCB) Test Blanket System (TBS) held by CN DA in early 2016, safety analysis becomes more and more important.
        As an important part of the HCCB TBS safety assessment, accident analysis will be presented in this paper with the updated identification of reference accidents based on the approved version of preliminary safety report by IO, and more scenarios will be simulated and then analyzed using the thermal hydraulics code RELAP 5 and MELCOR. The results comparison of RELAP 5 and MELCOR will be done. The code-to-code comparisons can help identify code issues or implication errors that could go unrecognized. To understand the expected impact of modeling and data uncertainties, the uncertainty analysis approach of RELAP5 is Best Estimate Plus Uncertainty (BEPU) and will be extended for HCCB TBS to provide a direct understanding of the contribution of variations to specific parameters. The inventory of tritium will also be calculated under normal operation and its release under maintenance. To estimate consequences of airborne radioactive releases after accidents, the atmospheric radioactive transport and related potential for exposure will calculated by MACCCS combined with MELCOR. The primary objective of the above analysis is to evaluate the consequential radiological doses outside the ITER facility in scenarios selected to envelope all conceivable events, and thereby demonstrate compliance with the General Safety Objectives of the project.

        Speaker: Dr Jiangtao Jia (Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences)
      • 1:40 PM
        Safety impact of the Be-steam reaction during in-box LOCA on the WCCB blanket for CFETR 2h

        The Water Cooled Ceramic Breeder (WCCB) blanket is one of the blanket candidates for Chinese Fusion Engineering Test Reactor (CFETR). In case of in-box LOCA, the cooling channels inside the blanket module are broken, causing leakage and vaporization of the high temperature and high pressure water coolant into the Beryllium pebble beds which are used as neutron multiplier. Then the Be-steam reaction will take place and impact the safety of the blanket system, as well as the fusion reactor. Therefore the safety characteristics of the WCCB blanket during in-box LOCA should be investigated to prevent serious damage. In this paper, RELAP5 which is a system analysis code, is employed to model the blanket modules and the Primary Heat Transfer System (PHTS) of blankets. And the in-box LOCA is simulated by RELAP5 to figure out the transient response of the blanket system with different break areas. Then parameters, such as the steam flow rate at the break, pressure and temperature are transferred to ANSYS CFX for the simulation of the Be-steam reaction in the Be pebble beds. The results show that the impact of hydrogen production is limited. And the system can response in time to mitigate the consequences. Some improvement measures for the WCCB blanket system are recommended.

        Speaker: Ms Xiaoman Cheng (Institute of Plasma Physics Chinese Academy of Sciences)
      • 1:40 PM
        Sensitivity Studies of Tritium Transport to WCSB of CFETR 2h

        Primary concept design of CFETR (Chinese Fusion Engineering Test Reactor) has finished. In the fusion reactor, tritium is bred by lithium and then be extracted. However, tritium will contaminate the reactor structures and be leaked out to the environment eventually. For realizing the environmental friendly and tritium self-consistency, a mathematical-physical model is established to analyze tritium transport in CFETR’s WCSB (Water Cooled Solid Blanket). Some sensitivity studies of parameters on tritium losses and inventories were performed, the results show that the thickness of tungsten covering the first wall, the hydrogen concentration in coolant water and the water concentration in purge gas have significant impacts on tritium transport process.

        Speaker: Yang Xiang (University of Science and Technology of China)
      • 1:40 PM
        Servo-Power-Controller Design Based on EPICS 2h

        Abstract—Based on the Experimental Physics and Industrial Control System (EPICS), a servo power controller is designed to be inserted into the existing power supply for testing its function of measurement and control in real time. The power supply controller uses double closed-loop feedback control, in which the outer-loop feedback control uses dead time modulation (DTM) technology to adjust the output current by tracking the external control signal. In this paper, the principles of the power supply and its controller are introduced, and the DTM method is studied and verified using MATLAB simulation and experiments to develop control technology for a tokamak power supplier.

        Speaker: Yinchi Duan (Institute of Plasma physics, Chinese Academy of Sciences)
      • 1:40 PM
        Simulation study of large power handling in the divertor for CFETR phase II 2h

        The Chinese Fusion Engineering Testing Reactor (CFETR) is the next device for the Chinese magnetic confinement fusion (MCF) program that aims to bridge the gaps between the fusion experiment ITER and the demonstration reactor DEMO. CFETR will be operated in two phases: Steady-state operation and tritium self-sustainment will be the two key issues for the first phase with a modest fusion power of up to 200 MW. The second phase aims for DEMO validation with a fusion power over 1 GW. For meeting both Phase I and Phase II targets and easily transitioning from Phase I to Phase II with the same machine, new design has been made by choosing a large machine with R = 6.6m, a=1.8m, BT= 6-7T since 2015. So far, most physics design are centered around the aims of phase I [1].

        The ability to exhaust the plasma power loss is a critical issue to the successful production of a fusion power reactor. In fact, for phase I fusion power of ~200MW is less than ITER, it would not be a serious challenge with a ITER-like W/Cu divertor. However, during Phase II of CFETR, exhausted thermal heat from the core plasma (Psep) is expected to be larger than 200 MW, which is increased for the steady-state operation scenario since larger current drive power is injected into the core plasma. At the same time, the divertor heat load will extremely exceed the material tolerable limit (~10MW/m2), which could prevent the long pulse or steady state operation. Externally seeded impurities can help partially radiate the heat before it reaches the divertor. The seeded impurities however cannot be so large as to negatively impact the plasma performance in the core. We have simulated the baseline operation scenario parameters by using SOLPS5.0 (B2.5-EIRENE) code package for a standard lower single null (LSN) divertor configuration. The modeling shows that Ar (or Kr) puffing is highly effective in mitigation of the divertor peak heat flux. In addition, the radiation loss fraction inside the separatrix will enhances and leads a reduction of the power across the separatrix entering into scrape-off-layer region Psep as impurities puff rate increase. The edge effective charge Zeff and Psep which are experimentally proved to be closely related with core confinement factor H98 [2,3] are studied with Ar and Kr seeding as well. Comparison simulations of different divertor geometries will be performed in this work to optimize CFETR design.

        Further work about advanced divertor configuration (field expansion and increase in connection length) together with technical improvement of heat load removal capacity will be study in future.

        This work is supported by National Magnetic Confinement Fusion Science Program of China under contract no. 2014GB11003, 2015GB101003, 2015GB101000; National Nature Science Foundation of China under contract no. 11305206.

        [1] Y. X. Wan et al., 26th IAEA Fusion Energy Conference, Kyoto, Japan 17-22 Oct. 2016
        [2] J. Schweinzer et al., Nucl. Fusion 51, 113003 (2011)
        [3] A. Loarte et al., Phys. Plasmas 18,056105 (2011)

        Speaker: Dr Xiaoju Liu (Institute of Plasma Physics, Chinese Academy of Sciences)
      • 1:40 PM
        Spectroscopic diagnostics for negative ion source test facility at ASIPP 2h

        In order to support the development of the negative ion based neutral beam injection system for next generation fusion experimental reactor, a negative ion source test facility with radio frequency source is being built at Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP). A full set of spectroscopic diagnostics was designed to meet the requirement of operation performance optimization. This paper describes the design of optical emission spectroscopy of the plasma source and negative ion absorption spectroscopy. Ion formation and recombination processes are accompanied by radiation of different characteristic spectral lines, the Balmer series and Cs line at 852.1nm. By measuring the intensity of spectral lines, the key parameters, such as temperature and density of electron and cesium quantity, will be estimated. Absorption spectroscopy, with a cavity ring-down technique, will provide a direct measure of the negative ion density. These diagnostics will provide strong support for the coupling of RF power and improvement of negative ion density.

        Speaker: lizhen liang (ASIPP)
      • 1:40 PM
        Study on considerable defects introduced in tritium breeding material Li2TiO3 by annealing in vacuum 2h

        Li2TiO3 is one of the most promising candidates for solid breeder materials. However, defects introduced in Li2TiO3 will affect tritium release. In the present study, vacuum-annealing defects in Li2TiO3 were investigated by means of electron spin resonance (ESR). The defects of E-centers were found to be introduced by vacuum-annealing in Li2TiO3. The color of Li2TiO3 samples becomes dark grey after annealing in vacuum. This color change suggests the change from Ti4+ to Ti3+ due to decrease in the oxygen content. And the color was observed to recover to initial color, white again after annealing in air. The concentration of vacuum-annealing defects reaches almost a constant when the pressure is lower than 10 Pa. The defect concentration increases as annealing temperature goes up and then decreases when the temperature reaches to a certain value. The amount of vacuum-annealing defects goes down and the color of Li2TiO3 samples recovers to white gradually when vacuum-annealing samples annealed in air at different temperatures. There are no defects and color change while Li2TiO3 samples anneal in air first and then transfer to vacuum tube rapidly for vacuum-annealing. More defects were introduced in Li2TiO3 samples immersed in water for 6 hours. This elucidates that the defects produced by vacuum-annealing are attributed to the reduction of water adsorbed in Li2TiO3. Mass of Li2TiO3 was found to vary after the change of the atmosphere from nitrogen to air investigated by thermogravimetry. X-ray diffraction (XRD) results indicate that there are no modifications on Li2TiO3 crystal phases.
        The authors acknowledge School of Materials Science and Engineering in University of Science and Technology Beijing for providing the experimental samples. This work supported by the National Natural Science Foundation of China under contract No. 11605230.

        Speaker: Dr Qiang Qi (Institute of Plasma Physics, Chinese Academy of Sciences)
      • 1:40 PM
        Study on the effect of Pressure on the Electrical insulation of Plasma-sprayed Alumina Coating 2h

        In this paper, 316LN austenitic stainless steel coated with plasma-sprayed alumina for low-voltage electrical insulation is reported, which used in ITER Magnent Support to prevent eddy current loops. The coating will endure the large pressure and experience the several thermel shocks under practical working environment. SEM shows that coating is dense and good adhesions with base materials, and the surface roughness and texture are uniform. After 50 times thermal shocks, the coating is free from cracks, flakes and debonding. The thermal shock has a little influence on the electrical resistance properities. In addition, though the surface contact resistivity decreases with the increase of pressure, it still more than 109Ω.m when the pressure inceased to 250MPa, which can be concluded that the coating is good electrical insulation enough and can be utilized in the Magnent Support of ITER.

        Speaker: Mrs Rongrong Luo (Southwestern Institute of Physics)
      • 1:40 PM
        Study on the Welding Process of the Vacuum Vessel Mock-up for CFETR 2h

        Chinese Fusion Engineering Testing Reactor (CFETR) is a superconducting magnet Tokamak, The vacuum vessel (VV) will provide a high-vacuum environment for the plasma, improve radiation shielding and plasma stability and provide support for in-vessel components. The Research and Development (R&D) of the key technologies to the VV manufacture have been carried out a few years ago by Institute of Plasma Physics Chinese Academy of Science (ASIPP), including Narrow-Gap welding, cutting and non-destructive testing. The manufacture of the PS for first 1/32 sector VV mock-up has been completed in 2015, the 4 PS sectors have been welded into a whole in 2016. This paper will describe the study of the welding process of the 1/32 Vacuum Vessel Mock-up for CFETR.

        Speaker: Mr ZHIHONG LIU
      • 1:40 PM
        Suppression of tungsten impurity by lithium injection in tungsten divertor on EAST 2h

        EAST has upgraded the upper graphite divertor to ITER-like W/Cu monoblock structure[1] with active water cooling in order to facilitate the high power and long pulse plasmas[2]. Without wall conditioning tungsten impurity accumulation has been usually observed in plasmas, which is a crucial impediment to achieving high power, long-pulse H-modes. Therefore, some wall conditioning technologies need be explored to suppress the tungsten impurity, such as lithium (Li) aerosol injection[3] and Li coating[4]. In 2016, plasma discharges are performed in tungsten (W) upper divertor, and some exciting results are obtained with Li aerosol injection.
        The Li evaporation system in EAST has been upgraded with three new ovens located in the horizontal D, J, O port on EAST, separated toroidally by 120 deg.. The new ovens have three apertures for Li evaporation, for improving the Li coverage uniformity. In addition, there are two lithium powder dropper systems mounted in the J upper port: one located above the upper X-point, the other one located radially outboard between the X-point and outer midplane. The amount of injected lithium aerosol is controlled by a resonating piezoelectric disk.
        The uniform Li coating with the new ovens effectively suppressed W impurity influx coming from W divertor to avoid impurity accumulation in the plasma core. Overall the Li coating provided an excellent wall conditioning for high performance plasmas on the W divertor, facilitating a 62s long H-mode. The real-time Li aerosol injection suppressed tungsten accumulation; plasma stored energy and confinement increased both in L- and H-mode. In addition the strength of the tungsten source decreased with the Li injection rate. Also, the inner target ion saturation current and electron temperature decreased at the inner target. Even after termination of Li aerosol injection, the core W intensity remained at a low level.
        These results are encouraging as a possible mechanism to control tungsten impurities in future fusion devices.

        1. D. M. Yao et al., Fusion Eng. Des. 98-99 (2015) 1692
        2. B.N. Wan et al., Nucl. Fusion 55(2015)104015
        3. J.S. Hu et al., Phys.Rev. Letts. 114 (2015) 155001
        4. G.Z. Zuo et al., Plasma Phys. Contr. Fusion 54 (2012) 0115014

        Speaker: Wei Xu (Institute of Plasma Physics)
      • 1:40 PM
        Test Results of ITER 52-kA HTS Current Lead Prototypes 2h

        The Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) is responsible for the construction and performance testing of the 30 pairs of ITER high-temperature superconducting (HTS) current leads. A first pair of Poloidal Field (PF) coil type, 52-kA HTS current lead prototypes was built and cold tested in ASIPP in mid-2016. The test results approved their excellent performance on low joint resistance, long loss-of-flow accident time and high current-sharing temperature. The overheating time, mass flow, and heat loads to 5-K ends also meet the expectation. This paper summarizes the major test results for the PF 52-kA HTS current lead prototypes.

        Speaker: Chenglian Liu (Institute of Plasma Physics, CAS)
      • 1:40 PM
        Test upgrade for ITER HTS current leads series production 2h

        Institute of Plasma physics, Chinese Academy of Siciencs (ASIPP) is being in charge of the the manufacturing and cold testing of the 60 single HTS current lead series production for ITER device in 2017. The norminal current for these HTS current leads are from 10 kA to 70 kA. The cold testing of all 3 types of ITER lead prototypes were qualificated by ASIPP from 2015 to 2016. This paper will firstly summarize the testing items and testing result of ITER lead prototypes, then based on the prototypes experiences, the final test piping & instrument diagram for ITER lead series production is presented and discussed. A additional new test platform in 5K powering test facility for the series is discussed here to optimize the test arrangement. The 30 kV high voltage isolated instrument for the cryogenic temperature and voltage are firstly used to protect the CODAC test system. The detail test procedures and test items for the series will be discussed. At present, the manufacturing of the first pair of ITER lead will be started in the early of 2017. Author would like to present the test results based on the test upgrade work in the conference.

        Speaker: Dr Chenglian Liu (Institute of Plasma Physics, CAS)
      • 1:40 PM
        The Design of DRAGON-V Loop for Key Technique Verification of Liquid PBLI Blanket 2h

        The liquid Lead Lithium (PbLi) blanket is one of the most promising blanket concepts for fusion reactors. Aiming to better develop the PbLi blanket technology and realize engineering application, key issues of PbLi blanket should be investigated such as material corrosion, the magnetohydrodynamic (MHD) effect and so on. In addition, the integrated tests and engineering design validation of PbLi blanket module should be performed as well. So, it is necessary to develop experimental platforms to study the key issues of PbLi technology for fusion reactor.
        At present, a series of PbLi experimental loops have been designed and built successfully by FDS Team such as DRAGON-I/II and DRAGON-IV. Some experiments have been conducted to investigate the corrosion behaviors of CLAM steel in magnetic field, the purification technology of liquid PbLi and MHD pressure drop test. To support the engineering design validation of DEMO blanket with the parameters covering the requirements of ITER-TBM and China DEMO, a dual coolant thermal hydraulic integrated experimental Loop DRAGON-V was designed, which can be used to study the integrated experiments under the multi physical field conditions for fusion reactor. It is composed of lead-lithium loop and helium loop. The maximum temperature in the test section is designed to be 1100 °C, the maximum flow rate of PbLi can reach 40 kg/s, and the magnetic fields is up to 5 T. The maximum helium pressure is 10.5 MPa. It can carry out the research of material corrosion under different magnetic fields, MHD test for components of liquid blanket and LOCA. The obtained findings can support the development of the key techniques in-pile and the engineering design of China DEMO reactor. Besides, it can also be used for the advanced Generation-IV reactors and civil application.

        Speaker: Dr zhiqiang zhu
      • 1:40 PM
        The Design of Real-time Communication System Based on RFM and MRG-Realtime for EAST 2h

        For the purpose of keeping a steady-state plasma and avoiding plasma instabilities,EAST plasma control system (PCS) needs more diagnostic data produced by the EAST distributed subsystems to perform complex control algorithms. To solve the transmission problem of the above data, a real-time communication system based on MRG-Realtime (Messaging, RealTime, and Grid) operating system and reflective memory (RFM) network is designed. It is expected to realize that the data acquisition (DAQ) systems of the EAST subsystems can transfer the diagnostic data to PCS in real time through RFM network during capturing data, then PCS uses the available data to carry out the necessary calculations according to control algorithms and sends the commands to the subsystems in every control cycle.
        Then a small real-time communication system according to the prototype of the EAST vacuum DAQ system with sampling rate of 10 KHz was built for a performance test. In this test, the vacuum DAQ system transferred the acquired data to PCS in per control cycle of 100 us, which preliminarily meets the above mentioned requirements. The test was performed for 1000s to get a stable and reliable result, which described in detail in the paper. In addition, the future work is also discussed.

        Speaker: Ms Li Chunchun (Institute of Plasma Physics, Chinese Academy of Sciences)
      • 1:40 PM
        The Design on Pulse Distributor and Its On-line Status Diagnosis for ITER PF Power Supply 2h

        The International Thermonuclear Experimental Reactor (ITER) Poloidal Field (PF) power supply is consisted by thyristor-based phase-controlled converters to supply megawatt power for six PF coils. Each bridge arm of ITER PF converter is paralleled with 12 thyristors to withstand 27.5 kA rated current. For avoiding the electromagnetic interference, the trigger signals are transmitted by optical fiber from the pulse generator. The pulse distributor is designed in this paper to transfer the optical signal to high current electrical trigger pulses with less than 2 us rising time, thereby triggering the paralleled thyristors reliably and isolating 20 kV voltage from converter to controller. In addition, the pulse distributor on-line status diagnosis is also the key function. The signals of crucial on-line status will be encoded and transmitted by optical fiber to controller thus guaranteeing the safe operation of ITER PF converter system. The pulse distributor has been experimented in ITER PF ac/dc converter for two years and all functions have been effectively verified.

        Speaker: Dr Xiaojiao Chen (a. Institute of Plasma Physics, Chinese Academy of Science(ASIPP) b. University of Science and Technology of China)
      • 1:40 PM
        The Development of a monitoring system for poloidal Field Power Supply 2h

        Inspired by the ITER Control, Data Access and Communication (CODAC), Experimental Physics and Industrial Control System (EPICS) has been chosen for the control system. The monitoring system is a important subsystem of ITER poloidal Field(PF) Power Supply(PS) system which can real-time monitor and control PF power supply running , timely warning, abnormal positioning, support multi-user configuration operation and has a good cross-platform portability and scalability. All nodes operate in a synchronized manner by using Time Synchronization Network (TSN) which based on IEEE-1588 protocol, are connected via a dedicated timing network. And moreover, the monitoring system also acts the role of Plant System Host (PSH), which helps non-EPICS controllers to keep working in ITER PFPS control system. EtherCAT Fieldbus is used in the field layer, which improves the real-time performance and reliability of the system. The asynDriver is used to develop interfacing device specific code to hardware driver. The test shows that the monitoring system has good performance during experiments and convenient human-machine interface to satisfy the requirements of all the experiments. In this paper, a description is given of the prototype ITER PFPS remote monitoring system that has been implemented on the experiments.

        Speaker: Dr Lili Zhu
      • 1:40 PM
        The dynamic testing and analysis of copper and copper alloy in divertor working temperature range 2h

        Abstract: Until now, the most advanced mature plasma facing unit (PFU) technology is the ITER W/Cu PFU, which is a monoblock structure composed by tungsten, copper and copper alloy (CuCrZr) as plasma facing material, interlayer and heat sink, respectively. Meanwhile copper and copper alloy play a role of structural materials in the divertor structure, on account of huge electromagnetic (EM) loads which are induced by eddy currents and halo currents in the magnetic field imposed on copper and copper alloy parts. The EM forces would give rise to material high strain rates because it can reach to a very large value. Moreover, sustained so high heat flux from high temperature plasma, temperature distribution of all divertor components is also complicated. In order to investigate the dynamic response of copper and copper alloy in divertors under EM loads, an experiment on CuCrZr used in EAST divertors was carried out employing a Split Hopkinson pressure bar (SHPB). As supplementary, quasi-static compression and tension tests on copper alloy and copper were performed to obtain basic stress-strain relationship using MTS hydro-servo system and DDL50 electronic testing machine separately. The true stress and true strain curves in relation to strain rates or temperatures are derived from above experiments, which would provide the necessary theoretical basis for evaluation on the dynamical response, fatigue life and damage evolution to divertor components under EM impact loads.

        Speaker: Mr X. Mao
      • 1:40 PM
        The effect of He nanobubble on inhibiting D trapping in radiation damaged tungsten 2h

        Previous study results show that He nanobubble has great effectiveness on reducing deuterium (D) retention by acting as D diffusion barrier in undamaged commercial ITER grade tungsten (W), however, this effect on radiation damaged tungsten has not been extensively studied. In this paper, He plasma exposure (ion flux: 1023 m-2 s-1, fluence: 1025 m-2) pre-treatment was performed to create a thin He nanobubble layer in tungsten at ~773 K, after which these samples were irradiated by 5 MeV Cu ions or 5 MeV C ions to induce average peak dpa of 0.001 to 0.1 at room temperature. Samples without He plasma exposure pre-treatment were irradiated by Cu or C ions at the same time. All the samples above were subsequently exposed to D plasma at 350 K to a fluence of 1026 m-2. Nuclear reaction analysis (NRA) was used to evaluate the D distribution profile in the near surface, while the total D retention was measured by thermal desorption spectroscopy (TDS). Possible mechanisms are proposed to interpret the experiment results.

        Speaker: Mr Quan Bai (SWIP)
      • 1:40 PM
        The enhancement of high temperature deformation resistance for V-4Cr-4Ti alloys 2h

        Vanadium alloys, especially those with the compositions of V-4Cr-4Ti, are important candidate materials for blanket in future fusion reactors. In the past one decade, to enhance the high temperature mechanical properties, which equates with safely increasing the operation temperature, is one of the main efforts made to V-4Cr-4Ti alloys. According to dislocation theory, to properly increase defects density is an efficient way to strengthen practical alloys. Moreover, for high temperature application, the thermal stability of such defects is required.
        This work presents various commercial techniques used for strengthening the V-4Cr-4Ti alloy, including alloying, cold work, aging and mechanical alloying (MA) as the highlighted topic. With characterization of the tensile and creep properties at elevated temperatures coupled with investigation on deformation mechanisms of the resulted V-4Cr-4Ti materials, more work is then focused on the nano-particle dispersion strengthening.
        In the experiments, different starting powders, carbide dispersion agents and MA routes are used. Results show mechanically alloyed V-4Cr-4Ti alloy with Ti3SiC2 addition exhibits promising strength at both room temperature and elevated temperatures. Especially, its steady creep rate is almost one order lower than melted V-4Cr-4Ti alloy. The mechanism is considered as the thermal stable nano-particles resisted dislocation motion at high temperature, and is worth being introduced to the strengthening of other structural materials.

        Speaker: Dr Pengfei Zheng (Southwestern Institute of Physics)
      • 1:40 PM
        The feasibility of application the existing IVVS concept to CFETR 2h

        The Chinese Fusion Engineering Test Reactor (CFETR) is the next generation fusion device of China for realization of fusion energy. Its mission aims to bridge the gaps between the fusion experimental reactor ITER and the demonstration reactor (DEMO). To carry out the maintenance work of CFETR, remote handling systems shall be employed. To inspect the position and status of in-vessel components of CFETR, a remotely operated in-vessel viewing system (IVVS) shall be developed. A kind of IVVS prototype with an articulate arm and a viewing system has built in China, which bases on the size of EAST Tokamak. To test, install and store the IVVS around the Tokamak, a cylindrical vacuum vessel with foundation support inside is designed and checked. To connect the vacuum vessel to the Tokamak, an ultra-high vacuum gate valve a bellow is utilized between them. To keep the vacuum degree of the Tokamak after connection, a set of vacuum unit with molecular pump is configured for the vacuum vessel. Various performance tests of the IVVS are performed on the remote handling test facility before it connects with EAST Tokamak. The test result shows that it is feasible to apply this IVVS concept to CFETR.

        Speaker: Mr ZIBO ZHOU (CASIPP)
      • 1:40 PM
        The forming die design and experimental research of CFETR Vacuum Vessel shells 2h

        Abstract:Vacuum Vessel is the main component of CFETR (China Fusion Engineering Test Reactor), a series of theoretical studies and engineering verifications should be done from the period of Vacuum Vessel structure design to analysis optimization, from manufacture and assembly process to the plasma experiment study. Notably, shells forming is relatively in the earliest phase, also one of the most important processes during Vacuum Vessel manufacturing. The profile errors caused by spring-back of shells at room temperature will directly influence the subsequent manufacturing techniques and experiment studies, so a set of dies whose profiles can be revised have been designed based on the spring-back theory. Differently from the regular casting press forming mold, the set of dies consist of surface plate and a framework structure of which the assembly and disassembly is easy to made according to the forming dimensional quality. Firstly, the empirical formula has been used to calculate the theoretical die profile value which is mostly determined by the mechanical properties of 316LN. Then, choosing three relevant nearby values to simulate the forming process in which the optimal profile errors of shell must meet the demand of ±2mm, Finally, the optimal dimensional value of the molds have been confirmed to guide the forming experiment study. After experiments, the actual profile errors, thickness reduction, maximum spring back, maximum deformation and surface residual stress have been measured. And those data are an accurate indicator of what from the FEA, verifying the reasonability of the manufacture process which can be used to guide the forming of the whole D shape shells.
        Key Words: CFETR Vacuum Vessel, Forming Process, Molds Design, Spring-back, Residual Stress

        Speaker: Yuncong Huang (Southwestern Institute of Physics)
      • 1:40 PM
        The in-vessel protection components for ITER First Plasma operation 2h

        ITER plasma operations begin with a limited 'first plasma', with the main purpose of concluding integrated commissioning of the tokamak. This first plasma is targeted to last at least 100 ms and with a minimum plasma current of 100 kA. As in other tokamaks, the first plasma is conducted prior to full installation of the baseline in-vessel components. Namely, the blanket and divertor are installed only after first plasma experiments are completed. To avoid potential adverse effects of start-up plasma on the vacuum vessel and other already-installed high-value components (in-vessel coils, cable looms, etc.), several temporary in-vessel components are installed to protect all in-vessel systems that cannot tolerate direct plasma interaction.

        Machine protection is in part provided by temporary limiters and divertor replacement structures, which together create a poloidal and toroidal guide to shelter the vacuum vessel from the plasma and possible fast particle beams. 72 temporary limiter structures are distributed around the tokamak as four poloidal loops of eighteen tiles that follow the first wall contour. For engineering safety during plasma commissioning, the system is capable of maintaining plasma pulses of up to 3 seconds, tolerating up to 30 MJ of thermal energy and 1 MA of plasma current. Divertor replacement structures complete the bottom periphery of the poloidal loop to interrupt any potential downward plasma movement.

        In addition to protection from the plasma, the vessel also requires protection from the electron cyclotron resonance heating (ECRH) beam. This beam injects 6 MW of microwave energy across the null region of the vacuum vessel to help energize and assist the breakdown of first plasma. To protect the vacuum vessel, an inboard mounted mirror reflects the ECRH beam energy from its upper port origin into an outboard equatorial port, where a second component (a beam dump) can effectively absorb the radiation.

        These first plasma protection components (FPPC) are unique to ITER, and fit the need of a simple and cost-effective system that is easily installed and un-installed prior to further plasma operation in such a way as to minimize later impact on machine operation. The initial FPPC effort matured these components to a conceptual design level in preparation for the Conceptual Design Review held in November 2016. This paper summarizes the FPPC design and analysis status resulting from this initial effort.

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

        Speaker: Dr Ryan Hunt (ITER Organization)
      • 1:40 PM
        The measurement of visible bremsstrahlung emissivity profiles on HL-2A 2h

        Reliable experimental determination of the mean effective charge (Zeff) is of great importance for the impurity control in high temperature plasma. The Zeff is usually calculated using data from line integrated bremsstrahlung measurements, electron density and temperature profiles, and the plasma geometry. The visible bremsstrahlung emissivity profiles are achieved with the spatial resolution of 1 cm on HL-2A tokamak for the first time. Light-absorption panels that face the telescope are installed on the inner vessel wall, aiming at reducing the influence of wall reflection on the bremsstrahlung measurements. High-resolution spectrometer is used to select the impurity-line free region instead of interference filters, owing to the fact that in discharges with impurity injection or auxiliary heating power injection, some unpredicted impurity lines may invade this region, giving rise to a deviation of profile measurements. The line integrated bremsstrahlung emissivity profile evolutions are obtained under different discharge conditions, such as NBI and/or ECRH injection, L-H mode transition, and impurity injection, etc. It is found that compared to the center-peaked profile shape in Ohm discharges, NBI power gives a different profile shape with another high peak in the vicinity of r=26cm.

        Speaker: Mr Liu Liang (Southwestern Institute of Physics)
      • 1:40 PM
        The numerical simulation for the heat transfer enhancement experiments of the HCCB-TBM first wall 2h

        The first wall (FW) of helium gas Cooled Ceramic Breeder (HCCB) Test Blanket Module (TBM) for ITER need bear the loads like high power density heat flux from plasma, nuclear heat from neutron deposition on the structure, and the transient high heat loads like plasma disruption. The average heat flux density on FW of HCCB-TBM is about 0.3MW/m2 and the maximum partial transient heat flux may reach up to 1MW/m2. In current design schemes in ITER, Reduced-Activation ferritic/Martensitic (RAFM) steel is mostly selected as structural material, several groups of “radial-circular-radial” flow channels are placed inside FW, and 8MPa high pressure helium gas is applied as coolant to remove the heat flux on the surface and the nuclear heat by neutron deposition. As for smooth channel, required heat transfer efficiency and structural security can only be achieved after the flow velocity of helium gas reaches 50~80m/s (operating temperature of RAFM steel structure is lower than 550℃). High flow velocity of helium gas consumes a large amount of pumping power which lowers the net output power of reactor and increases greatly the equipment cost. Although roughness (less than 10μm) technique on the flow channel surface enhance heat transfer efficiency to some extent, the average heat transfer coefficient increases by less than 10% (from 2700 W/m2K to 2900 W/m2K). To enhance the helium gas cooling efficiency and security in FW, heat transfer enhancement technology needs improving and optimizing for the design scheme of helium flow channel to meet the functional requirements of FW. Based on this objective, the filling-evacuating HPHCL (High-Pressure Helium-Cooled Loop) were build to test and prove the heat transfer enhancement schemes of helium gas cooling FW. In this paper, the design scheme of the filling-evacuating HPHCL is presented, and the key issues of engineering manufacture and the test cases are calculated and analyzed. As for the first step, the CFD numerical simulation method is adopted to simulate the test cases of filling-evacuating HPHCL. The sustainable evacuating time under different mass flow rate of the He gas are estimated. On the basis of calculating helium gas cooling scheme of FW smooth flow channel, FW structural temperature gradient, maximum wall temperature, average heat transfer coefficient, and pressure drop of flow channel are selected as evaluation indexes. Three dimension numerical simulation results are compared to acquire optimizational heat transfer enhancement schemes like placing transversal ribs and V-shaped ribs in the flow channel of front wall of FW. The helium gas turbulence intensity and the heat transfer area are improved through optimizing the distance and angle between V-shaped ribs and other coefficients to enhance heat transfer. The calculation results are used as reference for the next verification experiments, which the 8~10MPa high pressure helium gas will be selected as coolant.

        Speaker: Mr Desheng Cheng
      • 1:40 PM
        The reliability design of CFETR divertor 2h

        The design of divertor is one of the most challenges for CFETR as the max heat flux on the CFETR divertor would be more than 20MW/m2 and the max plasma current is 1MA. Surface and structural damage to divertor due to the frequent loss of plasma confinement remains a serious problem for the tokamak reactor concept. The deposited plasma energy during major disruptions, edge-localized modes (ELMs), and vertical displacement events (VDEs) causes significant surface erosion, possible structural failure, and frequent plasma contamination. The duty cycle of CFETR should be at least 0.3~0.5. This suggests a very high reliability for divertor and other key subsystems. At the pre-conceptual phase the structure of divertor has to be improved by system and reliability analysis. The plasma facing components, supports and their joints have been optimized by results of reliability design. The reliability design uses probabilistic methods to describe the loads and operation conditions and estimates the frequencies of plasma events.

        Speaker: Dr Lei Cao (Institute of plasma physics, Chinese Academy of Science)
      • 1:40 PM
        Thermal Analysis and Test for the Mockup of ITER Radial X-Ray Camera 2h

        Radial x-ray camera (RXC) is an important diagnostic device in the International Thermonuclear Experimental Reactor (ITER) tokamak. The function of x-ray camera is to provide a measurement of soft X-ray emission from plasma, which is necessary for the operation and control of plasma to support key physics researches. As a key device to measure the plasma in tokamak, it is required to meet the requirement of working temperature. In order to get the temperature distribution of the RXC in tokamak, the mockup of RXC is fabricated for test. In this paper numerical analysis is performed for the thermal loads on RXC mockup. The finite element model of RXC includes heat exchanger, detector, diagnostics shield module (DSM), cooling pipe, support and vacuum vessel. The heat source of RXC system comes from the high-temperature field of vacuum vessel, which causes temperature rise of RXC through heat conduction and radiation. To keep the RXC system in a reasonable working temperature, an efficient cooling system is designed. In the cooling system, helium is used as cooling medium to cool down RXC system through heat convection. The thermal analysis model combined of heat conduction, radiation and convection is performed for the conjugate heat transfer of RXC system. The analysis result shows a very similar temperature distribution of RXC with the test, which demonstrated the viability and accuracy of analysis method and test process and that present cooling system design meets the temperature requirement of RXC system. All the work done in this paper will provide an important reference for the RXC design.

        Speaker: jian ge (asipp)
      • 1:40 PM
        Thermal and mechanical analysis of the Wendelstein7-X cryo-vacuum pump plug-in 2h

        The function of the cryo-vacuum pump (CVP) system is basically the control of the plasma density by condensing undesirable gases together with a set of turbo molecular pumps. One CVP will be installed under each of the 10 units of the actively cooled divertor in Wendelstein7-X for the long pulse operation up to 30 minute duration scheduled in 2020. The 10 CVPs are independent and each one is operated with supercritical Helium (ScHe) at 3.5K and liquid nitrogen at 77K fed by a plug-in, which is installed inside a dedicated W7-X port of the plasma chamber. The plug-in made of stainless steel provides for the vacuum boundary between the plasma chamber and the torus hall atmosphere. The outer dimensions of the plug-in are: ~ 2 m long and ~ 90 mm. 4 pipes (12 x 1 mm) are positioned inside the plug-in: 2 for the inlet/outlet of ScHe and 2 for the inlet/outlet of nitrogen, respectively. On the supply interface side, the pipes are equipped with bellows to compensate the thermal elongation during operation. The connection to the CVP is equipped with flexible hoses to allow compensating of assembly tolerances and to accommodate the displacement of the plasma chamber during operation. The design needs to guarantee the feeding at the specified temperature of ScHe and nitrogen while minimizing thermal losses and thermal interactions between pipes. Inside the plug-in the vacuum level is 10^-3Pa at RT and 10^-5 Pa during operation. The pipes of the ScHe are shielded with a multi-layer super-insulation. In addition the cryogenic feed lines are protected with a cryo-shield against thermal loads in the port as well as in the plasma vessel. During baking, the relative displacement due to thermal expansion and mechanical load between the port and the cryostat could damage the plug-in and endanger the CVP feeding. This paper presents the thermal and mechanical analysis performed with ANSYS to check the selected design of the plug-in of the CVP.

        Speaker: Dr Zhongwei Wang
      • 1:40 PM
        Thermal Strain Measurement of EAST Tungsten Divertor Module with Bare FBG Sensors 2h

        Tungsten divertor is one of the most important plasma facing components in EAST device. However, it has complex structure and faces extreme work environment. Centralized thermal strain would cause leakages on some weak welding seams, which was harmful for divertor’s operation. To measure divertor’s thermal strain shall be a valued way to understand its service behavior and then optimize its design and manufacturing process. Fiber Bragg Graing (FBG) was a proven technique to measure temperature and strain in sensor field. Though no similar works have been done before, it was considered an appropriate and feasible method to measure thermal strain on such a high temperature and strong electromagnetic field work environment.
        In this work, a heat-resistant bare FBG sensor system had been introduced to measure surface thermal strain of one EAST tungsten divertor module. Ten FBG sensors made in four optical fibers were included in this system. Among the ten FBGs, seven were used for strain measurement and three for strain compensation. A logical compensation method had been adopted to make the results more credible. Heating procedure had been divided into four stages: fast heating stage, slow heating stage, heat preservation stage and cooling stage. Two thermocouples had been used for a feed-back loop to control heating rate and to record temperature.
        The strain measurement system had withstood as high as 210℃ temperature and finished the experiment successfully. Experiment results showed that three measurement areas were under tensile strain and four were under compressed strain. Detail strain values had been calculated approximately. The results also had been compared with the measurement results using electric resistance strain gauges. Through the comparison, major results on corresponding areas were found similar, which told that the measurement results were reliable to some extent. In general, areas under tensile strain were more possible to leak than areas under compressed strain. This experiment would be a meaningful reference for tungsten divertor’s optimization and maintenance.

        Speaker: Mr Xingli Wang (Institute of Plasma Physics, Chinese Academy of Sciences)
      • 1:40 PM
        Time Synchronization Network for Poloidal Field Power Supply Control System Based on IEEE 1588 2h

        This paper describes architecture and characteristics of time synchronization network of EAST poloidal field power supply control system. After analyzing the characteristics of IEEE 1588, the paper points out the advantages of IEEE 1588 and choses it as the implementation protocol for time synchronization network in EAST poloidal field power supply control system. With this method, a master node with a hardware clock synchronizes time of the slave nodes on the private time network in each second, and the maximum offset time between the master and the slave is less than 50ns. This time synchronized method is evaluated and tested with its feasibility, and also met the requirements for the time synchronization network of EAST poloidal field power supply control system.

        Speaker: Mrs Shiying HE (Institute of Plasma Physics, CAS)
      • 1:40 PM
        Tokamak Size Scaling 2h

        The size scaling is recently developed by Costley, Hugill and Buxton (CHB) for seeking highest tokamak performance within physics limits at fixing the fraction of Greenwald density limit fGW, normalized plasma pressure βN, and fusion power Pf by scanning device-major-radius, R. The size-cancelling effects of the density limit are found in the fusion triple product of Lawson criterion, nTτE and fusion power gain, Qf. In CHB scaling, cylindrical geometry plasma is assumed to be in the nearly-full-sized vacuum chamber of tokamaks, together with a minimum of 250 MWt fusion power output in a JET-sized machine [A.E. Costley 2016 Nucl. Fusion 56, 066003]. However the assumption meets the real low-power-gain case of JET experiments at less than 20 MW due to low burn rate of deuterium-tritium (DT) fusion. The compressed plasma is thus suggested for filling the power gaps of existing low-power-gain cases of tokamaks to high-power gain. Existing limitations of EAST tokamak are analyzed for accommodating and simulating the high-performance discharges, including the additional pulsed power suppliers and magnets. Possible operation scenarios of tokamaks are further analyzed for high-gain high-field (HGHF) fusion plasma suggested in [Li. G., Sci. Rep. 5, 15790 (2015)].

        Speaker: Prof. GE LI (Inst. of Plasma Physics, CAS)
      • 1:40 PM
        Transient stability analysis of a flexible generator used in fusion power plant 2h

        The China fusion engineering test reactor (CFETR) could output 50~200 MW fusion power for demonstrating power generation. The overall structure of fusion power plant in concept design is similar to a pressurized-water reactor (PWR), except a double-fed induction generator with flywheel (DFIGF) has been proposed to replace the traditional synchronous generator used in a PWR. The heat source of fusion station is not constant in long pulse operation mode of Tokamak. Hence it will lead to frequent inrush in synchronous generator and power grid. Although a high power electric heating steam pressure regulator can mitigate the inrush, it consumes a lot of energy. The proposed DFIGF has good power regulation ability and large energy storage capacity, which suits the application when instability of heat source exists. The simulation model has been established and dynamic response analysis has demonstrated that DFIGF can ride-through the burning interval smoothly and provide a flexible connection with power grid.

        Speaker: Dr Hua Li (Institute of Plasma physics, Chinese Academy of Sciences)
      • 1:40 PM
        Transverse velocity effect on Hunt’s flow 2h

        The study of magnetohydrodynamic (MHD) flow is important in the application of the liquid metal blankets in thermal nuclear fusion reactors. Hunt presented an analytical solution of the conducting fluid flow in a rectangular duct under a uniform transverse magnetic field with insulating walls parallel to the magnetic field and thin walls with arbitrary conductivity perpendicular the magnetic field. The case is called Hunt’s case II, which is recommended as a benchmark to verify and validate MHD codes related to fusion applications. Hunt’s analytical solution is based on two dimensional fully developed laminar flow assumption, which means that the velocity vector has only flow direction component as a function of planar space. However, the transverse velocity component vertical to the streamwise and magnetic field direction in three dimensional numerical simulation increases with the increasing of the Reynolds number and redistributes the streamwise velocity even the flow is laminar.

        Hunt’s case II with the same Hartmann number (Ha=500), wall conductance ratio (c=0.1) and wide range Reynolds number (2000≤Re≤10000) have been simulated using three dimensional MHD solver developed in OpenFOAM environment to investigate when the transverse velocity must be considered. The streamwise velocity and the pressure gradient obtained from numerical simulation are compared with Hunt’s analytical solutions, which are set as the standard results. If the relative difference of the maximum velocity or pressure gradient is more than 5%, it is defined that the two dimensional assumption breaks down. The results show that the relative difference of the velocity and the pressure gradient increases when Re≥4000. The relative difference of the velocity is 9.551% when Re=8000. The relative difference of the velocity and the pressure gradient is 14.152% and 6.229% respectively when Re=10000. Numerical simulation shows that the dimensionless transverse velocity normalized increases and rises from the order of 10^-5 to 10^-3 .

        The effects of the Hartmann number on the transverse velocity has been investigated by simulating Hunt’s flow at Re=4000, C=0.1 and 50≤Ha≤1000. It shows that the relative difference of the maximum velocity is more than 20% when the Hartmann number is less than 100, which indicates that the dimensionless transverse velocity is the order of 10^-3.

        Finally, the transverse velocity effect as a result of the wall conductance ratio is simulated. The difference of the pressure gradient is lower than 5% for c=0.01. When Re/Ha>12, the difference of the maximum velocity increases linearly with the Re/Ha increasing for any wall conductance ratio.
        In conclusion, the comparison of the three dimensional numerical simulation and the two dimensional analytical solution of the Hunt’s case II shows that the transverse velocity is not negligible when Reynolds number is more than 4000. The transverse velocity is also influenced by the wall conductance ratio and the Hartmann numbers. As a result, if the dimensionless transverse velocity is over the order of 10^-3, the two dimensional assumption becomes invalid and the analytical solution is no longer suitable for high Reynolds number MHD duct flow validation.

        Key words: Transverse velocity, Magnetohydrodynamic,Hunt’s flow, numerical simulation,three dimensional

        Speaker: Mr Hao Wang (School of Mechanical Engineering, Hangzhou Dianzi University,Hang Zhou, China)
      • 1:40 PM
        Tritium transport analysis for one water-cooled ceramic breeder blanket module of CFETR based on COMSOL 2h

        Tungsten and RAFM steel are respectively used as armor material of first wall (FW) and structure material. The Li2TiO3/Be12Ti mixed pebble bed is selected to fulfill tritium breeding for water-cooled ceramic blanket (WCCB) in China Fusion Engineering Test Reactor (CFETR). Tritium would be existed in all materials of the blanket because of its diffusivity and permeability. Thus, it is crucial to study the tritium retention and permeation in the main domains for the safe operation of blanket, especially considering the radioactivity of tritium. Based on the Finite Element Method (FEM), a two-dimensional tritium transport model is set up using COMSOL Multiphysics. The temperature field is constructed to display the temperature distribution within the model, considering the nuclear heat generated and the cooling of the coolant. The velocity field of He purge gas is simulated to analyze the ability of the helium purge gas making tritium entering into Tritium Extraction System (TES). Meanwhile, the processes of tritium diffused through the interface of different materials, permeated into the coolant and taken by purge gas are considered to calculate the tritium retention and permeation coupling with the temperature and velocity fields. The calculation results indicate that the permeation amount into the coolant of FW is 2.18×10-6mg/s, while that of four cooling plates is 2.70×10-6mg/s. In the purge gas, the tritium would be carried out at the speed of 1.27×10-3mg/s, assuming the existing HTO totally entering into TES. The retention in the inner pebbles is 0.03g. The accuracy of the result is also discussed in this paper.

        Speaker: Mr Dingyu Lao (Institute of Plasma Physics Chinese Academy of Sciences (ASIPP))
      • 1:40 PM
        Upgraded Design of EAST Lower Divertor 2h

        EAST is one of the most important experimental fusion devices of China, the design of each component has important reference significance for China Fusion Engineering Testing Reactor (CFETR). Divertor, as one of the most important in-vessel components on the EAST, has always been quite difficult and challenging for its design. With the completion of the upgrade of the upper divertor, EAST has achieved a series of good test results, and then, the thermal load capacity of the lower divertor has become a bottleneck that constrains EAST to obtain higher parameters. This article explains the upgrade of the lower divertor which referred to the structure of tungsten monoblock on upper divertor. The lower divertor uses the circular monoblock structure at hit point area, at the end of monoblocks, end boxes are applied. The role of the end boxes is rational distribution of water flow in the premise of maintain the existing water supply capacity (1.8KG/s), which will improve the thermal load capacity of lower divertor from 2MW/㎡ to 5MW/㎡. Finally, the structure of the target plate satisfies the requirements, at the same time, the existing space can also accommodate the installation of the support structure. The reconstruction of the lower divertor not only provides support for the subsequent physical experiments on the EAST, but also provides an important reference for the design of the CFETR divertor.

        Key Words:EAST,lower divertor,monoblock,tungsten,thermal load capacity

        Speaker: Mr Pengfei Zi ( Institute of Plasma Physics Chinese Academy of Sciences)
      • 1:40 PM

        One of the conceptual designs of the breeder blanket for a European DEMO is the Water Cooled Lithium Lead (WCLL) concept. Design development of the WCLL blanket in recent years has led to evolution of the neutronics model employing the MCNP code, with key changes in radial lengths and thickness as well as dimensional differences in inner structures such as breeding zone, shielding zones and manifold. Furthermore there has also been a significant increase in projected fusion power. Such changes were made with respect to its nuclear, thermohydraulic and thermomechanical performances. In this work neutronic characteristics of WCLL modules were analysed in terms of fusion power increase and dimensional changes of the blanket geometry presented in DEMO 2014 and 2015 models. For comparison, outboard and inboard blanket modules of equatorial region were selected. Investigated neutronic characteristics include activity, decay heat and contact dose rates. Numerical experiment was carried out with MCNP particle transport code and FISPACT activation calculation code using EAF-2010 nuclear data library.
        Keywords: nuclear fusion, FISPACT, MCNP, neutron activation, DEMO, WCLL

        Acknowledgements: 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: Andrius Tidikas (Lithuanian Energy Institute)
    • 3:40 PM 4:00 PM
      Break 20m
    • 4:00 PM 6:00 PM
      M.OP1: Plasma Operation and Control Salon 1

      Salon 1

      • 4:00 PM
        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.

        [1] G. Calabro, et al, Nucl. Fusion 55 (2015) 083005;

        Speaker: Bingjia Xiao (Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, 230031, China)
      • 4:20 PM
        Real-time control of MHD instabilities using ECCD 20m

        For ITER and also future devices like DEMO, the capability to control magneto-hydrodynamic (MHD) instabilities like sawteeth and neoclassical tearing modes (NTM) is required to ensure reliable high $\beta$ plasma operation. Such plasmas naturally encounter these phenomena, some of which are performance limiting at best but destructive in the worst case as NTMs may cause disruptions. A tried and trusted actuator with which these instabilities can be strongly influenced and eventually controlled is electron cyclotron resonance heating and current drive.

        ASDEX Upgrade is making a large effort to develop, operate and evaluate an ECCD based, generic solution to MHD control, easily portable to new devices like ITER. Depending on the control strategy with highest priority, be it sawteeth or NTMs, ECCD deposition may need to be targeted at different radial locations, but the general control scheme is very similar, hence a so-called supervisory controller can delegate tasks to lower-level controllers and achieve a globally ideal solution given the existing constraints. Moreover, thanks to its generality, it can easily be adopted by other plasma experiments. TCV, among others, has started similar programs.

        In order to achieve precise deposition control, a large number of real-time diagnostics and intelligent controllers work in unison, all coordinated by the discharge control system (DCS). In addition to the real-time equilibrium reconstruction, which is essential for our application, we require density profile measurements, real-time detection of MHD marker positions (rational surfaces, inversion radius, etc.) and some global plasma parameters ($I_p$, $\beta_{pol}$) which complement the dataset on which the controllers base their decisions.

        Using the system in closed loop operation with 4 completely independent actuators, we have achieved controlled stabilization of 3/2 NTMs at $\beta_N$ of 1.8 and preemption of NTM onset using the same control mode reaching $\beta_N$ of 2.3 without mode. The system can automatically identify magnetic islands and aim stabilizing ECCD at the appropriate rational surfaces. Newly introduced deposition sweeping schemes alleviate the deposition accuracy requirement for NTM stabilization such that even imperfectly measured flux surface geometry is not prohibitive for achieving the intended goal.

        The controller approaches maturity and is undergoing optimizations to improve its performance and reliability. For this step, we employ detailed data analysis with a beam tracing code to determine the physical limits of successful stabilization. In the case of NTMs these are dependent on the ratio of externally driven current to bootstrap current at the location of the magnetic island. Detailed analysis and a full system overview are being 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: Matthias Reich (IPP Garching)
      • 4:40 PM
        Real-time detection and localization of magnetic island used for neoclassical tearing mode control and disruption mitigation 20m

        It is well known that the neoclassical tearing modes (NTMs) can be destabilized by the perturbed bootstrap current, reducing the plasma confinement and even leading to a major disruption in both standard ELMy H–mode and advanced tokamak scenarios. In order to stabilize the NTMs with electron cyclotron resonant heating (ECRH) and electron cyclotron current drive (ECCD), the magnetic island is required to be localized accurately and then the EC beam power is deposited exactly inside the island. For the NTMs suppression on EAST tokamak, a real-time system to detect the magnetic island and trace its radial location has been developed. In this system, the diagnostic signals from electron cyclotron emission (ECE) and Mirnov coil measurement are acquired and processed in real time to obtain the frequency and amplitude of magnetic perturbation as well as the mode radial location; as an alternative, the soft-x ray signals are taken to deduce the mode location instead of the ECE diagnostic in the case that the low hybrid wave is applied to plasma heating and current drive. The construction and the algorithm implementation of the real-time system is introduced in this paper.
        As the outputs of the real-time system, the mode radial location is provided to the ECRH launcher to determine the angle of EC beam injection. The island amplitude is used to control the gyrotron power on and off, and meanwhile involves in the feedback control of the ECCD deposition position with respect to the island position. Furthermore, the island amplitude is also monitored to generate a disruption alarm to activate the massive gas injection valve for the disruption mitigation, since in some cases the NTM suppression by the ECCD could fail due to an ineffaceable misalignment, the insufficient EC power for the mode stabilization and so on, and then the magnetic island would grow further leading to a major disruption. An integrated control strategy available to both NTM control and disruption mitigation is being developed and is expected to be presented.

        Speaker: Ms Yang Zhang
      • 5:00 PM
        A first analysis of JET plasma profile based indicators for disruption prediction and avoidance. 20m

        Disruptive events still pose a serious problem for the protection of in-vessel components of large size tokamak devices, representing therefore a key aspect to be considered for the design of next step fusion devices such as ITER and DEMO. If an efficient mitigation is strongly required to avoid damage and preserve the structural integrity of the machine, efficient avoidance schemes are needed to possibly bring the plasma back to a safe operating condition. In this framework, disruption prediction plays a key role and in the last few years a substantial effort has been devoted to developing more sophisticated prediction systems and improving their performance both in terms of success rate and warning time. Many of the presently developed disruption predictors mainly rely on MHD markers related to still rotating modes and, especially, to locked modes, which are basically the final precursor of most of the disruptions. Nevertheless, in many cases the detection warning time is still unsatisfactory with respect to avoidance requirements, and a significant step forward needs to be taken.
        This work deals with the development of “plasma profile based indicators” for disruption prediction and avoidance in JET, where parameterized peaking factors have been implemented for electron temperature, density and plasma radiation profiles. The basic interplay of the time evolution of different profiles will be described in relation to the phenomenology characterizing specific disruption types together with the relevant time scales. Furthermore, a statistical analysis aiming to describe differences and boundaries between the safe and the disruptive space as well as among specific types of disruptions will be presented, discussing the implications in terms of disruption prediction and avoidance.

        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 and the ITER Organization.

        Speaker: Dr Alessandro Pau (Electrical and Electronic Engineering Dept- University of Cagliari, Cagliari, Italy)
      • 5:20 PM
        New control ability on EAST PCS for steady-state operation 20m

        EAST (Experimental Advanced Superconducting Tokamak), a toroidal device with a D- shaped poloidal cross section, aims at high confinement and steady-state operation with plasma current up to 1 MA and pulse length to 1000 s. To accomplish EAST physical targets, the plasma control system PCS, adapted from DIII-D PCS [1] and deployed on EAST in 2005, keeps in continuous development. Some new control abilities for steady-state operation has been achieved. One is the long pulse data acquisition/archiving using data segment technology of Mdsplus. The acquired raw data and calculated result of PCS can be read or analyzed by physical operators in real time, which will provide the possibility to adjust the control scenario during the plasma discharge [2]. Another is the loop voltage feedback controlled to realize the non-inductive operation. In 2016 EAST campaign, loop voltage is well controlled using low hybrid wave (LHW). Besides, another two control algorithms are implemented to reduce the divertor heat flux. One is radiation power control, which is successfully feedback controlled by using divertor inert gas puff and mid-plane supersonic molecular beam injection (SMBI). The other is quasi-snowflake (QSF) shape control using PEFIT/ISOFLUX, which shows significant heat load reduction to divertor target [3] according to the modeling and experiment result. In this paper, the strategy and implementation detail will be introduced. The steady-state ELM-free high confinement QSF discharge has been achieved with the pulse length up to 20s, about 450 times the energy confinement time. The present EAST PCS has become a huge system capable of long pulse, high performance advanced plasma control operation, which is ready to demonstrate ITER-like control contents.

        [1] J.R. Ferron, B. Penaflor, M.L. Walker, et al., Fusion Engineering, vol. 2, p870 (1996);
        [2] H. Wang, J.R. Luo, G.M. Li, P.J. Wei, Cryogenics and superconductivity, 34(1), p26 (2006);
        [3] G. Calabro, et al., Nucl. Fusion 55, 083005 (2015)

        Speaker: Dr Qiping Yuan (Institute of Plasma Physics, Chinese Academy of Sciences)
      • 5:40 PM
        ELM pacing with lithium granules injection in W divertor on EAST 20m

        Pellet ELM pacing is a baseline ELM control strategy for ITER. While reliable and effective ELM pacing has been achieved by injection of deuterium pellets into carbon-walled tokamaks, the reduction of peak heat flux with high frequency ELM pacing in metal-walled tokamaks has been marginal. In comparison, the use of non-fuel pellets such as lithium (Li) is desirable for ITER due to its decoupling ELM pacing from fueling. For ELM control, Lithium Granule Injection (LGI) experiments have been carried out on EAST [1] and DIII-D[2]. The injection of sub-millimeter Li granules to trigger and pace ELMs has demonstrated heat flux mitigation on EAST and DIII-D, each with a carbon wall. In 2016, the LGI experiment was performed in tungsten (W) divertor on EAST, and some exciting results of LGI applications were obtained. ELM pacing efficiency was studied by injecting Li granules of nominal diameter 0.3–0.9 mm, with injected speed of 50–120 m s−1. Robust ELM pacing with 100% efficiency of ELM triggering by Li granules injection was demonstrated in ITER-like wall plasmas during Li granule injection. ELM frequency was paced to ~130Hz from 300Hz, and high Z impurity accumulation was not observed. The experimental observations indicate that ELM triggering efficiency depends on many interwoven parameters, such as granule size, penetration depth, and heating scheme. Higher power discharges require larger granules for efficient triggering. A wide range of granule penetration depths was observed by two fast cameras. It was also observed Li granules injection shifted the density profile outward, which changed the characteristics of the edge fluctuations, i.e. more easily destabilizing the edge coherent mode. This work strengthens the basis for ELM pacing in future reactors.
        This research is funded by the National Magnetic Confinement Fusion Science Program under Contract No. 2013GB114004 and the National Nature Science Foundation of China under Contracts No. 11625524 and No. 11321092.

        [1]. D.K. Mansfield et al., Nucl. Fusion 53 (2013) 113023
        [2]. A. Bortolon et al., Nucl. Fusion(2016) 56 056008

        Speaker: Dr Zhen Sun (Institute of Plasma Physics, Chinese Academy of Sciences)
    • 4:00 PM 6:00 PM
      M.OP2: Materials I Salon 2

      Salon 2

      • 4:00 PM
        Application of Materials Science Advances to Fusion Energy 20m

        The development of practical fusion energy as a commercial energy source is widely acknowledged as one of the greatest scientific and technical challenges for the 21st century. Due to the extreme operating conditions in the first wall and blanket regimes, utilization of very high performance materials is vital for achieving a viable cost-competitive fusion reactor design. The materials to be utilized in ITER are based on relatively conservative engineering designs and 1990s-era (or earlier) materials. Many of the current reference materials for the structure and functional applications of proposed DEMO fusion reactors are similarly based on 1990s-era knowledge. Significant advances in materials science and engineering have occurred over the past 20 years, including the emergence of computational thermodynamics as an accurate tool for rapidly assessing phase stability in structural alloys that can enable the design of improved high performance materials. Similarly, improvements in advanced manufacturing such as additive manufacturing for fabrication of geometrically complex and/or multiple-material components and friction stir welding for joining melt-sensitive alloys can enable innovative new component designs that would have been impossible 20 years ago. Several examples will be reviewed to illustrate the potential for achieving ultra-high performance and radiation resistance in new generations of structural materials for high heat flux and blanket structural applications, including new creep-resistant copper alloys and reduced activation ferritic/martensitic steels. Opportunities for utilizing advanced manufacturing in DEMO reactor components will also be summarized.

        Research sponsored by the Office of Fusion Energy Sciences, U.S. Department of Energy

        Speaker: Dr Steven J. Zinkle (University of Tennessee)
      • 4:20 PM
        Thermomechanical properties of nanostructured W based coatings under ITER-relevant thermal loads 20m

        The full tungsten (W) divertor of ITER will suffer from extreme thermal loads during both steady and transient operating conditions. These thermal loads, together with energetic species bombardment from the plasma, induce W surface erosion, melting and recrystallization. The sputtered particles could migrate and form micrometric thick co/re-deposits on the peripheral regions of the divertor. These layers, together with recrystallized and solidified W, show different thermophysical properties from bulk W and must be opportunely characterized.

        Since we actually are not able to recreate the full ITER environment, it is mandatory to study at the lab-scale how W based components behaves in ITER-like operating conditions. In order to study the behavior of co/re-deposits expected in ITER, in previous works we exploited the versatility of Pulsed Laser Deposition (PLD) to deposit W based coatings, namely pure W coatings with tuned nanostructure and morphology and W-oxide coatings with different oxygen contents. These coatings, chosen as proxy of co/re-deposited W, have been exposed to ITER-relevant plasmas, and their deuterium retention properties, as well as their structural and morphological modifications after exposure, have been assessed [1, 2].

        In this work we characterize the mechanical properties (i.e. stiffness and ductility) of these PLD W coatings by Brillouin spectroscopy (BS), as function of nanostructure (e.g. crystallite size) and oxygen content. Thermal properties, i.e. coefficient of thermal expansion, are also assessed by an ad-hoc developed experimental setup based on substrate curvature measurement. In addition, we investigate their thermomechanical behavior under two different scenarios that mimic steady and transient ITER operating conditions. For the former case, we perform standard thermal annealing treatments at 200–1000°C on nanocrystalline-W (nano-W) samples, in order to study their behavior at ITER-relevant steady operating temperatures. We focus on the mechanical, structural and morphological properties modifications upon heating. BS is exploited to derive the mechanical properties, while samples structure is assessed by SEM and XRD analysis. We find that nano-W starts to crystallize at around 600 °C, which is well below the bulk W recrystallization temperature (i.e.1400°C); at this temperature, comparing to as-deposited nano-W, an increase by 60% of material stiffness with a corresponding loss of ductility by 30% is observed [4]. In addition, we expose the as-deposited coatings to nanoseconds laser irradiation. Nanoseconds lasers have been already exploited for mimicking thermal effects induced by ITER-like transient events (e.g. disruptions, ELMs)[3]. Here, exploiting the same Nd:YAG laser system we used for PLD, we look for thermal effects (e.g. cracks formation, melting) as function of laser energy fluence. The characterization is assessed by SEM morphological analysis. The fluence thresholds for the thermal effects are then compared with the ones obtained by the irradiation of bulk W plates, selected with different surface finishes. The measured experimental thresholds are compared to the ones obtained by numerical simulations using a 2D thermo-elastic code developed to this purpose.

        [1] M.H.J‘t Hoen, et al., J.Nucl.Mater. 463, 989(2015)

        [2] A.Pezzoli et al., J.Nucl.Mater 463, 1041(2015)

        [3] N.Farid, et al., Nucl.Fusion 54, 012002(2014)

        [4] E.Besozzi, et al., Mater.&Design 106, 14(2016)

        Speaker: Mr Edoardo Besozzi (Politecnico di Milano)
      • 4:40 PM

        Helium atoms, produced at high rates in steels in fusion environment, are inclined to be deeply trapped in small vacancy clusters and microstructural features due to its low solubility in metal. Eventually, the formation of He bubbles significantly degrades the mechanical properties of materials. Therefore, it is important to understand the nucleation of He bubbles in steels, both in the bulk and within microstructural features, especially under irradiation.
        In this presentation, the irradiation cascade damage process was simulated by molecular dynamics (MD) methods to investigate the formation and growth of He bubble in BCC iron under irradiation in which the energy of PKA is up to 200 keV. The effects of temperature and He concentration were analyzed. The temperature ranges from 300 K to 800 K. He atoms are randomly inserted into the iron matrix, either in tetrahedral or octahedral positions, and the corresponding He concentration is from 1000 appm to 3000 appm. The results distinctly show that the number of He-V clusters increases with increasing the PKA energy and dislocation loops with different types and sizes are produced. The formation and growth of He bubble is obviously faster with higher temperature and larger He concentration, respectively. Furthermore, the size of He bubble is almost distributed as the Gaussian distribution.

        Speaker: Jie Zhan (University of Science and Technology of China)
      • 5:00 PM
        Effects of high-energy C ions irradiation on the D retention behavior in V-5Cr-5Ti 20m

        Alloys based upon the V-Cr-Ti system (e.g. V-4Cr-4Ti, V-5Cr-5Ti) are attractive candidate structural materials in future fusion reactors because of their low activation properties, high thermal stress factor, good strength at elevated temperatures, and usable fabrication properties. However, the high hydrogen isotope retention in vanadium alloys has been a serious concern in the potential application as fusion structural material. In addition, the high fluence radiation of 14 MeV fusion neutrons will produce various kinds of defects, which could make the problem of hydrogen isotope retention in vanadium alloys even worse. So far, this issue has not yet been investigated systematically, due to both the extreme lack of 14 MeV neutron sources and the activation of the neutron irradiated samples. High energy heavy ion beam has long been used to simulate radiation effects of high-energy neutrons. In this paper, samples made of V-5Cr-5Ti alloy are irradiated by 5.5 MeV carbon (C) ions with dose of 2×10^14, 1×10^15, 3×10^15 C/cm2.
        To investigate the defect properties in the irradiated samples, doppler broadening spectrometry of positron annihilation (DBS-PA) tests are carried out at room temperature with an energy-variable slow positron beam. In the measurement, the doppler broadening spectrum of the annihilation radiation was examined by a high-purity Ge detector, recording the gamma rays with energy 499.5–522.5 keV. The S parameter in DBS-PA increases significantly while the W parameter decreases with the increase of the irradiation dose, which indicates that the vacancy-type defects are introduced by C ions irradiation.
        To characterize the effects of irradiation on the deuterium (D) retention property of V-5Cr-5Ti, the irradiated and unirradiated samples are implanted with D in an ECR linear plasma device at the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP). The electron temperature and density were Te ≈ 2.4 eV and ne ≈1.5×10^17 m-3, respectively. Thermal desorption spectroscopy (TDS) experiments are followed, and the D retention behavior in the irradiated and unirradiated samples are compared and analyzed.

        Speaker: Mr Yu-Ping Xu (Institute of Plasma Physics, Chinese Academy of Sciences)
      • 5:20 PM
        The experimental investigation of wetting property for liquid lead lithium alloy with breeder blanket materials 20m

        The dual-cooled lead lithium (PbLi) blanket is considered as one of the main options for the Chinese DEMO reactor. The liquid PbLi alloy is used as breeder material and coolant. The Reduced Activation Ferritic/Martensitic (RAFM) steel and the silicon carbide fiber (SiCf) are selected as its structural material and functional material respectively. In the present experimental investigation, the special vacuum experimental device has been built, and the ‘dispensed droplet’ modification of the classic sessile droplet technique has been used to investigate the wetting property and inter-facial interactions for PbLi/RAFM steel, PbLi/SS316L steel, PbLi/SiC and PbLi/SiCf couples. The contact angles were measured between the liquid PbLi and the various candidate materials under working temperature from 300 oC to 480 oC. The results could provide meaningful compatibility database of liquid PbLi alloy and valuable engineering design reference of candidate structural materials and functional material for future fusion blanket.

        Speaker: Prof. Weihua Wang (Institute of Applied Physics, Army Officer Academy)
      • 5:40 PM
        Design, synthesis and characterization of Li4SiO4-based solid solutions as advanced tritium breeders 20m

        Linjie Zhao*, Xiaojun Chen, Chengjian Xiao, Yu Gong, Heyi Wang, Xinggui Long, Shunming Peng

        Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621999, China

        *Corresponding author : zhaolinjie1989@163.com

        The breeding blanket is a key component of the fusion reactor since it involves tritium breeding and energy extraction, both of which are critically important for the development of fusion power. Different lithium based ceramics have been studied as attractive tritium breeder materials, Li4SiO4 has been selected as one of the most promising candidates for solid tritium breeding materials in fusion reactors because of its high lithium atom density, its high melting temperature and favorable tritium release behavior. Li4SiO4-based solid solutions: Li4+x(Si1−xAlx)O4 and Li4Si1-xTixO4 were prepared as advanced tritium breeder to improve the mechanical property, irradiation resistance and reduce the tritium retention. Different Li4SiO4-based solid solutions powders and pebbles containing aluminum and titanium were prepared by solid state reactions and Modified melt-spraying process. Phase analysis, microstructures and density of the ceramics were determined by XRD, SEM and Archimedes' method. Impedance spectroscopy was measured to evaluate the electrical conduction properties of the ceramics. The thermal conductivity was determined using a laser flash device. Tritium release performance in Li4+x (Si1−xAlx)O4 and Li4Si1-xTixO4 irradiated with thermal neutron was studied by out-of-pile annealing experiments. These facts would represent the following advantages to use Li4SiO4-based solid solutions in blanket system of D-T fusion reactor that the thermal conductivity is higher and tritium inventory is lower in Li4SiO4-based solid solutions than those in Li4SiO4.

        Keywords: Li4+x (Si1−xAlx) O4, Li4(Si1-xTix)O4, thermal conductivity, the mechanical property, tritium release performance

        Speaker: Linjie Zhao (China Academy of Engineering Physics)
    • 4:00 PM 6:00 PM
      M.OP3: Next Step Devices, DEMO, Power Plants Salon 3

      Salon 3

      • 4:00 PM
        Status of K-DEMO Design Concept Study 20m

        The conceptual study on the Korean fusion demonstration reactor (K-DEMO) has been carried out since 2012 [1]. K-DEMO is featured by the medium size tokamak (R = 6.8 m, a = 2.1), a high magnetic field (B$_{To}$ = 7.4 T) with steady-state operation. The primary candidate of coolant medium is the pressurized water. One unique aspect of K-DEMO is a two-staged development plan to mitigate the gaps between the present level of technology and the required technology level for the full functions of DEMO. At first, K-DEMO targets not only to demonstrate a net electricity generation (Q$_{eng}$ > 1) and a self-sustained tritium cycle, but also to function as a component test facility. Then, at its second stage, a major upgrade is expected to replace in-vessel components in order to demonstrate a net electric generation on the order of 500 MWe.
        A preliminary operating scenario using a combination of various H&CDs (heating and current drives) covering neutral beam, electron cyclotron, lower hybrid, and fast wave H&CDs has been derived. The total H&CD power is estimated approximately 110 MW. The main components of K-DEMO have been conceptualized. The superconducting magnets (toroidal field (TF), poloidal field, and central solenoid magnets) were developed. Key features of the K-DEMO magnet system include the use of two TF coil winding packs, each of a different conductor design, to reduce the construction cost and save the space for the magnet structure material. The CICCs (Cable-In Conduit Conductors) for each type of magnets were fabricated and tested. Divertor is adopting the monoblock-typed tungsten armors with the reference choice of a double-null operation. Solid ceramic pebble typed lithium orthosilicate (Li$_{4}$SiO$_{4}$) was primarily selected for the tritium breeder. Extensive mechanical and neutronic analyses have been carried out to support the developed design concepts and the results are presented.

        [1] K. Kim et al., “Design concept of K-DEMO for near-term implementation”, Nuclear Fusion 55 (May 2015) 053027 (9pp).

        Speaker: Dr Keeman Kim (National Fusion Research Institure, Republic of Korea)
      • 4:20 PM
        Building a Virtual Tokamak - Integrated Multi-Physics Modelling for Fusion Engineering 20m

        The design of any tokamak reactor presents one of the greatest engineering challenges in the world today, in particular for DEMO-class machines (which we define here as tritium self-sufficient, net electricity producing devices). By its very nature, such an endeavour requires the coordination of a vast effort spanning many fields, bridging physics and engineering disciplines. Typically, this activity is guided by the 0-1D systems code, PROCESS, which performs an extremely fast, preliminary, single-parameter optimisation of plant design parameters to meet a set of input constraints and requirements. This is done in a matter of seconds, and the design point generated forms the basis of all further design studies and analyses. These activities cover an extremely broad range of different areas (e.g. superconducting magnets, breeding blankets, remote maintenance, etc.) and typically last one to two years before meaningful results can be fed back to the systems codes and another design point can be generated.

        This work presents UKAEA’s approach to bridging the feedback gap between the ~1s 0-1D systems codes and the ~1-2 year discipline-specific design studies. We present case studies that illustrate the first steps towards the realisation of a UKAEA advanced parameterised engineering design tool enabling the rapid generation of optimised tokamak designs: a tool to build an in silico tokamak.

        The design for a parametrically engineered tokamak concept is presented, with a focus on the automated design of the superconducting toroidal field (TF) and poloidal field (PF) coils. We demonstrate some of the early capabilities of the code, including the capability to parametrically design, analyse, and optimise the superconducting coil cage of a tokamak (to the first order). A comparison study of the numbers of TF and PF coils is presented, and the resulting stored energy, superconductor volume, and cold mass calculated for each configuration. The impacts of different design philosophies are also assessed, such as adopting a super-X divertor geometry, or different first wall and vacuum vessel shaping strategies. Engineering tools for assessing structural integrity are combined with tools which as traditionally physics-based, such as equilibrium generators and coil placement optimisation routines, to evolve a viable plant concept in the form of automatically generated 3D geometry. This geometry is then used as a basis to measure the designs’ performance analytically using more detailed structural, thermal, and neutronics codes.

        The advanced parameterised engineering design tool is aimed at supporting engineering decision making, rapidly delivering the necessary substantiated designs and performance data to the designer, so that they may better understand the far-reaching implications of their design choices on the performance of DEMO-class tokamaks and fusion power plants as a whole.

        Keywords: DEMO, systems codes, TF/PF coils, plasma equilibria, super-X divertor

        Speaker: Mr Matti Coleman (UKAEA)
      • 4:40 PM
        Conceptual development of K-DEMO, highlighting maintenance and support details of in-vessel components 20m

        The Korean fusion demonstration reactor (K-DEMO) has progressed through early concept definition activities to establish machine parameters, an operating point and the definition of the major core components. A key part of the conceptual development activities centered on the in-vessel components and the concept definition of the blanket/shield system, its segmentation and support arrangement. These systems have a major influence in defining the overall K-DEMO configuration and planned maintenance scheme. Earlier concept details of in-vessel systems has been updated with the addition of planned heating and current drive details, added blanket penetrations and the addition of some of the external heating systems located outside of the device core. Further definition of the blankets and support systems also led us to revisit the structural analysis of the in-vessel system design performed earlier [1, 2]. With in-vessel systems further developed, an initial assessment of a remote maintenance approach to remove all in-vessel components through the vertical ports also was made. The results of this activity along with an overview of the latest K-DEMO general arrangement will be presented.


        [1] P. Titus, et.al, “Disruption Analysis of the Proposed K-DEMO Blanket support Structure”, 21st Topical Meeting on the Technology of Fusion “Energy, 2014, Anaheim, CA
        [2] P. Titus, et.al, “Structural Assessments of the K-DEMO Blanket Modules”, 29th Symposium of Fusion Technology, September 2016, Prague, Czech Republic

        Speaker: Mr Thomas Brown (Princeton Plasma Physics Laboratory)
      • 5:00 PM
        Preliminary Research on Reliability Index System of Fusion Power Plant 20m

        Nuclear fusion is one of the most promising options for generating large amounts of carbon-free energy in the future. Since fusion energy is innovative and fusion facilities contain unique and expensive equipment, the reliability issue is very important from their efficiency perspective. The evaluation of reliability is an important part in the safety study of fusion reactor. And the system reliability index is the premise and the basis of reliability evaluation.
        This paper aims to establish the reliability index system of fusion reactor. Firstly, the safety goals of fusion reactor were given in this paper. In this study, the safety goals were separated into quantitative safety goals and subsidiary numerical objectives. Quantitative safety goals are higher than the numerical objectives, which come from the two 0.1% risk limits defined by the United States Nuclear Regulatory Commission (USNRC). Subsidiary numerical objectives are actually developed under the quantitative goals and are more specific to the characteristics of fusion reactor. Secondly, the safety goals of fusion reactor were assigned to the components which performed safety functions. In the part of this study, the Probability Safety Assessment (PSA) was used to establish the risk models for fusion reactor. The PSA is an important method to evaluate the risk of system, which has rich experience applied in nuclear industry for fission power plants and other nuclear installations. Thirdly, the reliability index system was given based on the results of the risk analysis of fusion reactor.
        The validation of reliability index system is still on study. The reliability index system is expected to be the basis and the reference for the reliability evaluation of fusion reactor and nuclear safety monitoring in future.

        Speaker: Dr Dagui Wang (Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences)
      • 5:20 PM
        Initial concept for the plasma diagnostic and control system for the European DEMO tokamak reactor 20m

        The development of the plasma diagnostic and control (D&C) system for a future tokamak demonstration fusion reactor (DEMO) is facing unprecedented challenges. The DEMO D&C system has to operate with very high reliability, since any loss of plasma control may result in machine damage. On the same time, high accuracy of the D&C system is needed in order to allow for plasma operation near operational limits, where the fusion power is maximized. The implementation and performance of diagnostic components is however limited by space restrictions (optimization of the tritium breeding rate; integrity of the first wall and divertor against loads), and by adverse effects acting on the front-end components (neutron and gamma radiation, heat loads, erosion and deposition). Finally, the capabilities of the available actuators (poloidal field coils, external heating and fueling) are limited as well.

        As part of the European DEMO conceptual design studies, the development of the D&C system has recently been launched [1]. A preliminary suite of candidate diagnostics for DEMO have been selected, aiming to cover all the main plasma control quantities with some redundancy, and choosing types and locations for diagnostic front-end components such that long-term durability with minimum need for maintenance can be expected under typical loads (e.g. neutron radiation, particle fluxes and fluencies). Specifically, only robust metallic or ceramics diagnostic components shall be used inside the vacuum vessel, while the more sensitive components shall be located at more remote (protected) positions. This initial plasma diagnostic suite comprises microwave diagnostics, magnetic coil based and Hall sensors, passive spectroscopy and radiation power measurements, divertor thermo-current measurement, infrared interferometry/polarimetry and neutron-gamma measurements. In the first R&D phase the possibilities and conditions for integration of diagnostic sightlines and front-end components into the machine have been investigated, and an understanding of the required number of channels and components has been obtained.

        The conditions for controllability of the DEMO plasma are being analyzed by numerical simulations. To this purpose, the transport modelling code ASTRA, coupled to a radiation module, has been connected to the Simulink simulation framework, and the performance properties of diagnostics and actuators are being added [2]. Similarly, the existing CREATE equilibrium code is being amended to include limitations of diagnostics and actuators, such that the controllability e.g. of fast VDEs can be simulated under DEMO relevant settings. Furthermore, predictive control oriented models such as RAPTOR are being further developed as an alternative approach to analyze the controllability of DEMO. The common goal is to arrive at numerical simulations which closely mimic the control of the DEMO plasma such that the controllability based on the available diagnostics and actuators can be demonstrated quantitatively. Based on these R&D results, an initial version of the DEMO diagnostic and control concept has been elaborated and will be presented in this paper.


        [1] W. Biel et al., Fusion Engineering and Design 96–97 (2015) 8–15

        [2] F. Janky et al., Fusion Engineering and Design, 2017 (submitted)

        Speaker: W. Biel (Institut für Energie- und Klimaforschung, Forschungszentrum Jülich GmbH, Germany)
      • 5:40 PM
        Status of the US Virtual Laboratory for Technology 20m

        The Department of Energy Fusion Energy Sciences (FES) program created the Virtual Laboratory for Technology (VLT) in 1998 with the goal of establishing a single entity with central leadership that would connect all aspects of the FES Technology Program. Since its inception, the VLT has been successful in conducting high quality research in support of the fusion energy sciences program mission as well as representing the technology program within the fusion community. The VLT represents the diverse activities of US Universities, Laboratories, and Industry involved in fusion technology research and development, and is organized into technical program elements that span the spectrum of technologies required to carry out its mission.

        While technical work at each of the member institutions has continued, the VLT has been reinvigorated over the last year following a roughly two-year hiatus. An overview of the current activities, new activities that have recently been initiated, and discussions of plans will be presented.

        Speaker: Dr Phil Ferguson (Oak Ridge National Laboratory)
    • 7:00 PM 9:00 PM
      Women in Engineering Reception 2h Meeting Room 5

      Meeting Room 5

    • 8:00 AM 10:10 AM
      T.PLN: Plenary T Grand Ballroom

      Grand Ballroom

      • 8:00 AM
        Announcements 10m
      • 8:10 AM
        Plasma Instrumentation for Spaceflight Missions 40m

        Plasma measurements are an important part of spaceflight missions that seek to understand the formation and evolution of our solar system. Instrumentation has been designed for a wide variety of environments and measurement goals. We have developed plasma instrumentation that will fly within 9 solar-radii of the surface of the sun on NASA’s Solar Probe Plus Mission. At the other extreme, we developed an instrument to measure the tenuous solar wind around Pluto at the edge of our solar system for the New Horizons Mission. At Earth, the Magnetospheric Multiscale Mission employs four spacecraft flying in formation to study magnetic reconnection on a global scale making measurements at unprecedented rates. While at Jupiter, the Juno Mission makes an in-depth study of Jupiter’s polar magnetosphere to measure the effect of the precipitating particles on Jupiter’s ionospheric layers, to determine the composition and structure of the field-aligned currents, and to understand the mapping of these currents to the outer magnetosphere and other parts of the Jupiter system. The instrumentation developed for these measurements spans a broad range of energies from a few 10’s of eV up to 100’s of MeV. A wide variety of techniques and sensor technologies are employed to make the measurements, sometimes requiring special shielding and coincidence techniques to reduce background from the harsh space environment.

        Speaker: Mr Scott Weidner (Princeton University)
      • 8:50 AM
        Status and Progress of JT-60SA 40m

        JT-60SA is a highly shaped large superconducting Tokamak device. The project mission of JT-60SA is to contribute to early realization of fusion energy by supporting the exploitation of ITER and by complementing ITER in resolving key physics and engineering issues for DEMO with a variety of plasma actuators (heating, current drive, momentum input, stability control coils, resonant magnetic perturbation coils, W-shaped divertor, fuelling, pumping, etc) .

        Fabrication and installation of components and systems of JT-60SA procured by EU and Japan are steadily progressing towards start of operation in 2020. Up to April 2017, seven TF coils have been arrived at Naka from EU (ENEA in Italy, and CEA in France) after completing the careful cold test and preassembly with the outer inter-coil structure at CEA Saclay. The 340-degree part of the Vacuum Vessel (VV) and the thermal shield surrounding VV has been welded accurately, and the four TF coils have been installed around VV (Fig.1). Manufacture of all the six EF coils have been completed by QST with excellent accuracy of manufacture. Commissioning of the cryogenic system from EU was also completed in the Naka site. Sixteen High Temperature Superconducting current leads (in total 26) has been deliverd from Germany (KIT). Commissioning of the power supply system (ENEA, RFX, CEA and QST) has also been implemented smoothly. Manufacture of the Cryostat Vessel Body is also reaching its final phase in Spain (CIEMAT).

        Fig. 1  Four TF coils have been installed around the Vacuum Vessel of JT-60SA (Mar. 2017)
        Fig. 1 Four TF coils have been installed around the Vacuum Vessel of JT-60SA (Mar. 2017)

        The JT-60SA Research Plan (SARP) ver. 3.3 was issued in March 2016 by 378 co-authors (JA 160 (16 institutes), EU 213 (14 countries, 30 institutes) and the STP-PT (5). The main revision point of ver. 3.3 is the update of EU-DEMOs and JA-DEMO parameters. The revision made it clear that JT-60SA covers wide research areas for DEMO, both pulsed and steady-state operations. The fifth JT-60SA Research Coordination Meeting (RCM) was held at QST Naka in May 2016. Contribution of JT-60SA to ITER was emphasized in relation to the expected delay of ITER. The physics R&D priorities in JT-60SA fulfilling ITER needs were suggested by the ITER Organization and discussed by all of the participants. Critical issues, such as disruption mitigation, L-H threshold power, ELM mitigation, diagnostic R&D, should be tested in JT-60SA. It became a consensus of the JT-60SA research unit to modify the basic strategy of the Integrated Research Phase II (~2030) of JT-60SA so as to start with full coverage of Tungsten divertor and Tungsten first wall and accompany the initial heating experiments of ITER.

        Speaker: Y Kamada (National Institutes for Quantum and Radiological Science and Technology)
      • 9:30 AM
        Progress in the EU DEMO Research and Design Activity 40m

        As part of the Roadmap to Fusion Electricity, Horizon 2020, Europe initiated a pre-conceptual design study of a Demonstration Fusion Reactor Concept (DEMO) a few years ago, which targets the generation of a few hundred MW of net electricity and the demonstration of a closed tritium fuel cycle in the 2050s.

        The design and R&D approach adopted include some distinctive elements such as: 1) a strong philosophy of integrated design at an early stage to encourage a more ‘systems thinking’ culture and to bring major clarity to a number of critical design issues and overall integration challenges; 2) an improved understanding of system context as a foundation for informed plant design concept and technology development programmes; 3) a prudently modest extrapolation from the ITER physics and technology basis, in order to minimize programme/development risks and their associated mitigation costs; 4) multiple DEMO plant design architectures are studied in parallel (e.g. reactor configurations such as a double-null tokamak), as are major sub-systems or technologies for which there are particularly high technical risks or low maturity (e.g. the divertor, the breeding blanket, etc).

        The progress of the EU DEMO design and R&D activities to date is described, with a focus on the areas that are believed to have a strong hand in defining the conceptual layout of the DEMO device, and drive its performance. Recently, a number of external and internal developments have occurred that challenge some of the assumptions underpinning the original schedule. This includes the delay of ITER construction and DT operations and a greater appreciation of the ‘integration challenge’ required to define a robust plant architecture. A reasonable extrapolation from ITER results is maintained, and a provisional, updated DEMO schedule is discussed.

        The pulsed EU DEMO baseline design point continues to be the primary configuration studied (in particular for integration issues – many of which have broad applicability to other reactor designs); however a number of alternative reactor configurations are now also being studied in earnest. These include for example a double-null divertor machine, and a pulsed “flexi-DEMO” machine capable of transitioning to steady-state operation. Preliminary results of studies exploring the available design space and defining the main parameters and technical characteristics for these configurations are shown. The design strategy of the plasma-facing system is discussed, and the preliminary definition of a DEMO plant layout is presented, aimed at enabling further design integration studies as well as safety and cost analyses for the wider plant auxiliary systems.

        Design and technology down-selection will be of vital importance on the path reaching a DEMO concept and it is critical that a robust decision-making framework is established in the years to come to support future decisions. Thoughts on such a framework are presented here, and on its application to the fusion R&D programme in the future to progressively narrow down sub-system technologies and reactor architecture options.

        Speaker: Dr Gianfranco Federici (EUROfusion)
    • 10:10 AM 10:40 AM
      Break 30m
    • 10:40 AM 12:40 PM
      T.OA1: Diagnostics and Instrumentation I Salon 1

      Salon 1

      • 10:40 AM
        Design, Manufacturing, and Integrated Testing of the ITER Port Instrumentation 20m

        At ITER more than 50 different diagnostics are under development for the tokamak. The diagnostic systems are designed to be integrated within the interspace and port cell support structures of 27 upper, equatorial and lower ports. Basic instrumentation and control (I&C) is required to monitor the temperatures at selected locations of the port plug and interface support structure and for the electric heaters used for baking of windows and thermal stress compensation. Spare measurement channels have to be provided for future use.

        Currently the ITER project is transitioning from the detailed design phase to manufacturing, testing and integration in preparation for integrated commissioning. The focus of the work is on the first plasma diagnostics for which system integration in the equatorial ports 11 and 12 is essential. Since the port system I&C is required for many port systems, the development is already quite advanced with manufacturing and acceptance testing currently taking place. Furthermore the port system I&C is typical for industrial plant I&C and can therefore serve as an example for those plant systems.

        The design process starts with the requirement capture from all relevant sources, continues with a description of use cases and operating procedures, and is followed by the functional analysis including the definition of all the variables providing the interface with CODAC through its networks. The software implementation process is based on the CODAC Core System (CCS) and CODAC provided tools. The integrated testing follows a set of test campaigns starting from acceptance of the installed hardware and of the source code in the software repositories.

        This paper presents the development process of the port system I&C through all lifecycle phases from design to site acceptance and summarizes the test results.

        Speaker: Stefan Simrock
      • 11:00 AM
        Design and Analysis Progress of US ITER Integrated Diagnostic Upper Port 14 20m

        ITER is the world’s largest fusion device currently under construction in the South of France with over 50 diagnostic systems to be installed inside the port plugs (PPs), the interspace or the port cell region of various diagnostic ports. The Diagnostic First Wall (DFW) and Diagnostic Shielding Modules (DSM) are designed to protect front-end diagnostics from plasma neutron and radiation while providing apertures for diagnostic viewing access to the plasma. Four tenant diagnostic systems will be integrated into the upper port plug 14. The upper visible/IR wide angle viewing system (Vis-IR/Upper Cameras), or WAV system, is installed to provide visible and IR viewing of the inner vessel for machine component protection during plasma operations. The disruption mitigation system (DMS) is installed to mitigate the negative effects of plasma events due to sudden loss of plasma current or control by rapid injection of cryogenic pellets to mitigate the dissipation of the plasma thermal energy, the control of the plasma current quench, and the suppression of the generation of Runaway electrons. The Glow Discharge Cleaning (GDC) system is installed for reducing impurity and provides control of hydrogenic fuel out-gassing from plasma-facing components. The Plasma Position Reflectometry (PPR), for real time determination of the wall-plasma distance, is the ex-vessel tenant installed in the U14 Interspace region.

        The PP engineering design and multi-physics analysis has been performed following ITER upper port integration requirements including weight, neutron shielding (100 uSv/hr total dose limit), cooling layout, allowable deflections and structural integrity validation under single and combined load cases. Various DSM design configurations have been analyzed and resultant component integration and mass distribution is optimized to limit its impact to the DFW (IO scope) and in-port diagnostics, to mitigate significant impact from the undesirable VDE (Vertical Displacement Event) inertial loads. The DSM design maintains EM load distribution similar to that from a generic box-like shielding structure, still provides needed stiffness for the protection of on-board diagnostics and structural integrity. To moderate impact from inertial loads due to the Vacuum Vessel (VV) movements during asymmetric plasma VDEs, the rigid lock-in DSM interface was implemented into the U14 port integration analysis models for design validation. Structural integrity of U14 assembly is largely driven by the electromagnetic loads induced on the metallic structural components during plasma VDEs. The in-port diagnostics and the mounting supports, on the other hand, are largely driven by the steady-state thermal loads from volumetric nuclear heating, and the dynamic response of components attached to the DSM-PPS assembly under the VDE inertial loads due to the vessel movements. Progress on the U14 integrated design and analysis is reported. The tenant interface load transfer is also presented in details for in-port system attached to the DSMs as part of the design and analysis tasks for ITER PP engineering.

        Speaker: Dr Yuhu Zhai (PPPL)
      • 11:20 AM
        Novel multi-energy x-ray cameras for magnetically confined fusion plasmas 20m

        A compact multi-energy soft x-ray (ME-SXR) camera has been developed for time, energy and space-resolved measurements of the soft-x-ray emissivity in magnetically confined fusion (MCF) plasmas. Multi-energy x-ray imaging provides a unique opportunity for measuring, simultaneously, a variety of important plasma properties ($T_{e}$, $n_{e}^{2}Z_{eff}$, $n_{Z}$, $\Delta Z_{eff}$ and $n_{e,fast}$). Selecting an appropriate detector response eliminates the contamination introduced by the low- and high-energy line-emission from medium- to high-Z impurities facilitating temperature measurements in Ohmic and RF-heated scenarios (e.g. ICRH and LHCD) in agreement with conventional ECE and Thomson scattering systems. Impurity density measurements are also possible using the line-emission from medium- to high-Z impurities to separate background as well as transient levels of metal contributions. This novel imaging system developed at PPPL and tested first at Alcator C-Mod tokamak at MIT, combines the best features from both pulse-height-analysis (PHA) and multi-foil methods, and represents a very large improvement in throughput and spatial resolution thanks to present state-of-the-art pixelated PILATUS detectors with nearly 100k pixels. Being the first of its kind, this novel diagnostic will be used to resolve the impurity emission, study impurity transport and impurity-induced MHD and will become an essential part of a control algorithm coupled to physics and engineer actuators for minimizing impurity accumulation in tokamaks. This technique should be explored also as a burning plasma diagnostic in-view of its simplicity and robustness. Recent results from a detector sensitivity study including its response at high magnetic fields (up to 3.4 T and 3.0 T/s) and ITER-like neutron fluences (up to $10^{15}-10^{16}$ n$_{eq}$/cm$^{2}$) will be presented.

        Speaker: Luis F. Delgado-Aparicio (Princeton Plasma Physics Laboratory)
      • 11:40 AM
        Surface deterioration and recovery of CXRS first mirror in EAST 20m

        First mirror (FM) is the key element of the optical and laser diagnostic systems in fusion devices such as ITER. Facing the plasma directly, it has to operate in an extremely harsh environment and suffer from the sputtering by high energy ions and charge exchange atoms, the impurity deposition due to wall conditioning and the sputtered wall materials etc [1,2], which results in the deterioration of the reflectivity and shorter lifetime. Protective shutter and plasma cleaning are widely studied to mitigate the deposition during the FM operating and remove the impurities deposits afterwards in recent years [3,4] .
        The non-plane large size (300 mm×80 mm×40 mm) FM, made of 316L SS was used for the charge exchange recombination system (CXRS), which has been operated in EAST for 361 days in three experimental campaigns from 2014 to 2016. The FM was exposed to the plasma with a total discharge pulse of 12499 shots and a total duration time of 86036 s. During the exposure, a capsule holder and a motor shutter were used to mitigate the deposition. The scanning electron microscopy (SEM), electron energy disperses spectroscopy (EDS), laser beam injection spectrum and a self-made laser system were used to characterize the surface morphology, impurity composition and the reflectivity of the CXRS FM. The inhomogeneous deposition consisting of C, O, Si, W and Mo was detected on the FM surface due to the shadow of the holder on FM surface and the gap between the shutter and the holder. The sizes of the particles were about several micrometers to tens of micrometers. Due to the severe deposition, the reflectivity of the FM was strongly decreased from 71% to 15% at the wavelength of 532 nm. To recover the FM surface and address the cleaning effectiveness and homogeneous of the non-plane large size mirror as well as find a possible way for FM in-situ cleaning in EAST in the future, the Ar plasma driven by the 13.56 MHz RF capacitively coupled system was used to clean the FM. After 187.3 h cleaning, the inhomogeneous deposition was visibly unseen and uniformly removed. The SEM and EDS results indicated that the micro morphology was developed during the cleaning and few residual particles consisting of C and O were also remained and covered about 5.3% of the cleaned mirror surface. The reflectivity of the FM surface was recovered to 68% which demonstrated the cleaning effectiveness and homogeneous of the RF plasma cleaning the non-plane large mirror. To prolong the FM lifetime, the optimization of the capsule holder and motor shutter was suggested and the regular in-situ cleaning was proposed.

        [1] V. S. Voitsenya, A. E. Costley, V. Bandourko, et al.Rev. Sci. Instrum. 72, 475 (2001).
        [2] A. Litnovsky, P. Wienhold, V. Philipps, et al. J. Nucl. Mater. 363–365, 1395 (2007).
        [3] Ivanova D, Rubel M, Widdowson A, et al. Phys. Scr. T159 14011 (2014).
        [4] Lucas Moser, Roland Steiner, Frank Leipold, et al. J. Nucl. Mater. 463 , (2015) 940 - 943.

        Speaker: Yan Rong
      • 12:00 PM
        Integration Conceptual Study of Reflectometry Diagnostic for the Main Plasma in DEMO 20m

        The reflectometry diagnostic may present several advantages from the point of view of radiation robustness and components life time as compared to some other traditionally used diagnostics in large fusion devices. From the hardware perspective it does not contain front end elements such as mirrors or sensors which are expected to underperform earlier than the antennas and waveguides when subjected to similar radiation fluxes and deposition/erosion processes. On the other end stresses arising from thermal expansions and electromagnetic (EM) forces can be larger for the waveguides and are accommodated by design. The role of such diagnostic for DEMO is twofold: i) to provide the radial density profile at several poloidal angles (2D map) and ii) to provide data for the feedback control for plasma position. Several groups of antennas need to be distributed along the poloidal section in a number that can satisfy the DEMO control requirements. The study of diagnostic performance and control requirements definition is still being developed and the final number of diagnostic channels is not yet defined, nevertheless several aspects regarding integration can be readily assessed. This paper presents the first case study of integration of antenna groups and waveguides located at several poloidal angular positions covering a full poloidal section of the Helium Cooled Lithium Lead breading (HCLL) blanket. The integration design shall satisfy strong machine driven constraints (in addition to the physics performance). Diagnostic components installed in the blanket segments must: i) survive for the all period between blanket replacement, ii) be remote handling (RH) compatible with blanket, iii) behave thermomechanical as the blanket structure, iv) cross with integrity the vacuum and reference boundaries (vessel/cryostat/building) and tolerate their relative displacements and v) be compatible with the blanket shielding and cooling services. The present solution developed so far respects several of the main constraints namely, RH compatibility with the full blanket segment and its thermomechanical properties and cooling compatibility but also identifies important issues on the interfaces between the diagnostic antennae extensions and the pipe services at the vessel and also interfaces between vessel and cryostat requiring challenging RH and self-alignment solutions to be demonstrated. Monte Carlo neutronic simulations have been initiated in order to evaluate the heat loads and shielding capabilities of the system. The first results indicate that the cooling for the EUROFER diagnostic components (antennas and waveguides) can in principle be provided by the blanket cooling services (He is considered) via connection to the main Back Supporting Structure (BSS) and routed via the main diagnostic structure body to specific hot spots in the antennas.

        Speaker: R Luis (Instituto dos Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade de Lisboa, 1049-001 Lisboa, Portugal)
      • 12:20 PM
        Prototype manufacturing and testing of metalized ceramic printed circuit boards for ITER Bolometer cameras 20m

        The ITER bolometer diagnostic will be based on 110 five-channel metal resistive sensors distributed all around the vessel [1]. In order for the diagnostic to meet its operational and programmatic goals, a mechanically stable and reliable electric connection is required. Spring loaded connection of the sensor or crimped wires to connect the external cables, as used in current day fusion experiments, are not suitable to operate under ITER's challenging nuclear and thermal loads. In addition, the design must not exceed its space envelope and an efficient way how to customize the sensor position according to the final blanket module positions has to be demonstrated. Excellent thermal conductivity in these internal components is also a requirement in order not to overheat the sensor.

        In this paper, a design solution based on metalized ceramic aluminum nitride substrates providing the electrical interface between the internal bolometer sensor and the externally connecting macroscopic MI cables, is proposed. The substrates, also referred to as printed circuit boards (PCB), can be manufactured having a complex 3-D shape and can be coated on multiple sides with micrometer thick conductive tracks and pads.

        It is shown, that the sensor can be supported mechanically with an integrated design small enough to fit into the tight space envelope reserved for the ITER bolometer cameras. A solution, to allow flexibility in sensor positioning during assembly, will be explained. Investigations into bonding and micro welding techniques to provide a reliable electrical connection as well as the possibility to integrate a remote-handling compatible connection will be discussed.

        Prototypes based on a simplified PCB design have been manufactured by micro-dispensing, laser etching and laser activation in order to validate these technologies for the demanding ITER environment. To determine the exact specifications and design constraints for the final electrical interface, these simplified PCBs contain vias, bond pads coated with different material combinations (Au, Pt, Cu) and tracks running over 3-D shaped surfaces.

        Test results on mechanical stability and electrical properties of the different simplified PCBs before and after thermal cycling are discussed together with the analysis of the achieved electrical track widths and thicknesses and their impact on the diagnostic performance. Moreover, conclusions and considerations on cost-effective manufacturing will be presented. The paper concludes with an outlook describing the preliminary bonding specifications and challenges for the internal bolometer sensor and the external MI cables.

        [1] H. Meister et al.: Current status of the design of the ITER bolometer diagnostic, submitted to Fusion Engineering & Design, 2016.

        Speaker: Dr Florian Penzel (Max Planck Institute for Plasma Physics)
    • 10:40 AM 12:40 PM
      T.OA2: Divertors and PFCs: Tungsten Salon 2

      Salon 2

      • 10:40 AM

        Main focus of fusion engineering has been moved towards development of metal wall plasma facing components (PFCs) and corresponding interaction between plasma and metal wall. National Fusion Research Institute (NFRI) has started metal wall related research activities since 2012, which are closely related to major upgrade of KSTAR and research on K-DEMO. As the first step, metal bonding technology has been developed and tungsten brazed block samples with good bonding quality have been obtained. Bonding technology for tungsten monoblocks using HIP is currently under development. Two major issues on tungsten divertors with castellated structure in ITER and beyond, are steady state and transient power handling capability and fuel retention inside the gap. Monoblocks aligned perfectly to their neighbors have leading edges directly exposed to plasmas. Leading edges under high power ELMy H-mode can be melted in several seconds of plasma exposure time. In order to solve this issue, radiated divertor and shaping of castellated monoblocks are proposed: Optimization of the shape and the angle of the castellation structure can reduce significant amount of heat load on the PFCs. Tore Supra has found that fuel retention was dominated by co-deposition, especially at the gaps of tile blocks. In order to study those two issues, special tungsten block tiles with various shapes of castellated structure with leading edges were fabricated and installed on the central divertor of KSTAR. It is found that the leading edge heat load can be described by using simple optical approach without Larmor orbit effect. Results also indicate clearly that the shape-optimized block has more heat load handling capability compared with conventional one, and the maximum temperature under heat load is much lower. The contributions of ions and charge-exchange neutrals on the deposition inside the gap of various shapes and heights of castellation structures have been measured and a complete set of deposition profiles inside the gaps was obtained. 0.3 mm misalignment allowed in ITER shows no meaningful difference on deposition profile. Since KSTAR has not enough heating power to sputter tungsten atoms from the blocks, transport of tungsten atoms in plasmas cannot be studied. We have developed a gun type powder injector to put tungsten powders into L- and H-mode plasmas, which provides evaporation of tungsten powders releasing a large amount of tungsten atoms. Vacuum Ultra Violet (VUV) spectroscopy, whose wavelength is around 6 or 12 nm, is used for the tungsten line measurement. Tungsten powder injection experiment has been successfully performed and the core accumulation of tungsten atoms is measured. Obtained tungsten emission spectra show very similar features measured at ASDEX Upgrade indicating tungsten atoms were evaporated from the powder and penetrated into the core. This has opened a new research area in KSTAR despite of low heating power.

        Speaker: Dr Suk-Ho Hong (National Fusion Research Institute)
      • 11:00 AM
        Investigation of ITER-grade tungsten under very high heat loads. 20m

        Experiments carried out on advanced large tokamaks showed effective use of tungsten for making in-Vessel Components, interacting with the plasma. However, in reactor size fusion devices such as ITER and DEMO, are expected the critical loads on the divertor plates both in quasistationary stage and in pulsed events (disruption, VDE, ELMs et al.), High heat loads can cause not only increased erosion and destruction of material surface, but also strong absorption of tritium in erosion products. Usually, it's difficult to obtain divertor ITER-like power load in advance fusion devices with magnetic confinement. Therefore, to simulate ITER conditions powerful e-beam try to use, but it can’t replace the real simulation by plasma. As example, JET and ASDEX-U experiments with movement of the molten W droplet, can be explained by electron emission from this droplet in the magnetic field.

        On T-10 tokamak with a powerful ECR heating, were obtain regimes with nonambipolar energy flow on tungsten tiles of circular toroidal limiter. ITER-grade tungsten was use, which is intend for the ITER Dome divertor, manufactured by RFDA. The interiors of the limiter are heated to temperature exceed of 2000 0C and estimated heating power is more than 10 MW/m2. Spectroscopic line WI near this plates show exponential increasing, but total radiation power decrease from 50% to 15%, and radiation loss at the boundary increase 3-4 time.

        In this regime, there were deep and long cracks and powerful arcing occurred on W tiles. At that, cracks in ion side are perpendicular to tile edges and parallel to each other, as threads. The area of cracks coincide with the area of arcing. The edges of the cracks were melt and arc craters have been scattered not only across the surface but located along the cracks. All tiles surface was cover by resolidificated tungsten, on which there were many arc microcraters.

        The report discuss the nonambipolar mechanism of energy flow on metal surfaces, leading to self-heating in the presence of arcs, the ecton mechanism of arcing, mechanism of cracking and estimation of tritium absorption in such kind of cracking.

        Speaker: Mr Leonid Khimchenko (ITER RFDA)
      • 11:20 AM
        Tungsten monoblock concepts for the U.S. Fusion Nuclear Science Facility (FNSF) first wall and divertor 20m

        Next-step fusion nuclear devices require plasma-facing components that can survive a much higher neutron dose than ITER, and in many design concepts also require higher operating temperatures, higher reliability, and materials with more attractive safety and environmental characteristics. In search of first wall concepts that can withstand surface heat fluxes beyond 2 MW/m$^2$, we analyzed advanced “monoblock” designs using coolants and materials that offer more attractive long-term performance. These use tungsten armor and heat sinks, similar to previous designs, but replace the coolant with helium and the coolant containment pipe with either low-activation ferritic-martensitic steel or SiC/SiC composite. Two geometries of coolant containment pipe, round pipe and elongated slot (as in microchannels), were examined via 3D thermal and mechanical analysis, which was performed parametrically for optimization. The results of analysis show that helium-cooled steel can remove up to 5 MW/m$^2$ of steady-state surface heat flux and helium-cooled SiC/SiC can remove nearly 8 MW/m$^2$ while satisfying all materials and design requirements. This suggests that a He-cooled W/SiC monoblock could withstand divertor-like heat fluxes. More detailed results and conclusions are as follows.

        A monoblock with ferritic-martensitic steel round pipes is limited to a steady state surface heat flux of 2.1 MW/m$^2$, increasing to only 2.4 MW/m$^2$, with the use of advanced steels. The higher allowable temperature of advanced steel can not be fully exploited because in this case the stress limits performance. The use of a slotted “microchannel” geometry provides substantial additional heat flux handling capability. For “ordinary” ferritic steel, the heat flux limit rose to 3.7 MW/m$^2$, which roughly meets our original goal to double the performance of the previous He-cooled design with W pins. This value rises to 5.2 MW/m$^2$ using ODS steel. In this case, stresses did not constrain the performance.

        The use of SiC composite pipes to replace steel was considered in the context of large existing R&D programs developing advanced fission fuel cladding. Used inside of a W monoblock configuration, round pipes can satisfy temperature and stress limits up to ~5 MW/m$^2$ steady state surface heat flux, whereas a microchannel geometry can reach near 8 MW/m$^2$. This value of heat flux approaches the range expected in a tokamak divertor. Exact specifications of heat flux in the divertor of burning plasma devices are not available, but peak values in the range of 5-15 MW/m$^2$ are expected.

        Our SiC/SiC design variant provides a possible alternative to the He-cooled W-alloy divertor that has been explored in several design studies and R&D programs. The W-alloy divertor has been shown to allow very high performance and heat removal capability, but the availability of an acceptable alloy remains a major uncertainty for its continued development. While SiC composites at present do not achieve the higher heat flux capability of W-alloy, due to limited thermal conductivity, their commercial availability and existing database under neutron irradiation make them a more likely candidate for near-term applications in next-step devices.

        The authors would like to acknowledge the contribution of FNSF research group.

        Speaker: Mr Yue Huang (UCLA)
      • 11:40 AM
        Thermal Stress Evaluation on the Optimized Shaping Design for Tungsten Monoblock in EAST Divertor 20m

        Abstract—This paper investigates the issue of leading edge for EAST tungsten divertor monoblock which is also concerned in ITER project. Besides the positive effects like reduced risk of cracking, the castellation will lead to the increased probability of melting of the castellated divertor due to local power load on leading edge of the gap. That may introduce unacceptable amount of impurity into plasma and cause damage to the plasma facing components. The chamfering on monoblock is applied in order to avoid melting due to the local heating at leading edge. The previous research [1] had calculated the temperature distribution by employing finite element method and proposed an optimized chamfering geometry for the W monoblock in EAST, which can effectively reduce the maximum temperature under 10 MW/m2 heat load. In this work the stress of monoblock under the thermal load is further analyzed by means of finite element software ANSYS in order to evaluate the integrality and lifetime of monoblock. Both of the steady state and transient (e.g. ELM) thermal load are considered in the numerical calculation. According to the results of recent researches[2-4] the cosine law is applied in the calculation for steady state thermal load, and the ion orbit model is used for ELM condition. The behaviour of crack initialization is analyzed by using damage parameter curve which is obtained under creep and fatigue load. Moreover, crack propagation rate model is employed in the fatigue and creep lifetime evaluation for both shaped and unshaped monoblock. The result shows that the shaping for monoblock will lead to movement of the location of highest temperature toward the central region of top surface of monoblock. That will increase the gradient of temperature, so as to enhance the stress at W/Cu interface. The shaping for monoblock can reduce the risk of melting due overheating on leading edge. However, the life time of monoblock could be shortened due to increased stress.

        Keywords—W/Cu monoblock, leading edge, stress, lifetime
        [1] Xiahua Chen, et al. Numerical optimization of tungsten monoblock tile in EAST divertor Fusion Engineering and Design, Vol.108, pp98-103, 2016,
        [2] R. A. Pitts, et al. Numerical evaluation of heat flux and surface temperature on a misaligned
        JET divertor W lamella during ELMs, Journal of Nuclear Materials, 2011, Vol.415 Issue 1, pp S957-S964.
        [3] R. Dejarnac, et al. Physics basis and design of the ITER plasma-facing components, Nuclear Fusion, Vol.54, pp 123011, 2014
        [4] R. A. Pitts, et al. Physics conclusions in support of ITER W divertor monoblock shaping, 22th International Conference on Plasma Physics Surface Interactions in Controlled Fusion Devices, 2016

        Speaker: Mr Xiahua Chen (Institute of Plasma Physics Chinese Academy of Sciences, University of Science and Technology of China)
      • 12:00 PM
        Precipitation of transmutant elements in neutron irradiated tungsten 20m

        As the leading plasma facing material in fusion reactors, tungsten is confronted with extremely hostile environment, characterized by high temperature, and high fluxes of heat and particles (i.e., D, T, He, and neutrons). One of the primary concerns is the generation of transmutation elements (i.e., Re, Os) and the subsequent radiation-induced segregation and precipitation, and the resulting thermomechanical property degradation induced by the 14 MeV-peak neutron irradiation. In this study, we have used advanced electron microscope methods to explore the response of tungsten to high dose neutron irradiation in the High Flux Isotope Reactor, focusing on the detailed characterization of irradiation-induced W-Re-Os precipitates in polycrystalline tungsten through TEM, X-ray mapping in STEM, multivariate statistical analysis data-mining of the X-ray data and transmission Kikuchi diffraction. The association of voids and precipitates, the chemical compositions, crystal structures and phases of precipitates along the grain boundary and within the grains were identified. The results showed that the intragranular precipitates are sigma-phase while the precipitates along the grain boundaries are chi-phase. The kinetics process of transmutant elements and radiation defects were briefly discussed to reveal the formation process of the observed precipitates.

        In addition, we also investigated the hardening contribution of W-Re-Os precipitates. A dispersed barrier hardening model informed by the available microstructure data was used to predict the hardness. The results indicated that the formation of intermetallic second phase precipitate dominant the radiation-induced strengthening with a relatively modest dose (>0.6 dpa). The hardening strength factor of the transmutation-induced precipitates was also determined to be 0.6.

        The work presented in this paper was partially supported by Laboratory Directed R&D funds at ORNL. The research was also sponsored by the US Department of Energy Office of Fusion Energy Science under grants DE-AC05-00OR22725 with UT-Battelle LLC and by the US-Japan PHENIX project under contract NFE-13-04478, with UT-Battelle LLC.

        Speaker: Dr Xunxiang Hu (Oak Ridge National Laboratory)
      • 12:20 PM
        Defect production and deuterium bulk retention in quasi-homogeneously damaged tungsten 20m

        Tungsten (W) is foreseen as the leading plasma facing material (PFM) for future fusion reactors due to its advantageous thermal mechanical properties and relatively low solubility of tritium (T). W-PFM in fusion reactors will experience intense radiation by 14 MeV-peaked neutrons (n), which have long mean free paths on the order of centimeters in solids. T retention in W may greatly increase owing to the T trapping effects of defects created by neutrons throughout the W bulk. Therefore, T bulk retention in n-irradiated W becomes a significant safety concern. Recently, heavy ions are widely used as surrogates for neutrons to investigate the influence of n-produced defects on T retention. However, the damaged layer of heavy ions is usually limited to a few micrometers beneath the specimen surface and the damage profile is strongly peaked. Hence the effects of homogeneously distributed traps on T retention in W have not been fully understood. In this study, by using ultra-high energy ions and special sample irradiation techniques, we produced a quasi-homogeneous distribution of defects in bulk W; then the deuterium (D, a surrogate of T) retention mechanisms in the damaged W are investigated.

        The high-energy heavy-ion irradiation was performed at Heavy-ion Research Facility in Lanzhou. Annealed W foils were irradiated with 122 MeV 20Ne ions in a terminal chamber where an energy degrader for defect distribution tailoring was used. SRIM calculation showed that a quasi-homogeneous distribution of atomic displacement damage to 0.16 dpa within a depth of 50 m was produced in W. Then results from positron annihilation lifetime characterizing exhibited an extra long positron lifetime component of ~400 ps in the irradiated W, indicating the formation of large vacancy clusters. After that, the sample was exposed to D2 gas at 773 K. Thermal desorption spectra featured a high D release peak at ~1010 K and a broad D desorption temperature range (730-1173 K) for the irradiated W, which was very different from the non-tailored W sample (much narrow desorption window). Further transmission electron microscopy characterizing (under and over-focus pairs) showed a large amount of voids with a diameter of ~1 nm in the irradiated W. These voids could be the large vacancy clusters formed by heavy ion irradiation and should be the main reason to the high temperature desorption of D. However, whether the broad D desorption window is related to the quasi-homogenous distribution of voids should be further studied.

        Speaker: Dr Feng Liu (Institute of Plasma Physics, Chinese Academy of Sciences)
    • 10:40 AM 12:40 PM
      T.OA3: Blankets and Tritium Breeding: Liquid Breeders Salon 3

      Salon 3

      • 10:40 AM

        China has long been active in pushing forward the fusion energy development to the demonstration of electricity generation. As one of the most challenging components in DEMO, great efforts have been put on the development of breeder blanket and three blanket schemes were studied in China for fusion engineering test complementary with ITER. In this paper, the main blanket concepts developed in China will be summarized including two leading schemes of Dual Functional Lead Lithium (DFLL) and Helium Cooled Ceramic Breeder (HCCB).
        For ITER-TBM, the Procurement Agreements of HCCB-TBM has now been confirmed and led by three institutes, i.e. SWIP (Southwest institute of Physics), INEST (Institute of Nuclear Energy Safety Technology) and CAEP (Chinese Academy of Engineering Physics), each playing a different role. The other candidate scheme, DFLL-TBM has also been continuously supported by CN-MOST (Ministry of Science and Technology) and will be tested in ITER under international cooperation.
        And the technical challenges to ITER-TBM and also the DEMO mainly focus on fusion material development and testing, breeder/coolant technology and experiment validation, effective tritium production/extraction to achieve self-sufficiency, reliability and safety etc., which are the nuclear technology basis of DEMO blanket. In this paper, the recent R&D progress on DFLL-TBM is presented, including the progress on structural material fabrication technologies and properties of CLAM steel, PbLi/He coolant technology and safety issues, the RAMI analysis, small mockup neutronics experiments, and tritium behavior etc.. Based on the conceptual design and latest technical R&D progress, the entire test planning will be scheduled for DFLL concepts towards DEMO blanket.

        Speaker: Prof. Qunying Huang (Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences)
      • 11:00 AM
        WCLL breeding blanket design and integration: lessons learned in 2016 and follow-up 20m

        The Water-cooled lithium-lead breeding blanket (WCLL) is a promising viable option for European DEMO nuclear fusion reactor. The liquid lithium-lead is the breeder-multiplier flowing at low velocity and low temperature (i.e. about 330°C). Pressurized water is in charge to cool the structure and to transport the heat towards the power conversion system (PCS) and the energy storage system (ESS). The structural material is the EUROFER. The WCLL breeding blanket studied during 2016, in the framework of EUROfusion Project, is based on the single module segment approach. Basically, it is a breeder unit, which is repeated along the poloidal direction. The power is removed by means of radial-toroidal (i.e. horizontal) water cooling tubes in the breeding zone. The lithium-lead flows in radial-poloidal direction. A 100 mm thick plate will connect the breeding blanket segment with the vacuum vessel (VV), through an attachment system. All these components shall be designed to withstand the loads during normal operation and accidental conditions. Water and lithium lead manifolds are designed and integrated with a consistent primary heat transport system and the lithium lead system.
        The paper discusses the WCLL breeding blanket design features through selected relevant thermo-mechanic, thermo-hydraulic and neutronic analyses. The lessons learned from the design review will be pointed out in order to present the reasons for the improvements and the needed analyses.

        Speaker: Alessandro Del Nevo (ENEA FSN-ING-PAN)
      • 11:20 AM

        A conceptual design of a DCLL outboard equatorial module was produced according to the specifications of the EU-DEMO based on 16 sectors, 1572 MW fusion power, with the main objective of bringing into maturity this breeder blanket concept. Specification and design guidelines for the DCLL Blanket System were developed, identifying the main requirements needed for the initial design and producing a preliminary CAD model of an equatorial module in a DEMO outboard segment based on neutronics, thermal-hydraulics and thermo-mechanical calculations. During the definition of this first conceptual design of the DCLL a new version of the EU-DEMO (with 18 sectors, 2037 MW fusion power) was released. Thus, the blanket design has been adapted to this new scenario by reviewing its operational conditions and producing important differences in the CAD model. Thus, some changes have been implemented with respect to the previous design, looking for simplicity. One of the most important ones is the new PbLi routing inside the modules, implemented to facilitate the draining of the individual modules. Related to this point, the previous annular geometry of the connection between the modules and the Back Supporting Structure has been simplified to reduce strong MHD problems. A comprehensive transient structural analysis revealed the occurrence of high stress concentration at the connections of the FW with the radial plates in case of a LOCA, therefore suggesting that an increase in the number of radial stiffening walls is necessary, and therefore the number of internal PbLi circuits.
        Specific design elements have been consolidated, such us the thermal-hydraulic general scheme for the segments, the poloidal segmentation or the structural design. A MHD estimation of the convective heat transfer coefficient has been performed, and serves as input for the thermal-hydraulic and structural calculations.
        Finally, an integration of the DCLL blanket within the PbLi loop is made, including the outcomes from tritium transport modeling in order to understand the overall behavior of the DCLL, as well as the impact on the Tritium Extraction System.

        Speaker: David Rapisarda (CIEMAT)
      • 11:40 AM
        Integration of the Neutral Beam Injector System into the DCLL breeding blanket for the EU DEMO 20m

        The integration of plant systems involving penetrations into the in-vessel components, like H&CD, fuel cycle and diagnostics, is a complex task constrained by top level requirements of remote maintainability and high reliability. Within the EUROfusion PPPT Program, some activities are ongoing to assess the integration of different systems into the breeding blanket, specifically NBI, ECRH launchers, diagnostics sightlines, fueling lines and specific protections for the FW (like start-up limiters).
        This work describes the integration of the Neutral Beam Injector (NBI) system into the Dual Coolant Lithium-Lead (DCLL) breeding blanket for the EU DEMO. After identifying the major issues impacting the mechanical, thermal-hydraulic and neutronic behavior of the blanket, the integration efforts have been focused on minimizing the invasiveness of the NBI system and exploring different NBI options for the best compromise between plasma heating and breeding blanket performance. This paper describes the adaptation of the DCLL breeding blanket design to allocate the neutral beam duct. A particular attention is devoted to the redistribution of breeding and shielding functions, the new path of fluid circuits and the additional cooling needs.
        The consequences of design modifications on key neutronic aspects like Tritium Breeding Ratio (TBR) and shielding capability are addressed. Besides, after a brief discussion regarding the thermal loads transferred to the breeding blanket walls from the neutral beam and the plasma, a preliminary thermal assessment of the proposed integration solution is presented.

        Speaker: Dr Iván Fernández-Berceruelo (CIEMAT)
      • 12:00 PM
        Development, characterization and testing of a SiC-based material for Flow Channel Inserts in high temperature DCLL blankets 20m

        Flow Channel Inserts (FCIs) are one of the key elements in the high temperature DCLL blanket concept, one of those being considered for DEMO. FCIs must provide the required thermal insulation between the blanket steel structure and the hot liquid PbLi that is flowing inside them; the high PbLi temperatures (up to 700 ⁰C) allow a high reactor efficiency, but impose a considerable thermal gradient across the FCI’s walls, generating mechanical stresses that must be supported without damage during the operation time. Besides, they should provide enough electrical insulation to minimize MHD pressure drop, they must be inert in contact with PbLi preventing corrosion damage, and should present low tritium permeation. To develop a suitable FCIs material with these requirements is one of the main challenges in the development of a high temperature DCLL.
        In this research, a SiC-based sandwich material is proposed for FCIs, consisting of a porous SiC core covered by a dense CVD-SiC layer. SiC fulfils the operational requirements for FCIs including low activation and degradation by neutrons, and porous SiC is an attractive candidate to obtain a thermally and electrically low conducting structure; to prevent corrosion by PbLi and tritium permeation, a dense SiC coating is applied on the porous material. To produce the porous SiC core of the sandwich, a method consisting of combining the particle size of the starting SiC powder mixture with a carbonaceous sacrificial phase is proposed, being the sacrificial phase removed after sintering by oxidation. In this work, a description of the production method is presented as well as the properties of the resulting porous material after sintering and oxidation, like porosity, microstructure, thermal and electrical conductivity, and flexural strength. By using this technique, a wide range of porous SiC materials with different porosities and thus, conductivities and strength values, can be produced. According to thermomechanical calculations and FEM models, and assuming a SiC dense coating of 200 µm and a porous SiC core of 5 mm, the core material should present a thermal conductivity ≤ 7 W/m·K at 700 ⁰C and mechanical strength > 50 MPa to ensure the required insulation and mechanical integrity. In this work, a material with porosity near 45% and thickness ≈ 5 mm, thermal conductivity of 7 W/mK and flexural strength about 100 MPa is proposed as porous core. Porous SiC samples covered by a dense CVD SiC layer of ≈ 200 µm were tested under hot PbLi to study their response against corrosion. A first batch of samples was tested under static PbLi at 700 ⁰C during 1000 h, after which they did not show any sign of corrosion damage. Then, a second batch of samples was tested under dynamic PbLi flowing at velocity near 10 cm/s at 550 ⁰C during 1000 h. A magnetic field of 1.8T was applied during the test to some of the samples to study its possible effects on the corrosion behaviour. Results of all corrosion tests are presented and discussed.

        Speaker: Mrs Carlota Soto (CEIT )
      • 12:20 PM

        The breeding blanket is the key nuclear component for power extraction, tritium fuel sufficiency and radiation shielding in fusion reactors. Using pure lithium (Li) or Li-containing liquid metal (e.g. eutectic alloy lead-lithium, PbLi) in fusion blankets as breeder is a very attractive option due to their high heat removal, adequate tritium breeding ratio, relative simple design, potential attractiveness of economy and safety. All liquid-metal blankets have special features associated with the nature of liquid breeders, including their high chemical reactivity, and especially interaction with the plasma-confining magnetic field. Flowing liquid breeder under magnetic field would result in various magnetohydrodynamic (MHD) phenomenon such as huge MHD pressure drop, quasi-two dimensional turbulence. It would be an effective way to reduce MHD effects by reducing electric conductivity of liquid metal breeder.
        A potential technology to reduce electric conductivity is nano fluid technology, which adds functionalized nanoparticles into fluid to change its physical property. We demonstrated that it is possible to reduce electric conductivity of liquid metal by adding electrically insulating nanoparticles. The liquid metal we tested was eutectic alloy of GaInSn, which is liquid at room temperature. The nanoparticle we chose was SiO2, whose electric conductivity was several orders of magnitude lower than that of liquid metal and had a good wetting property with GaInSn. SiO2 nanoparticles smaller than 200 nm in diameter were added into liquid metal GaInSn, forming dilute suspensions called nanofluids, which aimed to reduce the electric conductivity of the liquid metal. The nanoparticle weight fraction dependences of electric conductivity for GaInSn with fractions 0.05%, 0.1%, 0.2%, 0.5%, 1% were investigated and the electric conductivity measured from electrochemical workstation monotonically decreased with increasing nanoparticle fraction. The nanoparticle scale dependences of electric conductivity for GaInSn with particle nanometer scales 10nm, 20nm, 50nm, 100nm, 200nm were investigated and showed a weak relation. The best result we got was the case of 10nm with 0.5% weight fraction, where the electric conductivity was reduced by 4.25 times.
        Based on this study, we evaluated the MHD pressure drop of typical liquid metal blankets such as DCLL and DFLL blanket, and the pressured drop was significantly reduced, which means that nano fluid technology was fit for liquid metal blankets. Further planning was scheduled for tritium breeder PbLi and Li.

        Speaker: Dr zi meng (Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences)
    • 12:40 PM 1:40 PM
      Lunch 1h Shanghai City Bistro/EZO Restaurant

      Shanghai City Bistro/EZO Restaurant

    • 1:40 PM 3:40 PM
      T.POS: Poster Session T Junior Ballroom

      Junior Ballroom

      • 1:40 PM
        0-D Physical Design for the Heating and Current Drive System of CFETR 2h

        As the next step for the fusion energy in China beyond ITER, the China Fusion Engineering Text Reactor (CFETR) aims to operate with duty time as 0.3~0.5, means that CFETR should operate at steady-state scenario. This provides a great challenge for the physical design of the heating the current driving system. In general, four different kinds of method as NBI, ECH, LHW and ICRH have been developed in worldwide for heating plasma and driving current. Considering the characteristics of each H&CD system, we provide two design solutions as the one with NBI and all-wave solution. For the solution with NBI, the total design power is 73MW with 33MW NBI, 20MW LHW and 20MW ECRH; For all-wave solution, the total design power is 80MW with 20MW LHW, 40MW ECRH and 20MW ICRH. Those two solutions can satisfy the heating and steady-state operating aims of the CFETR through the 0-D physical design.

        Speaker: Dr Defeng Kong (ASIPP)
      • 1:40 PM
        3D numerical simulations of hypervapotron geometry on Thermalhydraulic Performance 2h

        In order to satisfy the EAST first wall and divertor upgrade plan, a hypervapotron (HV) cooling concept is chosen to be developed as a candidate for the design of PFCs. The HV structure relies on internal grooves or fins and boiling heat transfer to maximize the heat transfer capability. The fabrication technology of W/Cu divertor has been developed at ASIPP (Institute of Plasma Physics Chinese Academy of Sciences), and one W/CuCrZr/316L HV component will be fabricated for high heat flux tests. Before fabrication, the relevant analysis was carried out to optimize the structure of HV component element. In this paper, numerical simulaitons with a 3D model of 490 mm × 50 mm × 20 mm have been performed using the CFD (computational fluid dynamics) analysis by means of ANSYS FLUENT code. In the model, W tiles with thickness of 2mm were selected as armor tiles considering that 2-mm-thick W tiles are being used in EAST upper divertor. And two fin designs were compared for optimization, then the advantages of slots on the fins were also discussed. Besides, several width and shapes of the groove between the fin and the side wall were also compared. And for each design, the comparison between subcooled boiling and single phase convection has been carried out, as well.

        Speaker: Mr Ran Wei (ASIPP)
      • 1:40 PM
        A digital signal processing system of digital Rogowski current transducer with comb filter 2h

        In ITER poloidal field (PF) prototype converter testing, the Rogowski current transducer is used to measure the current in a DC bus bar. When thyristors, which are parts of PF converter, are triggered on, they will produce electromagnetic noise around. The noise signal, which has a strong amplitude and fixed frequency, is easy to be coupled by cable between Rogowski coil and integrator, and transmitted to integrator. Then, the differential signal, which is produced by Rogowski coil and proportional to the current in DC bus bar, will be submerged. Consequently, it will lead to a very low signal-noise ratio, and the integrator cannot work. A digital signal processing system has been designed to solve the problem mentioned above. The design is based on the dual-ADC structure digital integrator which has been developed at ASIPP. A digital comb filter is utilized to filter out the electromagnetic noise signal, and measures are taken to weaken it from the hardware perspective. The experiment indicates that the method presented in this paper can decrease the amplitude of electromagnetic noise, increase the signal-noise ratio and improve the measurement accuracy.

        Speaker: Mr Zhen Zhang (Institute of Plasma Physics Chinese, Academy of Sciences)
      • 1:40 PM
        A flexible web visualization framework for nuclear fusion experiment data 2h

        As the fusion experiment goes to steady state and more sophisticated diagnoses are developed, the experiment data becomes larger and collaboration between researchers tends to be more frequent. So a well-designed flexible and easy to use web visualization framework is becoming more important.

        The new web visualization framework is designed and implemented based on ASP.NET MVC framework. It is part of the JCDB project, which is a database cloud for J-TEXT based on Cassandra,. In JCDB data are stored in form of matrix and can be read and written efficiently with cursor and writer. MongoDB is used to store data structure. Models are designed to works with the JCDB backend.

        In the controller, we design a RESTful web API which allows users to access and operate data through HTTP after authorized. For GRUD data operation, we provide actions with get, post, put and delete method. For large data transmission, the stream action and binary serialization can be chosen to reduce the network overhead and improve the performance.

        In the view layer, we adopt a modular interface which is flexible and highly user oriented. The tree module can present the whole experiment channel in the lazy loading way and allow users to design their experiment data structure. The visualization modules are responsible for data visualization for different channels. Different visualization modules are chosen automatically for different types of data. And users can save or share the setup anytime they like because the URL for the page keeps in sync with the page content and layout. Furthermore, all the modern browsers in intelligent terminals with different size are supported.

        This data visualization framework has been deployed and integrated in LogBook, which is a web system for experiment data management and visualization. The delay is usually small and the user experience is much better than that in traditional data visualization tool used in fusion community. With this web visualization tool, the researchers can visualize and analysis the experiment data wherever the Internet covers, and can save and share their experiment data more easily and efficiently.

        Speaker: Kuanhong Wan (Huazhong University of Science and Technology)
      • 1:40 PM
        A Maxmium Current Control Strategy for Three-phase PWM Rectifier for the ITER In-Vessel Vertical Stability Coil Power Supply 2h

        The required peak current of ITER in-vessel vertical stability (VS) coil power supply is up to 80 kA, so VS coil power supply needs a PWM rectifier to achieve high power factor operation under the highly transient power demand. A new maximum current control method for three-phase PWM rectifier based on its mathematical model in d-q coordinate has been discussed. The control method samples DC-voltage of power supply and changes the set-value of current-loop controller instantaneously at different voltage values, it meets the fast-charge demand of power supply and achieves a unity or high power factor operation. The feasibility of the control method has been verified by simulation and experiment.

        Speaker: Mr Kun Qian (Institute of Plasma Physics)
      • 1:40 PM
        A Method to Alleviate the Long History Problem Encountered in Monte Carlo Simulations via Weight Window Variance Reduction 2h

        Implementing weight window (WW) is a usual method for variance reduction (VR) of Monte Carlo simulation, however as for a complex and large model simulation it frequently encounter the long histories (LH in abbreviation) problem in parallel computing. LH behavior shows as the running time of a single particle history is significantly longer than that of normal histories. It would take a disproportionate amount of time for Monte Carlo simulation to accomplish and place a detrimental effect on the efficiency of parallel computing. In this paper, the investigation of reason that causing LH was carried out firstly. A simple dog-log model was constructed to observe and analyze the LH phenomenon. Then comparative tests were carried out on a 3D model of the Chinese Fusion Engineering Testing Reactor (CFETR) with three approaches these are: a) analog running without any VR techniques; b) normal weight window VR technique; c) a novel approach proposed in this paper of limitation of weight window splitting. The results show that a suitable set of parameters in the improved WW module significantly improves the efficiency of variance reduction performance in parallel calculation, making the long history problem tractable without biasing results.

        Speaker: Dr jia li
      • 1:40 PM
        A New User Front-End for EAST Remote Participation 2h

        To provide high-efficient and low-cost way to meet international collaboration requirements for the EAST, a web-based remote participation system (RPS), details of which were first reported in the 10th IAEA Meeting in 2015, has been developed. The RPS team focused on the extension, update and optimization for the RPS during last two years. The purpose of this paper is to provide an update of the Remote Participation System in EAST Tokamak. EAST RPS has established Apache-Flex based front-end components to provide a plant-cross user interface since 2012. However, some Web browsers, such as Firefox and Microsoft Edge, disable the flash player plugin by default and the Flex technology will become less relevant in the future. The front-end migration should be a priority to update the EAST RPS. The open source, cross-platform, maintainability and life-cycle are the key features the front-end platform must have. Bootstrap which was provide by twitter, was select to be the front-end platform for EAST RPS. Bootstrap is a HTML, CSS and JS framework for developing desktop and mobile projects on the web and the solutions are offered in this paper.

        Speaker: Dr Xiaoyang Sun (Institute of Plasma Physics Chinese Academy of Sciences)
      • 1:40 PM
        A numerical model of RF ion source for the ITER-relevant NBI 2h

        With the development of magnetic confinement fusion, the new requirements and challenges are emerged for ITER NBI[1]. Briefly, It is required that the ion source of the neutral beam injection system should produce a uniform large volume high density plasma with the capability of long pulse steady state and long service life. Based on the EAST-NBI bucket ion source[2] where the main structure characteristics of large area high current ion source are introduced. In order to understand the radio frequency (RF) ion source this candidate for fusion NBI, here a numerical model of RF ion source is introduced, where the transport properties of electrons and ions are described based on the drift diffusion theory, The power coupling of RF power and plasma is analyzed, The influence of the external magnetic field on the plasma transport is also investigated.

        Speaker: Dr Xingquan Wu
      • 1:40 PM
        Advanced shape design with F2EQ code in CFETR 2h

        The Chinese Fusion Engineering Test Reactor (CFETR) is the next device in the roadmap for the realization of fusion energy in China, which aims to bridge the gaps between the fusion experiment ITER and the demonstration reactor (DEMO) [1]. CFETR will be operated in two phases: Steady-state operation and self-sufficiency with a modest fusion power of up to 200MW in Phase I, and DEMO validation with 1GW fusion power in Phase II[1]. A key challenge facing high-power steady-state operations and self-sufficiency is to maintain a fusion plasma with adequate performance while preventing damage to the vessel walls, especially divertors. Except the plasma detachment from a divertor target methods, other solutions that optimize the magnetic field topology in the divertor region have been proposed to reduce the power loads [2.3]. These advanced configurations will be a potential choice for CFETR.
        In order to optimize the advanced plasma divertor shape, a new tool dubbed the F2EQ code was developed to fit CFETR shape design. F2EQ is a MATLAB toolbox with a series of scripts and functions, which can be used to solve the fixed-boundary and free-boundary plasma equilibrium problems. It provides the flexibility to set multiple plasma configuration, especially the advanced diverted configuration, like snowflake-plus. Starting from a conventional single null configuration, second order null constraint and multiple first order null constraint on X-points was enforced to obtain the desired advanced plasma shape. In order to satisfy the different concepts and ideals in concept design phase, multiple divertor coils inside or outside the vessel with optimized currents are considered. It is easy to switch any coil on or off with any desired plasma equilibrium in different shape and other parameters, like and . Based on a series calculation with F2EQ code, the optimized plasma equilibrium with two in-vessel divertor coils with modest current in snowflake-plus configuration gives factor 4 flux expansion then the used case with two external DC coils [4].

        [1] Y.X. Wan, et al. Overview of the present progress and activities on Chinese Fusion Engineering Test Reactor 26th IAEA Fusion Energy Conference, Kyoto, Japan, 2016 (Submitted to Nuclear. Fusion)
        [2] D.D. Ryutov, et al, Phys. Plasmas, 15 (2008) 092501
        [3] P.M. Valanju, et al. Phys. Plasmas, 16 (2009) 056110
        [4] Z.P. Luo, et al. IEEE Trans. Plasma Sci. 42 (2014) 1021

        Speaker: Dr Zhengping Luo (Institute of Plasma Physics, Chinese Academy of Sciences)
      • 1:40 PM
        An Integration method of Hybrid Power Filter for Specific Harmonic Suppression in Tokamak Power System 2h

        This paper deals with an integration method of hybrid power filter in Tokamak Power System. The integration method not only takes full advantage of SVC which originally installed in Tokamak but also gives a combination of an active power filter to suppress the selective harmonics. Due to the rich harmonic spectrum in the AC side of non-linear load in Tokamak Power Systems, along with most reactive power compensation and harmonic filtering platforms in Tokamak barely notice low-order harmonics especially below the third harmonic, these low-order harmonics can be resonate with the capacitive impedance and inductive impedance in the circuit. The resonance will do harm to the grid system and the Tokamak system. Theoretical analyses and simulation results obtained from the EAST power system evaluate the effectiveness of the Integration method and stability of the whole hybrid power filter system. In addition, the simulation results are validated by experiments based on a testing platform.

        Speaker: Mr Jing Lu (ASIPP Hefei Anhui)
      • 1:40 PM
        Analysis and derivation of the EU-DEMO high level plant requirements 2h

        Ultimately, a DEMO fusion power plant must mature the physics, design, and engineering/technology basis for a future fusion power plant (FPP). In doing so, it must also enable an extrapolable assessment of the economic performance of an FPP. Following discussions with stakeholders and fusion reactor experts, these goals have been further refined into a set of high-level plant requirements. In this work we analyse the implications of these requirements, and derive further requirements which can begin to be allocated to the various plant sub-systems.

        In particular, the following requirements are discussed and analysed in more detail:

        • Targeted overall plant availability (30%)
        • Maximum shutdown duration for maintenance (250 days)
        • Tritium self-sufficiency
        • Provision oftritium for a fusion reactor beyond DEMO

        The implications of the plant availability and unplanned shutdowns on the required tritium breeding ratio and tritium start-up inventory are assessed. The requirement for tritium self-sufficiency necessarily leads to a requirement to provide a tritium stockpile buffer in the event of unforeseen shutdowns. An attempt is made to define a term for this required stockpile of tritium and the implications on the TBR are shown. The presently unconfirmed and ill-defined requirement for DEMO to provide tritium for a future FPP is discussed and attempts are made to reach a coherent and reasonable definition of this requirement. A preliminary assessment of the impact such a requirement would have on the required TBR is made. Sensitivity studies are performed on the maintenance shutdown durations to determine their impact on other high level requirements. Requirements for the overall plant availability are refined and preliminary attempts are made to sub-divide and allocate the availability budget to sub-systems in the form of a lifetime reliability target.

        Speaker: Matti Coleman (UKAEA / EUROfusion)
      • 1:40 PM
        Analysis and experimental study of impedance matching characteristic of RF ion source on neutral beam injector 2h

        The neutral beam injector (NBI) is one of the plasma heating methods on fusion device, which has highest plasma heating efficiency and the clearest heat physical mechanism. The high power ion source is one of the key parts of NBI system. Compare to the traditional hot cathode ion source, the radio frequency (RF) ion source have many merits, such as higher lifetime because of no filaments, simpler mechanical structure, lower cost due to the cheaper power supply, and power supply on ground potential due to a transformer used. It is also the reference ion source for ITER. The impedance matching is the important unit for the RF ion source, which is used to match the parameters of the RF generator and ion source antenna. It can helps to transfer the maximum RF power to the RF antennal of ion source and gets stable plasma. Due the plasma impedance will be changed before and after the plasma generation, the impedance characteristic is not easy to be calculated and measured. So, it also need more experimental study. In this paper, the impedance matching unit was analyzed and designed according to the principle of RF ion source. The matching characteristic was studied during the experiment, and got the best impedance matching characteristic. It also verified the design of impedance matching unit. Based on the results of impedance matching study, high RF power of 50 kW was coupled into the plasma and got long pulse stable plasma discharge.

        Speaker: Dr Caichao Jiang (ASIPP)
      • 1:40 PM
        Analysis of non-inductively high-performance discharges 2h

        EAST research program aims at achieving steady-state long-pulse operations, which have been obtained with fully non-inductively current drive and heating, maintaining around zero loop voltage for nearly the entire plasma current flat-top in about 60s at EAST shot #67341 recently. Based on the analysis of non-inductive current fractions, high βp is desirable in order to sustain steady-state high performance discharges on EAST in the near future. The effect of bootstrap current relates to the nonlinear component of vertical magnetic field judged by Maxwell equations. Furthermore, the quasi-linearity relationship in flat-top phase between vertical magnetic field and line-averaged plasma density lays the theoretical basis for radial compression. An increase in magnetic strength will allow high density, high beta, high bootstrap current fraction and high fusion gain to be reached, offering an attractive regime for compressed plasma to approach the Lawson parameter, especially for steady state operation of the designed CFETR — Chinese Fusion Engineering Testing Reactor. Existing limitations of EAST tokamak are analyzed for accommodating and simulating the high-performance discharges.

        Index Terms— EAST tokamak, non-inductive current drive, high beta, vertical magnetic field, compressed plasma

        Speaker: Ms Qin Hang
      • 1:40 PM
        Analysis on Phase array ultrasonic signals of the ITER PF jacket inspection 2h

        ITER magnet system consists of 18 Toroidal Field (TF) coils,a Central Solenoid (CS),6 Poloidal Field (PF) coils 9 pairs of Correction Coils (CCs). These four types of the coils are relied on Cable-In-Conduit Conductors (CICCs). The CICC for the PF Coils of ITER is referred to as PF conductors. This paper is mainly presented the Phase array ultrasonic test (PAUT) results on PF conductor jackets and analyzed the typical signals based on the statistics. Wavelet Transform (WT) method are proposed to extract signal waveform feature which will be provide reference basis for characterization of defects.

        Speaker: Ms xiaochuan Liu (ASIPP)
      • 1:40 PM
        Application of automatic ultrasonic testing system based on joint robot in Fusion Engineering 2h

        The articulated arm robot has precise mechanical link, similar to the human arm and it is integrated with the ultrasonic testing system, which can provide users with a flexible automatic ultrasonic testing scheme. The Institute of Plasma Physics, Chinese Academy of Sciences, has introduced a multi joint mechanical arm ultrasonic testing system from the French M2M company, and developed a variety of detection methods based on the system. This paper focuses on the application of the technology in ITER Feeder explosive welding composite plate, EAST W/Cu divertor tube plate weld, CFETR vacuum chamber austenitic stainless steel welding seam detection. The results of test show that the automatic ultrasonic inspection system which based on joint robot has the characteristics of flexible system, high detection efficiency and good repeatability and it will have a wide application prospect in fusion engineering.

        Speaker: Dr rui wang
      • 1:40 PM
        Application of laser-induced breakdown spectroscopy (LIBS) for in situ characterization of lithium deposition layer on EAST tokamak 2h

        Lithium wall conditioning has been a routine method to reduce fuel recycling and impurity deposition, which significantly improves the plasma performance in EAST tokamak [1]. In the 2016 EAST experimental campaign, with the help of intensive lithium wall conditioning, one-minute steady state long-pulse H-mode discharge was obtained. However, the time and amount of lithium used for the daily wall conditioning were from previous experience. There are no effective methods for in situ and real time characterizing of wall conditioning situation on the first wall, especially the thickness and the local growth rate of deposited lithium layer as well as the hydrogen isotopes retention in the lithium layer. Laser-induced breakdown spectroscopy (LIBS) is a promising candidate for in situ characterization of the first wall. Recently, an in situ and remote LIBS system has been established for the first wall condition monitoring in the EAST tokamak [2].

        In this work, the growth rate of the lithium layer was in situ and real time monitored by LIBS during the lithium wall conditioning. The results showed that the growth rate of the lithium layer was fast at the beginning of lithium conditioning and the growth rate becomes slower with time. According to post LIBS analysis in the laboratory, about 100 nm deposition layer ablated by one laser shot at the same energy density. About 2 um lithium layer was estimated deposited on the first wall by lithium wall conditioning by 200 minutes in EAST. The thickness of the coating layer showed consistency with the amount of lithium for wall conditioning. The thicknesses of lithium coating layers were measured after wall conditioning and after a whole day plasma discharge for comparison. The results showed that about 500 nm lithium deposited layer was removed by EAST plasma discharges per day. The hydrogen isotopes were measured as well. The H/(H+D) ratio in the deposited layer after lithium conditioning was lower than that after EAST discharge, which means the deuterium was saturated with the reducing of the deposited layer and D-D discharge. The investigation of lithium layer and the hydrogen isotopes by LIBS in EAST will help to optimize and predict the wall conditioning for EAST operation and demonstrate the potential using LIBS in ITER.

        [1] Wan B., J. Li, H. Guo, et al., Nucl Fusion, 2015. 55(10):104015.

        [2] Hu Z, Li C, Xiao Q, et al., Plasma Sci. Technol.,2017. 19(2): 025502.

        Speaker: Dr Zhenhua Hu
      • 1:40 PM
        Application of PAUT in CFETR vacuum vessel austenitic stainless steel welding R&D 2h

        Full penetration welding and 100% volumetric examination are required for all welds of pressure retaining parts of the CFETR(China Fusion Engineering Test Reactor) Vacuum Vessel (VV) according to the design manual. But not every welding joint can be tested with RT due to the structure and welding position. Therefore the ultrasonic testing (UT) has been selected as an alternative method.
        Considering the misjudgment and undetectable in the austenitic stainless steel welding by the traditional ultrasonic testing method, this paper introduce the application of PAUT(phase-array ultrasonic technique) in the CFETR VV R&D. Base on the ultrasonic simulation and dynamic focus, the precision of the defect position and the signal/noise(S/N) can be improved. The PAUT show excellent detectability and applicability for the austenitic stainless steel weld in the CFETR VV.

        Speaker: wang rui (University of Science and Technology of China)
      • 1:40 PM
        Application of the voltage control mode of second-generation EAST active feedback power supply 2h

        The ability of magnetic confinement to plasma can be improved by elongating plasma cross-section in EAST (Experimental Advanced Superconducting Tokamak). But elongated plasma has vertical displacement instability, without control, plasma will dash against wall of vacuum vessel and disrupt, that will cause failure of plasma discharge. So feedback control system is needed to restrain plasma vertical displacement. PCS (Plasma Control System) detects the vertical displacement of the plasma and calculates the value of signal sent to power supply, the signal is real-timely tracked and linearly amplified to generate a fast-changing magnetic field, which will suppress the vertically unstable displacement of the plasma.
        The analog control was adopted in the first-generation active feedback power supply, which worked in current tracking mode. The conventional proportional regulator which guarantees satisfactory control accuracy of the output current was adopted.
        DSP (Digital Signal Processing) was adopted as the main control chip. To achieve the maximum current’s rising rate, second-generation EAST plasma vertical displacement active feedback power supply applies voltage control mode while retaining the first-generation current tracking mode. Its average output voltage value is linear to given voltage signal. For a given voltage signal of 10V, the power supply outputs 1600V, and -1600V corresponds -10V. Compared with current mode, voltage mode achieves a significant increase in rising rate of load current. However PCS cannot be fully counted on to detect the real-time load current and change the polarity of the given signal before the current exceeding its limitation, so power supply system itself must possess perfect over-current protection function.
        The driving signal is blocked when the output current reaches the protection threshold value and resumed after falling below a certain set value. In this status, the stored energy within the load coil inductance can only be released through the inverse parallel diode of the IGBT (Insulated Gate Bipolar Transistor) to storage capacitors on DC side, which will lead to continuous increase of DC voltage. When the power supply outputs high voltage, the accompanied frequent over-current protection will return the power to DC side, which leads to over-voltage protection. To solve this problem, the current limiting control mode is adopted. The current limiting mode will stay till the polarity of input voltage signal changes.
        In 2014 EAST experiment, voltage mode was applied to plasma vertically unstable displacement by the second-generation active feedback power supply. In 52444th experiment, active feedback system exhibits great control ability to vertical displacement of plasma. Even plasma reaches vertical displacement of 4.6cm and growth rate is 530/s, the active feedback system is still able to pull it back to equilibrium position, while the first-generation active feedback power supply can only deal with 1.9cm of plasma vertical displacement and growth rate is 150/s at most.
        Through exploration of voltage mode, combination of voltage open loop control and current limiting control is present and the control effect was verified by EAST experiment, which will provide new idea to control vertical displacement of the plasma.

        Speaker: Prof. haihong huang (hefei university of technology)
      • 1:40 PM
        Application of ZD REDOX Detection Technology for Measuring Hydrogen Isotopes in Tritium Extraction System 2h

        It is limited for thermal conductivity detectors(TCDs) to measure hydrogen isotopes with gas chromatography in a large amount of helium gas environment. Then a new detection technology of ZD REDOX combined with chromatographic column was investigated in order to measure the low concentration of hydrogen isotopes in the atmosphere of tritium extraction system. The estimated detection limit for H2/D2 gas was 1 ppm in the mixed gases with 99.9% He, and the measuring precision of relative deviation was less than 5%.

        Speaker: Mr xiao chengjian (Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics)
      • 1:40 PM
        Charactarization of Low Energy Plasmas in the device PG-QRO-1 2h

        The Plasma Gun PG-QRO-1 is a Coaxial electrode plasma discharge device with Mather type geometry. This geometry has been used in the past to develop Plasma Foci, which produce a large spectrum in energy, with energies ranging up to tens of keV. The study of the interaction of magnetized plasmas with candidate materials for fusion reactors, is a main topic in fusion research. The PG-QRO-1 device has been tailored to produce plasmas with relevant densities but limiting the high energy spectrum in order to use it for plasma-wall-interaction studies. We present here the study of plasmas of low energy produced with this device. The energy profile of the plasma is determined form the depth profile of samples of different materials exposed to deuterium discharge. The deuterium retention profiles in the materials are very shallow with penetration depths of the order of tens of nm.

        Speaker: Dr Gonzalo Ramos (Instituto Politecnico Nacional)
      • 1:40 PM
        Comparison of Deformation Models of Flexible Manipulator Joints for use in DEMO 2h

        A hybrid kinematic manipulator (HKM) is being designed at RACE (Remote Applications in Challenging Environments) to handle the large breeder blanket segments for DEMO. The payload of this HKM is around 80 tonnes, and its trajectory requires stringent position accuracy as it passes key points, in order to manoeuvre the blanket into and out of position in the vacuum vessel. The TARM (Telescopic Articulated Remote Mast) at RACE is also under upgrading, and it is necessary to investigate its’s deformation displacement due to its massive weight and the payload.

        From the past experience of heavy duty robotic machines, it is noticed that deformation of the manipulator joints contribute significantly to the end-effector displacement. In order to compensate such end-effector deformation displacement in the control system, it is necessary to develop computation-effective deformation model of the flexible joints. In addition the deformation model can be further utilized to optimize the end-effector trajectory by using the iterative algorithms.

        In order to support the large payload, the joints of the manipulator are complex, making it unreasonable to employ the truss and beam simplifications from the structural mechanics. The finite element analysis (FEA) method can estimate the deformation of a complex structure with high accuracy given the payload, however, its computation consumption makes it prohibitive to apply to the control system and in the iterative algorithms.

        The paper proposes two approaches to model the joint deformations: a non-parametric ANN (artificial neural network) model and a parametric model using the Bayesian Markov Monte Carlo method. Both models are trained and identified off-line using a basic dataset from the FEA of the target joints. After the models are well established, they can be used in the control system or iterative optimization algorithms in real-time. In practice, the proposed methods can also be carried out to model the deformation of joints incorporating the transmission mechanisms, based on real on-site measurement data.

        The comparative results of applying proposed deformation models on different joints are presented in the paper. The validation of the non-parametric ANN model and the stochastic process based parametric model are conducted, individually, by comparing with the results of applying the FEA on several joints of HKM and TARM. The study can provide a good premise for constructing the entire computation-effective deformation model of manipulators that will be employed in the DEMO.

        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 Ming Li (Lappeenranta University of Technology)
      • 1:40 PM
        Comparison of radiative divertor behavior in Ar and Ne seeded plasmas in EAST 2h

        In ITER and future fusion devices, a high radiation level for power exhaust will be mandatory to avoid thermal overload of divertor targets. Increasing divertor radiation by injecting impurities is a general and effective method to reduce scrape-off layer heat flux and to cool the divertor plasma to detachment. Impurities such as nitrogen (N2), neon (Ne) and argon (Ar) have been widely used in radiative divertor experiments on several tokamaks. Last two years, Ar and Ne impurities were seeded respectively as the radiator from EAST upper divertor which upgraded into ITER-like full tungsten PFCs in 2014 to investigate their effects to plasma behavior, especially in the divertor region.
        According to the cooling factor of Ar and Ne, which is closely associated with electron temperature, mixture of Ar/D2 was firstly seeded from the upper divertor region as a radiator. To compare with Ar impurity, then experiments under similar plasma parameters’ condition and the same gas puff position with Ne seeding, including pure Ne and Ne/D2 mixture, were carried in 2016 campaign. In this work, both Ar and Ne impurity showed the high efficiency in reducing particle flux and heat load on divertor targets. After impurity seeding, saturation ion current, Is, electron temperature, Te, and heat flux on divertor target, qt, decreased rapidly. In this case, the inner divertor first entered the detached state and the outer divertor followed the inner one soon. However, these two impurities showed clearly different radiation behavior. Compared with Ar impurity, the rise of radiation in Ne seeded plasma more located in the divertor region. It was more difficult for Ne to enter the plasma core region than Ar because the former belongs to a kind of low-Z impurity and has a lower cooling factor in the core region. After the gas puffing was terminated, it took 1~2s for the rise of radiation caused by Ar impurity to gradually drop down to the initial state, while the radiation after Ne seeding remained in a rising state until plasma burned out. The reason may be that Ar impurity quickly ionized but Ne impurity stayed in fluctuated ionization-recombination state due to the cooled plasma near the divertor target region where the low electron temperature as low as below 8 eV. With regard to impurities, there were notable increases and decreases of Li, C and tungsten impurity after the Ar impurity was seeded. These impurities, observed by the divertor impurity spectroscopy, dominated over all other kinds of impurity. However, these impurities presented a relatively low level in Ne seeded plasma. Therefore, it is indicated that using Ne as radiator preferentially in controlling PWI issue in radiative divertor experiments.
        In addition, through Supersonic Molecular Beam Injection (SMBI) and divertor piezo valve collaborative control, we obtained the active feedback controlled radiative divertor operation last December. In this case, the rate of radiation loss, frad, could reach 40%, which is of great significance for the goal of long pulse high performance operations in EAST.

        Speaker: Mr Jingbo Chen (Institute of Plasma Physics, Chinese Academy of Sciences)
      • 1:40 PM
        Computational study of the elastic modulus of mixed pebble beds for WCSB 2h

        Primary concept design of CFETR (Chinese Fusion Engineering Test Reactor) has finished. In the fusion reactor, tritium is bred by mixed pebble bed in CFETR’s WCSB (Water Cooled Solid Blanket). In this study, the discrete element method (DEM) was used to study the mechanical behaviors and elastic moduli of mixed pebble beds. The effect of cyclic pressure p within the granular system in different sizes were considered. Besides that, we re-confirmed the nonlinear elastic stress-strain relation, the much lower elastic moduli of a granular system than that of solid materials and the faster growth of moduli than the p1/3 law predicted by the effective medium theory (EMT). We also observed that the cyclic pressure (mechanical excitation) would stiffen the granular system, but this effect would be smaller as the cycle number increases. This indicates that the stability of system could be stronger under this effect. Besides that, the subtle difference in grain sizes was observed to soften the system although it caused a little higher packing density. Furthermore, although the effective moduli of different granular materials Es are diverse, they were found to nearly collapse to the different distribution when both Es and p are non-dimensionalized by particle moduli Ep, and the mixed pebble bed cannot obey the EMT theory.

        Speaker: Mr Yuanjie Li (University of Schience and Technology of China)
      • 1:40 PM
        Cooling Needs and Thermal Hydraulic Design Studies of Diagnostic Shielding Module of US ITER Port Plugs 2h

        ITER machine will install a set of 45 diagnostics to ensure controlled plasma operation. Many of them are positioned in the upper & equatorial ports. US ITER diagnostics scope includes the design and integration of 2 equatorial port plugs (E03/09) and 2 upper ports (U11/U14). Each port contains three different zones starting from the in-vessel: the port plug zone, the interspace zone and the port cell zone. The diagnostic components in the port plug zone are installed to a large metallic structure assemblies, called diagnostic port plug, consists of three components: Diagnostic First Wall (DFW), Diagnostic Shielding Modules (DSM) and port Plug Structure (PPS). The DFW protects the diagnostic components from plasma neutron and radiation and provides the diagnostic apertures to peer into the plasma. The DSM is designed to support the DFW structures providing neutron shielding together with the DFW. Therefore, the DSM design will cope with the design drive loads from the harsh thermal and electromagnetic environment, especially in the front end. The water channel within the DSM will be designed to allow sufficient cooling during normal operation and for heating during bake-out. The DSMs and its tenant diagnostic systems require the well-distributed balance to limit the maximum temperature range and gradients of various interfaces to ensure the structural integrity. Despite of the challenging design constraints due to various interface requirements, to obtain the optimized cooling water mass flow rates and thermal hydraulic performance will be particularly investigated during the port integration. This paper highlights the study of the cooling needs and thermal hydraulic design for the DSM as one of the design engineering and integration tasks of the US ITER ports.

        *This work is supported by US DOE Contract No. DE-AC02-09CH11466. All US activities are managed by the US ITER Project Office, hosted by Oak Ridge National Laboratory with partner labs Princeton Plasma Physics Laboratory and Savannah River National Laboratory. The project is being accomplished through a collaboration of DOE Laboratories, universities and industry. The views and opinions expressed herein do not necessarily reflect those of the ITER Organization.

        Speaker: Dr Yuhu Zhai (Princeton Plasma Physics Laboratory)
      • 1:40 PM
        Curent Status Concerning Tritium Removal Technology and its Implementation at Cernavoda NPP(ROMANIA) 2h

        In CANDU fission reactors and also in fusion reactors, tritium should be recovered from large amounts of effluents for environmental and staff protection, for safety and for various applications.
        The combined cryogenic distillation(CD) with catalyzed isotopic exchange between deuterium and liquid tritiated water(LPCE) it’s one of the most suitable technology for removal and its recovery.
        For LPCE process, the key issue and driver force consists of in a very efficient and stable contact element which has to work in direct contact with liquid and vapor water for long time with high separation performances. In order to check and to prove CD-LPCE technology for tritium removal from tritiated heavy water from Cernavoda CANDU Power Plant, an experimental pilot plant for tritium removal (ExpTRF), has been built at ICSI Rm-Valcea and tested within comprehensive program.
        Based on the authors’ experiments and results, the present paper presents the current status and key aspects of activities concerning the operation of Tritium Removal Technology and its implementation at Cernavoda NPP in Romania. A comparison between present ExpTRF and the future Industrial Tritium Removal Facility (IndTRF) is shown and discussed.
        The paper presents also a critical analysis on main contact elements used in LPCE module. The critical analysis it’s focused on:
        - selected types of hydrophobic catalysts and hydrophilic packing;
        - methods and conditions for manufacture;
        - key aspects in operation of TRF
        - improvement of the performances of the proposed catalysts for industrial nuclear applications;
        - extrapolation of research results at industrial scale;
        As result, a new improved contact element, more compact, has been developed and it’s still under testing at ICSI Rm-Valcea. This new improved contact element has been selected to equipped the LPCE column within Industrial Pilot Plant for Tritium Removal Facility at Cernavoda NPP.
        This new improved contact element could be an option in the process of selection of catalytic mixed packing for Water Detritiation System(WDS) and Isotopic Separation System (ISS) from the ITER reactor.

        Speaker: Dr Gheorghe Ionita (ICSI Rm-Valcea)
      • 1:40 PM
        Degradation of Neutral Beam heating & current drive by Alfvénic instabilities 2h

        Neutral beam injection in tokamaks results in a population of energetic particles (EP) that can drive instabilities in the Alfvén frequency range. In turn, instabilities can lead to redistribution or loss of EPs, thus affecting the controllability and predictability of quantities such as neutral beam (NB) current drive efficiency and radial profile of the non-inductive current fraction. In this work, examples from NSTX and NSTX-U discharges featuring robust Alfvénic activity are discussed to investigate the reduction of NB current drive by instabilities. Recent improvements to the tokamak transport code TRANSP enable quantitative, time-dependent simulations of NB-heated plasmas in the presence of EP-driven instabilities. In particular, a new physics-based model has been implemented in TRANSP to account for the resonant interaction between EPs and instabilities, which results in more reliable simulations than previously achieved using a simple, ad-hoc diffusive model. Results show that instabilities can strongly affect the 
EP distribution function. Modifications with respect to ‘classical’ EP behavior (i.e., in the absence of instabilities) propagate to macroscopic quantities such as the profiles of NB-driven current and of the local EP power transferred to the thermal plasma species through thermalization. For scenarios with multiple unstable EP-driven instabilities, the computed reduction in NB current drive efficiency can be as high as 40% with respect to classical simulations.

        Speaker: Mario Podesta (Princeton Plasma Physics Laboratory)
      • 1:40 PM
        Design and Analysis of CFETR CSMC Cooling Loop 2h

        The Central Solenoid Model Coil (CSMC) of China Fusion Energy Test Reactor (CFETR) is currently in the design and manufacture process. CSMC assembly consists of the winding pack, an outer NbTi coil,a middle Nb3Sn coil, an inner Nb3Sn coil and a pre-load structure. The highest field of the model coil is 12T, while the highest change rate of magnetic field of the conductor is 1.5T/s. Due to the AC losses during charging, a huge heat load will be produced in the model coil.In order to make the coil work properly in normal condition, a well-designed and precisely-analyzed cooling loop plays an important role.
        In this paper, the design of the cooling loops is based on the calculation results of the AC losses deposited on the model coil. The length of the cooling channels, together with the thermo-hydraulic parameters such as inlet pressure, temperature, mass flow rate are optimized.In addition,thermal hydraulic analysis for the cooling loop located in the worst condition of the model coil was conducted to recognize the temperature and mass flow rate change over time. The hydraulic model, the material properties and the heat loads involved in the analysis are given, and the results of the analysis are presented.

        Speaker: Dr Qiangwang Hao (ASIPP)
      • 1:40 PM
        Design and Analysis of “Filling-Evacuating” High-Pressure Helium-Cooled Loop 2h

        The breeder blanket and divertor are crucial plasma facing components (PFC) in a fusion reactor. The helium cooled blanket and divertor concepts have exhibited the best potential to come up to the highest safety requirements and therefore been chosen for the development object. As a result of high heat flux radiated from the plasma in the fusion reactor and high power density nuclear heat deposited by high-energy neutrons, the cooling of the First Wall (FW) and the discharge of nuclear heat have become one of the major technical challenges. To demonstrate and verify the helium-cooled technology and tools of China Test Blanket Module (TBM), and explore the feasibility and key technology of thermal hydraulics process of helium-cooled divertor, we have creatively adopted the “filling-evacuating” approach to design and fabricate the High-Pressure Helium-Cooled Loop (HPHCL), in which a mock-up of reduced-scale helium-cooled blanket module is designed and manufactured as a test section. Based on different experimental cases, the operating pressures of helium at mock-up range from 3 to 10 MPa and the maximum mass flow rate can reach up to 0.21Kg/s. In this paper, the design scheme of the HPHCL is presented, and the key issues of engineering manufacture and the test cases are calculated and analyzed. The helium gas flow and the heat transfer are calculated according to the test working conditions of the referenced ITER TBM’s FW surface heat flux, using ANSYS fluid dynamics software FLUENT. The results will provide support for the follow-up fabrication of test system and implementation of the tests.

        Speaker: Dr Haifei Deng (Institute of Plasma Physics, Chinese Academy of Sciences)
      • 1:40 PM
        Design and Analysis Progress of US ITER Integrated Diagnostic Equatorial Port 09 2h

        ITER is the world’s largest fusion device currently under construction in the South of France with over 50 diagnostic systems to be installed inside the port plugs (PPs), the interspace or the port cell region of various diagnostic ports. The Diagnostic First Wall (DFW) and Diagnostic Shielding Modules (DSM) are designed to protect front-end diagnostics from plasma neutron and radiation while providing apertures for diagnostic viewing access to the plasma. Three tenant diagnostic systems will be integrated into the equatorial port plug 09 (E09). The toroidal interferometer and polarimeter, or TIP system, is installed in the left drawer (DSM3) for measuring the plasma density so to control fuel inputs. The electron cyclotron emission (ECE) system is installed in the middle drawer (DSM2) to provide high spatial and temporal resolution measurements of the electron temperature evolution and electron thermal transport inferences. The visible/infrared wide angle viewing system will be installed in the right drawer (DSM1, right looking from plasma) to provide visible and IR viewing and temperature data of the first wall for its protection in support of the machine operation.

        The PP engineering design and multi-physics analysis has been performed following ITER port integration requirements including weight limit (45 tons total), neutron shielding (100 uSv/hr total dose limit), cooling layout and structural integrity validation. Mass distribution for the TIP and ECE DSMs has been optimized to meet the weight limit by the new design of B4C shielding pockets. The lightened DSM maintains its front-end EM load distribution with better protection of on-board diagnostics; while still provides sufficient front-end stiffness for structural integrity. To moderate impact from VDE inertial loads due to the Vacuum Vessel (VV) movements during asymmetric plasma Vertical Displacement Events (VDEs), the rigid lock-in DSM to PP structure interface was implemented into the E09 port integration analysis models for design validation. The structural integrity of E09 PP assembly is largely driven by the electromagnetic loads induced on the metallic structural components during plasma disruptions. The in-port diagnostics and their mounting supports, on the other hand, are largely driven by the steady-state thermal loads from volumetric nuclear heating, and the dynamic response of components attached to the DSM-PP structure assembly under the VDE inertial loads. Progress on the E09 integrated design and analysis is reported. The tenant interface load transfer is also presented in details for in-port system attached to the DSMs as part of the design and analysis tasks for ITER PP engineering.

        *This work is supported by US DOE Contract No. DE-AC02-09CH11466. All US activities are managed by the US ITER Project Office, hosted by Oak Ridge National Laboratory with partner labs Princeton Plasma Physics Laboratory and Savannah River National Laboratory. The project is being accomplished through a collaboration of DOE Laboratories, universities and industry.

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

        Speaker: Dr Yuhu Zhai (Princeton Plasma Physics Laboratory)
      • 1:40 PM
        Design and Fabrication Process of Toroidal Field Coil for HL-2M 2h

        HL-2M is a medium-sized copper-conductor tokamak under construction at the Southwestern Institute of Physics (SWIP). The designed plasma parameters are as follows, plasma current=3 MA, toroidal field = 3 T, major radius =1.78 m, minor radius = 0.65 m, flux-swing > 14 Vs, plasma pulse ~ 5 s, with a plasma shape of elongation = 2 and triagularity > 0.5. The demountable toroidal field coils structure has been selected in order to removal of the vacuum vessel and poloidal coils integrally. The toroidal field coil is designed for a maximum field of 3 T at R = 1.78 m and consists of 140 turns (20 bundles, each bundle 7 turns) with a maximum current of 191 kA. Under normal operation condition, the current is 140 kA and the field is 2.2 T. Each turn is composed of inner, upper and outer arc part which are connected by finger joint and bolted joint. The report introduces the structure of toroidal field coils, selection and fabrication process of conductor materials for toroidal field coils, the parameters of conductor materials such as the hardness, electrical conductivity, yield strength, tensile strength, and fatigue strength firstly. Then introduce fabrication technology for toroidal field coils, such as welding cooling water pipe technique, turn to turn insulation technique, prestressed epoxy-glass cylinder technology.

        Speaker: Mr Yin Qiu
      • 1:40 PM
        Design and implement of Varying Frequency Three-phase Synchronous Signal processing system Based on modern signal processing 2h

        In HL-2M the magnetic field power supply includes CS power supply and sixteen poloidal field power supplies. Each power supply is consists of three-phase full bridge thyristor converters and phase control is used to fire thyristor. The AC power of magnetic field power supply is provided by a six-phase motor generator with two Y windings of shifting 30°. The AC voltage waveform is distorted due to heavy loads, and the frequency of generator outputs is changed during the pulse of plasma shot. So it is difficult to obtain clean and precise synchronous voltage for thyristor firing system. In order to enhance control precision and reliability of power supply, a new three-phase synchronous signal process test platform is developed. The simulation results in test platform show that the method is feasible. Then, the new synchronous processing system is founded. Through real-time data acquisition system three-phase synchronous AC signal are inputted. Digital filter technology is used to deal with input signal and according real-time frequency and phase bias compensation phase is realized by FPGA. And different phase of synchronous signal can be gained through this new system. The experimental results show that the three-phase synchronous signal processing system meets the design requirement.

        Speaker: Dr Weibin Li (SWIP)
      • 1:40 PM
        Design and Installation of Small Angle Slot (SAS) Divertor in DIII-D 2h

        Divertor solutions to efficiently disperse heat from fusion reactors are critical because the maximum steady-state power load is limited to qt ≤ 5-10 MW/m2 to the divertor target. This may pose a special challenge for next-step Advanced Tokamaks (AT), which will have lower plasma density than ITER for high performance long pulse or high duty cycle operations. A new Small Angle Slot (SAS) divertor concept has been developed to address this critical issue. The SOLPS-EIRENE edge code analysis shows that a SAS divertor can achieve strongly dissipative/detached divertor plasmas at a significantly lower upstream plasma density, thus potentially providing a power handling solution for long pulse ATs.
        During the vent of the DIII-D vessel in late 2016, a graphite tile SAS divertor was installed. The design of the SAS divertor enables us to test the new slot divertor without affecting the geometry of the existing pumped divertor that is used for high performance advanced tokamak research. The new divertor tiles are mounted to an existing water cooled baffle structure that presently serves as the support structure for the graphite armor tiles for the pumped divertor region. The profile of the new tiles includes a narrow slot that is located outboard of the existing divertor target and divertor pump entrance and does not have any pumping.
        Material for the SAS tiles was chosen to be Graftech XTJ-15. XTJ-15 graphite is an isotropic graphite material which is Graftech’s replacement for ATJ. ATJ is the material that is primarily used in DIII-D for the graphite armor tiles, but is no longer produced.
        This new SAS divertor has been operationally tested during the 2017 DIII-D physics campaign. Special design considerations were required to include Langmuir probes and thermocouples in the slot region of the new divertor.
        Details of the tile design, modelling and installation will be presented. Work supported by the U.S. DOE under DE-FC02-04ER54698.

        Speaker: Chris Murphy (General Atomics DIII-D)
      • 1:40 PM

        Wendelstein 7-X (W7X) is a large, superconducting stellarator with modular coils and an optimized magnetic field.
        A multi-purpose manipulator (MPM) system has been developed and installed on the W7-X vessel, aimed at investigating the edge plasmas of the stellarator. It is a flexible tool for integration of a variety of different diagnostics as e. g. electrical probes, probing magnetic coils, material collection, or material exposition probes, and gas injection. The system is designed as user facility for many diagnostics, which can be mounted on a unique interface without breaking the W7-X vacuum. The manipulator system, located in the equatorial plane, transports the inserted diagnostic probe to the edge of the inner vacuum vessel. From there the probe can be moved over a maximum distance of 350 mm to different positions inside the plasma with a maximum acceleration and deceleration of 30 m/s2.

        In the framework of the EUROfusion S1 work program for the preparation and exploitation of W7-X campaigns, a diagnostic insertable probe head called HRP (High Resolution Probe) was developed by Consorzio RFX in collaboration with IPP Greifswald, to study the electrostatic and electromagnetic features of turbulence in the edge region of W7-X using the MPM. In particular the aim of the HRP head is to provide information on parallel current density associated to L-mode filamentary turbulent structures as well as on ELMy structures in H-mode. Furthermore the possibility to measure the time evolution of radial profiles of flow was considered as a further interesting part of the study, given the strong interplay expected between the turbulent fluctuation and the average flows.

        The paper reports the design development of the HRP head, from the choice of the sensors to the engineering design. The assumptions and evaluations supporting the main design choices, together with the R&D tests carried out to check the most critical parts, are described in detail.

        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: Piero Agostinetti (Consorzio RFX)
      • 1:40 PM
        Design of a robust linear and rotary sensor compatible with hostile environmental conditions 2h

        Large machines for fusion research and the fusion rectors have hostile environmental conditions characterised by high level of radiation, vacuum and temperature both in the vessel and in the port cell and the close area, commercial displacement sensors like linear and rotary encoders and their electronics are not suitable with this conditions. This problem has been approached and solved by developing an optical and optoelectronic concept for the reading section of rotary and linear encoders able to resist to the hard environmental conditions. In this concept the driving electronics of the system is placed far from the hostile environment and the connections to the sensor are realised by means of suitable optical fibers. Some of these devices was developed by us to instrument more complex systems, they have been also validated and tested in vacuum, temperature and radiation over long periods.
        The paper describe a recent activity of design and simulation that has been performed in order to overcame the sole critical aspect of such methodology that is related to the very close reading distance between the static part and moving part of the encoder. The new design increases the reading distance of many times to relax the requisites of the associated mechanics thus making the system more reliable in case of large temperature gradients and levels. An upgraded concept of the optical and optoelectronics reading section has been conceived; being based on micro optics and micromechanics, before the implementation we have chosen to develop a custom 3D simulation tool both to validate the concept and to optimize the micro optical and micromechanical design. The tool is based on ray tracing algorithms developed using the 3D vector geometry. It was implemented using Mathematica and an optimization phase has been necessary to be able to perform simulations of hundreds of thousands of rays in useful time. Different choices in the design has been simulated in order to perform a trade-off analysis based on criteria of robustness, low cost, commercial availability of the components and depending on the environmental conditions. A brief description of the ray tracing algorithm is presented, followed by the numerical results obtained for the different choices of the configuration that have been evaluated in the trade-off analysis. The optimal design chosen for the implementation is discussed and presented in the paper and the expected characteristics are reported. An analysis of the possibility of increasing the resolution and the accuracy of the sensor by means of inter-mark interpolation is included. An outline of the driving electronics and its main characteristics is also presented showing that the new design optimizes also this part. The design is presented with the aim to show a suitable solution for displacement and rotary sensors, which can be adopted in diagnostics and/or in remote handling systems of large fusion machines or in other fields where similar hostile environmental conditions are present.

        Speaker: Dr Carlo Neri (ENEA)
      • 1:40 PM

        Tearing Mode (TM) creates magnetic islands in the tokamak. Using external resonant magnetic perturbation (RMP) coils is a convenient method to affect magnetic islands. To avoid mode locking and major disruption, the stabilization of TM control by using RMP is a promising method. A new method for applying modulated magnetic perturbation is presented to suppress magnetic island and accelerate island rotation. The phase difference between TM and external RMP is denoted by Φ. RMP has a stabilizing (destabilizing) effect on island when 0.5π<ϕ<1.5π (-0.5π<ϕ<0.5π) and an accelerating (decelerating) effect when π<ϕ<2π (0<ϕ<π). Moreover, a net suppression effect has been proved by numerical simulation result when π<ϕ<2π. Based on this mechanism, if RMP is applied to the phase region of π<ϕ<2π, magnetic island can be suppressed and accelerated in every island rotation period.
        J-TEXT tokamak has a set of RMP system which contains four sets of in-vessel saddle coils. To achieve the mechanism above, a bipolar current-pulse power supply with magnetic island phase detected system is applied to TM control. In the phase region of π<ϕ<2π, the power supply gives positive current-pulse to accelerate island, and in the region of 0<ϕ<π, it gives negative current-pulse to double the effect. The island phase detection should be accurate and current-pulse power supply should have rapid current changing edges to have expected effect. In this paper, the working principle of the current-pulsed power supply is elaborated. The power supply contains a H-bridge inverter using IGBTs to provide high power high frequency bipolar current for inductive load. A six-pulse rectifier with a LC filter is used for DC source. A DC/DC chopper is added on bus to have a faster response of adjusting load current amplitude. Before the H-bridge, a set of boost capacitors with a diode is designed to steepen current changing edges. It will store the energy of inductive load on current falling edge and boost voltage on current rising edge. To ensure phase region accurate, current edge changing should be less than 100 us. The current frequency should follow the TM frequency changing from 1 kHz to 7 kHz and amplitude should be 3 kA in maximum. Based on calculation and simulation results, the capacitance should be suitable to keep the balance between current changing speed and capacitors voltage. A power supply prototype has been made to obtain experiment results. Because of the leakage inductance on bus, there is a voltage spike at the IGBT turn-off moment. A snubber circuit is designed for inverter to reduce the voltage spike.

        Speaker: Dr Li Mao (Huazhong University of Science and Technology)
      • 1:40 PM
        Design of the optical emission spectroscopy diagnostic system and preliminary experimental results in RF negative ion source 2h

        The development of radio frequency (RF) negative ion sources for neutral beam systems requires knowledge of the plasma parameters. Optical emission spectroscopy (OES) is a non-invasive and in situ diagnostic tool, so optical emission spectroscopy diagnostic system are designed to be applied to the measurements of the RF negative ion source, and diagnostic principle and simplified analysis methods for plasma parameters are introduced. A preliminary results of a variety of plasma parameters are obtained based on the part of the local thermodynamic equilibrium (PLTE) state. When the discharge power is 25kW and the discharge operates pressure is 0.5Pa, the electron temperature is about 0.83eV and the positive hydrogen ion density is 2.7×1018/m3.

        Speaker: yan wang
      • 1:40 PM
        Development of Rotational Speed Control Equipment And Brake Equipment for 300MVA Pulse Generator 2h

        For supplying enough power for 2M-HL Tokamak, a new 300MVA pulse generator has been developed, the new generator with 400 tons of rotor to stored energy will be driven by an 8500kW asynchronous motor . In order to reduce the large starting current, a high voltage variable frequency converter has been developed as the starting device because of the large inertia. Liquid resistors in series with motor rotor as the standby starting equipment has been developed. Two sets of equipment start the generator through the switch . In this paper, a simulation model of the high voltage variable frequency converter is built by the MATLAB/SMULINK. Calculation are made for motor rotor series of liquid resistors. The maximum series resistors , the starting current and starting time are obtained.
        The working speed of 300MVA generator is 500RPM. It costs more than one hour to stop freely the unit. So a energy-consumed braking equipment and mechanical braking equipment are developed. These two equipments are analyzed in this paper. The braking resistor, excitation current and braking time are calculated. the mechanical brake pressure and brake time are calculated also.
        During the debugging the unit, the actual running data and calculation data are compared. The analysis and calculation are more conform to the actual running situation.

        Speaker: Mr Haibing Wang
      • 1:40 PM
        Development of Neural-Network Potentials for Atomistic Modelling of PWI 2h

        Artificial neural-networks (NN) have been used to model the potential-energy surfaces of, for example, bulk silicon or copper surfaces. NNs may reach the structural and energetic quality of density-functional theory (DFT) at a small computational cost [1][2]. The authors intend to develop NN potentials based on ab initio data as an alternative approach to empirical potentials for the atomistic modeling of plasma wall interaction processes (PWI).
        At the stage of the ‘training’ process, we obtain data for Be-W surfaces and small molecular clusters of BenWm, BenHm, WnHm with n + m ≤ 4 including also the pure species with m = 0.

        In the present contribution, we focus on a comparison of quantum chemical methods for BenWm, BenHm, WnHm species. Second order perturbation theory (MP2) and coupled cluster CCSD(T) theory are compared with plane-wave and atomic orbital DFT calculations, and with available experimental data. The plane-wave calculations were carried out with the VASP code using the PBE functional for the description of exchange and correlation of valance electrons, and the projector augmented wave approximation for inner shell electrons. The atomic orbital calculations have been performed with the Gaussian code, where we employ the double hybrid B2PLYD3 and the dispersion corrected ωB97X-D functional and also compare the performance of various ECP basis sets. With preliminary results and the knowledge from literature, we can already conclude that a single method will probably not be sufficient to deal with molecular clusters and surfaces alike.

        [1] Jorg Behler and Michele Parrinello, PRL 98, 146401 (2007)
        [2] Nongnuch A., Behler J, PHYSICAL REVIEW B 85, 045439 (2012)

        Speaker: Mrs Lei Chen (University of Innsbruck, Austria)
      • 1:40 PM

        The ITER Tokamak, designed to study deuterium-tritium fusion reactions and to demonstrate its viability as a sustainable and clean energy source, is currently being built in South France, on the Cadarache site. Its vacuum system, one of the largest and most complex vacuum systems ever to be built, requires several hundred vacuum sensors for pressure monitoring of its high vacuum systems.

        High vacuum gauges operating under magnetic fields as high as 300mT, with gamma radiation in excess of 1MGy and significant neutron fluency, will be necessary in order to fulfill the pressure measurements of the torus vessel, neutral beam injectors, cryostat vessel, diagnostics, cryogenic distribution and heating systems of the ITER Tokamak.
        Following an international call for tender and the signature of a Strategic Agreement between ITER and the company INFICON, specific pirani and cold cathode gauges with remote controller have been developed and qualified to operate under the difficult ITER environment.

        In this paper the ITER specific environmental conditions and requirements for pressure measurements are reminded. The standardization process for ITER passive vacuum gauges and controllers is then described and emphasis is given toward the products development and qualification testing. Final gauges performances are then detailed and successfully commercialized ITER standard products are lastly exposed.

        To complete the picture, highlight is given on additional vacuum sensing development required to complete the ITER vacuum instrumentation portfolio and achieve an operationally safe design.

        Speaker: Bastien Boussier (ITER Organization)
      • 1:40 PM
        Diversification of the position sensing instrumentation for the JET neutral beam calorimeters 2h

        The JET neutral beam injection system incorporates a calorimeter in each beamline, comprising 2 large cooled copper panels (about 2.5m x 1m) instrumented with thermocouples to provide diagnosis of the beam shape and alignment. The panels are rotated out of the beam path to allow the beam to enter the torus; they are inertially cooled and can only sustain full beam power for a fraction of a second, hence it is essential they are fully withdrawn during plasma operation.

        Calorimeter position is monitored with in-vacuum micro-switches close to the limits of travel, but these have proved unreliable in the past; furthermore the panels are known to twist due to a combination of bearing friction, water bellows reaction torque and actuation from the top and as a result may not always reach the switches. This has led to periods of operation where the bottom of the panel has unknowingly scraped the edge of the beam and in 2013 this resulted in melting of the edge of one panel and a large water leak.

        A procedure has been implemented to check the calorimeter position and thus avoid a repeat of the melting incident; however in 2015 an independent review panel examined NBI reliability and recommended that a diversity of methods should be used to detect the positions of the calorimeters. This paper summarises the methods considered and details the option selected for installation during the 2017 shutdown.

        Key constraints on the technology choice were:

        • Compatibility with ultra-high vacuum
        • Presence of sputtered copper
        • High neutron level (particularly during the planned D-T operation)
          meaning no active electronics close to the sensor
        • Magnetic fields during pulsing
        • High levels of vibration
        • Measurement accuracy better than 5mm

        Many technologies were considered, the most promising being:

        • Rows of mechanical or magnetically actuated reed switches, to
          indicate a series of discrete positions
        • Bespoke inductive proximity sensor
        • A spring element deflected by the calorimeter movement, instrumented
          with strain gauges

        The latter 2 were investigated in more detail through laboratory experiments and both considered suitable, however the spring element was finally selected on the basis of being considered lower risk.

        The detailed design of the final sensor is described, along with the laboratory work on an inductive sensor; this technique was only rejected on the basis of requiring more development work and hence presented a higher risk given the limited time available to design and manufacture a sensor. It may have applicability to other in-vacuum position sensing requirements where more development time and resources are available.

        This work has been carried out within the framework of the Contract for the Operation of the JET Facilities and has received funding from the European Union’s Horizon 2020 research and innovation programme. The views and opinions expressed herein do not necessarily reflect those of the European Commission.

        Speaker: Mr Peter Blatchford (Culham Centre for Fusion Energy)
      • 1:40 PM
        Electrical and Magnetic Analyses and Design of New NSTX-U PF1A Coil 2h

        Abstract – The PF1A coil is one of the Poloidal Field (PF) shaping coils on the NSTX-U machine. It is critical for shaping highly elongated, and high triangularity plasmas current. In July 2016, the failure of the PF1A Upper coil resulted in shutdown of the NSTX-U experiment. As part of the causal analysis, it was discovered that several passive structures around the PF1A coil had an adverse effect on the electrical and magnetic behavior of the coil system under AC conditions, more than was expected. This effect, although not a direct cause of the failure, significantly increased the harmonic ripple in the coil current as well as the plasma current beyond the design target, and also caused problems with the magnetic diagnostics. Therefore, several analyses were conducted to understand the electrical and magnetic behavior of the coil system at under AC conditions and to account for it in the new design.
        A finite element analysis was first performed to map the magnetic field around the coil and capture the eddy current and magnetic coupling effect of the surrounding passive structures on the effective coil resistance and inductance over a range of AC frequencies. The calculated AC impedance is first compared to field measurement, and then the resistance and inductance, at the characteristic power supply frequency, is put into a detailed electrical model, which includes a detailed representation of the power supply system and electrical network, to simulate the electrical behavior during operation. Results from the electrical simulation were then compared to operational records for verification.
        Using results from these analyses, the new PF1A coils have been designed, which includes an external reactor to account for the passive structure effect. Same methodology can also be applied to design of the other new PF coils.

        Key Words: NSTX-U, PF Coil, Magnetic Coupling, Eddy Current, Thyristor based power supply, effective resistance, effective inductance

        Speaker: Zhi Gao (pppl)
      • 1:40 PM
        Electromagnetic Analysis of the ITER Glow Discharge Cleaning Electrode in Equatorial Port No.12 2h

        Glow discharge cleaning (GDC) shall be used on ITER device to reduce and control impurity and hydrogenic fuel out-gassing from in-vessel plasma facing components. After first plasma, permanent electrodes (PEs) will be used to replace Temporary Electrodes (TEs) for subsequent operation. These PEs will be used inside vacuum vessel and during plasma operation, major disruption and vertical displacement events may cause huge electromagnetic forces on these PEs and destroy them. Especially for equatorial ports, the PE is closer to plasma, so analysis has to be done to understand the specification of these EM loads, and these loads will be used as input for structural analysis. This report presents results of transient electromagnetic analysis of electrode structural components during plasma disruption for the seven cases. The calculation is based on the ITER Global Model (IGM) for the EM analysis of vessel components. The method is inserting the model of GDC into the IGM, which provides a certain region for the GDC insertion. The eddy current distribution and the time-varying Lorentz force and torque moment acting on the GDC electrode are presented. According to the calculation results of 7 cases, the maximum moment is about 3.3kNm. From the results the severe case(s) can be determined and only this severe case will be considered if a structure design is updated and this will save lots of resources avoiding consider 7 cases again. Below is a picture showing the moments and forces under 7 different cases.

        Speaker: Dr Lijun Cai
      • 1:40 PM
        Endoscope Emulator Test Stand for ITER Dust Monitor Diagnostic 2h

        E. Veshchev1*, Ya. Sadovskiy2, L. Begrambekov2, O. Bidlevitch2, O. Gordeev2, P. Shigin1, G. Vayakis1, M. Walsh1, R. Walton1

        1 ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex, France

        2 National Research Nuclear University (MEPhI) - NRNU MEPhI

        ITER is a basic nuclear installation and as such, safety is one of the most important drivers in the design. The ITER Licensing agreement requires that the quantity of dust in the vacuum vessel must remain below given limits. The maximum amount of mobilizable dust in the vessel is 1000 kg. A technique based on a flexible endoscope was selected as a tool for diagnostic of dust in ITER. The diagnostic will consist of two tools – one for fine viewing of dust with a resolution down to a few tens of microns in few mm spot and another one for dust collection. Both of endoscopes will have coarse viewing with resolution of few hundred microns over wider area to allow the possibility for more general inspection of the surrounding environment. The Endoscope will have to work in a harsh environment where the activation limit reaches a few hundreds of Gy/h, at a magnetic field of about 8T, and at a temperature of 100C. Ideally it will have to work in vacuum in order to allow inspections of the tokamak between shots or after disruptions. The Endoscope will have to go up to 20m deep inside the tokamak to the inspection region. Due to the specific design features of ITER, the endoscope will have to go upward on an inclined surface for inspection about 18 meters away from the insertion point. In order to ensure that the endoscope gets to the desired region of inspection it will be pushed through guide tubes having a number of bends along their length. Initial estimations of endoscope jacket materials, endoscope stiffness and push/pull forces were defined experimentally. This paper will give a brief reminder of the overall strategy for Dust/Erosion/Tritium monitoring in ITER and the role of the dust monitor in this context. It focuses on experimental results of real-size tests inside guide tubes of the behaviour of different endoscope emulators under various conditions.

        *Corresponding author: tel.: +33 4 42 17 65 20, e-mail: Evgeny.Veshchev@iter.org

        Speaker: Evgeny Veshchev (ITER Organization)
      • 1:40 PM
        Establish full covering liquid metal film flows under poor wettability conditions for liquid divertor of fusion reactor 2h

        Investigation on effects of wettability on liquid metal free surface film flow states have been performed by numerical simulations and experiments to establish a film flow which can cover the whole bottom solid surface. The effects of density of fluid, inlet film thickness and the width of bottom solid surface on the film flow states under poor wettability conditions have been investigated by numerical simulations, the results show that the rivulet flow is easily developed when the initial film thickness is small; it is more easily developed to rivulet flow when the fluid destiny becomes smaller; the covering bottom surface becomes big with the increase of the bottom surface width. But for liquid lithium it is difficult to get the film flow which can cover the whole bottom solid surface by increasing the bottom surface width and inlet film thickness. A new method by using a multi-curve bottom surface has been proposed to solve above problem. Firstly an experiment of the film of GaInSn alloy flow through a chute with a multi-curve bottom surface has been done to validate above solving method, it is indicated that this method is effective and the experimental results are in agreement with numerical results. Secondly numerical simulations have been performed to get the lithium film flows which can cover the whole bottom surface, it is shown that a full covering lithium film flow can be obtained by optimizing the shape of above mentioned multi-curve bottom surface. Above results is valuable for the design of liquid divertor of magnetic fusion reactor.

        Speaker: Dr Xiujie Zhang (Southwestern Institute of Physics)
      • 1:40 PM

        In Neutral Beam Injector (NBI), ions are accelerated to desired energy (from ~10 kV to ~MV range) by an electrostatic multi-aperture grid system of an ion source. Accelerated ions subsequently neutralized in a gas cell called neutralizer. To maintain the electrostatic lens configuration, grid plates are placed closely packed (~mm distance) parallel to each other. Paschen-breakdown (here it is called as grid breakdown) between the grid plates occurs routinely during system conditioning phases due to the presence of high voltage (HV) and sufficient gas (@ sub-atmospheric pressure). When grid-breakdown occur the stored energy of HVDC transmission line is dumped into the grids of the ion source at the breakdown location and possesses a danger to damage the grids by melting and even puncturing the spot.
        In Twin Source [1] 120meter long transmission[2] line is designed to connect accelerator power supply system -35kV ,15 A and extraction power supply system -11kV ,35A with Plasma Grid(PG) , Extraction Grid (EG) and Ground Grid(GG) .The major contribution for the stored energy emanates from the inter conductor capacitances or stray capacitances of the HVDC transmission line. This paper discusses the methodologies for estimation of the inter conductor capacitance and thus stored energy. The exercise helped to get the optimized possible transmission line configuration to ensure low stored energy to avoid grid damage. The analytical calculations of aforesaid configuration is validated with simulation, performed on COMSOL platform and the corresponding obtained results is further confirmed from capacitance measurement of 1m long prototype HVDC transmission line in similar configuration. The stray capacitances for 1m long HVDC transmission line between PG Line, EG Line and GG Line i.e. CPG EG , CEG GG and CPG GG estimated analytically are 13.41pF, 19.67pF and 8.6pF respectively. The simulated values are 19.41 pF, 18.45pF and 10.46pF respectively. The measured values are 16.86pF, 15.52pF and 9.67pF respectively. The estimated stored energy is 12.85mJ for 1m length of HVDC transmission line.

        [1] M. Bandyopadhyay, R.Pandey, S.Shah, G.Bansal, D.Parmar, A.Gahlaut, J.Soni, R.K.Yadav, D.Sudhir, H.Tyagi, K.Pandya, K.G.Parmar, H.S.Mistri, M.Vuppugalla and A.K.Chakraborty, “Two RF Driver Based Negative Ion Source Experiment”, IEEE Transactions on Plasma Science, 42, 624, March (2014).

        [2] Deepak Parmar, V. Mahesh, A. Gahlaut, K.G. Parmar, B.Prajapati, H. Shishangiya, M N Vishnudev, M. Bandyopadhyay, R. Yadav, J. Soni, R. Pandey, G. Bansal, K. Pandya, J.Bhagora, S.Shah, Dass Sudhir Kumar, H. Tyagi, A. Chakraborty, “Design & Development of Electrical System for TWIN Source”, Proceedings of Symposium of Fusion Engineering (SOFE) (2015).

        Speaker: Mr Vishnudev M N (ITER India, Institute for Plasma Research)
      • 1:40 PM
        Evaluation of the distribution of C5+ and Li2+ by the VUV imaging system on EAST 2h

        The Chinese Fusion Engineering Test Reactor (CFETR) is designed as the next fusion device in China aiming to bridge the gaps between the fusion experimental reactor ITER and the demonstration reactor (DEMO). The current EAST tokamak will provide a long-pulse, high power test bench for advanced operation scenarios under actively cooled metal wall condition for CFETR [1]. To achieve long-pulse and high power steady state operation on EAST, impurity accumulation is one of the key issues should be considered. Therefore, studies on impurity transport become important.

        Carbon is one of the major intrinsic impurities in EAST. Additionally, Lithium may exist through the dedicated Lithium-related experiment by Li-pellet injection in the experiment or through wall conditioning by lithiation. On EAST, a new vacuum ultraviolet (VUV) imaging system is developing. It selectively measures the emission with 13.5 nm in wavelength, which is mainly contributed by C VI (n = 4-2 transition), or the Ly-α line emission from Li III on EAST. In this work, a new method is proposed to evaluate the distribution of C5+ and Li2+ from the VUV imaging data. In this proposal, the evolutions of the impurity profiles are calculated by the 1D transport model. With the measured electron temperature and density profiles, the emissivity can be estimated. Then, the VUV images can be simulated. Finally, the impurity distributions can be obtained by fitting the simulated images and experimental images.

        This work is supported by the Natural Science Foundation of China under Contract No. 11605244 and the National Magnetic Confinement Fusion Science Program of China under Contract No. 2014GB106000, 2014GB106001 and 2013GB106000.

        [1] Y. X. Wan et al., 2016 26th IAEA Fusion Energy Conference, Kyoto, 17-22 Oct. 2016, Paper No. OV/3-4.

        Speaker: Mr Fan Zhou (Institute of Plasma Physics, Chinese Academic Sciences, Hefei, China; Science Island Branch of Graduate School, University of Science and Technology of China, Hefei, China)
      • 1:40 PM
        Experimental Investigation on the Second Commutating Process of a Quench Protection Switch 2h

        The quench protection switch (QPS) is an indispensable component to ensure the safety of the magnet coils of a superconductive tokomak when a quench happens. The two most important functions of a QPS are to carry high direct current during normal operation and to interrupt the high direct current when a quench occurs. In this paper, the second commutating process of a QPS based on artificial current zero is investigated. In this process, the current, which has already transferred from the by-pass switch to the main circuit breaker (vacuum circuit breaker), is forced to commutate from the vacuum circuit breaker to the dump resistance by the counter current. A LC oscillating circuit is applied to generate oscillating current to simulate the direct current near its peak which is in the range of 4-20kA. The counter current with frequency of 500Hz and 1000Hz is provided by a pre-charged capacitor bank. The equivalence of the interrupting process between practical direct current source and LC oscillating source is analyzed. The vacuum interrupter of the vacuum circuit breaker adopts a pair of contacts generating transverse magnetic field. The evolution of vacuum arc in the interrupting process is investigated by a high-speed camera with exposure time of 2μs.The experiment results indicate that the initial process and the motion of the vacuum arc before injecting the counter current have crucial impacts on the interruption performance.

        The research was supported in part by the National Magnetic Confinement Fusion Energy Research Project under Project No. 2015GB121005, and in part by the National Natural Science Foundation of China under Project Nos. 51322706 and 51325705.

        Speaker: Mr Sheng Li
      • 1:40 PM
        Experimental Study on Multilayer Liquid Metal Film Flow Characteristics under Horizontal Magnetic field 2h

        The liquid lithium has been considered as a suitable selection for the plasma facing component (PFC) materials in fusion reactors because of its plenty of advantages for removing the heat fluxes, incident tritium and impurity effluxes, and enabling a lithium wall fusion regime. However, under a complex magnetic field, the flowing liquid metal will exhibit complicated flow characteristics induced by the action of extra Lorentz force, which named magnetohydrodynamic (MHD) effect. Some preliminary experimental and numerical studies have proven that the flow resistance increased dramatically, the surface wave of film flow changed greatly, and liquid film cannot cover the whole solid surface with the existence of magnetic field. Because of the above deficits, it is hardly to form a uniform, stable lithium film in the Tokamak environment. In this paper, we make an attempt to make up the disadvantages on liquid metal film under magnetic field. A newly built liquid film generator with eight outlets distributed in different heights to form eight layers short liquid films, which connect one after another to form a long liquid film, is used to test the feasibility of our idea on the production of ideal liquid film. The Galinstan, at liquid sate in room temperature and low toxicity, is chose to substitute the lithium in our experiments. Experimental results show that this kind of film generator can enhance the spreading performance of liquid metal on solid surface and reduce the flow resistance induced by a magnetic field. A preliminary analysis is also carried and an evaluation of using this kind of liquid film generator as PFC in real fusion device has been conducted to lead some further studies.

        Speaker: Dr Yang Juan-Cheng (Xi'an Jiaotong University)
      • 1:40 PM
        Experimental Study on the Liquid Lithium Film Flow characteristics under Spanwise direction Magnetic field 2h

        Using liquid lithium film as the plasma facing component (PFC) is the prospective scheme in the future magnetic confinement fusion to withstand the plenty heat flux and improve the plasma performance. Under the magnetic field along the film spanwise direction, the lithium film flow will exhibit a complicated flow characteristics induced by the action of extra Lorentz force, which named magnetohydrodynamic (MHD) effect. Because of the lithium physical properties (e.g high melting point, high chemical activity and so on) and the space limitation of the magnet, some preliminary experimental studies are carried out by using the room temperature liquid metal Galinstan, which proved that the spanwise magnetic field could thicken the film, suppress the film flow turbulence, detach the flow away from the side wall, change the film surface wall, and so on. However, these experimental results can not apply to lithium entirely because of the great differences in physical properties (i.e the density of the lithium is small than one tenth of the Galinstan). In the present paper, the experimental facility of liquid lithium film flow under uniform spanwise magnetic field has been established at UCAS (University of Chinese Academy of Science), in order to study the spanwise magnetic influence on the lithium film behaviors. The test section, which is made by stainless steel, with the upside visible, enables to produce the lithium film flow with a width of 60mm and inclined angle of 6o. The flow rate, which is driven by the high pressure argon, could be adjusted from 0 to 10 L/min. The strength of the external magnetic field, which is generated by an electromagnet, is varied from 0 to 2T, with maximal unevenness lower than 5%. Experimental results show that the lithium film flow in the stainless-steel test section is significantly changed by the spanwise direction magnetic field, the surface waves are suppressed and became more stable. Finally, some quantitative analyses are also carried in present paper.

        Speaker: Prof. Ni Ming-Jiu (University of Chinese Academy and Sciences)
      • 1:40 PM
        Experimental study on vacuum control method for Paschen tests of the superconducting magnet 2h


        In order to verify the integrity of the insulation of superconducting magnet, it is needed to perform the Paschen test on the insulation after its manufacture. The vacuum vessel for the Paschen test is required to keep for an adequately long time under the pressure values of different degrees. To achieve the low pressure of the vacuum vessel, isolating the pumping unit from the vessel is not applicable as degassing inside the vessel will eventually ruin the pressure. Therefore, one dynamic balance control system for low pressure control is designed. The feature of the system is that adjusting the opening of intake valve by Proportion Integration Differentiation (PID) control system automatically, while the vacuum pump is working constantly. The results show the dynamic balance control system can keep the pressure value of 1±0.05 Pa, 10±0.15 Pa, 100±0.2 Pa and 1000±0.5 Pa, respectively and the holding time of each vacuum degree is more than 2 hours, which satisfy the basic vacuum requirement for the Paschen test of the superconducting magnet.

        Keywords: Superconducting magnet, Paschen test, Vacuum control, Dynamic balance, PID control system

        *Corresponding author. H. Wu. Fax: +86 55165591310.
        E-mail address: hwu@ipp.ac.cn (H. Wu).

        Speaker: Dr Zhirong Zhang
      • 1:40 PM
        Fast Boundary Reconstruction from Tangentially Viewed Visible Images for Plasma Control in EAST 2h

        The fast plasma boundary reconstruction is usually used for real-time control of tokamak plasma. In EAST experiment, the time consuming for boundary reconstruction should be within 1ms to meet the need of real-time control. A fast algorithm based on visible imaging diagnostics is developed in EAST to reconstruct the plasma boundary directly and independently. Compared to the results of EFIT, the overall average error is within 1.5cm, the average error at the lower X point is within 1cm, and the average error at the outermost and innermost points of LCFS are below 0.5cm. The causes of the deviation are discussed, and the methods for decrease are presented. For an image with the size of 680×544, the algorithm implemented by C++ with OpenCV can complete the computation in 0.9ms , achieving an acceleration of 300 times, when compared with parallel MATLAB. Furthermore, when the pixels of camera sensor is not saturated, the algorithm is robust for different intensities of the discharge images.

        Keywords: Plasma boundary reconstruction; Visible imaging diagnostics; EAST; OpenCV; Plasma control

        Speaker: Mr Heng Zhang
      • 1:40 PM
        Flow Test at Factory for ITER Thermal Shield 2h

        Thermal Shield (TS) is to be installed between vacuum vessel/cryostat and superconducting magnet in ITER tokamak. The TS plays a role in minimizing thermal radiation load onto the magnet. The TS has to be cooled by flowing 80 K helium gas inside cooling pipes welded on the TS surface. The helium coolant is supplied from the cryoplant via manifold pipes and distributed to all TS segments. Flow through each TS segment should be fully characterized to accurately predict the flow distribution in TS flow network for ensuring reliable operation of the TS.
        This paper describes how the manufactured TS segments are validated by factory flow test. Instead of applying cryogenic helium flow in the test, high pressure and room temperature nitrogen gas passes through the cooling pipe on TS segment. Equivalent test flow rate is determined by matching test Reynolds number with that of actual operating condition of TS based on similarity principle. Flow rate is controlled by a thermal mass flow controller and pressure drop between the inlet and the outlet of the pipe routing is measured by a differential pressure gauge. Test results are compared with calculated ones by incompressible pipe flow analysis to check the validity of pipe and elbow loss coefficients for the real manufactured TS segments. Orifice elements are to be connected to several TS segments for mass flow balancing of the TS flow network. The orifices are also tested separately by the flow test apparatus. Correlation for the orifice loss coefficient is derived from the test results.

        Speaker: Dr Kwanwoo Nam (ITER Korea)
      • 1:40 PM
        Heat transfer and Structural analyses of a water cooled tube under one-sided heating conditions for fusion reactor divertor 2h

        It is necessary to take effective cooling methods and remove the high heat flux from the divertor which sustains one-sided high heat fluxes over 10 MW/m2 in steady state. In order to ensure the reliability and security of the divertor in the fusion reactor, it is necessary to conduct the heat transfer and structural assessment of cooling tube in the divertor. In this paper, the heat transfer enhancement of the subcooled flow boiling in vertically upward screw cooling tubes, the tube with twisted tape and plain tube were carried out by using the Fluent software. Then, this paper conducted the structural analyses of the three cooling tube. The ranges of the working parameters of water are as follows: pressure P=4.5 MPa, water temperature at the inlet T_in=473 K and mass flux G=8653 kg/m2s. The 3Sm rule, pipe material temperature and the assessment of critical heat flux have been compared with investigated for the three cooling tube, which can guide the optimum for the heat transfer and structural assessment of the monoblock divertor.

        Speaker: Ms Liu Ping
      • 1:40 PM
        Inertia load analysis of ITER equatorial and upper port plug EPP9 and UPP14 2h

        Inertia load analysis of ITER equatorial and upper port plug EPP9 and UPP14

        Han Zhang1, Yuhu Zhai1, Julio Guirao2, Silvia Iglesias3, Peter Titus1

        1Prineton Plasma Physics Lab,Princeton, New Jersey, hzhang@pppl.gov
        2ITER Organization, Route de Vinon sur Verdon, 13067 Saint-Paul-lez-Durance, France
        3AETEC, Advanced Engineering Technologies S.L. 33206 Gijon, Spain

        The work presented in this paper mainly focuses on the response spectrum analysis of ITER diagnostic equatorial port plug (EPP) and upper port plug (UPP) structure assemblies to extract dynamic behavior of PPs and the in-port diagnostic systems due to transient vacuum vessel (VV) movements during plasma vertical displacement events (VDEs) and seismic loading. The generic port plug structural models were provided by ITER Organization (IO). Based on the generic EPP models, the US ITER equatorial port #9 Diagnostic Shielding Module (DSM) with in-port systems such as the Electron Cyclon Emission (ECE) was integrated in and the latest design of closure plate was used to replace the simple plate in the generic model too to ensure structural integrity. For the UPP14 model, diagnostic first wall (DFW), Glow Discharge Cleaning (GDC) system, wide-angle viewing system (WAV) and DSM shielding blocks etc. are updated to latest design based on the generic UPP model.
        Two types of response spectrum analysis (RSA) were performed: The floor response spectra (FRS) analysis based on random vibration (power spectrum density (PSD)) is to provide the input response spectra for the RSA of next level components in their system integration design and inertial load calculation. The RSA based on deterministic method (Multi-Point Response Spectrum (MPRS)) is to compute the steady state inertial loads of the components currently in our design. Four load cases were simulated and results are provided in this paper: plasma vertical displacement events (VDE-II, VDE-III, VDE-IV) and seismic event (SL-2).
        However, RSA gives only steady state result. For VDEs which last only for a very short time, less than half second, the system may not reach steady state so that the energy may not accumulate to reach maximal response as RSA calculated. Thus a transient run would be better to determine the dynamic behavior of the system. There are infinite numbers of time histories that are compatible to a given spectrum. Currently many commercial softwares exist to generate time histories for seismic qualification but almost nothing available for plasma VDEs. This paper provides the typical process to generate artificial time history for VDEs. For our model, two time histories are created and used to run the UPP14 model. Although the two results have relatively big difference at this time, when we have more information on the real disruption behavior of the VV, i.e. how the magnitude gradually increases, holds and decreases etc., this method can be improved.

        Speaker: Mrs Han Zhang (Princeton Plasma Physics Lab)
      • 1:40 PM

        The Neutral Beam (NB) system for ITER is composed of two heating neutral beam injectors (HNBs) and a diagnostic neutral beam injector (DNB). A third HNB can be installed as a future up-grade. This paper will present the design solution of the sealed interface between the components so called ‘Neutral Beam Front End Components’. The components to be considered are the Drift Duct, the Vacuum Vessel Pressure Suppression System box (VVPSS box), the Absolute Valve, and the Fast Shutter. These components connect the Neutral Beam vessels of the injectors to the Tokamak Vacuum Vessel. These components are connected with circular flanges bolted together. They are all first confinement barrier and by the way Safety Important Components classified. They must comply with stringent requirements in term of leak tightness and robustness. They are all classified RH class 3 that means that their life time shall comply with ITER life time without planned maintenance. In case of unlikely incidents or accidents, and regarding the results of Neutronic analysis in the Neutral Beam cell, the safety approach is to consider that all operations will be done fully remotely. It is not likely that it will be acceptable to allow human intervention.

        The paper will describe the design of the interface solutions which have to be implemented between these components regarding the primary vacuum confinement and the remote maintenance operations. The IO baseline solution to ensure the confinement and the leak tightness at these interfaces was the lip seal weld. The design and manufacturing code chosen for the NB FEC is the RCC-MR. The Lip seal weld solution raised two main concerns which are the compliance with the RCC-MR code and the feasibility of the full remote maintenance operations. The cutting and re-welding operations of the lip seal weld have never been demonstrated without a human intervention. And the RCC-MR is not directly applicable (or not relevant) for the lip seal weld. It appears that the design of the lip seal weld raises a lot of open issues like feasibility, tests and full RH operations inside the NB cell. The solution with two metallic seals with a pumped interspace will solve all lip seal weld concerns. A complete study has been carried out to demonstrate the compliance of this solution in term of leak tightness requirement, feasibility of a full remote maintenance operations and improvements of testing and monitoring these key interfaces.

        Speaker: Mr Marc Urbani (ITER Organization)
      • 1:40 PM
        Investigation on the Effect of Tritium Production using Temperature Control for DEMO Blanket 2h

        The tritium production is a key issue in the fuel recycle for DEMO blanket. It is affected by the temperature field inside the blanket interior due to the temperature requirement of the tritium release and the recovery. This paper discusses the tritium breeding ratio issues based on a PWR water-cooled blanket module. In particular, the variation trend of TBR is explored with the change of blanket interior. The tritium distribution is studied with the blanket temperature field. It is found that the pipe bore affects the local TBR sensitively, and the pipe bore with 9mm reaching the maximum local TBR for a blanket module. Tritium distribution calculations indicates there will be a large quantity of tritium generated in the area near the cooling pipes if the pipe bore is designed larger than 9mm. This will lead to the low tritium release efficiency for the blanket module due to the cooling effect of the pipes. Finally, the optimal range of the design parameters is obtained in view of TBR and tritium release performance.

        Speaker: Mr Yang Qiu (Institute of Plasma Physics Chinese Academy of Sciences)
      • 1:40 PM
        IPSE DIXIT: A User-Friendly Software Tool for the Design and Operation of Tokamak Power Supplies 2h

        The design and the operations of a tokamak often require to assess the feasibility of a desired experimental scenario with the available power supplies (PSs). After the definition of the evolution of the current waveforms in the supplied conductors (active elements, AEs) and in the plasma, it is necessary to verify if the PS equipment is able to produce them, also taking into account the conductive structures (mainly of the Vacuum Vessel) that are not directly fed by a PS (passive elements, PEs).
        In an operating tokamak, this process aims to estimate, with a reasonable trade-off between computation time and accuracy, the achievable experiments, the consequent stress on the components and the power demanded from the external grid. In the design of a new facility, this is essential to identify the component ratings and specifications and the impact on the electrical distribution systems.
        A user-friendly software tool was developed to answer the question “Is Power Supply Equipment Designed Implementing Current Scenario In Tokamak?” and was named IPSE DIXIT after the acronym of such a question.
        The main input data (mutual inductance matrix, resistance vector, currents in the AEs) can be entered with any mesh and time resolution. The effect of the PEs can be obtained by including their inductances and resistances in the input data. A more refined simulation of the plasma and of the AEs can be implemented by modeling them as distributed axisymmetric current filaments. Worst-case or random waveforms can be selected to include the influence of the coils for the feedback control of the plasma position and instabilities.
        Some functional models are available for the most common devices and system used in tokamaks, as AC/DC converters, thyristor bridges, switching network units (SNUs) and boosters. The current actually flowing in each thyristor of parallel and back-to-back bridge configurations can be estimated, also taking into account the limit firing angles and the circulation currents. The knowledge of the heatsink characteristics allows the calculation of the thyristors’ power loss and junction temperature. The verification of the single SNU resistor elements is also possible.
        The electrical contribution of the systems for plasma additional heating and of the plant auxiliary services can be estimated by simplified models.
        The final result consists in the estimation of the active, reactive and apparent powers expected for each tokamak operation.
        IPSE DIXIT was used to design the PSs of the Divertor Tokamak Test (DTT) facility moving from a reference single-null scenario. The results obtained considering only the AEs were compared to those obtained introducing the PEs by meshes of increasing complexity (at least 140 elements). A similar analysis was carried out also on JT-60SA, where the PS components are already defined. In this case, specific thresholds can be set to verify a selected scenario.
        The target of the research is to provide a free environment that could be adopted independently of the considered tokamak, covering a wide set of PS topology and components with a possible interaction with existing tools for structural analysis of experimental scenarios.

        Speaker: Alessandro Lampasi (ENEA)
      • 1:40 PM
        Manufacture and Electrical Properties of Instrumentation Wire Extraction Specimens for the ITER Feeder HV Insulation 2h

        For the quench detection of the superconducting busbars and joints in the ITER Feeder system, several high voltage (HV) instrumentation wires should be led out from the electrical insulation layers to transmit the voltage potential signal to the data acquisition systems. The penetration of the HV instrumentation wires through the intact insulation layers introduces a potential defect in the insulation, which could open up a path from the HV potential to ground; if this were to happen Paschen discharge would inevitably occur, which is the destructive accident for the safe operation of the Feeders.. Since 2015, ITER Organization and ASIPP have collaborated on the architecture and technology of instrumentation wire extraction, and an R&D program was undertaken in ASIPP. In this paper, the detailed HV wire extraction design is presented, along with specimens validating the design. The electrical properties of the wire extractions specimens, including DC hipot test, Paschen test and Partial discharge test, are presented and discussed.

        Keywords: high voltage, instrumentation wire, Paschen test, ITER, joint

        Speaker: Fang Linlin
      • 1:40 PM
        Measurement system of PSM HVPS for neutral beam injection on HL-2A 2h

        In order to improve the experimental parameters on the HL-2A device and meet the requirements of the HL-2A modification, the capability and pulse duration of the auxiliary heating system should be improved greatly. The high voltage power supply (HVPS) which is based on PSM technology is a method of controlling the total output voltage of many identical DC choppers connected in series by means of managing the on and off of modules step by step and modulating their pulse widths in a certain sequence . The neutral beam injection auxiliary heating system should have high output power (80kV/200A,), high accuracy of the output voltage (1%), flexible control, very high overall efficiency, very low amount of stored energy ( when loads are arcing ) and can be switched off immediately (< 15us) . It is necessary to design stable and precise measurement system that is basic premise and important guarantee for these features, especially for the sampling data, system control and protection reliability. The paper describes the measurement system (the slow signal bandwidth is about 0-20kHz, the fast signal bandwidth is about 0-250kHz) which is composed of various kinds of transducers (HV divider, AC and DC current transducer), isolation transmission, industry control machine and PCI based data boards. The experiment results show that the measurement system is flexible and stable.

        Speaker: yali wang (SWIP)
      • 1:40 PM
        Micro perspective on anti-fatigue performance enhancement of PFC metal welding interface with MD simulation 2h

        As one of key technical specification, the anti-fatigue performance of plasma-facing components (PFC) in fusion reactor receives widely concerns. Many researchs concluded that the micro structure on PFC metal connecting interface greatly affects its fatigue performance of interface, especially for some micro/nano scale pories in connection zone, which are difficult to be detected by ordinanry techniques and easy to be negleceted. In this paper, a method of ultilizing impacting stress wave to elimate internal nano pories on welding interface of PFC component is proposed. To examine the feasibility and results of this method, moleculer dynamics models are established. Then the healing process of a half sphere nano cavity under the impacting of different stress wave at different temperture conditions are computed and observed. To remedy the dislocation of grain after impacting stress wave, a kind of post heat treatment process is simulated based on different nano cavity-healed results at different conditions. Finally, the fatigue strengths of different cases are compared. The results show a prominent increase of fatigue strength for case treated by proposed method.

        Speaker: Mr Xuan Wang
      • 1:40 PM
        Modeling and analysis on the six-phase generator - converter system as the magnetic field power supply of HL-2A/M 2h

        The flywheel six-phase generator operated in pulse mode supply power of the magnetic field on HL-2M tokamak. The precise and stable control of six-phase generator is essential for obtaining high beta, steady state plasmas in HL-2M.

        Concretely for the power supply of the toroidal field coils on HL-2A, it consists of six-phase synchronous generators with their excitation systems, diode rectifiers, and the toroidal field coils. The diode rectifiers connect the terminals of the generators from AC side, and the toroidal field coils from DC side. During a shot of plasma discharge, the energy stored mechanically in the shafting of generator is transferred to the toroidal field coils by firing the exciter of the six-phase synchronous generator. In this case, modeling on the six-phase synchronous generator operated in pulse mode is the one of key issues for the precise and stable control.

        The state space realization of the six-phase generator with related exciters is specifically developed by dual DQ transformation. Based on the electromagnetic dynamics which describes the flux linkage changes in DQ frame, the six-phase generator is represented by exciting voltage controlled current source. On the other hand, the electromagnetic torque is calculated from the interaction between the flux linkages and currents in DQ frame, this torque drives the shafting speed to decrease from the initial speed, i.e., the energy is released.

        The dynamics of the toroidal field current scenario is simulated on the basis of state space realization of six-phase generator built in DQ frame. The results show the consistency with the experimental results of HL-2A, and the realization is effective for the toroidal field current control. Furthermore, this model can be adopted for the preprogrammed feedback control of power supplies on HL-2M. Moreover, it is also suited for the simulation and analysis of synchronous machine - converter systems.

        Speaker: Dr Xiaolong Liu (Southwestern Institute of Physics, Chengdu, China)
      • 1:40 PM
        Multi-scenario evaluation and electromagnetic loads on CFETR VV mockup during MD event 2h

        China Fusion Engineering Test Reactor (CFETR), as a new tokamak device to bridge the gap between ITER and DEMO, is developed for further research fusion power plant by China National Integration design Group for Magnetic Confinement Fusion. As a key component to maintain the reliability in run of high-temperature plasma in tokamak, vacuum vessel (VV) has a direct influence on the operation security of the total device. In order to establish the fabrication technology of VV, CFETR vacuum vessel mockup is constructed by Institute of Plasma Physics Chinese Academy of Sciences (ASIPP), its design parameters come from the China Fusion Engineering Test Reactor (CFETR). CFETR magnet system is required to meet the requirement of three scenarios of coil currents, which are used to realize the ITER-like, snowflake and Super-X plasma equilibrium shapes, respectively. In this paper numerical analysis is performed for the electromagnetic loads on CFETR VV mockup corresponding to three different current scenarios shapes during the MD event, respectively. The finite element model for electromagnetic analysis include a 22.5° VV sector and a magnetic system including 2 halves toroidal field (TF) coils, 6 poloidal field (PF) coils, 6 central solenoid (CS) coils, 2 divertor coils (DC) has been built, and a detailed CFETR VV mockup finite element model is established which consists of inner shell, outer shell, reinforcing ribs, ports and magnet coils, etc. The current loads are applied by current density method. The influence of plasma equilibrium configuration on the eddy current and electromagnetic force is also analyzed. The electromagnetic loads on VV during major disruption (MD) will provide a technical support for the future structural design and loads evaluations of CFETR VV.

        Speaker: Dr Ni Xiaojun (ASIPP)
      • 1:40 PM
        Multiphysics Modeling of the FW/Blanket of the U.S. Fusion Nuclear Science Facility (FNSF) 2h

        The dual coolant lead-lithium (DCLL) blanket concept, which is utilized in the Fusion Nuclear Science Facility (FNSF) conceptual design, is based on a helium-cooled first wall and blanket structure with RAFS (Reduced Activation Ferritic Steel) and a self-cooled LiPb breeding zone. The objective of this work is to develop a multiphysics modeling process in order to optimize the design and achieve long lifetime, maintainability, and high reliability. 3D finite element multiphysics modeling of the DCLL first wall and blanket (midplane of one sector) has been performed using COMSOL 5.2. The multiphysics aspect of the design is demonstrated via coupling of Computational Fluid Dynamics (CFD), conjugate heat transfer and solid mechanics. Both normal and off-normal loading conditions have been analyzed. The results of velocity, pressure, and temperature distributions of helium flow, as well as the primary and thermal stress of the structure were obtained. This was followed by determination of the factors of safety along three critical paths based on the ITER Structural Design Criteria for In-vessel Components (ISDC-IC). We show here that the structural design meets the ITER-ISDC design rules under both normal and off-normal operating conditions, though the safety factors under off-normal condition with 8 MPa helium pressure are marginal. Thus simple design optimization was conducted based on a parametric study on first wall dimensions to improve the design.

        Speaker: Yue Huang (UCLA)
      • 1:40 PM
        NBImag: a useful tool in the design of magnetic systems for the ITER Neutral Beam Injectors 2h

        NBImag is a code suitable for the design and optimization of complex magnetic field configurations, such as that of a multi-aperture, multi-stage negative ion source and accelerator. The NBImag code has been developed for the design of the ITER Neutral Beam Injector (NBI), whose full-size prototype, MITICA, is presently under construction in Padova, Italy. The ITER injector shall produce a focused beam of neutral particles (H or D) having an energy of about 1 MeV and a total power of 16.5 MW, for 3600 s continuous operation.

        The accelerator is constituted by a system of 7 conductive grids having different potential (from -1 MV to ground), each including 1280 apertures, with the purpose of forming a bundle of accelerated H- or D- beamlets with a total current up to 46 A or 40 A, respectively.
        Since none of the available commercial (or freeware) codes was suitable for efficiently modelling such a complex magnetic field configuration with acceptable detail level and computation time, the code has been developed and used for optimizing the magnetic field configuration in the ion source and accelerator, so as to comply with the constraints existing in different regions:

        • minimal magnetic field in the RF drivers for effective plasma start-up in the plasma source;
        • reduction (filtering) of the fast electrons in the plasma source for efficient production and extraction of negative ions in proximity of the plasma-facing grid;
        • optics quality (aiming and focusing) of 1280 negative ion beamlets ;
        • disposal of co-extracted and stripped electrons and minimization of the heat loads on accelerator grids (by early deflecting co-extracted and stripped electrons).

        A combination of a weak horizontal "long range" magnetic field produced by currents flowing in the plasma facing grid and in suitably arranged bus-bars, and of a strong vertical "local" magnetic field, produced by 5616 permanent magnets embedded in the accelerator grids, proved to be the most efficient configuration on the basis of an automated optimization procedure.

        The NBImag code is based on an integral formulation and allows an efficient calculation of any static magnetic field configuration on the basis of the geometry of the magnetic sources, with linear material and permanent magnets. NBImag also includes magnetic force and inductance calculation, based on the same formulation. Thanks to the capability of efficiently describing a large number of permanent magnets with limited computational effort, NBImag has also been integrated with different automatic optimization procedures for the solution of inverse magnetic problems.

        This paper describes the formulation of the code and of the optimization algorithms, the validation against analytical models and experimental measurements, and the application to the design of MITICA.

        Speaker: Daniele Aprile (INFN-LNL)
      • 1:40 PM

        The He-cooled Lithium Lead breeder blanket could be developed with the utilization of relatively mature material technology, which is used by Reduced Activation Ferritic / Martensitic (RAFM) steel as the structural material in China. It is necessary to analyse the first wall structure heated because of the thermal stress from two coolants in the blanket which would directly affect the blanket life and the safe operation coefficient, and indirectly carries on affection of the enhancement of thermal efficiency from electricity generation.
        The helium flow in the First Wall and LiPb flow with a transverse magnetic field in vertical channels in the blanket are investigated. The specially numerical MHD code based on the CFD software has been developed for analysis of the LiPb flow. The helium flow with four kinds of design scheme have been calculated and simulated. The three-dimensional temperature distributions of the LiPb flow in heating duct have been given. The analysis of the flow field and temperature gradient in the boundary layer of the duct have been performed. The heat transfer boundary condition of helium flow duct was determined by means of liquid-solid coupled method. The analysis for the structural stresses of the LiPb flow channel have been performed. The effect of the ratio of thermal load on the heat transfer characteristics of the helium and LiPb flow have been calculated and performed.

        Speaker: Hongyan WANG (Nanjing Institute of Technology)
      • 1:40 PM
        Optimization and Design of Divertor Langmuir Probe Diagnostic System on the EAST Tokamak 2h

        A flush-mounted Langmuir probe system has been built on the lower graphitic divertor targets on the EAST tokamak in 2016, which is transformed from the previous divertor Langmuir probes, aimed to reduce the erosion from high energy particle and strong heat flux on the probe surfaces exposed in the plasma, and explore new structure application. During the 2016 EAST campaign, the flush-mounted probe system has measured the plasma parameters by using single probe and triple probes respectively, to obtain electron density, electron temperature, particle and heat fluxes, and compared with the previous domed probe in the same plasma discharge condition. The results show that the plasma parameters measured by different measuring methods or different probe shapes are basically consistent, and demonstrate the flush-mounted probe system has been successfully used as a reliable diagnostic tool in the EAST divertor. Meanwhile, the design of the divertor Langmuir probe system is put forward and discussed, for the next generation of EAST lower divertor, which will be upgraded to full tungsten divertor with active water cooling, by optimizing the successful design of the Langmuir probe system on the ITER-like top tungsten divertor and the flush-mounted Langmuir probe system.

        We would like to acknowledge the support and contributions from the rest of the EAST probe team, collaborators and individuals for design and fabrication of the divertor probe diagnostic system. This work was supported by National Magnetic Confinement Fusion Science Program of China under Contract Nos. 2013GB107003, 2015GB101000, 2013GB106000 and National Natural Science Foundation of China under Grant Nos. 11575236, 11422546, 11575235 as well as the Thousand Talent Plan of China.

        Speaker: Mr Jichan Xu (ASIPP; USTC)
      • 1:40 PM
        Pebble Bed Thermo-mechanical Modeling for Water Cooled Ceramic Breeder Blanket for CFETR 2h

        The beryllium pebble bed and Li2TiO3/Be12Ti mixed pebble bed are selected to realize neutron multiplication and tritium breeding in the Water-cooled ceramic breeder blanket (WCCB) of China Fusion Engineering Test Reactor (CFETR). In order to evaluate and improve the performance of WCCB, studies of the thermo-mechanics of the concerned pebble beds are necessary.
        In the current research, a numerical model was constructed by using distinct element meth-od (DEM) to analyze behavior of prototypical blanket pebble bed. A thermal contact model based on SZB model was developed to analyze heat transfer in pebble bed. Besides, a numerical analysis program for uniaxial compression test was performed to estimate the macro-meso mechanical behaviors of pebble beds. The suitability and validity of the current numerical model were evaluated by comparing with the previous experimental or numerical data. According to the current calculations, the results of both effective thermal conductivity estimation and pebble bed loading/unloading analysis agree well with the previous experiments.
        Finally, the model was extended to the pebble bed analysis of WCCB. A series of numerical simulation work, including steady-state thermal analysis, uniaxial compression test were con-ducted to obtain basic pebble bed characteristic parameters, such as effective thermal conductivi-ty and strain-stress relation. This study will be dedicated to present the heat transfer features, macro-meso mechanical behaviors and the thermo-mechanical coupling characteristics of the blanket pebble beds, especially the Li2TiO3/Be12Ti mixed pebble bed for WCCB.

        Speaker: Mr Lei Chen (Institute of Plasma Physics, Chinese Academy of Sciences(ASIPP))
      • 1:40 PM
        Performance analysis on the VUV imaging system in EAST tokamak 2h

        Performance analysis on the VUV imaging system in EAST tokamak
        Z.J.Wang1, 2, T.F.Ming1, X.Gao1, 2, Y.M.Wang1, F.Zhou1, 2, F.F.Long1, 2 and EAST Team
        1Institute of Plasma Physics, Chinese Academic Sciences, Hefei, China
        2Science Island Branch of Graduate School, University of Science and Technology of China, Hefei, China

        In the present fusion research, magnetically confined Tokamak device is one of the most promising candidates for future commercial fusion reactor. The Experimental Advanced Superconducting Tokamak (EAST), the first fully superconducting tokamak with D-shaped poloidal cross-section, can be operated under similar configurations with ITER. It aims at high-performance plasma for long-pulse operation scenarios under actively cooled metal wall condition. During the past few years, a lot of significant progress and advances in both physics and technology has been made on EAST tokomak [1]. Additionally, studies on EAST will play an important role on both basic physics and key technologies for the Chinese Fusion Engineering Test Reactor (CFETR) as well.
        A tangentially viewing vacuum ultraviolet (VUV) high-speed imaging system, based on an inverse Schwarzschild-type optic system is developing to measure the edge plasma emission (including the pedestal region) in EAST. The telescope system consists of two multilayer mirrors: a convex mirror and a concave mirror. The mirrors are made of layers of molybdenum and silicon, which can selectively reflect 13.5nm (∆λ ∼ 1 nm) vuv light [2]. With this diagnostic, two dimensional (2D) structures of the edge magnetohydrodynamic (MHD) instabilities can be evaluated, which may be helpful on the physical understandings. In this work, the performance of this imaging system is discussed, including the image quality, estimation of spatial resolutions and noise levels, etc.

        This work is supported by the Natural Science Foundation of China under Contract No. 11605244 and the National Magnetic Confinement Fusion Science Program of China under Contract No. 2014GB106000, 2014GB106001 and 2013GB106000.

        [1] Y.X.Wan.et al 2016 26th IAEA Fusion Energy Conference, Kyoto, 17-22 Oct. 2016, Paper No. OV/3-4
        [2] T.F. Ming et al. Plasma and Fusion Research, Vol 6, 2406120 (2011)

        Speaker: Ms Zhijun Wang
      • 1:40 PM
        Physics and Geometry Design of Lower Divertor Upgrade in EAST Tokamak 2h

        Abstract-Experimental Advanced Superconducting Tokamak(EAST) device is a D-shaped full superconducting tokamak with actively water cooled plasma facing components. Before this upgrade, three generations of divertors, which, respectively, are steel divertor and carbon divertor and International Thermonuclear Experimental Reactor (ITER)-like divertor have designed. To achieve long pulse and high β H-mode plasma, new plasma position and shape are calculated and optimized in 2016 for EAST. The new geometry of lower divertor heavily relies on numerical simulations of the plasma in EAST.
        New divertors are designing to fit the high β H-mode plasma and endure the heat flux up to 10 MW/m2. To solve this issue, the lower carbon divertor will be upgrading in the future in EAST, which is in conceptual design phase. In consideration of the structure profile and function in EAST tokamak, in conceptual design phase these questions will be solved as follow. Firstly, the divertor should be better designed with advanced physical operation mode. Secondly, the divertor should be advanced geometry and high efficient cooling structure. The cooling circuit and the support systems of the component are installed on the vacuum vessel. The size of the space under the divertor should be consideration, which is very important. Thirdly, the cooling structure and maintenance of the divertor are also introduced in the paper.
        In consideration of physics and geometry design of lower divertor upgrade in EAST Tokamak, in the paper, mainly introduce the research progress of the fourth generation of divertors in EAST, much effort was focus on the divertor configuration and geometry.

        Speaker: Houchang Xu (Institute of Plasma Physics, Chinese Academy of Sciences, University of Science and Technology of China, Hefei University. )
      • 1:40 PM
        Power control system of 4.6GHz Lower hybrid wave for experimental advanced superconducting tokmak 2h

        The 6 MW/4.6 GHz lower hybrid current drive (LHCD) system as an effective approach for auxiliary heating and noninductive current drive has been designed and installed with twenty-four 250 KW/4.6 GHz high power klystron amplifiers in the experimental advanced superconducting tokamak(EAST). The power control system of 4.6GHz lower hybrid wave (LPCS) in continuous wave mode has been set up, which can control the lower hybrid power and protect the LHCD system. In this paper, the system architecture and software of the LHPCS are presented. The LPCS of 4.6GHz LHCD included the microwave pre-amplifier system, directional coupler, high reflected power protection subsystem, data acquisition and power control subsystem. The microwave pre-amplifier system contain master oscillator box, two power dividers box and 24 pre-amplifier box. There were two set high reflected power protection systems. They were installed to make sure klystron, ceramic window and other devices in safety once high reflection occurs. Data acquisition and power control computer is set up on the basis of national instruments CompactRIO and PXI system. The software for high reflected power protection subsystem, data acquisition and power control subsystem were based on LabView. Moreover, the experiment of measurement of incident power, High reflected power protection were dexcribed here in detail. Finally, High power CW operation and power modulation experimental with feedback controlled low power microwave source to results in EAST were show here in detail.

        Speaker: Mr Wendong Ma (Institute of Plasma Physics, Chinese Academy of Sciences)
      • 1:40 PM
        Prediction of departure from nuclear boiling in the first wall of WCCB blanket for CFETR 2h

        The Water Cooled Ceramic Breeder blanket(WCCB), which employs the operating conditions of Pressurized Water Reactor(PWR), namely inlet/outlet temperature of 285/325℃ and pressure of 15.5MPa, is being comprehensively researched in the Institute of Plasma and Physics Chinese Academic of Sciences. As an important component of the Water Cooled Ceramic Breeder blanket(WCCB), the first wall faces the plasma directly. It removes away the high heat flux and nuclear volumetric heat by coolant water flowing through the internal cooling passages. The departure from nuclear boiling(DNB) is the typical crisis in the reactor which uses water as coolant, especially for PWR, because it can make the water near heated wall dry out, at which point the local temperature can have a excursion exceeding the limits and the integrity of the structure is damaged. The DNB can easily happen in the first wall(FW) channel for the reason that the enhanced radial transport and edge-localized modes(ELMS) in the fusion reactor can increase the heat flux as high as several MW/m2. Therefore, the investigation on the departure from nuclear boiling is necessary.
        In this paper, the DNB is numerically analyzed by the CFD approach, which has the capacity of solving the Eulerian two-phase equation with Rensselaer Polytechnic Institute (RPI) wall boiling model. Responding to the excursion of heat flux, the main issue concerned is to determine the ultimate heat flux when the boiling instability happened during the normal operation, indicating the DNB occurs. Furthermore, the influence of different structure design on DNB is also investigated. The FW containing the parallel channels is modeled, in which the velocity of each channel is obtained from the previous thermal hydraulic analyses on the blanket under the normal condition, namely the heat flux of 0.5MW/m2. Besides, the detailed flow behavior and distribution of two-phase are also revealed. All these results are beneficial for the further safety operation of fusion reactor.

        Speaker: Dr Kecheng Jiang (Institute of Plasma Physics Chinese Academy of Science;University of Science and Technology of China)
      • 1:40 PM
        Preliminary Cooling Channel Design and Thermal-hydraulic Analysis of GDC PE in UPP14 2h

        The GDC PE in UPP14 is one of the plasma facing components in the ITER tokamak device. It will get a large number of thermal loads such as radiation and charge-exchange, neutron heating. So it would need the active cooling requirements from Diagnostic Port Plugs(DPP).
        The electrode model consists of three parts: the electrode head, the electrode rod and pipes.The first layer channel shape is a rectangle (11mm×15mm) except for the first and the last one due to the irregular head shape.The shape of the other channels is all circular and many channels are designed in parallel way for two reasons: The thermal load on the second half is less than the front part; Reduce the flow resistance.
        To obtain the reasonable cooling needs, the preliminary thermal-hydraulic simulation and analysis has been done, which is based on the turbulence model.
        The pressure loss in the fluid channel is about 0.452MPa(allowable maximum pressure drop 1.35MPa) during POS. The temperatures of the electrode volume and the electrode temperature peak is 277.5℃, which is less than limit peak stainless steel temperature 450℃. Therefore, all the results meet the thermal design under the flow rate 0.7kg/s, and they could provide some references for the next design optimization, such as pressure drop matching with other components, etc.
        Key words: ITER, GDC, thermal-hydraulic

        Speaker: Yong LU (SWIP)
      • 1:40 PM

        HL-2M tokamak is considered as one of the most important short pulse device for future fusion research in China which is being built at Southwestern Institute of Physics. In the vacuum vessel of the HL-2M, The first wall in weak magnetic side is designed to protect the vacuum vessel, cryopump, RMP coils and diagnostic components from the plasma particles and heat loads. Currently, the preliminary design for the first wall in weak magnetic side is in progress.
        Considering the risk of leakage and complexity of design, a passive cooling structure is adopted in the first wall of the weak magnetic side. In order to enhance thermal transfer, the first wall is made of copper alloy (CuCrZr) and graphite tile. Transient thermal analysis has been used to predict heat load for normal operating scenarios in a day. The maximum temperature of this first wall is about 307℃ which is engendered on the graphite tile. After a day of operating, the temperature of the passive cooling first wall can be reduced to 54℃.
        As a consequence of the high temperature, the stress between graphite tile and copper alloy need more attention. Spring washer and pressure bar have been carried out to optimize the mechanical joint. Flexible copper sheet is placed in joint faces to increase thermal contact resistance.

        Speaker: Dr Tao Lin (Center for Fusion Science of Southwestern Institute of Physics)
      • 1:40 PM
        Preliminary mechanism analysis of HyperVapotron experiment for high heat flux components 2h

        The ITER Tokamak Cooling Water System (TCWS) is designed to provide cooling and baking for the Primary Heat Transfer Systems (PHTSs) and relevant systems including the first wall/blanket, vacuum vessel, divertor, etc. For its characters of promising to enhance the heat transfer performance and increase critical heat flux (CHF), HyperVapotron (HV) elements has been put forward as heat removal of high heat flux components in nuclear fusion research facilities.
        In our study, a hypervapotron loop test facility was built to conduct some experiments of heat transfer. Phenomena of HV were observed using the techniques of planar laser induced fluorescent (PLIF) , high speed photography, particle image velocimetry (PIV), etc. According to the design of the ITER cooling water system(CWS), the flow and condition parameters were utilized: (1) CuCrZr alloy material, (2) rectangle fin with height 8mm and width 3mm , (3) inlet subcooling temperature of 298K, (4) channel flow speed of around 6ms-1, (5) maximum heat flux of around 1.5MWm-2.
        The relation of the heat transfer coefficient (HTC) between rectangle fin 3mm×8mm and rectangle fin 3mm×3mm will be presented under the identical tested conditions. Furthermore, the preliminary mechanism of the heat transfer will be explained by the coupling effect of the vortex flow with air bubble in the fin pitch, and the heat transfer efficiency is extraordinary dependent on the maintain time of vortex forming between the fins.

        Speaker: Delin Chu
      • 1:40 PM
        Pumping Performance Calculation of HL-2M in-vessel Cryopump based on Monte Carlo method 2h

        Sufficient pumping speed and good pumping performance should be guaranteed for the HL-2M Tokamak.Technical parameters were obtained by directly simulating with the Monte Carlo method for HL-2M in-vessel cryopump under molecular flow conditions. The predicted pumping speed of bare pump is 51.29 m3/s for H2, 38.04 m3/s for D2 and 24.94 m3/s for He. An advanced divertor system will be installed, located on the floor of the HL-2M vessel. Its pumping speed contained conductance of divertor is 20.11 m3/s for H2, 15.14 m3/s for D2 and 12.50 m3/s for He. It shows that the pumping effectiveness of the cryopump be affected by the structure of divertor greatly. The pumping speed of the cryopump can be influenced by the sticking coefficient, results be obtained by analyzing different sticking coefficient (vary ±0.05). The numerical and deduction results show that the sticking coefficient has small effect on pumping speed for H2 and D2, but obviously affects the pumping speed for He. The pumping process dynamic evaluation results show that the cryopump has a quick time response.

        Speaker: Mr Yong Li (Southwestern Institute of Physics)
      • 1:40 PM
        Qualification of ITER PF6 Helium Inlet 2h

        The Poloidal Field (PF) coils are one of the main sub-systems of the 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 signed between ASIPP and Fusion for Energy (F4E).
        The ITER PF6 winding pack is composed by stacking of 9 double pancakes. Each double pancake is supplied with supercritical helium with 2 inlets located on the innermost turn, in the middle of the layer joggles where the conductor goes from a pancake to the other. The helium inlet will undergo huge cyclic electromagnet loads during Tokamak operation, thus needs to be qualified with rigorous procedures.
        This paper describes PF6 helium inlet qualification. In the qualification process, helium inlet hole drilling and stainless steel wrapping removal were carried out. Helium inlet welding with full penetration by automatic welding machine was performed, temperature measurement during welding was implemented and was under 250 °C. Leak test and X-ray test were applied to ensure no defect was exist. The qualification samples passed the 600,000 cycles fatigue test and laminography test. Micro and macro inspections were been done to finally check the welding quality. PF6 helium inlet qualification was approved by ITER IO before Dummy double pancake manufacturing.

        Speaker: Mr shuangsong Du
      • 1:40 PM
        R&D of linear plasma facilities for PMI research at ASIPP 2h

        Linear plasma devices can produce low energy, high flux plasmas to simulate boundary conditions in tokamaks. Recently, two new linear plasma facilities have been built at the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP). To understand the hydrogen isotopes behavior in first wall materials, the PREFACE (Permeation and Retention Evaluation FACility for fusion Experiments) machine has been constructed. Another effort is to achieve reactor-relevant divertor plasma parameters in laboratory using RF-based technology. For this purpose, the HPPX (Helicon Physics Prototype eXperiment) machine has been built to address some scientific and technical issues of steady-state helicon plasma discharge.
        The main mission of PREFACE is to perform hydrogen plasma-driven permeation (PDP) experiments on plasma facing materials. The PREFACE facility is equipped with a 6 kW@2.45 GHz electron cyclotron resonance (ECR) source and plasmas with a diameter of 40 mm can be produced. The typical electron temperature and density are Te = 2-6 eV and ne = 1E16-1E17 m-3, respectively. For PDP experiments, the plasma density should not be too high to avoid the melting of sample membrane. The basic diagnostics includes a Hiden Langmuir probe, an Avantes spectrometer (197-717 nm) and several thermal couples. Hydrogen isotopes permeation and retention data have been taken for materials like tungsten, reduced activation martensitic/ferritic steels and copper alloys in PREFACE.
        The HPPX facility has a 4 m long vacuum chamber, which consists of four 1 m sectors with an inner diameter of 0.5 m. Modularization design has been applied so that the vacuum vessel can be easily extended for other research purposes in the future. At present, a 13.56/27.12 MHz RF source has been connected to the machine and the maximum power is 50 kW. A steady-state plasma density of >1E19 m-3 is expected. The electron temperature and density will be further increased by extra plasma heating. An ECR source with a power of over 100 kW has already been proposed.

        Speaker: Dr H.-S. Zhou (Institute of Plasma Physics, Chinese Academy of Sciences)
      • 1:40 PM
        Recent progress of pellet injection system in Experimental Advanced Superconducting Tokamak 2h

        Pellet injection, which is regarded as the most promising technique for the research of edge plasma physics, can control the edge localized mode (ELM) [1] and reduce the power threshold of L-H transition [2]. Since the installation of the 10 Hz pellet injection system [3] in Experimental Advanced Superconducting Tokamak (EAST), lots of experiments have been carried out. The 10 Hz pellet injection system in EAST can continually produce pellets with both diameter and length of 2 mm, containing ~3.78 × 1020 atoms in each pellet. Except for the normal fueling effect, high-confinement (H-mode) plasma was achieved by injecting frozen deuterium pellets in EAST. Interesting phenomena of simple and two-stage low-high confinement (L-H) transitions are observed in EAST with radio frequency heating after shallow pellet injection. The results of the L-H transitions induced by pellets are discussed in detail with different theories. It is found that pellet injection in EAST can reduce the power threshold of H-mode. Furthermore, the pellet-induced edge density gradient is one of the important parameters affecting the L-H transition. Comprehensive researches will be carried out in the next campaign with the development of a new 50 Hz pellet injection system recently [4, 5] in EAST, which is capable of injecting pellets with different sizes. Besides, it is also observed in EAST that a deep penetration pellet can cause severe snake-like perturbation in the core plasma region [6]. This snake phenomenon, which was clearly monitored by the soft X-ray diagnostic, had a long life time of ~1 s. These investigations prove that pellet is a powerful tool to investigate not only the edge plasma physics but also the core plasma physics. This research is funded by the National Nature Science Foundation of China under Contracts No. 11625524, No. 11321092, and No. 11605246 and the National Magnetic Confinement Fusion Science Program under Contract No. 2013GB114004.
        [1] Lang P. T. et al 2003 Nucl. Fusion 43 1110
        [2] Gohil P. et al 2001 Phys. Rev. Lett. 86 644
        [3] Li C. Z. et al 2012 Fusion Engineering and Design 89 99
        [4] I. Vinyar et al 2015 Fusion Engineering and Design 98-99 1898
        [5] Yao X. J. et al 2017 Fusion Engineering and Design 114 40
        [6] Yao X. J. et al 2016 Plasma Phys. Control. Fusion 58 105006

        Speaker: Dr Xingjia Yao (ASIPP)
      • 1:40 PM
        REFMULF: 2D Full-wave FDTD Full Polarization Maxwell Code 2h

        An important tool for the progress of reflectometry is numerical simulation, able to assess the measuring capabilities of existing systems and to predict the performance of future ones in machines such as ITER and DEMO. A novel 2D full-wave FDTD code, REFMULF, presented here is able to cope with full polarization waves, coupling the Transverse-Electric Mode (TE, X-mode) with the Transverse-Magnetic Mode (TM, O-mode) via a linear vectorial differential equation for $\mathbf{J}$ with a generic external magnetic field $\mathbf{B_0}$. This equation, coupling wave propagation, described by Maxwell curl equations to the plasma media, is solved using a modified Xu-Yuan kernel [1], [2] with extended long-run stability. The external magnetic field components of $\mathbf{B_0}$ lying on the propagation plane are responsible for linking the TE and TM modes. For a $\mathbf{B_0}$ purely perpendicular to the propagation plane the code describes simultaneously o-mode and x-mode propagation. This code enlarges the possibilities of simulation of microwave reflectometry, including depolarization processes in turbulent plasmas, offering capabilities unavailable in present day 2D reflectometry codes, closing the gap to the much sought-after computationally affordable 3D code. Being a parallel code is able to cope with real size problems. Although originally written with reflectometry in mind the code can is useful to simulate other diagnostics such as Collective Thomson Scattering, or Electron Cyclotron Resonant Heating.

        [1] Lijun Xu, Naichang Yuan, IEEE Antennas And Wireless Propagation Letters 5, 335-338 (2006).

        [2] F. da Silva, M. Campos Pinto, Bruno Després and Stéphane Heuraux, Journal of Computational Physics 295, 24-45 (2015).

        Speaker: Dr Filipe da Silva (Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade de Lisboa)
      • 1:40 PM
        Repair of the cracked surface of W using high energy pulsed laser 2h

        W is a promising plasma facing material for fusion devices. It is expected to suffer from the transient heat flux during normal ELMs, abnormal VDEs and plasma disruption events. In the next fusion device such as the ITER, the transient high heat flux can reach up to several MJ/m2 in a very short pulse (~ms), which is as high as enough to cause the surface damages especially in form of cracks [1]. The cracks, to some extent degrade material perfomance. To alleviate the influence of the cracks, the repair of the cracked surface using high energy pulsed laser has been proposed and investigated [2].
        In the present work, the repair of cracked surface of W was performed in the laser and wall material evaluation device. The W was pre-damaged by the transient heat flux exposure in EMS-60 with parameters of 400MW/m2 for 1ms and 100 cycles. The net-like macro cracks were successfully generated as expected. Then, the damaged surface of W was repaired in the laser and wall material evaluation device with a high vacuum circumstance. The damaged W was preheated to a elevated temperature exceeding DBTT of W, then the high energy pulsed laser with a wavelength of 532 nm, a energy of 0.2~0.8 kJ, a frequency of 10 Hz and a circular beam diameter about 0.4 mm was scanned on the damaged area with the adjustable single spot repetition numbers and overlapping ratio between adjacent spots from 10-50%.
        After the repair process, the net-like cracks successfully disappeared at the laser scanned areas, meanwhile, there were no any other type cracks founded. It is should note the high energy pulsed laser can also cause other type of crack patterns at room temperature. Thus, the pre-heating process suppressed the cracks formation by laser shocks. The residual trace for the net-like cracks could be also distinguished by micro observation. In addition, the surface seemed to become rough from micro perspective, identifying that the surface underwent the plastic deformation during laser scanning. The repair mechanism may different with the laser re-melting method with an initial room temperature [3]. The single spot repetition numbers, the overlapping ratio between adjacent spots, the laser energy and spot dimension have the important influence on the repair effect and need in-depth optimization. Moreover, the behavior and properties of the repair surface under the subsequent plasma and heat exposure is unknown and need future relevant tests.
        [1] Xiang Liu, Youyun Lian, Lei Chen, et. al., Experimental and numerical simulations of ELM-like transient damage behaviors to different grade tungsten and tungsten alloys, J. Nucl. Mater. 463 (2015) 166-169.
        [2] Y. Ueda, Pulsed heat load effects on plasma facing materials, 2015 ITER International School, Hefei, China, 2015,14 - 18
        [3] Th. Loewenhoff, J. Linke , J. Mat ˇejíˇcek, et. al., Laser re-melting of tungsten damaged by transient heat loads, Nuclear Materials and Energy 9 (2016) 165–170

        Speaker: Mr Dahuan Zhu (Institure of Plasma Pyhsics, Chinese Academic of Sciences)
      • 1:40 PM
        Research and design of microwave diagnostics on CFETR 2h

        Microwave diagnostics including reflectometry and electron cyclotron emission are the candidates for tools to measure the basic parameters such as electron density and temperature profile on CFETR. They have high spatial and temporal resolution with the radial coverage of the entire plasma. Nevertheless, because of the transmission of signal relies on the waveguide which can survive in the neutron environment on CFETR, they have need for reduced access, front-end robustness. However, due to the high temperature of the target plasma especially in the core region based on the scenarios, realistic effect can change the position of the cutoff layers and downshift the ECE frequency. Based on the newly developed scenarios, the cutoff frequencies and the electron cyclotron frequencies are carefully calculated taking the realistic effect into consideration. The spatial coverage and resolution are evaluated under the developed scenarios of CFETR.

        Speaker: Mr Hao Qu (Institute of Plasma Physics, Chinese Academy of Sciences)
      • 1:40 PM
        Research on the Method of Reactive Power Detection for Tokamak Coil Power Supply Based on AC/DC System Active Power Balance 2h

        A new method of real-time reactive power detection is proposed, which aims to be applied in reactive power for Tokamak coil power supply volatility, high random and the contradiction between accuracy and real-time of traditional method. It based on AC/DC system active power balance principle, and considering the electric network voltage sag, distortion and excitation current of transformer. The characteristics of the detection method are high real-time and precision, which not affected by electric network time-varying parameter. The method has been proved correctness and effectiveness by the experimental of reactive power detection for poloidal field power supply in EAST.

        Speaker: Mr Yanan Wu (ASIPP)
      • 1:40 PM
        RIPER: An irradiation facility to test Radiation Induced Permeation and Release of deuterium for fusion reactor materials. 2h

        For successful future Fusion Power Plant operation, tritium self-sufficiency is an essential element of the multiple technical challenges facing the fusion programme. In particular all the different types of candidate blankets will have to make use of different functional materials, such as SiC for flow channel inserts (FCI), ceramic coatings on steel for liquid metal blankets, and Li-ceramic breeders for the helium cooled pebble blanket (HCBP). For all these advanced materials radiation enhanced tritium permeation and retention are of concern. Also thermochemical, and in particular radiation stability, must be taken into account. Radiation induced changes in composition and microstructure may alter either production and/or extraction of tritium, permeation, and retention, hence seriously affecting the achievable tritium breeding ratio (TBR). The validation of these advanced materials requires experimental data to be obtained under as near as possible reactor operating conditions.

        At CIEMAT (Research Centre for Energy, Environment, and Technology) during the various Euratom and Broader Approach (BA) agreements, different experimental systems have been developed in the beam line of a 2 MV Van de Graaff electron accelerator. These allow one to study in situ numerous radiation enhanced and induced effects such as electrical, luminescence, and diffusion properties in fusion insulating and breeding blanket materials. Within this framework the Radiation Induced Permeation and Release (RIPER) facility has been developed to provide essential tritium related data. The facility consists of several special irradiation chambers and corresponding experimental systems to measure deuterium permeation for ceramic coated metals during irradiation at different irradiation temperatures and gas pressures, as well as deuterium release from Li-ceramic breeders during irradiation. In the same beam line facility it is also possible to measure hydrogen isotope (H and D) transport under relevant conditions, where deuterium adsorption, absorption, and desorption, including thermally induced desorption (TID) can be measured under controlled ionizing radiation and temperature conditions. The system also allows one to determine possible decomposition such as lithium vaporization/release during irradiation and/or heating of the Li-ceramic materials. All the above gas release processes are monitored using a Pfeiffer Smart Test commercial gas leak detector (He, D2 sensitivity ≥ 10-12 mbar.l/s) and a Pfeiffer PrismaPlus QMG 220 residual gas analyser – quadrupole mass spectrometer (sensitivity ≥ 10-14 mbar) connected to the vacuum system.

        The paper will give a detailed description of the above experimental systems as used to test ionizing radiation and temperature effects on the functional properties of candidate breeding ceramics, silicon carbide for FCIs, and radiation induced permeation through alumina coated stainless steel.

        Speaker: Dr Alejandro Morono (CIEMAT)
      • 1:40 PM
        Shutdown Dose Rate Calculation for the Preliminary Concept of K-DEMO Equatorial Port Area 2h

        The Korean fusion demonstration reactor (K-DEMO) will be operated in a highly irradiated condition by 14 MeV neutrons from D-T plasma. During this condition, irradiated materials generate radioactive nuclides. The nuclides emit decay gammas during operation and even after the shutdown of the tokamak reactor. One of the important safety-related maintenance areas in the tokamak reactor is the outboard equatorial port area. Although it is close to highly irradiated plasma facing components, the human access is necessary for the maintenance. Thus, the reliable result for the shutdown dose rate calculation has to be presented to assure the human safety. The preliminary concept of K-DEMO equatorial port was developed and then, it was transported into the K-DEMO neutronic analysis model [1]. This model adopted the labyrinth structure to prevent neutron leakages between the equatorial port structure and neighboring components. The shutdown dose rate calculations have been performed in the vicinity of the equatorial port area based on the rigorous 2-step (R2S) method [2]. This method couples transport and activation codes of the MCNP [3] and FISPACT [4]. The shielding calculation by changing shield thickness has also been performed to provide adequate neutron and radiation shields to reduce the dose level at the equatorial port interspace. The preliminary analysis results indicate that the dose level in this area is below the design target value of 100 μSv/h at 12 days after shutdown.

        [1] J. S. Park, K. Im, and S. Kwon, “Development of the Advanced Neutronic Analysis Model for the K-DEMO with MCNP Code”, IEEE Transactions on Plasma Science 44 (2016) 1751-1757.
        [2] Y. Chen and U. Fischer, “Rigorous MCNP based shutdown dose rate calculations: computational scheme, verification calculations and application to ITER”, Fusion Engineering and Design 63 (2002) 107-114.
        [3] D. Pelowitz, (Ed.), MCNP6TM User’s Manual, Version 1.0, Los Alamos National Laboratory LA-CP-13-00634, Rev. 0, May 2013.
        [4] R. A. Forrest, FISPACT-2007 User manual EURATOM/UKAEA fusion association (2007)

        Speaker: Jongsung Park (National Fusion Research Institute)
      • 1:40 PM
        Simulation of turbulent plasma heat flux to the DEMO first wall 2h

        First wall (FW) of the DEMO reactor should protect the breeding blanket and mechanical construction elements of the burning plasma exposure. The plasma impacts the wall surface by heating and by energetic particles. The heat load on FW during steady state burning mainly consists of the following factors: (i) the plasma photonic radiation, (ii) the plasma heat flux along the magnetic field lines, (iii) charge-exchange neutrals, (iv) alpha particles produced by fusion reaction and partially leaked into the scrape-off layer (SOL).
        Assessment of the FW heat load is one of the key design issues determining the DEMO reactor, because the heat flux there is a challenge for the FW armor material both due to high operation temperature and considerable erosion rate by sputtering. Cooling system of the first wall in DEMO should provide stable operation in the wide range of surface heat fluxes: from 0.3 MW/m2 up to 3-5 MW/m2.
        In this paper a model for the FW heat load caused by the plasma turbulent heat flux associated with plasma blobs is developed. Plasma blobs, or plasma blob filaments, are localized regions of isolated enhanced plasma density and temperature of a few cm cross-field sizes, spanned along magnetic field in SOL from wall to wall. The blobs propagate in radial direction from separatrix to the wall with rather large velocity, depositing heat flux by electron thermoconductivity and by ion convection parallel to the magnetic field at the intersections with the wall. A model of the blobs moving with constant radial velocity and depositing heat flux to the wall due to these processes has been developed and implemented into the TOKES code, developed over the past decade at FZK-KIT for integrated 2D simulations of transient events in tokamaks.
        First simulations of the intermittent turbulent plasma wall heat load due to the blobs for the DEMO-I tokamak reactor design have been performed. Poloidal profile for the heat flux to the toroidally symmetric first wall has been calculated.

        Speaker: Dr Sergey Pestchanyi (KIT)
      • 1:40 PM
        Some Properties of Beryllium Pebbles Produced by Powder Metallurgy for HCPB Breeding Blanket Application 2h

        Beryllium is planned to be used as the neutron multiplier in helium cooled pebble-bed (HCPB) breeding blanket concept for DEMO power plants. Under neutron irradiation a large amount of helium and tritium is produced in beryllium. The key issues of neutron irradiation of beryllium are helium-induced swelling and tritium retention and release. Because of safety requirements, the in-pile tritium release should be sufficiently high to avoid risk to personal in case of a serious accident in a fusion power plant leading to abrupt release of all accumulated tritium.
        In the present HCPB breeder blanket design, beryllium is used in the form of pebbles with diameter of ~1 mm, having inherently large grains (in the 500-1000 $\mu$m range) due to the fabrication by “Rotation Electrode Method” or “Fluoride Reduction Process”. However, it is expected that in beryllium with fine grain structure (average grain size of a few tens micrometers) helium and tritium release can be improved significantly. In order to produce the pebbles with a fine grain structure, some R&D were performed in Bochvar Institute. Several experimental batches of Be pebbles with average pebble size of 1.2 – 1.3 mm and different grain sizes (from ~ 13-14 $\mu$m up to ~615 $\mu$m) have been fabricated by powder metallurgy and then characterized.
        This paper presents the results of investigation of three batches of beryllium pebbles with average pebble size of 1.2 – 1.3 mm and different average grain sizes (13-14 $\mu$m, ~50 $\mu$m and ~615 $\mu$m). Microstructure and chemical composition of produced beryllium pebbles are presented as well as packing density and pebble size distribution. The influence of grain size on tritium release and retention in Be pebbles during temperature programmed desorption (TPD) after high-temperature loading of tritium/hydrogen gas mixture are also described.

        Speaker: Igor Kupriyanov (A.A. Bochvar High Technology research Institute of Inorganic Ma)
      • 1:40 PM

        The debate on which will be the operative mode of a future commercial fusion power plant, steady state or pulsed, is still open even in the on going pre-conceptual design phase of the European demonstration power plant, DEMO.

        As part of a series of studies on the economics of alternative options of commercial fusion power plant carried out with the FRESCO (Fusion Reactors Simplified Cost) code, the work here presented provides further insights on the issue.

        FRESCO is based on simplified models of physics, engineering and economics of a TOKAMAK-like pulsed or steady state fusion power plant. While the model of the BoP (Balance of plant) is derived from the PPCS study and is usually kept fixed, the assumptions on the plasma physics and technological choices can be adapted to the case of study. The assessment of the economics of the power plant is aimed at estimating the effects of specific plant features on the cost of electricity (COE) rather than formulating forecasts. Then, stochastic analyses based on the Monte Carlo method are carried out in order to assess the weight on the COE of uncertainties on multiple aspects (e.g. on the costs of both raw materials and components themselves, on the actual lifetime of plasma facing components and the power plant as a whole, on the power plant financing, and so on).

        Two alternative DEMO-like power plants are modelled with FRESCO. Both provide 500MWe power and rely on the same plasma model as the pulsed DEMO1 and the steady state DEMO2, the current European design options. As an extension of the previous analyses carried out with FRESCO, which estimated the effects of the pulse duration and H&CD efficiency on the COE of a pulsed power plant, here the effects of the uncertainties related to each specific operative mode are evaluated through stochastic analyses. For example, the uncertainties on the maximum number of cycles the power plant components can withstand deeply affect the COE range. Discussions on which conditions could lower the COE are provided along with considerations on the related probability of this event to occur.

        Speaker: Prof. Giuseppe Zollino (Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete SpA), University of Padova, Department of Industrial Engineering )
      • 1:40 PM
        Structural design and analysis of the feeder in the CFETR CS model coil 2h

        According to the CFETR CS model coil properties test requirements, a large cryogenic test platform will be set up. The feeder, as one of the important components of the platform, should be studied to afford the much valued help for the basic test facilities. The paper focuses on the structural design (location, conductor type, insulation, supports and so on) and FE analysis based on the gravity load, heat load and induction electromagnetic force at room temperature/low temperature.

        Speaker: Liang Guo
      • 1:40 PM
        Structural Integrity Report of Neutron Flux Monitor at occluded EqP#07 (PBS 55.B4.D0) 2h

        ITER is one of the most ambitious energy projects in the world today. The Neutron Flux Monitor (NFM) diagnostics module will be installed on ITER, it measures the total neutron emission, providing the evaluation of the fusion power, and will be positioned in the occluded Equatorial Port #07, more exactly mounted on the inner Bio-shield wall and within penetration from Tokamak Pit to the NB Cell.
        This article describes the overall design of NFM, defines relevant failure modes, criteria of the structural integrity assessment, gives an overview of the structural design criteria used in the structural integrity assessment of the ITER NFM#07system. In the article, NFM is analyzed by the finite element tool ANSYS and the simulation result is evaluated based on analytical design, then summarizes the structural assessments related to each failure mode and the further research about the stress distribution at the large stress intensity area is performed.
        The results and outcomes of this assessment will be taken into account for the design and the future fabrication of the NFM#07 system and its components.

        Speaker: Mr Jun Li
      • 1:40 PM
        Structural Stress Analysis of the CFETR CS Model Coil 2h

        CFETR (China Fusion Engineering Test Reactor) CS (Central Solenoid) model coil made with CICC (Cable in Conduit Conductor) superconductor had been developed in Institute of Plasma Physics, Chinese Academy of Sciences. The highest field of CS model coil is 12T, and the largest magnetic field change rate is 1.5T/S. CS model coil mainly consists of two Nb3Sn inner coils and three outer NbTi coils, buffer zone, feeders and joints, preload supports and so on. The inner diameter of the coil is 1500 mm, and the outer diameter is 3520 mm. Preliminary stress analyses were performed using coupled solver for simultaneous structural, thermal, and electromagnetic analysis. A global finite element model was created based on the initial design geometry data, and it was used to calculate the stresses and deformations of components. Numerical simulations were performed for room temperature condition, cool down to 4.5 K, and the operating current with 47 kA. Computational analysis led to the structural design of the coil, while the optimization was done during design process to verify structural integrity.

        Speaker: Aihua Xu (institute of plasma physics Chinese academy of sciences)
      • 1:40 PM
        Study of electromagnetic effects induced by huge plasma current variations for EAST CS coils quench detection 2h

        The quench detection for EAST superconducting CS coils is considered the most difficult quench detection work because of pulsed operation and the strong coupling with pulsed coils and huge plasma current. The coupling coefficient between superconductor and plasma is not fixed unlike the coupling with pulsed coils because the plasma current configuration is constantly changing. It lead to false quench triggers in case of big disruption due to ignore it in the original compensation system design which means a real challenge.
        In order to discriminate inductive voltage induced by huge plasma current more thoroughly, the active plasma noise compensation system (APC) has been studied and developed on EAST tokamak. Due to the inductance between the plasma and the CS coils is time-varying with different plasma shape and density distribution, the main task of the APC is to get the dynamic compensation coefficient by calculating the time-varying inductance quickly and efficiently.
        In the past few months, calculations and studies, taking different parameters (plasma shape, density distribution, fast plasma events etc.) into consideration, have improved voltage compensation greatly.

        Speaker: Mr Teng Wang (ASIPP)
      • 1:40 PM
        Study of fire impact on detritiation of atmosphere in tritium handling facility: catalytic oxidation of fume gas produced by cable burning 2h

        Fire is one of main scenario for accidental tritium release in tritium handling facilities. Preventing this tritium escape to the environment requires maintaining sub-atmospheric pressure in the affected rooms and detritiation of the gas prior to its discharge. In all gas detritiation systems designed to process a large gas flow (in JET, ITER) first operation stage is catalytic conversion of tritium in hydrogen-containing gases to form of tritiated water. Then the tritiated water is either removed from the gas steam by its drying or detritiated by phase isotopic exchange with liquid water. For handling accidental tritium leak to atmosphere of the room affected by fire the challenge is to ensure operability and efficiency of catalytic recombiner. Gaseous hydrocarbons unavoidably produced during fire present a source of fuel for catalytic recombiner. Their oxidation in exothermic reactions will result in rise of the catalyst’s temperature. Because power supply and I&C cables are most common fire load this study focused on catalyst behavior in oxidation of fume gas produced in cable’s burning in air atmosphere. Cables for power supply and I&C of low-halogen ALSECURE type from NEXANS S.A., France were used in experimental tests. Their burning was performed in electrical furnace under purge with constant flow of ambient air.
        An analysis of the flue gas’s composition showed that the combustion of polymeric materials in cable’s electrical insulation occurs under oxygen starvation conditions. The flue gas contains a large number of different organic products of insulation’s thermal cracking in addition to carbon monoxide. Aerosols were removed from the gas stream by its filtration through HEPA filter. Prior to injection to the catalytic recombiner the gas stream was mixed with an additional flow of atmospheric air and heated to operation temperature of the recombiner (473K). Mixing of gas stream from the furnace with an additional air stream was at several ratio, 1:3.5, 1:8, 1:27 and 1:80. Temperature of of gas stream at recombiner’s inlet and catalyst’s temperature at various points of the recombiner were measured continuously. The recombiner was filled with catalyst which contains 0.5weight % of platinum on alumina.
        It was observed that increase of catalyst temperature depends on mixing ration of the gas from furnace with stream of additional air. For example, at ratio 1:3.5 catalyst temperature rised from 473K to reacged 1570K and the gas’s temperature from 473K to 1270K. At ration 1:8 highest catalyst’s temperature exceeds upper limit for the thermocouples. Recombiner investigation after this test reveals that internal components made of stainless steel were melted down. With further increase of the mixing ratio to 1:27 rise of catalyst’s temperature felled down to 970K. At mixing ratio of 1:80 no temperature rise was detected.
        The experimental data were compared with mathematical modeling of the process. Heat transfer parameters of the recombiner needed for the model were evaluated by measuring temperatures rise in tests with air containing constant concentration of hydrogen in air. Comparison of the experimentally measured and calculated temperatures indicates a satisfactory adequacy of the model for the process interpretation.

        Speaker: Mr Mikhail Rozenkevich (D. Mendeleev University of Chemical Technology of Russia)
      • 1:40 PM
        Study of plasma density effects on the divertor power width of EAST by SOLPS5.0/B2.5-Eirene 2h

        Edge plasma code package SOLPS5.0 is employed to study the effects of upstream density on divertor power width λq for EAST L-mode discharges. The divertor power width, which is an important physical and engineering parameter for diverted tokamak fusion devices, is determined by the parallel and perpendicular transport in the SOL region. Upstream density scan is implemented in the simulation to obtain a wide divertor operational regime from attached divertor regime to detachment. It is found that the divertor power width tends to increase with the increase of plasma density, in consistent with the EAST and multi-machine experimental results. Further analysis shows that the line radiation loss power of CII and CIII in the divertor region move from the far SOL towards the strike point with increasing plasma density. The CII and CIII line radiation dominate the power loss at divertor region. The shift of line radiation loss for CII and CIII may be the main reason for the positive correlation between the edge plasma density and divertor power width. The mechanisms of the changes in carbon impurity radiation and the effects of plasma density on the edge plasma radial and parallel transports will be studied and included in this work to provide a better understanding of the effects of plasma density on divertor power width.
        Further work focusing on the effects of other major plasma parameters such as plasma current and heating power on divertor power width will be carried out.
        This work is supported by National Magnetic Confinement Fusion Science Program of China under contract No. 2014GB11003, 2015GB101003, 2015GB101000; National Nature Science Foundation of China under contract No. 11305206.

        Speaker: Mr Guozhong Deng (Institute of Plasma Physics, Chinese Academy of Sciences)
      • 1:40 PM
        Study on dynamic behavior of EAST upper divertor with vertical displacement events 2h

        The upper divertor of EAST had been upgraded as ITER-like monoblock configuration after 2012 experimental campaign. Compared to previous graphite divertor, it can greatly improve handling capability which is capable of offering an opportunity to study ITER-related plasma physics and engineering and to study the performance of the advanced tungsten divertor at the same time. During plasma operation, the divertor will not only suffer complicated thermal load but also experience huge electromagnetic (EM) force. Considering that the EM forces induced by transient events like current quench (CQ) or vertical displacement events (VDEs) will deposit onto the divertor surface for several millisecond, it will bring a large damage to the divertor which maybe reduce the lifetime of the divertor or break the divertor directly. Eddy current induced by CQ and VDEs will greatly decrease since the monoblock structure is chosen as the upper divertor module. Therefore, halo current induced by VDEs which can reach up to about 50 percent of plasma current is the predominant source of the EM force. And it turns into one of the remarkable threats to capability of the tungsten divertor. Aiming at investigating the response and studying the dynamic behavior of the tungsten divertor components with VDEs, dynamical experiments on W and CuCrZr employed in EAST upper divertor had been done already. And based on the experiment results, constitutive equations involving five material constants were built by using Johnson-Cook model to describe the dynamical properties of these two materials. In this paper, three different halo current distributions on EAST divertor module were calculated respectively according to the halo configurations of vertical displacement events on ITER. Then the EM force and statics analysis were completed by importing the current results for each cases. Finally, dynamic behavior analysis on monoblock structure of the inner plates with 1 MA plasma current was performed and strength analysis and life prediction were done to evaluate the effects on divertors with VDEs based on dynamical properties gotten from the experiments and foregoing calculated EM force. This work was supported by National Magnetic Confinement Fusion Science Program of China (Contract No.2014GB101001).

        Speaker: Mr Xinyuan Qian (School of Nuclear Science and Technology, University of Science and Technology of China)
      • 1:40 PM
        Test results about simple CDA+MIK quench detection method on EAST for ITER Superconducting CS Coils 2h

        The test about simple CDA+MIK quench detection method implemented a total of about 100 charging runs on EAST in the past two years, aiming at verifying this new method whether apply it to ITER cs coils or not.This project is supported by the ITER Organization (IO).
        To obtain this target, the whole instrumentation hardware have been developed and installed on EAST and used to generate experiment data.
        It is real challenge to achieve the plasma discharge due to CS module triad connection configuration, the limitation of peak coil current &voltage,very high accuracy high voltage measurement etc.
        In verification experiments , the low loop voltage plasma discharge with the assistance of the
        electron cyclotron resonance frequency (ECRF) and low hybrid wave (LHW) heating for
        pre-ionization and to following burn-through was the first accomplished successfully plasma scenarios with around 3 second and 250kA pulse plasma current.
        Moreover the numerical quench detection model have been designed and the artificial signals
        created through it should be verified with experimental signal .
        The test have shown that CDA +MIK technology is applicable to the ITER cs coils as a backup method although the co-wound voltage tap sensor has obviously better noise rejection ratio.
        But the sensitivity of this system should be improved greatly in order to meet experiment requirements for ITER operation in the future.
        This paper introduces the test program, typical achieved operation, and the results of preliminary analysis.
        Key words: CDA +MIK technology,EAST, verification experiments

        Speaker: Prof. HU Yanlan (Asipp)
      • 1:40 PM
        The Analysis of Socio-economic Impact on Big Science R&D: Focusing on Fusion R&D Program in Korea 2h

        This paper is focused on the analysis of socio-economic benefits of the ongoing R&D program on big science such as nuclear fusion in Korea. The spillover effects are understood here as positive externalities of publicly funded R&D activities that may be revealed at the companies’ level in the form of newly created knowledge stock; development of innovative products/ processes with broader market applications; strengthening of R&D, manufacturing and marketing capabilities; etc. In addition, this study critically reviews the literature on the socio-economic benefits of publicly funded big science R&D. In that literature, two main methodological approaches have been adopted —surveys and case studies. These studies have also highlighted the importance of spillovers and the existence of localization effects in research. From the literature based on surveys and on case studies, it is clear that the benefits from public investment in big science R&D can take a variety of forms. We classify these into seven main categories, reviewing the evidence on the nature and extent of each type. The results demonstrate that fusion R&D programs have relatively outstanding performance in seven categories: (1) increasing the stock of useful knowledge; (2) training skilled graduates and researchers; (3) creating new scientific means and methodologies; (4) forming networks and stimulating social interactions; (5) reinforcing the capacity for scientific and technological problem-solving; (6) creating new firms; and (7) access to scientific facilities. In particular, those projects were observed to form an industrial ecosystem for nuclear fusion that extends to the accelerator sector, in the category of creating new firms, while making a significant contribution to training talented researchers and expanding social networks as well. We reconsider the rationale for government funding of big science R&D, arguing that the traditional ‘market failure’ justification needs to be extended to take account of these different forms of benefit from big science R&D. The article concludes by identifying some of the policy implications that follow from this review.

        Speaker: Wonjae Choi (National Fusion Research Institute)
      • 1:40 PM
        The Design of a 70kA/20kV Two-section Pyrobreaker for Quench Protection 2h

        Due to the fast response, high reliability and one-time-use property, explosive driven circuit-breaker known as pyrobreaker(PB) has been applied in several high power supply systems. This paper presents the designing process of a new two-section PB with the capability of opening a DC current of 70 kA under a voltage up to 20kV. In accordance with the CFETR (China Fusion Engineering Test Reactor) specifications, this switch will be applied as a back-up breaker for quench protection. Related simulations and calculations about the steady temperature rise, explosion process and commutation process are described in this paper, which fill the gap in the theoretical analyze for the designing of this kind of PBs. It will simplify the existing design method, which normally involves a great number of time-consuming and money-costing experiments. Several tests, including the steady state test and operation test, are conducted on the prototypes. The results verify the feasibility of the new model and demonstrate the reliability of the presented designing process.

        Speaker: Mr Jun He (Institute of Plasma Physics)
      • 1:40 PM
        The deuterium retention behavior in helium irradiated tungsten after plasma exposures in EAST 2h

        Tungsten (W) is considered as the most attractive plasma facing material for future fusion devices attributed to its outstanding properties. During the operation of plasma, helium ions produced by deuterium and tritium fusion reaction will impinge on W, which will raise more complicated tritium retention behavior in W. In the present work, the deuterium retention behavior in helium irradiated W is studied.
        Helium ion irradiation experiments with the fluencies of 3×1015, 3×1016 and 3×1017 ions/cm2 have been performed on recrystallized W, respectively. Transmission electron microscope (TEM) observation suggests that large numbers of dislocation loops are generated in W and the size of the dislocation loop increases with irradiation fluence. To understand the deuterium retention behavior in helium irradiated W formed in tokamak environment, the samples are exposed to the EAST tokamak plasma by Material and Plasma Evaluation System (MAPES). The results of thermal desorption spectroscopy (TDS) indicate that the total deuterium retention increase with the irradiation fluence which can be attributed to the defects induced by helium irradiation. The release peak for deuterium at around 403 K and 994 K are observed in all of the samples. Finally, the deuterium behavior in helium irradiated W is discussed in combination with the TEM results.

        Speaker: Mingzhong Zhao (Institute of Plasma Physics, Chinese Academy of Sciences)
      • 1:40 PM
        The Disturbance Analysis for Ultrasonic Doppler Profile Measurements Through Numerical Simulation 2h

        The liquid PbLi and PbBi alloys have been selected as one of the most promising coolant materials in fusion blanket and accelerator driven subcritical (ADS) reactor, respectively. The velocity distribution in the flow channel of blanket will greatly affect the efficiency of heat transfer and materials corrosion, so it’s necessary to carry out relevant researches to obtain the detailed flow field. Ultrasonic Doppler velocimetry (UDV) is considered as one of the best methods measuring opacity fluid.
        In the process of UDV measurement, the measuring accuracy is one of the most important problems and large bias is observed in measuring results. The reason for the bias is that there are three factors, intersection between measuring volume and wall, refraction of the ultrasonic, ringing effects of piezoelectric element, which mainly influence the measuring accuracy. However, it’s well known that the ultrasonic is a kind of mechanical wave, which will lead to vibration of the medium in the process of propagation. The vibration of medium will cause disturbance to flow field and then lead to measuring bias, so it should be also considered as a key factor in flow field measurement. However, the influence of vibration is usually ignored and no researches were performed on it. Hence, it’s necessary to research about it.
        In this paper, ANSYS analysis is adopted to investigate the effect of vibration of medium. The vibration of medium leads to the acoustic pressure variation of fluid which can deduce the vibration velocity. The acoustic pressure in flowing fluid is calculated through numerical simulation and the two factors, excitation voltages and backing layer, are analyzed. The results show that a large disturbance velocity up to 10mm/s is generated in heavy liquid metal and it’s proportional to excitation voltages. The disturbance is slowly growing up with the thickness of backing layer increasing from 2mm to 6mm, but no obvious variation is obtained when the thickness of backing layer is larger than 6mm. So, it’s very meaningful for bias analysis in flow field measurement based on the results. Besides, it can support the design of UDV sensor and conduct the parameters setting in experiments.

        Speaker: Dr wangli huang
      • 1:40 PM
        The influence of heat transfer on MHD flow in the blanket at high Hartmann Number 2h

        In the breeding blanket fusion reactor, the dynamic viscosity of liquid metal (LM) is influenced by heat transfer, leading to the change of velocity distribution. The effect of heat transfer on magnetohydrodynamic(MHD) flow in a rectangular duct at high Hartmann Number is investigated by a coupling method. In this method, the velocity field is calculated through a second-order projection method, coupling with the temperature distribution calculated by a finite volume method. The numerical result without temperature influence is validated by Hunt's and Shercliff's analytical solutions, and shows very good accuracy. On the basis of the numerical code, the velocity distribution of a Hunt's case with temperature influence is simulated. The simulation result indicates that the velocity field is different from the benchmark solution as the result of the influence of heat transfer.

        Speaker: Jiajia Han
      • 1:40 PM
        The Influences of irradiation defects on mechanical properties for ceramic breeder material Li2TiO3 2h

        Tritium breeder materials are significant for blanket design of fusion reactor. However, during blanket operation, the ceramic breeder materials will be subjected to neutron irradiation which could be detrimental to mechanical properties. Because of its good chemical stability and available tritium release behavior, Li2TiO3 is becoming one of candidate ceramic breeder materials.
        In this study, Li2TiO3 samples are irradiated by 120keV deuterium ions. For sample characterization, the phase composition is investigated by using X-ray diffraction (XRD) before and after irradiation. After deuterium irradiation, the Electron spin resonance (ESR) experiments are employed to investigate the irradiation defects. Micro-hardness measurement is applied to study the changes of mechanical properties. XRD results indicate that Li2TiO3 crystals are damaged by deuterium irradiation, but no new phases are produced. According to ESR experiment, the defect type after deuterium irradiation is E-center which are vacancies trapping one electron. From Vickers hardness measurement, size effect of micro-hardness is observed. The Meyer coefficient obtained in the experiment is 1.65 which is less than 2. The Vickers hardness increases as applied loads decrease which are consistent to Meyer theory. And the Vickers hardness of the Li2TiO3 decreases as irradiation doses increase. The details of this condition are under investigation.

        Speaker: jing wang
      • 1:40 PM
        The offline simulation module of J-TEXT Real-Time Framework 2h

        Tokamak as the most promising way to achieve fusion energy, the reliability of its plasma control system (PCS) is of great importance. J-TEXT Real-Time Framework (JRTF) is the next generation real-time framework for tokamak plasma control of J-TEXT. Introducing object-oriented programming (OOP) technology, JRTF is an advanced framework and can be a future candidate for developing PCS real-time framework of China Fusion Engineering Test Reactor (CFETR). For such huge device, efficient plasma operation and safety is essential. Thus, the complicated PCS needs to be tested to verify the reliability, robustness and availability before plasma operation. This paper proposed an offline simulation module based on JRTF to validate the PCS. OOP technology and flexible configuration using XML files contributes a lot to flexibility and the reusability of the offline simulation module. In the paper, the PCS of J-TEXT tokamak is briefly introduced and analyzed. Then, the conception design of creating offline simulation module’s model is presented. In this section, we introduce a module to simulate the input data through history log and experiment data. The output is record by a simulated output module. Under such circumstance, the program is the same with the online program and only small changes in XML configure files is needed to import such offline simulation modules instead of real hardware I/O modules of reflective memory modules. In addition, a test in vertical field control system is conducted to proving the availability of offline simulation module of J-TEXT Real-Time Framework, an online experiment result has been compared with the offline simulation. Then, a test in horizon field control system is carried out in order to confirm the reusability of the offline simulation module. Also, a Simulink module converter has been developed. It can convert a Simulink designed module into a JRTF offline simulation module, which dynamically generates data based on Simulink designed models instead of using historical data.

        Key Words: JRTF; OOP; offline simulation; J-TEXT; PCS

        Speaker: Mr Yang Li (Huazhong University of Science and Technology)
      • 1:40 PM
        The Protection Strategy Design and Implementation for ITER PF Converter System 2h

        The International Thermonuclear Experimental Reactor (ITER) Poloidal Field (PF) power supply has 14 thyristor based ac/dc converter units to feed six PF coils.PF1 and PF6 coils are both fed by one four-quadrant converter unit respectively, while the PF2-PF5 coils are supplied power by three four-quadrant converter units connecting in series and under sequential control to reduce reactive power for each. The rated parameters for each converter unit is ±55 kA and ±1.05 kV. On account of the complex operation modes and the huge power of converter unit, any fault in converter might lead to escalation of fault and then damage the equipments in case of improper protective action or out of protection. Consequently, the protection strategy is definitely an indispensable and important part for ac/dc converter system. In this paper, the protection strategy is carefully designed based on the fault analysis including the current unbalance between two sharing current bridges, circulation current out of control, bridge and DC terminal over current, and other internal fault, etc. The ITER PF ac/dc converter has been manufactured and its control system prototype has been finished. All protection strategies have been experimented and effectively verified on ITER PF ac/dc converter in Institute of Plasma Physics, Chinese Academy of Sciences(ASIPP).

        Speaker: Dr Liansheng HUANG
      • 1:40 PM
        The vacuum ultraviolet imaging system and its application on EAST 2h

        The Chinese Fusion Engineering Test Reactor (CFETR) is the next device scheduled in the roadmap to realize fusion energy in China[1]. It aims to bridge the gaps between ITER and DEMO. Steady-state operation is one of the key issues of CFETR. The EAST tokamak will provide a long-pulse, high power test bench for advanced scenarios under actively cooled metal wall condition, which will play an important role on supporting the steady-state operation of CFETR.
        In CFETR, operation scenarios of H factor over ELMy H mode are around or higher than 1.0, as listed in Ref.[1]. For long-pulse ELMy H mode discharge, it is a big challenge for the divertor plate to hold the high-level transient heat flux due to the quasi-periodic ELM event. Therefore, ELM control is necessary to realize steady state operation. It is known that ELMs are strongly related with the dynamics of the so-called pedestal region, where steep pressure gradient exists. But the mechanism is still an open topic in fusion research. Experimental studies on the pedestal may be helpful on the understanding of the related physics and benefit the development of efficient method on ELM control.
        A vacuum ultraviolet (VUV) imaging system is developing on EAST tokamak. It aims to measure the evolution of the spatial structures of the pedestal, by selectively measuring emission of 13.5 nm in wavelength, which mainly comes from C VI (one of the intrinsic impurities in EAST). It has been installed on EAST to view the plasma perpendicularly and has been operated in the 2016 experiment campaign. ELM dynamics can be studied by the combination of VUV imaging and the existing visible imaging system, which mainly monitors the bottom of the pedestal and SOL region on EAST. In this work, the hardware of the VUV imaging system and the first results from the VUV imaging data will be presented. In addition, the upgrade of the optics is scheduled for the next campaign, which can be operated to view the plasma tangentially. The proposals of the upgrade will be discussed as well.

        This work is supported by the Natural Science Foundation of China under Contract No. 11605244 and the National Magnetic Confinement Fusion Science Program of China under Contract No. 2014GB106000, 2014GB106001 and 2013GB106000.

        [1] Y.X.Wan.et al 2016 26th IAEA Fusion Energy Conference, Kyoto, 17-22 Oct. 2016, Paper No. OV/3-4

        Speaker: Dr Tingfeng Ming (Institute of Plasma Physics, Chinese Academic Sciences)
      • 1:40 PM
        Thermal hydraulic analysis for one water cooled blanket module of CFETR based on RELAP5 2h

        The Water Cooled Ceramic Breeder blanket (WCCB) is one of the blanket candidates for Chinese Fusion Engineering Test Reactor (CFETR). The conceptual of WCCB for CFETR under 200MW fusion power has been designed based on pressurized water cooled reactor (PWR) technology. RELAP5 code, which is mature and often used in transient thermal hydraulic analysis in PWR reactor, is selected as the simulation tool. In this paper, the nodal model for RELAP5 is developed corresponding to typical WCCB module, i.e. the coolant passages inside the module were nodalized as hydrodynamic components and the associated module components were simulated with one dimensional heat structures. The steady state characteristic under full power is analyzed. The stable fluid and wall temperature distributions and pressure drops are studied. The results are agree with those of the two dimensional CFD analysis by FLUNET. Furthermore, the transient characteristics under three accidental scenarios, in-vessel loss of coolant accident (LOCA), in-box LOCA and ex-vessel LOCA, are analyzed, respectively. Simulation results show all the temperature of structure wall are within the design limitation and the decay heat can be removed by radiation heat transfer in the three LOCA scenarios, also the pressure of the related volume is within the limits.

        Speaker: Mr Shuang Lin (Uiniversity of Sicence and Technology of China)
      • 1:40 PM
        Thermal-hydraulic analysis of high temperature superconducting magnets in CFETR 2h

        The China Fusion Engineering Test Reactor (CFETR) is the next device in the roadmap for the realization of fusion energy in China, which aims to bridge the gaps between the fusion experimental reactor ITER and the demonstration reactor (DEMO). CFETR will be operated in two phases: Steady-state operation and self-sufficiency will be the two key issues for Phase I with a modest fusion power of up to 200 MW. Phase II aims for DEMO validation with a fusion power over 1 GW.[1]
        For saving the cost of construction and meeting both Phase I and Phase II targets with achievable technical solutions, a new design has been made by choosing a larger machine with R =6.6m,/a=1.8m, BT= 6-7T. Over 1GW fusion power can be achieved technically and it is easy to transfer from Phase I to Phase II with the same machine. In order to obtain the maximum magnetic flux of 224 VS from the CS coils in Phase II, the high temperature superconductors of Bi2212 material are used for the CFETR reactor.[2]
        In order to evaluate the feasibility of high temperature superconducting magnets used in CFETR, the 4C code is employed in this paper to analyze the thermal-hydraulic state of the coils.
        The inlet and outlet pressure of helium cooling loops and operational temperature of the magnets is designed. The temperature margin of the superconducting magnets for the reference scenario of plasma discharge is estimated.

        [1] Yuanxi Wan, Jiangang Li et al , Overview of the present progress and activities on the Chinese Fusion Engineering Test Reactor, submitted to Nuclear Fusion.
        [2] Zheng J.X. et al 2016 IEEE Trans. Appl. Supercond. 26(7) 4205505

        Speaker: Dr Junjun Li (Institute of Plasma Physics Chinese Academy of Sciences)
      • 1:40 PM
        Thermo-Hydraulic Performance Testing for Plasma Facing Components by 3D Metal Printing Technology 2h

        3D metal printing technology was selected for the development of fusion divertor research, and the optimization of thermo-hydraulic performance with a water cooling in a Korean heat load test facility by using electron beam (KoHLT-EB). The various cooling design for ITER and DEMO divertor have been fabricated for the enhancement of cooling performance, such as swirl tube and hypervapotron. The main target of this work is the overcoming of fabrication limitations in such cooling devices and the development of new cooling mechanism by using 3D metal printing. And 3D printed divertor mockup was designed and fabricated based on the optimization of 3D cooling structure. The high heat flux test facility KoHLT-EB was used to evaluate the enhancement of cooling capacities. KoHLT-EB was modified in water cooling system for the performance test and the experimental evaluation of the divertor mockups. High heat load for the divertor mockup was applied up to 10 - 20 MW/m2. Also, Thermo-hydraulic and thermomechanical analysis with ANSYS-CFX were performed to determine the test conditions and performance of 3D printed mockups. Present research results will contribute the development of Korean fusion reactor and DEMO program.

        Speaker: Dr Suk-Kwon Kim (Korea Atomic Energy Research Institute)
      • 1:40 PM
        Thermomechanical Assessment of the K-DEMO Divertor Target Applying CuCrZr and RAFM as Heat Sink Materials 2h

        Divertor is one of the most challenging and important components in DEMO plants, since the enormous heat load from plasma applied onto the divertor target must cool down. In a conceptual study of the Korean fusion demonstration reactor (K-DEMO), a water-cooled divertor concept applying the tungsten monoblock type was of primary consideration. The target peak heat flux of 10 MW/m2 was set in steady state operation. To faithfully cool down the heat load, the selection of materials that the divertor is composed of is important as well as the decision of design parameters. Especially, the choice and design of the heat sink material in the divertor target are quite significant because the heat sink directly interfaced with the coolant. Reduced activation ferritic martensitic (RAFM) steel and CuCrZr have been considered the most promising candidates as the heat sink material. The preliminary designs of the high heat flux (HHF) units operating within materials’ own allowable temperature were derived by accomplishing thermohydraulic analyses for RAFM and CuCrZr. Based on the designs of HHF units with a support structure, thermomechanical analyses were carried out. In mechanical analyses, the mechanical loads including the body force, the pressure caused by the coolant, the electromagnetic force were considered as well as the thermal load imported from computational fluid dynamics calculation. In this study, the structural stability of the divertor target applying RAFM and CuCrZr heat sink was accessed by performing the elasto-plastic analysis

        Speaker: Dr Sungjin Kwon (National Fusion Research Institute)
      • 1:40 PM
        Three confinement systems - Spherical Tokamak, Advanced Tokamak and Stellarator: A comparison of key component cost elements 2h

        Since the 1950’s Next Step fusion devices and power plant studies have been developed for a number of magnetic confinement systems but an open question remains…can a magnetic fusion device be simplified to the point where it will be cost competitive and operate with high availability? Concept designs based on the advanced tokamak (AT), spherical tokamak (ST) and the quasi-axisymmetric stellarator (QAS) option have progressed in recent years through a series of PPPL studies with an underlying intent to improve the engineering feasibility of each, giving special attention to concepts that simplify the device configuration and improve maintenance features. For the spherical tokamak option, design details centered on a 3m Fusion Nuclear Science Facility concept that evolved to incorporate vertical maintenance, HTS magnets, a small inboard DCLL blanket and a liquid metal divertor. In collaboration with the K-DEMO and CFETR concept study teams the AT design has evolved to increase plasma component access within a vertical maintenance approach using enlarged TF coils incorporating a low and high field Nb3Sn winding pack that can provide a peak field of 16T. A recent PPPL stellarator study focused on simplifying the stellarator winding topology to improve access to in-vessel components; combining coil optimization with winding surfaces that incorporated geometry constraints specified by engineering. This study centered on a 1000 MW power plant design with a tokamak like vertical maintenance scheme that allows access to remove large segmented internal blanket sectors.
        Results of these three confinement studies will be presented to highlight concepts that simplify each device configuration and improved their maintenance features. Scaling each option to a common 1000 MW net electric power plant mission allows comparisons to be made of key cost elements such as to major core component sizes, sizing of the test cell or external facilities needed for on-site construction or facilities to handle and store activated in-vessel components.

        Speaker: Mr