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27th IAEA Fusion Energy Conference - IAEA CN-258

Mahatma Mandir Conference Centre

Mahatma Mandir Conference Centre

Gandhinagar (nearest Airport: Ahmedabad), India
    • O/1 Opening
      • 1
        Traditional Lightning of the Lamp
      • 2
        Opening Address
      • 3
        Welcome Address
        Speaker: Host Country Representative (India)
      • 4
        Importance of Energy and the role of Nuclear Energy in India’s Energy Mix

        For the past ten years, cumulative average growth rate for electricity generation in India has been close to 6%. During the year 2016-17, total electricity generation was about than 1430 billion kW-hour or TW-hour. It will be more than 1500 TW-hour in 2017-18. Considering rate of economic growth, linkage between economic growth and electricity requirements, increasing urbanisation and current low per capita electricity availability, electricity generation in India is likely to exceed 8000 TW-hour by the middle of this century. Environmental sustainability enjoins on India to generate a significant fraction of the total generation by low-carbon technologies that is nuclear, hydro, solar and wind. Considering that total potential of hydro, solar and wind is only about one-fourth of the projected electricity requirements, nuclear must play a dominant role.

        The talk will explain near- and medium-term plans to accelerate growth in installed nuclear capacity, and provide a glimpse of ongoing research and development aimed at directing growth in installed capacity in the long-term.

        Speaker: Ravi B. Grover (Homi Bhabha National Institute (HBNI), Anushakti Nagar, Mumbai 400094, India)
      • 5
        India’s Quest for Fusion Energy & Road to ITER

        Recognizing the limitations of currently available resources, India’s quest for new energy sources is common for all nations, which are in a state of rapid growth and aspire to seek a respectable place on the global canvas of peaceful and sustainable co-existence. Lack of adequate energy denies opportunity to lead a developed and precludes realization of human potential into what it could have been. The global impact can be gauged from the fact that among the 17 Sustainable Development Goals, spelt out after an extensive study by the UN, the 7th Goal is about ‘affordable and clean energy’. Today or in very near future – the whole world is or will be in a situation that will require every conceivable energy source to be tapped, improved in efficiency, made cost-effective and be equipped with a method to mitigate any adverse impact on the environment.

        In spite of India’s taking significant steps towards tapping every bit of both conventional and renewable energy sources, the demand is much higher than what is currently available and is still growing. If one takes a grand challenge of bringing parity with the world-average for the per capita consumption, the capacity has to be trebled! How fast can one add ~400 GW? No matter what we do and however staggering this figure is, there is no going back from this target. So, an equally challenging problem that emerges is how do we manage to grow on sub-optimal energy supply in the interim period. Techniques to reduce energy consumption by increasing efficiency of various processes need to be developed. For this one needs new tools, materials and research-infrastructure to innovate, improvise and harness the benefits of improvement on a mass-scale. Scales matter; even a tiny saving/improvement for a nation with billion people is quite impactful.

        Advanced technologies like fusion hold the promise but have been traditionally considered too far away for any serious investment so far. The ‘fear factor’ of failure can be overpowering for policy makers. However, it’s time to turn it around and ask ourselves: What difference will it make if fusion reactor works as desired? Well, it will make a tremendous difference. It deserves a try, just for that hope we have. The ITER Project is a collective expression of this global quest for energy in the form of the largest scientific endeavour involving more than half of the world population. The task however, is complex and embeds challenges of extreme kind. But fusion research is also all about innovative ways and can continue to provide the world with spin-offs while it graduates from hydrogen plasma to D-T and from there on to power-reactors.

        India has come a long way in both fusion-science and technology via its well-conceived indigenous as well as collaborative measures. India’s journey began in 1982 and it has grown in several areas of plasma and fusion research. A number of developments has taken place: tokamaks with copper-coils (ADITYA) and with superconducting coils (SST-1) have been built indigenously in the Institute for Plasma Research, Gandhinagar. The scientists have gained enormous experience in plasma operations of these tokamaks as well as in SINP-tokamak, which is located in the Saha Institute of Nuclear Physics, Kolkata. Now, an upgraded ADITYA-U is in place capable of experiments with shaped plasma. A host of auxiliary technologies have been developed and tested with the test-beds created in-house.

        India needs to sustain the momentum of its fusion research to be able to reap the benefits from participation in ITER and to quickly channelize the success of ITER in its vision. The ITER participation has been followed in India with the blanket and the divertor technology development initiatives. Industrial applications of the plasma have come off age and last but not the least, the human resource development has taken place with a strong academic back-bone. In this talk, the above-mentioned developments are overviewed and an outline of the future plan --and how it blends with ITER participation is also presented.

        Speaker: Shishir Deshpande (India)
      • 6
        EU R&D Energy policy and the role of fusion research

        Research and Innovation contributes to several of the ten priorities of the European Commission for 2015-19. The EU's energy research policy contributes, in particular, to provide its citizens and businesses with secure and affordable energy, while also addressing the causes of climate change.
        The next Research and Innovation Programme, covering the period 2021-2027, will build on the success of the current Programme (Horizon 2020) under the guiding principle of 'evolution, not revolution'.
        Intensified international cooperation under the next Programme will ensure that European researchers and innovators have access to and benefit from the world’s best talent, expertise and resources. This will, inter alia, enhance the supply and demand of innovative solutions and promote reciprocal international research partnerships.
        In the area of Fusion Research the implementation of the 'Roadmap to the Realisation of Fusion Energy' will continue to be the priority focus, with a strong and continued support for the construction of ITER and a significant research effort to prepare for DEMO.

        Speaker: Patrick Child (European Commission)
      • 7
        Fusion is our Future: Readiness of the Fusion Technology and the 4th Industrial Revolution

        The time and cost of further increasing the overall readiness level of fusion energy, which requires testing materials under extreme environment, data collection, analysis and new designs, can be significantly reduced with the advent of the fourth industrial revolution. The fourth industrial revolution is on its way. Known as Industry 4.0, it represents the current trend to use automation and data exchange technologies that include cyber-physical systems, the Internet of things, cloud computing and cognitive computing. These technologies are rapidly being developed to perform industry activities. Components of future fusion reactors are expected to be designed and manufactured by using advanced simulation technologies and advanced manufacturing methods. The costs will be further reduced as there will be increased harmonisation of codes and standards. IAEA have already taken steps to ensure that the design rules are harmonised before the technology is commercialised. In case of the fission technology there was commercialisation before harmonisation but for fusion technology it will be harmonisation before commercialisation.

        Speaker: Nawal Prinja (AMEC Foster Wheeler)
    • 10:15 AM
      Coffee Break
    • OV/1 Overview Magnetic Fusion
      • 8
        Progress of ITER-India activities for ITER deliverables: Challenges & Mitigation Measures
        The responsibilities of ITER-India include a mix of precision, heavy, R&D intensive and interface intensive systems, under built-to-print and functional systems category. In several systems, components fall under the category of first of its kind or of the largest kind. The uniqueness of specifications lead to a challenging situation – namely that neither the existing labs or potential suppliers have ever done or encountered such scale-up (either in size/volume, capacity, precision etc.) and do not have even the R&D infrastructure to match the requirements. Under a graded approach a full-scale prototype or at an appropriate scale needs to be developed apart from the testing infrastructure. Facilities have been established to demonstrate the integrated and functional performance in the first if its kind and R&D intense systems, as a risk mitigation strategy. These include, for IC system, an extension of the successful prototype results to demonstrate the production of 2.5 MW in a double chain configuration with a combiner at the output. For Neutral Beams, development of the ion source to realize the stringent parameter space for DNB and development of special technologies, involving special copper alloy Cu-Cr-Zr and special manufacturing technologies, involving high precision of <50 micron over ~ 1 m, as the first of its kind. Additionally, development of SIC compliant isolators and ultimately, setting up of a test facility with an unique attribute to test for the beam transmission Setting up of a special cryogenic test facility to test the performance, against the designed performance for the 4 K, 50 K and 80 K Helium lines with multi process pipes. Development of a SIC compliant 140 kV class feed-through to feed 100 kV for the DNB High voltage power supplies. It is demonstrated that engineering efforts invested at the stage of prototyping have led to a significantly reduced effort in the resolving the technical issues encountered at the stage of production and manifests as a primary risk mitigation strategy in the management of ITER-India procurement. The paper presents the technical achievements and the overall status with an emphasis on the special developments for the first of its kind components to meet the challenging specifications.
        Speaker: Arun Kumar Chakraborty (ITER-India, Institute for Plasma Research)
      • 9
        Progress toward ITER’s First Plasma
        ITER reached 50% completion of the work required to achieve First Plasma in November 2017. Progress has been made on ITER infrastructure since the 2016 FEC, most visibly the construction of many key buildings. The tokamak assembly building and the tokamak bioshield have been completed. The tokamak building will be ready for equipment in 2020. The cryogenic plant and the magnet power supply buildings are complete, and these systems begin commissioning in 2019. The power conversion and distribution area is complete and the component cooling water system building has started construction. Commissioning of these systems starts in 2018. Thus, the physical plant is moving rapidly toward completion, and key systems are entering the commissioning phase. Equally impressive is progress toward manufacturing components of the ITER tokamak. The base and lower cylinder of the cryostat have been assembled on the ITER site. The first of the six modules of the central solenoid has been wound, and three of the six poloidal field coils are presently being wound. The first winding pack of the toroidal field magnets is complete, as is the first casing, which has been verified to meet the high tolerances required (<0.5 mm). The first complete set of parts comprising a vacuum vessel sector has been fabricated and demonstrated to meet strict tolerances (<1 mm). Therefore, the major components of the tokamak have passed into the fabrication phase. The Heating and Current Drive systems (NB, ECH and ICH) are also in the final design phase. The sequence of ITER operation from First Plasma (FP) to the achievement of the Q = 10 and Q = 5 project goals has been consolidated in a Staged Approach. This is a stepwise installation of components and ancillary systems, with all systems installed before the start of the FPO operational phase. The ITER Research Plan has been revised in 2017 to be consistent with the systems available in each phase. Physics R&D focuses on the Disruption Mitigation System, design of the ITER tungsten divertor, and modelling of ITER plasma scenarios. An international Task Force has been established to coordinate R&D on disruption mitigation. Modelling concentrates on the initial phases of the Research Plan and on the Q = 10 scenario, especially plasma termination. The focus is on scenarios that access the H-mode regime in the PFPO-1 and PFPO-2 phases.
        Speaker: Mr Bernard Bigot (ITER Organization)
      • 10
        Overview of the JET preparation for Deuterium-Tritium Operation
        Europe has elaborated a Roadmap to the realisation of fusion energy in which ‘ITER is the key facility and its success is the most important overarching objective of the programme’. We review the contribution of the recent JET experiments with the ITER first wall materials mix, and, the underlying physics understanding to mitigate the scientific risks identified in the ITER research plan. Indeed, together with the ITER scenario development, a strong focus on JET is pursued for addressing ITER needs and developing a sound physics basis for the extrapolation through first principle and integrated modelling: plasma wall interaction, disruption mitigation (installation of a third mitigation valve), H mode access, W-control with higher electron heating (ICRH ITER-like antenna re-instated), pellet ELMs pacing with the optimised vertical high field side track. The JET ITER-Like Wall experiment provides an insight in the coupling between tokamak-plasma operation and plasma-surface interaction in the unique Be/W material environment and acts as test-bed to verify models and modelling tools for ITER. Disruptions are considered as the highest programmatic risk in the ITER Research Plan and experimental and modelling effort in Europe and JET are reviewed. High spatial resolution Doppler backscattering measurements have revealed novel insights into the development of the edge transport barrier. The operational constraints of a metal wall can prevent reaching plasma energy confinement required for QD-T=10 on ITER. Progress on JET to mitigate this risk is reported aiming at maximizing the core and pedestal performance in stationary condition with the W divertor constrain. The measured D-D neutron fluence and gamma dose rates have been successfully compared with simulations performed with the codes used for ITER nuclear safety analyses. Finally, the benefit to further use JET beyond 2020 to train the international ITER team with an upgrade tungsten divertor and with the ITER control tools will be 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 Emmanuel Joffrin (CEA/IRFM)
      • 11
        DIII-D Research Towards Establishing the Scientific Basis for Future Fusion Reactors
        DIII-D research is addressing critical challenges in preparation for ITER and the next generation of fusion devices through a focus on scientific investigation of plasma physics fundamentals, integration of disparate core and boundary plasma physics, and development of attractive scenarios. Fundamental studies show that including the energetic particle “kick” model in transport codes dramatically improves agreement with the measured beam ion profile during strong Alfvénic activity, while dimensionless parameter scaling studies of intrinsic rotation lead to a predicted ITER rotation profile with significant turbulence stabilization. Hard X-ray spectra measurements show that anomalous dissipation of runaway electron (RE) beams is strongest for low energy RE populations, likely due to interactions between the low energy RE population and RE-driven kinetic instabilities. Core-boundary integration studies show that the small angle slot divertor achieves detachment at lower density and extends plasma cooling across the divertor target plate, which is essential for controlling heat flux and erosion. A rotating n=2 RMP combined with a stationary n=3 RMP has demonstrated access to ELM suppression with lower 3D field strength, while at the same time dynamically controlling the divertor heat and particle flux. Other edge studies show that the higher L-H power threshold with RMP fields is potentially due to both 3D density gradient modifications and changes in ExB shear layer topology. Super H-mode experiments in the presence of ELMs have achieved near-record pedestal pressures and record stored energies for the present DIII-D configuration with βN,ped≈1.3, H98y2≈1.6-2.4 and IP≤2.0 MA. In scenario work, the ITER baseline Q=10 scenario has been advanced by adjusting the early current density profile evolution to obtain reproducibly stable operation with ≈0 external torque and without n=1 tearing modes. In the wide pedestal QH-mode regime that exhibits improved performance, the startup counter torque has been eliminated so that the entire discharge uses ≈0 applied torque and the operating space is more ITER-relevant. Finally, the high βP scenario with large-radius ITB has been extended to IP~1 MA (q95~6) with high confinement H98y2~1.6 from both Shafranov shift and negative magnetic shear. Work supported by the USDOE under DE-FC02-04ER54698.
        Speaker: Dr Craig Petty (General Atomics)
    • 12:30 PM
    • OV P1-P8 Overview Posters
      • 12
        DIII-D Research Towards Establishing the Scientific Basis for Future Fusion Reactors
        Speaker: Dr C. Craig Petty (General Atomics)
      • 13
        ELM and ELM-control Simulations
        Speaker: Dr Stanislas Pamela (CCFE - UKAEA)
      • 14
        Experiments in Disruption Avoidance for ITER Using Passive and Active Control
        Speaker: Edward Strait (General Atomics)
      • 15
        NSTX-U Theory and Modeling Results
        Speaker: Dr Jonathan Menard
      • 16
        Overview of first Wendelstein 7-X high-performance operation with island divertor
        Speaker: Prof. Thomas Klinger (Max-Planck Institute for Plasma Physics)
      • 17
        Overview of HL-2A Recent Experiments
        Speaker: Mr Min Xu (Southwestern Institute of Physics)
      • 18
        Overview of new MAST physics in anticipation of first results from MAST Upgrade
        Speaker: Dr James Harrison (CCFE)
      • 19
        Overview of Operation and Experiments in the ADITYA-U Tokamak
        Speaker: Mr Rakesh Tanna (Institute For Plasma Research)
      • 20
        Overview of Physics Studies on ASDEX Upgrade
        Speaker: Dr Hendrik Meyer (UK Atomic Energy Authority)
      • 21
        Overview of Research Results from the Alcator C-Mod Tokamak
        Speaker: Dr Earl Marmar (Mass. Inst. of Technology)
      • 22
        Overview of the First Deuterium Experiment in LHD
        Speaker: Dr Tomohiro Morisaki (National Institute for Fusion Science)
      • 23
        Overview of the JET preparation for Deuterium-Tritium Operation
        Speaker: Dr Emmanuel Joffrin (CEA)
      • 24
        Overview of the KSTAR research progress and future plan toward ITER and K-DEMO
        Speaker: Prof. Hyeon K. Park (UNIST)
      • 25
        Overview of the Validation Activities of IFMIF/EVEDA: LIPAc, the Linear IFMIF Prototype Accelerator and LiFus6, the Lithium Corrosion Induced Facility (
        Speaker: Masayoshi Sugimoto (National Institutes for Quantum and Radiological Science and Technology)
      • 26
        Overview of TJ-II stellarator results
        Speaker: Dr Enrique Ascasibar (CIEMAT)
      • 27
        Physics research on the TCV tokamak facility: from conventional to alternative scenarios and beyond
        Speaker: Dr Stefano Coda (CRPP-EPFL)
      • 28
        Progress of Indirect Drive Inertial Confinement Fusion in the US
        Speaker: Dr John Kline (LANL)
      • 29
        Progress of ITER-India activities for ITER deliverables: Challenges & Mitigation Measures
        Speaker: Arun Kumar Chakraborty (ITER-India, Institute for Plasma Research)
      • 30
        Progress of JT-60SA Project
        Speaker: pietro barabaschi (f4e)
      • 31
        Progress of the CFETR Design
        Speaker: Dr Guoqiang Li
      • 32
        Progress toward ITER’s First Plasma
        Speaker: Mr Bernard Bigot (ITER Organization)
      • 33
        Recent advances in EAST physics experiments in support of steady-state operation for ITER and CFETR
        Speaker: Dr Baonian Wan (Institute of Plasma Physics, Chinese Academy of Sciences)
      • 34
        The Strategy of Fusion DEMO In-vessel Structural Material Development
        Speaker: Dr Hiroyasu Tanigawa (Japan Atomic Energy Agency)
      • 35
        Tokamak research in Ioffe Institute
        Speaker: Mr Nikolai Bakharev (Ioffe Institute)
    • OV/2 Overview Magnetic Fusion
      • 36
        Overview of Physics Studies on ASDEX Upgrade
        The ASDEX Upgrade (AUG) programme, jointly run with the EUROfusion MST1 task force, continues to enhance significantly the physics base of ITER and DEMO. Here, the full tungsten wall is a key asset for extrapolating to future devices. The high overall heating power and flexible heating mix and comprehensive diagnostic set allows studies ranging from mimicking the scrape-off-layer (SOL) and divertor conditions of ITER and DEMO at high density to fully non-inductive operation (q95=5.5, betaN<=2.8) at low density. Higher ECRH heating power <=8 MW, new diagnostics and improved analysis techniques have enhanced the capabilities of AUG. Stable high-density H-modes with Psep/R<=11 MW/m with fully detached strike-points have been demonstrated. The ballooning instability close to the separatrix has been identified as a potential cause leading to the H-mode density limit. Density limit disruptions have been successfully avoided using a path-oriented approach to disruption handling and progress has been made in understanding the dissipation and avoidance of runaway electron beams. ELM suppression with resonant magnetic perturbations (RMP) is now routinely achieved reaching HH98(y,2)<=1.1 giving new insight into the field penetration physics, in particular with respect to plasma flows. Modelling agrees well with plasma response measurements and and a helically localised ballooning structure observed prior to the ELM is evidence for the changed edge stability due to the RMP. Fast measurements of Ti and Er show that the dominantly neoclassical character of Er holds through the ELM recovery. Good agreement of 3D nonlinear MHD modelling with measured ELM crash dynamics is achieved. As type-I ELMs (even mitigated) are likely not a viable operational regime in DEMO studies of no ELM regimes have been extended. Stable I-modes up to n/nGW<=0.7 have been characterised using beta feedback. Despite the sub-Alfvenic beam energy nonlinear energetic particle modes have been observed allowing modelling comparisons under burning plasma conditions. First measurements of the eddy tilt angle of ne fluctuations using correlation Doppler reflectometry as well as the radial correlation and cross-phase angles of Te fluctuations have been achieved, showing good agreement with Gyrokinetic simulations. Dedicated matches of H, D and He discharges (core/edge) highlight important isotope physics.
        Speaker: Dr Hendrik Meyer (UK Atomic Energy Authority)
      • 37
        Recent advances in EAST physics experiments in support of steady-state operation for ITER and CFETR
        Significant progress in the development of plasma control mechanism and understanding the related physics for steady-state advanced high-performance H-mode plasmas have been achieved on EAST since the last IAEA FEC in 2016. First demonstration of >100 seconds time scale long-pulse steady-state scenario with a good plasma performance (H98(y2) ~ 1.1) and a good control of impurity and heat exhaust with the tungsten divertor has been successfully achieved on EAST using the pure RF power heating and current drive. The synergy effect between the ECH and two LHW systems (2.45GHz and 4.6GHz) on EAST has been investigated for enhanced current driven and improved confinement quality. ELM suppression using the n=1 and 2 RMPs has been achieved in EAST and applied for development of the long-pulse H-mode scenario. Reduction of the peak heat flux on the divertor was demonstrated either in a QSF configuration or using the active radiation feedback control. A fully non-inductive steady-state QSF plasma with a duration of 21s has been obtained with a reduced factor of 2.5 on the outer divertor target. Divertor particle and heat flux control using a low n rotating RMP field has been confirmed. Suppression of the W sputtering has been achieved by lowering the edge medium-Z impurity content (C, O, etc) and forming a mixture deposition on the surface of divertor target after the application of lithium wall conditioning. Disruption mitigation experiments have been studied on EAST with the application of the massive gas injection of helium or argon on the termination of initial stable target plasmas. A further increase in the total heating power and improvement of the plasma confinement are expected when using a 0-D model prediction for high bootstrap fraction (fbs~50%) regime. Preliminary 1.5-D simulations suggest that the on-axis ECRH will enhance the deposition of LHW power in the core region, which is beneficial to the effective core heating of the plasma. A new designed lower ITER-like tungsten divertor with active water-cooling is reported. With this upgrade, EAST will be capable to access the high-triangularity small-ELM H-mode regimes and also to perform the target plasma in an advanced X-divertor configuration with assistance from two new water-cooled internal PF coils in support of steady-state operation for ITER and CFETR.
        Speaker: Dr Baonian Wan (Institute of Plasma Physics, Chinese Academy of Sciences)
      • 38
        Overview of the KSTAR research progress and future plan toward ITER and K-DEMO
        The KSTAR research efforts have been focused on expansion of the KSTAR operating space for ITER and K-DEMO[1], validation of critical physics and search of new physics. The operating regimes are high bp (up to ~2.8) non-inductive long pulse up to ~8s, high bN up to ~ 4.3, and k of ~2.16 and a long H-mode discharge over a minute. An improved underlying physics of resonant and non-resonant magnetic perturbation (RMP and NRMP) with the IVCC coils resulted in a long pulse edge localized mode (ELM)-crash suppressed H-mode discharge (~34s).[2] The ELM-crash suppression dependence on critical physical parameters, such as q95, d, andn was extensively investigated. Identification of the role of turbulence induced by RMPs in suppression of the ELMs identified the turbulence flow (ω_(⊥e)) physics during the RMP ramp up and down periods.[3] The study of L/H transition threshold power (Pth) dependence on the non-axisymmetric field found that the Pth is significantly affected by RMPs while NRMP components had no influence on Pth.[4] New physics of interaction between the macroscopic fluctuation (2/1 island) and micro turbulences [5] and validation of q0 issue in sawtooth instability has been explored.[6] Also the misaligned RMP configurations are used to test the divertor heat flux dispersal. [6] A major upgrade plan in KSTAR will be initiated in ~2021 for stable higher beta long pulse operation. Emphasis will be placed on a new actively cooled tungsten divertor possibly with new first wall materials and current drive (LHCD/Helicon). For the search of metal wall materials in the KSTAR upgrade plan, test of specially designed castellated tungsten block tiles of various shapes,[7] impurity transport physics experiments via injection of trace Ar and Kr gases and tungsten dust were performed. *This work is supported by the KSTAR research project funded by Korea Ministry of Science, ICT and Future Planning. References: [1] Y.K. Oh et al., FED 84 344 (2009) [2] Y. In et al., NF 55, 043004 (2015) [3] J. Lee et al., PRL, 117 (7), 075001 (2016), J. Lee et al., ibid (2018) [4] W.H. Ko et al., APS bulletin (2017) [5] J.M. Kwon et al., ibid (2018), M.J. Choi et al., NF 57, 126058 (2017) [6] Y. In et al., ibid (2018) [7] S.H. Hong et al., ibid (2018)
        Speaker: Prof. HYEON KEO PARK (UNIST)
      • 39
        Overview of Research Results from the Alcator C-Mod Tokamak
        Alcator C-Mod has been the only divertor tokamak in the world capable of operating at magnetic fields up to 8 T, equaling and exceeding that planned for ITER. Because of its relatively compact size, C-Mod accesses regimes of extreme edge power density (1 MW/m2 average through the surface of the plasma. H-modes on C-Mod have achieved world-record tokamak volume-average and pedestal plasma pressures (

        above 0.2 MPa, Pped ~ 80 kPa). The highest pedestals are obtained by accessing the super H-mode regime predicted by EPED enabling C-Mod to demonstrate Pped at 90% of the ITER target. Data from a multi-machine database shows that the boundary heat flux width scales inversely with Bp, independent of machine size. The most recent data have extended this scaling to Bp=1.3 T, beyond that envisioned for ITER, and the 1/Bp scaling persists. Based on these results, it is clear that power handling in reactors will be an even bigger challenge than in ITER, arguing for the urgent need for one or more dedicated Divertor Test Tokamaks (DTT). Laser blow-off induced cold-pulses, an enigmatic transient phenomenon that has challenged the standard local-transport paradigm, has been explained by a new local turbulent transport model. Results from the TRANSP power balance code, coupled to the quasilinear transport model TGLF-SAT1, with a new saturation rule that came about from cross-scale coupling physics, and that captures the nonlinear upshift of the critical gradient, are shown to describe the cold-pulse, including the existence of core temperature inversions at low density and disappearance at high density. A Random Forests Machine Learning algorithm, has been trained on thousands of C-Mod discharges to detect disruption events. Disruption evolution time scales on C-Mod are relatively short, and this approach gives reliable warning no more than a few ms before disruption. Warning time-scales on larger plasmas are generally longer, good news for reactor applications. Steady-state tokamak reactors will need high bootstrap fraction, supplemented by RF current drive. Lower Hybrid Current Drive is among the most efficient non-inductive techniques. Recent modeling indicates that moving the launch point to the high field side can have many benefits, including accessibility at lower n|| for higher efficiency.

        Speaker: Dr Earl Marmar (Mass. Inst. of Technology)
      • 40
        Progress of Indirect Drive Inertial Confinement Fusion in the US

        Indirect drive converts high power laser into x rays using small high-Z cavities called hohlraums. X rays generated at the hohlraum walls drive a capsule filled with DT fusion fuel. Recent experiments have produced fusion yields exceeding 50 kJ where alpha heating provides ~3x increase in yield over PdV work. Comparison of the results to the common Lawson criterion suggests the current implosions performance is ~30% from conditions expected to initiate thermonuclear gain. Improvements to close the gap on the last ~30% are challenging requiring optimization of the target/implosions and the laser to extract maximum energy. The US program has a three-pronged approach to maximize target performance each closing some portion of the gap. The first item is optimizing the hohlraum to couple more energy to the capsule while maintaining symmetry control. Novel hohlraum designs are being pursued that enable larger capsule to be driven symmetrically to both reduce 3D effects and increase energy coupled to the capsule. The second issue being addressed is capsule stability. Seeding of instabilities by the hardware used to mount the capsule and fill it with DT fuel remains a concern. Work such reducing the impact of the DT fill tubes and novel capsule mounts such as three sets of two single wire stands forming a cage, as opposed to the thin membranes currently used, are being pursed to reduce the effect of mix on the capsule implosions. There is also growing evidence native capsule seeds such as micro-structure may be playing a role on limiting capsule performance and dedicated experiments are being developed to better understand the phenomenon. The last area of emphasis is the laser. As technology progresses and understanding of laser damage/mitigation advances, increasing the laser energy to as much as 2.6 Megajoulse at 351 nm and increasing the laser power to 600 TW seems possible. This would increase the amount of energy available to couple to the capsule and allow larger capsules potentially increasing the hot spot pressure and confinement time. The combination of each of these focus areas have the potential to produce conditions to initiate thermo-nuclear ignition. The current understanding, status, and plans for near term research in each of these areas will be presented in the context of what is believed to be needed to obtain burning plasmas on NIF.

        Speaker: Dr John Kline (LANL)
    • OV/P P1-P8 Overview Posters
      • 41
        Activity of Indian High Heat Flux Test Facility
        Plasma facing components (PFCs) of ITER-like tokamak are expected to subject high heat loads up to 10MW/m^2 during the tokamak operation in steady state condition. Selection of plasma facing materials/components required extensive qualification and testing for tokamak application. High Heat Flux Test Facility (HHFTF) plays an important role for the qualification and estimation of the life of the component under defined heat load condition. HHFTF with heat flux generated by an electron beam system having 200kW power and 45kV maximum acceleration voltage is in full-fledged operation since 2016. HHFTF is dedicated for high heat flux testing of numerous materials and plasma facing components (small & medium sized) for several thousands of thermal cycles at different heat loads. The facility is equipped with high vacuum pumping systems with pressure regulation, high pressure high temperature water circulation loop and several diagnostics devices such as pyrometers, IR-cameras, video cameras, flow, pressure and temperature sensors . This paper describes the main capabilities of the HHFTF and glimpse of various test performed on plasma facing materials and components.
      • 42
        Advances in Fusion-Relevant Physics on the Large Plasma Device
        Studies of turbulence and transport in the Large Plasma Device (LAPD) have: documented the role of drift-Alfvén waves and flows in avalanche events; revealed a new instability in the edge of increased β plasmas; and demonstrated an interaction between ICRF waves and edge turbulence, leading to strong modulation of the former and enhancement of the latter. Intermittent collapses of the plasma pressure profile (avalanches) are observed with off-axis heating in LAPD and are associated with unstable drift-Alfvén waves. Flows play a critical role in the dynamics, in particular in the onset of the drift-Alfvén waves and avalanches through the interplay of the stabilizing flow shear and the destabilizing pressure gradient. Active control of the flows is obtained using biasing; this leads to control over the size and frequency of avalanches. With controlled flows, a regime is found in which avalanches are absent. Strongly electromagnetic turbulence, identified as being due to a new instability, the Gradient-Driven Drift Coupling Instability (GDC), is observed in the edge of increased β plasmas in LAPD. As the plasma beta is increased (up to 15%), magnetic fluctuations are observed to increase substantially, with δB/B ~ 1% at the highest β, while density fluctuations decrease slightly. Parallel magnetic fluctuations are observed to be dominant at the highest β, with δB∥/δB⊥~2. Comparisons of the experimental data with linear and nonlinear GENE simulations of the GDC yield good qualitative and quantitative agreement. An experimental campaign on the physics of ICRF waves on LAPD has established a correlation between strong modulation of core coupled fast waves and edge density fluctuations, both of which increase with antenna power. Strong low-frequency modulation of coupled fast wave power is observed via direct measurement of the core RF waves with magnetic probes. This modulation is well correlated with low-frequency edge density fluctuations associated with drift waves (measured with Langmuir probes). The amplitude of the RF modulation and the amplitude of edge density fluctuations in the drift wave frequency range both grow with increasing RF power, suggesting some nonlinear coupling between the edge drift waves and large amplitude fast waves in the core region.
        Speaker: Prof. Troy Carter (University of California, Los Angeles)
      • 43
        Design, development and recent experiments at the CIMPLE-PSI device
        It is important to understand how the plasma with unparalleled heat (~10MWm-2) and ion (~1024 m-2s-1) flux will interact with the tungsten walls in the ITER tokamak, more specifically at the Divertor region of the fusion machine. Several linear magnetized plasma devices have been developed worldwide that reproduced ITER Divertor like extreme conditions for studies on relevant plasma surface interaction (PSI) issues under simulated plasma conditions. The “CPP-IPR Magnetized linear Divertor Plasma Experiment for Plasma Surface Interaction” or CIMPLE-PSI is one of the few Tokamak Divertor simulator devices that successfully reproduces both ITER-like ion and heat flux values, whose design, development and recent commissioning will be presented in this paper. A segmented plasma torch produced high-density plasma jet collimated with a maximum 0.45 Tesla axial magnetic field propagates at few Pascal chamber pressure that is maintained by four numbers of roots vacuum pumps with 14,000 m-3/h pumping capacity that interacts with a remotely placed tungsten target under controlled experimental conditions. The paper will report detailed diagnostics of the plasma jet through optical emission spectroscopic techniques (1.33 m McPherson spectrometer), a retractable Langmuir probe and water calorimeters while operating the plasma with helium and hydrogen mixture of gases. During recent PSI experiments in this device under irradiation of pure helium plasma (exposed for 1800 seconds under 0.3T magnetic field, target biased to -45 V), we had witnessed (FESEM, HRTEM) formation of surface nanotendrils in profuse amounts, recent characterization results from which also will be presented here.
        Speaker: Dr Mayur Kakati (Centre of Plasma Physics-Institute for Plasma Research, Sonapur, Assam, India)
      • 44
        Formation of Hot, Stable, Long-Lived Field-Reversed Configuration Plasmas on the C-2W Device
        TAE's research has been devoted to producing a high temperature, stable, long-lived field-reversed configuration (FRC) plasma state by neutral-beam injection (NBI) and edge biasing/control. C-2U experiments have demonstrated drastic improvements in particle and energy confinement properties of FRC's, and the plasma performance obtained via ~10 MW NBI has achieved plasma sustainment of up to 5 ms and plasma lifetimes of 10+ ms [1]. The emerging confinement scaling, whereby electron energy confinement time is proportional to a positive power of the electron temperature T_e, is very attractive for higher energy plasma confinement; accordingly, exploration of the observed scaling law at 10× higher T_e is one of the key research objectives. TAE's new experimental device, C-2W (also called "Norman"; the world's largest compact-toroid device), has been constructed with the following key subsystem upgrades from C-2U: (i) higher injected power (up to ~21 MW), optimum and adjustable energies (15-40 keV), and extended pulse duration (up to ~30 ms) of the NBI system; (ii) installation of inner divertors with upgraded edge-biasing electrode systems, which allow for higher biasing voltage and longer pulse operation (30+ ms); (iii) increased overall stored energy in the FRC formation pulsed-power system; (iv) fast external equilibrium/mirror-coil current ramp-up capability for plasma ramp-up and control; (v) installation of trim/saddle coils for active feedback control of the FRC plasma; and (vi) enhanced overall diagnostic suite. A remarkable side note is the fact that TAE spent only ~1 year to construct the C-2W device and produce its first plasma. C-2W experiments have already produced a dramatically improved initial FRC state after translation and merging. As anticipated by design and also in our simulations, the merged initial FRC state exhibits much higher plasma temperatures (T_e >250 eV; total electron and ion temperature >1.5 keV) and more trapped flux, providing a very attractive target for effective NBI. Edge biasing/control experiments have also demonstrated stabilization of the FRC, thereby improving plasma confinement and prolonging FRC lifetime (up to ~10 ms), in which overall plasma performance is already equivalent to or better than that obtained in C-2U. [1] H. Gota et al., Nucl. Fusion 57, 116021 (2017).
        Speaker: Hiroshi Gota (TAE Technologies, Inc.)
      • 45
        Fusion Energy Development Applications Utilizing the Spherical Tokamak and Associated Research Needs and Tools
        The spectrum of scientific and technological gaps that must be closed to achieve practical fusion energy using magnetically confined plasmas has been extensively documented. A common barrier to narrowing or closing these gaps is the scale and cost of fusion facilities needed to address the gaps. The low-aspect-ratio “spherical torus/tokamak” (ST) is being explored world-wide as a potentially attractive configuration for closing scientific gaps and demonstrating technical achievements on a path toward a demonstration power plant and as a more compact and/or modular fusion power source in its own right. The international fusion community is presently assessing the suitability of the ST for applications to advance fusion energy development including: developing solutions for the plasma-material-interface (PMI) challenge,fusion-fission hybrid systems, developing fusion components capable of withstanding high fusion neutron flux and fluence including breeding blankets, demonstrating electricity break-even from a pure fusion system, and electricity production at industrial levels in modular and/or larger-scale fusion power plants. This range of fusion energy development applications utilizing the ST will be described, common application-driven research needs discussed, upcoming and recently achieved ST facility capabilities and relevant highlights described, and near-term prioritized ST research directions supporting longer-term fusion energy development applications presented.
        Speaker: Dr Jonathan Menard (Princeton Plasma Physics Laboratory)
      • 46
        ITER-relevant research on the COMPASS tokamak
        In the years 2016-17 the research on the COMPASS tokamak was focused on support of solution of the key challenges for the design and operation of ITER and next-step devices. This included mainly installations and upgrades of state-of-art edge plasma diagnostics, such as the new divertor probe array and the High Resolution Thomson scattering. Strong emphasis was placed on development of relevant scenarios: discharges with impurity seeding at different locations in the divertor were focused on accessing partially detached plasmas. It was demonstrated that such regime can be achieved, when nitrogen is injected at the outer target, although drop of upstream pressure was also observed. Measurements of peak ELM energy densities in the divertor complemented the existing scaling by Eich et al. and confirmed the validity of proposed model. The same set of probes mounted on the horizontal reciprocating manipulator allowed to perform upstream measurements of power decay length during ELMs. It was observed, that the power decay length exhibits a significant broadening (factor of 4) compared to the inter-ELM value. Dedicated campaigns were focused on experiments with runaway electrons (RE), studying the role of different gases (Ar, Ne, D) on the generation and mitigation of the RE beam. It appeared that an intensive injection of D may significantly slow down the current decay of RE beams triggered by Ar or Ne injection in the discharge phase with practically zero external loop voltage. On request of the ITER Organization, a unique system of COMPASS High-field-side (HFS) Resonant Magnetic Perturbation (RMP) coils was used to study the effects of Error Fields (EF) originating from misalignment or inclination of central solenoid on the plasma performance like L-H transition, H-mode performance degradation, locked modes, etc., The experimental observations are compared to predictions of the ideal MHD code IPEC. This study is being carried out in collaboration with ITER Organization and Princeton Plasma Physics Laboratory, USA. In all the aforementioned fields, a significant progress under the joint EUROfusion effort has been achieved in 2016-17 and the results complemented and broadened the existing databases.
        Speaker: Dr Michael Komm (Institute of Plasma Physics of the Czech Academy of Sciences)
      • 47
        Overview and Status of Direct-Drive Inertial Confinement Fusion in the United States
        Direct-drive (DD) inertial confinement fusion (ICF) offers a potential path for high yield and ignition. Two approaches – Laser Direct Drive (LDD), being pursued primarily at the OMEGA laser and the National Ignition Facility (NIF), and the Magnetized Liner Inertial Fusion (MagLIF), being pursued primarily at the Sandia National Laboratories, will be discussed in this talk. In LDD nominally identical laser beams are used to drive an imploding cryogenic shell on OMEGA to obtain high pressures and temperatures in a hot spot surrounded by a cold fuel. The goal is to obtain ignition-relevant hot-spot pressures in OMEGA-scale cryogenic deuterium–tritium layered implosions. Hot-spot pressures up to 567 Gbar have been demonstrated in these implosion experiments. In addition, recent implosion results when scaled to NIF energies are predicted to produce fusion yields approaching 300 kJ. Experiments on the NIF are additionally used to address the MegaJoule-scale physics such as laser coupling and preheat from energetic electrons. In the MagLIF approach, a 1kJ, 1TW laser pulse is used to preheat the plasma just as the 16 MA current begins to quasi-adiabatically compress the pre-magnetized deuterium. Promising ion temperatures (~ 3KeV) and neutron yields (5x1012 DD neutrons) have been obtained with MagLIF experiments at relatively low implosion speeds of ~7  106 cm/s, indicating successful magnetic flux compression and decreased thermal conductivity losses required for ignition. Ignition remains a challenge for both the direct-drive approaches, including improving understanding of the plasma conditions, controlling nonuniformity, improving laser coupling, and developing enhanced diagnostics. The motivation, challenges, and status of direct-drive research in the United States is presented in this talk. This material is based upon work supported by the Department of Energy National Nuclear Security Administration under Award Number DE-NA0001944, the University of Rochester, and the New York State Energy Research and Development Authority. The support of DOE does not constitute an endorsement by DOE of the views expressed in this article.
        Speaker: Bahukutumbi P. Radha (University of Rochester)
      • 48
        Overview of diagnostics upgrade and experiment progress on KTX
        The Keda Torus eXperiment (KTX) is a new built middle-size reversed field pinch (RFP) device at the University of Science and Technology of China. After the long time conditioning, the favorable wall condition is achieved for implementing experiment on KTX. In present, the maximum plasma current can reach 200kA, the discharge length is beyond 20ms and the duration of typical reversed field pinch state is 2.0ms. The diagnostics on KTX has been greatly developed: 1) Total number of DAQ channel has been upgraded to 960; 2) Terahertz interferometer has been upgraded to 7 chords to obtain density and current profiles; 3) Thomson scattering with 3Joule Laser is undergoing commissioning; 4) 3D Langmuir probe system has been developed for the electromagnetic turbulence measurement; 5) 3D double-foil soft X-ray diagnostics are mounted on two poloidal sections for 3D MHD research; 6) Edge capacitive probe has been installed for the radial electrical field measurement; 7) multi-channel spectrograph system has been built for detecting impurities of carbon and oxygen. After the wall condition improvement and diagnostics upgrade, many early research such as the 3D RFP physics and electromagnetic turbulence, etc., have been conducted on KTX. The forward scattering is observed by the interferometer system which shows the potential for turbulence research with wider spectrum after improving the beam size and acceptance angle of the diagnostic beam through plasma. The research on MHD activities related with 3D RFP physics on KTX is intensely carried out with the capability upgrade of magnetic field measurement, soft X-ray tomography and high-speed visible imaging system. The electromagnetic turbulence is tentatively investigated on KTX. The 3D spectra characters of electromagnetic turbulence are firstly measured with classical two-point technique by the 3D Langmuir probe arrays, particularly in the small wavenumber range, providing the new prospect of electromagnetic turbulence in RFP plasmas. The confinement improvement of turbulence suppression is achieved in Biasing electrode experiment. The resistive MHD modelling of QSH state using NIMROD is setup in the KTX regimes. In the next step, higher performance plasma of KTX with larger plasma current, higher temperature and longer energy confinement time is expected with the capacity upgrade in the second phase.
        Speaker: Prof. Wandong Liu (University of Science and Technology of China)
      • 49
        Overview of the FTU results
        Since the 2016 IAEA FEC Conference, FTU operations have been mainly devoted to experiments on runaway electrons and investigations about a tin liquid limiter; other experiments have involved the elongated plasmas and dust studies. The tearing mode onset in the high density regime has been studied by means of the linear resistive code MARS and the highly collisional regimes have been investigated. New diagnostics, such as a Runaway Electron Imaging Spectroscopy system for in-flight runaways studies and a triple Cherenkov probe for the measurement of escaping electrons, have been successfully installed and tested, and new capabilities of the Collective Thomson Scattering and the Laser Induced Breakdown Spectroscopy diagnostics have been explored.
        Speaker: Dr Gianluca Pucella (ENEA)
      • 50
        Overview of the Recent Experimental Research on the J-TEXT Tokamak
        Recent J-TEXT research has highlighted the significance of the role that non-axisymmetric magnetic perturbations, so called 3D magnetic perturbation (MP) fields, play in fundamentally 2D concept, i.e. tokamak. In this paper, the J-TEXT results achieved over the last two years, especially on the impacts of 3D MP fields on magnetic topology, plasma disruptions, MHD instabilities, and plasma turbulence transport, will be presented. On J-TEXT, the resonant MPs (RMPs) system, capable of providing either a static (DC) or a high frequency (up to 6 kHz) rotating (AC) non-axisymmetric MP field, has been upgraded by adding a new set of 12 in-vessel saddle coils, and the total number of in-vessel RMP coils increases from 12 to 24 (3 rows × 8 columns). The new capabilities advance J-TEXT to be a forefront of international magnetic fusion facilities, allow a flexible study of 3D effects in a tokamak. Both density and plasma rotation dependences of the m/n = 2/1 locked mode threshold, *B*_r,(2,1)^c, have been investigated systematically on J-TEXT. Recent experimental results showed the *B*_r,(2,1)^c scales linearly on the toroidal rotation, and depends weakly on plasma density, n_e. The fast rotating RMP field has been successfully applied for avoidance of mode locking and the prevention of plasma disruption. Remarkably, the rotating tearing mode was completely suppressed by the electrode biasing (EB) in addition to the RMP field. The impacts of 3D magnetic topology on the turbulences have been investigated on J-TEXT. It is found that the fluctuations of electron density, electron temperature, and plasma potential can be significantly modulated by the island structure, and a larger fluctuation level appears at the X-point of islands. The suppression of runaway electrons (REs) during disruptions is essential to the operation of ITER, and it has been reached by utilizing the 3D magnetic perturbations on J-TEXT. The NIMROD simulation indicates that the strong stochastic in the whole plasma cross-section expel out the runaway seed and results in runaway free disruptions on J-TEXT. This may provide an alternative mechanism of runaway suppression for large-scale tokamak and ITER.
        Speaker: Dr Nengchao Wang (International Joint Research Laboratory of Magnetic Confinement Fusion and Plasma Physics (IFPP), AEET, SEEE, CnHUST)
    • 4:10 PM
      Coffee Break
    • FIP/1 ITER Technology
      • 51
        Completion of the first TF Coil Structure of ITER
        This paper reports the completion of the first Toroidal Field Coil Structure (TFCS) of ITER of which Japan Domestic Agency (JADA) takes 100% share on procurement responsibility. The major technical challenges of the TFCS of ITER are (i) new material development for high ductility under cryogenic temperature (4K), (ii) application of partial penetration welding (PPW), (iii) welding deformation control, (iv) special ultrasonic test (UT) development considering attenuation by weldment of austenitic stainless steel and (v) fitting of large (16m x 9m) complex D-shape structure for closure welding (CW) within tight tolerance of a range of 0.5mm. Developed solutions for these challenges lead us to the successful completion of the first TF coil structure. ITER TFCS requires both high sterength and high ductility at cryogenic 4K temperature. For this, special austenitic stainless steel was newly developed. This developed material is used at inboard straight leg which supports most severe electro-magnetic force of 600MN. JADA also developed new method to keep fracture toughness requirement finding strong correlation between fracture toughness and Md30. The PPW is applied to attachments with fracture mechanics assessment using data of crack propagation under cryogenic temperature. JADA developed new UT method to apply for PPW joints, which is to obtain the position of tip of discontinuity with its continuous length and then to assess the area of deviation from nominal position compared with maximum allowable defect area obtained by fracture mechanics. The mechanical property at cryogenic temperature was checked for fracture toughness as well as yield strength of welded joints. The welding deformation was controlled by special welding process to keep balance of angular distortion. The attenuation of UT beam in the weld is compensated by transfer correction factor obtained by welding test piece made with actual material and weld metal. The TFCS was finally machined on its closure welding root that has a range of 0.5mm gap and misalignment tolerance between two welding edges. For this, quite precise control was performed such as temperature control/compensation or setting of the machining position target based on as-machined data of the other parts. As a result of fitting test, CW roots were fit and the first TFCS of the ITER was successfully completed.
        Speaker: Dr Masataka Nakahira (JpQSTNFI)
      • 52
        Completion of 1st ITER gyrotron manufacturing and 1 MW test result & Outcome of R&D program for ITER ICRF Power Source System & Recent progress in the development of the European 1 MW, 170 GHz CW gyrotron for ITER
        A. This paper presents a summary of recent progress pertaining to the manufacturing and inspection of ITER gyrotrons and their operation system in QST. Major achievements are as follows: (i) The final design of ITER gyrotron was accomplished and manufacturing of 2 ITER gyrotrons was finished. Then their factory acceptance test (FAT) in QST has started with ITER relevant high voltage power supply configuration. The 1st ITER gyrotron has achieved 1 MW output power for 10 s pulse and 200 kW operation for 300 s which suggests thermally stable condition and sufficient cooling performance for 1 MW long pulse operation; (ii) The coupling function of gyrotron power into the transmission line (TL) waveguide was improved and calculation result of coupling efficiency was increased as high as 96.9% for the fundamental mode purity in waveguide inlet which could produce the sufficient LP01 mode purity in whole EC H&CD system. These results lead to success of ITER EC H&CD system construction toward first plasma. B. As a part of in-kind contribution, India is responsible to deliver nine numbers (1 Prototype + 8 series production) of RF Sources to ITER system, each having power handling capability of 2.5 MW/CW at VSWR 2:1 in the frequency range 35 – 65 MHz or 3.0 MW/CW at VSWR 1.5:1 in the frequency range 40 – 55 MHz, along with other stringent requirements. As there is no such amplifier chain able to meet the output power specifications as per ITER need, the RF source consists of two parallel three-stage amplifier chains, with a combiner circuit on the output side. This kind of RF source is unique in terms of its stringent specifications. A voluntary R&D program by India has been initiated for establishing the high power technology prior to Prototype and series production, using Diacrode and Tetrode tubes. In this program, single chain experimentation at 1.5MW for 2000s is conducted for the frequency range 35-65 MHz up to VSWR 2:1, with any phase of reflection coefficient. The main objective for the R&D test is to confirm the system performance for the power, duration and frequency range as per ITER need and to check the reliability of both the tube and the amplifier with matched as well as with mismatched load (up to VSWR 2:1), which essentially simulates the plasma load condition. To support the R&D program, a dedicated high power test facility has been developed at ITER-India to test RF amplifiers based on both the technologies. For Diacrode based system, high power ITER relevant tests completed in 2016 and reported elsewhere [1]. Over the past two years, assembly and integration of R&D RF source using Tetrode technology at Indian test facility is completed with validation of all the relevant sub-systems/systems as standalone mode. The high power RF test using Tetrode based RF amplifier achieved 1.7MW of power for 3600s duration at 36 MHz. For other ITER operating frequencies, the system was operated at 1.5MW/2000s successfully. This paper reports commissioning of RF amplifier using Tetrode technology with various operating scenarios, dissipation limit, safety system and challenges faced during high power operation at Indian test facility and describes the final outcome of R&D activity. C. The European 1 MW, 170 GHz industrial prototype CW gyrotron for ECRH&CD on ITER is a conventional (hollow-cavity) gyrotron, which is being developed by the European GYrotron Consortium (EGYC) in cooperation with the industrial partner Thales Electron Devices (TED), under the coordination of the European Joint Undertaking for ITER and the Development of Fusion Energy (F4E). The CW industrial prototype was extensively tested in the short-pulse regime (with pulse length up to 10 ms) and operated under long-pulse conditions with pulse lengths of up to 180 s, which is the limit at the High-Voltage (HV) power supply currently available at KIT. In this contribution we report on the performance of the tube during the long-pulse operation at the KIT test facility, details regarding the operating points are presented and the long-pulse phase of the experiments with pulses up to 180 s is analyzed.
        Speaker: Dr Yasuhisa Oda (National Institute for Quantum and Radiological Science and Technology)
      • 53
        Outcome of R&D program for ITER ICRF Power Source System
        As a part of in-kind contribution, India is responsible to deliver nine numbers (1 Prototype + 8 series production) of RF Sources to ITER system, each having power handling capability of 2.5 MW/CW at VSWR 2:1 in the frequency range 35 – 65 MHz or 3.0 MW/CW at VSWR 1.5:1 in the frequency range 40 – 55 MHz, along with other stringent requirements. As there is no such amplifier chain able to meet the output power specifications as per ITER need, the RF source consists of two parallel three-stage amplifier chains, with a combiner circuit on the output side. This kind of RF source is unique in terms of its stringent specifications. A voluntary R&D program by India has been initiated for establishing the high power technology prior to Prototype and series production, using Diacrode and Tetrode tubes. In this program, single chain experimentation at 1.5MW for 2000s is conducted for the frequency range 35-65 MHz up to VSWR 2:1, with any phase of reflection coefficient. The main objective for the R&D test is to confirm the system performance for the power, duration and frequency range as per ITER need and to check the reliability of both the tube and the amplifier with matched as well as with mismatched load (up to VSWR 2:1), which essentially simulates the plasma load condition. To support the R&D program, a dedicated high power test facility has been developed at ITER-India to test RF amplifiers based on both the technologies. For Diacrode based system, high power ITER relevant tests completed in 2016 and reported elsewhere [1]. Over the past two years, assembly and integration of R&D RF source using Tetrode technology at Indian test facility is completed with validation of all the relevant sub-systems/systems as standalone mode. The high power RF test using Tetrode based RF amplifier achieved 1.7MW of power for 3600s duration at 36 MHz. For other ITER operating frequencies, the system was operated at 1.5MW/2000s successfully. This paper reports commissioning of RF amplifier using Tetrode technology with various operating scenarios, dissipation limit, safety system and challenges faced during high power operation at Indian test facility and describes the final outcome of R&D activity. [1] Aparajita Mukherjee et. al., Progress in High Power Test of R&D Source for ITER ICRF system, 26th IAEA FEC 2016, 17-22 Oct 2016, Kyoto, Japan
        Speaker: Dr Yasuhisa Oda (Japan Atomic Energy Agency)
      • 54
        Recent progress in the development of the European 1 MW, 170 GHz CW gyrotron for ITER
        The European 1 MW, 170 GHz industrial prototype CW gyrotron for ECRH&CD on ITER is a conventional (hollow-cavity) gyrotron, which is being developed by the European GYrotron Consortium (EGYC) in cooperation with the industrial partner Thales Electron Devices (TED), under the coordination of the European Joint Undertaking for ITER and the Development of Fusion Energy (F4E). The CW industrial prototype was extensively tested in the short-pulse regime (with pulse length up to 10 ms) and operated under long-pulse conditions with pulse lengths of up to 180 s, which is the limit at the High-Voltage (HV) power supply currently available at KIT. In this contribution we report on the performance of the tube during the long-pulse operation at the KIT test facility, details regarding the operating points are presented and the long-pulse phase of the experiments with pulses up to 180 s is analyzed.
        Speaker: Dr Yasuhisa Oda (Japan Atomic Energy Agency)
      • 55
        Technologies for realization of Large size RF sources for –ve neutral beam systems for ITER -Challenges, experience and path ahead & Progress in the ITER Neutral Beam Test Facility & Demonstration of 1 MV vacuum insulation for the vacuum insulated beam source in the ITER NB system

        A. Technologies for manufacturing of small and medium size Ion source (upto four RF driver) for positive and negative neutral beam systems have been evolved over last many decades and such ion sources are being successfully operated at various experimental facilities across the world. However, as the need arises for the larger size ion sources (eight driver) for ITER diagnostics and heating neutral beam systems, several existing manufacturing technologies and considerations have to upgraded and re-evaluated to qualify them for (1) highest vacuum quality class (2) nuclear environment.
        Diagnostic Neutral Beam (DNB) source is the first candidate in a family of such big size ion sources, being manufactured according to the ITER specification with ‘re-evaluated’ manufacturing technologies and it throws light on many unforeseen challenges as manufacturing progresses. The nature of challenges are mainly related to usage of the material with radioprotection requirement (i.e restricted contents of Co wt%0.05, Nb wt %0.01 and Ta wt %0.01), special requirements on weld joint configuration to enable full penetration with 100% volumetric inspectability, dissimilar material welding technologies, machining process development to meet stringent dimensional accuracies (in the range of 10-50 microns) of individual ‘angled’ grid segment to achieve overall alignment of +/-0.2mm, electro-deposition of copper with thickness>3mm over the angled surfaces with control over distortion, vacuum brazing, restricted usage of Silver for brazing and plating purpose, development of electrical isolators with customized electrostatic shield, threaded connection between metal and alumina load carrying capacity of 10kN with electrical isolation of 140kV in vacuum.
        The paper shall present experience gathered in development of above mentioned manufacturing technologies, the methodology adopted for mitigating the practical limitations, prototyping to establish and qualify the manufacturing procedure, evaluating the non-conformities, assessment of deviation proposals, in compliance with ITER specifications. In summary, the experience generated during the manufacturing of DNB Beam source, presented here, is aimed to help in generating the recipe manufacturing and providing the ‘re-evaluated’ technical specifications for upcoming ITER neutral beam sources.

        B. The ITER Heating Neutral Beam (HNB) injectors, one of the tools necessary both to achieve burning conditions and to control plasma instabilities, are characterized by such demanding parameters as to require the construction of a test facility dedicated to their development and optimization. This facility, called NBTF, is in an advanced state of realization in Padua (Italy), with the direct contribution of the Italian government, through the Consorzio RFX as the host entity, IO, the in kind contributions of three DA’s (F4E, JADA, INDA) and the technical and scientific support of various European laboratories and universities. The NBTF hosts two experiments: SPIDER and MITICA. The former is devoted to the optimization of the HNB and DNB ion sources and to the achievement of the required source performances. It is based on the RF negative Ion Source concept developed at IPP (Garching). MITICA is the full size prototype of the ITER HNB, with an ion source identical to the one used in SPIDER. The construction and installation of SPIDER plant systems was successfully completed with their integration into the facility, followed by integrated commissioning with control (CODAS), protection and safety systems. The mechanical components of the ion source have been installed inside the vessel and connected to the plants. Finally, the integrated commissioning of the whole system ended positively and the first experimental phase began. Also the realization of the MITICA project is well advanced, although the completion of the system and its entry into operation is expected in 2022 due to the long procurement times of the in-vessel mechanical components. In particular, the power supply designed to operate at 1MV are in an advanced phase of realization, all the high voltage components have been installed and the complex insulation test phase has begun in 2018. Furthermore, all the other auxiliary plant systems are being installed and / or undergoing testing. This paper gives an overview of the progress of the NBTF realization with particular emphasis on issues discovered during this phase of activities and to the adopted solutions in order to minimize the impact on the schedule while maintaining the goals of the facilities. Finally, the first results obtained with SPIDER experimentation and with the 1MV insulation tests on the MITICA HV components will be presented.

        C. For the ITER neutral beam (NB) system, a measure to achieve the 1 MV vacuum insulation of the beam source have been developed. For this purpose, design basis for 1 MV vacuum insulation has been developed by integrating previous empirical scaling for plane and coaxial electrodes and new scaling for area with locally-concentrated electric field. Consequently, as the measure, the beam source is surrounded by more than three intermediate electrostatic shields instead of single gap to sustain 1 MV. Effectiveness of the shields designed by the design basis was experimentally verified by using a part of the beam source. The voltage holding capability has been significantly improved from 0.7 MV to 1 MV. This result ensures the 1 MV vacuum insulated beam source in the ITER NB system.

        Speaker: Mr Jaydeepkumar Joshi (ITER-India (Institute for Plasma Research))
      • 56
        Progress in the ITER Neutral Beam Test Facility
        The ITER Heating Neutral Beam (HNB) injectors, one of the tools necessary both to achieve burning conditions and to control plasma instabilities, are characterized by such demanding parameters as to require the construction of a test facility dedicated to their development and optimization. This facility, called NBTF, is in an advanced state of realization in Padua (Italy), with the direct contribution of the Italian government, through the Consorzio RFX as the host entity, IO, the in kind contributions of three DA’s (F4E, JADA, INDA) and the technical and scientific support of various European laboratories and universities. The NBTF hosts two experiments: SPIDER and MITICA. The former is devoted to the optimization of the HNB and DNB ion sources and to the achievement of the required source performances. It is based on the RF negative Ion Source concept developed at IPP (Garching). MITICA is the full size prototype of the ITER HNB, with an ion source identical to the one used in SPIDER. The construction and installation of SPIDER plant systems was successfully completed with their integration into the facility, followed by integrated commissioning with control (CODAS), protection and safety systems. The mechanical components of the ion source have been installed inside the vessel and connected to the plants. Finally, the integrated commissioning of the whole system ended positively and the first experimental phase began. Also the realization of the MITICA project is well advanced, although the completion of the system and its entry into operation is expected in 2022 due to the long procurement times of the in-vessel mechanical components. In particular, the power supply designed to operate at 1MV are in an advanced phase of realization, all the high voltage components have been installed and the complex insulation test phase has begun in 2018. Furthermore, all the other auxiliary plant systems are being installed and / or undergoing testing. This paper gives an overview of the progress of the NBTF realization with particular emphasis on issues discovered during this phase of activities and to the adopted solutions in order to minimize the impact on the schedule while maintaining the goals of the facilities. Finally, the first results obtained with SPIDER experimentation and with the 1MV insulation tests on the MITICA HV components will be presented.
        Speaker: Mr Jaydeepkumar Joshi (ITER-India (Institute for Plasma Research))
      • 57
        Demonstration of 1 MV vacuum insulation for the vacuum insulated beam source in the ITER NB system
        For the ITER neutral beam (NB) system, a measure to achieve the 1 MV vacuum insulation of the beam source have been developed. For this purpose, design basis for 1 MV vacuum insulation has been developed by integrating previous empirical scaling for plane and coaxial electrodes and new scaling for area with locally-concentrated electric field. Consequently, as the measure, the beam source is surrounded by more than three intermediate electrostatic shields instead of single gap to sustain 1 MV. Effectiveness of the shields designed by the design basis was experimentally verified by using a part of the beam source. The voltage holding capability has been significantly improved from 0.7 MV to 1 MV. This result ensures the 1 MV vacuum insulated beam source in the ITER NB system.
        Speaker: Mr Jaydeepkumar Joshi (ITER-India (Institute for Plasma Research))
      • 58
        Diagnostic mirrors for ITER: research in a frame of International Tokamak Physics Activity
        Mirrors will be used as first plasma-viewing elements in optical and laser-based diagnostics in ITER. Deterioration of the mirror performance due to e.g. sputtering of the mirror surface by plasma particles or deposition of plasma impurities will hamper the entire performance of the affected diagnostic. Specialists Working Group on First Mirrors (FM SWG) in the Topical Group on Diagnostics of the International Tokamak Physics Activity (ITPA) plays a crucial role in finding solutions for diagnostic first mirrors. Sound progress was achieved during the past decade. Single crystal (SC) rhodium (Rh) mirrors became available. SC Rh and molybdenum (Mo) mirrors survived in conditions corresponding to ~ 200 cleaning cycles without a degradation of reflectivity. These results are important for a mirror cleaning system, based on sputtering of contaminants by plasma. Efforts are invested to the physics understanding of a cleaning discharge. Ion energy distribution and flux in radiofrequency (RF) discharge have been studied. Repetitive cleaning was tested on several mirror materials. Experiments comprised contamination/cleaning cycles. The reflectivity SC Mo mirrors was preserved after 34 cycles. First in-situ cleaning was conducted in EAST with a mock-up mirror of ITER Edge Thomson Scattering using RF plasma. Contaminants from the mock-up mirror were removed. Mirror contamination can also be suppressed by a protecting diagnostic duct. A Deposition Mitigation duct system was exposed in KSTAR. The real-time measurement of deposition in the diagnostic duct was pioneered during this experiment. Results evidenced the dominating effect of the wall conditioning and baking on contamination inside the duct. A baffled cassette with mirrors was exposed in the main wall of JET ILW for 23,6 plasma hours. No significant degradation of reflectivity was measured on mirrors in the ducts. Predictive modeling was advanced. A model for the particle transport, deposition and erosion inside the port-plug was used in selecting an optical layout of the ITER core Charge-Exchange Recombination Spectroscopy diagnostic. These achievements contributed to the focusing of the first mirror research, accelerating the diagnostic development. Predictive modeling requires more efforts to be invested. Ensuring the progress in the remaining crucial areas will be a focus of the future work of the FM SWG
        Speaker: Dr Andrey Litnovsky (Forschungszentrum Juelich, Germany)
      • 59
        Integration of Thomson scattering and laser-induced fluorescence in ITER divertor: engineering and performance analysis
        This paper describes the benefits and challenges of divertor Thomson scattering 55.C4 (DTS) and laser-induced fluorescence 55.EA (LIF) integration in the divertor port #8 of ITER. One of the main challenges for the DTS system is to measure extremely low electron temperatures in the vicinity of the divertor plates. The cool and dense divertor plasma leads to pronounced collective effects and significant distortions of the TS spectra. Therefore, standard TS signal processing, valid for light scattering on a swarm of free electrons, is already not valid. To examine the real DTS performance, we apply a special simulation technique based on synthetic experiments. The estimated measurement accuracies of electron temperature and density are quite better than the specified technical requirements, in spite of the pronounced collective effects. On the contrary, in the case of low electron density, when the classical TS spectrum is expected, the diagnostics performance degrades significantly, though still satisfying the technical requirements. Currently, the LIF diagnostic is to measure density of Helium atoms with a collisional-radiative model (CRM) describing a relation between the fluorescence and plasma parameters. Required for CRM electron parameters are taken from DTS diagnostics. The temporal forms of the Helium fluorescence are dependent on electron parameters and the pumping laser pulse characteristics. Therefore, LIF can measure electron density in the range of 10^18 10^20 m^-3 analyzing the temporal behavior of Hellium fluorescence with the Helium CRM. This technique helps to expand the measurable range of electron density. The main advantage of this LIF measurements is that calibration of the collection system spectral and / or absolute sensitivity is not required, contrary to the DTS approach. Both DTS and LIF are laser aided diagnostics; hence, it seems attractive to develop universal laser and probing optics, which is the most sophisticated and expensive part of any ITER optical diagnostics. The engineering solutions discussed and challenges of the DTS and LIF integration includes collinear combination of DTS and LIF lasers, laser mirrors, collection mirrors, etc. Although the proposed solutions are considered in terms of ITER divertor compatibility, their use in currently operating magnetic confinement devices is also under discussion.
        Speaker: Dr Eugene Mukhin (Ioffe Institute)
      • 60
        Current Design and R&D Progress of CN HCCB TBS
        As the testing mockup of tritium breeding blanket for DEMO, Chinese Helium Coolant Ceramic Breeder (HCCB) Test Blanket System (TBS) are under developing by China and will be tested in ITER to verify the key tritium breeding blanket technologies. After the approval of conceptual design by ITER Organization in 2015, the design optimization and more R&D activities for HCCB TBS have been under implementation for preliminary design phase. As the structural material of TBM module, eight tons RAFM steel (CLF-1) plates and forgings have been fabricated and a certification of 3.2 requested by EU Pressure Equipment Directive 97/23/EC (PED) has been obtained for CLF-1 steel. The fabrication techniques for the functional materials, beryllium pebble and Li4SiO4 pebble, have also been developed and the properties have tested. The new manufacture facility for Li4SiO4 pebble is under construction and the manufacture facility for beryllium pebble was upgraded to achieve production rate 10kg/batch. Recently the TBM-set design was significantly optimized and the whole integration method of TBM and the fabrication procedure plan has been updated. The results show that the total heat deposition in TBM was similar with conceptual design, while the tritium production ratio was slightly higher. The fabrication technology of TBM is under development. Following the fabrication procedure plan of TBM, semi-prototype of TBM is under fabrication to verify the final integration plan. Ancillary systems have been optimized considering the review comments, safety and interface requirements. Accordingly the Process Flow Diagram (PFD) and Pipe & Instrumentation Diagram (PID) have been updated, but still some interface issues with ITER facility have been identified and have to be solved. The system performance has been assessed to optimize the operation control plan and equipment requirements. Several test platforms for breeding blanket technology development have been constructed and started experiments to test components, processes and get the operation data. At same time, the safety assessment of HCCB-TBS has been updated and it shows that HCCB TBS has not over-temperature issues for all accident cases. Considering the limited inventories and multiple confinement barriers, no major safety consequences had been identified through accident assessments.
        Speaker: Prof. Xiaoyu WANG (Southwestern Institute of Physics)
    • Reception
    • FIP/1 P1 Posters
      • 61
        Completion of 1st ITER gyrotron manufacturing and 1 MW test result
        Speaker: Dr Yasuhisa Oda (Japan Atomic Energy Agency)
      • 62
        Completion of the first TF Coil Structure of ITER
        Speaker: Dr Masataka Nakahira (National Institutes for Quantum and Radiological Science and Technology)
        Summary Slide
      • 63
        Current Design and R&D Progress of CN HCCB TBS
        Speaker: Prof. Xiaoyu WANG (Southwestern Institute of Physics)
      • 64
        Demonstration of 1 MV vacuum insulation for the vacuum insulated beam source in the ITER NB system
        Speaker: Mr Atsushi Kojima (National Institutes for Quantum and Radiological Science and Technology)
      • 65
        Diagnostic mirrors for ITER: research in a frame of International Tokamak Physics Activity
        Speaker: Dr Andrey Litnovsky (Forschungszentrum Juelich, Germany)
      • 66
        Integration of Thomson scattering and laser-induced fluorescence in ITER divertor: engineering and performance analysis
        Speaker: Dr Eugene Mukhin (Ioffe Institute)
      • 67
        Outcome of R&D program for ITER ICRF Power Source System
        Speaker: Mr Rajeshkumar Gajanan Trivedi (ITER-India, Institute for Plasma Research)
      • 68
        Progress in the ITER Neutral Beam Test Facility
        Speaker: Dr Vanni Toigo (Consorzio RFX)
      • 69
        Recent progress in the development of the European 1 MW, 170 GHz CW gyrotron for ITER
        Speaker: Dr Gerd Gantenbein (Karlsruhe Institute of Technology, Institute for Pulsed Power and Microwave Technology)
      • 70
        Technologies for realization of Large size RF sources for –ve neutral beam systems for ITER -Challenges, experience and path ahead
        Speaker: Mr Jaydeepkumar Joshi (ITER-India (Institute for Plasma Research))
    • OV/3 Overview Magnetic Fusion
      • 71
        Progress of JT-60SA Project
        The JT-60SA project was initiated in June 2007 under the framework of the Broader Approach (BA) agreement and Japanese national fusion programme for an early realization of fusion energy by conducting supportive and complementary work for the ITER project towards supporting the basis for DEMO. In 2009, after a complex start-up phase due to the necessity to carry out a re-baselining effort to fit in the original budget while aiming to retain the machine mission, performance, and experimental flexibility, the detailed design of the project could start immediately followed by the start of manufacturing of the long lead items. Components and systems of JT-60SA are procured by the implementing agencies (IAs): Fusion for Energy in EU and QST (previously JAEA) in Japan. With the project now in an advanced implementation stage, the early defined approach for its implementation has proven to be successful and hence continues to be employed. This is underpinned by the very close collaboration between QST in Japan, F4E in Europe, and all other European stakeholders: the EU Voluntary Contributors (EU-VCs) and EUROfusion. The employed management model follows the early establishment of a single Integrated Project Team (IPT) that operates in accordance to an agreed Common Quality Management System, defining resources and processes crossing the lines between organizations. For JT-60SA the same management model strategy is planned also for the period beyond 2020, that is when the facility will be jointly operated and enhanced by the EU and JA.The paper will overview the progress of the manufacturing and assembly of the JT-60SA machine, the outlook towards First Plasma, and progress in preparing for the scientific exploitation of JT-60SA following this milestone.
        Speaker: pietro barabaschi (f4e)
      • 72
        Progress of the CFETR Design

        The Chinese Fusion Engineering Test Reactor (CFETR), complementing ITER, is aiming to demonstrate fusion energy production up to 200 MW initially and to eventually reach DEMO relevant power level, to manifest high duty factor of 0.3~0.5, and to pursuit tritium self-sufficiency with tritium breeding ratio (TBR) > 1. The key challenge to meet the missions of the CFETR is to run the machine in steady state and high duty factor. Recently, a self-consistent steady-state scenario for CFETR with fully sustained non-inductive current drive is developed using a multi-dimensional code suite with physics-based models. In addition, results from the experimental validation conducted by a recent EAST steady-state experiment with off-axis current drive enhance confidence in the performance prediction from the integrated modeling. Finally, a fully non-inductive reverse-shear scenario scaled to R = 6.7 m, βN~3, H98 ~ 1.5 and fBS ~ 0.75 with the performance that meets the high gain CFETR mission is demonstrated. The scenario presents a self-consistent solution for the CFETR transport, equilibrium and pedestal dynamics.
        At present, the CFETR physics design focuses on optimization of the third evolution CFETR (R = 7 m, a = 2 m, kappa = 2, Bt = 6.5-7 T, Ip = 13 MA) consistent with steady-state or hybrid mode and a radiative divertor. Listed below are the main tasks we needed to tackle in the near-term, e.g. to demonstrate compatibility with the alpha particle stability and transport, and to quantify the tritium burn-up rate during the steady-state burning plasma phase in order to find a solution to meet the central fueling requirement, and so on. The details will be given in this meeting

        Speaker: Guoqiang Li (Institute of Plasma Physica, CAS)
      • 73
        Overview of the Validation Activities of IFMIF/EVEDA: LIPAc, the Linear IFMIF Prototype Accelerator and LiFus6, the Lithium Corrosion Induced Facility
        In this report, the latest results of the validation activities of the IFMIF/EVEDA project under the Broader Approach agreement are overviewed. For the Linear IFMIF Prototype Accelerator (LIPAc) to demonstrate the 9MeV/125mA D+ beam acceleration, the beam qualification study of the injector was completed with the emittance of 0.16 pi mm mrad smaller than required 0.3 pi mm mrad, and the maximum vane voltage in the RFQ cavity was achieved at 143kV exceeding the required 132kV. These components and other subsystems of LIPAc are ready to inject the beam to RFQ to provide the 5MeV D+ beam. The Superconducting RF linac necessary for the 9MeV D+ beam is close to the end of manufacturing phase to start its final assembly in Rokkasho. For the liquid lithium loop activities, 4,000 hours lithium corrosion test of the Reduced Activation Ferritic/Martensitic steels using the LiFus6 were completed and verified that the corrosion rate can be kept under control and well below the requirement of 1 micro-m/y, after achieving a good purity of lithium, < 30ppm N.
        Speaker: Masayoshi Sugimoto (QST-Rokkasho)
      • 74
        The Strategy of Fusion DEMO In-vessel Structural Material Development
        The structural material development for the breeding blanket in a future fusion reactor is regarded as the most challenging technical issue due to the significance of 14 MeV DT fusion neutron irradiation that induces high displacement damage with a significant amount of the transmutation formed gas elements such as helium and hydrogen. The strategy of fusion in-vessel structural material development toward fusion DEMO is addressed with special emphasis on the current status and the limitations due to the reliability of data. A major issue in developing and validating structural materials for a fusion DEMO reactor are missing facilities where materials can be tested under the real in-vessel conditions of deuterium-tritium (DT) fusion. Ideally, neutron irradiation induced changes are expected to be negligible or “minor”. The reality is, however, that irradiation effects are neither “negligible” nor “minor”. Thus, it is essential to define the negligible and maximum level of irradiation-induced changes which could be incorporated into safety factors that are defined “empirically”, and the most significant technical challenges are to develop and qualify materials based on the knowledge and data acquired in experiments not performed under “real” fusion environment but in fission neutron irradiation and various simulation irradiation experiments, and to develop and verify a framework of DEMO reactor design criteria for in-vessel components (DDC-IC). Here we propose a new strategy based on probabilistic approaches, where the probability of failure is calculated based on the probability density function of postulated load distribution and material property distribution, as a part of the design methodology in order to mitigate the uncertainties caused by multiple sources. It is essential to conduct statistical analyses on material property data to make the data applicable to the probability based design method. Consequently, the vast amount of fission neutron irradiation data which fulfill the statistical requirements should be developed up to some critical irradiation dose levels at which the irradiation effects caused by fusion neutron spectra are expected to become very different from fission data.
        Speaker: Dr Hiroyasu Tanigawa (National Institutes for Quantum and Radiological Science and Technology)
    • P1 Posters
      • 75
        ACTYS Code System: Towards Next Generation Nuclear Activation Codes for Fusion Reactors
        Nuclear activation and subsequent radiological response of structural materials within fusion reactors like ITER and beyond need to be studied for operational, safety and radiological waste management reasons. The future fusion machines should be equipped with low radioactive materials optimized for expected neutron environment. Numerical tools with extended capabilities are needed for this kind of analysis. A project named ACTYS-Project is initiated at Institute for Plasma Research to meet the requirements stated above. This effort so far developed more than five states of art codes and few innovative computational tools for analysis and design of fusion reactors. The details of all the codes and tools will be presented in this paper. ACTYS is the first code within the project. It is a single-point neutron activation code and computes nuclide inventories and other radiological responses within materials when exposed to neutron flux through either continuous, pulse irradiation or mixed. It solves coupled first-order LODEs using Bateman solution for linearized chains. An ‘exponential convergence’ algorithm and ‘chain weighing’ termination technique is developed in-situ for this purpose. These two methods lend ACTYS an added edge over typical linear chain solvers. ACTYS is well validated and the details of the same will be presented. Highly resolved nuclear activation analysis and radiation waste classification are warranted for large-sized fusion machines with a wide variety of materials. To ensure a fast multi-point activation analysis without sacrificing accuracy, inherent changes must be done to single-point activation codes like ACTYS. To this end, a multipoint activation code-named ACTYS-1-GO is developed. Recently, it has been coupled with transport code ATTILA by developing a subsidiary module, activation source generator. One of the important features is that nearly 1 million meshes can be computed in less than few hours. A mathematical formulation to account for the contribution of the parent constituents of any irradiated material towards the radiological responses was derived and implemented. The first order derivatives of Bateman linear chain solution with respect to the decay and cross sections constants are generally used for the sensitivity analysis. A simplified and improved set of recursive relations are developed for these derivates.
        Speaker: Dr Subhash P.V (ITER-India, Institute for Plasma Research, Gandhinagar, Gujarat)
      • 76
        Alignment and Calibration Schemes for ITER CXRS-Pedestal Diagnostic
        Charge eXchange Recombination Spectroscopy (CXRS) diagnostic shall provide the key measurement for ITER advance plasma control and physics studies. ITER CXRS-pedestal has a primary role of edge ion temperature, plasma rotation (toroidal and poloidal velocity) and impurity density measurements in the pedestal region (r/a=0.85 -1.0). Meeting the measurement requirements for these parameters in ITER is more challenging than the present tokamaks due to restricted access for diagnostics components in addition to the harsh environment of ITER. Some of these challenges are like the calibration offset that limits the velocity measurement accuracy requirement. As well as precise alignment required because of the lower angle between the line of sight with a toroidal plane that introduces additional error in the measurement. Therefore, to meet the measurement requirements in ITER, robust calibration and alignment schemes are being developed. CXRS-pedestal shall cover broad wavelength range, the emission of (He, Be, Ne, Ar, C) recombining lines (460-532nm) and Beam Emission Spectroscopy (BES) Hα (656.3nm) spectral line simultaneously, compatible with the spatial resolution of 20 mm (that demands a fine alignment) with 5Hz DNB modulation: 100ms exposure with DNB ON, 100ms background exposure. Statistical and systematic errors including atomic modeling along with low light signal due to strong attenuation of the diagnostic neutral beams require better light transmission path and high throughput spectrometer detection system. To access this requirement, preliminary performance assessment carried out using Simulation of Spectra (SOS) code to see the dependency of the measurement accuracy on SNR. In this contribution, details of the design and development of the ITER CXRS-pedestal diagnostic system in view of alignment & calibration in the suitable transmission system, this includes the optimum light transmission path analysis using ZEMAX ray tracing tool. This will ensure the required alignment for accurate measurement from the DNB and plasma cross-section area of the pedestal region. The various calibration and alignment schemes are studied and shall be developed to test the performance in the ITER-India Lab to meet the ITER requirement.
        Speaker: Mr Gheesa Lal Vyas (ITER-India, Institute for Plasma Research)
      • 77
        Application of Finite Element Techniques in Simulation of Mechanical Design and Performance Assessment of Different Components of a Neutral Beam Systems
        Accelerators, Ion dumps and beam transport system for Neutral Beam application are designed to manage high heat loads in the range of 2-10MW/m2. The performance of these components under various damage criteria are assessed for their thermo-mechanical stability under various operating and faulty conditions. Due to the pulse nature (3s ON/ 20s OFF with 5 Hz modulation) of beam operation, components often exposed to cyclic thermal loads. Further, the above system is incorporated with large number of flexible elements (e.g. bellows, etc.) to absorb the thermal movements. For systems like accelerator and electrostatic residual ion dumps, there is an additional need of non-metallic components, like ceramics, which functions as electrical isolation as well act as structural elements. To assess diverse nature of such systems with complex loading requirements, Finite Element Analysis tools (e.g. ANSYS, CFX, SYSWELD, etc.) have been employed as part of design evolution and results are verified according to codes and standards (ASME / RCC-MR / EJMA). The experimental validation of effectiveness of these assessment have been also performed by prototype testing and performing the tests on the real manufactured products. It is also important to note that, the tools are also useful to address the in-process manufacturing modification those may arise due to feasibility constraints. The paper shall present some of the important simulations results on 10MW/m2 capability of Heat Transfer elements, functional tests on 100 kV post insulators, bellows assessment in water lines, CFD simulations for beam source components, etc.
        Speaker: Mr VENKATA NAGARAJU MUVVALA (ITER-India, Institute for Plasma Research)
      • 78
        Automated Testing of ITER Diagnostics Scientific Instrumentation
        ITER requires extensive diagnostics to meet the demands for machine operation, plasma control, protection, safety and physics studies. Most diagnostics require high performance scientific computing for the processing of complex algorithms for the measurements. The most stringent requirements are found in the more than 50 diagnostics measurement systems in terms of high performance data acquisition, data processing and real-time data streaming from distributed sources to the plasma control system as well as large amounts of raw data streaming to scientific archiving. While most of these requirements have been achieved individually the challenge for ITER will be the integration of these state-of-the-art technologies in a coherent design while maintaining all of the performance aspects simultaneously. The instrumentation and control (I&C) systems for each diagnostic must meet around 500–700 functional and non-functional requirements which include also the requirements from the ITER handbooks such as the Plant Control Design Handbook (PCDH), Electrical Engineering Design Handbook (EDH) and the Radiation Compatibility Handbook. While the diagnostics I&C system engineering methodology (Figure 1) is well established for requirements management, detailed design, and implementation, the acceptance testing, demonstrating the compliance of the I&C system with the requirements needs further elaboration. This includes the definition of the test scenarios, detailed test procedures, and well-defined pass-fail criteria for each test. Since compliance validation against a large number of requirements can be very time consuming a high degree of automation during testing is desirable. This paper presents the elaboration of the pass/fail criteria, the acceptance testing procedures for diagnostics plant system I&C, and describes the design and implementation for automated testing. First results will illustrate the reduction in testing time for obtaining a detailed compliance evaluation.
        Speaker: Dr Stefan Simrock (ITER)
      • 79
        Baking System of Aditya Upgrade Tokamak

        In tokamaks, baking of vacuum vessel and first wall components is a prerequisite in order to obtain impurity free plasmas. Baking is performed to remove impurities viz. H2, H2O and Hydro-Carbon from the vessel and first wall components. ADITYA tokamak has been upgraded ADITYA-U tokamak to achieve shaped plasmas. The ADITYA-U is equipped with a comprehensive baking system for heating the SS vacuum vessel, pumping systems, associated diagnostics along with the graphite limiter and diverter tiles up to 150 C. The DC Glow discharge cleaning is also carried out in presence of baking to achieve better wall conditioning for high performance plasma operation. Due to space limitation between vessel and Toroidal field coils at the high-field side, 1.5 mm thick silicon heaters has been designed and procured. In-situ installation of heaters has been quite challenging due to structural complexity. For efficient heat insulation, 6 mm thick silicon jacket designed, fabricated and installed according to vessel profile. A detail analysis carried out in ANSYS for its optimum performance and to examine its effect on vessel, especially on the several weld joints. Whole baking system consists of ~80 heaters installed on different sectors of the vessel, pumps and diagnostics. The heaters are controlled in close loop by in-house developed Programmable Logic Controller (PLC) based automatic control system. It comprises of three main phases, temperature ramp-up, constant heating and ramp-down to room temperature. All these phases are individually controlled as required. The entire baking system has been tested thoroughly for its automatic operations for long hours(~48 hr.), integration, ruggedness, reliability, small form factor. The detailed hardware concept, software design and prototype testing and its regular operation in presence and absence of GDC will be discussed. Partial pressure of impurities is monitored in every baking cycle which decides the controls of the baking temperature and duration automatically. Further, the potency of lithiumization carried out before, during and after baking has been compared for the first time in ADITYA-U by estimating the lithium lifetime on the walls with plasma operation. The improved wall conditioning with baking and its effect on plasma operation along with technical challenges faced during installation will be presented in this paper.

        Speaker: Mr Kaushal Patel (Institute for Plasma Research)
      • 80
        Basic studies of the interaction of blobs with suprathermal ions and millimetre-wave beams in the TORPEX device
        The fundamental interactions between turbulent structures or blobs and suprathermal ions, injected by Li6+ beams in plasmas created by microwaves at 2.45GHz with ne~1015-1017m-3 and Te~2-10eV, are extensively investigated on the TORPEX toroidal device. Comparisons between fully validated numerical simulations and experimental 3D time-averaged suprathermal ion profiles reveal an entire spectrum of non-diffusive suprathermal ion transport: super-diffusive, diffusive, or sub diffusive, depending on particle energy and turbulence amplitude. 3D time-resolved measurements of 30eV and 70eV ions, exhibiting super- and sub-diffusive transport respectively, show that in all cases the ions are subject to bursty displacement events and that intermittency, quantified by the skewness of the time-traces, is present to some degree in all profiles, also for intermediate energies, including in the sub-diffusive cases. We develop an analytical model that links the time averaged-profile of the ion current and the profile of the statistical moments of the fluctuations. In fusion devices, externally injected beams in the electron cyclotron (EC) frequency range are employed for heating and current drive, and to stabilize neoclassical tearing modes. EC beams must propagate through the Scrape-Off Layer where blobs may scatter the incoming wave by locally modifying the plasma permittivity. This may lead to a loss of efficiency in EC heating and mode stabilization. To understand the effect of plasma turbulence and its structures on the propagation of millimeter waves (mmw), we measure wave scattering in TORPEX by blobs of size comparable to the wavelength. A low-power beam is launched at 29.7GHz in the X-mode from the top of the device using a pyramidal horn antenna. The X-mode component of the transmitted power is detected at the bottom using a pyramidal horn antenna and a Schottky diode, whose position can be radially adjusted. A conditional sampling technique averages the effect of several thousand individual blobs. Combining these scattering measurements with first principle full-wave simulations using COMSOL, we show that density fluctuations associated with plasma blobs, with δne as small as ~10-3 of the critical density, can significantly defocus the mmw-beam in the wake of the blobs, resulting in mmw-power fluctuations that increase monotonically with the blob amplitude.
        Speaker: Dr Ivo Furno (EPFL- CRPP)
      • 81
        Comparative modeling of plasma boundary corrugation due to the application of 3-D fields with ELM control coils in various ITER scenarios
        The plasma response to the 3-D external resonant magnetic perturbation (RMP) fields, applied for controlling type-I edge localized modes (ELMs) in ITER, is systematically computed in terms of the normal displacement of the plasma surface, in other words the 3-D corrugation of the plasma boundary. Five representative ITER H-mode plasma scenarios, ranging from an initial hydrogen plasma discharge in pre-nuclear phase to the Q = 10 nuclear phase DT operation. The plasma surface corrugation, computed using the MARS-F code, is used as a basis to understand the capability and robustness of the type-I ELM control in these ITER scenarios. A key aspect of this study is to assess effects of variation/uncertainty of pedestal plasma rotation on the plasma response. For each plasma scenario, a set of the toroidal rotation - both amplitude and radial profile - is generated by the transport code ASTRA, assuming different Prandtl numbers as well as different ratios of the toroidal momentum to thermal confinement times. Toroidal modeling results show that, (i) the plasma response is similar for the two DT scenarios with 15 MA/5.3T plasmas but with different fusion gain factors (Q=5 versus Q=10); (ii) the other plasma scenarios, with similar rotation profiles, have different plasma boundary corrugation; (iii) the effect on ELM control performance by utilizing 2 or 3 rows of coils, with the coil phasing optimization, varies depending on the availability of the ELM control coil power supplies. The plasma response database, generated in this study, can also be used for further studies such as the divertor footprint and heat load, or energetic particle losses due to RMP.
        Speaker: Dr Li Li (Donghua University, Shanghai, China)
      • 82
        Completion of DC 1 MV power supply system for ITER neutral beam test facility
        Technologies of DC 1 MV insulation with water and DC 1 MV vacuum insulation have been developed. As the result, manufacturing of DC 1 MV power supply components to produce 1 MeV negative ion beams have been completed for the ITER neutral beam test facility (NBTF). For the transmission line (TL), insulating tubes for hot and cooling water have been developed by clarifying resistivity of high-temperature water and properties of insulation material with water absorption. Based on these results, the integrated configuration of the TL has been established through electrical and thermo-mechanical analyses. For the HV bushing, 1 MV vacuum insulation was achieved based on the empirical scaling for the vacuum insulation. Then, all components have been manufactured and shipped to the NBTF site in 2017. The installation is on-going toward the integration test in 2018. These achievements contribute to push forward with a start of the NBTF operation and a realization of the ITER NB.
        Speaker: Dr Hiroyuki Tobari (National Institute of Quantum and Radiological Science and Technology)
      • 83
        Conceptual Design Study for Heat Exhaust Management in the ARC Fusion Pilot Plant
        The ARC pilot plant conceptual design study [1] is extended to explore options for managing ~525 MW of fusion power generated in a compact high field ($B_0$ = 9.2 T) tokamak about the size of JET ($R_0$ = 3.3 m). Exploiting ARC’s demountable high temperature superconductor toroidal field (TF) magnets, poloidal magnetic field (PF) coils located inside the TF, and vacuum vessel (VV) immersed in molten salt FLiBe blanket, this follow-on study identifies novel robust power exhaust solutions. The superconducting PF coil set is reconfigured to create double-null plasma equilibria that include an X-point target divertor geometry. Modeling shows that such long-leg configurations enhance power handling and can achieve passively-stable detachment fronts that stay in the divertor leg over a wide power window [2,3]. The VV is modified to include the divertors while retaining original core plasma volume and TF magnet size. The molten salt FLiBe blanket shields all superconductors, functions as an efficient tritium breeder, and, with augmented forced flow loops, serves as a single-phase, low-pressure coolant for the divertor and VV. MCNP neutronics calculations show a tritium breeding ratio of ~1.08. The neutron damage rate of the remote divertor targets is ~3 times lower than that of the first wall, which is beneficial because high neutron damage often leads to degradation in thermal performance. The demountable TF magnets allow for vertical maintenance schemes and replacement every 1-2 years, increasing tolerance for neutron damage. The divertor has tungsten swirl-tube cooling channels capable of exhausting 12 MW/m$^2$ of heat flux, which includes a factor of ~8 safety margin over anticipated steady state heat loads. Novel diagnostics supporting the heat exhaust mission compatible with the neutron environment are proposed, including the use of Cherenkov radiation emitted in FLiBe to measure fusion reaction rate, microwave interferometry to measure divertor detachment front location, and IR imaging through the FLiBe blanket to monitor divertor “hotspots.” *The authors acknowledge support from the MIT Nuclear Science and Engineering Department and the PSFC.* [1] B.N. Sorbom, *et al.*, *Fus. Eng. and Des.* **100** (2015): 378-405. [2] M.V. Umansky, *et al.*, *Phys. of Plasmas* **24** (2017): 056112. [3] M. Wigram, *et al.*, *Conts. to Plasma Phys.* (2018)
        Speaker: Ms Elizabeth Tolman (UsPSFC)
      • 84
        Consorzio RFX Contribution to the JT-60SA Project in the Frame of the Broader Approach Agreement
        The JT-60SA satellite tokamak is now under advanced assembly phase in Naka (Japan). The majority of the new power supplies are provided by Europe, and the Italian National Research Council (CNR), acting through Consorzio RFX, has contributed in particular with two systems: the Quench Protection Circuits (QPC) for the superconducting magnets and the Power Supply System for RWM control. The procurement of both the systems has been successfully carried out: the QPCs were delivered to Naka site in autumn 2014; the installation, commissioning and acceptance tests were completed in July 2015, fully in line with the schedule agreed in 2009. The protection system for the superconducting coils is composed of thirteen units: three for the TF circuit and ten for the PF circuits. Their duty is to conduct the coil current in normal operation and commutate it into a dump resistor in case of quench or other faults by means of a dc Circuit Breaker (CB). The nominal currents to be interrupted and the maximum reapplied voltages are 25.7 kA and 2.8 kV for the TF QPCs and ±20 kA and ±5 kV for PF QPCs. As for the RWM-PS system, we are very close to the completion too, with the delivery on site and closure of the procurement expected in autumn 2018. This system consists in an input rectifier stage and 18 power amplifiers, one for each coils, capable to supply a peak current of 300 A and an output voltage of 240V and satisfy strict dynamic requirements in terms of latency and current bandwidth (50 s, 3 kHz) thanks to the adoption of new hybrid Silicon-Silicon Carbide (Si-SiC) power semiconductors for the power amplifiers and to the development of a new sophisticated control board, based on the combination of a fast microcontroller and a FPGA running optimized firmware. A summary of the studies for the development of both the systems, of the main phases of their procurement and relevant results will be presented. The innovative aspects of their design will be highlighted: JT-60SA QPC represents the first application of hybrid mechanical-static technology for protection of superconducting magnets in fusion experiments and RWM-PS is the first PS system in fusion experiments adopting SiC semiconductors. The future work will be also discussed; outcomes from the operation of these systems, useful for ITER and DEMO, are expected.
        Speaker: Dr Elena Gaio (Consorzio RFX)
      • 85
        Control of NTMs and integrated multi-actuator control on TCV
        Detailed experiments have been performed on TCV with its flexible electron cyclotron heating/current drive system to investigate reliable and efficient control of NTMs. For example, a novel sinusoidal sweeping technique has been studied in detail and we have shown for the first time that it is efficient for both NTM stabilization and preemption. This method is important for future devices as it relaxes the demand on the accuracy of the mode location estimation and the beam deposition calculation, and circumvents the need for extra diagnostics or many shots for tuning. Comparison between NTM preemption and stabilization has been achieved with sweeping and it shows that preemption can be more than twice as efficient as stabilization in terms of the necessary power. The reliable, efficient and generic control of NTMs allows the development of a controller working for all the scenarios and independent of the specialties of TCV, facilitating the integration with other real-time (RT) algorithms. RT control of NTMs, beta and model-estimated q profiles have been achieved simultaneously on TCV for the first time with a generic integrated control framework that consists of a hierarchy of state estimation/prediction, plasma event monitoring, supervision, high-level (HL-) actuator management (AM), generic controllers and low-level (LL-) AM. In an integrated control test, RT diagnostics are used with RT simulations to reconstruct the plasma state. We will show how RT analyses of magnetic signals are used to provide details of the mode, the RAPTOR observer to reconstruct electron temperature and q profiles, the RAPDENS-observer to generate density profiles and RT-TORBEAM to calculate beam depositions. This information is then used by the plasma event monitor to produce a finite-state representation of the plasma state based on user-defined thresholds. The supervisor prioritizes various tasks, activates relevant controllers and interfaces with a generic control layer. In this layer, controllers send requests for each task to the HL-AM that optimizes the actuator allocation and sends back the actuation capability to the controllers and the LL-AM. The LL-AM sends controller commands to the tokamak-specific actuators. Importantly, the control layer has been made tokamak agnostic to facilitate its reuse in other devices and to provide a layer of abstraction for operators.
        Speaker: Mengdi Kong (ChSPC)
      • 86
        Design and Development of 500 kV, 100 mA DC High Voltage Power Supply for Particle Accelerators at IPR
        At IPR Neutral Beam Injection (NBI) facility to heat and drive the plasma current in Tokamak is been built by accelerating the positive / negative ion beam of energy around 100 keV. Under the current R&D plan the projection is to develop the technology for future Mega Volt range DC Power Source facility to accelerate ion beam of energy to the tune of 1 MeV and power of the order of few MW. To meet this objective a compact 500 kV, 100 mA DC upgradable to 1000 kV Power supply is being designed and developed as a first step. This power supply shall also be used for several other applications within IPR related to particle accelerator. The 500 kV, 100 mA, 50 kW DC particle accelerator power supply is being designed using a symmetrical Cockcroft-Walton (CW) voltage multiplier topology owing to its design simplicity and economical construction. Other advantages of such cascade generators are: (a) low voltage rating of components, (b) balanced voltage w.r.t. ground, (c) gradual build-up of voltage, and (d) modular construction. The use of a high frequency power source gives an added advantage of low stored energy, less ripple, better regulation and faster response. A 415 V, 50 Hz 3-phase AC input source is converted into single phase high frequency (i.e. 20 kHz) source using IGBT based full H bridge inverter power supply rated for 100 kVA, 400 V (rms). The high frequency power supply charges the symmetrical CW voltage multiplier through a high voltage high frequency (HVHF) step-up center-tap ferrite core transformer rated for 80 kVA, 400 V / 25 kV – 0 – 25 kV (rms). The output voltage and current of the voltage multiplier unit are controlled by controlling the output voltage of the front end inverter operating in close loop control. This paper will present the design and simulation results of 500 kV, 100 mA DC Power Supply modeled in MATLAB Simscape toolbox. The paper will explain the optimization / sensitivity study performed in selecting and sizing of various active / passive components of CW voltage multiplier, inverter and step-up transformer taking into account the possible operational difficulties and future expansion. Both steady state and transient study results will be explained. This paper will briefly cover the engineering assembly design aspects of voltage multiplier unit in general and of a 250 kV prototype voltage multiplier developed.
        Speaker: Mr ASHOK MANKANI (IPR)
      • 87
        Design and Development of Safety control system of Indian Test Facility (IN-TF) for ITER DNB
        Indian Test Facility (IN-TF) [1] is being built in IPR to characterize Diagnostic Neutral Beam [2] in cooperation with ITER Organization. INTF is a complex system which consists of several plant systems like beam source, gas feed, vacuum, cryogenics and mechanical systems. To ensure successful operation INTF, integrated operation involving all the constituent plant systems is required. The experimental phases involve application of HV power supplies and High power RF power (~800KW) which will produce considerable amount of power (~6 MW) within the facility for longer durations. Hence the entire facility will be exposed high heat fluxes and RF radiations. For ensuring occupational safety for working personnel, it is of prime importance that a mature Safety control system [3] be developed and commissioned for INTF. The design of safety control system (SCS) is based on ITER PCDH guidelines and industrial standards for programmable safety systems (IEC 61511 and IEC 61508). The process of detailed design includes identification of safety instrumented functions (SIF), sensor selection and prototype development. The control hardware includes fault tolerant Siemens PLC with distributed interface on Profisafe protocol and safety software which is developed using Siemens safety programming environment. The SCS has to interface with the conventional INTF Control system (which is based on CODAC core system) for non critical data exchange. The SCS also dictate the overall mode of INTF operations. This paper describes the design methodology involved in arriving at final design with details of application of safety standards for identifying the Safety integrity levels (SIL) of SIFs and details of software level interface. The overall integrated system configuration and test results are also discussed. References: 1. M.J. Singh (2011, October). An Indian test facility to characterise diagnostic neutral beam for ITER, Fusion Engineering and Design.[Online].86,pp.732-735 2. A. Chakraborty (2010, March). Diagnostic Neutral Beam for ITER—Concept to Engineering. IEEE Trans. Plasma Sci.[Online].38(3), 3. H.Tyagi (2016, November). Preliminary design of safety and interlock system for Indian test facility of diagnostic neutral beam. Fusion Engineering and Design.[Online].(112).pp. 766-770
        Speaker: Mr Himanshu Tyagi (ITER-India,IPR)
      • 88
        Design and development of the Articulated Robotic Inspection Arm (ARIA) for fusion machine
        Remote Handling (RH) systems for maintenance and inspection of in-vessel components have been addressed in great detail for fusion machines around the world. Maintaining high availability of fusion machine and minimizing the maintenance time require robust and dependable RH systems. Such RH systems, being electro-mechanical in nature, requires research and development in various areas such as structural design, kinematic and dynamic modelling, efficient real-time control, and Virtual Reality (VR) based monitoring. Adding to the aforesaid requirements, is criticality of investment protection of the sophisticated in-vessel components and their size and weight scales. The Articulated Robotic Inspection Arm (ARIA) has been indigenously developed at IPR, India as a proof-of-concept for in-vessel maintenance. The paper presents, in detail, the design and development of the ARIA and associated VR based monitoring and control system. ARIA is a 6-Degrees of Freedom manipulator with a cantilevered payload capacity of ~25kg at 2meters distance. ARIA is controlled using a VR based user interface that immerses the ARIA model into the working environment. The effective 1:1 scale mapping of the VR model with the manipulator hardware makes provision for task planning and executing of the control commands from a remote location. The theoretical calculations with structural analysis of components like links, shafts, couplers, lugs and bearings are elaborately discussed. Results for payload sensitivity analysis during dynamic behavior are also presented. The system is optimized and developed to incorporate efficient commercially available servo actuators, bearings and gear-boxes, to maintain a high degree of accuracy and repeatability. Experimental validation and test results on a mock-up facility show that the system can be controlled with an end-effector positional accuracy within 2mm. The design and integration methodology, presented here, lays foundation to develop efficient RH systems with greater reach and payload capacity for future fusion machines.
        Speaker: Mr Krishan Kumar Gotewal (Institute for Plasma Research)
      • 89
        Design Validation of ITER XRCS Survey Spectrometer with Nuclear Code RCC-MR
        In the ITER, systems are classified in the different safety categories as per their function in the machine; Protection Important Components (PIC) needs more attention during the design and analysis for better safety margins. The French Nuclear Code RCC-MR (2007) is employed in the design, analysis and the manufacturing, applicable to the ITER protection important mechanical components. It is always a challenge to the designers to develop and qualify the design for a PIC system under ITER loading conditions. This becomes even more stringent when the system is exposed to high nuclear radiation and performing the confinement function of radioactive tritium as in the case of X-Ray crystal spectroscopy (XRCS) Survey system. XRCS Survey diagnostics is an ITER PIC system, located in Equatorial port-11, will be used to monitor impurities in the highly ionized state and measure line emission from plasma in the X-ray range (0.1 to 10 nm). This system is connected with the Port Plug flange, due to its specific nature and exposed to complex environments of neutron radiation, high heat flux, electromagnetic forces, etc. To ensure the structural integrity of XRCS Survey from the constant loading (P Type damage), repeated loading (S type damage); we have studied various loads and associated load responses. These loads are broadly categorized in the following three types i) ITER Generic loads ii) Accidental loads and iii) Radiation loads. FE (ANSYS) analysis has been performed and design is validated using the French Nuclear design rules RCC-MR (2007). This paper describes results obtained from structural damage analysis of XRCS Survey system, and their compliance with relevant design rules given in the French Nuclear code RCC-MR validating the design. Topic name: Fusion Engineering, Integration and Power Plant Design
      • 90
        Deuterium Depth Profile Measurement in Pre and Post Irradiated Tungsten

        Tungsten (W) will be used in ITER as a Plasma Facing Material (PFM) in divertor due to its capability to handle high heat flux while having a low Hydrogen (H) isotope affinity. However in presence of fusion neutrons and alpha particles, tungsten can accumulate radiation damage, which might significantly enhance its H retention property. In order to investigate the effects of radiation damage on Deuterium (D) trapping in tungsten, we have carried out experiments using D beam in pre and post irradiated polycrystalline tungsten foils. In this paper we present the comparison of D depth profile measurements using Elastic Recoil Detection Analysis (ERDA) and Secondary Ion Mass Spectroscopy (SIMS) technique.

        Polycrystalline tungsten foil samples of size 8mmx8mmx0.1mm foils were mechanically polished and annealed at 1838 K to release the stress and to minimize the defects. These foils were further irradiated with gold ion (80 MeV), boron ions (10 MeV) to create defects. These samples were then exposed to a D beam of 100keV energy for a fluence of 5x1017 ions/cm2. The trapped D was measured using ERDA and SIMS, and the depth profiles were modelled using binary collisions Monte Carlo method by including the surface roughness. The preliminary results show the enhancement in amount of trapped D in pre-damaged tungsten samples in contrast to the undamaged ones. The effect of Helium (He) on D trapping in sample was also analyzed and it was observed that D trapping is reduced in presence of He. The details of experiments and the analysis will be presented.

        Speakers: Dr Anil Tyagi (ITER-India, Institute for Plasma Research), Mr Matteo Barbarino
      • 91
        Development of a High Temperature Black Body Source for ITER ECE Diagnostic
        For ITER Electron Cyclotron Emission (ECE) diagnostic, there is a requirement of high-temperature black body radiation source operating at atmospheric pressure. This source needs to be operated at high temperature (~ 500 0C) having a microwave emissivity > 0.95 in the frequency band 100-500 GHz, and > 0.75 for 500-1000 GHz. Moreover, the radiation surface should have temperature uniformity within ±10 0C. This source will be utilized for characterizing the ITER ECE measuring instruments like Michelson Interferometer and radiometer. For this purpose, a radiation source has been designed and developed. The radiation source consists of a heater and an emissive surface. The emissive surface is made of silicon carbide (SiC), as it has high thermal conductivity, low thermal coefficient of expansion, excellent machinability, good vacuum compatibility and high emissivity in the mm-wave region. The diameter of the emissive surface is 150 mm.The suitable heating element has been used having high resistivity and good oxidation resistance nature. This paper deals with the design, analysis, and characterization of the developed high-temperature black body radiation source in the frequency range 100 to 1000 GHz. The Finite Element Method based software, “COMSOL”, has been used to analyze and optimise the heating coil design to get desired temperature uniformity of the emissive surface. Experimentally, the temperature uniformity is measured by an IR camera and microwave emissivity is measured by the Michelson interferometer.The operating temperature of 500 0C is achieved in the developed source with temperature uniformity within ±10 0C. The short and long-term temperature stability up to ± 2 0C and ± 10 0C respectively has also been achieved. Further, the microwave emissivity of ~0.8 - 0.9 has been observed over wideband 100-1000 GHz. The above measured values are in compliance with ITER requirement.
      • 92
        Development of High Power Gyrotrons for Advanced Fusion Devices and DEMO
        Megawatt (MW) gyrotrons with a wide frequency range from 14 to 300 GHz are being developed in a collaborative ECH study for advanced fusion devices and a DEMO. (1) Detailed designs of a 14 GHz 1 MW gyrotron has been started for actual fabrication. For a 14 GHz RF beam with high divergence, a calculated transmission efficiency of 85% to the corrugated waveguide coupling position was initially obtained by minimizing the RF transmission path. (2) In the experimental tests of a new 28/35 GHz dual-frequency gyrotron, the cooling characteristics of an optimal-structure double-disk sapphire window was evaluated. We confirmed that operating at 0.4 MW with a continuous wave (CW) at 28 GHz is possible, which is two times the output power reported in previous studies. (3) A 77/51 GHz dual-frequency gyrotron with an output of over 1 MW is presented. (4) In an experiment with a 300 GHz gyrotron, the influence of the reflected wave from the window was reduced by tilting the output window, and mode competition in the cavity was suppressed. An output power of 0.62 MW with a pulse width of 1 ms, which is the new record in this frequency, was obtained.
        Speaker: Dr Tsuyoshi Kariya (Plasma Research Center, University of Tsukuba)
      • 93
        Development of the far-infrared laser polarimetry for current profile measurement on ITER
        The authors are demonstrating the key technology necessary for the ITER poloidal polarimeter (PoPola) in order to measure the plasma current profile in ITER. The entire optical train of a prototype channel was made to evaluate the performance of the laser alignment system and the stability of the polarization measurement. The PoPola system injects multiple far-infrared (FIR) laser beams into the plasmas (wavelength is 119 μm) and those probing beams are reflected by retro-reflectors (RRs). The polarization state of the FIR laser beams returning to the diagnostic room are measured by means of the rotating waveplate Stokes polarimeter (RWS polarimeter). The RWS polarimeter technique measures both orientation angle (θ) and ellipticity angle (ε) of the polarization state. Changes in θ and ε, which are mainly associated with the Faraday and the Cotton-Mouton effects, provide information of electron density, electron temperature and magnetic field. Equilibrium reconstruction of PoPola measurement data together with other ITER diagnostics data provides the current profile. Since the RWS polarimeter technique does not use interference signal of a probing and a reference beam, it does not need to take care about wave front distortion of laser beams and change of path length difference between the probing and the reference beam. However, the RWS polarimeter technique needs higher power (~10 μW) of the laser beam returning to a detector than other polarimeter based on interferometer. Key technologies for getting high power of the returning laser beam are a retro-reflector and a laser beam alignment system. Prototypes of the tungsten RR was made of a tungsten mono-block by machining and the angle between orthogonal mirrors was 89.9167°. Taking into account the thermal expansion during the plasma operation, the achieved manufacturing tolerance is promising. We developed the laser beam alignment method in order to minimize the loss due to shading at the vacuum window and RR. When the laser beam is tilted within +/- 1 mrad for the sake of searching the RR center, the beam position displacement at the vacuum window was 2.0 mm or less. The alignment error above leads to the laser power loss of 4 % owing to shading and acceptable.
        Speaker: Dr Ryota Imazawa (National Institutes for Quantun and Radiological Science and Technology)
      • 94
        Dielectric windows as front-end diagnostic elements in ITER
        The performance of front-end elements of optical diagnostics in ITER under long-term operation and with limited access for their maintenance is in the focus of extensive R&D program involving laboratory study and testing in working tokamaks. The requirements to the front-end element design are driven by high-energy neutron and gamma radiation, intense particle fluxes and thermal loads at the element location on the one hand and necessity to provide periodic or continuous surface recovery on another. The insulating diagnostic window as an alternative to commonly accepted first mirror option is discussed in the presentation. The approach implementation is illustrated for the divertor Thomson scattering (DTS) optical scheme using front-end windows for injection of laser beam and collecting of scattered light. Surface recovery techniques based on plasma cleaning and laser ablation are described with the focus on the performance of the windows under laser and plasma treatment. The windows made from fused silica glass KU-1 and Al2O3 were tested. Plasma cleaning experiments have been performed for clean windows and windows coated with Al films. As was shown by the means of optical microscopy, XPS and AFM the dominant mechanism of window optical degradation is surface roughening. The development of surface relief becomes more intensive after deposition and removal of Al. The clear indication of the dependence of surface degradation rate on the initial polishing quality was also obtained for the windows with and without Al deposition. Laser experiments reveal the decrease in laser-induced damage threshold by the factor of~ 3 for both window materials under continuous tungsten deposition. In the case of Al droplets spraying, damage threshold is about 6 times as low as that of pure KU-1 window. The experiments on the long-term laser cleaning under continuous contamination showed that the evolution of tungsten film stops over the first hundreds of pulses and further exposition has no effect on the film thickness. The steady-state thickness of tungsten deposit in the beam spot was found to be ~ 5 nm for the deposition rate of ~10nm/min and laser (3 ns) energy density of ~ 2 J/cm2, forming almost transparent coating in visible and near-IR regions. Radiation-induced effects in silica glass and sapphire and corresponding limitations are also discussed.
        Speaker: Mr Alexey Razdobarin (Ioffe Institute)
      • 95
        Divertor impurity seeding experiments at the COMPASS tokamak
        Partial detachment is the desired regime for the baseline burning plasma scenario in ITER and next–step devices, as it allows to convert the majority of the energy carried by charged particles through the scrape-off-layer (SOL) into isotropic radiation and thus avoids localized heat flux deposition in the divertor region. In order to maintain relevance to ITER and DEMO, a concentrated effort has been initiated at the COMPASS tokamak to achieve detached operation by means of impurity seeding. Series of experiments with impurity injection in the range of 2-9x1020 molecules per second at different locations in the divertor were performed with the aim to cool the plasma and influence the particle and heat transport in the divertor region and provoke partial detachment. Previously reported results [1] were largely extended by injection of nitrogen at the outer divertor target and also by attempts to seed the plasma with neon. The effects on SOL and divertor plasma conditions were monitored by means of horizontal reciprocating probe manipulator located at the outer midplane and by arrays of divertor Langmuir and Ball-pen probes. The radiation in the edge plasma was observed by AXUV bolometers and fast visible cameras. Experiments in L-mode discharges with nitrogen injected at the outer divertor target have shown that the presence of radiating impurity leads to drop of pressure in the divertor. Depending on the magnitude of the seeding, the upstream pressure can be also affected, suggesting possible penetration of nitrogen into the confined plasma region. The target pressure, however drops at faster rate than upstream, which allows to reach the regime of partial detachment. Similar results were obtained by the HFS nitrogen injection, however the change in divertor pressure was more generally more abrupt and was less sensitive to the amount of injected nitrogen. References: [1] M. Komm et al., proceedings of the 44th EPS conference, Belfast (2017) P1.118
        Speaker: Dr Michael Komm (Institute of Plasma Physics of the Czech Academy of Sciences)
      • 96
        Dynamic Simulation of Loss of Insulation Vacuum Event for ITER Cryodistribution System
        The Auxiliary Cold Boxes (ACBs) of the ITER cryodistribution system has multiple cryogenic process volumes as well as interfaces with cryolines with isolated vacuum spaces. The cryogenic process volumes inside a single vacuum space have different temperature level, 4 K and 80 K as well as different operating pressure, 0.5 MPa and 1.8 MPa. The cryogenic process volumes including interfacing cryolines are protected with safety relief valves (SRVs). In the event of Loss of Insulation Vacuum (LIV) of any particular vacuum space, the incidental heat load of the order of ~6.5 kW/m2 results in rapid pressurization of the cryogenic process volume and pressure must be relieved through SRV. As per the safety requirements of the ITER, the maximum helium inventory inside the tokamak building is restricted and therefore a common relief header is necessary for collecting the release of helium through SRV and carrying it outside the tokomak building. The sizing of the SRVs is performed for the various scenarios as per applicable standard; however, due to the long length of relief header, the required information regarding back pressure on the SRVs is not known in advance. The back pressure is an important parameter to be considered for the sizing and selection of SRV and is a function of geometric condition of relief header, process condition of relieving process volume and relieving mass flow rate. Estimation of back pressure considering steady state condition and maximum mass flowrate through SRV may results in a conservative and unrealistic value of back pressure. Dynamic simulation of the safety relief event along with the complete model of process volume, correct boundary conditions as well as geometric detail of relief header is developed and analyzed based on pressure flow solver model in Aspen HYSYS®. Results are presented for the most demanding scenario viz. LIV event of the largest cryoline and comparison made with the two approaches of back pressure prediction, one with steady state and the other with dynamic simulation model. Results obtained from the dynamic simulation of the entire safety relief system gave useful results at various locations; moreover, the back pressure on SRV is almost half of the back pressure resulted from the steady state approach. Certainly, the dynamic simulation provided valuable inputs for the overall system configuration.
      • 97
        Early identification of disruption paths for prevention and avoidance
        Disruption prevention in the perspective of high performance, high current, long duration plasma discharges requires a substantial evolution of the schemes applied in most of the present tokamaks. An efficient prevention scheme requires the early identification of the nature of the off-normal behavior possibly leading to a disruption and the automatic selection of the appropriated countermeasure, either avoidance or mitigation. The development of such comprehensive scheme is being pursued in a coordinate effort. For the purpose of the avoidance, on which this paper is focused, the disruption can be seen as the result of the interplay of the physical events and of the control system responses to them and to the technical failures. The building blocks of such description should include the integration of several sets of plasma scalar data, plasma profile data, magneto-hydrodynamics indicators and engineering data. Previous work has shown the potential of the Generative Topographic Mapping (GTM) [Bishop C., Neural Comp. 10, 1998] algorithm for identification and discrimination of the disruptive operational space in tokamak devices [Pau A.,Ph.D Thesis,, 2014; Cannas B., PPCF56, 2014] . In this paper it is shown that the magnetic fluctuations associated with rotating MHD modes can be characterized using a set of observables derived from the Singular Value Decomposition applied to the data collected by an array of pick-up coils. They can provide an input to the GTM analysis such that a clustering separating disruptive and non-disruptive timeslices can be found. A further source of information comes from the analysis of the sequences of events recorded by the machine control system. The analysis of such sequences shows that disruptions and non-disruptive terminations generally follow different paths, i.e. are not populating equally the same sequences. Moreover, the time analysis of the most populated disruptions paths shows that in most of the cases the sequence can be recognized with an advance ranging about from 0.15s to 1.5s with respect to the disruption time. Such information is readily available to the control system and can contribute to the early triggering of the avoidance action. Details of such combined analysis and application to different databases of JET, TCV and AUG tokamaks will be discussed in the paper.
        Speaker: Dr Carlo Sozzi (Istituto di Fisica del Plasma - CNR, Milano, Italy)
      • 98
        Effect of magnetic shear and the finite banana-orbit width on the neoclassical toroidal viscosity in perturbed tokamaks
        The effect of magnetic perturbations on the rotation profile in tokamak has been studied both experimentally and theoretically, since the prediction and control of the plasma rotation is one of the key issues for the stable operation. The NTV torque caused by magnetic perturbations is evaluated by adopting either a local or global drift-kinetic models. In the local models, the finiteness of the drift orbit width is neglected, and the magnetic-shear dependence of the precession frequency $\omega_B$ has been omitted. However, recent studies have found that the NTV evaluated from the global simulations, which keep the finite-orbit-width (FOW) effect and the magnetic-shear effect, are different from what the local models predict. Therefore, understanding the reason of this discrepancy in the NTV calculations is important. Here, by comparing global and local simulations, the FOW effect and the magnetic-shear effect are investigated. To study these two effects separately, we prepared two local simulation models, one neglects the magnetic-shear effect while the other keeps it in the evaluation of $\omega_B$. For electrons, it is found that the NTV profiles from the global and local codes are similar. Strong resonance of drift motions with the perturbed field occurs if $\omega_B\simeq 0$, which causes the strong NTV in low-collisionality regimes. $\omega_B$ depends on the magnetic moment and the local shear. The resonant condition in the velocity space approaches to the trapped-passing boundary as the local magnetic shear becomes more positive. In the positive-shear case, the resonant orbits are easily disturbed by small collisions and therefore the NTV evaluated by the global model tends to be smaller than that by the local one. Opposite tendency can be seen in the negative-shear case. For ions, it is found that the difference in NTV between local and global simulation becomes significant and is caused not only by the magnetic-shear effect but also by the FOW effect. In the global calculation, trapped particles see the spatial variation of the magnetic perturbations along the perturbed drift motions, while the local model assumes that a trapped particle bounces along an unperturbed field line. The non-local effect causes a significant difference in the ion NTV and in the rotation profile predicted from global and local simulations.
        Speaker: Dr Shinsuke Satake (National Institute for Fusion Science, Japan)
      • 99
        Electromagnetic Particle Injector (EPI) as a Fast Time Response Disruption Mitigation Concept

        The Electromagnetic Particle Injector (EPI) has the potential for delivering the radiative payload to the plasma center on a 3-4 ms time scale, much faster, and deeper, than what can be achieved using present methods. Predicting and controlling disruptions is an important and urgent issue for ITER. While a primary focus is the early prediction and avoidance of conditions favorable to a disruption, it is understood that some disruptions may be inescapable. For these cases, a fast time response method is essential to protect the ITER facility. Experimental tests on a proto-type system have been able to verify the predicted rapid response capability of the EPI system by accelerating a 3.2 g sabot to 150 m/s in 1.5 ms.

        The primary advantage of the EPI concept over present systems is its ability to meet short warning time scales while accurately delivering a radiative payload composed of acceptable low-Z materials such as Be, B or BN. This is done at velocities of ≥ 1 km/s required to achieve core penetration in high power ITER discharges, thus providing thermal and runaway current mitigation. This capability will provide the means for initiating a controlled plasma termination that originates at the plasma center, rather than from the outer periphery. This added capability, in addition to the fast time-response capability, should provide greater flexibility in controlling tokamak disruptions.

        *This work is supported by U.S. DOE Contracts: DE-AC02-09CH11466, DE-FG02-99ER54519 AM08, and DE-SC0006757.

        Speaker: Dr Jonathan Menard (Princeton Plasma Physics Laboratory)
      • 100
        Erosion and deposition in the JET divertor during the ITER-like wall campaigns
        During JET operation with all carbon walls prior to 2010 (JET-C) massive re-deposition of previously eroded carbon was observed in the divertor and in remote divertor areas. This massive carbon re-deposition was accompanied by a high retention of hydrogen isotopes trapped by co-deposition. Extrapolations of these results to ITER predicted very high potential tritium retention, resulting in the decision to remove carbon from the ITER divertor. One aim of the JET ITER-like wall (JET-ILW) project was to study plasma-surface interactions in a carbon-free beryllium/tungsten environment comparable to the ITER material configuration. All divertor tiles were manufactured either from tungsten coated carbon-fibre composite (CFC) material or from bulk tungsten. Erosion and deposition in the JET divertor were studied during the campaigns JET-ILW1 (2011-2012), ILW-2 (2013-2014) and ILW-3 (2015-2016) by using specially prepared divertor marker tiles using W/Mo marker layers, which were analysed before and after the campaign using elastic backscattering of 3 and 4.5 MeV incident protons and nuclear reaction analysis using 0.8 to 4.5 MeV $^3$He ions. The erosion/deposition pattern observed with the JET-ILW configuration shows partly drastic changes compared to the pattern observed with JET-C: The total material deposition rate in the divertor decreased by a factor of 4–9 compared to the deposition rate of carbon in JET-C. This decrease of material deposition in the divertor is accompanied by a decrease of total deuterium retention inside the JET vessel by a factor of about 20. The erosion/deposition pattern observed during JET ILW-2 was qualitatively comparable to JET ILW-1, the observed D inventory was roughly comparable to the inventory observed during JET ILW-1. The results obtained during JET ILW-2 therefore confirm the positive results observed in JET ILW-1. Early results from JET ILW-3 also indicate agreement; more details will become available in summer 2018.
        Speaker: Dr Matej Mayer (Max-Planck-Institut für Plasmaphysik)
      • 101
        Exhaust Behavior and Mass Balance of Tritium in Large Helical Device
        The control and management of tritium in a fusion test facility is one of the important issues from the viewpoints of radiation safety and public acceptance. As for the tritium control in a fusion test device, understanding of tritium behavior in the exhaust gas would give us new knowledge into the characteristics of the tritium release and inventory. In the deuterium plasma experiment on the Large Helical Device (LHD) which has the stainless based first wall, a small amount of tritium is produced by deuterium-deuterium reaction in the core plasma and it can be used as a tracer. A portion of produced tritium is exhausted from the vacuum vessel via the vacuum pumping system. To investigate the tritium behavior, the tritium in the exhaust gas was monitored by a water bubbler system for discriminating chemical forms and an ionization chamber. In the exhaust gas from LHD, the chemical forms of tritiated hydrogen gas was more than 95% and the tritiated hydrocarbons was a few %. Since the divertor tiles are made of carbon, a part of tritium was incorporated into the hydrocarbons by chemical sputtering. The ratio of tritiated hydrocarbon exhaust gas was less than that in JT-60U which has carbon-based first wall. On the other hands, the tritium in the plasma facing component was released by the He and D2 glow discharge cleaning operation. The tritium release mechanism was supposed to the hydrogen isotope exchange reaction and diffusion limited process. The tritium exhaust rate was gradually increased with the progress of deuterium experiment. Then, the total amount of exhausted tritium was approximately 35.5% of produced tritium at the end of the plasma experimental campaign. It suggested that two-thirds of produced tritium would be implanted in the first wall. The ratio of exhaust tritium during plasma experiment in LHD was about 1.5 times larger than that of JT-60U. Thus, the metal first wall would reduce the tritium inventory in the fusion machine. The tritium tracer study in the first deuterium plasma experiment in LHD revealed that (i) the tritium on the surface was removed by hydrogen isotope exchange reaction and the tritium release from plasma facing component was diffusion limited process, and (ii) The metal wall is one of key factors to control the tritium inventory and to reduce the tritium compounds in exhaust gas.
        Speaker: Dr Masahiro Tanaka (National Institute for Fusion Science)
      • 102
        Experimental Measurements of Cryogenic Heat Loads on SST-1 Helium Cryogenic Plant
        The SST-1 cryostat houses 130 thermal shields cooled using liquid nitrogen, 16 toroidal field (TF) coils, 9 poloidal field (PF) coils and their associated support structures. Superconducting Magnets System (SCMS) of the SST-1 consisting of TF and PF coils is designed to cool with forced flow supercritical helium (SHe) at 4 bar (a), 4.5 K and a mass flow rate of 300 g/s using helium refrigerator-cum-liquefier (HRL) of 1.3 kW equivalent cold power at 4.5 K. Last several campaigns, we have observed that the TF and PF coils could not be simultaneously cooled to 4.5 K due to heat loads from SCMS exceeding the installed cryogenic capacity of HRL. In order to cool the TF coils system at desired conditions of 4.5 K, we had to isolate PF coils as well as TF Case hydraulics from HRL at intermediate temperatures of ~ 20 K. In this specific case, the PF coils and TF Case surfaces would be at elevated temperatures in the range of 40 K – 50 K. To ascertain overall heat loads from SCMS, its associated supports structure along with the cryogenic distribution system under different cooling scenarios on SST-1 helium cryogenic plant, we have recently conducted a dedicated campaign. In this experiment, we demonstrate cool down of TF magnets in single phase supercritical helium mode to ~ 5 K for the first time. Helium supply pressure, temperature and mass flow rate are measured at the outlet of HRL before it is fed to SCMS while helium return temperature and pressure from SCMS are recorded at return line of HRL. This gives a clear picture of equivalent heat loads on HRL system. The cryogenic heat load is found to be ~ 1286 - 1350 W (+/-3%) at 5.5 K under single phase flow conditions. In the same campaign we have succeeded to cool all the nine PF coils to ~ 5 K by isolating TF coils from HRL for the first time. In this work, we report the experimental measurement procedure, instrumentation details and heat load data analysis. These results serve useful purpose in assessing the net cooling power requirement for the simultaneous cooling of the TF and PF coils and facilitate long duration plasma experiments in future.
        Speaker: Mr Nitin Bairagi (Institute for Plasma Research)
      • 103
        Experimental studies of pressure and plasma current profiles for equilibria calculations during AC transition in the ISTTOK tokamak

        In general, the operation of AC discharges in small tokamaks requires the control of a few external parameters such as vertical and horizontal fields, external heating (where available), chamber conditioning and gas puff. The dynamics and type of control used are mostly based on experimental empirical learning, with different combinations of actuators depending on the tokamak device. Experimental studies performed during the AC operation in the ISTTOK tokamak have addressed the influence of several control parameters in the success of the AC transition. Although the link between the different external actuators and plasma discharge evolution could be verified, successful AC transitions above 4 kA plasma current could not be achieved. In order to build a more predictive control of the AC transition it would be useful to develop a first principles model which interprets the experimental observations. Such model would need to combine experimental data and calculations on the equilibria and stability in several time stamps of the transition, current profile evolution, ramp-up and runaway generation, drift electrons, and the electro-technical properties of the tokamak during AC operation. The output of such model would inform the discharge controller how to balance evolution of the external actuators during the AC transition.
        The present paper presents an initial step towards the development of a deeper understanding of the equilibria and current profile during the AC transition in ISTTOK. The goal of the present study is to identify the topology of flux surfaces based on experimental pressure-like measurements and matched current profiles, the existence (or not) of antiparallel plasma currents during transition and the existence of drifting electrons and their role during current ramp-up. There is also experimental evidence on the presence of fast electrons (possibly a significant run-away fraction) playing an important role during the initial stages of the discharge immediately after the transition. This will be further investigated using colisonless numerical simulations to determine the maximum lifetime of the drift electrons and their response to H-V fields. It is important to use this electron population in combination with gas puff to produce a more efficient Townsend avalanche during the current ramp-up.

        Speaker: Dr Matthew John Hole (Australian National University)
      • 104
        Exploring Deuterium Beam Operation and Behavior of Co-Extracted Electron in Negative-Ion-Based Neutral Beam Injector
        Deuterium beam operation of the negative-ion-based neutral beam injector (N-NBI) was initiated in the Large Helical Device (LHD) in 2017. Both hydrogen (H) and deuterium (D) neutral beams were generated by changing the operation gas using the same accelerator. Comparison of the beam properties such as the extracted negative ion current and the co-extracted electron current, obtained with $\rm H_{2}$ and $\rm D_{2}$ gases, will clarify the production and extraction mechanism of the negative ions. Remarkable results are as follows: (i) 46 A deuterium negative ion current ($I_{D^-}$) has been extracted with the averaged negative ion current density of 190 A/m$^2$ by two negative ion sources in the injector. (ii) The current ratio of co-extracted electrons to negative ions ($I_e/I_{D^-}$) was 0.39 using 0.43 Pa source gas pressure. Although the configuration of the ion source is not optimized for D, the observed current of D$^-$ ions reached 82 % of the LHD requirement and those results were comparable to the ITER-NBI specification ($I_{D^-}$ = 40 A with the current density of 200 A/m$^2$ at 0.3 Pa). (iii) Linear dependence of the minimum value of the $I_e/I_{D^-}$ on the arc-discharge power is found, and is stronger in the D$^-$ operation than $I_e/I_{H^-}$ in the H$^-$ operation. The degradations of the negative ion current and the increase in the co-extracted electrons are probably caused by decrease of the surface production rate of D$^-$ ions which strongly depends on the incident D$^0$ atom velocity to the plasma grid (PG) surface. In addition, caesium (Cs) sputtering became enhanced in the deuterium discharge. This Cs behavior suggests that larger energy transfer by the deuterium ions impinging onto the PG surface removes the Cs layer required for surface production of the negative ions. These features could be a technical issue in D$^-$ beam operation in future NBI where a higher power and a longer pulse duration are required.
        Speaker: Dr Katsunori Ikeda (JpNIFS)
      • 105
        Extension of the Operating Space of High-“β_N” Fully Non-inductive Scenarios on TCV Using Neutral Beam Injection
        The fully non-inductive sustainment (“V_loop~0") of high normalized beta (“β_N”) plasmas is a crucial challenge for the steady-state operation of a tokamak reactor. In order to assess the difficulties associated with such scenarios, steady-state regimes have been explored on TCV using the newly available 1MW Neutral Beam Injection (NBI) system. Compared to the past [O. Sauter et al Phys. Rev. Lett. 84(15) 3322 (2000), S. Coda et al Phys. of Plasmas 12 056124 (2005)], plasmas closer to those expected in ITER, i.e. with significant NBI and ECRH current drive, bootstrap current and fast ion fraction, have been investigated. The operating space has been explored by carefully scanning the total auxiliary power “P_aux=P_EC+P_NB”, the NB power fraction (“P_NB/P_aux”) and the radial deposition location of the NB and EC heating and current drive. “β_N” values up to 1.4 and 1.7 at “V_loop~0” have been reached in L-mode and H-mode plasmas, respectively. Fully non-inductive operation was not achieved with NB alone, whose injection could even increase “V_loop” in presence of EC waves. Internal Transport Barriers, which are expected to maximize the boostrap current fraction, were not formed in either the electron or the ion channel in the plasmas explored to date; and this despite a significant increase in the toroidal rotation and Fast Ion (FI) fraction with NBI, which are known to reduce turbulence [J. Garcia et al Nucl. Fusion 55 (2015) 053007]. The possibility that these plasmas are Trapped Electron Mode (TEM) turbulence dominated is being analysed in dedicated transport analyses. A strong contribution of bulk and FIs to the total plasma pressure has been experimentally evidenced and confirmed by modelling (ASTRA, NUBEAM). Interpretative simulations further predict that FI charge-exchange reactions are the main loss channel for NB heating efficiency. Similar results were also obtained in inductive L-mode plasmas in a circular limited configuration at TCV [B. Geiger et al Plasma Phys. Control. Fusion 509 115002 (2017)]. Interpretative transport analysis with TRANSP coupled to NUBEAM is carried out to quantify the role of NBI losses and of the anisotropy in the FI velocity space distribution in the NBCD efficiency. A complete understanding of this evidence is crucial to the development of fully non-inductive plasmas
        Speaker: Dr Chiara Piron (ItRFX)
      • 106
        Extrapolation of Be Erosion Modelling from JET and PISCES-B to ITER
        Beryllium (Be) erosion data is one of the key issues for ITER including the first wall (FW) life time predictions [1], which undergo a re-visit based on the recent studies at the existing devices: tokamak JET equipped with the ITER-like wall (ILW) and linear plasma device PISCES-B. The extrapolation of physical and chemically assisted sputtering data is based on interpretive and predictive numerical modelling by the 3D plasma-surface interaction and impurity transport ERO code. One of the key elements is the proper treatment of the sputtering ion trajectories in the magnetic sheath, determining the angle and energy distributions by impact with the surface, and, thus the effective local sputtering yield. This recent improvement has helped to resolve the discrepancy in the normal incidence part of the factorized physical sputtering yields, which were interpreted using ERO from the JET-ILW and PISCES-B measurements. The uncertainties due to plasma-facing surface temperature and fuel e.g. deuterium (D) content in the wall are considered. The D content in plasma-wetted areas was shown to be large (the yields based on the assumed 50%D surface content, which are smaller by about a factor 3-4 than for the pure Be, lead to the best agreement with experiments). This means that the most optimistic of ITER life time predictions [1] of 4200 baseline Q=10 yields discharges based on the lowest yields (50%D) is confirmed, though somewhat corrected down due to the improved sheath model. It is important, however, to emphasize that the zero order uncertainty in these FW net erosion predictions originates in the background plasma specification which remains significant for the ITER far-SOL plasma. The advantages of the new massive-parallel ERO2.0 which allows treating the whole of JET-ILW or ITER volume, and thus providing self-consistent treatment of self-sputtering and magnetic shadowing are an additional motivation for the revisit of [1]. Furthermore, ERO2.0 enables cross-check between diagnostics, e.g. spectroscopic sightlines and filtered images from 2D cameras characterizing Be influx and plasma content, or IR images mimicking heat load distributions. Related predictive simulations of Be impurity light emission can assist in designing (sensitivity and stray light issues) the ITER visual range spectroscopy systems. [1] D.Borodin et al., 2011 Phys. Scr. T145 14008
        Speaker: Dr Dmitriy Borodin (Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung IEK-4: Plasmaphysik)
      • 107
        First Mirror Test in JET for ITER: Complete overview after three campaigns in JET with ITER-like wall
        Metallic first mirrors are essential plasma-facing components (PFC) in all optical spectroscopy and imaging systems used for plasma diagnosis. First Mirror Test (FMT) has been carried out at the JET tokamak with the ITER-like wall (JET-ILW). Over 120 test Mo mirrors were exposed in JET during the entire project. The aim is to provide an overview of results obtained for mirrors exposed during: (i) the third ILW campaign, ILW3, 2015-2016, 23.6 h plasma; (ii) all three campaigns, i.e. ILW 1-3: 2011-2016, 62h in total and (iii) a comparison to results in JET-C. Examinations were done by optical, electron and ion beam techniques. The total reflectivity of all mirrors in the main chamber has decreased by 2-3% from the initial value. All of them are coated by a very thin co-deposit (5-15 nm) containing D, Be, C and O. This has affected the optically active layer (15-20 nm on Mo) and led to increased diffuse reflectivity. No W and N have been found on the surface. All mirrors from the divertor lost reflectivity by 20-80%. There are significant differences in the surface state dependent on the location and exposure time. Reflectivity loss is connected predominantly with the co-deposition of Be and some C species. The thickest layers have been found in the outer divertor: 850 nm after ILW1-3, indicating the average growth rate of 4 pm s-1. The layers thickness is not directly proportional to the exposure time. Nitrogen, tungsten and nickel are on all mirrors from the divertor. The highest N and W contents are in the inner divertor: N reaches 1×1017 cm-2, W is up to 3.0×1016 cm-2, while the greatest Ni content is in the outer leg: 2.5×1017 cm-2. The results obtained for the main chamber mirrors allow some optimism regarding the diagnostics reliability in ITER. Tests done in JET-C and JET-ILW show that the degradation of optical properties in a machine with metal PFC is distinctly smaller than in the carbon surrounding. However, a long-term exposure and off-normal events may change surface properties of the mirrors. Laser- or plasma-induced cleaning techniques of tokamak mirrors have not brought any positive results. There are some indications that single crystal mirrors may be cleaned more efficiently than polycrystalline. Search for engineering solutions for mirror exchange in a reactor should not be abandoned especially for the divertor mirrors.
        Speaker: Mr Stefan Jachmich (BeLPP)
      • 108
        High fusion performance at high Ti/Te in JET-ILW baseline plasmas with high NBI heating power and low gas puffing
        This paper presents the transport analysis of high density baseline discharges in the 2016 experimental campaign of the Joint European Torus with the ITER-Like Wall (JET-ILW), where a significant increase in the Deuterium-Deuterium (D-D) fusion neutron rate (~2.8 x 1016 sec-1) was achieved with stable high Neutral Beam Injection (NBI) powers of up to 28MW and low gas puffing. Increase in Ti exceeding Te were produced for the first time in baseline discharges despite the high electron density, and this enabled a significant increase in the thermal fusion reaction rate. As a result, the new achieved record in fusion performance was much higher than the previous record in the same heating power baseline discharges where Ti=Te. In addition to the decreases in collisionality and the increases in ion heating fraction in the discharges with high NBI power, Ti > Te can also be attributed to positive feedback between the high Ti/Te ratio and stabilisation of the turbulent heat flux resulting from the Ion Temperature Gradient (ITG) driven mode. The high Ti/Te ratio was correlated with high rotation frequency. Among the discharges with identical beam heating power, higher rotation frequencies were observed when particle fuelling was provided by low gas puffing and pellet injection. This reveals that particle fuelling played a key role for achieving high Ti/Te, and the improved fusion performance. The impact of particle fuelling on high Ti/Te has an important implication for 2019 D-T experimental campaign, as it can provide a further increase in the fusion performance with the present heating power capability.
        Speaker: Dr Hyun-Tae Kim (EUROfusion Consortium JET)
      • 109
        Hollow pellets for magnetic fusion
        Motivated by edge localized mode (ELM) control in H-mode plasmas, we summarize experimental and theoretical progress in MHD physics of plasma interaction with small pellets ranging from 10s of microns to a few mm in size. Layered spherical structures with a hollow core (“hollow pellets”) are attractive in comparison with solid spheres and gas puffing. Theoretical results based on multi-fluid calculations of pellet-induced cold plasmoid formation and interactions with background plasmas are given. The experimental results include a new dual-spectroscopy technique for imaging of ELMs and fabrication of prototype hollow pellets.
        Speaker: Dr Zhehui Wang (Los Alamos National Laboratory)
      • 110
        Impact of Neon Injection on Electron Density Peaking in JET Hybrid Plasmas
        Impact of low-mid Z impurity injection on plasma transport and confinement has been observed and reported in several Tokamak experiments. Understanding particle transport in mixed species plasmas is crucial for reactor relevant conditions where control of DT mixture along with control of He concentration will be necessary. In this paper we present the analysis of experimental electron density profile evolution in JET hybrid scenario discharges with increasing level of Neon seeding. The measured electron flux is compared with fully predictive transport simulations in search for the possible existence of a particle inward pinch proportional to the light impurity concentration as predicted by first principle gyro kinetic simulations. These seeding experiments, performed for power exhaust mitigation studies, offered the opportunity to study systematically the effect of Neon on density peaking and to compare it with theory predictions. The database includes hybrid discharges at IP = 1.4 MA, BT = 1.9 T, βN = 2.2, additionally heated by 16.5 MW of Neutral Beam Injection power (NBI). A few of the above discharges had a small amount (< 1 MW) of Ion Cyclotron Resonance Heating (ICRH). The current ramp-up, overshooting the plateau value was used to produce a central qo ≈ 1 broad low shear region for better confinement and NTM avoidance. Neon was injected at the start of the NBI heating phase and it was already present during the transition to H-mode: when the central density reached its top value (≈ 4 s later) the Neon contribution to the total number of injected electrons ranged from 5% to 40%. Un-seeded reference discharges were also performed with the same engineering parameters. In the seeded discharges, the core density profile peaking, defined as the ratio between the central (ρ=0.25) and the pedestal density, increases up to npeak/ped ≈ 2 depending on the amount of injected Neon. Interestingly, in this database, the density peaking did not increased with the average as previously described for un-seeded discharges. Fully predictive transport simulations, carried out with JETTO code proved that the introduction of an inward particle pinch proportional to the effective charge and the ion temperature gradient, as predicted by microturbulence theory, was needed to match the data.
        Speaker: Dr Domenico Frigione (ENEA)
      • 111
        Improvement of ITER equatorial EC launcher design for poloidal steering compatibility
        This report describes the key development of the ITER equatorial ECH/ECCD launcher (EL) for poloidal steering compatibility. The steering direction of the EL has been changed from toroidal to poloidal in order to enhance the current drive capability. The design modification is being progressed toward the design finalization in 2019. The concept of upper launcher (UL) steering mechanism of steering-mirror assembly (SMA) is adapted for EL for poloidal steering. However, the redesign of the SMA for EL is needed and the key of the design is the torque balance between the bellows actuator, the coil springs and spiral pipe for mirror. Since the heat load of the steering mirror is larger than that of UL, the pipe diameter of the spiral-cooling water channel must be larger to provide more cooling water, which increases the torque of the spiral pipe. In order to compensate the increased torque, the design change of the coil spring is performed. Another of the redesign issues is the thermal stress at the Blanket Shield Module (BSM) for poloidal steering configuration. The thermal analysis shows the peak stress of the cooling channel is 820MPa, which exceeds the allowable stress limit (370MPa). By separating the first wall from the integrated shield structure, more cooling water channels can be routed close to the surface, which reduces the thermal stress of the cooling channel to around 300MPa. The mirror and waveguide unit are attached to the closure plate by rectangular flanges in the poloidal steering configuration. Because the surface pressure at the corner of the rectangular flange is high, it is impossible to keep homogeneous pressure to the rectangular vacuum seal. A simulation of the vacuum seal compression shows the necessary load for the bolts is 67.8kN, which exceeds the stainless steel bolt limit. In order to solve this problem, the introduction of the Inconel 718 bolt is considered. The 8 RF beams radiated from 8 waveguides are injected to the large parabolic steering mirror and focused to plasma. Therefore, injection angles of each beam are slightly different, which gives modified RF absorption profile compared to expected profile. In order to improve this situation, a ray tracing code is integrated with the EL optical system optimization program.
        Speaker: Dr Ken Kajiwara (National Institutes for Quantum and Radiological Science and Technology)
      • 112
        In-Vessel Inspection System: Design progress of high vacuum and temperature compatible remote handling for fusion purposes
        The plasma facing components (PFCs) in a tokamak are subjected to high heat flux and high temperature during plasma operation, which causes erosion of the first wall. There is also hot spot formation on the PFC due to physical phenomenon like thermal electron emission. In addition to fore-mentioned phenomenon, the events such as Edge Localized Mode (ELM), vertical displacement event (VDE) are serious concern for the fatigue damage of the PFCs. Therefore, health monitoring of the PFCs is an essential requirement in any tokamak, which is met by periodic inspection of the PFCs. The periodic inspection can be performed during the tokamak shutdown period or during plasma operation. The latter is most desirable as it allows quick and frequent in-service inspection of the PFCs between the plasma shots without breaking the vacuum. The work presented in this paper covers the conceptual design of In-Vessel Inspection System (IVIS) and storage chamber to carry out in-service visual inspection of SST-1 like tokamak under vacuum in between the plasma shots. The designed IVIS manipulator is ~2m long with 04- Degrees of Freedom (DOF), comprising of three rotary joints and one linear motion for deployment within the tokamak. The manipulator is designed to handle a cantilevered payload of ~1kg with a positional accuracy of <2mm. IVIS is initially stowed in a 4m long Ultra-High Vacuum (UHV) storage chamber isolated from the VV by an UHV gate valve. During (one quarter i.e. + 90°) viewing, the gate valve will open so that IVIS can be deployed inside the VV, complete the viewing procedure and return back to its initial position outside the VV. Issues like choices of the structural materials to minimize the out-gassing under vacuum and high temperature during conditioning are discussed with feasible solutions. Improvements to enhance IVIS operation under temperature and vacuum conditions for SST-1 like machine are reviewed. Results for theoretical calculations, kinematic and structural integrity analyses are presented in detail along with ways to optimize the design.
        Speaker: Mr Manoahstephen Manuelraj (Institute for Plasma Research, Gandhinagar, Gujarat-382428, INDIA)
      • 113
        Influence of Magnetic Field on Plasma Energy Transfer to Material Surfaces in ELM Simulation Experiments with QSPA-M
        Features of plasma energy transfer to the material surfaces during the plasma-surface interaction in presence of strong magnetic field are investigated within recently developed quasi-stationary plasma accelerator QSPA-M. This novel PSI test-bed facility is able to reproduce the ELM impacts, both in terms of heat load and particle flux to the surface, and to provide plasma transportation in external magnetic field, which mimics the divertor conditions. Investigations of energy transfer to the material surface have been performed for varied plasma heat load and external magnetic field value. Calorimetry, optical emission spectroscopy and a high speed imaging were applied for PSI characterization. For perpendicular plasma incidence, it has been shown that the transient plasma layer is formed in front of the surface by stopped head of plasma stream even for rather small plasma heat loads, which not resulted in surface melting. The plasma density in this near- surface layer is much higher than in the impacting stream. It leads to the arisen screening effect for the energy transfer to the surface. For B=0, the thickness of screening layer is less than 2 cm, but it increases up to 10 cm when B= 0.8 T. Reducing the size of the target leads to growth of the fraction of plasma energy, which is absorbed by the surface. For plasma exposures of tilted target surfaces, the thickness of transient plasma layer is found to be essentially non-uniform. It is maximal for downstream part of the target while the upstream surface area remains completely unprotected. The impacting plasma shifts significantly the screening layer along the surface and also generates oblique shock wave from the protruding edge. This shock wave together with available shift of plasma layer along the target provides an additional shielding for the downstream part of the exposed surface. The important contribution of external magnetic fields to the plasma energy transfer to the material surfaces is also discussed. It has been found that presence of strong external magnetic field leads to decrease of the energy, which is transferred to the exposed surface, due to the growing plasma density in near- surface layer and its increasing thickness.
        Speaker: Prof. Igor Garkusha (NSC KIPT)
      • 114
        Influence of neutral-plasma interactions on 3D scrape-off layer filaments
        Filaments are field aligned, non-linear density perturbations, which have been observed in most plasmas. In tokamaks they can carry a significant amount of particles and heat from the last closed flux surface to the far scrape-off layer (SOL). This highly non diffusive transport mechanism can cause a significant heat load onto first wall materials. It is important to understand the motion of filaments, particularly in regard to the design of future fusion devices. Recent experiments on several machines have shown that the plasma density of the SOL can have a significant influence on the dynamics of filaments. We have carried out non-linear, 3D seeded filament simulations, with the focus on neutral-plasma interactions, using the BOUT++ library. The model is an extension of the STORM code, which is a two fluid model, including thermal electrons. In order to study the influence of neutrals, 1D background profiles are computed. By varying particle and heat influx, different profiles are generated. The filaments of critical size showed an increasing radial velocity with increasing upstream temperature, as expected from scaling laws. The filament further showed a decreasing radial velocity with increasing plasma density. In these conditions, the neutrals interaction resulted in a reduced radial velocity. It was further observed that the filaments radial velocity had a strong dependency on the target temperature, resulting in an increasing radial motion for an increasing target temperature. As higher neutral densities could affect the strong sheath currents, studying the neutrals filament interaction at higher densities is of interest. In the current study the density was further increased, as the previous simulations showed an increasing influence of the neutrals on the filaments with increasing background plasma density and temperature on the filament. The purely diffusive neutral model in STORM was extended to enable the modelling of higher density conditions towards detachment. This has been validated against other neutral simulation codes. 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.
        Speaker: Mr David Schwörer (Dublin City University, Ireland)
      • 115
        Installation And Initial Run Of 96kV 7.2MW Acceleration Grid Power Supplies
        Acceleration Grid Power Supplies (AGPS) provides 8MW power at (-) 96kV to the beam source of DNB (Diagnostic Neutral Beam) and SPIDER (Source for Production of Ion of Deuterium Extracted from Rf plasma) for acceleration of negative ions with specific modulation. High Voltage Power supplies (HVPS) based on PSM (Pulse Step Modulation) topology has already demonstrated its ability for broadcast transmitters, accelerators of RF source, neutral beam injectors. PSM based 96kV/75A AGPSs have been developed to feed the Acceleration Grid of Beam sources. Design redundancy ~15% allows for tolerating SPS modules failure without leaving the ongoing campaign. The AGPS is designed to turn off in a time much lower than 100 µs to minimize the energy (20 J) delivered to the arc in case of short circuit or breakdown. AGPS mainly composed of Multi-Secondary Transformers (3nos. 2.8MVA each), Switched Power Supply (SPS) Modules (150nos., 60kW each), FPGA/Real Time based controller and other auxiliaries including passive protection devices; factory tested in witness of IO. Novel, state of the art technologies for HV insulation such us multiple bushings integrated on large resin insulators and building feedthroughs have been developed. To ensure described functionalities a single AGPS is controlled by 9 powerful synchronous FPGAs managed by real time controller which support high performance requirement of PSM based HVPS like low ripple, high resolution, programmable rise time, fast dynamics, full depth modulation, fast switching off and fast (~few milliseconds) re-application in case of breakdowns. Deliveries of AGPSs are sequenced to allow early operational drills at ITER-India lab while other unit is being installed at NBTF site. Present article describes operational drills including protection functions, insulation test and specified behaviour of AGPS on dummy load at ITER-India lab. This allows for offering DNB-AGPS for extended Factory acceptance testing.
        Speaker: Mr Narinder Pal Singh (Institute for Plasma Research)
      • 116
        Integrated modelling of core and divertor plasmas for DEMO-FNS hybrid facility
        The steady state regime for tokamak based neutron source DEMO-FNS with parameters R/a=3.2 m/1m, B=5Т, Ip=4-5 MA, PNBI=30MW and РECR=6МW is studied using a consistent modeling of the central and divertor plasma. In our formulation, the divertor plasma state is determined by the values of heat flux PSOL and the pressure of the neutrals in the divertor pn. As boundary conditions for the central plasma we use values of density and temperatures of ions and electrons on the separatrix and the neutral flux through the separatrix toward to the central plasma column. In divertor region all values calculated by the program SOLPS4.3 for a set of operating points (~30 in our case) with different values of PSOL and pn, and then the calculation results are approximated by analytical formulas. Heat transport in the central plasma is calculated using the ASTRA code and sets the scaling for the confinement time of energy IPB(y,2) with variation of H-factor. The simplified physical model for the description of the pedestal in H-mode inside the separatrix is used, based on the scalings for width and pressure at the pedestal. The density of the plasma (electrons or ions of deuterium and tritium) is modelled taking into account sources of neutrals coming from divertor region, as well as the injection of fast atoms and/or pellet injection. The neon injection is modeled to reduce the heat loads to divertor plates, that would able to radiate up to 60% of input power. The Helium plasma dilution is taken into account to estimate the maximum permissible helium confinement values. The simulation determines the window of plasma parameters DEMO-TIN, in which the heat load on divertor plates remain at an acceptable level, and the divertor plasma does not go into “detachment” mode. The dependence of these conditions on the radiation power, the impurity level, fraction of alpha-particles is investigated.
        Speaker: Dr Alexey Dnestrovskiy (NRC Kurchatov institute)
      • 117
        Isotope Dependence of Confinement in JET Deuterium and Hydrogen Plasmas
        Heat, particle and momentum confinement in L- and H-mode in deuterium (D), hydrogen (H) and in D/H mixtures have been investigated in JET. In L-mode (3T/2.5MA) at fixed density (2.5x10^19m^-3) a weak positive scaling of stored energy with ion mass, tau_Eth~A^0.15, is found [1], consistently with multi-machine scaling tau_Eth~A^0.2 [2]. Core temperature profiles are stiff with Ti~Te, and R/LTe~8 at mid-radius [1]. Flux-driven core transport modelling with TGLF show ITG’s to be dominant and predict no isotope scaling as a result of the Ti profile stiffness. A fuelling rate ~30%, higher in H than in D, was necessary to achieve the same density as in D, indicating a difference in particle confinement which was confirmed by EDGE2D/EIRENE simulations near the LCFS [1]. In type I ELMy H-mode (1T/1MA, 1.7T/1.4MA and 1.7T/1.7MA, Paux in the range 3 to 17MW) it was not possible, except in a couple of cases, to establish the same densities in H as in D, despite gas fuelling rates several times higher in H, showing a strong reduction of particle confinement. The best regression for the thermal stored energy for ELMy H-mode is obtained as W_Eth ∝ A^0.38 P^0.64 Ip^0.89 n^0.5 G^0.21 where A is the ion mass and G the fuelling rate. The mass scaling is twice that of IPB98(y,2). GENE gyrokinetic calculations in H-modes show ITG’s to be dominant in both species. The observed negative dependence of momentum confinement on the gas fuelling rate suggests that edge fuelling leads to a direct deterioration of ion heat transport. Dimensionless identity experiments for H and D pairs provided good matches for the kinetic profiles in L-mode, but not in H-mode. In H-mode the scaled confinement time in D was 30% higher than in hydrogen for the best approximate match. The evidence from these experiments suggests that the isotope scaling in these experiments, as well as the absence of good dimensionless matches in H-mode, have their origin in the pedestal and boundary region, which are sensitive to atomic physics, fuelling and recycling. [1] C.F. Maggi et al, Plasma Physics and Controlled Fusion 60 (2018) 014045 [2] ITER Physics Basis, Nuclear Fusion 39 (1999) 2175
        Speaker: Mr Henri Weisen (SPC EPFL)
      • 118
        Machine Control System for Large Volume Plasma Device: Current Status and Future Directions
        The Large Volume Plasma Device (LVPD) [1] is a cylindrical shaped pulsed plasma device dedicated in carrying investigations relevant to fusion and magneto-spheric plasma. For meeting requirements for its up gradation, efforts are in progress towards enhancing plasma duration (tpulse ~9-50ms), to cater need of controlled experiments on Electron Temperature Gradient(ETG) turbulence, a major source of plasma loss in fusion devices by suitably varying the density gradient scale length. The Machine control system (MCS) has the responsibility of protected and integrated operation of the device using standardized interface. MCS consists of namely, 1) PXIe based data acquisition system [2], 2) Modbus based process automation system [3.4] and 3) RAID configured based data handling system. The PXIe based data acquisition system is already implemented and its upgradation for data processing to convert raw signals of various diagnostics to plasma parameters and up gradation of hardware for non-linear structure study are underway.The Modbus bus has been selected for process automation of the device. Currently, process automation has been carried out for high current filament power supply and radially movable probe positioning system (~ 12 numbers). The efforts are going on for extension of the automation for 3-axis probe drive system, camera based surveillance system, axial probe positioning, vacuum production system and different power supplies. A RAID configured server is under procurement for hosting MDS+ based data handling system. The LabVIEW has been selected as supervisory data acquisition and control system for development.The novelty of the work lies in integration and handling of heterogeneous I&C controllers under single console. The paper will discuss the results obtained for integration and operation of machine control system. References: 1. S. K. Mattoo, S. K. Singh, L. M. Awasthi, et al., Physical Rev. Lett.108, 255007(2012). 2. R. Sugandhi, P. K. Srivatava, Prabhakar Srivastav, et al., 7th Int. Conference on Cloud Computing, Data Science and Engineering, IEEE conference series, 804(2017). 3. R. Sugandhi, P. K. Srivastava, A. K. Sanyasi, et al., Fusion Engineering and Design 112, 804 (2016). 4. R. Sugandhi, P. K. Srivastava, A. K. Sanyasi, et al., Fusion Engineering and Design 115, 49(2017).
        Speaker: Mr Ritesh SUGANDHI (Institute for Plasma Research, Gandhiangar, India)
      • 119
        Minimising power load asymmetries during disruption mitigation at JET
        The high thermal loads caused by a disruption of an ITER baseline scenario pulse potentially stored thermal energy of 350MJ and magnetic energy inside the vessel of 400 MJ pose a severe threat to the first wall components [1]. Massive gas injection (MGI) into a disrupting plasma has been shown to be capable of reducing the energy deposited onto the plasma facing components by increasing the radiation. However, the uneven distribution of the radiated power following a single local massive gas injection leads to highly localised radiation and hence to significant thermal loads due to the radiation “flash” [2]. In addition, the presence of the n=1 mode during the disruption produces toroidal and poloidal radiation asymmetries. Depending on the phase relationship between the n=1 mode and the MGI-location, this effect can be enhanced or diminished. In order to address this issue, JET has installed three MGI-valves at poloidal and toroidal positions similar to ITER. Single or a combination of two MGI-valves have been fired into a locked error field mode, whose toroidal O-point position was imposed by applying an external n=1 magnetic perturbation field. By measuring the radiated power at two separate toroidal locations and varying the toroidal phase of the perturbation field a toroidal peaking factor TPF, defined as the ratio of the maximum radiation to the average value, could be estimated. For a single injection TPFs in the range of 1.5 up to 1.8 have been found, depending on the type of impurity gas used. Optimising the time delay between two MGI-valves, which are toroidally at opposite locations, allowed a reduction of the TPF down to 1.2. The measured radiated power asymmetries are sensitive to small variations of the delay between the two MGI valve triggering times in the order of less than a millisecond. In this contribution the experimental findings of radiation asymmetries during mitigated disruptions caused by a seeded error field mode and the comparison with a heuristic model will be presented and the implications for the ITER disruption mitigation system discussed. [1] M. Lehnen et al., Journ. Nucl. Mat. 463 (2015), 39. [2] R. Pitts et al., Journ. Nucl. Mat. 463 (2015), 748.
        Speaker: Mr Stefan Jachmich (BeLPP)
      • 120
        New results in stellarator optimisation
        The ROSE code was written for the optimisation of stellarator equilibria. It uses VMEC for the equilibrium calculation and several different optimising algorithms for adjusting the boundary coefficients of the plasma. Some of the most importand capabilities include optimisation for simple coils, the ability to simultaneously optimise vacuum and finite beta field, direct analysis of particle drift orbits and direct shaping of the magnetic field structure. ROSE was used to optimise quasi-isodynamic, quasi-axially symmetric and quasi-helically symmetric stellarator configurations.
        Speaker: Dr Michael Drevlak (Max-Planck-Institut für Plasmaphysik)
      • 121
        Nuclear Performance Analysis and Optimization Study of Indian Solid Breeder Blanket for DEMO
        The tritium breeding blanket is the essential part of a fusion reactor which provides the tritium fuel self-sufficiency to the reactor. India under its breeding blanket R&D program for DEMO is focusing on the development of two breeding blanket concepts viz. Lead–Lithium cooled Ceramic Breeder (LLCB) and Helium Cooled Ceramic Breeder (HCCB). The study presented in this paper focuses on the neutronic design analysis and optimization of HCCB blanket which is having an edge on configuration and is one of the variants of helium cooled solid breeder blanket concepts proposed by several other countries. Indian HCCB blanket aims at utilizing the low energy neutrons at the rear part of the blanket and has RAFMS as the structural material, Lithium Titanate (Li2TiO3) as tritium breeder with Beryllium (Be) as neutron multiplier. The aim of the optimization is to minimize the radial blanket thickness, while ensuring tritium self-sufficiency and provide data for further neutronic design and thermal-hydraulic layout of HCCB blanket. Several parametric studies have been performed considering, different 6Li enrichment, varying composition of Be & Li2TiO3 in the breeder blanket and radial length of the breeder zone, as well as different arrangements of Be & Li2TiO3 layers in the blanket. The cases provided tritium self-sufficiency and sufficient shielding of the TF-coils have been identified. Neutronic calculations are performed using the 1-D discrete ordinate code ANISN with FENDL-2.1 nuclear cross section data library to assess the overall nuclear performance of HCSB blanket. The inboard and outboard blanket thicknesses of 40 cm and 60 cm respectively can give TBR > 1.3, with 60% 6Li enrichment which is assumed to be sufficient to cover potential tritium losses and uncertainties. It is found that optimal multiplier to breeder material volume fraction ratio obtained is around 3:1. The results also demonstrated that Be packing fraction has more profound impact on the TBR as compared 6Li enrichment and packing fraction of Li2TiO3. Other improvements on the TBR are seen by introducing a 10 mm breeder layer before multiplier layer behind the first wall.
        Speaker: Mr Deepak Aggarwal (Institute for Plasma Research)
      • 122
        Numerical Diagnostic to Investigate Poloidal Asymmetry in Three-Dimensional Magnetic Configurations
        Some experimental observations show poloidal asymmetry in the turbulence measurements, which can affect the plasma transport, so detailed spatial structures must be clarified. In Large Helical Device (LHD), an up-down asymmetry has been observed by the PCI diagnostic. Complicated configurations make it difficult to capture the entire structures of fluctuations in helical plasmas, so three-dimensional (3-D) turbulence simulations are necessary for understanding the mechanism. We are developing the Turbulence Diagnostic Simulator (TDS), and carry out the numerical diagnostics in helical plasmas for understanding the plasma turbulence. In this case, the gyrokinetic simulation code GKV-X provides turbulent fluctuations in 3-D configurations, and then, the TDS calculates its line-integration along the line of sight (LS) as in phase contrast imaging (PCI) to give numerical observation signals. There is a problem to resolve the local values from the line-integrated signal. The pitch angle of the magnetic field is used to help the identification of the local spectrum. A finite resolution in the local wavenumber spectrum deteriorates the reconstruction. The ITG modes have a characteristic wavelength and frequency, and difference in the spectrum can be distinguished at different radial positions, considering the spatial resolution. Characteristics of turbulence can be estimated by this analysis. The results of the TDS application indicate three factors to induce the poloidal asymmetry; 3-D magnetic configuration with the realistic LS, effect of signal processing techniques, and inherent inhomogeneity of the turbulence itself. The original data includes only small up-down asymmetry, because this is given from a single flux-tube data. The effect from the 3-D configuration generates the asymmetry, and tends to be enhanced by signal processing, but is not comparable with the experimental results. Artificial reduction of the fluctuation amplitude in the bottom half of the region can give the comparable asymmetry. This result indicates an inherent asymmetry of the turbulence. This asymmetry may be attributed to the dependence on the field line label, which is being confirmed by the GKV-X code.
        Speaker: Dr Naohiro Kasuya (Kyushu University)
      • 123
        On the role of radial electric fields on turbulence spreading in the plasma boundary of fusion devices

        Turbulence spreading is the transfer of free turbulent energy from strongly driven (i.e., unstable regions) to weakly driven locations [1]. The net effect of this phenomenon is the radial redistribution of turbulent energy, modifying local plasma features. It has been pointed out that spreading may be important in setting the Scrape-Off Layer (SOL) width. The peak heat load onto the divertor is intimately related to the SOL width, and the understanding of the mechanisms setting this width is fundamental for a reliable prediction of the SOL decay length for ITER. In this work, we report on measurements of turbulence drive and turbulent spreading, as defined by Manz, P. et al [2], from the near edge to the far SOL region of TJ-II. A 2-D Langmuir probe array [3] was used to measure both parameters as well as the profiles of floating potential, plasma density, radial turbulent particle flux, effective radial velocity, potential turbulence correlation time and phase velocity of the fluctuations. The radial electric field in the edge was modified by a biasing electrode, inserted into the edge of the plasma ($\rho \approx 0.85$), delivering a voltage $\pm$ 350 V (with respect to the wall), with a square 40 Hz waveform. All the parameters were modulated by the biasing. At -350 V, the velocity shear reached its maximum, resulting in a strong suppression of turbulent transport and the effective radial velocity fluctuations, not only at the shear layer, but also in the far SOL. Moreover, the ion saturation profile steepened at the shear layer location and was reduced in the SOL. The local turbulence drive and turbulence spreading were also impacted by the biasing. The driving term was strongly reduced in the shear layer, and only slightly reduced in the SOL. Turbulence spreading was mainly modified in the SOL when the $E_r\times B$ shear reached values close to the inverse of the turbulence correlation time in the vicinity of the Last Close Flux Surface (LCFS). In summary, biasing was found to reduce edge-SOL coupling by decreasing turbulence spreading, thus affecting the ion saturation current profile, which may have an impact on the SOL width. [1] X. Garbet et al., Nucl. Fusion 34 (1994) 963. [2] P. Manz P. et al., Phys. Plasmas 22 (2015) 022308. [3] J. Alonso. et al., Nucl. Fusion 52 (2012) 063010.

        Speaker: Dr Carlos Hidalgo (CIEMAT)
      • 124
        Overview of disruptions with JET-ILW
        This paper presents an analysis of disruptions occurring during JET-ILW plasma operations covering the period from #80128 up to #92504. The total number of disruptions was 1951 including 466 MGI (massive gas injection), VDE (vertical displacement event) and Error Field Correction Coil experiments, which led to intentional disruptions; hence the average disruption rate is 16.1%. MGI has been routinely used in protection mode both to terminate pulses when the plasma is at risk of disruption, and to mitigate against disruptions, in total 896 shots were ended by MGI. The subset of 913 natural disruptions, which were not affected by special dedicated experiments or MGI protection, was used for analysis of pre-disruptive plasma behaviour. The pre-disruptive plasma parameters of the natural disruptions are Ip=(0.82-3.14)MA, toroidal field Bt=(0.98-3.36)T, q95=(1.52-9.05), li=(0.58-1.86), betap=(0-1.1), volume average plasma density n_e=(0.2-8.5)10^19 m^-3, X-point (317 shots) and limiter (596 shots) configurations. Apart from 21 exceptional cases, the MGI was triggered by n=1 locked mode (523 shots) or by the disruption itself, specifically by dIp/dt (207 shots) or by toroidal loop voltage (145 shots). On JET only the locked mode was treated as either a precursor or the cause of disruptions. However, long lasting locked modes (≥100ms) do exist prior to disruption in 75% of cases. Though, 10% of non-disruptive pulses have a locked mode which eventually vanished without disruption. The plasma current quench (CQ) may result in 3D equilibria, termed as asymmetrical disruptions, which are accompanied by sideways forces. Unmitigated VDEs generally have significant plasma current toroidal asymmetries. The unmitigated disruptions also have large plasma current asymmetries presumably because there is no plasma vertical position control during CQ. However, MGI is a reliable tool to mitigate 3D effects and accordingly sideways forces. The vessel structure loads depend on the force impulse and force time behaviour or rotation. The toroidal rotation of 3D equilibria is of particular concern because of potential resonance with the natural frequencies of the vessel components in large tokamaks such as ITER. The amplitude-frequency interdependence is important, since a simultaneous increase of amplitude and frequency would potentially create the most challenging load conditions.
        Speaker: Dr Sergei Gerasimov (CCFE)
      • 125
        Overview of ITPA R&D Activities in Support of ITER Diagnostics
        The International Tokamak Physics Activity (ITPA) Topical Group (TG) on Diagnostics has been conducting R&D activities to support improved ITER diagnostic performance. In this paper, highlights of the Topical Group activity are overviewed: mitigation of first mirror degradation in optical systems, mirror cleaning techniques have been progressed; in-vessel stray-light has been investigated to reduce its impact on diagnostics; diagnostics of escaping particles, feasibility test of the activation probe technique has progressed under a multi machine joint experiment. Diagnostic systems are essential for machine protection, reliable machine operation and comprehensive understanding of burning plasma behavior in ITER [1]. In order to achieve the above aims, more than fifty sub-systems will be developed for measurement of plasma and plasma facing components in the harsh ITER environment, e.g. higher neutron/ -ray irradiation and lower accessibility/maintainability compared to that of existing fusion devices. ITPA Diagnostics TG has addressed common physics issues in diagnostics development [2]. The TG activity is mainly directed to High Priority research areas (HP); HP-1: Optimization of the life-time of plasma facing mirrors used in optical systems, HP-2: Assessment of impact of in-vessel wall reflections on diagnostic systems, HP-3: Develop methods of measuring the energy and density distribution of escaping alphas HP-4: Plasma control system measurement requirements HP-5: Develop diagnostic calibration techniques/strategies compatible with the burning plasma environment and Joint Experiments for Diagnostics (JEX-DIAG) under a framework between ITPA and the Implementing Agreement on Co-Operation of Tokamak Programs of the International Energy Agency (IEA); JEX-DIAG-2: Environmental tests on first mirrors, JEX-DIAG-5: Field test of an activation probe, JEX-DIAG-6: Cross comparisons of Charge Exchange Recombination Spectroscopy and X-Ray Imaging Crystal Spectroscopy, JEX-DIAG-7: Distributed monitoring of microwave power density, JEX-DIAG-8: Benchmark of Wall reflections, JEX-DIAG-9: Spectral MSE (MSE-LS) experiments as design driver for ITER JEX-DIAG-10: Minimizing microwave absorption in vacuum windows JEX-DIAG-11: Determination of the runaway electron distribution function by spectral Bremsstrahlung measurements in the gamma-ray energy range.
        Speaker: Dr David Brower (University of California Los Angeles)
      • 126
        Performance Evaluation of 1.3 kW at 4.5 K Helium Refrigerator/ Liquefier (HRL) at IPR
        At IPR, 1350 W at 4.5 K helium cryo plant is dedicated to facilitate the cooling requirements of SST-1 machine. Since 2004, helium refrigerator / liquefier (HRL) (Make: M/s. Air Liquide, France) is operational in mixed mode equivalent to 650 W (refrigeration power) and 200 lh-1 (liquefaction capacity) at 4.5 K. The HRL can be operated in two phase (1.3 – 1.5 bar (a) at 4.5 K – 4.7 K) as well as single phase supercritical helium (at 4 bar (a) and 4.5 K with nominal mass flow rate of 300 gs-1 ) modes of operation. The refrigeration capacity of the HRL is 650 W at 4.5 K used to make TF and PF coils superconducting whereas the remaining capacity of 200 lh-1 is utilized for powering the vapor cooled current leads system of SST-1 at rated current of 10 kA. To ensure the availability of the HRL and its best performance as per the needs of long duration SST-1 experiments, we carry out preventive maintenance of the different cryogenic components and subsystems as per defined schedule. These activities result in increasing the life span of the HRL as well as ensure its maximum availability during SST-1 operation. M/s. Air Liquide envisaged to carry out every five years preventive maintenance of the HRL for all the sub-systems and components. After major maintenance, it is desirable to have performance test on the HRL. We have carried out major preventive maintenance of the HRL and measure the HRL capacity during 2009-2010. Recently, we have further carried out the maintenance ourselves and carried out the performance test. The equivalent cold power of HRL found to be 1160 W (in pure refrigeration mode), 10.7 gs-1 (in pure liquefaction mode) and 1300 W equivalent (in mixed mode) at 4.5 K. These values matches with our last experimental measurements during HRL maintenance performed in 2010 and as expected considering the operational hours of HRL after thirteen years of operation. These results are quite satisfactory from the HRL performance point of view. The HRL capacity strictly depends on the different modes of operations. In this paper, we report the performance evaluation of cold capacity of HRL at IPR since it commissioning to till date.
        Speaker: Mr Rohitkumar Panchal (Institute for Plasma Research)
      • 127
        Performance of the plasma source and heating concept for the Prototype-Material Plasma Exposure eXperiment (Proto-MPEX)
        The Material Plasma Exposure eXperiment (MPEX) is a planned linear plasma device to address plasma-material interactions for future fusion reactors. Its concept does foresee the capability to expose apriori neutron irradiated material samples to fusion reactor grade divertor plasmas. This new capability will be unique world-wide addressing important research needs in the area of fusion nuclear science. It will be an evolution to current operating steady-state linear plasma devices, which are limited either in plasma fluxes they can deliver to the material targets or plasma temperatures (for ions and electrons) they can reach in front of the material targets. The concept of MPEX does foresee a combination of a high-power helicon plasma source with microwave electron heating and ion cyclotron resonance heating. This source and heating concept is being tested on the Prototype-Material Plasma Exposure eXperiment (Proto-MPEX). With 100 kW helicon power a plasma density of 8e19 m-3 was achieved, which is about a factor 2 more than required for MPEX. Electon heating was pursued with a 28 GHz gyrotron. A maximum power of 50 kW was delivered to the plasma, which is produced by the helicon. At this frequency, the plasma is overdense in the plasma center (> 1e19 m-3). Maximum electron temperatures of 20 eV have been achieved under those overdense plasma conditions with Electron Bernstein Wave EBW) heating. This is almost the electron temperature required for MPEX (25-30 eV). Ion cyclotron heating (ICH) was performed in the frequency range of 6 - 12 MHz with a low power ICH antenna able to launch about 25-30 kW of power. Without ICH, the ion temperature is about 2-4 eV. With ICH ion temperatures of 8-12 eV were measured. The ion fluxes to the target are about 5e23 m-2s-1. The plasmas produced by the helicon antenna have been modeled extensively with a fluid plasma code, coupled to a Monte-Carlo neutral code (B2-Eirene). The plasma transport can be well explained by this fluid approach and a radial diffusion coefficient consistent with Bohm-like transport. The transport of auxiliary heated plasmas (ECH/EBW and ICH) is currently being investigated and experimental results of this investigation will be presented.
        Speaker: Dr Juergen Rapp (Oak Ridge National Laboratory)
      • 128
        Plasma and diagnostics preparation for alpha-particle studies in JET DT
        A deuterium-tritium (D-T) experimental campaign DTE2 on JET scheduled for 2019-2020,will be done in the Be/W vessel and will address essential operational, technical, diagnostic and scientific issues in support of ITER [1]. In preparation for the campaign, developments were performed on JET aiming at studies of alpha-particles. For studying AEs driven entirely by alpha-particles, a scenario similar to the TFTR beam “afterglow” [2] was developed for JET. In DT plasmas, after NBI is switched off, alpha-particles will be the only energetic ions during time interval between slowing-down times for NBI-produced ions and alpha-particles. Detection of alpha-driven AEs in this time window may help in diagnosing the temporal evolution of the pressure profile and slowing-down time of alpha-particles. JET advanced tokamak scenarios with q ≈ 1.5-2.5 were chosen and discharges have been successfully developed. The transport modelling extrapolated to DT predicted that alpha-particle beta of ≈ 0.1% could be achieved comparable to that in successful TFTR experiments. In “hybrid” scenario plasmas with q0 ≥ 1, fast ion losses in the MeV energy range were observed during n=1 fishbones driven by a resonant interaction with D beam ions in the energy range ≤ 120 keV [3]. The losses are identified as an expulsion of D-D fusion products, 1 MeV tritons and 3 MeV protons. A mode analysis with the MISHKA code combined with the study of nonlinear wave-particle interaction with HAGIS show that the loss of toroidal symmetry strongly affects the confinement of high energy tritons and protons by perturbing their orbits and expelling them in a good agreement with experiment. The extrapolation to the case of alpha-particles in DTE2 hybrid scenarios with similar fishbones has shown an additional alpha-particle loss of ~ 1% [3]. References: [1] Weisen H et al., Fus. React. Diag., AIP Conf. Proc. 1612, 77-86 (2014); [2] Nazikian R et al., PRL 78, 2976 (1997); [3] Fitzgerald M et al, submitted to Nucl. Fusion (2018).
        Speaker: Dr Sergei Sharapov (CCFE)
      • 130
        Predict-First Analysis and Experimental Validation of MHD Equilibrium, Stability, and Plasma Response to 3D Magnetic Perturbations
        An integrated-modeling workflow has been developed to predict equilibria and response to 3D magnetic perturbations in tokamak experiments. Starting from an equilibrium reconstruction from a past experiment, the workflow couples together the EFIT Grad-Shafranov solver, EPED model for pedestal stability, and NEO drift-kinetic-equation solver (for bootstrap current calculations) in order to generate equilibria with self-consistent pedestal structures as the plasma shape and various scalar parameters (e.g., normalized beta, pedestal density, $q_{95}$) are changed. These equilibria are then analyzed using automated M3D-C1 to compute the MHD plasma response to 3D magnetic perturbations. The workflow was created in conjunction with a DIII-D experiment studying the effect of triangularity on plasma response, showing excellent agreement between the analysis of the workflow's equilibria and equilibria reconstructed from the experiment. Various versions of the workflow demonstrated that the details of the edge current profile were not important for these cases, while $q_{95}$ and the details of the global pressure profile had a significant impact on the results. A predict-first study was then carried out for a DIII-D experiment examining how plasma response varies between single- and double-null shapes. The predicted equilibria were used to guide experimental planning and the predicted response was found to agree well with the perturbed magnetic field measured on the high-field-side midplane. Applications of this workflow to KSTAR and EAST experiments will also be explored. This work forms the basis of predictive scenario development across current and future devices (e.g., ITER), allowing for higher-fidelity predictions of MHD stability and 3D plasma response. Work supported by the US DOE under DE-FG02-95ER54309 and DE-FC02-04ER54698, along with NFRI, South Korea.
        Speaker: Dr Brendan Lyons (General Atomics)
      • 131
        Preliminary Design of IN-DA Diagnostic Plant Instrumentation & Control
        In ITER, plant Instrumentation & Control (I&C) components are exposed to harsh environment. Hence I&C for plant system is one of the most challenging requirements to fulfil the ITER demands. IN DA is responsible for the delivery of several diagnostics systems including (1) X-Ray Crystal Spectroscopy (XRCS) edge & survey to measure the plasma impurity for machine protection and basic control as well as well measure the profile of plasma parameters for advanced control. (2). Electron Cyclotron Emission (ECE) – to provide temperature profile, fluctuation and the power due to emission in 70 -1000 GHz for ITER plasma and (3). UP#09 for housing the diagnostics in the port plug and integration with rest of its system. (4) CXRS-pedestal – to provide the ion velocity and temperature of plasma for advanced control & physics studies. Out of these the XRCS diagnostics and UP#09 are classified as protection important components (PIC) and hence need special attention to achieve safety functions. ITER plant I&C systems are being developed according to industrial systems engineering standards, and compliant with the standards, specifications and interfaces defined in the Plant Control Design Handbook (PCDH) and its satellite documents. Enterprise Architect (EA) is the tool is to be used for diagnostic I&C documentation. EA helps to trace high-level specifications for the analysis, design, implementation, test and maintenance models using Unified Modelling Language (UML), SysML and other open standards. I&C deliverable documents are being developed according to the ITER diagnostic guidelines. This paper describes the detailed preliminary design of plant I&C for diagnostics system a) Operation procedures, b) functional analysis including variable definition, c) hardware architecture and signal d) cubicle configuration and e) the plant system operating state machine (PSOS) for automation including the mapping to Common Operation States (COS).
      • 132
        Preliminary development on a conceptual first wall for DEMO
        For a DEMO reactor, the first candidate material for plasma facing material(PFM) is tungsten(W) and current available structure material is reduced activation ferritic/martensitic (RAFM). And tungsten coating material is promising to be applied on first wall. Since chemical vapor deposition tungsten (CVD-W) has a higher density, less porosity and better thermal shock resistance, thick CVD-W coating is used as the plasma facing material here. Southwestern Institute of Physics (SWIP) has developed a new RAFM material, which is called CLF-1. The new conceptual first wall for DEMO in this work is designed and developed with CVD-W and CLF-1. Due to different thermal expansion coefficients of tungsten and steel, CVD-W will detach from CLF-1 steel under heat load and plasma exposure if it is coated onto the CLF-1 directly. As a result, an interlayer must be applied to mitigate the stress between CVD-W and CLF-1. Furthermore, the tungsten will generate cracks under steady state and transient heat loads in reactors and crack in the tungsten will make tritium to penetrate into the substrate rapidly. Tritium accumulation is a critical parameter for reactors which is very important for safety and steady state operations. The new conceptual first wall consists of CVD-W, CLF-1 and an interlayer between them. The interlayer is required to have good bonding property and tritium prevention, which is crucial for controlling the inventory buildup and maximizing the fuel efficiency. SiC and TiN applied as the interlayer between W and CLF-1 in the first wall are investigated. In order to figure out the influence of fabrication technology, layer thickness and coating rate, a series of material samples are fabricated and tested. The SiC interlayer on the CLF-1 substrate is made by three means of coating technologies including physical vapor deposition (PVD), chemical vapor deposition (CVD) and Chemical Vapor Infiltration (CVI) while TiN interlayer is obtained by CVD. On the top of the interlayer SiC or TiN, a CVD-W layer with the thickness of 1mm is coated with the rate of 0.5mm/h at the temperature of 450-550 oC. The material analysis and mechanical tests on those samples present that SiC by CVI and TiN by CVD and TiN by CVD have sufficient adhesiveness as an interlayer between W and CLF-1, which show good bonding property and no obvious detachment or delamination is found.
        Speaker: Dr Laizhong Cai (Southwestern Institute of Physics)
      • 133
        Preliminary Pipe Stress Analysis of High Pressure, High Temperature Experimental Helium Cooling System
        Experimental Helium Cooling Loop (EHCL) is a high pressure-high temperature helium gas system. EHCL is similar to the First Wall Helium Cooling System (FWHCS) of LLCB TBM and in this loop First wall mock ups up to one fourth (¼) size of TBM can be tested. EHCL modelling consists of equipment arrangement, pipe routing, support, cable tray routing, instrumentation arrangement and tube routing. EHCL lab floor dimensions are 18m x 18m length and width respectively while the vertical height is 5 meter. The lab is divided in three major areas: process area, control room and free space for maintenance activities. The process and control room covers 9m x 9m and 14m x 5m floor area respectively. The EHCL is designed to operate with helium gas at 8.0 MPa (gauge) pressure and at 300-400 C temperature. The flow rate varies from 0.2 kg/s to 0.4 kg/s. The selected size for the connection pipes is DN 50. The high temperature pipes in this loop are at 400 C and at 8 MPa pressure, and these pipes are connected to equipment in a limited space. The detailed flexibility analysis was carried out, to ensure safety of the piping system and to maintain the structural integrity under loading conditions (both external and internal), which may occur during the lifetime of the system. SS 316L is used as structural material for piping and equipment. This poster presents the modelling of EHCL and the results of detailed flexibility analysis of EHCL pipes. To carry out the analysis, the entire piping system of the loop was modeled and the static and dynamic analysis was carried out in CAESAR II software. For the floor response spectra, the floor level in two horizontal and one vertical direction was computed. As IPR lies in seismic zone –III, and the process loop is planned to be located at ground level at IPR campus, accordingly the FRS was used to find out the induced stress in the process loop. The dynamic effect and weight effects are considered in the design so that the stresses created by the combined loads do not exceed the allowable stresses prescribed by the design codes. Finally the piping layout satisfying the code requirements along with the results are presented in the poster.
        Speaker: Mr Aditya Kumar Verma (Institute For Plasma Research)
      • 134
        Preliminary results of prototype Martin-Puplett Interferometer and transmission line developed for ITER ECE Diagnostic
        The ECE Diagnostic system in ITER will be used to determine the electron temperature profile evolution, the high frequency fluctuation of the plasma electron temperature, the characterization of runaway electrons and the radiated power in the electron cyclotron frequency range (70-1000 GHz). These measurements will be used for advanced real time plasma control (e.g. steering the electron cyclotron heating beams) and the ITER plasma physics studies. In view of the ITER requirements, an ultra-wide band (70 – 1000 GHz) transmission line coupled to a fast scanning, broadband spectrometer are required to estimate the ECE radiated power loss and to study the behavior of runway electrons. Typically, the transmission lines and spectrometers are not operated in vacuum and there are consequently significant losses at certain frequencies due to water vapor line absorption over this large frequency range. To avoid these losses, both the transmission line and the spectrometer must be operated in vacuum. Further, producing an efficient high etendue long wavelength spectrometer with extremely high scan speeds in vacuum is a major challenge. Also long distance (~ 43 meters) transmission of very low in-situ calibration source power (~ nW level) with an ultra-wide frequency range is another challenge for the transmission line development. For the purpose, a prototype polarizing Martin–Puplett interferometer has been developed to operate in a low vacuum with high throughput and excellent time resolution of 10 ms with scanning length of 15 mm. And also a prototype transmission line to be used in vacuum is developed. An experimental set up has been established at ITER-India lab to test the performance of various prototype subsystems of the ECE diagnostic. The experimental set up consists of the high temperature black body source in this frequency range, transmission line and the Martin-Puplett interferometer with data acquisition system. This paper describes the experimental set up and preliminary results of subsystems developed for ECE diagnostic.
        Speaker: Dr HITESH KUMAR PANDYA (Institute for Plasma Research)
      • 135
        Progress in Development and Fabrication of the JT-60SA ECH/CD System
        Development of the ECH/CD system for JT-60SA has been progressed. Successful results on the JT-60SA gyrotron development for multi-frequency, high-power, long-pulse oscillation such as 1 MW/100 s at both 110 and 138 GHz, 1.9 MW/1s and 1.5 MW/5s at 110 GHz, 1.3 MW/1.2 s at 138 GHz and 1 MW/1 s at 82 GHz were reported in IAEA FEC in 2014 and 2016. The development of the high-power, long-pulse and multi-frequency JT-60SA ECH/CD system is now focusing on the launcher, transmission line (TL), control and power supply. In addition, design, fabrication and testing of a part of these components have been progressed toward start of the first plasma experiment of JT-60SA. The main results achieved in this time are as follows. (i) A full length (~7 m) mock-up of the mirror steering structure of the launcher has been successfully tested in vacuum. The required life without maintenance, which is 10^5 cycles for the poloidal steering range of 60° and 10^4 cycles for toroidal beam steering range of 30°, has been achieved. A newly introduced solid lubricant enabled the smooth movement of the steering shaft by reducing the sliding resistance between balls and rail/block of the linear guide used in the steering structure. (ii) The temperature rise distribution of aluminum waveguides has been measured at high-power of 0.5 MW. It is in the range from 0.2 to 1.2 °C per 1 MJ transmission and acceptable for 1 MW/100 s (100 MJ) transmission required in JT-60SA. (iii) The preparation of the JT-60SA ECH/CD system is progressing as planned. For instance, an ECH/CD control system has been designed with a layered and distributed structure to achieve sufficient flexibility for upgrading and for easy optimization depending on the experimental purposes. In addition, fabrication of TL components including waveguides, cooling and vacuum system has been started. Moreover, the new power supplies for two gyrotrons (1 MW/100 s each) have been designed and the fabrication has started by F4E as a part of broader approach activities. The above discussed progresses in the launcher/waveguide developments and the design /fabrication of the JT-60SA ECH/CD system components significantly contribute to smooth start of the JT-60SA experiment and improve the plasma performance with high reliability and flexibility.
        Speaker: Dr Takayuki Kobayashi (National Institutes for Quantum and Radiological Science and Technology)
      • 136
        Progress on Lithium Ceramic breeder materials development, Characterization and R&D activities in IPR
        Several materials have been developed and being investigated for reliable and sustainable breeder candidate material. Lithium meta-titanate (Li2TiO3) and Lithium ortho-silicate(Li4SiO4) are the prominent among the suitable candidate materials for breeders. India has proposed lithium meta-titanate (Li2TiO3) as the tritium breeder materials in the form of pebble bed for LLCB TBM. Li2TiO3 power was prepared by solid state reaction using LiCO3 and TiO2 followed by ball-milling and calcination. Li2TiO3 pellets and pebbles are prepared from this powder followed by high temperature sintering. Effect of sintering time and temperature on the properties of pebbles has been studied. At every stage of preparation, extensive characterizations are being carried out to meet the desired properties of these materials.The geometry and loading conditions of the breeder blankets makes the analysis complex. For a robust design of blankets requires a thorough understanding of the thermo-mechanical response of the breeder materials at different loading conditions. In this context, the material characterization plays a vital role in determining the breeder response.It is essential to measure the mechanical and thermo-mechanical properties of pebble bed. Experimental set ups have been built indigenously at IPR for the measurement of effective thermal conductivity of pebble bed using steady state-axial heat flow and transient hot wire methods. The effective thermal conductivity (keff) of pebble beds is an important parameter for the design and analysis for a fusion tritium breeder blanket. The keff of Li2TiO3 pebble bed is measured as a function of average bed temperature from RT to 500 °C in different environment (vacuum, helium gas etc.).Initial results obtained from these experiments will be discussed in this paper. Details of lithium ceramic breeder material development, their characterizations and related R&D activities will be discussed in this paper.
        Speaker: Dr Paritosh Chaudhuri (Institute for Plasma Research)
      • 137
        R&D status of Indian Test Facility for ITER DNB characterization
        Indian Test Facility (INTF) is a R&D facility under development at Institute for Plasma research (IPR), Gandhinagar as a part of the neutral beam development from negative ion source (NNBI) program. The major advantage of the INTF besides developing the beams from large ion sources is to characterize and benchmark the ITER Diagnostic Neutral Beam (DNB) to the desired specifications over transport lengths of ~ 21m, a unique feature of this test bed. Such a study will enable establish the expected power to be delivered by DNB into the ITER plasma, an important parameter to estimate the S/N ratio expected from the He ash measurements by CXRS in ITER plasmas. The INTF beam line has a one to one correspondence with the DNB in ITER in terms of the components, their placement and the inter component distances. However, the 9 m long 4.5 m dia vacuum vessel with a top openable lid and with double O rings seals for the vacuum is different from the rectangular vessels envisaged at ITER. The other difference is the 12 number of modular cryopumps providing the same pumping surface as the single panel ITER cryopumps. In addition, the beam characterization at 21 m is planned with a second calorimeter housed in the vacuum vessel connected to the end of the duct. The Data acquisition and Control system is developed using ITER CODAC platform and integrates around 800 channels from all plant systems for enabling safe remote operations. Extensive physics and thermomechanical calculations for various types of operational heat loads and loads due to various accidental scenarios have been performed to finalise the component design. Adequate choice of materials, manufacturing and jointing processes compatible to ITER safety standards has been made in order to make the components adhere to the safety and quality classification thereby ensuring that the components survive the ITER life time while operating in harsh nuclear environments. The components are currently in various phases of manufacturing and the first operations INTF are anticipated in Q4 of 2019. The experiments on INTF are supported by single driver test bed, ROBIN, and the two driver TWIN source. The paper will describe the R&D status of different components and auxiliary systems of Indian Test Facility (INTF), the envisaged experimental program of operation and some results from operational test beds.
        Speaker: Prof. Mahendrajit Singh (ITER - India Institute for Plasma Research Bhat Gandhinagar Gujarat 382428 India)
      • 138
        Recent Progress on the Production and Testing of the ITER Central Solenoid Modules
        Several key milestones have been completed recently in the fabrication of the Central Solenoid (CS) modules for ITER. The qualification coil has been completed and tested with many lessons learned that have now been incorporated into the processing of the production modules. Currently four modules are in production with the first module scheduled for completion in 2018 followed by full current testing at 4.5K. Shipment of the first module to Cadarache is scheduled for 2019, arriving in advance of its need date. The Central Solenoid is a key component of the ITER tokamak providing the inductive voltage to initiate and sustain the plasma current and to position and shape the plasma. The design of the CS has been a collaborative effort between the US ITER Project Office (USIPO), the international ITER Organization (IO) and General Atomics (GA). GA is fabricating seven 110 tonne CS modules (one is a spare). After arrival at the ITER site, the six modules will be stacked in the Assembly Hall, the structure added and transferred in a single lift to the ITER tokamak. In a dedicated facility in Poway, California, USA, GA is currently fabricating the modules, with each one requiring approximately 22 months start to finish. Following fabrication a series of tests including high voltage testing of the insulation, full current testing of the conductor at 4.5K and a repeat of the high voltage tests at room temperature are performed. The testing duration is an additional five months for each module and is the program critical path. Recently, the qualification coil was completed, electrically tested, and cooled to 4.5K by supercritical helium. While at 4.5K, a series of tests were performed which simulated those tests that will be performed on the modules to validate the test methods and equipment. After the tests were completed, the mockup coil was dissected to determine the quality of the resin injection. This paper describes some of the challenges in accomplishing the recent milestones in completing the qualification coil fabrication and testing, the implications on the module production, and the status of the module production.
        Speaker: Mr John Smith (General Atomics)
      • 139
        RFQ Commissioning of Linear IFMIF Prototype Accelerator (LIPAc)
        The IFMIF project aiming at material tests for a future fusion DEMO reactor is under the EVEDA phase in the BA Agreement of fusion program between Japan and EU. As the accelerator activity, the installation and commissioning of the Linear IFMIF Prototype Accelerator (LIPAc) is at the second stage of demonstration of the feasibility of the low energy section of an IFMIF deuteron accelerator up to 9 MeV with a beam current of 125 mA, CW. The installation of injector, RFQ, MEBT, D-Plate and LPBD for LIPAc with 8 coaxial high-power transmission lines and RF power system was just done in 2017 at Rokkasho, Japan. After that, the RF conditioning of RFQ for beam commissioning is underway. The beam commissioning of RFQ with H+/D+ and the acceleration demonstration up to 5MeV-125mA-0.1% duty cycle with D+ will be done.
        Speaker: Dr Atsushi Kasugai (Japan Agency for Quantum and Radiological Science and Technology (QST), Rokkasho Fusion Institute)
      • 140
        Runaway electron beam stability and decay in COMPASS
        Runaway electrons (REs) as one of the yet unsolved threats for ITER and future tokamaks are a topic of intensive research at most of the tokamaks. The experiments performed on COMPASS are complementary to the experiments at JET and MST (Medium-Size Tokamaks), building on the flexibility of the diagnostics set-up and low safety constraints at this smaller (R0=0.56 m, a=0.23 m) device. During the past couple of years two different scenarios with the RE beam generation triggered by gas injection have been developed and investigated. The first one is based on Ar or Ne massive gas injection (MGI) into the current ramp-up phase leading to a disruption accompanied by runaway plateau generation, while the second uses smaller amounts of gas in order to get runaway current dominated plasmas. The successful generation of the beam in the first scenario depends on various parameters, including the toroidal magnetic field. The generated beam is often radially unstable, and the stability seems to be a function of various parameters, including the value of current lost during the CQ. Surprisingly, the current decay rate of the stable beams is rather similar in most discharges. The second scenario is much more quiescent, with no observable fast current quench. This allows to better diagnose the beam phase and also to apply secondary injections or resonant magnetic perturbations (RMP) to assist the decay of the beam. In this regard, interesting results have been achieved using secondary deuterium injection into a runaway electron beam triggered by Ar or Ne injection and also using n=1 error field generated by top and bottom RMP coils. While D dilution is clearly able to almost stop the beam decay, RMPs help to accelerate the beam decay. The effect of RMPs seems to be very different when acting on Ar and Ne background plasmas. Very interesting effects have been observed also by the high-speed cameras, including filamentation during the application of the RMPs and a slow local variation of the light intensity similar to turbulence during the beam decay.
        Speaker: Mr Ondrej Ficker (CzIPP, CzFNSP)
      • 141
        Seismic Analysis Of High Power Amplifier in ITER ICRF Range
        ITER-India is responsible for delivery of 8+1(prototype) RF sources to ITER project. Each RF source will provide 2.5MW of RF power at VSWR 2:1 in the frequency range of 35 to 65MHz. Eight such RF sources will generate total 20MW of RF power. Two RF chains containing three high power amplifiers (HPA1, HPA2 and HPA3) need to be combined to build an RF source. HPA2 and HPA3 are RF tube based amplifiers while HPA1 is a solid state power amplifier. This paper covers detailed seismic analysis of High Power Amplifiers for worst case seismic loading condition. A SL-2 seismic event has been analyzed to determine potential areas that will require inspection and/or replacement. According to the design basis, a Response Spectra Analysis (RSA) has been performed for the frames and cavity of high power amplifiers which includes the self-weight of all structural members, platform dead weight and reactions from the base. The RSA requires a modal analysis to be performed which is used to determine the rigidity of the support structures. The accelerations of the Zero Peak Acceleration (ZPA) are applied in order to account for all masses. All structures and components must respect the requirement that there must be no failure that would prevent a SIC-1 or SIC-2 component from performing its safety function. The ANSYS software is used for Modal analysis and Response Spectrum analysis. This paper will also point out the maximum stressed link in structure and modifications may be proposed to achieve the required strength.
        Speaker: Mr Rohit Anand (ITER India , IPR)
      • 142
        Shattered Pellet Injection Technology Design and Characterization for Disruption Mitigation Experiments
        The technology of forming high-Z cryogenic pellets mixed with D2 that are shattered upon injection into a plasma has been developed at ORNL for mitigating disruptions and has been selected as the basis for the baseline disruption mitigation system on ITER. In these shattered pellet injectors (SPIs), large pellets of neon and argon mixed with D2 are formed from gas and are shattered upon impact with a bent tube just before entering into disrupting plasmas in order to radiate the plasma energy to mitigate possible damage to in-vessel components [1]. In support of disruption mitigation research for ITER, SPI systems have been designed and fabricated for use on thermal mitigation and runaway electron dissipation experiments on DIII-D and JET. These systems have common features of 3 barrels of different size pellets that are formed in-situ and collimated into a single injection line. The shatter tubes are bent stainless steel tubes that are mounted inside the vacuum vessel of the tokamak. The large pellets are formed in-situ from the low pressure gas feed into the barrels that are cooled with liquid helium and held intact ready to fire until needed. Pressurized gas is also used to accelerate these pellets with gaps in the injection lines to remove as much of the gas as is practical to avoid influencing the plasma shutdown. Solid pellets of argon in particular present a challenge to fire the pellet because of high shear stress, thus mechanical punches have been developed that can apply higher impact to release these pellets. Punches using high pressure gas and solenoid drivers have been developed. Tests of gas punches have revealed that argon can be released and achieve speeds up to 160 m/s for 8 mm size pellets. The slower pellet speeds achieved with a punch have been found to result in larger fragment sizes, which is appealing for deeper penetration in high performance plasmas. Higher speed pellets that are achieved with high pressure gas and high deuterium content in the same shatter tube result in finer particles and higher gas content in the resulting shatter material spray. [1] N. Commaux, et al., Nucl. Fus. 50 (2010) 112001. [2] L. R. Baylor, et al., Fus. Sci. Tech. 68 (2015) 211.  This work was supported by ORNL managed by UT-Battelle, LLC for the U.S. Department of Energy under Contract Nos. DE-AC05-00OR22725 and DE-FC02-04ER54698.
        Speaker: Dr Larry R. Baylor (Oak Ridge National Laboratory)
      • 143
        Singlet Breakdown Optimisation to a Doublet Plasma Configuration on the TCV Tokamak
        This paper presents a fresh attempt on TCV to optimise plasma breakdown in the break-less (45 micro-Ohm impedance) vacuum vessel culminating in a double breakdown and the formation of a doublet configuration. A statistical analysis of legacy single pole breakdown and early plasma current ramp failures helped modify vessel current estimators together with PSU command and control issues to obtain reliable plasma initiation +/-30cm in TCV’s 3:1 elongated vacuum vessel. Although precise control of the vacuum null was achieved, control of the high plasma ramp rate proved complex since the highest (~10v) loop voltage was necessary for reliable breakdown and, through trial, the acceptable range of pre-fill pressures was limited. A double breakdown with simultaneous, separated, magnetic nulls was then achieved. Initial ohmic heating alone was limited by lobe separation instabilities with the upper lobe merging into the lower lobe after ~15ms. Plasma multipole control was attempted using two X2 gyrotrons, aimed at each lobe’s, core to modify each lobe’s resistivity and thus current. A transport barrier in the mantle surrounding the doublet configuration was observed with both lobes seemingly heating independently of the ECH heating location. To date a combined plasma current of 260kA after 20ms was obtained for which Thomson density and temperature profiles indicate two clear plasma lobes. Doublets are predicted to offer increased Beta limits, vertical stability and the potential of a novel solution to divertor exhaust where the entire mantle, surrounding the plasma, may be available for exhaust dissipation.
        Speaker: Dr Basil Duval (Ecole Polytechnique Fédérale de Lausanne – Swiss Plasma Center (SPC), Association Euratom-Confédération Suisse(EPFL) CH-1015 Lausanne, Switzerland)
      • 144
        SOL transport and detachment in alternative divertor configurations in TCV L- and H-mode plasmas
        The effect of magnetic geometry on scrape-off layer (SOL) transport and detachment behavior is investigated on the TCV tokamak with the goal of assessing the potential of alternative divertor geometries and for the validation of theoretical models. L-mode experiments reveal that increasing connection length and hence divertor volume by either increasing poloidal flux expansion or divertor leg length have different effects on the boundary plasma. In attached conditions, the SOL heat flux width l_q inferred from target infrared thermography measurements is weakly dependent on poloidal flux expansion but increases approximately with the square root of the divertor leg length. The divertor spreading factor S shows no clear trend with leg length but decreases with flux expansion. TOKAM3X turbulence simulations of the leg length scan are in qualitative agreement with the experiment and can explain observations by a strongly asymmetric (ballooning) transport at and below the X-point. Evidence for increased transport in the region of low poloidal field is obtained in the Snowflake minus geometry. The presence of an additional X-point in the low-field side SOL increases the effective SOL width by approximately a factor two. Increasing flux expansion and leg length both result in enhanced divertor radiation levels, with the effect being much larger in the latter case. This behavior, together with the observed trend in l_q, is consistent with a substantial drop in the density threshold for divertor detachment with increasing leg length and a weak variation with flux expansion. Novel spectroscopic techniques reveal that the drop in target ion current and access to detachment is caused by a reduction of the divertor ionization source due to power starvation, while volume recombination is only a small contributor. This interpretation is confirmed by SOLPS modeling. TCV alternative divertor studies are being extended to neutral beam heated H-mode plasmas. The H-mode power threshold is found to vary weakly between standard, X-, and Super-X geometries. In all cases, ELMy H-mode is obtained at intermediate current, while the discharges are ELM-free at high current. Signs of detachment have so far only been observed in the latter case. Ongoing experiments further investigate H-mode detachment in these plasmas and will be extended to Snowflake configurations.
        Speaker: Prof. Christian Theiler (EPFL-SPC)
      • 145
        Steady states for nonaxisymmetric rotating toroidal plasmas
        Small applied nonaxisymmetric magnetic fields have been demonstrated to have strong and complex effects on otherwise axisymmetric toroidal fusion plasmas. Their importance raises the question of the best ``steady state'' plasma configuration to use for their analysis. A steady state that is valid on fast time scales of a few Alfvén times is needed to invert and interpret experimental measurements and as an initial state to study slower-developing plasma instabilities and plasma processes. It should possess a magnetic flux function Psi with B dot grad(Psi)=0 and a well-confined boundary surface that confines the magnetic field lines. It contains free functions and parameters that must be taken from observations or outside models. The simplest choice is ideal MHD. Axisymmetric and helical MHD plasmas with zero plasma flow possess a good flux function, the plasma pressure, which in axisymmetry is equivalent to the poloidal magnetic flux psi. Axisymmetric states with plasma rotation have two functions, psi and the centrifugally shifted plasma mass density, which represent electron and ion surfaces, respectively. The shifted density modifies the mapping of experimental density to magnetic flux surfaces and allows larger density gradients at the large-R boundary of the torus. Magnetic nonaxisymmetry due to external fields couples the two functions. In single-fluid MHD, the coupling can be shown to impose strong and probably unrealistic constraints on the allowable variation of the rotation and density relative to the magnetic field. Two-fluid models decouple the electron and ion motions and allow greater freedom that removes the restrictions. They also have other properties that reflect experimental observations. The proposed solutions will be studied for experimental cases with rotation and nonaxisymmetry, by numerical simulation with the nonlinear extended MHD code M3D [1], using the real nonaxisymmetric fields. The results will also be compared to the nonlinear evolution. *Work partially supported by U.S. DOE OFES contract DE-SC0007883. [1] Park, W. et al. 1999, Phys. Plasmas 6, 1796; Sugiyama, L.E., and Park, W. 2000, Phys. Plasmas 7, 4644.
        Speaker: Linda Sugiyama (M.I.T.)
      • 146
        Study of Corrosion Properties ITER In-Wall Shield (IWS) Fasteners and Structural Integrity of IWS
        In-Wall Shield(IWS) Blocks will be inserted between inner and outer shell of ITER Vacuum Vessel. These blocks comprise of number of plates of Stainless Steel stacked together using fasteners of XM-19 and M30 size. Plates are tightened with pretension of 107 kN to withstand EM force of 1.83 × 104N during ITER operation. These bolts are spot welded with blocks to lock any type of rotation. There are approximate 1500 such bolts exposed to vacuum in one vessel sector with approximate surface are of 70.5 m2. Hence, surface condition of these fasteners play an important role while leak testing of VV. XM-19 material is very corrosive resistant but, if the fasteners are exposed to normal or humid environment for a long time its surface may get oxidised and catch the corrosion which may further impact the ITER operation in three ways: (a) Reduced Structural Integrity of blocks (b) Gas load due to Outgassing(c) Generation of corrosion products in Cooling Water System. This corrosion has been assessed by (a) Measuring the Corrosion rate (CR) of XM-19 fasteners (exposed in natural environment) and (b) XM-19 washer exposed to water with ITER operating temperature and pressure. This study is carried out using Scanning Electron Microscope (SEM) and Electrochemical Polarization Technique. For SEM analysis, samples were polished and corrosion depth was measured and accordingly CR was calculated. In Electrochemical Polarization Technique, samples were induced with corrosion at room temperature and high temperature in water medium. Pt electrode was used as cathode and Ag-AgCl3 as reference electrode. CR was calculated with the help of corrosion current. Tafel curves of corroded samples show that, reverse polarization path do not intersect the forward path and indicate no tendency of pitting corrosion. Maximum corrosion observed by using Tafel curve is 0.1067 mpy. Outgassing rate of naturally corroded XM-19 bolt was measured 6.06 E-8 Pam3s-1m-2 which is less than the acceptable limit for IWS. Total corrosion product for one Vessel sector was calculated with the help of CR and surface area in one vessel sector and found 3.20 Kg/year. It can be removed by appropriate filters. Study shows that corrosion and out gassing properties of corroded XM-19 fasteners are acceptable for ITER IWS. Detailed experimental set up and results of corrosion study will be presented in this paper.
        Speaker: Ms Abha Maheshwari (ITER-India, Institute for Plasma Research)
      • 147
        Subdivertor fuel isotopic content detection limit for JET and impact on the control of ICRH for JET-ILW and JET-DT operation
        In preparation for JET Deuterium-Tritium Experiments 2 (DTE-2) and to assure readiness to provide fuel cycle-relevant measurements, the subdivertor fuel isotopic ratio detection limit, as determined by Penning optical gas analysis (OGA) [1, 2], was recently researched. Reevaluation of OGA data from DTE-1 [1] revealed a 1% uncertainty (error bar) at the 1% T/(H+D+T) concentration level. A similar detectability limit (at ~1% concentration) was found for H/(H+D) when evaluating a more recent JET ICRH-specific dataset. This analysis also shows a persistent ~ 1% systematic offset of the OGA with respect to divertor spectroscopy values. These studies are in support of substantial diagnostic upgrade for DTE-2 aiming to assure this isotopic detectability, as well mitigate gradual deterioration partly caused by coating of viewport windows by the OGA’s own Penning discharge. The importance of resolving isotopic concentrations at the ~1% level during ICRH plasmas was also explored. When (H/(H+D)) is reduced from 2% to <1%, an increase of core plasma Ti and a decrease of Te are measured [3, 4]. This is consistent with full wave ICRH modelling indicating that when the concentration of the minority species is low enough, 2nd harmonic D absorption becomes dominant over the fundamental H minority absorption; if the plasma density is large enough, it provides collisional bulk ion heating rather than the typical electron heating observed with H minority absorption. The higher background Ti in conjunction with the RF acceleration of the D NBI ions to supra-source energies leads to an increase of the neutron yield by 30% in the case explored. For the same case, the increase of the energy of the fast H tail at small minority concentrations also contributes to sawtooth stabilization. This would imply that the ability to measure, and ultimately control, the fuel isotopic content down <~ 1% concentration level is important for optimizing the performance of a given ICRH scheme in fusion devices. The ability of the OGA technique to act as a global diagnostic of the isotopic mix is of great consequence for ITER, where divertor spectroscopy is unlikely to work, at least for such low concentrations [2].-- [1] D.L. Hillis et al., RSI 70 (1999) 359; [2] C.C. Klepper et al., 2017 JINST 12 C10012; [3] E. Lerche et al., 2016 NF 56 036022; [4] M Goniche et al., 2017 PPCF 59 055001
        Speaker: Dr C Christopher Klepper (Oak Ridge National Laboratory)
      • 148
        Survey on Hot Isostatic Pressing Technique for development of Tokamak components
        Hot Isostatic Press (HIP) equipment is basically an electric furnace which is contained in a pressure vessel. In HIP, the component is subjected to elevated temperature (generally over 1000 degree centigrade) and pressure (generally over 1000 bar) which results in fully isotropic material properties. As per 2012 estimate, approximately 1000 HIP systems have been installed worldwide. Around 50% of these HIP installations were for R&D applications. HIP is used to eliminate pores (and remove casting defects), consolidation of powder and diffusion bonding of dissimilar metals or alloys. The components are often of net shape or near net shape. HIP eliminates inspectibility issues, enables new alloy system and enhances weldability. HIP improves fatigue properties, creep properties, ductility and impact strength. It provides an alternate supply route for long lead time components. Hot Isostatic Pressing of Austenitic Stainless Steel Powders for pressure retaining applications is reported in The American Society of Mechanical Engineers (ASME) proceedings. The technology has developed over the last 20 years and HIP can now produce twice as much product using the same type of machine as they could twenty years ago. The capability of producing full dense near net shape product can be utilized for multilayered plasma facing components fabrication. Joining of various dissimilar materials is possible, such as tungsten to copper joining, Copper to copper alloy, SS to CuCrZr material etc. using HIP. The fabricated joints are reported to be satisfactory. Many fusion components are also fabricated through powder metallurgy route using HIP technique. In this paper, we have performed a survey on applications of HIP in various R&D in fusion community. Some offshore applications, interesting applications in science projects and application for additive manufactured components etc. shall also be discussed.
        Speaker: Mr Gautam Vadolia (Institute for Plasma Research)
      • 149
        The concept of lithium based plasma facing elements for steady state fusion tokamak-reactor and its experimental validation
        The modern results on the implementation of the Russian strategy in the development of designs of long-operating plasma facing element for steady state fusion reactors are considered and analyzed on an example of liquid metal limiters of tokamaks Т-15 and FTU. The experimental validation of this strategy is presented and results on liquid metal CPS behavior in tokamak conditions, effective heat removal up to 12 MW/m2 with low pressure heat transferring medium (0.2 MPa) on the basis of a gas-water spray are considered. The promising scheme of liquid metal divertor target plate for DEMO reactor is presented and discussed.
        Speaker: Mr Alexey Vertkov (SC "Red Star")
      • 150
        The impact of poloidal flux expansion on JET divertor radiation performance

        For a burning plasma device like ITER, radiative power removal by seeded impurities will be inevitable to avoid divertor damage. Increasing divertor radiation by injecting low-Z impurities such as nitrogen, to reduce scrape-off layer heat flux and to cool the divertor plasma to detachment, is put forward as the primary method to achieve this goal. Here, the possibility of increasing the radiative fraction is assessed by using poloidal magnetic flux expansion. Initial ohmic and nitrogen seeded H-mode High Flux Expansion (HFE) experiments, characterized by the presence of 2-nearby poloidal magnetic field nulls and a contracting geometry near the inner target plate have been recently achieved at JET tokamak In this contribution the physics of the dependence of radiative volume and total radiated power on flux expansion variation at JET, equipped with ITER-like Wall (ILW), will be addressed. EDGE2D-EIRENE simulations have already shown that the divertor heat fluxes can be reduced with N2-injection, qualitatively consistent with experimental observations, by adjusting the impurity injection rate to reproduce the measured divertor radiation. Through EDGE2D-EIRENE code modelling, a detailed analysis of the power balance has been set up to physically investigate the reason of the increase of the radiated power for HFE discharges. An increase of charge exchange losses has been related to an increase of connection length and flux expansion both at X-point at strike points position. Spectroscopy data suggests that there is evidence of a detachment front moving towards the X-point from both the movement of the electron density and the low charge nitrogen charge states as the flux expansion increases. Initial experiments with a second null, on the high field side, forming a configuration with significant distance between the two nulls and a contracting geometry near the target plates have been performed leading to an increase of the main magnetic divertor geometry parameters. In addition, nitrogen seeded H-mode experiments have been set-up showing an increase of the total radiated power of the same factor of the flux expansion increase. Further experiments will be devoted to varying the divertor coils polarities to move the secondary x-point on the low field side region and consequently increase the outer flux expansion both in the x-point and strike point region.

        Speaker: Prof. Gianmaria De Tommasi
      • 151
        The influence of displacement damage and helium on deuterium transport and retention in tungsten
        Among many other favorable properties of tungsten its low intrinsic fuel retention makes it a promising candidate as plasma facing material. However, during operation defects in the tungsten lattice will evolve that will trap hydrogen isotopes. While for present day devices this increased retention is only limited to the near surface it will take place throughout the whole bulk in future nuclear devices as a consequence of the neutron irradiation. There is not yet a microscopic understanding that would allow to describe the processes that will prevail in a reactor environment quantitatively, where damage creation and hydrogen retention will mutually influence each other. Present day predictions are only based on extrapolation of data collected from non-nuclear machines. Hence, the influence of hydrogen on defect production, the influence of defects on hydrogen isotopes transport as well the influence of the presence of helium (directly implanted close to the material surface from the plasma as well as produced throughout the bulk by tritium decay and transmutation) is not considered in these extrapolations. Implantation of different ion species with energies in the MeV range can be used to simulate the displacement damage neutrons will cause. Contrary to neutron irradiation, ion beam irradiation is fast and does not activate the samples. Likewise, the influence of He on transport and retention can be studied by implanting He with MeV-energy deep into the material. In this way surface effects can be separated from bulk effects. These kind of experiments allow dedicated parameter studies under well controlled conditions. In this contribution such benchmark experiments on transport and retention of deuterium in displacement damaged and helium containing tungsten are presented that allow to test and extend present day understanding on a quantitative level. The dependence of deuterium retention on the damage level, the influence of damage rate as well as the influence of the specific ion used to create the displacement damage will be shown. Results from hydrogen isotope exchange experiments are presented that reveal the dynamics of hydrogen transport which is a chain of trapping and detrapping processes. Rate equation modelling without free parameters is used to describe the observed uptake during plasma exposure as well as the release during annealing.
        Speaker: Dr Thomas Schwarz-Selinger (Max-Planck-Institut für Plasmaphysik)
      • 152
        The influence of Fe-ion irradiation on the microstructure of reduced activation ferritic-martensitic steel Eurofer 97
        The reduced-activation ferritic-martensitic steel Eurofer 97 is the European benchmark structural material for in-vessel components of fusion reactor. Experimental data on neutron irradiated Eurofer 97 material have shown decrease in plasticity and radiation hardening at irradiation temperatures about 300 °C. Formation of dislocation loops and α' pre-precipitates is considered as the main reason of this phenomenon. In this work Eurofer 97 steel was irradiated with Fe ions up to 10^16 ions/cm^2 at 250, 300 and 400 °C. The irradiated samples were characterized by TEM and APT. TEM study of ion irradiated samples revealed nucleation of dislocation loops. The pair-correlation analysis of APT data detected an initial stage of solid solution decomposition. The hardening of ion irradiated Eurofer 97 was calculated with DBH model taking into account radiation-induced dislocation loops to comparison with the change of yield stress for neutron irradiated Eurofer 97. According to obtained results it can be supposed that the formation of dislocation loops plays the main role in the low temperature radiation hardening of Eurofer 97 at the dose level up to ~10 dpa.
        Speaker: Dr Sergey Rogozhkin (SSC RF ITEP of NRC “Kurchatov Institute”)
      • 153
        The Operation, Control, Data Acquisition System of ASDEX Pressure Gauge for Neutral Pressure
        The Bayard-Alpert (BA) type hot cathode ionisation gauge is widely used to measure neutral pressure precisely in vacuum system below 10-3 Torr pressure. Neutral pressure measurement in magnetic confinement fusion experiments is quite challenging for standard BA type gauge due to higher pressure limitation and its ionisation is affected by high magnetic and electrical fields. To overcome above limitations, A special hot cathode ionisation gauge, named ASDEX Pressure Gauge (APG) system has been developed by G. Haas at the Max-Planck-Institute, Germany \cite{Haas}. The APG system works in high magnetic field upto 6 Tesla and high temperature plasma environment with broad pressure measurement range from 10-1 to 10-6 mbar with fast response (<10 msec) and good noise immunity. For ADITYA Tokamak, A customised system of operation, control and data acquisition for standard APG system has been designed and developed to measure real time neutral pressure during high temperature plasma discharges. The developed system can achieve synchronous control of gauge controller using GPIB and data acquisition of ion and emission current of gauge head using PCI based data acquisition module. Initially, the APG calibration with standard BA type hot cathode ionization gauge had been carried out on the test setup of low magnetic field and ultra high vacuum system with different gases like H2, Ar, He. For APG calibration in various pressure range of different gases, precise gas feed control system has been developed using controller based hardware and LabVIEW software. After successfully testing and calibration, the APG was installed on ADITYA tokamak and calibrated under high magnetic field of ADITYA Tokamak. The developed APG control system has been configured to set the gauge parameter before the plasma discharge and acquired real time analog signal acquisition using simultaneous sampling by analog to digital convertor (ADC) during plasma discharge. The acquired raw data and processed real time pressure measurement gives valuable neutral density information to tokamak plasma.
        Speaker: Mr Kiran PATEL (Institute for Plasma Research)
      • 154
        The Software and Hardware Architecture of the Real-Time Protection of In-Vessel Components in JET-ILW
        The JET ITER-like wall (JET-ILW) combines plasma-facing components (PFC) made of bulk beryllium for main chamber limiter tiles and of bulk tungsten as well as tungsten coated CFC tiles for divertor tiles. The risk of damaging the metallic PFCs caused by beryllium melting or cracking of tungsten owing to thermal fatigue required a new reliable active protection system. To address this issue, a real time protection system comprising newly installed imaging diagnostics, real time algorithms for hot spot detection and alarm handling strategy has been integrated into the JET protection system. This contribution describes the design, implementation, and operation of the near infrared (NIR) imaging diagnostic system of the JET-ILW plasma experiment and its integration into the existing JET protection architecture. The imaging system comprises four wide-angle views, four tangential divertor views, and two top views of the divertor. Regions of interest (ROI) on the selected PFCs of different materials are analysed in real time and the maximum temperature measured in each ROI is sent to a real time algorithm called vessel thermal map (VTM) to determine the likely cause of the overheating and to request an appropriate response from the plasma control system. Post-pulse data visualization and advance analysis of all types of imaging data is provided by the new software framework JUVIL (JET Users Video Imaging Library). The hot spots formation at the re-ionization zones due to impact of the re-ionised neutrals as well as due to RF-induced fast ion losses is recognized as a big threat due to quick surface temperature rise. Because it could trigger the protection system to stop a pulse, it is important to identify the mechanisms and conditions responsible for the formation of such hot spots. To address this issue a new software tool Hotspot Editor has been developed. Future development of the JET real time first wall protection is focused on the D-T campaign and the ITER relevant conditions which will cause failure of camera electronics within the Torus hall. To provide the reliable wall protection, two more sensitive logarithmic NIR camera systems equipped with new optical relays to take images and cameras outside of the biological shield have been installed on JET-ILW and calibrated with in-vessel calibration light source.
        Speaker: Mrs Valentina Huber (Forschungszentrum Jülich GmbH)
      • 155
        Thermal Analysis of Protection Important Components of ITER XRCS-Survey Diagnostic System

        In the ITER, an important aspect of qualifying the components to the mandatory regulatory requirements, the system developers have a challenge to first design the components fulfilling guidelines of the ITER recommended French nuclear code RCC-MR (2007) and later on demonstrate to the regulator. It is even more involving for systems that are extending primary vacuum to the interspace and port-cell as these zones are accessible by a human. The paper addresses such requirements in the thermal design of the X-Ray Crystal Spectroscopy-Survey (XRCS-Survey) system, which is a first plasma diagnostic.
        The XRCS-Survey is a broadband (1 - 100 Å) X-ray crystal spectrometer for real-time monitoring of absolute concertation and in-flux of the plasma impurities. For measurements, the transport of x-ray emission is done using a nearly 10m long sight-tube directly connecting the spectrometer to the closure plate of the port-plug. The sight-tube components, classified as Protection Important Components due to their function in confinement of radioactive tritium and dust, are subjected to various thermal loads while machine operations. These loads are mostly due to baking to achieve ultra-high vacuum inside the ITER vacuum vessel. Furthermore, the components are also subjected to a neutron, gamma radiation of D-T fusion.
        For reliable performance and safe operation of XRCS-Survey diagnostic, a preliminary engineering design and ANSYS analysis of the XRCS-Survey sight-tube components have been performed, with and without radiation shielding, in order to analyze the behavior of components under baking heat loads, operational heat loads and also accidental fire heat loads.
        The paper presents an optimized design layout for the sight-tube of XRCS Survey and results of the thermal analysis; defining temperature limits to observe compliance with safety criterion defined by ITER regulatory guidelines on PIC (class SIC-1) components as well as providing inputs to the structural integrity analysis of the system.

        Speaker: Prof. Sanjeev Varshney (ITER-India, IPR)
      • 156
        Thermal-hydraulics and Structural analyses of LLCB TBM set
        India is developing Lead-Lithium cooled Ceramic Breeder (LLCB) Test Blanket Module (TBM) for testing in ITER for the validation of fusion blanket design tools, tritium breeding performance and high grade heat extraction capability relevant to Indian DEMO. The LLCBTBM will be tested from the first phase of ITER operation (H–H phase) in one half of the ITER port no-2. LLCB TBM set consists of TBM and its shield along with supports and piping.The LLCB TBM consists of U shaped helium cooled first wall (FW) with back plate enclosing internal components covered by top and bottom plates. The TBM internals consist of four ceramic breeder canisters(Li2TiO3) in the form of pebble bedwith Pb-Li flowing around these canisters to cool the internal structure.The TBM is supported at TBM shield by supports. The back side of Shield is welded to TBM set flange, which is bolted to the ITER port plug frame.TBM Shield made of SS 316 L (N)-IG located behind the TBM is composed of steel and water with a combination of 50:50 to shield neutrons. It consists of two symmetrical parts that have grooves to accommodate pipes. The neutronic heat generated inside shield structure is extracted by water flowing inside the shield. The detailed thermal –hydraulics of TBM set has been performed based on the heat flux on FW and neutronic heat generation on TBM set. The temperature distribution obtained from thermal analysis has been used for thermo-structural analysis. CFD analysis of helium flow inside the FW channels and manifolds has been carried out to estimate temperature, pressure drop and heat transfer coefficient. The distribution of flow inside the different flow circuits of FW from manifolds and water flow in TBM Shield will also be described in this paper. Structural analysis has been performed on TBM set based on load combinations as per ITER load specifications. RCC-MR 2007 code has been used for the structural assessment for the prevention of p-type and s-type damages and calculation of safety margins. The structural analysis results of different components of TBM set which include TBM, back plate, supports,process pipes and TBM shield will be discussed in detail in this paper.
        Speaker: Mr Deepak Sharma (Institute For Plasma Research)
      • 157
        Thermo-Mechanical Experiments On Lithium Titanate Pebble Bed
        Among the various lithium ceramics, Li2TiO3 is one which has been received much attention due to its very excellent properties, such as reasonable lithium atom density, low activation, excellent tritium release performance and chemical stability, etc. Lithium Titanate [Li2TiO3] pebbles with the diameter of 1mm was widely used for the experiments after successful completion of variety of modeling and experiments. In the present study we have prepared lithium titanate from its high pure raw material of lithium carbonate and titanium dioxide by solid state reaction in the stoichiometric ratio. The reaction temperature has been estimated from the thermo-gravimetric and differential thermal analysis (TG-DTA) and the same scenario has been executed for the bulk production using high temperature furnace. The phase and phase stability at different temperature were analyzed by using powder X-ray diffractometer. The pebble preparation has been carried out from this raw material after once again ground them to fine powder and addition of PVA as a binder for the preparation of green pebbles using extruder-spheronizer technique. The green pebbles were sintered at high temperature to attain desired density for further studies. The details of the Li2TiO3 powder and pebble fabrication and their characterizations like XRD, density, porosity, crush load, SEM analysis, Young’s Modulus and creep will be discussed in the paper.
        Speaker: Dr Riscob Bright (Institute for Plasma Research)
      • 158
        TRIGA Integral Activation of Mn foils, Li2O and LiF as Potential Tritium Production Monitors for Fusion Applications
        In the future fusion reactors, such as ITER or DEMO, tritium will be produced by bombardment of lithium atoms with neutrons and several types of special Tritium Breeder Modules (TBM) will be installed in the ITER reactor to demonstrate the self-sufficiency of tritium production. LiF pellets commercially produced as Thermo-luminescent detectors (LiF - TLDs) can be used to measure tritium production. The similarities between the sensitivity profiles of the neutron reaction of tritium production in 6Li(n,t) and those of the 55Mn(n,g)56Mn reaction in the TBMs indicated that the latter reaction could be used as a tritium production monitor, at least for short-term monitoring, the half-life of 56Mn being 2.579 h. However, experimental verification and improvements and validation of the Mn cross-sections are needed in order to meet the required accuracy. In the scope of the Fusion for Energy (F4E) project of the European Commission, foils of certified reference materials Al-1%Mn and Al-0.1%Au, as well as TLD(LiF) and Li2O samples were irradiated in the JSI TRIGA research reactor, both bare, and under Cd and boron-nitride to study the potential use of Manganese detectors for monitoring the tritium production in fusion machines. In order to obtain complementary information for data validation purposes, the irradiations were performed in different neutron spectra, i.e. in the Central Channel, the Pneumatic Tube in position F24 in the outer “F” ring of the reactor core, in position F19 and in the IC-40 irradiation channel in the graphite reflector. Bare, and under Cd and boron-nitride irradiations were needed for the subtraction of epi-thermal neutron contribution in the 55Mn(n,g)56Mn reaction. Two series of measurements was performed, in 2014 and 2017. The transport calculations were performed using the Monte Carlo transport code MCNP6.1 with a detailed model of the TRIGA reactor including the irradiation capsules. The uncertainties involved in the measurements and the calculations were carefully evaluated. The principle objective was to study the energy response of the 55Mn(n,)56Mn reaction and correlations between the Mn and TLD / Li2O measurements. Good consistency between the measured and calculated reaction rates, in most cases within the uncertainty bars, was observed and will be reported in the paper.
        Speaker: Prof. Ivan Kodeli (Jozef Stefan Institute)
      • 159
        Turbulence and radial electric field asymmetries measured at TJ-II plasmas
        Dedicated experiments have been carried out for a systematic comparison of turbulence and radial electric field measured at poloidally separated positions in the same flux-surface in the stellarator TJ-II. The rationale behind this study is twofold, verification of the spatial localization of instabilities predicted by the Gyrokinetic simulations in stellarators and verification of the electrostatic potential variation on the flux surface as calculated by Neoclassical codes and its possible impact on the radial electric field. Poloidal asymmetries in the turbulence wavenumber spectrum and in the Er profile have been found that depend on density, heating conditions and magnetic configuration. These quantities have been measured using a Doppler reflectometer that covers the radial region from rho = 0.6 to 0.9, at different perpendicular wave-numbers of the turbulence: 1-14 cm-1, and at two plasma regions poloidally separated. Different plasma scenarios have been studied with different profile shapes. These include, high power on-axis ECH heated plasmas vs. low power off-axis ECH heated plasmas; ECH vs. NBI heated plasmas; standard vs. high iota magnetic configurations, and Hydrogen vs. Deuterium dominated plasmas. Differences in the turbulence intensity are found when comparing the k-spectra measured at poloidally separated positions in the same flux-surface, in ECH heated plasmas in the standard magnetic configuration. However, almost no asymmetries are found in NBI heated plasmas, i.e. higher density, lower electron temperature, where very similar turbulence intensity and spectral shape are measured at both plasma regions. Besides, no significant differences have been found when comparing Hydrogen and Deuterium dominated plasmas. The asymmetry in the turbulence intensity found in the standard magnetic configuration reverses in the magnetic configuration with high rotation transform. Radial electric field profiles measured at the two plasma regions show pronounced differences in low density plasmas, i.e. plasmas in neoclassical electron root confinement. At higher densities the Er asymmetry gradually decreases and almost disappears in ion root plasmas. The detailed comparison of the k-spectra and Er profiles under different plasma scenarios are presented, providing valuable information for comparison with Gyrokinetic and Neoclassical simulations.
        Speaker: Dr Teresa Estrada (CIEMAT)
      • 160
        Validation of global gyrokinetic simulations in stellarator configurations
        In this contribution, recent simulations carried out in stellarator configurations with the global gyrokinetic code EUTERPE and the ongoing validation effort are presented. The linear relaxation of zonal flows (ZFs) has been studied in global simulations in many stellarator configurations. The code has been verified by comparing results with both other gyrokinetic codes and analytical estimations. Furthermore, these calculations were validated against experimental measurements obtained during pellet injection experiments in TJ-II. The oscillatory relaxation of potential measured by the HIBP was compared to simulations including impurity ions, with a good quantitative agreement in the frequency and damping rates. This is the first experimental confirmation of the ZF oscillation in stellarators, accurately described by the linearized gyrokinetic equation. The electrostatic micro-instabilities have been studied numerically in the stellarators TJ-II and W7-X and an effort to validate simulations against experimental turbulence measurements from Doppler Reflectometry (DR) has been done. The model validation has been pursued at different levels of detail, including the density fluctuation level, frequency spectra, and the localization of instabilities along the flux surface. In dedicated experiments in TJ-II, the power and deposition location of the ECRH heating were changed, thus modifying significantly the density and temperature radial profiles. The experimental measurements from the DR system in TJ-II have been compared to simulations. The relevant wave-numbers and the radial variation of experimental density fluctuations spectra are consistent with the range of unstable wave-numbers and the location of maximum instability found in simulations. No dependency of the power spectra with the bulk ion mass is observed experimentally, which is consistent with the kind of unstable modes (electron-driven) found in simulations. A systematic difference is found between the density fluctuation spectra measured by the DR system at poloidally separated positions on the same flux-surface, which is largely affected by the rotational transform. The localization of instabilities in simulations is also influenced by the rotational transform, however, a discrepancy between the location of maximum fluctuation level in simulations and experiments is found so far.
        Speaker: Dr Edilberto Sánchez (Laboratorio Nacional de Fusión, CIEMAT)
      • 161
        Verification tests for remote participation at ITER REC
        The ITER Remote Experimentation Centre (REC, [1]) in Rokkasho is one of the projects implemented within the Broader Approach (BA) agreement [2] as part International Fusion Energy Research Centre (IFERC). The long-term objective of the REC is to allow researchers to take part in the experimentation on ITER from a remote location. On a shorter time scale, before ITER will be operated, the REC facility will be used to test technologies for remote participation in collaboration with existing European tokamaks [3] and with JT-60SA, whose first operations are envisaged in 2020 [4]. Other than setting up and equipping the REC control room (see also [5]), during the first phase of the REC (2013-2017) the scope of the project included activities aimed at developing and evaluating software tools for fast data transfer, remote participation, data analysis, and plasma simulation. These activities have been carried out by an Extended-Integrated Project Team (E-IPT). Indeed, due to the characteristics of the remote experiment, the collaborations with experts in other institutes providing experimental data, network infrastructure, data transfer protocol, and experiences on inter-continental data transfer and data acquisition were essential for the success of the REC activities. As a consequence, members from different Japanese and European institutions were invited to join in E-IPT; among the various contributors, there are members of the ITER project and of the Satellite Tokamak Programme (STP), members of the National Institute of Informatics (NII) and the National Institute for Fusion Science (NIFS) in Japan, and members of JET and WEST in Europe. In this paper, we report on the results of the REC verification tests that have been carried out in 2017. These tests were mainly aimed at assessing the functionalities of the REC control room (i.e., the configurability of the room layout, the capabilities of the video wall, etc.), as well as the functionalities of the software tools for remote participation that have been developed during the first BA period. A report on the preliminary remote participation tests carried out in collaboration with the JET tokamak will be also given, together with a description of the tests with both JET and WEST that have been planned for 2018.
        Speaker: Dr Susana Clement Lorenzo (Fusion for Energy)
      • 162
        Visual Servo of Tokamak Relevant Remote Handling Systems using Neural Network Architecture
        Tokamak inspection and maintenance requires different Remote Handling (RH) systems such as long reach planar manipulators, multi-DOF hyper-redundant arms etc. As no structural support can be provided inside the tokamak, these RH systems are usually cantilevered and have a number of articulations to traverse the toroidal geometry of the tokamak. The kinematic configuration is thus different for conventional manipulators. Due to long cantilevered length, heavy payload handling, structural deformations, gearbox backlash and control system inaccuracies the final pose of the end effector may vary from the desired pose when only a servo feedback loop is used. Such inaccuracies can only be eliminated by using Visual Servo (VS) technique, where the inverse kinematics and trajectory planning are done based on visual feedback from cameras mounted on the RH system. The paper gives a fresh approach to visual servo for tokamak RH systems using artificial neural networks (NN) architecture. A multi-layered feed-forward NN is trained using the joint angle vector as input and the corresponding feature vector(s) of markers in a sample tile as output. The trained NN can thus predict the joint configurations for given features vectors. This eliminates the requirement of closed-form inverse kinematic solution of the manipulator and camera calibration. The NN architecture and proposed controller are validated and presented using simulation on 5DOF remote handling manipulator. Real time implementation methodology for NN based controller are also discussed.
        Speaker: Mr Pramit Dutta (Institute for Plasma Research)
    • 10:15 AM
      Coffee Break
    • EX/1-TH/1 Energetic Particles
      • 163
        Strongly non-linear energetic particle dynamics in ASDEX Upgrade scenarios with core impurity accumulation
        In 2017 a new scenario on ASDEX Upgrade for the dedicated investigation of energetic particle (EP) physics has been developed. This scenario is unique in two aspects: firstly, the neutral beam (NB) induced fast-ion beta is comparable to the background plasma β, and secondly, the ratio of the fast ion energy to the thermal background is of the order 100. At ASDEX Upgrade we reach this previously unexplored regime by NB off-axis heating only and by letting impurities accumulate in the core. Due to strong radiation losses the background tempera- tures and pressures of both ions and electrons stay low, despite 2.5 − 5MW NB heating. In the stable flattopphaseanunprecedentednumberofvariousEP-driveninstabilities(despitevEP/vAlfvén ≈0.4≪ 1) is simultaneously observed: EP-driven geodesic acoustic modes (EGAMs), beta-induced Alfvén eigenmodes (BAEs), reversed shear Alfvén eigenmodes (RSAEs) and toroidal Alfvén eigenmodes (TAEs), that are modulated by transient q = 2 sawtooth-like crashes, NTMs and ELMs. The physics reasons for these strong mode activity are discussed. During the stable flat-top phase meaningful EP distribution function measurements (FIDA) and analysis (TRANSP/FIDASIM) can be performed. First results indicate that the EP profiles differ significantly from neoclassical predictions. Bicoherence analysis using an advanced toolset for non-stationary processes reveals that non-linear coupling processes between different frequency bands exist. In addition, TAE bursts are observed to trigger the onset of EGAMs which indicates coupling of these modes via the velocity space (EP avalanches). Linear and non-linear tools (HAGIS/LIGKA, ORB5, XHMGC) are used for mod- eling mode onset and non-linear phases. These experiments facilitate the experimental study of the interaction of AEs, zonal modes and turbulence and thus serve as an ideal validation opportu- nity for various non-linear analytical and numerical models. In addition, the observed onset of EP avalanches can be quantified. The investigation and understanding of these - so far not accessible - physics elements is a prerequisite for a reliable prediction of the self-organisation of a burning plasma.
        Speaker: Dr Philipp Lauber (IPP Garching)
      • 164
        Simulations of energetic particle driven instabilities and fast particle redistribution in EAST tokamak
        Instabilities driven by energetic particles including fishbones and Alfven eigenmodes, together with fast particle loss and redistribution due to resonant magnetic perturbations (RMPs), have been investigated numerically with codes M3D-K, MEGA, and GYCAVA in EAST tokamak. Firstly, hybrid simulations with the global kinetic-magnetohydrodynamic (MHD) code M3D-K have been carried out to investigate the beam-driven fishbone in EAST experiment. The results are consistent with the experimental measurement with respect to mode frequency and mode structure. Nonlinear simulations show that the frequency of the fishbone chirps up and down with corresponding hole-clump structures in phase space, consistent with the Berk-Breizman theory. In addition to the low frequency fishbone, a high frequency beta-induced Alfven eigenmode (BAE) is excited during the nonlinear evolution. Secondly, two kinetic-MHD codes, namely MEGA and M3D-K, have been applied to study fast ion driven toroidal Alfven eigenmodes (TAEs) in EAST tokamak. Parameter scans show that the frequency and growth rate of TAEs simulated by the two codes agree well with each other. The analysis of the resonant interaction between the TAE and fast ions shows that the TAE exchanges energy with the co-current passing particles with parallel velocity $|v_∥ |≈V_{A0}/3$ or $|v_∥ |≈V_{A0}/5$, where $V_{A0}$ is the Alfven speed on the magnetic axis. Moreover, the TAE destabilized by the counter-current passing ions has much smaller growth rate than that driven by the co-current ion. Thirdly, the effects of RMPs on the loss and redistribution of passing ions are investigated numerically by the orbit following code GYCAVA for EAST tokamak. The loss fraction and the loss region of passing ions increase with the amplitude of RMPs. For the energetic passing ions, the extra loss induced by RMPs can be comparable to the loss induced by the magnetic drift. The extra loss of passing ions induced by RMPs is related to the drift island structure induced by RMPs and the magnetic drift, and the stochasticity induced by overlap of magnetic islands. The dependence of the loss fraction and loss region on the toroidal mode number of RMPs is related to the safety factor. Finally, the pitch angle and energy of particle can impact the loss of energetic passing ions. These results would provide guidance for future EAST experiments.
        Speaker: Dr Wei Shen (Institute of Plasma Physics, Chinese Academy of Sciences)
      • 165
        Reduced energetic particle transport models enable comprehensive time-dependent tokamak simulations
        The inclusion of the reduced-physics energetic particle (EP) *kick model* for EP transport in TRANSP has resulted in a dramatic improvement of interpretive and predictive capabilities for time-dependent tokamak simulations including the effects of EP transport by instabilities. The kick model has recovered the measured toroidal Alfvén eigenmode (TAE) spectrum on NSTX-U and has reproduced details of the fast ion diagnostic data measured on DIII-D for EP modes and tearing modes. Being able to predict the occurrence and effect of those instabilities is one of the grand challenges for fusion and a necessary step to mitigate their negative effects. The kick model has proven the potential of phase-space resolved EP simulations to unravel details of EP transport for detailed theory/experiment comparison and for scenario planning based on optimization of NBI parameters. Work is also ongoing to complement the kick model approach with the RBQ1D model based on the resonance-broadening quasi-linear theory to develop a self-consistent, numerically efficient predictive EP transport model. On NSTX-U, the kick model successfully reproduces the stability of co- and counter-propagating TAEs driven unstable by NB injection. The model successfully reproduces the transition from a co-TAE dominated scenario to one with coexhisting co- and counter-TAEs. Based on the analysis, strategies for mitigating the instabilities are developed through TRANSP by varying the NB injection parameters. The phase space resolution implemented in the model is also crucial for its successful validation against fast ion diagnostics data from Fast Ion D-Alpha (FIDA) and neutral particle analyzers (NPA). For DIII-D discharges with strong Alfvénic activity, the amplitude of the instabilities used in the simulation is first adjusted to match the measured neutron rate. The inferred FIDA and NPA signals based on the simulation are then compared with the experimental data for validation, showing excellent agreement. Initial analysis via the RBQ1D model gives similar results, indicating its potential for predictive simulations. Enhancements to TRANSP via the inclusion of reduced EP transport models are playing an important role in scenario development including realistic treatment of fast ion transport by instabilities, e.g. to optimize the scenario by tailoring NB injection power and voltage.
        Speaker: Dr Mario Podesta (Princeton Plasma Physics Laboratory)
      • 166
        Critical fast ion distribution in phase space for the synchronized sudden growth of multiple Alfvén eigenmodes and the global transport of fast ions
        Alfvén eigenmodes (AEs) driven by fast ions in tokamak plasmas and the fast ion distribution formed with the AEs, neutral beam injection (NBI), and collisions are investigated with hybrid simulations for energetic particles and a magnetohydrodynamic (MHD) fluid [1]. The multi-phase simulation [2], which is a combination of classical simulation and hybrid simulation, was applied for various beam deposition power (P_NBI) and slowing-down time (t_s). In the classical simulation, energetic particle orbits are followed in the equilibrium magnetic field with NBI and collisions while the MHD perturbations are turned off. The physical parameters other than P_NBI and t_s are similar to those of a TFTR experiment [3]. For P_NBI=10MW and t_s=100ms, which are similar to the TFTR experiment, the AE bursts take place with a time interval 2.7ms and the maximum amplitude of radial MHD velocity normalized by the Alfvén velocity vr/vA=3x10^-3, which are close to the TFTR experiment. With increasing volume-averaged classical fast ion pressure, the fast ion confinement degrades monotonically due to the transport by the AEs. The fast ion pressure profile resiliency, where the increase in fast ion pressure profile is saturated, is found for the cases with the AE bursts. In this work, we have clarified the physical process of the AE burst in toroidal plasmas. Before the AE bursts occur, multiple AEs become unstable, and grow to low amplitude. The low-amplitude AEs gradually and locally flatten the fast ion distribution in phase space leading to the formation of a stepwise distribution. The stepwise distribution is a “critical distribution” where the further beam injection leads to the higher AE amplitude, the broadening of the locally flattened regions, and their overlap. This resonance overlap of the multiple AEs [4] brings about the AE burst, the global transport of fast ions, and the saturation of the distribution. [1] Y. Todo, New J. Phys. 18 (2016) 115005. [2] Y. Todo et al., Nucl. Fusion 54 (2014) 104012. [3] K. L. Wong et al., Phys. Rev. Lett. 66 (1991) 1874. [4] H. L. Berk, B. N. Breizman, and M. Pekker, Phys. Plasmas 2 (1995) 3007.
        Speaker: Dr Yasushi Todo (National Institute for Fusion Science)
      • 167
        Impact of ECH/ECCD on Fast-ion-driven MHD Instabilities in Helical Plasmas & Excitation mechanism of the energetic particle driven resistive interchange mode and strategy to control the mode in Large Helical Device
        A. We discuss the effect of electron cyclotron heating (ECH) and current drive (ECCD) on fast particle (FP)-driven MHD instabilities in stellarator/heliotron (S/H) plasmas obtained in LHD, Heliotron J and TJ-II. We demonstrate that FP-driven MHD instabilities including energetic particle modes (EPMs) and Alfvén eigenmodes (AEs) can be controlled by means of magnetic shear s modified by EC-driven plasma current. EPMs can be controlled by changing continuum damping rate, which is the main damping mechanism of the EPM and depends on s. AEs are significantly affected by the change of structure of the shear Alfvén continuum which can be modified by s. We also find that ECH (non-ECCD) can impact FP-driven MHD instabilities. Candidates to explain the ECH effect on FP-driven MHD instabilities are the variation in the fast ion profile and/or the trapped electron collisional damping. B. The helically-trapped energetic-particle (EP) driven resistive interchange mode (EIC) observed in the Large Helical Device (LHD) causes large amount loss of EPs. It is destabilized when the precession motion of the helically trapped EP resonates with the pressure driven mode. A velocity modulation caused by the toroidicity of the magnetic field produces this resonance. Strategy and the initial results to suppress the EIC mode based on the knowledge of the EP orbit effects, by the ECH heating and by the RMP application, are presented. EPs having perpendicular velocity components are trapped in the weak magnetic field region of the LHD and making precession motion helically. The rotation frequency of this precession motion is slow enough to interact with the pressure driven MHD modes. If the energy transfer from the EP to the mode is estimated by evaluating the correlation of the fluctuating component of the precession motion and the MHD mode, a resonance is found when the MHD mode rotates poloidally -1.2 times of the poloidal component of the heliccally trapped EP motion. This resonance disucssed here is consistent with the following observations found in the hydrogen / deuterium experimental campaign. 1) MHD mode rotates in the electron diamagnetic drift direction while the EP moves in the ion diamagnetic drift direction. 2) The mode frequency is almost the same with the precession frequence of the initial velocity of the NB-injected EPs. The EIC modes are succesufully suppressed by the ECH injection and RMP application. The physical mechnism of the stabilization will be discussed.
        Speaker: Dr Satoshi Yamamoto (Institute of Advanced Energy, Kyoto University)
        oral presentation
        Summary Slides
      • 168
        Excitation mechanism of the energetic particle driven resistive interchange mode and strategy to control the mode in Large Helical Device
        The helically-trapped energetic-particle (EP) driven resistive interchange mode (EIC) observed in the Large Helical Device (LHD) causes large amount loss of EPs. It is destabilized when the precession motion of the helically trapped EP resonates with the pressure driven mode. A velocity modulation caused by the toroidicity of the magnetic field produces this resonance. Strategy and the initial results to suppress the EIC mode based on the knowledge of the EP orbit effects, by the ECH heating and by the RMP application, are presented. EPs having perpendicular velocity components are trapped in the weak magnetic field region of the LHD and making precession motion helically. The rotation frequency of this precession motion is slow enough to interact with the pressure driven MHD modes. If the energy transfer from the EP to the mode is estimated by evaluating the correlation of the fluctuating component of the precession motion and the MHD mode, a resonance is found when the MHD mode rotates poloidally -1.2 times of the poloidal component of the heliccally trapped EP motion. This resonance disucssed here is consistent with the following observations found in the hydrogen / deuterium experimental campaign. 1) MHD mode rotates in the electron diamagnetic drift direction while the EP moves in the ion diamagnetic drift direction. 2) The mode frequency is almost the same with the precession frequence of the initial velocity of the NB-injected EPs. The EIC modes are succesufully suppressed by the ECH injection and RMP application. The physical mechnism of the stabilization will be discussed.
        Speaker: Dr Satoshi Yamamoto (Institute of Advanced Energy, Kyoto University)
    • 12:30 PM
    • EX/1-TH/1 P2 Posters
      • 169
        Critical fast ion distribution in phase space for the synchronized sudden growth of multiple Alfvén eigenmodes and the global transport of fast ions
        Speaker: Dr Yasushi Todo (National Institute for Fusion Science)
      • 170
        Excitation mechanism of the energetic particle driven resistive interchange mode and strategy to control the mode in Large Helical Device
        Speaker: Dr Satoshi Ohdachi (National Institute for Fusion Science)
      • 171
        Impact of ECH/ECCD on Fast-ion-driven MHD Instabilities in Helical Plasmas
        Speaker: Dr Satoshi Yamamoto (Institute of Advanced Energy, Kyoto University)
      • 172
        Reduced energetic particle transport models enable comprehensive time-dependent tokamak simulations
        Speaker: Dr Mario Podesta (Princeton Plasma Physics Laboratory)
      • 173
        Simulations of energetic particle driven instabilities and fast particle redistribution in EAST tokamak
        Speaker: Dr Wei Shen (Institute of Plasma Physics, Chinese Academy of Sciences)
      • 174
        Strongly non-linear energetic particle dynamics in ASDEX Upgrade scenarios with core impurity accumulation
        Speaker: Dr Philipp Lauber (IPP Garching)
    • OV/4 Overview Magnetic Fusion
      • 175
        Overview of first Wendelstein 7-X high-performance operation with island divertor
        The optimized superconducting stellarator device Wendelstein 7-X restarted operation after the assembly of a graphite heat shield and an inertially cooled island divertor. This paper reports on results from the first high-performance plasma operation. Plasma densities of $1-4\cdot 10^{19}\,\mathrm{m}^{-3}$ with electron temperature $5-10\,\mathrm{keV}$ were routinely achieved with hydrogen gas fuelling, eventually terminated by a radiative collapse. Up to $1.4\cdot 10^{20}\,\mathrm{m}^{-3}$ plasma density was reached with repetitive hydrogen pellet injection. Here, the ions are indirectly heated, and at a density of $8\cdot 10^{19}\,\mathrm{m}^{-3}$ temperatures $T_e\simeq T_i = 3.4\,\mathrm{keV}$ were accomplished, which corresponds to $nT\tau_E = 6.4\cdot 10^{19}\,\mathrm{keVs/m}$ with peak diamagnetic energy $1.1\,\mathrm{MJ}$. Stable $25\,\mathrm{s}$ long-pulse helium discharges with $2-3\,\mathrm{M}$W ECRH power and up to $75\,\mathrm{MJ}$ injected energy were created routinely for equilibrium and divertor load studies, with plasma densities around $5\cdot 10^{19}\,\mathrm{m}^{-3}$ and $5\,\mathrm{keV}$ electron temperature. The divertor heat loads remained far below the limits. The O/C impurity concentration ratio has decreased in comparison to the previous limiter operation and no intrinsic impurity accumulation along with high edge radiation were observed in stationary plasmas. During pellet-fuelled hydrogen discharges, full detachment was observed with divertor target heat flux reduction by more than $\times 10$. Both X2 and O2 mode ECRH schemes were applied and electron cyclotron current drive (ECCD) experiments were conducted. During co-ECCD injection experiments with axial currents up to $13\,\mathrm{kA}$, frequent fast crashes were observed mainly in the core electron temperature, suggesting a fast magnetic reconnection mechanism. The radial electric field measured with (Doppler) and correlation reflectometry changes sign at the plasma edge from $+10\ldots+20\,\mathrm{kV/m}$ to $-10\ldots-5\,\mathrm{kV/m}$, fairly independent of discharge parameters and heating power. Edge and scrape-off layer turbulence was measured with both Langmuir probes and reflectometer diagnostics. Core turbulence was measured with a phase contrast imaging diagnostic and different levels of broad band turbulence as well as coherent Alfvén mode activity were observed.
        Speaker: Prof. Thomas Klinger (Max-Planck Institute for Plasma Physics)
      • 176
        Overview of the First Deuterium Experiment in LHD
        In the first deuterium (D) experiment, LHD established one of the most important milestones towards the realization of the helical fusion reactor, ion temperature Ti of ~ 10 keV. This is the highest record among stellarator/heliotron devices. Clear reduction of the ion thermal diffusivity in both core and edge regions in D discharge from hydrogen (H) was identified, indicating the effect of the isotope mass. This experimental result was supported by the initial results from gyrokinetic simulations including multi-species of ion. By measuring the neutron flux from D plasma, energetic particle (EP) behavior trapped in the helical ripple could directly be estimated, which is quite important for heliotron devices, because demonstration of the EP confinement is essential to realize the burning condition. Precise measurement of the tritium exhaust demonstrated the tritium mass balance including the evacuation system.
        Speaker: Dr Tomohiro Morisaki (National Institute for Fusion Science)
      • 177
        Overview of TJ-II stellarator results
        The flexibility of TJ-II together with its unique plasma diagnostics makes it an ideal laboratory to study the relationship between magnetic topology, electric fields, transport and model validation. **Zonal flows and heat transport.** HIBP measurements of zonal electrostatic potential relaxation are consistent with EUTERPE gyrokinetic (GK) simulations. The width of the oscillating zonal flow (ZF) radial electric field (Er) structures depends on its frequency. Additional GK simulations predict the localization of density fluctuations, in line with Doppler Reflectometry (DR) measurements. Transfer Entropy technique-based analyses shows that transport is not smooth and continuous but rather occurs in a stepwise fashion. **Impurity and particle dynamics**. Neoclassical (NC) theory results show how a negative Er field can coexist with an outward impurity flux. Flux-surface variations of electrostatic potential can have a significant impact on high-Z impurity radial fluxes. Probe measurements of plasma potential asymmetries on magnetic flux surfaces and DR measurements of poloidal asymmetries in Er fields, are consistent with NC simulations. Plasma core fuelling experiments with pellets show that the radial redistribution of particles can be understood qualitatively from NC predictions. Thermal neutrals react to low frequency plasma fluctuations. **NC and turbulent transport**. Zero frequency Er fields as well as low frequency ZF-like global oscillations have been identified during the Low to High (L-H) transition in H and D plasmas. No evidence of the isotope effect was observed in the L-H transition. **Power-exhaust physics.** The TJ-II programme on liquid metals address fundamental issues such as the self-screening effect driven by liquid lithium evaporation and the tritium inventory control. **Stellarator optimization**. Explicit expressions for the radial NC fluxes have been calculated in low collisionality regimes and have been included in a numerical code to deal with magnetic configurations close to omnigeneity. The relaxation of the constraint of periodicity imposed by the external confining magnetic field coils in a Helias configuration produces weak periodicity-breaking deformations of the plasma. The conditions of quasi-isodynamicity are not significantly altered by the periodicity-breaking distortions.
        Speaker: Dr Enrique Ascasibar (CIEMAT)
      • 178
        ELM and ELM-control Simulations

        Future devices like JT-60SA, ITER and DEMO require quantitative predictions of pedestal density and temperature levels, as well as divertor heat fluxes, to improve global confinement capabilities while preventing divertor erosion/melting in the planning of future experiments. Such predictions can be obtained from non-linear MHD codes like JOREK, for which systematic validation against current experiments is necessary. In this paper, we show the validation of ELM simulations with JOREK using quantitative comparison against JT-60U experiments. Note this is the first JOREK validation of ELM simulations at exact Spitzer resistivity. In addition, we demonstrate the essential importance of the separatrix position, required for a successful agreement with experimental data. On the basis of this validation, we propose estimates of ELM size, ELM-induced divertor heat-fluxes, and pre-ELM pedestal pressure, for future JT-60SA scenarios.

        Speaker: Dr Stanislas Pamela (CCFE - UKAEA)
      • 179
        Experiments in Disruption Avoidance for ITER Using Passive and Active Control

        Key plasma physics and real-time control elements needed for robustly stable operation of high fusion power discharges in ITER have been demonstrated in US fusion research. Optimization of the current density profile has enabled passively stable operation without n=1 tearing modes in discharges simulating ITER’s baseline scenario with zero external torque. Stable rampdown of the discharge has been achieved with ITER-like scaled current ramp rates, while maintaining an X-point configuration. Significant advances have been made toward real-time prediction of disruptions: machine learning techniques for prediction of disruptions have achieved 90% accuracy in offline analysis, and direct probing of ideal and resistive plasma stability using 3D magnetic perturbations has shown a rising plasma response before the onset of a tearing mode. Active stability control contributes to prevention of disruptions, including direct stabilization of resistive-wall kink modes in high beta discharges, forced rotation of magnetic islands to prevent wall locking, and localized heating/current drive to shrink the islands. These elements are being integrated into stable operating scenarios and a new event-handling system for off-normal events in order to develop the physics basis and techniques for robust control in ITER.

        Work supported by US DOE under DE-FC02-04ER54698, DE-SC0008520, DE-SC0016372, DE-FG02-04ER54761, DE-AC52-07NA27344, DE-SC0015878, DE-SC0014264, DE-FG02-99ER54524, DE-FOA-0001498, DE-AC02-09CH11466, DE-FC02-99ER54512, DE-SC0010720, DE-SC0010492, and the DOE Computational Science Graduate Fellowship, and by the EUROfusion Consortium with funding through FuseNet 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 Edward Strait (General Atomics)
    • P2 Posters
      • 180
        A promising grassy ELM regime for high-performance steady-state operations with metal wall in EAST and CFETR
        A highly reproducible stationary grassy ELM regime has been achieved in the EAST superconducting tokamak with water-cooled metal wall, exhibiting good energy confinement, H_98y2~1.1, strong tungsten impurity exhaust, and compatibility with low rotation, high density and fully non-inductive operations. It offers thus a highly promising operational regime in EAST, potentially applicable to future steady-state tokamak fusion reactors, such as the Chinese Fusion Engineering Test Reactor (CFETR). Recent linear and nonlinear simulations using ELITE and BOUT++ codes have uncovered, for the first time, the underlying physics of this grassy ELM regime. Both grassy and type-I ELMs are triggered by the marginally unstable intermediate-n peeling-ballooning modes (PBMs). However, the radial width of the linear mode structures cannot explain the small ELM size. The nonlinear simulations indicate that the pedestal current-profile relaxation is much slower than the pressure-gradient collapse. For the type-I ELMs, the high current density and gradient can still drive the kink/peeling-dominated low-n PBMs unstable even when the pressure gradient is significantly reduced, thus the collapsing front propagates radially inward, leading to large ELMs, as observed by Lithium BES on EAST. In contrast, for grassy ELMs, the pedestal current density and gradient are inherently lower and the operational parameter space can intrinsically improve the pedestal stability against the low-n PBMs. Hence, the instabilities quickly die away when the pressure gradient is just slightly reduced, leading to small ELMs. Some important features of the EAST grassy ELM regime are expected in future steady-state reactor-level plasmas, such as the relatively lower pedestal density gradient, higher SOL density and wider pedestal at high betap and low rotation. The desired edge density profile can be self-consistently generated by the strong cross-field particle transport driven by the high-frequency grassy ELMs. In particular, the pedestal density gradient in reactor-level plasmas could be even lower, since the plasma temperature and density at the separatrix are high so that the penetration of recycling neutrals into the pedestal is almost negligible. This may facilitate access to the grassy ELM regime in future devices, thus opening a potentially new avenue for next-step steady-state fusion development.
        Speaker: Dr Guosheng Xu (Institute of Plasma Physics, Chinese Academy of Sciences)
      • 181
        A Transmission Electron Microscopy Investigation of Defects Induced in Tungsten Foils by Gold (Au) and Boron (B) Ion Irradiation
        Tungsten is a promising candidate for first-wall material in fusion reactors and its use as a plasma-facing material is being investigated in both tokamaks as well as laboratory experiments [1, 2]. In fusion environment tungsten will be exposed to neutron, helium and hydrogen isotope implantation along with the heat flux which will lead to material damage. Irradiation by charged particles such as H, D, T, He, Au, W etc. is employed to surrogate the experiment of high energy and high flux neutron irradiation in tungsten. Present work concerns the study of ion mass in meso-scale defects created in tungsten using Transmission Electron Microscopy (TEM) after irradiated by (1) high energy heavy mass gold (Au of 80MeV) and (2) low mass boron (B of 10MeV) ions with a fluence of 1.3x1014 cm-2. Prior to irradiation tungsten foil samples of 100 µm thickness (99.96 % pure), procured from Princeton scientific corp. USA, and were recrystallized at 1838 K under 10-3 mbar base pressure in 200 mbar Ar+ 8% H2 environment. Defects created by Au and B ions irradiation in the recrystallized foil were characterized for the types of defect such as defect clusters, dislocation lines, loops etc., and are quantified in terms of dislocation line length, dislocation loop size and their densities using transmission electron microscopy. The small defect clusters in Au irradiated samples and dislocations segments and dislocation loops were observed in B irradiated samples. Furthermore, the Au ion irradiation has led to the formation of dislocation lines density lesser than that of B irradiated foil. [1] E. E. Bloom, Structural Materials For Fusion Reactors, Nuclear Fusion, 30, (9), 1879-1896, 1990. [2] M . Rubel, Structure Materials in Fusion Reactors: Issues Related to Tritium, Radioactivity and Radiation-Induced Effects, Transactions of Fusion Science and Technology, 53, 459-467, 2008.
        Speaker: Dr Prashant Sharma (ITER-India, IPR Gandhinagar)
      • 182
        Advanced energetic ion and impurity ion physics in 2D and 3D magnetically confined plasmas
        The VENUS-LEVIS code [D. Pfefferle, *et al*, Compt. Phys. Communs. **185**, 3127 (2014)] has been optimised for full orbit and guiding centre simulations in fully 3D electromagnetic fields. Curvilinear flux coordinate systems are deployed, with analytic (Fourier) representation of the fields for accurate simulation over slowing down time scales of fast particles and heavy impurities. An important recent application includes the viability of ICRH [J. Faustin, *et al*, Nucl. Fus. **56**, 092006 (2016)] and synergetic NBI-ICRH in Wendelstein-7X. Optimisation of fast ion generation and core heating is identified via variation of magnetic configuration, and methods of heating and associated properties (e.g. 3-ion species heating, minority heating, ICRH heating of NBI minority ions etc). Higher harmonic ICRH is a recent upgrade of the SCENIC ICRH package [M. Jucker, *et al*, Comp. Phys. Commun. **182**, 912 (2011)] that will permit various heating scenarios to be validated in advance of experiments. Recent updates to the VENUS-LEVIS code include higher order drift effects [S. Lanthaler, et al, Plasma Phys. Control. Fusion 59, 044014, (2017)], and advanced switching between full orbit and higher order drift orbit approximation during particle motion, as required in order to maintain accuracy and numerical efficiency. Applied for example to the current European DEMO design it is found that the magnetic ripple associated with 16 toroidal field coils has a weak affect on the radial transport of alpha particles, increasing the power flux due solely to prompt losses by a factor of about two. In addition, higher order guiding centre modelling has facilitated implementation of a 5 1/2D ICRH modelling scheme into SCENIC, which has many advantages over the standard quasi-linear operator approach. The VENUS-LEVIS code has also been updated [M. Raghunathan, *et al*, Plasma Phys. Control. Fusion **59**, 124002 (2018)] to include strong toroidal plasma rotation and the neoclassical effect of collisions in the frame of the diamagnetic flows of thermal ions in three dimensions. This upgrade has been applied to the transport of tungsten in JET hybrid scenarios susceptible to *m=n=1* continuous modes. Neoclassical collisional transport effects in 3D rotating magnetic fields can cause strong core accumulation of tungsten.
        Speaker: Dr Jonathan Graves (Ecole Polytechnique Federale de Lausanne (EPFL), Swiss Plasma Center (SPC), CH-1015 Lausanne, Switzerland)
      • 183
        Advances in Plasma-Wall Interaction Control for H-mode Operation over 100s with ITER-like Tungsten Divertor on EAST
        Managing excessively high divertor power and particle fluxes and related plasma-wall interactions (PWI) is one of the most critical issues for the steady-state operation of the EAST superconducting tokamak and future fusion devices, such as ITER and CFETR. A world record long pulse H-mode operation of 101.2 seconds with H_98=1.1 and total power injection of 0.3 GJ has been successfully achieved in EAST with ITER-like top tungsten (W) divertor, which has steady-state power exhaust capability of 10 MWm-2. The peak temperature of W target T≈500 oC and a heat flux ≈3 MWm-2 was maintained stably. Great efforts have been made to simultaneously control peak heat flux and particle/impurity exhaust towards the long pulse of 100 s time scale. Particle exhaust was optimized by preferentially directing the plasma flow toward the outer target with the ion Bx∇B drift away from the W divertor and improving divertor pumping with the top cryo-pump. Effective power dispersal was achieved by tailoring the three dimensional (3D) divertor plasma footprint using lower hybrid wave (LHW) through induced edge magnetic topology change and broadened plasma wetted area, thus reducing peak heat flux and W sputtering. Extensive lithium coating was employed to lower edge recycling, low-Z impurity content and W sputtering. In addition, divertor detachment in H-mode for PWI handling was achieved for the first time with W divertor in EAST. Compared with previous L-mode in EAST, in H-mode the detachment has a higher density threshold with n_e/n_G~ 0.65. Active feedback control of radiative divertor with neon impurity seeding was successfully achieved with f_rad ~ 18 - 36%, and a slight loss of plasma stored energy ~ 7-11%, offering a promising technique for steady-state divertor radiation and heat flux control. The upgrade plan and status of EAST bottom divertor from graphite to water-cooled W to accommodate more challenging PWI for steady-state H-mode over 400 s and L-mode operation over 1000 s will also be presented.
        Speaker: Dr Liang Wang (Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP))
      • 184
        Analysis of energetic particle driven toroidal Alfven eigenmodes in CFETR baseline scenario
        For burning plasmas in fusion reactors, Energetic Particles (EP) generated from plasma heating and D-T reaction can destabilize Alfven Eigenmodes (AE). Alfven eigenmodes can conversely induce transport and loss of energetic particles. It is one of the crucial issues to study the interaction between EPs and AEs for CFETR (China Fusion Engineering Test Reactor). Eigenanalysis of AEs in CFETR baseline scenario is taken by using AWEAC (Alfven Wave Eigen-Analysis Code), a developing code similar to NOVA/NOVA-k but dealing with asymmetric configuration of tokamaks. Linear simulations of TAEs driven by EPs are performed using the hybrid-kinetic MHD module in the NIMROD code. This HK-MHD module includes the kinetic effects of EPs through the coupling between a δf particle-in-cell (PIC) model for EPs and the 3D MHD model for the bulk plasma. The CFETR equilibrium used is obtained from the EFIT code based on self-consistent core-pedestal coupled OMFIT workflow. The “slowing down” distribution is used to model the equilibrium distribution of energetic ions from α particles produced by fusion. The frequency of TAEs generated by EPs in NIMROD simulation are consistent with the eigen-analysis results from AWEAC, which are within the range 40-100 kHz. For TAEs/EPMs driven by α particle from D-T fusion, the growth rate increases with both the toroidal mode number and EP beta fraction. Global 2D twist structures of TAEs/EPMs in CFETR baseline scenario, especially RSAE (Reverse Shear Alfven eigenmode) structure for some cases, are obtained for the first time using NIMROD. These results may be helpful for the future design of CFETR operations. Acknowledgments: This work is supported by the National Magnetic Confinement Fusion Science Program of China grant Nos. 2014GB124002 and 2015GB101004, and by the Natural Science Foundation of China grant No. 11205194. One of the authors P. Zhu also acknowledges the supports from U.S. DOE grant Nos. DE-FG02-86ER53218 and DE-FC02-08ER54975. This research used the computing resources from the Supercomputing Center of University of Science and Technology of China, the National Energy Research Scientific Computing Center in US, and local clusters in USTC, such as HPC, Lenovo, Inspur, and HWC.
        Speaker: Dr Yawei Hou (University of Science and Technology of China)
      • 185
        Burning Plasma Simulation with Alpha-Particle Heating
        To achieve self-sustained ignited operation in a high energy Tokamak, it is important to understand and maximize the energy confinement time, which falls in the domain of transport theory. To analyze and understand dynamics of plasma in Tokamak, performing a one-dimensional transport simulation is still one of the best approaches. In our work we focus on burning plasma simulation and study the alpha particle heating in high energy Tokamaks like ITER. Transport simulations can be performed by solving 1D transport equations using codes such as LCPFCT (Laboratory for Computational Physics Flux-Corrected Transport)[1], which is used to solve 1D generalized coupled continuity, momentum and energy equations along with Maxwell’s equations. The transport equations are solved in flux coordinates by coupling with 2-D tokamak equilibrium. In this model, the effects of fusion reactions, coulomb collisional losses, radiation losses, alpha-heating, auxiliary heating and neo-classical Ware pinch are included. This will predict the performance of tokamak based fusion reactor for obtaining the steady state operation. This model is being developed and will be bench marked with published results. This will be use to predict the performance of SST2-like [2] and ITER-like [3] cases and results will be presented in this paper. References: [1] Boris, J.P.; A.M. Landsberg; E.S. Oran; and J.H. Gardner. 1993. "LCPFCT - A Flux-Corrected Transport Algorithm for Solving Generalized Continuity Equations." NRL Memorandum Report 93-7192. [2] R Srinivasan and the Indian DEMO Team, Fusion Engineering Design, 112 (2016) 240 [3] Progress in the ITER Physics Basis, Nuclear Fusion, 47 (2007)
        Speaker: Mr Udaya Maurya (InIPR)
      • 186
        Characterization of Particle Growth and Enhancement of Sputtering Yields in a Co-generated Dusty Plasma
        Most of the tokamaks including ITER, a significant part of the plasma-facing component including diverters, limiters, etc is comprised with graphite material. In the fusion plasma environment, the graphite gets bombarded by hydrogen and its isotope (deuterium and tritium) ions and erode graphite to a significant extent. Since such carbon particles can retain large amounts of hydrogen, dust contributes to the problem of inventory of radioactive tritium inside the fusion machine. Another impact of the dust particles in the operation of a fusion device is the possible degradation of the discharge performance. Such particles penetrating in the core plasma region can lead to discharge disruption. Thus, in order to perform successful fusion experiments it is important to assess and understand the processes by which dust is formed and by which it interacts with the fusion device and its plasma. Instead of understanding processes that exactly happen inside a fusion reactor, it is always better to match some aspects of graphite-hydrogen interaction in a plasma environment in small laboratory devices, and study the physical processes. To address some of this issues, we have performed an experiment to examine the particle growth and sputtering yields in a DC glow discharge plasma in between the graphite electrodes.
      • 187
        Comparison of energetic particle radial transport between single-n and multiple-n simulations of Alfvénic modes

        The results of a set of simulations of Alfvén modes driven by an energetic particle population are presented, with the specific aim of comparing energetic particle radial transport between single-n and multiple-n simulations. The hybrid reduced O($\epsilon^3_0$) MHD gyrokinetic code HMGC is used, retaining both fluid (wave-wave) and energetic particles nonlinearities. The code HMGC retains self-consistently, in the time evolution, the wave spatial structures as modified by the energetic particle (EP) term.
        A model equilibrium has been considered, rather than a specific experimental device, with the aim of studying how the dynamics of the EP driven Alfvénic modes changes when considering single-n or multiple-n simulations, while keeping all the other parameters fixed. A circular, shifted magnetic surface, static equilibrium has been considered, characterized by a large aspect ratio ($\epsilon_0= 0.1$) and a parabolic safety factor profile with $q_0=1.1$ and $q_a=1.9$ being, respectively, the on-axis and edge safety factor. A bulk ion density profile $n_i(r)$ ~ $(q_0/q(r))^{2}$ has also been assumed, in order to have the toroidal gap radially aligned, for all the mode considered. Regarding the EPs, an isotropic Maxwellian distribution function has been considered.
        Simulations with toroidal mode numbers 1≤n≤15 have been considered. A variety of modes are observed (TAEs, upper and lower KTAEs, EPMs) during the linear growth phase. All the strongly unstable modes (4≤n≤12) exhibit pronounced (both up and down) frequency chirping at saturation. Nevertheless, no appreciable global modification of the energetic particle density profile is observed at saturation for the unstable modes.
        On the contrary, multiple-n simulations, with the same Fourier toroidal mode spectrum of the set of single-n simulations, exhibit an appreciable broadening of the energetic particle radial density profile at saturation, thus showing an enhanced radial transport w.r.t. the single-n simulations. Moreover, the sub-dominant modes are strongly modified by the nonlinear coupling, which results both from the MHD and from the energetic particle terms. The present nonlinear simulations show that all the toroidal modes saturate almost simultaneously, after inducing an enhanced energetic particle radial transport. No evidence of the so-called "domino" effect is observed.

        Speaker: Gregorio Vlad (ENEA, Fusion and Nuclear Safety Department, Frascati, Italy)
      • 188
        Comprehensive magnetohydrodynamic hybrid simulations of fast ion losses due to the fast ion driven instabilities in the Large Helical Device
        In the LHD, the fast ion confinement has been investigated by using three tangentially injected neutral beams (NBs) with 180 keV fast ions and/or two perpendicularly injected NBs with 40-80 keV fast ions. The Alfvén eigenmodes (AEs) are observed during the tangential-NB injections. The fast ion driven instabilities enhance the fast ion losses. It is important to identify the instabilities and clarify the properties of the lost fast ions due to the instabilities. A hybrid simulation code for nonlinear magnetohydrodynamics (MHD) and energetic-particle dynamics, MEGA, has been developed to simulate recurrent bursts of fast ion driven instabilities including the energetic-particle source, collisions and losses. In order to identify the instabilities and to clarify the process of the fast ion losses in the LHD experiments, MEGA is applied to the LHD plasmas, where fast ion driven instabilities and lost fast ion properties are investigated by using tangential-NBs, with the realistic conditions close to the experiments. In a plasma with tangential-injected neutral beams (NBs), the Alfvén eigenmode (AE) bursts with m/n=2/1 occur recurrently. As a result, the stored fast ion energy is saturated at a lower level than that of a classical slowing down calculation where the magnetohydrodynamic (MHD) perturbations are neglected. Fast ion losses occur during the AE bursts. The fast ion losses brought about by the AE bursts are proportional to the square of AE amplitude, which reproduces well the LHD experiment. This indicates the emergence of stochasticity in the fast ion loss process. The fast ions deposited well inside the plasma up to the magnetic axis are significantly lost for the counter-injected fast ions. We present the first self-consistent simulations that reproduce and clarify the fast ion loss properties in the LHD experiments.
        Speaker: Dr RYOSUKE SEKI (National Institute for Fusion Science)
        Summary Slides
      • 189
        Design and Qualification of Precision Support Structure for Diagnostics
        The ECE diagnostic is planned to be used for the measurements of plasma electron temperature profile with good spatial and temporal resolutions. Secondary objectives are to obtain information on non-thermal electron populations and the power loss due to ECE. One of the major requirements of ITER like Fusion Device is to study the plasma parameter to ascertain and control the fusion reaction. These diagnostic systems need to be assembled in the constrained space around the machines with tight tolerance for optical accuracy in many cases. The ECE diagnostic is planned to be used for the measurements of plasma electron temperature profile with good spatial and temporal resolutions. Secondary objectives are to obtain information on non-thermal electron populations and the power loss due to ECE. This diagnostics system has about 40m long multiple wave-guides to transmit the signal from the ITER Diagnostic building for data acquisition and assessment. For which the Design and qualification of Wave-guide Support Structure has been carried out. This paper elaborates on the Cost effective Design and Qualification of precision alignment cum support structures for the wave-guides which needs to be aligned accurately. The cost effective design has been developed using off the shelf components. This design reduces the ±25mm tolerances on the building to only ±0.5mm on the assembled wave-guide with sufficient stability. The Support Structure qualification has been done using the Design by analysis approach of the ASME code and the stresses are assessed using the ANSYS tool
        Speaker: Mr Shrishail Padasalagi (ITER-India, IPR, HBNI)
      • 190
        E_rxB shear effect on cross phase mitigates ELM at high collisionality
        A non-stationary, effective edge localized modes (ELMs) mitigation / suppression regime has been recently obtained by counter NBI heating at high collisionality on the Experimental Advanced Superconducting Tokamak (EAST). Our results show that counter NBI can significantly enhance the reversed toroidal rotation as well as the E_r×B flow shear of the pedestal. With the increased E_r×B flow shear, the ELM sizes can be suppressed by nearly 80%. The increased E_r×B flow shear can also broaden the power spectrum of the pedestal turbulence and enhance the amplitude of modes with high frequency (f>100kHz). The bispectrum study indicates that the nonlinear mode coupling of the pedestal turbulence also increases in counter NBI case, which can interrupt the linear growth of the peeling mode, thus leading to the suppression of ELM. When power of counter NBI is high enough, an ELM-free H mode can even be achieved on EAST. During the ELM-free H mode, the line averaged density as well as the amplitude of resistive ballooning mode keeps increasing until the H-L back transition. Those observations may link with the density limit in H mode discharge. BOUT++ simulations have been applied to study the characteristics of edge-localized mode at fixed high collisionality for different E_r structure. The simulation result reveals that the increased E_r×B shear suppresses the ELM size and delays the pedestal crash, which is consistent with the observations on EAST. Analysis of the cross-phase spectrum of potential and pressure perturbations indicates that the increased E_r×B shear can shorten the phase coherence time τ_c and flatten the spectrum of τ_c, which is and limited by nonlinear mode interaction. Thus, the peeling-ballooning mode doesn't get enough time to allow growth to large amplitude, which can be supported by the bispectrum study on EAST that increased ErxB flow shear can enhance the nonlinear interaction. Besides the collisionality, our simulations suggest a new way (Er shear) to control the ELM size, which is consistent with observed ELM suppression at larger E_r×B shear in high collisionality plasmas on EAST.
        Speaker: Dr Defeng Kong (Institute of Plasma Physics Chinese Academy of Sciences)
      • 191
        Effect of Cathode Geometry on Magnetically Coupled Hollow Cathode Plasma Source
        A direct current (dc) plasma source consisting of hollow cathode geometry and a constricted anode is presented. The effect of a hollow cathode geometry on radial density distribution of a magnetized plasma column has been studied in a low-pressure (approximately 1.4Pa) argon discharge. The plasma column is characterized using Langmuir probe and the radial density distribution for two different 'inside' profiles of a hollow cathode is discussed. Probe measurement indicates that conical-profile hollow cathode produces a plasma column with centrally peaked plasma density whereas cylindrical-profile hollow cathode forms plasma column with off-centered density peak. Thus overall dynamics of perpendicular and oblique cathode sheaths behind the sustenance of magnetized plasma column has been discussed. **Keywords:-** constricted anode, conical hollow cathode, cylindrical hollow cathode, Langmuir probe, magnetized plasma column, radial density distribution, oblique cathode sheaths.
        Speaker: Mr Montu Bhuva (Institute for Plasma Research, INDIA)
      • 192
        Effect of the Controlled Density Gradient on Equilibrium and Confinement in a Simple Toroidal Device with two plasma sources
        A simple toroidal device (SMT) is a toroidal device in which plasma is confined by the application of toroidal and vertical magnetic field only resulting in absence of a conventional effective rotational transform. Such devices provide a simple and well diagnosable test-bed for studies related to equilibrium, fluctuations and particle confinement for Tokamak edge. The device BETA at the Institute for Plasma Research (IPR) is one such SMT with a plasma major radius of 45 cm and minor radius of 15 cm and a maximum toroidal field of 0.1 Tesla. Quasi-static equilibrium in an SMT is controlled by the nature of fluctuation and flow [1, 2]. As observed in hot cathode discharges studied earlier [1, 2], density gradient provide fluctuation in the plasma and hence the instabilities [2]. Whereas radial electric field provides poloidal flow. Thus, the conditions are akin to Tokamak edge. To experimentally understand the effect of the density gradient, it is desirable to be able to control the local gradient at the outboard side by an additional plasma source. To this end, a new microwave source of frequency 2.4 GHz and power about 0.5 kW has been developed [3]. Hot cathode and microwave sources are used in tandem such that the upper hybrid resonance falls at the outboard density gradient region, which in turn allows us to control the density gradient locally. The details of the experiment will be presented. References [1] T. S. Goud, Thesis, Institute for Plasma Research, Gandhinagar, Gujarat, India (2012). [2] Umesh Kumar, Shekar G Thatipamula, R. Ganesh, Y. C. Saxena and D. Raju, Phys. Plasmas 23, 102301 (2016). [3] Umesh Kumar, R. Ganesh, K. Sathyanarayana, Y. C. Saxena, S. G. Thatipamula, D. Raju , Manuscript under preparation
        Speaker: Mr Umesh Kumar (Institute for Plasma Research, Gandhinagar, Gujarat, India)
      • 193
        Experimental Study of Multi-scale Interaction between (Intermediate, Small)-scale Microturbulence and MHD modes in EAST Plasmas
        Understanding plasma transport in phases with significant MHD activities (especially during plasma current ramp-up/down and disruption) in tokamak plasmas is crucial for predicting and thus controlling plasma behavior for future fusion devices, e.g. ITER. Since microturbulence plays an important role in driving anomalous plasma transport, the interactions between MHD modes and microturbulence is thought to be important in determining anomalous plasma transport [1]. Recent theoretical results in the literature show that microturbulence can nonlinearly interact with macro-instabilities such as kink/tearing mode through nonlinear cascade process or through temperature and/or density profile modulation from macro-instabilities. Due to the huge temporal and spatial scale separation between microturbulence and MHD modes, it is impossible for the present-day supercomputers to simulate their nonlinear interactions in a self-consistent way. In this talk, we present evidence of multi-scale interactions between (intermediate, small)-scale (kρ_i~2-6) microturbulence and MHD modes in EAST plasmas, including the first experimental identification of nonlinear coupling between microturbulence and an MHD mode during the current ramp-down phase in a set of L-mode plasmas in EAST [2] and the effects of 2/1 classical tearing mode on microturbulence [3] in the core of another set of EAST L mode plasmas using the EAST CO_2 laser collective scattering diagnostic in forward mode and far-forward mode. We demonstrate the nonlinear coupling between microturbulence and MHD mode with bispectral analysis [4] and envelope method [5], showing statistically significant bicoherence and modulated turbulent density fluctuation amplitudes correlated with the MHD mode. We also show that microturbulence spectral power is correlated to the 2/1 tearing mode and modulation effects on microturbulence by the 2/1 tearing mode. [1] P.J. Sun et al 2018 Nucl. Fusion 58 016003 [2] P.J. Sun et al 2018 Plasma Phys. Control. Fusion 60 025019 [3] Kim Y C and Powers E J 1979 IEEE Trans. Plasma Sci. PS-7 120 [4] Y. Nagashima et al 2005 Phys. Rev. Lett. 95095002 *Work supported by the National Natural Science Foundation of China with Contracts Nos.11475222, 11505228, 11735016, 11575238
        Speaker: Dr Pengjun Sun (Institute of Plasma Physics, Chinese Academy of Sciences)
      • 194
        Fast-ion studies in high performance fully non-inductive discharges on EAST
        On the EAST tokamak, one hundred seconds steady state H-mode (H98y2~1.1) discharge has been achieved by RF-only (LHW+ECRH+ICRH) heating with improvement of the auxiliary heating and current drive systems on actively cooled ITER-like mono-block tungsten divertor. Towards EAST high performance advanced state-steady operation regimes, fast-ion related physical issues become crucial for achieving EAST scientific objectives with both co-Ip and counter-Ip neutral beam injections [1]. Accordingly, EAST several complementary fast-ion measurements [2] have been developed and validated, e.g. fast-ion D-alpha (FIDA), fast-ion loss detectors (FILDs), neutral particle analyzers (NPA), neutron spectrometers and TOFED, etc. In recent experiments, compared with RF-only discharge, NBI and RF plasmas has a higher βp and H98y2, although the bootstrap current fraction fBS is nearly the same [3], TRANSP analysis shows that it is mostly due to fast ions, and fast ions do not contribute significantly to fBS. To obtain high performance plasma and improve confinement and transport on EAST, key related parameters (e.g. density, plasma current, beam energy, etc.) need to optimize further to reduce the fast ion slowing down time and prompt loss. To investigate fast-ion distribution function and prompt loss, different beam voltage and plasma current are investigated as well. Experimental results show that prompt loss from counter beams is large and can be reduced by reducing beam voltage and increasing plasma current, which is consistent with simulations. The relationship of the fast ion loss and distribution to the different beam settings and plasma parameters will be reported in this paper, which is very helpful to understand energetic particle physics in long pulse H-mode plasmas on EAST and contributes to ITER.
        Speaker: Ms Juan Huang (CnIPPCAS)
        Summary Slide
      • 195
        Global gyrokinetic multi-model simulations of ITG and Alfvenic modes for tokamaks and the first operational phase of Wendelstein 7-X
        Results from a hybrid approach (CKA-EUTERPE code) which couples an MHD code with a gyrokinetic code are presented. Although perturbative, it offers a relatively fast way to investigate the destabilisation of Alfven modes by fast particles. TAE saturation amplitudes and their scaling with growth rate and collisionality were investigated in a tokamak as well as in Wendelstein 7-X. Full volume linear electrostatic gyrokinetic simulations for an OP 1.1 Wendelstein 7-X scenario showed modes driven by the strong electron temperature gradient with negligible influence from trapped particles. Using a Fourier solver approach, long-time fully kinetic runs of damped GAEs and TAEs could be performed. Super-resolution methods allowed to accurately resolve the continuous Alfven spectrum.
        Speaker: Dr Ralf Kleiber (Max-Planck-Institut für Plasmaphysik)
      • 196
        Imaging of SST-1 plasma with LHCD power
        Plasma imaging is an essential diagnostics system for any tokamak as it can provide vital information on various plasma parameters. These systems are ones of the first diagnostics installed and are basic not only at start-up stage but also in subsequent operations. Imaging system generally consists of at least two cameras, one of them is a high speed camera and another one is slow speed camera. The first one provides study of fast processes in plasma and plasma-wall interaction. The second camera ensures video image for general plasma operation monitoring. Generally, imaging systems make it possible to plasma monitoring, plasma formation and start-up: break down and ramp-up, study and observation of the magnetohydrodynamic (MHD) instability-edge localized modes (ELMS), multifaceted, asymmetric radiation from the edge (MARFE), displacements study dust migration and deposition study, plasma wall interaction, plasma position control. A Tangential viewing optical imaging system is installed on SST-1. Plasma images are transferred through coherent optical imaging fibre and coupled to the CCD camera placed outside the SST1 machine. The CCD camera used with this system operates at 30 frames/sec to acquire plasma images. The data from the CCD camera is transferred through gigabit Ethernet cable to acquisition PC placed in diagnostics Lab. The whole system is fully automated for operation and data acquisition of the imaging data. In this paper we are presenting observations during LHCD power launching in the SST1 machine. The LHCD pulse was launched into the plasma at various instant of time and varying in pulse length. Plasma images exhibit change in distribution of visible radiation during the interaction of LHCD with the plasma. This increase in emission may be attributed to the enhance in plasma wall interaction as the plasma moves outwards which results in increase of plasma wall interaction. Decrease in plasma size is also observed during interaction of LHCD pulse with plasma.
        Speaker: Dr Manoj Kumar (Institute for Plasma Research)
      • 197
        Investigations on Temperature Fluctuations and Energy Transport in ETG Dominated Large Laboratory Plasma
        Extensive measurements are carried out on micro turbulence because of their possible role in causing anomalous particle and energy transport in fusion devices [1]. Outcome from past investigations suggest that the Electron Temperature Gradient (ETG) driven turbulence is considered presently as a major source of anomalous plasma transport in fusion devices, as transport by ion scale turbulence is largely understood. Direct measurement of ETG is extremely difficult in fusion devices because of its extremely small scale length ($\sim \mu m $ ). In this background, efforts were made in Large Volume Plasma Device (LVPD), to produce plasma suitable for carrying out investigations on ETG turbulence ( $ \sim mm $ ). Introduction of Electron Energy Filter (EEF) divides LVPD plasma into three distinct regions of Source, EEF and Target plasmas. In the core region of target plasma ( $ x \le 45 cm $ ), unambiguous, identification of ETG turbulence is successfully demonstrated [2, 3]. Simultaneous measurement of fluctuations in electron temperature ( $ 10 \% - 30 \% $), plasma density ($ 5 \%-10 \% $) and potential ($ 0.5 \% -2.5 \% $) are carried out. Particle and energy transport are estimated from $ < \tilde{n}_e \tilde{E}_{\theta}> $ and $ < \tilde{T}_e \tilde{E}_{\theta} > $ correlations. It was observed that electrostatic particle transport agrees well with theoretical estimates [4] while, electromagnetic particle flux satisfies the relationship ( $ \Gamma_{es} \sim 10^{-5} \times \Gamma_{es} $ ). Strong negative correlation is observed between fluctuations of density and temperature with potential fluctuations, showing correlation coefficients, $ C_{\tilde{n}_e,\tilde{\phi}} \sim -0.8 $ and $ C_{\tilde{T}_e,\tilde{\phi}} \sim -0.7 $ respectively. This paper will present results on work carried out for energy transport due to ETG turbulence. Details on adopted diagnostic methods, for accurate measurement of temperature fluctuations will also be presented. A comparison will be made of experimentally derived energy transport with theoretically estimated values.
        Speaker: Mr Prabhakar Srivastav (Institute For Plasma Research, Bhat Gandhinagar India-382428)
      • 198
        Leak Width in a Multi-cusp Field Configuration: A Revisit with a Versatile Experimental device
        The cusp configuration reduces the plasma losses to boundary by diverging plasma to the narrow regions where the magnetic field lines intersect the wall. The efficiency of multi dipole cusp confinement depends on the plasma losses through cusp loss area, widely known as leak width. The Multi-line cusp Plasma Device (MPD) used electromagnets for plasma confinement and gives opportunity to vary the magnetic field strength which controls the plasma loss area. We discuss the scaling of leak width with different magnetic field strength to understand its role in the particle confinement for such configurations
        Speaker: Ms Meenakshee Sharma (Institute for Plasma Research)
      • 199
        Model-Predictive Kinetic Control for Steady State Plasma Operation Scenarios on EAST
        Robust model-predictive control (MPC) algorithms based on extremely simple linear data-driven models have been recently developed for plasma kinetic control on EAST. This paper shows, for the first time, that MPC can be performed using a two-time-scale approximation, considering the kinetic plasma dynamics as a singular perturbation of a quasi-static magnetic equilibrium, which itself is governed by the flux diffusion equation. This technique takes advantage of the large ratio between the time scales involved in magnetic and kinetic transport, and is applied here to the simultaneous control of the safety factor profile, q(x), and of the poloidal beta parameter, beta_p, on EAST. MPC results in a much faster and more robust control than the so-called near-optimal control algorithms that were tested previously [D. Moreau, et al., Nucl. Fusion 55 (2015) 063011]. The models are state-space models identified with datasets obtained from fast nonlinear METIS simulations (METIS includes an MHD equilibrium and current diffusion solver, and combines 0-D scaling laws and ordinary differential equations). For a given operation scenario, the identified model is augmented with an output disturbance model, which is used to estimate the mismatch between measured and predicted outputs and ensures robustness to model uncertainties. An observer provides, in real time, an estimate of the system states and disturbances, and the controller predicts the behavior of the system over a prediction horizon, taking the actuator constraints into account. For plasma parameters typical of the high-beta_p steady state operation scenarios on EAST, nonlinear closed-loop simulations show that the desired q(x) profiles and beta_p can be obtained in about 2.5 s and 0.5 s, respectively, and with a monotonic approach to their target values. This is essential for avoiding MHD instabilities during the build up of the plasma equilibrium. In these control simulations, the actuators are the LHCD system at 4.6 GHz, the ICRH system, and optionally the plasma surface loop voltage. Various examples are shown, with negative shear or monotonic q-profiles, and with different beta_p target waveforms. The actuators adjust in order to reach the various beta_p targets while maintaining the q-profile in steady state, with the desired shape (or as close as possible if the q(x) and beta_p targets are not achievable).
        Speaker: Prof. Didier Moreau (CEA/DSM/IRFM)
      • 200
        Modeling studies of X-divertor configuration on SST-1 tokamak using SOLPS5.1
        To solve the challenging problem of heat removal in a tokamak based fusion reactor, several advanced divertor configurations have been proposed and studied. We present here the reults of our studies of one of the more promising configurations, the X-divertor, conducted for the parameters of the Indian tokamak SST-1. Using the equilibrium code CORSICA, we develop the appropropraite magnetic geometry and then study its performance using the Scrape-Off Layer plasma transport code package SOLPS5.1/B2.5-Eirene. One of the main motivation was to find out if the X-divertor (XD) could boost the heat handling capacity as compared to the standard divertor (SD) configuration for SST-1 that designed to handle 1 MW of total input power. In order to compare the performance XD with the existing SD, we first ensured the core equivalence of both configurations. An additional poloidal field coil was placed behind the divertor target to produce XD configuration. The plasma equilibrium for SD and XD are generated. The divertor index (DI) is varied from 2 to 13. For a plasma operation with P=1 MW input power and plasma edge density ne = 1x1019 m-3 , the peak heat load on the target plates in the X-divertor configuration has reduced by 50% as compared to standard divertor; the heat flux profile near the separatrix was also broadened due to flaring of field lines. The latter increases the plasma-wetted area at the targets. This is a preliminary demonstartion that XD will allow SST-1 to operate at higher input power.
        Speaker: Mrs Himabindu Manthena (IPR)
      • 201
        Nonlinear decay and plasma heating by toroidal Alfvén eigenmodes
        Gyrokinetic theory of nonlinear mode coupling as a mechanism for toroidal Alfv\'en eigenmode (TAE) saturation and thermal plasma heating in the fusion plasma related parameter regime is presented, including 1) parametric decay of TAE into lower kinetic TAE (LKTAE) and geodesic acoustic mode (GAM), and 2) enhanced TAE coupling to shear Alfv\'en wave (SAW) continuum via ion induced scattering. Nonlinear decay of TAE into a GAM and a LKTAE with the same toroidal/popoidal mode number is investigated due to its crucial implications on TAE nonlinear saturation, improved confinement, as well as energetic particle (EP) power channeling, including fusion-alpha power density to bulk thermal plasma heating. The parametric dispersion relation is derived and analyzed, and the parameter range for this process to occur and dominate over other mechanisms is discussed. The nonlinearly generated LKTAE and GAM can be dissipated via electron and ion Landau damping, respectively, leading to anomalous EP slowing down and channeling of EP power to thermal ion heating. The thermal plasma heating rates are also estimated. Furthermore, the nonlinearly generated GAM, as the finite frequency zonal flow, could contribute to regulating drift wave turbulence and consequently, improved confinement. The TAE frequency cascading via nonlinear ion induced scattering and saturation due to enhanced coupling to SAW continuum is also investigated. The wave-kinetic equation for the TAE spectrum evolution in the continuum limit is derived using nonlinear gyrokinetic theory, which is then solved to obtain the saturation spectrum of TAE, yielding a lower fluctuation level than previous drift-kinetic theoretical estimates, as a consequence of the enhanced nonlinear couplings in the short wavelength limit. The bulk ion heating rate from nonlinear ion Landau damping is also calculated. Our theory shows that, for TAE saturation in the parameter range of practical interest, several processes with comparable scattering cross sections can be equally important.
        Speaker: Dr Zhiyong Qiu (Zhejiang University)
      • 202
        Numerical simulations of GAE stabilization in NSTX-U
        Simulations of confinement-limiting Alfvén eigenmodes in the sub-cyclotron frequency range show a robust physical stabilizing mechanism via modest off-axis beam injection, in agreement with experimental observations from National Spherical Torus Experiment (NSTX-U). Experimental results from NSTX-U have demonstrated that neutral beam injection from the new beam sources with large tangency radii deposit beam ions with large pitch, which can very effectively stabilize all unstable Global Alfven Eigenmodes (GAEs). Beam-driven GAEs have been linked to enhanced electron transport in NSTX, and the ability to control these modes will have significant implications for NSTX-U, ITER, and other fusion devices where super-Alfvénic fast ions might be present. Nonlinear simulations using the HYM code have been performed to study the excitation and stabilization of GAEs in the NSTX-U right before and shortly after the additional off-axis beam injection. The simulations reproduce experimental finding, namely it is shown that off-axis neutral beam injection reliably and strongly suppresses all unstable GAEs. Before additional beam injection, the simulations show unstable counter-rotating GAEs with toroidal mode numbers and frequencies that match the experimentally observed modes. Additional of-axis beam injection has been modelled by adding beam ions with large pitch, and varying density. The complete stabilization occurs at less than 10% of the total beam ion inventory.
        Speaker: Dr Elena Belova (PPPL)
      • 203
        Observations of Plasma Stimulated Electrostatic Sideband Emission and Harmonic Distortion: Evidences of Over-dense Plasma Generation inside a Microwave Discharge Ion Source
        Microwave discharge ion source (MDIS) is used in many applications including accelerators based neutron generators on suitable target through D_D or D_T fusion. The electromagnetic (EM) pump wave (ω_0) can propagate beyond cut off plasma density by changing its polarity and/or decomposing into different daughter waves through which it transfer its energy thus producing over dense plasma. Role of electric field on power coupling through different decay channels during density jump from under-dense to over-dense is obtained by theoretical modeling. This is validated with experimentally obtained spectral features in the ion plasma frequency range. In the present experiment, the plasma stimulated emission spectra was measured in the frequency range 0.5ω_0to 3ω_0 to understand the different probable energy decay channels role; e.g. Electron Bernstein waves (EBWs), Ion cyclotron waves (ICWs), Lower hybrid oscillations (LHOs), Ion Bernstein waves (IBWs) and Ion Acoustic Waves (IAWs) etc. The energy decays through different ion-type waves by parametric instability is studied by observing the different side-bands generation around the pump frequency and also the electron cyclotron (EC) harmonic frequencies. The intensity and growth rate of IAWs/ICWs and harmonics (up to 3rd) from parametrically decayed ordinary (O) mode pump wave was used to get an estimate of electric field and localized electron temperature. The density threshold of each electrostatic IAWS/ICWs was measured by stepping pump wave amplitude and external magnetic field. The IAWs lines appear at lower density threshold than the ICWs emission lines. The measured IAWs and ICWs ranges from 317-397 kHz and 410-555 kHz respectively with a density jump from 9.3x1016/m3 to 4.9x1017/m3.At higher density (>3.3x1017/m3), the electrostatic ICWs lines dominates the IAWs thereby yielding negligible damping through ion waves.
        Speaker: Mr Chinmoy Mallick (institute for Plasma Research)
      • 204
        Preliminary Results of Wall Conditioning Experiments using High Power ICRH System on SST-1 at Different Toroidal Magnetic Fields
        Proper wall conditioning has turned out to be an essential element for achieving the highest possible plasma performance in present-day fusion devices. The main issues are controlling the generation of plasma impurities, liberated by plasma-surface interactions. Superconducting fusion machines need efficient wall conditioning techniques for routine operation in between shots in the presence of high toroidal magnetic field for wall cleaning to control the in-vessel impurities. Ion Cyclotron Wall Conditioning (ICWC) is fully compatible with steady-state tokamak in presence of magnetic field. Here we report the preliminary results of ICRF wall conditioning experiments done on Steady State Superconducting Tokamak (SST-1) using High Power Ion Cyclotron Resonance & Heating (ICRH) System indigenously developed including MW RF generator, Transmission Line with Matching System, Vacuum Transmission Line (VTL) and Fast Wave Poloidal Antenna with Faraday shield. In the first stage, the experiments are conducted to condition the complete system and antenna by introducing low power RF pulses in the SST-1 machine. It is observed that the conditioning pulse removes gas species from Antenna and VTL. In the second stage, the wall conditioning experiments are conducted at 0.2 - 0.4 T and in third stage the wall conditioning experiments are conducted at 1.5 T in Helium gas. The diagnostics used are the visible camera, spectroscopy, Residual Gas Analyzer (RGA) etc. More than 600 RF pulses of 150 kW with 0.5 seconds on time and 0.8 Seconds off time were introduced and significant impurity generation is observed from antenna and vacuum vessel. It is observed that RF conditioning at low pressure releases H2 and other gas species. The previous ICWC experiments done on Aditya tokamak show that in presence of toroidal magnetic field (0.45 T) conditions as well as with 20% Helium gas in a hydrogen plasma is found more effective in releasing wall impurities like water & methane as half an order (~ 5) of initial vacuum condition. The preliminary results on SST1 show that the ICWC in the presence of magnetic field seems to be effective and can be used an alternative method for vessel wall conditioning. In this paper, the above-mentioned experiments and results will be discussed.
      • 205
        Progress towards Development of Long Pulse ITER Operation through RF Heated H-mode Experiments on EAST and HL-2A
        Recent long pulse experiments in EAST have resulted in a new world record of 100 s long H-mode discharge, sustained by the radiofrequency (RF) systems, predominantly Lower Hybrid Current Drive (LHCD). In parallel, experiments in HL-2A have demonstrated successful LH wave coupling in H-mode plasmas with an ITER-relevant passive-active multijunction (PAM) LHCD launcher. These two achievements, obtained as a part of the specific EU-China collaboration, show the viability of LHCD as a successful method for heating and current drive in high performance H-mode plasmas. Experimental comparison of the two LHCD systems in EAST shows that the current drive efficiency is higher for the 4.6 GHz system than for the 2.45 GHz system. Higher power was therefore systematically used on the 4.6 GHz launcher in the long pulse experiments. Increasing the radial distance between the plasma and the launchers (up to 8 cm) was employed as method to optimizing the density in front of the launchers and to avoiding hot spots during the long H-modes. Lithium evaporation showed to have a beneficial effect on the LH current drive efficiency. An increase in efficiency from  = 0.8×1019 AW-1m-2 to 1.1×1019 AW-1m-2 was observed when the accumulated Lithium in the EAST vessel was above 150 g. Good agreement between experimental results and simulations with C3PO/LUKE is obtained for EAST fully non-inductive discharges. C3PO/LUKE can well reproduce the experimental values of the internal inductance, as well as the non-inductive current profile obtained from equilibrium reconstruction constrained by interfero-polarimetry. In HL-2A, a 3.7 GHz LHCD system with four klystrons and an ITER-relevant PAM launcher has been successfully brought into operation and used in H-mode experiments. Coupling of LH waves in ELMy plasmas has thus been demonstrated with an ITER-relevant launcher for the first time. The maximum coupled LH power has reached 1 MW in L-mode and 0.9 MW in H-mode. H-modes were triggered and sustained with LHCD together with ~ 700 kW NBI power. In H-modes with ne > 2.5×1019 m-3, a reduction in ELM amplitude and increase in ELM frequency were observed for injected LH power > 300 kW. The divertor peak heat load released by the ELMs was strongly reduced during this phase, which suggests that the LH power can be used for controlling ELMs.
        Speaker: Dr Annika Ekedahl (CEA, IRFM)
      • 206
        Radial Characteristics of a Magnetized Plasma Column
        The cross-field transport of electrons/ions across magnetic field is fundamentally important as it determinines the characteristics of plasma wetted area in the scrape of layer region and particle confinement in magnetically confined plasma devices. The electrically biased objects in the edge region inside tokamaks as well as in Linear plasma devices are known to influence the dynamics of charge particles. The external electrodes in magnetized column can introduce long range electric fields in the plasma column. This leads to either excitation/ suppression of the instabilities resonsible for such transport. In this paper we present experimental results on radial plasma characteristics obtained of a cylindrical plasma column produced in a Linear Device. The magnetized plasma column at one end is terminated with conducting electrodes which are deliberately biased with respct to the plasma. The nature of the long range perturbation during application of electric bias on the electrodes have been investigated using electric probes and its impact on the radial characteristics have been qualitatively explained.
        Speaker: Mr Satadal Das (Institute For Plasma Research)
      • 207
        Recent finding in fusion studies using table top and miniaturized dense plasma focus devices operating from hundred joules to less than one joule
        In a dense plasma focus (DPF) the plasma is compressed into a hot-warm dense pinch. Since, 50 years ago and during the first three decades the dense plasma focus (DPF) was studied as a possible device to produce dense transient plasmas for fusion research. The trend was to produce bigger devices over MJ stored energy and MA current through the plasma pinch, in order to increase the efficiency of fusion neutron production. Unfortunately, the neutron production suffers saturation in devices operating at MA. Alternatively, in our group we have been studying how scale a dense plasma focus to very low energy of operation, keeping the nuclear fusion reactions and neutron emission. Several dense plasma focus devices under kJ stored energy (400J, 50J, 2J, and 0.1J) were designed and constructed in our laboratory. In all of them nuclear fusion reactions are obtained. In fact, recently we reported the evidence of nuclear fusion in a plasma focus operating in deuterium at only 0.1J. Despite these devices are far to produce net energy, these studies have contributed to learn that it is possible to scale the plasma focus in a wide range of energies and sizes keeping the same value of ion density, magnetic field, plasma sheath velocity, Alfvén speed and temperature. However, the plasma stability depends on the size and energy of the device. Recent findings related nuclear fusion studies are presented, including: a) evidence of nuclear fusion in an ultraminiaturized plasma focus operating at 0.1J; b) observations of plasma filaments and its role in the neutron emission; c) characterization of the plasma ejected after the pinch in table top and small DPF devices (50J, 400J and 900J) and their use to study the effects on materials relevant to the first wall of fusion reactors; and d) studies of the plasma interacting with a target material on front of the anode using digital optical refractive diagnostics and visible spectroscopy. In addition, how to increase the current in the pinch plasma, increasing the number of fusion nuclear reactions and neutron production, in a regime of enhanced stability is discussed. Supported by ACT-172101 CONICYT and FONDECYT 1151471 Chile grants.
        Speaker: Prof. Leopoldo Soto (Chilean Nuclear Energy Commission)
      • 208
        Reconstruction of MHD modes for energetic particle dynamics studies in toroidal equilibria with arbitrary q profiles
        The interaction of energetic particles with MHD modes of different types is a major concern for the next generation of experiments involving burning plasmas. This issue arises in different contexts such as particle redistribution due to current driven instabilities (involving or not magnetic reconnection), activation of Alfven eigenmodes (AE) due to wave-particle interaction or loss of confinement caused by neoclassical tearing modes (NTM). The physics involved in these processes is varied and complex. However, the construction of adequate models to study particle redistribution is usually simplified by assuming that the modes affect the particle dynamics through the perturbation of the equilibrium fields. Thus, the knowledge of the total field, equilibrium plus perturbation, produced in each case enables the calculation of the particle redistribution. In previous works, a model employing a fixed equilibrium and internal modes reconstructed from experimental data was developed and successfully applied to study alpha particle redistribution in the presence of kink modes and sawteeth with partial reconnection. To be able to tackle a larger number of problems, in this work, we extend the method to allow for the use of MHD equilibria with arbitrary safety factor (q) profiles. Again, external data either from experiments or simulations may be incorporated to estimate the structure of the modes. The resulting model is flexible and can be employed to study the effect of MHD modes on test particles in a variety of situations. As a first example, the redistribution of energetic particles caused by the sawtooth crash is considered. Several scenarios are investigated including full and partial reconnection in usual tokamak equilibria as well as configurations with an extended region of low magnetic shear at the plasma core.
        Speaker: Dr Pablo Garcia-Martinez (CONICET - Centro Atomico Bariloche)
      • 209
        Self-consistent gyrokinetic description of the interaction between Alfven modes and turbulence
        It is getting increasingly clear that many tokamak plasma phenomena which have traditionally been investigated separately, are actually intrinsically linked. One outstanding example along these lines - which is investigated in the present contribution - is the the interaction between Alfven modes (AM), turbulence, and zonal structures (ZS), like zonal flows and geodesic acoustic modes. Recently, a strong interest was raised in the fusion community by the possibility of generating ZS via nonlinear interaction with global modes like Alfven instabilities. In this work, the interaction of AM, turbulence and ZS is studied with the code ORB5. This model treats ions and electrons respectively as gyrokinetic and driftkinetic. ORB5 is a nonlinear global particle-in-cell code, developed for turbulence studies [1] and extended to its electromagnetic multi-species version [2] for the investigation of Alfven dynamics [3]. Recently, the importance of the kinetic electron effects in the ZS dynamics has also been emphasized with ORB5 [4]. ORB5 has also accomplished a verification/benchmark phase for AMs and has been used for the study of the nonlinear wave-particle interaction [5]. The competition between the different excitation mechanisms of ZS is the main focus of this work. When an EP population is added to the electromagnetic turbulence, the perturbed saturated field is observed to be modified by the presence of AMs. The effect of the different players are described separately, and in particular: wave-particle nonlinearity, wave-wave nonlinearity, effect of turbulence on AMs, effect of AMs on turbulence, for example via ZS generation, and bulk plasma omega-star effects on the AM growth rate and saturation. Comparisons with analytical theory and other models like the gyrokinetic Eulerian code GENE [6,7] are also done. [1] Jolliet S., et al. 2007, Comput. Phys. Comm. 177, 409 [2] Bottino A., et al. 2011, Plasma Phys. Controlled Fusion 53, 124027 [3] Biancalani A., et al. 2016, IAEA Fusion Energy Conference, Kyoto, Japan, TH/4-2 [4] Novikau I., et al. 2017, Phys. Plasmas 24, 122117 [5] Cole M. D. J., et al. 2017, Phys. Plasmas 24, 022508 [6] Jenko F. et al. 2000, Phys. Plasmas 7, 1904 [7] Goerler, T. et al. 2011, J. Comput. Phys. 230, 7053
        Speaker: Dr Alessandro Biancalani (Max-Planck-Institut für Plasmaphysik)
      • 210
        Simulation of Toroidicity-Induced Alfven Eignenmode Excited by Energetic Ions in HL-2A Tokamak Plasmas
        The toroidicity-induced Alfven eigenmode (TAE) excited by energetic ions was first simulated by using GTC code based on HL-2A experimental configuration. The simulation results show that the fraction of energetic (fast) ions in HL-2A experiments is about 3%. The TAE eigenmode frequency is around 211 kHz and is inversely proportional to the square root of electron density, which is quantitatively in agreement with the experimental observation. The real frequency of TAE modes increases with both temperature of energetic ions (beam energy) and toroidal model numbers increasing thanks to the toroidal precession resonance is dominant, but almost keeps constant when the density of energetic ions changes. The growth rates of TAE modes increase with increasing density as well as density gradient of fast ions. The amplitude of the vector potential A// exponentially increases with time for linear TAE mode. Besides, the low n (toroidal mode number) TAE modes, such as n=1 can also be driven by energetic ions when off-axis heating with higher beam energy is employed during HL-2A NBI experiment. The half width of radial mode structures for low n modes is usually wider than those for high n modes. The perpendicular wave vector of the TAE modes and Larmor radius of ions satisfy the relation . At the same time, the polarization of the TAE mode shows that the perturbed parallel electric field is zero. Thus, the TAE mode is close to an ideal MHD mode.
        Speaker: Dr Hongda He (Southwestern institute of physics)
      • 211
        Simulation Study of Heat Transport with On-Off Axis ICRH in Thailand Tokamak Using BALDUR Code
        Self-consistent simulations of plasma in a proposed tokamak design of Thailand Tokamak. (major radius = 65 cm, minor radius = 20 cm, plasma current = 100 kA, toroidal magnetic field = 1.5 T) are carried out using the 1.5D BALDUR integrated predictive modeling code. The simulations are used to investigate plasma transport with on and off axis positions of ion cyclotron resonance heating (ICRH) in the range of 0.3 – 5 MW. The core transport is predicted using the combination of Multimode (MMM95) or Mixed Bohm/gyro-Bohm (Mixed B/gB) anomalous core transport model and NCLASS neoclassical transport model. It is found that the electron temperatures obtained from both simulations are in the range of 0.3 - 1 keV which agree with the HT-6M experimental results. When the ICRH is applied, ion and electron thermal transport increase. Consequently, ion and electron temperature and plasma stored energy increase. During ICRH for both MMM95 and Mixed B/gB model, the electron temperature at the center (Te(0)) ranges from 1 to 1.5 keV with on axis and from 1 to 1.9 keV with off axis. The ion temperature at the center (Ti(0)) ranges from 0.7 to 25 keV with on axis and 50 eV to 7 keV with off axis.
        Speaker: Ms Jiraporn Promping (Thailand Institute of Nuclear Technology)
      • 212
        Simulations of the Sawtooth-Induced Redistribution of Fast Ions in JET and ITER
        Results of simulations of the sawtooth-induced redistribution of fast ions in JET and ITER with the code OFSEF are presented. The dependence of the redistribution on the particle parameters (energy and pitch angle) is studied. The redistribution of the trapped and marginally passing particles is found to exhibit barrier-like behaviour at the separatrix between the trapped and passing particles: the particles with high energies cannot pass the radial coordinate corresponding to the separatrix. The algorithm and structure of the rapid code developed on the basis of the OFSEF calculations are discussed. Simulations of the sawtooth effect on fusion alpha particles in ITER are carried out; they show that when the shape of the q-profile is non-parabolic (which is expected, for example, in the hybrid mode), the post-crash radial profile of the alpha particle distribution function can change significantly. Determining the parameters of a sawtooth crash --- the sawtooth mixing radius and the sawtooth crash duration --- from observations of the electron cyclotron emission in the equatorial plane of a tokamak is discussed; examples for JET sawtooth crashes are presented. Results of simulations of the sawtooth effect on the neutron emission in several recent JET discharges are presented. In most JET discharges, neutrons are mainly born by deuterons of the NBI (neutral beam injection) beam consisting mainly of passing particles with energies ~100 keV. However, in discharges with the third-harmonic ICRH (ion cyclotron resonance heating), a significant fraction of neutrons is produced by the ICRH tail of trapped deuterons in the MeV energy range, which provides an opportunity to verify the theory predictions.
        Speaker: Dr Yurii V. Yakovenko (Kyiv Institute for Nuclear Research, Kyiv, Ukraine)
      • 213
        Simulations of two types of energetic particle driven geodesic acoustic modes and the energy channeling in the Large Helical Device plasmas
        Energetic particle driven geodesic acoustic modes (EGAMs) in the Large Helical Device (LHD) plasmas are investigated using MEGA code. MEGA is a hybrid simulation code for energetic particles (EPs) interacting with a magnetohydrodynamic (MHD) fluid. In the present work, both the conventional and extended models of MEGA are employed. In the conventional model, only the EPs are described by the kinetic equations, while in the extended model not only the EPs but also the thermal ions are described by them. The simulations are conducted based on realistic parameters. The energy of neutral beam injection (NBI) is 170 keV. A Gaussian-type pitch angle distribution is assumed to model the NBI energetic ions. Using MEGA with a conventional model, it is found that the transition between low frequency EGAM and high frequency EGAM is decided by the slope of EP velocity distribution. Also, the phase difference between the bulk pressure perturbation $\rm \delta P_{bulk}$ and EP pressure perturbation $\rm \delta P_{EP}$ are analyzed. For the low frequency EGAMs, $\rm \delta P_{bulk}$ and $\rm \delta P_{EP}$ are in anti-phase. They cancel each other out, which reduces the restoring force of the oscillation leading to the low frequency. While for the high frequency EGAMs, $\rm \delta P_{bulk}$ and $\rm \delta P_{EP}$ are in the same phase. They enhance each other, and thus the frequencies are higher. Using MEGA with an extended model, the low frequency EGAMs are reproduced. The mode structure, mode number, and mode frequency are not only consistent with the results of conventional MEGA model but also consistent with theory and experiment. Also, the energy transfer of various species is analyzed and the bulk ion heating during the EGAM activity is observed. The ions obtain energy when the EPs lose energy, and this indicates that an energy channel is established by EGAM. The EGAM channeling is reproduced by simulation for the first time. From t = 0 to t = 0.36 ms, the energy transferred from EP is 63 J. About half of this energy (51%) is transferred to bulk ions (34%) and electrons (17%), while another half is dissipated. The heating power of bulk ions around t = 0.1 ms is $\rm 3.4~kW/m^3$ which is close to the value $\rm 4~kW/m^3$ evaluated from the experiments.
        Speaker: Dr Hao WANG (National Institute for Fusion Science)
      • 214
        The combined effect of neoclassical tearing modes and ELM control coils on fast-ions: validation in AUG and extrapolation for ITER
        This contribution aims to broaden the understanding of the interplay between the internal and external 3D perturbations on the fast ions in tokamak plasmas. At first, we used simulations using the ASCOT suite of codes to analyze an ASDEX Upgrade discharge showing clear sign of the interplay between a (3,2) neoclassical tearing mode (NTM) and external RMP coils on the fast ion loss detector (FILD) signal of neutral beam ion losses. At this context, also a code- code benchmark with the LOCUST code is presented. The same set of analysis tools is then used to predict both the alpha particle and neutral beam ion losses in the ITER 15 MA standard H-mode scenario in the presence of (2,1) and/or (3,2) NTM and ELM control coils (ECC). Magnetically confined fusion relies on that the fusion-born alpha particles will be well con- fined, thus providing significant plasma heating and keeping the first-wall intact. Recent nu- merical simulations indeed show that this is the case for most planned ITER scenarios [1]. However, these simulations were carried out assuming that the transport is fully neoclassical, and that the plasma is MHD-quiescent. Both of these assumptions should be relaxed before making the final verdict on the fast-ion confinement in ITER. In this contribution we partly relax the MHD-quiescence condition by adding NTMs in our simulations. Although a significant up to 100% increase in the total power losses for ITER was observed, so far no direct risk for the first wall was found. In this study both the NTM and the RMP perturbation was assumed to be static, thus maximizing the interaction between the two. Without further increased transport, by for example toroidal Alfven waves or turbulence, the fast ion power loads stay within the engineering limits.
        Speaker: Mr Antti Snicker (Aalto University)
      • 215
        Transport induced by energetic geodesic acoustic modes
        Energetic particles naturally exist in a tokamak due to either fusion reactions or external heating such as ICRH or NBI. These energetic particles need to be well-confined in order to transfer their energy to thermal particles and achieve this way a regime with self-sustained fusion reactions. However, energetic particles excite modes that tend to de-confine the particles themselves. This is the reason why energetic particle mode excitation and saturation need to be understood and controlled. We focus our analysis on a special class of energetic particle modes, called energetic geodesic acoustic modes (EGAMs). In this work, we present highly resolved full-f global gyro-kinetic 2-species simulations using GYSELA code that evidence the formation of chain of islands in phase space during the nonlinear saturation of EGAMs. Those islands appear at the predicted positions using linear and nonlinear wave-particle interaction theory. By means of a test-particle tracing method we solve the particle equations of motion using the self-consistent electrostatic potential obtained from 2-species GYSELA simulations and show that, even for weak fractions of energetic particles the EGAM island can interact with the trapping/de-trapping region characteristic of toroidal devices. In particular, counter-passing particles can be trapped and eventually de-confined, in agreement with experiments and with previous full-orbit particle simulations. Also, the nature of the transport induced by the energetic modes has been analysed. For this purpose, statistical analysis of 20000 counter-passing particles around the EGAM resonance has been performed. The variance of the particle displacement in phase space shows a super-ballistic transport. When the EGAM saturates the losses increase following also a power law and the transport becomes sub-diffusive.
        Speaker: Dr David Zarzoso (CNRS)
      • 216
        Tungsten control in NBI-dominant H-mode discharges in EAST tokamak
        In EAST tokamak, H-mode discharges have been obtained without a basic change at various heating conditions after installation of tungsten monoblocks at upper divertor. Recently, a reproducible long pulse H-mode operation with sufficient tungsten suppression has succeeded for both electron cyclotron resonance and lower-hybrid wave heated discharges and various experimental approaches are also attempted for the tungsten suppression. In discharges dominantly heated by NBI, however, the long pulse H-mode operation has been often restricted by appearance of the tungsten accumulation. Therefore, an exploration of experimental scenarios capable of avoiding the tungsten accumulation is urgently necessary for achievement of the long pulse H-mode discharge with NBI heated high-performance plasma. In the present work, control of the tungsten accumulation in the H-mode discharge with NBI-dominant heating is studied in EAST by measuring tungsten spectra and those radial profiles in extreme ultraviolet (EUV) range at 20-500Å. In order to control the tungsten accumulation in NBI H-mode discharges, experiments have been done by superimposing the LHW heating. One of the experiments is carried out by changing the 4.6GHz LHW power intermittently injected in the NBI H-mode discharge. When the LHW pulse is switched on, plasma particles immediately start to pump out. The tungsten concentration is largely reduced in the plasma core, while the tungsten concentration in the plasma outer region does not change so much. Similar behavior is also observed in the radiation loss. In addition, two-dimensional radiation distribution show that the tungsten accumulates at a very narrow region in plasma core (ρ<0.2) during the NBI phase and considerably flattens during the LHW pulse. These results clearly indicate a change in the tungsten transport in the NBI H-mode discharge. A series of experiments are completed by changing the LHW injection power in the NBI H-mode discharge. As a result, a sufficiently reduced tungsten concentration is obtained at P_{LHW}/P_{NBI} ~1.0, e.g. by an order of magnitude. The beneficial role of LHW injection observed for the first time in EAST is very similar to results of on-axis ECRH and ICRH in ASDEX-U and JET. The tungsten transport in the present experiment is being analyzed with a simulation code.
        Speaker: Dr Ling Zhang (CnIPPCAS)
      • 217
        Verification and Validation of Integrated Simulation of Energetic Particles in Toroidal Plasmas
        Energetic particle (EP) pressure gradients in fusion plasmas can readily excite mesoscale EP instabilities such as the Alfven eigenmodes (AEs) and energetic particle modes that drive large EP transport, which can degrade overall plasma confinement and threaten the machine’s integrity. EP could strongly influence thermal plasma dynamics including the microturbulence and macroscopic magnetohydrodynamic (MHD) modes. In return, microturbulence and MHD modes can affect EP confinement. We have developed first-principles capability for global integrated simulation of nonlinear interactions of multiple kinetic-MHD processes by treating both EP and thermal plasmas on the same footing. Verification and validation have been carried out for the gyrokinetic toroidal code (GTC) simulations of EP interactions with thermal plasmas in a DIII-D NBI-heated plasma. GTC kinetic-MHD simulations of EP interactions with thermal plasmas focus on the DIII-D discharge #159243, which is a NBI-heated plasma with many small-amplitude reversed shear Alfven eigenmodes (RSAE) and toroidal Alfven eigenmodes (TAE), significant flattening of the EP profile, and large amplitude microturbulence. GTC linear simulations using EFIT equilibrium and experimental profiles find that the most unstable AE is RSAE with significant growth rate for toroidal mode number n=3-6. The most unstable RSAE is n=4 and has a radial domain of ⍴=0.3 - 0.6 (square-root of normalized toroidal flux function). These results are in good agreement with other gyrokinetic and gyrokinetic MHD-hybrid codes, as well as experimental data. Consistent with experimental observation, GTC simulations also find that weaker TAE exist at the outer radial domain of ⍴=0.6 - 0.9. The most unstable TAE mode is n=5. Finally, GTC simulations find strong driftwave instability excited by thermal plasma pressure gradients in the core. The most unstable ion temperature gradient (ITG)-like mode is n=20. The linear ITG-like mode amplitude peak at ⍴=0.3, but large fluctuations nonlinearly spread to the whole radial domain. These results indicate that RSAE and TAE in this DIII-D experiment could interact nonlinear with each other and with the microturbulence.
        Speaker: Zhihong Lin (UC Irvine)
      • 218
        Kink Mode Study in EAST High β_{P} Plasma

        Two types of kink modes, fishbone and long-lived mode are experimentally and numerically studied at EAST tokamak. In high β_{P} plasma, sawtooth instability was replaced by a saturated 1/1 internal kink mode which either manifests itself as a periodical burst energetic ion related fishbone or as a long-lived mode which is associated to the core safety factor at q_0~1. The present of those 1/1 internal modes are beneficial to the sustain of hybrid scenario with extended regions of low-magnetic shear profile and q_0~1, because of that they can expel high-Z impurity and can make flux pumping. The mechanism responsible for the flux pumping caused by kink mode was numerically in nonlinear 3 D magnetohydrodynamic simulations using the M3D code. Furthermore, M3D+K code hybrid simulation shows a good agreement to the fishbone activity in EAST.

        Speaker: Dr Liqing Xu
    • 4:10 PM
      Coffee Break
    • OV/5 Overview Magnetic Fusion
      • 219
        Overview of HL-2A Recent Experiments
        Experiments on the HL-2A tokamak have been aimed at physics issues involved in advanced tokamaks and ITER since the last IAEA FEC. In particular, significant progresses have been made in the following areas: techniques and physics of ELM control, energetic-particle physics, MHD, disruption, multi-scale interactions, physics of advanced tokamak scenario, edge turbulence. Regarding to techniques and physics of ELM control, intensive experiments for controlling ELMs have been performed in HL-2A with several tools, including RMP, LHCD, LBO-seeded impurities (Al, Fe, W) and impurity SMBI (Ar, Ne). The observed ELM mitigation with pedestal turbulence enhancement and radial spectral shift due to the pedestal velocity shear reduction can be qualitatively simulated by a turbulent heat transport model. Toroidal Alfvén eigenmodes (TAE) driven by energetic-ion had been observed on HL-2A. Progress has been made in understanding the physics of instabilities that may interacts with turbulence causing strong influence on cross-field transport and in developing strategies to control them, including neo-classical tearing modes and core-localized Alfven eigenmodes. The stabilization of m/n=1/1 ion fishbone activities by ECRH were found on HL-2A. The experimental results confirmed the stabilization of m/n=1/1 fishbone depends not only on the injected power but also on the radial deposition location of ECRH. Disruption mitigation experiments with a new fast SMBI gas injection system have been recently performed. In HL-2A, advanced tokamak scenario with central q close to 1 was achieved. Auxiliary heating (mainly NBI) during the current rise phase was used, creating ITBs with a weak magnetic shear in the plasma centre. In ITB plasmas with weak magnetic shear, kinetic electromagnetic instabilities were confirmed and investigated. For the study of edge turbulence and flows, a signature of incoherent phase slips was evidenced by the study on the interaction between E×B shear and cross phase between radial velocity perturbation and poloidal ¬velocity perturbation. In the pedestal region, the dynamics of the plasma flows, turbulence and pedestal formation across the L-I-H transition were studied by Doppler reflectometry. The electromagnetic character of filamentary structure was measured in the scrape off layer of HL-2A for the first time.
        Speaker: Dr Min Xu (CnSWIP)
      • 220
        Physics research on the TCV tokamak facility: from conventional to alternative scenarios and beyond
        The research program of the TCV tokamak ranges from conventional to advanced tokamak scenarios and advanced divertor configurations, to exotic plasmas driven by theoretical insight, exploiting the device’s unique shaping capabilities. The facility is operated intensively both domestically and with EUROfusion support. The new 1-MW NBI has expanded the parameter range, now encompassing ELMy H-modes in an ITER-like shape, stationary non-inductive discharges sustained by ECCD and NBCD, and negative-triangularity diverted plasmas. Disruption avoidance by real-time locked mode prevention or unlocking with ECRH was thoroughly documented, using magnetic and radiation triggers. Runaway generation with high-Z noble gas injection and runaway dissipation by subsequent Ne or Ar injection were studied for model validation. Turbulence is reduced in the core at negative triangularity, consistent with increased confinement and in accord with global gyrokinetic simulations. The GAM, possibly coupled with avalanche events, has been linked with particle flow to the wall in diverted plasmas. In H-mode, the pedestal pressure and plasma stored energy are insensitive to fueling, whereas nitrogen seeding moves the pedestal outwards and increases the stored energy. High fueling at high triangularity (0.54) is key to accessing the attractive small-ELM (type-II) regime. Detachment, SOL transport, and turbulence were studied in L- and H-mode in both standard and alternative configurations (snowflake, super-X, and beyond). The L-H transition threshold is independent of the divertor topology. In the attached L-mode phase, an increase in flux expansion or divertor leg length reduces the power exhausted at the outer strike point and increases radiation. The detachment process is caused by power “starvation” reducing the ionization source, with volume recombination playing only a minor role. The SOL density shoulder observed at high collisionality is correlated with increased blob size. A doublet plasma, featuring an internal X-point, was achieved successfully, if only transiently, and a transport barrier was observed in the mantle just outside the internal separatrix. In the near future variable-configuration baffles and cryopumping will be introduced to investigate the effect of divertor closure on exhaust and performance, and 2-MW ECRH and 1-MW NBI heating will be added.
        Speaker: Dr Stefano Coda (CRPP-EPFL)
      • 221
        Overview of Operation and Experiments in the ADITYA-U Tokamak
        Ohmically heated circular limiter tokamak, ADITYA has been upgraded to a tokamak named ADITYA Upgrade (ADITYA-U) having open divertor configuration with divertor plates. Experiment research in ADITYA-U (R0 = 75 cm, a = 25 cm) has made significant progress, since last FEC 2016. After successful commissioning of ADITYA-U, the Phase-I plasma operations have been conducted from December 2016, with graphite toroidal belt limiter. Filament pre-ionization assisted purely Ohmic discharges with circular plasma have been obtained. Hydrogen gas breakdown has been obtained in each of ~ 700 discharges without a single failure. Repeatable plasma discharges of plasma current ~ 80 kA – 95 kA, duration ~ 80 – 100 ms with toroidal magnetic field (max.) ~ 1T and chord-averaged electron density ~ 2.5 x 10^19 m^-3 has been achieved. Later, the discharge duration has been enhanced up to ~ 180 ms with the application of negative converter along with better wall conditioning, achieved by implementing the Glow Discharge Cleaning (GDC) with Ar: H2, He: H2 gas mixture and with intense short plasma pulses in ECR produced plasma background. Being a medium sized tokamak, runaway electron generation, transport and mitigation experiments have always been one of the prime focus of ADITYA-U. MHD activities and density enhancement with H2 gas puffing has also studied. The Phase-I operation was successfully completed in March 2017. The Phase-II operation preparation in ADITYA-U includes, calibration of magnetic diagnostics followed by commissioning of major diagnostics and installation of baking systems. After repeated cycles of baking the vacuum vessel up to ~ 130°C, the ADITYA-U Phase-II operations have been resumed from February 2018 and is continuing in order to achieve plasma parameters close to the design parameters of circular limiter plasmas using real time plasma position control. Several experiments, including the fueling with Supersonic Molecular Beam Injection, H2 gas puffing for runaway control during current flat-top and disruptions, Neon gas puff assisted radiative improved confinement and the experiments related to plasma shaping is undergoing. The complete upgradation including dismantling of ADITYA and reassembling of ADITYA-U along with experimental results of Phase-I and Phase-II operations from ADITYA-U and overall progress will be discussed in this paper.
        Speaker: Mr Rakesh Tanna (Institute For Plasma Research)
      • 222
        Tokamak research in Ioffe Institute
        Research of various aspects of tokamak physics is conducted on small tokamaks at Ioffe Insitute in a wide range of experimental conditions: R/a=1.6, Bt=0.5(1.0) T, Ip=250(500) kA – Globus-M(M2), R/a=2.4, Bt=1.0 T, Ip=150 kA – TUMAN-3M, R/a=7.0, Bt=3.0 T, Ip=25 kA – FT-2 tokamaks. Results obtained in final Globus-M experimental campaign (before upgrade shutdown) with the 25% toroidal magnetic field and plasma current increase up to 0.5 T and 250 kA respectively are presented. In these experiments an overall improvement in plasma performance was observed. Energy confinement time study was performed in both OH and NBI heated H-mode plasma. Strong tau_E dependence on both Ip and Bt was observed, while the dependence on density and absorbed power was similar to the conventional H-mode scaling IPB98(y,2). The lifetime of modes with ITB reached a few confinement times before the q=1 resonant surface appeared in the plasma. Plasma confinement was also studied in the compact TUMAN-3M tokamak. No noticeable isotope effect in particle confinement in hydrogen and deuterium ohmic L-mode was observed. On the contrary, in the ohmic H-mode particle confinement was approximately 1.5 times higher in deuterium than in hydrogen. Study of TAEs on Globus-M was performed at increased magnetic field. The mode character and influence on the fast ions changed with the increase of the Bt and Ip. At TUMAN-3M Ion Cyclotron Emission in OH and NBI heated discharges was studied. Application of the NBI revealed central location of ICE, excitation by sub-Alfvénic beam ions and fine structure of the emission spectral lines. New diagnostics, designed for Globus-M2, were installed and tested on Globus-M. At the FT-2 tokamak the ELMFIRE global gyrokinetic modeling of the OH discharge is compared to the experimental data using the specially developed fast linear version of the X-mode DR synthetic diagnostics. The anomalous absorption of the pump wave in the ECRH experiments due to the parametric excitation of trapped UH waves in the vicinity of the density or magnetic field profile local maximum is considered.
        Speaker: Dr Nikolai Bakharev (Ioffe Institute)
      • 223
        Overview of recent progress in understanding NSTX and NSTX-U plasmas & Overview of new MAST physics in anticipation of first results from MAST Upgrade

        A. The mission of the spherical tokamak NSTX-U is to explore the physics that drives core and pedestal transport and stability at high-β and low collisionality, as part of the development of the ST concept towards a compact, low-cost ST-based Pilot Plant. NSTX-U will operate at up to 2 MA and 1 T with up to 10 MW of Neutral Beam Injection (NBI) power for 5 seconds with up to 4 MW of High Harmonic Fast Wave (HHFW) power. In this parameter space, electromagnetic instabilities are expected to dominate transport. Furthermore, beam-heated NSTX-U plasmas will be able to explore the energetic particle (EP) phase space that is relevant for both α-heated conventional and low aspect ratio burning plasmas. A further objective is to develop the physics understanding and control tools to ramp-up and sustain high performance plasmas in a fully-noninductive fashion for pulse lengths up to 5 s. NSTX-U began research operations in 2016, producing 10 weeks of commissioning and scientific results. However, a number of technical issues, including the failure of a key divertor magnetic field coil, resulted in the suspension of operations and initiation of Recovery activities. During the Recovery outage, there has been considerable work in the area of analysis, theory and modeling with a goal of understanding the underlying physics to develop predictive models that can be used for high-confidence projections for both ST and higher aspect ratio regimes. The studies have addressed issues in thermal plasma transport, indicating the importance of non-local and multi-scale effects, EP-driven instabilities at ion-cyclotron frequencies and below, studying the wave-particle interactions and development of descriptive predictive models, and heat flux width modeling and the role of turbulence broadening. NSTX-U is expected to resume operations during CY2020.

        This work was supported by US Department of Energy Contract No. DE-AC02-09CH11466

        B. MAST Upgrade will operate in 2018 with unique capabilities to explore plasma exhaust and alternative divertor configurations to address this key issue for DEMO. Modelling of the interaction between filaments with BOUT++ indicates filaments separated by more than 5x their width move independently, and their velocity is slightly perturbed by if their separation is 1 width, suggesting radial density profiles can be modelled as the superposition of filaments. Secondary filaments on MAST are found up to 1ms after type-I ELMs that correlate with plasma interaction with surfaces near the X-point. A quiescent region devoid of filaments near the X-point has been routinely observed, extending from the separatrix to a normalised flux of 1.02. Counter-streaming flows of doubly ionised carbon along field lines, generated by localised gas puffing, have been observed and reproduced in EMC3-EIRENE simulations. MAST-U will be an excellent facility for understanding detachment onset and control in closed divertors. SOLPS modelling predicts the upstream density needed to reach detachment will be over 2x lower in the Super-X configuration compared with the conventional divertor due to increased total magnetic flux expansion. Analytic modelling predicts detachment control in a Super-X is more amenable to external control. Detailed measurements of transport through the edge have been made in MAST L-mode plasmas to characterise a Geodesic Acoustic Mode 2cm from the separatrix. Interpretation of plasma potential profile measurements using ball-pen probes have been improved through kinetic modelling, showing that electrons polarise the material around the probe, leading to ExB drifts of ions to the probe.
        Measurements of the effects of sawteeth on fast ion confinement on MAST indicate that passing and trapped particles are equally redistributed by the sawtooth crash. There is no apparent energy threshold for redistribution, indicating redistribution due to a mechanism resonant with the m=1 perturbation. Gyrokinetic simulations of ETG turbulence in MAST are in close agreement with the measured collisionality dependence of the energy confinement time. Beam emission spectroscopy measurements show that flow shear leads to eddy tilting in up-down symmetric plasmas and skewed density fluctuations. First results from MAST Upgrade operations will be presented.

        Speaker: Dr Jonathan Menard
      • 224
        Overview of new MAST physics in anticipation of first results from MAST Upgrade

        MAST Upgrade will operate in 2018 with unique capabilities to explore plasma exhaust and alternative divertor configurations to address this key issue for DEMO.
        Modelling of the interaction between filaments with BOUT++ indicates filaments separated by more than 5x their width move independently, and their velocity is slightly perturbed by if their separation is 1 width, suggesting radial density profiles can be modelled as the superposition of filaments. Secondary filaments on MAST are found up to 1ms after type-I ELMs that correlate with plasma interaction with surfaces near the X-point. A quiescent region devoid of filaments near the X-point has been routinely observed, extending from the separatrix to a normalised flux of 1.02. Counter-streaming flows of doubly ionised carbon along field lines, generated by localised gas puffing, have been observed and reproduced in EMC3-EIRENE simulations. MAST-U will be an excellent facility for understanding detachment onset and control in closed divertors. SOLPS modelling predicts the upstream density needed to reach detachment will be over 2x lower in the Super-X configuration compared with the conventional divertor due to increased total magnetic flux expansion. Analytic modelling predicts detachment control in a Super-X is more amenable to external control.
        Detailed measurements of transport through the edge have been made in MAST L-mode plasmas to characterise a Geodesic Acoustic Mode 2cm from the separatrix. Interpretation of plasma potential profile measurements using ball-pen probes have been improved through kinetic modelling, showing that electrons polarise the material around the probe, leading to ExB drifts of ions to the probe.
        Measurements of the effects of sawteeth on fast ion confinement on MAST indicate that passing and trapped particles are equally redistributed by the sawtooth crash. There is no apparent energy threshold for redistribution, indicating redistribution due to a mechanism resonant with the m=1 perturbation.
        Gyrokinetic simulations of ETG turbulence in MAST are in close agreement with the measured collisionality dependence of the energy confinement time. Beam emission spectroscopy measurements show that flow shear leads to eddy tilting in up-down symmetric plasmas and skewed density fluctuations. First results from MAST Upgrade operations will be presented.

        Speaker: Dr Jonathan Menard
    • Nuclear Fusion Board Meeting
    • IFE/1 Inertial Fusion Experiments & Theory
      • 225
        Two-colors mixed petawatt laser designed for fast ignition experiment

        Here we report a novel design of a heating laser for the fast ignition, combining fundamental and second harmonics lights. Such a two-colors laser is expected to heat a dense core more efficiently than a laser only with a fundamental light. We chose a LBO (LiB3O5) crystal which can convert a focusing beam due to its large acceptance of phase matching angle. We experimentally demonstrated the second harmonic conversion with efficiency of 60% at the maximum. The LBO crystal shows a high damage threshold more than 5 J/cm2 with a down-scale LFEX beams. A full size (10 cm×10 cm×2 mm) LBO crystal was manufactured completely and is ready to install for the full-scale LFEX operation.

        Speaker: Dr Yasunobu Arikawa (Insituteof Laser Engineering Osaka University)
      • 226
        Production of keV-Temperature Plasma Core with Magnetized Fast Isochoric Heating
        The quest for the inertial confinement fusion (ICF) ignition is a grand challenge, as exemplified by extraordinary large laser facilities like National Ignition Facility (NIF) [J. Lindl et al., Phys. Plasmas 11, 339 (2004), J. Lindl et al., Phys. Plasmas 21, 020501 (2014)]. Although scientific break-even, the energy released by fusion reaction exceeds the energy contains in the compressed fusion fuel, was achieved on NIF [O. A. Hurricane et al., Nature 506, 343 (2014)], the pathway to the ignition is still unclear. Fast isochoric heating, also known as fast ignition, of a pre-compressed fuel core with a high-intensity laser is an attractive and alternative approach to the ICF ignition [M. Tabak et al., Phys. Plasmas 1, 1637 (1994)] that avoids the ignition quench caused by the hot spark mixing with the cold fuel, which is the crucial problem of the currently pursued ignition scheme. High-intensity laser-plasma interactions efficiently produce relativistic electron beams (REB). However, only a small portion of the REB collides with the core because of its large divergence. Here we have demonstrated enhanced laser-to-core coupling with a magnetized method to confine the REB in a narrow transport region resulting in efficient isochoric heating. The method employs a laser-produced kilo-tesla-level magnetic field [S. Fujioka et al., Sci. Rep., 3, 1170 (2013)] that is applied to the transport region from the REB generation point to the core which results in guiding the REB along the magnetic field lines. We have created successfully a 1.6 ± 0.2 keV-temperature plasma core having 1 Gbar of energy density by using the MFI scheme with 7.7 ± 1.3% of an efficient laser-to-core energy coupling [S. Sakata et al., ArXiv 172.06029 (2017)]. We should emphasize that our result can be explained by a simple model coupled with the comprehensive plasma diagnostics, while several ICF experiments relay heavily on computer simulations due to difficulties of diagnosing micro-scale phenomena occurred in the small and complex plasma. The simplicity may secure scalability of this scheme to the ignition. 15% of the laser-to-core coupling is achievable for an ignition-scale high area density core (0.3 - 0.5 g/cm2) according to the model. The ignition target based on the MFI scheme is being designed by using multi-scale and multi-dimensional simulations.
        Speaker: Prof. Shinsuke Fujioka (Institute of Laser Engineering, Osaka University)
      • 227
        Liquid DT Layer Approach to Inertial Confinement Fusion

        The baseline approach to high gain ICF involves the implosion of capsules containing a layer of DT ice [1]. DT ice layer designs require a high convergence ratio (CR > 30) implosion, with a hot spot that is dynamically created from DT mass originally residing in a thin layer at the inner DT ice surface. Although high CR is desirable in an idealized 1D sense, it amplifies the deleterious effects of realistic features and asymmetries [2]. An alternative ICF concept uses DT liquid layers [3]. DT liquid layers allow for much higher vapor densities than are possible with DT ice layers. The wide range of vapor densities that are possible with DT liquid layers provides flexibility in hot-spot CR (12 < CR < 25), which, in turn, will provide a reduced sensitivity to asymmetries and instability growth. Given enough vapor mass, the hot spot can be formed from the mass originally residing in the central vapor region. Recent experiments at the National Ignition Facility (NIF) have demonstrated cryogenic liquid DT layer ICF implosions, along with the associated flexibility in the hot spot CR [4,5].
        There are tradeoffs involved in high CR ice layer and reduced CR liquid layer designs. With reduced CR, hot spot formation is expected to have improved robustness to instabilities and asymmetries [2-5]. In addition, the hot spot pressure (Pr) required for self-heating is reduced if the hot spot radius (Rhs) is increased (Pr α Rhs^-1). With a reduction in the hot spot Pr requirement, the implosion velocity and fuel adiabat requirements are relaxed. On the other hand, with larger hot spot size, the hot spot energy requirement for self-heating (Ehs) is increased (Ehs α Rhs^2), and the required capsule absorbed energy is increased. In this presentation, we will summarize the recent liquid layer experiments at the NIF and will discuss the hot spot energy, hot spot pressure, cold fuel adiabat, and capsule-absorbed energy requirements for achieving self-heating and propagating burn using liquid layer capsules with hot spot CR<20.

        [1] S. W. Haan et al., Phys. Plasmas 18, 051001 (2011).
        [2] B. M. Haines et al., Phys. Plasmas 24, 072709 (2017).
        [3] R. E. Olson and R. J. Leeper, Phys. Plasmas 20, 092705 (2013).
        [4] R. E. Olson et al., Phys. Rev. Lett. 117, 245001 (2016).
        [5] A. B. Zylstra el al., Phys. Plasmas, to be published (2018).

        Speaker: Dr Ray Leeper (Los Alamos National Lab)
      • 228
        Thermo-mechanical and Atomistic Assessment of First Wall and Optics in non-protective chamber in Inertial Fusion Energy

        Different Inertial Fusion Energy (IFE) First Wall (FW) protections have been proposed in diverse conceptual designs that lead to very different irradiation conditions and macroscopic effects. A review is needed to understand their behavior. Some years ago a European proposal projected the possibility of non-protective FWs considering W and nano-tungsten. This work is describing in detail the behavior of a W and nano-tungsten first wall under pulsed irradiation conditions predicted for the different operational scenarios of that laser fusion project by using advanced engineering modeling tools. Starting with the calculations of the time-dependent pulsed radiation fluxes, assuming 3D geometrical configurations, we estimate the irradiation-induced evolution of first wall temperature as well as, the thermo-mechanical response of the material. Finally, we carry out crack propagation calculations. Results allow us to define operational windows and to identify the main limitations for operation. The atomistic effects of irradiation in the FW are the other key magnitude to determine available lifetime. The role of grain boundaries on the radiation-induced damage and light species behavior is studied both experimentally and computationally, also under pulsed conditions. Important differences are observed in the density of vacancies between nanostructured and coarse-grained samples as well as the preferential places for H accumulation concluding with the influence of temperature.
        Optics damage is a great concern in IFE; a new full conceptual final focusing system based on silica transmission lenses for dry wall chambers was designed assuming pulsed conditions based on a temperature control system by using a heat transfer fluid. Optical response of composite materials containing metal nanoparticles was investigated and optimized. Highly concentrated silver colloidal nanoparticle solutions were produced thanks to fs laser ablation and it was demonstrated that such embedded plasmonic nanoparticles may be viable candidates to reduce damages produced on optics by swift heavy ions due to the change of their shape under irradiation.

        Speaker: Prof. Jose Manuel Perlado (Instituto Fusion Nuclear / Universidad Politécnica Madrid)
      • 229
        Demonstrations of foam shell and infrared heating methods for FIREX targets
        We study fuel layering for Fast Ignition Realization Experiment (FIREX) cryogenic targets according to two strategies: a foam shell and Infrared (IR) heating. Foam is a porous material and would soak up a liquid fuel uniformly by capillarity. The method has the difficulty to form void-less solid fuel because of the density difference between the liquid and solid phases. We have demonstrated the residual void fraction of ~1 % in a foam wedge. ANSYS simulations have represented that the technique would be applicable to a FIREX target. We examine the simulated process using a dummy foam shell target and succeed to form an ice layer with a reduced void fraction. The IR heating technique has originally been developed for central ignition targets, which requires spherical symmetry. We modify it for an axisymmetric FIREX target. We have developed the dedicated layering system with additional temperature control of the cone. To date, the sphericity of a formed ice layer reaches 95 %.
        Speaker: Dr Akifumi Iwamoto (National Institute for Fusion Science)
    • P3 Posters
      • 230
        2D and 3D modelling of JT-60SA for disruptions and plasma start-up
        The JT-60SA is a superconducting tokamak device being built as a joint international project between Japan and Europe in the frame of the broader Approach agreement. One of the main goals of JT-60SA is to study practical and reliable plasma control schemes in view of the power plant. Plasma electromagnetic modelling is one of the essential tools for plasma operation in a fusion device and they require detailed models for ensuring an accurate preparation of the magnetic controllers. To achieve this goal, suitable models are needed at different level of details. 2D plasma nonlinear equilibrium codes are used to develop the operational scenarios and to perform breakdown studies. Furthermore, three-dimensional modelling permits the assessment of 3D vessel structures on the plasma behaviour, e.g. during disruptions, as well as to study non-axisymmetric plasma instabilities. On the other hand, engineering-oriented models are essential for the commissioning of the magnetic diagnostics, and the design of control algorithms. In this context, a set of alternative modelling tools based on the CREATE 2D equilibrium codes have been developed as additional benchmark for magnetic modelling. These tools have been exploited to perform breakdown studies and to design a preliminary functional architecture of the plasma magnetic control system. Furthermore, several studies of the impact of three-dimensional structures on plasma evolution have been carried out, ranging from pure electromagnetic analysis of the magnetic field produced by the non-axisymmetric coils, to nonlinear evolution of n=0 instabilities. In this paper, we report on the activities that have been carried out exploiting the CREATE modeling tools. In particular, 2D modelling has been exploited to study the magnetic configurations for the EC assisted breakdown, while 3D tools have been used to evaluate the effect of three-dimensional structures on evolutionary equilibrium of axisymmetric plasmas.
        Speaker: Prof. Gianmaria De Tommasi (Università degli Studi di Napoli Federico II)
      • 231
        3MW Dual Output High Voltage Power Supply Operation: Results for Accuracy, Stability and Protection Test
        High temperatures inside tokamak for fusion research is achieved from auxiliary heating systems like neutral beam injectors (NBI), or RF heating devices, viz., ion cyclotron (IC), electron cyclotron and lower hybrid systems where High Voltage Power Supply (HVPS) is an essential requirement. ITER requires 20 MW of ICRF for heating and driving plasma current. A cascaded chain of amplifier is a practical solution due to limiting level of power with available vacuum tubes. Each chain of amplifier has to provide 1.5MW power in frequency range of 35- 65 MHz for 3600 seconds. The system must be capable to operate both at matched and mismatched load condition (VSWR 2) [1]. HVPS based on pulse step modulation (PSM) topology has already demonstrated its ability for broadcast transmitters, accelerators using radio frequency (RF) source and neutral beam injectors. A novel concept of tapping two outputs from single PSM based HVPS is attempted for the first time. A PSM based HVPS is developed with dual output to feed driver and end stages of a high power RF amplifier [2]. Developed dual output HVPS is capable of providing 14 - 18 kV, 250 kW to driver stage and 16-27 kV, 2800 kW to end stage of a RF amplifier chain, simultaneously [3]. Present article covers the validation of dual output HVPS for integrated operation with RF Amplifier system. It includes wire burn test conducted at the output of HVPS, demonstrating tight synchronization among both stages. Test set up, gauge/length for fuse wire to meet the critical energy limit qualifications is presented. HVPS performance parameters viz. ripple, regulation and stability over extended duration of 3600 seconds are presented for various scenario of RF Amplifier operation. Implemented scheme for protection against overvoltage and overcurrent is also discussed. [1]Aparajita Mukherjee et al., “Progress in High Power Test of R&D Source for ITER ICRF system”, unpublished, FEC 2016. [2]A.Patel et al., “Development of 3 MW Dual Output High Voltage Power Supply for ICRH System”, International Power Modulator and High Voltage conference (IPMHVC-2016), San-Francisco, July 5-9, 2016 [3]A.Patel et al., “Initial operation of 3 MW dual output high voltage power supply with IC RF system”, Fusion Engineering and Design, Volume 126, January 2018, Pages 59–66.
      • 232
        A Concept of Self-Cooled Breeding Blanket with Advanced Molten Salt Flinak for High-Efficiency and Long-Life Operation

        An advanced molten salt (AMS), in which powders of hydrogen-soluble and chemically reactive metals such as titanium are mixed, is investigated as a potential self-cooled breeding blanket material. It is shown that hydrogen isotope uptake in a vanadium plate in molten salt FLiNaK is suppressed by the addition of Ti powders into the salt. In addition, the corrosion of candidate structural materials in FLiNaK with HF is also suppressed by the addition of titanium powders. Considering these result, tritium formed in the molten salt in fusion blanket will be trapped by the Ti powders, not being trapped by the structure materials (vanadium alloy) and not corroding the structure materials. Neutronics and tritium mass balance calculations are also performed and it is showed that FLiNaK based Be-free blanket is feasible.

        Speaker: Mr Juro Yagi (Institute of Advanced Energy, Kyoto University)
      • 233
        A Multi-Parameter Optimization technique considering temporal and spatial variation in nuclear response of materials in Fusion devices
        Structural materials present in and round any fusion device will face stringent conditions due to the high-energy, high-intensity neutron emitted from the fusion plasma. This will have significant life-limiting impacts on the reactor components of both experimental and commercial fusion devices. The neutrons interact with the material initiating nuclear reaction leading to the production of radioactive isotopes, gas molecules and related defects. These gases, particularly helium, can cause swelling and embrittlement of the material. Furthermore, the radioactive isotopes produce would cause heating in the material. These isotopes may have long lives which would contribute towards the radwaste produced in the fusion devices. Hence designing of low activation materials for fusion devices is warranted. At Iter-India, Institute for Plasma Research a number of computational tools are being developed to estimate the nuclear response of the materials and to optimize accordingly. ACTYS-1-GO, a multipoint neutron activation code which can calculate radiological responses of materials located at various positions in a fusion reactor efficiently is developed. Also, a mathematical framework is developed for accessing the relationship of radiological quantity with the initial elements present in the material. Such framework helps in identifying and minimizing the fraction of most dangerous elements/isotopes from the material composition. In the present study both the methodologies are efficiently coupled for a complete material optimization. Quantities responsible for various radiological effects (like activity, dose, heat, and radwaste) and related defects in the material are considered and their contributing elements are optimized accordingly. Also, since a single material faces a gradient of neutron flux over its entire volume, all such optimization is carried out over the entire range of neutron flux faced by that material. This provides a comprehensive picture of the response of the material to neutron irradiation, enabling the assessment of structural integrity of components in a fusion device.
        Speaker: Ms Priti Kanth (Institute for Plasma Research, HBNI)
      • 234
        Advanced capabilities of multi-functional calculation program SuperMC3.2 for complex nuclear system
        Super Multi-functional Calculation Program for Nuclear Design and Safety Evaluation, SuperMC, is a full-function neutronics simulation software system including inner-coupled calculations among efficient radiation transport, depletion, activation and shutdown dose. Its advanced capabilities include CAD/image-based accurate modeling for complex irregular geometry, intelligent data analysis based on multi-D/multi-style visualization and network collaborative nuclear analysis on cloud computing platform. Besides, several advanced radiation transport methods such as global weight window generator (GWWG) were proposed to solve the key problems for radiation protection in fusion system, such as deep penetration problem, sky scattering problem. SuperMC has been verified and validated by more than 2000 benchmark models and experiments including HCPB mock-up experiments in SINBAD, IAEA-Activation Calculation Benchmark (ACB), FNG-ITER SDR experiment and so on. And it was also applied in the neutronicis analysis of ITER, DEMO, etc.
        Speaker: Dr Lijuan Hao (Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031, China)
      • 235
        Application of ANSYS FLUENT MHD Code for Liquid Metal MHD Studies
        Magneto Hydro Dynamic (MHD) phenomena plays an important role in governing liquid metal flow characteristics under strong transverse magnetic field and has, therefore, gained the attention of fusion community for the design of liquid breeder blankets. In presence of plasma confining toroidal magnetic field, the flow of electrically conducting liquid metal (Li/Pb-Li), typically used for coolant and/or tritium carrier, is greatly affected due to flow opposing Lorentz force, which arises due to interaction between magnetic field and induced current in the liquid metal. For the successful design and development of liquid breeder blankets, detailed MHD analysis is highly desirable to understand various effects of MHD, such as change in velocity profile, pressure drop, heat transfer etc. The liquid metal MHD studies are being carried out using both analytical and numerical approaches. The analytical solutions, derived under 2D fully developed flow approximations, are limited to the simple flow geometries and hence they are not applicable for the analysis of complex blanket flow configuration, which consists of bends, transition zone, multichannel flow etc. Numerical simulation techniques are, therefore, used extensively to perform MHD analysis in such complex flow configuration and various MHD codes, either newly developed or commercially available are being reported. The MHD code, however, needs to be benchmarked extensively and validated before its application to complex flow configuration in liquid breeder blanket. In the present work, three MHD benchmark problems of ref. [1] has been successfully analyzed using ANSYS FLUENT MHD code and results are compared with available literature data. The selected problems are (i) 2D fully developed laminar steady MHD flow, (ii) 3D laminar, steady developing MHD flow in a non-uniform magnetic field and (iii) MHD flow with heat transfer (buoyant convection). The results have provided more confidence in using FLUENT as a promising MHD analysis tool for fusion application. The numerical model, analysis, methodology and simulation results of each benchmark problem will be discussed in detail. References: 1. S. Smolentsev, S. Badia , R. Bhattacharyay etal, FED 100 (2015) 65–72.
        Speaker: Mrs Anita Patel (Institute For Plasma Research)
      • 236
        Artificial Neural Network for Yield Strength Prediction of Irradiated RAFM Steels
        Structural materials to be used in proposed fusion reactor will be exposed to hostile neutronic environmental conditions. These steels will interact with high energy neutron particles. The interaction is expected to degrade structural material properties such as loss of ductility, increase of yield strength and DBTT temperature. Artificial neural network (ANN) with back-propagation (BPN) technique is used in this work to develop a numerical model which predicts the change in yield strength of irradiated steels at various irradiation condition. More than 15,000 material related parameters such as composition, temperature, yield strengths are obtained from literature. These experimental results are used to generate more than 100 networks after proper training, testing and validation. A statistically validated neural network is used to predict the change in yield strength of RAFM steel in the range of 290 K − 900 K and 0 − 80 DPA. For instance, at 673 K and 300 K of test temperature and irradiation temperature, the yield is first found to increase and then remain constant after 50 DPA. Again at the same test temperature and higher irradiation temperature of 700 K, the yield strength is first found to increase till 25 – 30 DPA and then decreases thereafter. In the work we plan to present such kind of behavior at different temperatures and DPA conditions.
        Speaker: Mr Agraj Abhishek (Institute for Plasma Research)
      • 237
        Characterization of Isotope Effect on Confinement of Dimensionally Similar NBI-Heated Plasmas in LHD
        Energy confinement and thermal transport has been widely regarded as gyro-Bohm in tokamak as well as stellarator-heliotron for a single kind of ion. However, this gyro-Bohm model predicts confinement degradation in deuterium (D) plasmas because of larger normalized gyro radius than in hydrogen (H) plasmas, which conflicts with major experimental observations. This study aims to quantify a peculiarity in dependence on normalized gyro radius in H and D plasmas in order to address this unresolved issue. The first deuterium plasma campaign in LHD reveals definitive characteristics of isotope effect on NBI-heated plasmas from elaborated experiments. Stationary uneventful plasmas, which are accompanied by neither ITB nor transition, have been assessed here. Thermal energy confinement time gives the regression expression scaling with the isotope mass (A) as A to 0.15, which shows moderate improvement in D plasmas. This positive isotope dependence contradicts with gyro-Bohm and is similar to the recent result from L-mode plasmas in JET-ILW. Operational flexibility of magnetic field, density, and heating power enables adjustment of three major non-dimensional parameters, those being normalized gyro radii, collisionality and beta , and dimensionally similar plasmas of H and D in all these three parameters can be obtained. Then TASK3D-a / FIT3D is used for analysis of heating power deposition, power balance and local thermal transport. If gyro-Bohm nature predominates in these plasmas, thermal diffusivity normalized by Bohm diffusion should be the same in a pair of dimensionally similar plasmas of H and D. Different characteristics have been found in electron and ion loss channels. Electron heat diffusivity normalized by Bohm diffusion in H is lower than that in D and even lower by a factor of 1 over square root of 2 which means net improvement. This trend is robust and insensitive to parameters such as normalized gyro radii, collisionality, beta, scale length of density gradient, etc. In contrast, ion thermal diffusivity shows a same characteristics as in the electron channel in low collsionality regime while that in D compared with the case with H degrades with the increase of collsionality. These results have shown definitively that the gyro-Bohm nature is violated in the comparison of H and D plasmas in a large scale stellarator-heliotron.
        Speaker: Prof. Hiroshi Yamada (National Institute for Fusion Science)
      • 238
        Conceptual design of Neutron Activation System for IN-LLCB TBM
        Neutron Activation System (NAS) is the primary neutron diagnostics for Indian Lead-Lithium Ceramic Breeder (LLCB) Test Blanket Module (TBM) in ITER. The main objective of NAS is to measure spatial distribution of neutron flux and energy spectra and in-situ measurement of tritium production rate inside the TBM. These measurements will be utilized for validation of neutron transport tools (software codes) and tritium breeding predictions used for breeding blanket systems design. NAS for LLCB TBM mainly consists of transfer station, capsule loader, transfer lines, foil gamma activity measurement system and irradiation ends. The irradiation of capsules consisting of foils is positioned inside the LLCB TBM at mid-plane location. The conceptual design of TBM along with NAS irradiation piping has been developed and its engineering design is in progress. All the components of NAS will be kept inside tritium building level L-2 of ITER building. The capsules are pneumatically transferred to irradiation end of piping located inside the TBM. After irradiation, the capsules are transferred back to counting station for foil activation measurement. This paper will present the conceptual design of NAS system along with preliminary engineering analysis and sequence of operations.
        Speaker: Ms Shailja Tiwari (Institute for Plasma Research)
      • 239
        Contribution of fusion energy to low-carbon development under the Paris Agreement and accompanying uncertainties
        The Paris Agreement requires deep reduction of greenhouse gas emissions. The world is toward rapid transition not only for climate change mitigation but also for sustainable development. Fusion energy has outstanding characteristics of plentiful resources, no nuclear runaway and zero-carbon emission, and its development has made a remarkable progress thanks to large investment for more than 50 years. However, long-term strategies for fusion energy development will become critically important in order to promote future DEMO projects by another large-scale investment and gain social acceptance. In this study, we assessed potential contribution of fusion energy to low-carbon development which is prescribed in the Paris Agreement under the combination of uncertainties of future socioeconomic development, the 2°C target and future commercial fusion power plants. We analyzed global energy systems up to 2100 in consideration of uncertainties by combining socioeconomic scenarios, global CO2 emission pathways, and fusion power plants by using a global energy systems model: DNE21+. We used three Shared Socioeconomic Pathways (SSPs) to express the uncertainty of future socioeconomic development. Assumptions and parameters for DNE21+ were harmonized with the SSP narratives. Four global CO2 emission pathways were used to simulate the uncertainty of the long-term targets of the Paris Agreement. For the uncertainty of fusion energy development, we set three scenarios, i.e., No Fusion, Conventional R&D and Advanced R&D which have different assumptions on parameters of fusion power plants. The parameters were set by considering potential and achievable cost reduction and performance improvement on the extension of DEMO concept design. Global negative CO2 emission in 2100 by drastic decarbonization of energy systems is required in order to achieve the 2°C target, and fusion power plants will be installed in the latter half of the 21st century mainly in the countries which have limited potentials of zero-emission energy sources such as Japan, Korea and Turkey. If inexpensive power plants could be developed by enhanced R&D and advanced design in DEMO projects, fusion power plants will also be deployed in the EU28, India and China. This study could be implicated in long-term strategy planning for fusion energy development.
        Speaker: Dr Keii Gi (Research Institute of Innovative Technology for the Earth)
      • 240
        Core transport improvement in stable detachment with RMP application to the edge stochastic layer of LHD
        Significant core plasma transport improvement is observed in the detachment divertor operation, which is stabilized by application of resonant magnetic perturbation (RMP) to the edge stochastic layer of LHD. Pressure profile becomes peaked and the heat transport coefficient,chi_eff, estimated from transport analysis, reduced in the entire confinement region. The RMP amplitude scan experiments show change of detachment transition density and of resulting chi_eff, while attained divertor particle flux reduction and radiated power are independent of the RMP amplitude. The results are new systematic study of RMP effects on detachment as well as on the core plasma transport. It suggests compatibility of good core plasma performance with divertor power load reduction in 3D magnetic field configuration with RMP application. Compatibility of good core plasma performance with enhanced edge radiation to mitigate the divertor power load is a crucial issue for magnetically confined fusion reactors. It is, however, commonly observed that the core confinement degrades with increasing radiation fraction. It is also not clear yet how RMP affects the core plasma transport during detachment, where no systematic RMP amplitude scan experiments have been performed so far in either tokamaks or stellarators in this respect. In LHD, stable detachment control is realized with RMP application of m/n=1/1 mode, where the core plasma transport is found to improve in the detached phase. The present paper reports, for the first time, analysis of the core transport, edge radiation, and divertor particle/power flux reduction with systematic scan of RMP amplitude (B_r/B_0). The RMP application creates a magnetic island of m/n=1/1 in the edge stochastic layer, where the impurity radiation is enhanced due to increased volume of cold plasma region. The divertor power and particle flux exhibit n=1 mode pattern in toroidal direction. With RMP application, the radiated power increases at lower density compared to the no-RMP case, and thus it results in earlier detachment transition. There appears a plateau region of radiation against density rise. This leads to a wide density operation range and thus provides a stable detachment control. RMP amplitude scan experiments show celar change of detachment transition density and resulting energy transport coefficients.
        Speaker: Dr Masahiro Kobayashi (JpNIFS)
      • 241
        Dependence of RMP penetration threshold on plasma parameters and ion species in helical plasmas
        We investigate the penetration threshold of the RMP (Resonant Magnetic Perturbation) by the external coils in the LHD (Large Helical Device) for the various configurations. In a configuration of the LHD, it has qualitative similar dependence with that in Ohmic tokamak plasmas. However, the qualitative dependence on the collisionality is opposite to that in a high plasma aspect configuration, which is a quite unique property, and first found in the LHD. Also, we investigate the threshold on the ion species, and find that the threshold of deuterium is quite smaller than that of hydrogen. In the above cases, the RMP penetration thresholds are higher as the poloidal rotation is faster, which is qualitatively consistent with the torque balance model between the electro-magnetic and the poloidal neoclassical viscous torque.
        Speaker: Dr Kiyomasa WATANABE (National Institute for Fusion Science)
      • 242
        Design and Simulation Studies of Calorimetric Dummy Load for Gyrotron System
        High microwave power is generally measured and characterized by calorimetric dummy loads, which are designed to suit the exiting modes of the gyrotron / HPM. The output mode of the gyrotron is converted to a Gaussian mode-HE11 mode after passing through series of mode converters. The objective of this study is to design and fabrication of Calorimetric Dummy Load with efficient cooling medium which absorb maximum power of 200 kW at 42±0.2GHz frequency applied for 3 seconds, suited for microwave power propagating in HE11 mode. There is rigorous requirement of proper cooling channel or cooling medium over the dummy load system for the dissipation of heat in the quickest manner. As an effect of high microwave energy (maximum heat), internal heat buildup in the dummy load system which could results in a catastrophic failure or decrease in the life span of the dummy load. This research envisaged the thermal effect of microwave energy on a reflecting structure incorporated to transfer microwave energy to heat absorber media concatenating the effect of heat conduction via multi flow path technique. In this manuscript, CFD analysis using ANSYS has been carried out to find the temperature contour, velocity contour, pressure contour for water passing through the helical tubes and thermal analysis has also been carried out for reflecting medium and microwave absorber material inside the enclosure. Details of these analyses results and their optimizations will be discussed in this paper.
        Speakers: Mr Axat Patel (Assistant Professor), Mr Maulik Shah (Assistant Professor)
      • 243
        Design Progress of Advanced Fusion Neutron Source for JA/DEMO Fusion Reactor
        Based on results from the IFMIF/EVEDA project in the Broader Approach (BA) activities, a conceptual design of the Advanced Fusion Neutron Source (A-FNS) in Rokkasho aiming at obtaining material irradiation data up to 20 dpa for a fusion DEMO reactor is presented in this paper. The A-FNS is composed of an accelerator facility with a 40 MeV and 125 mA deuteron beam, a test facility and a post irradiation examination facility. A particular attention in the design is paid on an integration of the test facilities by adopting a newly designed test specimen module for A-FNS. Recently, the nuclear analysis of test module has been progressed to optimize the irradiation of test pieces and then it was clarified that our original module enabled the test pieces to be irradiated uniformly.
        Speaker: Dr Kentaro Ochiai (QST)
      • 244
        Deuteron Beam Commissioning of the Linear IFMIF Prototype Accelerator Source and LEBT

        The Linear IFMIF Prototype Accelerator aims to operate in Rokkasho Fusion Institute a 125 mA/cw deuteron beam at 9 MeV In order to prove the technical feasibility of the IFMIF accelerators concept.

        A 2.45 GHz ECR ion source developed by CEA-Saclay is designed to deliver 140 mA/100 keV CW D + beam. The low energy beam transfer line (LEBT) relies on a dual solenoid focusing system to transport and match the beam into the next accelerating section which is a Radio-Frequency Quadrupole (RFQ). At the end of the LEBT, the normalized RMS emittance has to be lower than 0.3$\pi$ mm.mrad in order to reach the optimal beam transmission through the RFQ.

        This contribution will present the different commissioning phases of LIPAC ion source and LEBT. The experimental results that have been obtained will be reported. In particular, beam emittance measurements as a function of ion source extraction voltage gaps, total extracted current from the source and solenoid tunings will be showed.

        In order to model as well as possible the beam transport thought LEBT, intensive beam dynamics simulations that take into account space charge compensation have been performed using a self-consistent particle-in-cell code. Simulation results will be discussed and compared to experimental data.

        Speaker: Dr Masayoshi Sugimoto (JpQSTRFI)
      • 245
        Development and Qualification of Passive Active Multijunction (PAM) Launcher for LHCD System of ADITYA -Upgrade Tokamak
        A Passive Active Multijunction (PAM) antenna is designed and developed and would be commissioned on the ADITYA-U tokamak. The PAM antenna has many advantages over the grill antenna such as exhibiting a lower reflection coefficient at the plasma densities close to its corresponding cut-off density. The PAM antenna along with its transmission line components are designed to deliver RF power up to 250 kW for 1 second and its design is validated using COMSOL Multiphysics and CST studio. This paper describes the fabrication protocols of each component of the PAM launcher and its transmission line components along with its low power test method