book of abstracts - soft 2018 › ?download=book-of-abstracts.pdf · 30th symposium on fusion...

30TH SYMPOSIUMON FUSION TECHNOLOGY

SEPTEMBER 16-21, 2018Giardini Naxos, Italy

www.soft2018.eu

SOFT30th

2018

ITALIAN NATIONAL AGENCY FOR NEW TECHNOLOGIES,ENERGY AND SUSTAINABLE ECONOMIC DEVELOPMENT

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30TH SYMPOSIUM ON FUSION TECHNOLOGY

SEPTEMBER 16-21, 2018Giardini Naxos, Sicily - Italy

B O O K O F A B S T R A C T S

www.soft2018.eu

Contents

I1 Invited Talks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

I2 Invited Talks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2

I3 Invited Talks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5

I4 Invited Talks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8

I5 Invited Talks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .11

I6 Invited Talks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14

O1 Oral session . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .20

O2 Oral session . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .41

O3 Oral session . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .62

P1 Poster session . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .83

P2 Poster session . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 315

P3 Poster session . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 550

P4 Poster session . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 787

Author index . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1022

ii

30th edition of the Symposium on Fusion Technology (SOFT 2018) | 16th to 21st September 2018 | Giardini Naxos (Messina), Sicily, Italy

Page 1

I1 .1

ITER construction and manufacturing progress toward first plasmaBigot, Bernard

The ITER Organization, St Paul-lez-Durance Cedex, France

ITER reached in November 2017 completion of 50% of the work required to achieve First Plasma . Progress is most visible in the completion of many key buildings, such as the tokamak assembly building, the cryogenic plant, and the magnet power supply building have been completed . The tokamak building will be ready for equipment in 2020 and the bioshield is already to full height . Key systems begin commissioning in 2018, including the steady-state electric network and the component cooling water, while the cryogenic system and magnet power supply commissioning begins in 2019 . Thus, the physical plant is moving rapidly toward completion, and key systems are entering the commissioning phase . Manufacturing progress is equally impressive . The base and lower cylinder of the cryostat have been assembled on the ITER site . Three of the six poloidal field coils and the first of the six modules of the central solenoid are being wound or finished . The winding pack and casing for the first toroidal field magnet are complete and verified to meet the high tolerances required (


Page 2

I2 .1

Status European procurement for ITERSchwemmer, Johannes

Fusion for Energy, Barcelona, Spain


Page 3

I2 .2

European roadmap to fusion energyDonné, Tony

EUROfusion, Garching, Germany

The European Roadmap to the realisation of fusion energy breaks the quest for fusion energy into eight missions . For each mission, it reviews the current status of research, identifies open issues, proposes a research and development programme and estimates the required resources . It points out the needs to intensify industrial involvement, to educate the fusion scientists and engineers of the future, and to seek all opportunities for collaboration outside Europe .

A long-term perspective on fusion is mandatory since Europe has a leading position in this field and major expectations have grown in other ITER parties on fusion as a sustainable and secure energy source . China, for example, is launching an aggressive programme aimed at fusion electrici-ty production well before 2050 . Europe can keep the pace only if it focuses its effort and pursues a pragmatic approach to fusion energy . With this objective the present roadmap has been elaborat-ed . The roadmap covers three periods: The short term which roughly covers the period until ITER comes into operation and the DEMO Conceptual Design is completed, the medium term which runs until ITER is in routine operation at high performance and the DEMO Engineering Design is completed and the long term .

ITER is the key facility of the roadmap as it is expected to achieve most of the important mile-stones on the path to fusion power . Thus, the vast majority of resources proposed in the short term are dedicated to ITER and its accompanying experiments . The medium term is focussed on taking ITER into operation and bringing it to full power, as well as on preparing the construction of a demonstration power plant DEMO, which will for the first time supply fusion electricity to the grid . Building and operating DEMO is the subject of the last roadmap phase: the long term . It might be clear that the Fusion Roadmap is tightly connected to the ITER schedule . A number of key milestones are the first operation of ITER, the start of the DT operation, and reaching the full performance at which the thermal fusion power is 10 times the power put in to the plasma .

DEMO will provide first electricity to the grid . The Engineering Design Activity will start a few years after the first ITER plasma, while the start of the construction phase will be a few years after ITER reaches full performance . In this way ITER can give viable input to the design and development of DEMO . Because the neutron fluence in DEMO will be much higher than in ITER (atoms in the plasma facing components of DEMO will undergo 50-100 displacements during the full operation life time, compared to only 1 displacement in ITER), it is important to develop and validate materials that can handle these very high neutron loads . For the testing of the materials a dedicated 14 MeV neutron source is needed . This DEMO Oriented Neutron Source (DONES) is therefore an important facility to support the fusion roadmap .

The presentation will focus on the strategy behind the fusion roadmap and will describe the ma-jor challenges that need to be tackled on the road towards fusion electricity . Encouraging recent results will be given to demonstrate the outcome of the focused approach in European fusion re-search .

The author is indebted to the whole European fusion community that is together working to make fusion a reality . This work has been carried out within the framework of the EUROfusion Con-sortium 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 .


Page 4

I2 .3

Role of Italian DTT in the Power Exhaust implementation strategyMazzitelli, Giuseppe

FSN, ENEA, Frascati (RM), Italy

In the European road map towards the realisation of fusion energy, one of the challenges is the power exhaust for DEMO . If the ITER baseline strategy can’t be extrapolated to DEMO, tens of years will delay the realization of a fusion plant . So in parallel to ITER exploitation, it is manda-tory to test alternative solutions for the heat loads on the divertor as risk mitigation for DEMO .

In the last years two schemes have been proposed as possible solutions: alternative magnetic con-figurations and the use of liquid metal divertors . Up to now these solutions have been tested at proof of principle level in devices with a plasma current not exceeding 1 MA and SOL parameters significantly different from reactor conditions . To implement one of these concepts on DEMO it is necessary to make another intermediate step, otherwise the extrapolation to DEMO is too large by upgrading existing facilities or by building a dedicated Divertor Tokamak Test (DTT) facility .

In this framework, the Italian fusion community with the involvement of European Labs has pro-posed a new device, the Italian DTT, in which one or more alternative magnetic configurations or/and liquid metals can be tested in DEMO relevant conditions .

To fulfil the DEMO requirements the device have to operate at relevant heat load on the divertor maintaining high core performance, i .e . operations in an integrate scenario with core and edge pa-rameters as close as possible to ITER and DEMO .

The Italian DTT project has fully supported by the Italian Government that has also identified and is implementing a funding scheme .

DTT will be a high field superconducting toroidal device (6 T) carrying plasma current up to 6 MA in pulses with length up to 100s, with an up-down symmetrical D-shape defined by major radius R=2 .15 m, minor radius a=0 .7 m, and an elongation around 1 .7 .

The status of the project will be illustrated, highlighting the reviewed design addressed to increase the flexibility by allowing for fully double null operation .

The site, the time schedule, and the cost estimation will be presented too .


Page 5

I3 .1

European Integrated Programme in support to ITER: Overview of JET and Medium Size Tokamaks resultsLitaudon, Xavier

EUROfusion Consortium JET, Culham Science Centre, Abingdon, United Kingdom

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’ . EUROfu-sion has seized the unique opportunity to develop an integrated programme on devices of different sizes, i .e . on EU Medium-Size Tokamaks (MSTs), and, on JET in order to provide a step-ladder approach for extrapolation to ITER . In addition, the ITER Organization has issued a detailed analysis of the risks to ITER operation and has identified the main R&D needs to mitigate those risks in the revised ITER research plan . In this context, this paper will provide an overview of the recent coordinated contributions of the EU programme to optimise ITER operation .

Disruptions are considered as the highest operational risk in the ITER Research Plan . The high priority physics studies on JET and MSTs consist of disruption prediction, avoidance, mitigation and associated modelling (including multi-machine run-away electrons model validation) . A new shattered pellet injection system is being installed on JET to compare with massive gas injec-tion and elucidate the differences in run-away electrons beam mitigation in view of impacting the design of the ITER disruption mitigation system . The JET and the MSTs programmes have concentrated on the preparation of ITER operating scenarios and on providing a physics basis for optimising fusion performance operation with metallic first wall materials . It is found on JET and ASDEX Upgrade, that plasma performance is significantly affected when plasma boundary condi-tions are modified which will affect the strategy to achieve the fusion performance in the coming JET deuterium-tritium campaign and ITER QDT=10 main mission . In addition, preparation of the ITER non-active phase has been carried out in hydrogen on JET, and, in hydrogen and helium in the MSTs . The recent progress will be reviewed on plasma surface interaction with ITER first wall materials (e .g . beryllium and tungsten erosion/migration, helium and tungsten interaction), scaling of L to H mode power threshold, Scrape-Off-Layer physics, core and pedestal confinement with different hydrogen isotopes and helium, control of detached divertor scenarios using extrinsic impurity seeding, and options for ELMs control with pellets or with resonant magnetic perturba-tions, RMPs . ELMs control with RMPs has been established on ASDEX Upgrade in helium using methods developed for deuterium plasmas, addressing a ITER issue of the transferability of ELMs control methods .

To conclude, the success of ITER operation will also require integrating the experimental progress made in different fusion facilities through theory-based first principle and integrated modelling . The European Transport Simulator, ETS, for integrated modelling has undergone major develop-ment and has been benchmarked against TRANSP . The strategic movement towards the adoption of the ITER integrated modelling and analysis suite (IMAS) has been pursued by the continued support and validation of the IMAS infrastructure and extension of the EUROfusion experimental databases in IMAS .

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

1 See author list of “X . Litaudon et al ., 2017 Nucl . Fusion 57 102001”

2 See author list of “H . Meyer et al ., 2017 Nucl . Fusion 57 102014”


Page 6

I3 .2

JT-60SA Program Contribution to Fusion EnergyKamada, Yutaka

Naka Fusion Institute, National Institutes for Quantum and Radiological Science and Technology, Naka-shi, Ibaraki-ken, Japan

JT-60SA is a highly-shaped large superconducting Tokamak under construction by EU and Japan . The mission of JT-60SA is to support ITER and to complement ITER towards DEMO by resolv-ing key physics and engineering issues . Fabrication and installation of components of JT-60SA by EU-Japan Integrated Project Team are progressing on schedule towards the first plasma in Sep . 2020 . On the Cryostat Base made by CIEMAT, the 340-gegree-part of the Vacuum Vessel (VV) has been placed and welded accurately by QST . By Feb . 2018, all 18 TF coils have been manufac-tured and cold-tested by ENEA and CEA and 14 TF coils have been assembled to the tokamak by QST . Manufacture of all 6 EF coils have been completed by QST . Commissioning of the cryogenic system was completed by CEA in Naka . High Temperature Superconducting current leads have been delivered by KIT . Commissioning of the power supply system (ENEA, RFX, CEA and QST) has also been implemented smoothly . The Cryostat Vessel Body has been delivered by CIEMAT .

The JT-60SA Research Plan (SARP) ver . 3 .3 was issued in March 2016 by 378 co-authors (JA 165 (16 institutes), EU 213 (14 countries, 30 institutes): Using ITER- and DEMO-relevant plasma regimes and its sufficiently long discharge duration, JT-60SA enables studies on all the key physics issues for ITER and DEMO . From ~2030, the first wall will be changed from carbon to full tung-sten-coated carbon . By integrating these studies, the project provides ‘simultaneous & steady-state sustainment of the key plasma performances required for DEMO’ . Such JT-60SA research activity includes consolidation of a “Plant Simulator” . As for the first plasma and heating experiments, JT-60SA will start earlier than ITER by five years . Therefore, experiences and achievements in JT-60SA are expected to contribute to reliable operation of ITER .


Page 7

I3 .3

W7-X: Technology progress of the experimental campaign with divertor plasmasFellinger, Joris

Max Planck Institute for Plasma Physics, Garching bei München, Germany

Wendelstein 7-X (W7-X), a fivefold symmetric stellarator located at the Max-Planck-Institute for Plasma Physics in Greifswald, Germany, was successfully taken in operation with short pulse limiter plasmas in 2015 . Hereafter, ten symmetrically positioned un-cooled graphite divertors were installed, the plasma facing wall was refurbished with graphite tiles and various auxiliary systems and diagnostics were upgraded . The reinforcements allow for an increase of the energy input from 4 to 80 MJ .

The experimental campaign with island divertor plasmas was launched in August 2017 . The main goal of this campaign is to demonstrate the capability of the un-cooled test divertor in high den-sity and high power plasmas . In the island divertor concept, the divertor is positioned in front of the pumps (for the neutrals exhaust) and it only intersects the relatively cold resonant islands around the core plasma . In this way, convective heat loads are limited to 10 MW/m² and neutral pumping is effective, despite the fact that the divertor area represents only 10 % of the plasma facing surface .

Second goal of the experimental campaign is to prepare for the installation of the final water cooled divertor . The water cooled divertor is planned for 2020 and has to be installed with great care and high accuracy . It relies on the experience gained with the installation of the un-cooled divertor .

To improve the density in comparison to the first campaign with limiter plasmas, an injector of frozen hydrogen pellets was installed and a fast hydrogen gas inlet with piezo-valves mounted in cut-outs of the divertor were taken in operation . In addition, various diagnostics were added or enhanced before start of the divertor campaign .

The tests divertor campaign is split into two parts: During a short break a the middle of the ex-perimental campaign, two so-called scraper elements are installed in front of their corresponding divertor to shield the sensitive edges of the divertor near the pumping gap . Aim of the scraper program is to compare the edge loads on the divertor with and without a scraper and to evaluate the impact of the scraper on the neutrals pumping efficiency . The scraper elements are monitored by additional diagnostics . The break was also used to take the first of two neutral beam injectors (NBI) into operation and to add or harden several diagnostics .


Page 8

I4 .1

SPIDER in the roadmap of the ITER Neutral BeamsSerianni, Gianluigi1; Toigo, Vanni1; Team, NBTF1; Boilson, Deirdre2; Rotti, Chandramouli2; Bonicelli, Tullio3; Chakraborty, Arun4; Fantz, Ursel5; Kashiwagi, Mieko6; Tsumori, Katsuyoshi7

1Consorzio RFX, Padova, Italy2ITER Organization, St. Paul-lez-Durance, France3Fusion for Energy, Barcelona, Spain4Institute for Plasma Research, Gandhinagar, India5IPP, Max-Planck-Institut für Plasmaphysik, Garching bei München, Germany6National Institutes for Quantum and Radiological Science and Technology, Naka, Japan7National Institute for Fusion Science, Toki, Japan

To reach fusion conditions and control plasma configuration in ITER, the next step towards es-tablishing nuclear fusion as viable energy source, suitable combination of additional heating and current drive systems is necessary . Among them, two Neutral Beam Injectors (NBI) will provide 33MW hydrogen/deuterium particles electrostatically accelerated to 1MeV; efficient gas-cell neu-tralisation at such beam energy requires negative ions, obtained by caesium-catalysed surface conversion of hydrogen/deuterium atoms in the ion source . ITER NBI requirements have never been simultaneously attained; so a Neutral Beam Test Facility (NBTF) was set up at Consorzio RFX (Italy) . Experiments will verify continuous NBI operation for one hour, under stringent re-quirements for beam divergence (


Page 9

I4 .2

A smart architecture for the DEMO fuel cycleDay, Christian1; Butler, Barry2; Ploeckl, Bernhard3; Giegerich, Thomas1

1Institute of Technical Physics (ITEP), Karlsruhe Institute of Technology (KIT)-CN, Eggenstein-Leopoldshafen, Germany2Culham Science Centre (CCFE), Abingdon, United Kingdom3Max-Planck-Institute for Plasma Physics (IPP), Garching, Germany

In the framework of the EUROfusion DEMO Programme the EU has elaborated a completely nov-el and most innovative fuel cycle architecture, driven by the need to reduce the tritium inventory to an absolute minimum .

To achieve this goal, batchwise processes used in the fusion fuel cycle so far were replaced by continuous processes wherever possible . This includes the change from discontinuous cryopumping to mercury based continuous vacuum pumping with practically zero demand on cryoplant power, and the introduction of thermal cycling ab- and adsorption processes for isotope rebalancing in the tritium plant instead of large cryogenic distillation columns with tritiated liquid hold-ups . To further reduce inventory, the well-known approach to route all exhaust gas through the tritium plant has been abandoned in favor of a three-loop architecture . There, superpermeable metal foils are introduced in the divertor ports to separate a pure DT stream which is then immediately recycled to feed the pellet injection systems . To increase the core fueling efficiency, optimization potentials in the design of the high field side pellet injection systems are being exploited . Finally, a unified fuel cycle simulator is under development on a commercial software platform in order to identify optimization potentials within the fuel cycle, to allow impact studies, and on a long term to support the development of tailored control and operational strategies .

The talk will present the first integrated and consolidated design point of the fuel cycle based on the 2017 European DEMO baseline . It is shown how the DEMO requirements are picked up and affect system level performance . Examples are given for integration issues and how they were solved (remote handling, divertor integration, plasma control) . Finally, a roadmap is delineated which illustrates the remaining R&D efforts needed to achieve at a validated and complete con-ceptual design until the mid 2020s .


Page 10

I4 .3

High Temperature Superconductors for Future Fusion Magnets and Industrial High Current ApplicationsWolf, Michael J .

Institute for Technical Physics, Karlsruhe Institute of Technology, Karlsruhe, Germany

High-temperature superconductors (HTS) have the potential to enable the operation of a future fu-sion reactor at higher magnetic fields (> 14 T) or at higher temperatures compared to conventional low- temperature superconductors . In particular, the operation at high magnetic fields with good temperature margin is perceived to be an important advantage of HTS in a fusion power plant .

Fusion magnets require high current superconductors, which are embedded in a stainless-steel jacket for mechanical support against Lorentz forces and actively cooled by a forced flow of cool-ant . Design challenges and cable proposals for winding packs based on HTS conductors will be presented .

HTS CrossConductor (HTS CroCo) is a high current HTS conductor with high current density to be fabricated in long lengths . The recent progress of HTS CroCo fabrication will be shown and a toroidal field coil winding pack design based on such HTS CroCos will be presented as an exam-ple to demonstrate the principle feasibility of HTS for future fusion magnets .

The knowledge gained in the design of high-current conductors for fusion magnets and the experi-ence to fabricate HTS CroCo strands in different geometries enable the design and construction of HTS DC cables for industrial high current applications as well . Key design aspects and promising fields of applications of HTS high current cables beyond fusion magnets will be presented .


Page 11

I5 .1

Present Progresses and Activities on the Chinese Fusion Engineering Test ReactorWan, Yuanxi1; Li, Jiangang1; Liu, Yong2; Wang, Xiaoling3

1Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, China2Southwest Institute of Physics, Chengdu, China3Chinese Academy of engineering, Mianyang, China

The Chinese Fusion Engineering Testing Reactor (CFETR) is the next device for the Chinese mag-netic confinement fusion (MCF) program which aims to bridge the gaps between the fusion exper-iment ITER and the fusion power plant . CFETR detail engineering design and R&D project have been approved by Chinese government . The activities have been started in the end of last year . CFETR new design focus on the high magnetic field by using high performance Nb3Sn wire for TF and 2212 HTc CICC for CS, and large size with major radio 7m . Steady-state operation and trit-ium self-sustainment will be the two key issues for the first phase with a modest fusion power up to 200 MW . The staged operation for late phases will explore for DEMO validation with a fusion power over 1 GW and Q over 20 . Operational scenarios with L-mode, hybrid H-mode and steady state advanced H-mode physics will introduced in this presentation together with R&D activities for H&CD, diagnostic, VV, divertor, superconducting magnets, T-plant and related technology, material, remote handling, physical validation on EAST tokamak aiming high performance steady state operation, and future developing plan .


Page 12

I5 .2

European Materials Development: Results and PerspectivesPintsuk, Gerald1; Diegele, Eberhard2; Gorley, Mike3; Henry, J .4; Rieth, Michael5

1Forschungszentrum Jülich GmbH, Jülich, Germany2EUROfusion Consortium, Garching, Germany3Culham Centre for Fusion Energy, Culham Science Centre, Oxfordshire, United Kingdom4CEA, Saint-Paul-Lez-Durance, France5Karlsruhe Institute for Technology, Karlsruhe, Germany

This paper reviews the material strategy of the EU fusion roadmap and the recent progress of activities within the EUROfusion materials research program . It highlights, both, the characteri-zation and validation of in-vessel components baseline materials, i .e . EUROFER97, CuCrZr and tungsten as well as the development and characterization of advanced structural and high heat flux materials for DEMO and beyond . In support of engineering design activities the primary focus is on compilation of data and the supply and release of material property handbooks, material as-sessment reports complemented with the development of design criteria and material design limits appropriate for DEMO thermal, mechanical and environmental conditions .

Data are presented and discussed with respect to DEMO operating specifications for selected sub-topics, which include (a) advanced steels optimized towards low and high temperature extension of the current operational window, (b) heat sink materials, i .e . copper based alloys and compos-ites and (c) plasma facing materials, i .e . tungsten based composites . Based on the discussion and conclusions drawn, perspectives for the required materials performance and future research and validation steps will be given .

A first glance on these perspectives, the most far-reaching progress by now in the EUROfusion materials program and the first step of a necessary and more extensive qualification program, is provided by the launch of nine neutron irradiation campaigns within the last two years . Baseline structural and high heat flux materials are irradiated up to medium neutron dose levels for con-tinuously filling gaps in the materials property handbook and advanced material options at lower neutron fluence for screening, down-selection and increase of fundamental knowledge of n-damage . The results will guide the future materials research and validation program as well as design op-tions for blanket and divertor components .


Page 13

I5 .3

Towards the EU fusion-oriented neutron source: the Preliminary Engineering Design of IFMIF-DONESBernardi, Davide1; Arbeiter, F .2; Cappelli, Mauro3; Fischer, U .2; García, A .4; Heidinger, R .5; Krolas, W .6; Martin-Fuertes, F .4; Miccichè, Gioacchino1; Muñoz, A .7; Nitti, Francesco Saverio1; Pérez, M .4; Pinna, Tonio3; Tian, K .2; Ibarra, a .4

1ENEA, Brasimone, Italy2KIT, Karlsruhe, Germany3FSN, ENEA, Frascati (RM), Italy4CIEMAT, Madrid, Spain5F4E, Garching, Germany6IFJ PAN, Krakow, Poland7Empresarios Agrupados, Madrid, Spain

The need of a high-intensity, 14 MeV-peaked neutron source for the qualification of materials under fusion-relevant conditions has been recognized in the European (EU) fusion programme as an essential step towards the design and licensing of DEMO and future commercial fusion power plants . This need has pushed the EU to support the development of a Li(d,nx) neutron source called IFMIF-DONES (International Fusion Materials Irradiation Facility-DEMO Oriented Neu-tron Source) based on and taking advantage of the results obtained in the IFMIF/EVEDA (Engi-neering Validation and Engineering Design Activities) project conducted in the framework of the bilateral EU-Japan Broader Approach Agreement .

The design activities and the supporting R&D work of the DONES facility are presently being carried out in the framework of the Work Package Early Neutron Source (WPENS) of the EU-ROfusion consortium in close collaboration with Fusion for Energy agency, with the main goal of consolidating the underlying technology and developing a sound design basis in order to be ready for IFMIF-DONES construction at the early beginning of the next decade .

In this paper, the main engineering advances achieved during the first three years of the WPENS project and included in the recently released IFMIF-DONES Preliminary Engineering Design Report as an important milestone of the project are presented, focusing in particular on the main design evolutions from the previous phases and on the critical aspects to be further developed in the near future .

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 agree-ment No 633053 . The views and opinions expressed herein do not necessarily reflect those of the European Commission, Fusion for Energy, or of the authors‘ home institutions or research funders .


Page 14

I6 .1

An Overview of the EU Breeding Blanket Strategy as Integral Part of the DEMO Design EffortFederici, Gianfranco

Power Plant Physics and Technology, EUROfusion, Garching, Germany

As an important part of the Roadmap to Fusion Electricity, Europe is conducting a pre-conceptual design study of a DEMO Plant to come in operation around the middle of this century with the main aims to demonstrate the production of few hundred MWs of net electricity and to demon-strate feasibility of operation with a closed-tritium fuel cycle .

This paper provides and overview of the newly revised design and development strategy of the breeding blanket in Europe that has been defined to take into account the input from the DEMO pre-conceptual design activities the findings and recommendations of a thorough technical and programmatic assessment of the breeding blanket program and the EU TBM program, for ITER recently conducted by an independent expert panel [1] . This was conducted to identify, among the available options, the most mature and technically sound breeding blanket concepts to be potentially used as “driver” blanket in DEMO [2] and as “advanced” blanket (to be in-stalled and tested in properly designed segments), having the potential to be more attractive for a First-of-a-Kind (FoaK) reactor, the remaining technical gaps and to align and strengthen the supporting R&D Program . To ensure a coherent and efficient Program, a change of the EU TBM options to be tested in ITER is proposed in order to obtain important and useful information from the two considered breeders (solid and liquid) and the two coolants (helium and water) .

[1] M . Gasparotto et al ., TBM/DEMO BB Programs Review, Final report September 2017 .

[2] i .e ., the near-full coverage blanket to be installed by day-1 to achieve electricity production and to achieve tritium self-sufficiency


Page 15

I6 .2

Safety aspects on the road towards fusion energyPerrault, Didier

Institut de Radioprotection et Sûreté Nucléaire (IRSN), Villeneuve les Avignon, France

On the road toward fusion energy, ITER is the first fusion installation which will have enough ra-dioactive inventory to be potentially dangerous for the public and the environment . As such, ITER has a licensed nuclear facility status and ITER Organization, the operator, has to follow a licensing process through which it has to demonstrate to the regulator that the installation is safe at all stages of its operation . In practice, the operator has to define technical, organizational and human provisions such as to prevent or adequately limit the risks of accidents and the disadvantages (exposure to ionizing radiation, environmental releases and waste) that the installation presents .

All installations which will be created after ITER in order to develop (DEMO . . .) then make use of fusion energy (PROTO . . .), will clearly also be installations for which the safety aspects will have to be taken into account . If this is not correctly done, it could be an obstacle (in terms of delay or additional cost) or a stop (no licensing) on the way to fusion energy . Perfectly mastered, it could also be a benefice for fusion power compared with other choices .

This paper begins with describing the safety features of fusion installations (first confinement bar-rier surrounded by large energy sources, plasma disturbances, explosion hazard . . .) . Then, a state of progress of the ITER safety demonstration is provided: solved issues (the tokamak support design, accident within the neutral beam cell . . .), issues under IRSN’s expertise (Explosion within the vac-uum vessel, New Vacuum Vessel Pressure Suppression System . . .), issues to be solved (Detritiation system efficiency, radioprotection, hot cells, tritium and waste buildings . . .) . Finally, a chapter is dedicated to the possible evolutions of the safety issues for the installations succeeding ITER on the road to fusion energy (decay heat removal, exposure to ionizing radiation, environmental re-leases . . .) . General comments about dealing with these evolutions close this paper (safety from the earliest design stage, lessons learned, involvement of the regulator . . .) .


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I6 .3

Neutronic challenges on the way from ITER to DEMOVillari, Rosaria1; Angelone, Maurizio1; Batistoni, Paola1; Fischer, Ulrich2; Loughlin, Michael3; Pampin, Raul4; Taylor, Neill5

1FSN, ENEA, Frascati (RM), Italy2Karlsruhe Institute of Technology, Karlsruhe, Germany3ITER Organization, Saint Paul Lez Durance, France4Fusion for Energy, Barcelona, Spain5Centre for Fusion Energy, Culham Science Centre, Abingdon, United Kingdom

Reliable neutronic assessments are essential for the design and the safe operation of high perfor-mance fusion facilities . For the under construction ITER device, accurate and complete evalua-tions of the nuclear responses from the various radiation sources are mandatory to optimize the shielding design, guarantee a sufficient protection of critical components and minimize the occu-pational exposure of workers . To reduce uncertainties of the computational predictions and the associated risks for the ITER operations, benchmark experiments are in preparation for the future DT campaign at JET within the EUROfusion Consortium, aimed at validating neutronics tools and data used in ITER nuclear analyses . In parallel, design studies are underway in the EU for the development of a demonstration fusion power plant (DEMO) aiming at the production of elec-tricity and the operation of a closed fuel cycle . The DEMO design and related R&D activities will benefit from the ITER experience . Nevertheless, DEMO should demonstrate full tritium breeding capability to achieve self-sufficiency, which is only partially addressed with ITER through the Test-Blanket-Module (TBM) program . Furthermore, the DEMO operative irradiations conditions will be significantly more demanding and the issues related to the degradation and changes of ma-terials properties, including gas production, activation, erosion and corrosion products, contamina-tion, etc . ., and the resulting radiation dose loads, will be more severe than in ITER . The significant in-vessel material damage and activation necessitates the development of proper neutron resistant and low activation structural materials which need to be qualified and tested under intense fusion neutron source . Thus, accurate and reliable nuclear analyses to address DEMO design and safety requirements require specific efforts on the improvement and optimisation of simulation tools and nuclear data supported with dedicated experimental activities .

This paper reviews the main ITER neutronic issues, lessons learnt and implications for the DEMO nuclear design and safety, as well as the current R&D activities on codes and nuclear data devel-opment and the supporting experimental program . Further needs and challenges to cover the tech-nological gap between ITER and DEMO and efforts to reduce uncertainty margins are discussed .


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I6 .4

Diagnostics for plasma control - from ITER to DEMOBiel, Wolfgang1, 2; Albanese, Raffaele3; Ambrosino, Roberto3; Ariola, Marco3; van Berkel, Matthijs4; Cecconello, Marco5; Conroy, Sean5; Dinklage, Andreas6; Duran, Ivan7; Dux, Ralph8; Entler, Slavomir7; Fable, E .8; Farina, Daniela9; Finotti, Claudio10; Franke, Thomas8, 11; Giacomelli, Luca12; Giannone, L .8; Gonzalez, W .1; Hjalmarsson, Anders5; Hron, Martin7; Janky, F .8; Kallenbach, Arne8; Kogoj, J .13; König, Ralf14; Luis, Raul15; Malaquias, Artur15; Marchuk, Oleksandr1; Marchiori, Giuseppe10; Mattei, Massimiliano3; Maviglia, Fabio3, 11; De Masi, Gianluca10; Mazon, Didier16; Meister, Hans8; Moutinho, R .15; Mlynek, Alexander8; Nowak, Silvana9; Piron, Chiara10; Pironti, Alfredo3; Policarpo, Hugo15; Quental, P .B .15; Rispoli, Natale9; Rohde, V .8; Sergienko, Gennady1; El Shawish, Samir17; Siccinio, M .8, 11; Silva, A .15; Da Silva, F .15; Sozzi, Carlo9; Tardocchi, Marco10; Tokar, Mirkhail1; Treutterer, Wolfgang8; Vale, Alberto15; Zohm, Hartmut8

1Forschungszentrum Jülich GmbH, Jülich, Germany2Ghent University, Ghent, Belgium3Consorzio CREATE, Napoli, Italy4DIFFER institute, Eindhoven, Netherlands5Uppsala University, Uppsala, Sweden6Max-Planck-Institut für Plasmaphysik, Greifswald, Germany7Czech Academy of Science, Praha, Czech Republic8Max-Planck-Institut für Plasmaphysik, Garching, Germany9IFP-CNR, Milano, Italy10Consorzio RFX, Padova, Italy11EUROfusion, Garching, Germany12Università degli Studi di Milano-Bicocca, Milano, Italy13Cosylab, Ljubljana, Slovenia14Magnetic sensor laboratory, Lviv, Ukraine15Universidade de Lisboa, Lisboa, Portugal16CEA, Saint Paul-lez-Durance, France17Jožef Stefan Institute, Ljubljana, Slovenia

The development of the plasma diagnostic and control (D&C) system for a future tokamak demon-stration fusion reactor (DEMO) [1] faces significant challenges [2] . These comprise the required reliability of operation, the high accuracy to which the plasma parameters are to be controlled, and the robustness of components and methods against any adverse effects or disturbances .

The ongoing developments for the ITER D&C system represent an important starting point for progressing towards DEMO . ITER diagnostic development is guided by a measurement require-ments table, in which different categories of diagnostics for machine protection, basic and ad-vanced control, as well as for evaluation and physics studies are being distinguished . Since ITER is an experiment aiming to explore and optimize in detail the physics of a burning plasma, ambitious targets for space and time resolution as well as for the measurement accuracies have been defined . These are pushing the ITER diagnostic design towards using sophisticated methods and aiming for large coverage and high signal intensities, forcing to mount many front-end components in


Page 18

forward positions . This results in many cases in a rapid aging of diagnostic components, so that additional measures like protection shutters, plasma based mirror cleaning or modular approaches for frequent maintenance and exchange have to be developed .

Under the even stronger fluences of plasma particles, neutron/gamma and radiation loads on DEMO, high reliability and long lifetime of diagnostics can only be achieved by selecting the methods with

regard to their robustness, and retracting vulnerable front-end components into protected loca-tions in the machine . Based on this approach, an initial DEMO D&C concept has been elaborated, which is covering all major control issues by main control parameters to be derived using at least two different diagnostic methods . Within this paper, on overview on the current status of D&C development for DEMO will be provided .

References:

[1] G . Federici et al ., Fusion Engineering and Design 109-111 (2016) 1464-1474

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


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I6 .5

A PROs-industry joined effort for the ITER construction: evaluating the impactPuliga, Gloria1; Manzini, Raffaella1; Batistoni, Paola2

1LIUC – Università Cattaneo, Castellanza (VA), Italy2FSN, ENEA, Frascati (RM), Italy

In modern innovation ecosystems, academic and practitioners’ studies point out the importance of contracts and relationships between innovation actors . Among others, science-based partners, namely universities and Public Research Organizations (PROs), are valuable external sources of knowledge and fundamental pillars of the innovation ecosystems . Although consensus exists on the fact that linkages between science-based partners and industry are crucial for both parts, measur-ing their impact is still daunting . In this regard, the following research question arises: What is the impact for industry when collaborating for ITER?

A mixed methodology has been adopted to answer this question . The unit of analysis is the single firm that collaborates in the context of ITER project, also with the involvement of ENEA . A to-tal of 26 Italian contractors and subcontractors were identified and analyzed . Data were collected among secondary sources for each firm (financial report and employment data available on AIDA, patent data available on Orbit database, macro data involving the economic and employment situation of the regions analyzed) . At the same time, 6 case studies were conducted with selected firms, and were analyzed by content analysis . Case studies enable to drill down the correlations emerged with statistical analyses, suggest explanations for not significant correlations and provide further insights to detect impacts that cannot be grasped by secondary data .

Descriptive statistics and regression analyses show a positive correlation between the starting of the contract with the return on activities (ROA) and the net financial position . Descriptive sta-tistics show the positive impact of the contracts for the financial performance EBITDA/Sales, not fully confirmed by linear regressions . As well, from general descriptive statistics, the impact on the employment growth appears, but linear regressions do not confirm this evidence . However, case studies support the correlations above, even those not confirmed by linear regressions, and bring into evidence other relevant dimensions of impact: market, innovation, learning, social impact . For SMEs, case studies point out an important strategic impact . The ITER project gives firms awareness of their “know-how” and of the need to stress their competencies and to diversify their businesses in a long-term view .

This contribution provides an overview of a scantly analyzed collaboration: PROs-industry at a firm-level dimension . Policy makers should consider the several dimensions of impact and sup-port firms to transform their huge investment into a long-term value . As well, the study helps managers to be aware of the investments required and set a long-term vision on how to transform them in operative returns . Particularly for SMEs, the award of big ITER contracts implies the change of the vision because it offers the awareness to be competitive and the cash flow to invest also in other businesses .


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O1A .1

Plasma control for EAST long pulse non-inductive H-mode operation in a quasi-snowflake shapeXiao, Bingjia1, 2; Luo, Zhengping1; Li, Jiangang1, 2; Yuan, Qiping1; Wu, Kai1; Wang, Yuehang1; Gong, Xianzu1; Wang, Liang1; Calabro, Giuseppe3; Albanese, Raffaele4; Ambrosino, Roberto4; De Tommasi, Gianmaria4; Crisanti, Flavio3; Pironti, Alfredo4

1Division of Control and Computer Application, Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, China2School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, China3ENEA Unità Tecnica Fusione, C.R. Frascati, Frascati, Roma, Italy4CREATE, Universittà di Napoli Federico II, Universittà di Cassino and Universittà di Napoli Parthenope, Napoli, Italy

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

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


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O1A .2

Implementation and exploitation of jet enhancements in preparation for dt operation and next step devicesMurari, Andrea

Programme Management Unit, EUROfusion Consortium, Abingdon, United Kingdom

JET presents some unique capabilities: the reactor fuel, ITER wall materials and the capability to confine the alphas . JET next T-T and D-T experimental campaigns can therefore address ma-jor physics and technological gaps for the development of fusion energy: the isotopic effects on confinement, the access the H mode and ELM behaviour . The total yield of the final D-T phase is expected to be 1013 n/s·cm2, a factor of six higher than the previous DTE1 . In this context, three main aspects of JET capability have been recently improved: 1) scenario development to enhance performance 2) the quality of the measurements to maximize the scientific exploitation 3) specific technologies for ITER and DEMO .

With regard to the scenarios, the performances of JET with a carbon wall have been reproduced up to a current of 3 MA, which supports the prediction of 15 MW fusion power in full DT . More-over improved control systems (wall load protection, simultaneous control of ELM frequency and beta, plasma mixture) insure that the plasma configurations are compatible with the wall proper-ties, from melting to retention and dust production .

In terms of general diagnostic capability, JET can now deploy much better resolution diagnostics, particularly for the edge quantities, and a consistent set of techniques to diagnose the fast par-ticles, from redistribution to losses, using techniques ranging from gamma ray spectroscopy to a scintillator probe and Faraday cups . A full calibration of the neutron diagnostics for the 14 MeV neutrons has just been completed successfully .

With regard to ITER and DEMO relevant technologies, specific programmes are being pursued to investigate: the radiation field, the induced activity and dose rates and the radiation damage of materials . Dedicated studies for DEMO, including the tests of a new tritium pumping cycle and a tritium breeding blanket mock-up, are also almost completed .


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O1A .3

Electromagnetic FEM studies of disruptions and engineering consequences for the power supply and coils design of planned upper divertor at ASDEX UpgradeTeschke, Markus; Herrmann, Albrecht; Pautasso, Gabriella; Vierle, Thomas; Zammuto, Irene

Max Planck Institute for Plasma Physics (IPP), Garching, Germany

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

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


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O1A .4

Integrated current profile, normalized beta and NTM control in DIII-DPajares, Andres1; Schuster, Eugenio1; Wehner, William P .1; Eidietis, Nicholas2; Welander, Anders2; La Haye, Robert2; Ferron, John2; Barr, Jayson2; Walker, Michael2; Humphreys, David2; Hyatt, Al2

1Lehigh University, Bethlehem, United States2General Atomics, San Diego, United States

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

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


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O1A .5

Conceptual design of the COMPASS-U tokamakPanek, Radomir1; Havlicek, Josef1; Hron, Martin1; Dejarnac, Renaud1; Komm, Michael1; Urban, Jakub1; Weinzettl, Vladimir1; Adamek, Jiri1; Bilkova, Petra1; Bohm, Petr1; Casolari, Andrea1; Ficker, Ondrej1, 2; Grover, Ondrej1, 2; Horacek, Jan1; Imrisek, Martin1, 3; Jaulmes, Fabien1; Peterka, Matej1, 3; Kripner, Lukas1, 3; Markovis, Tomas1, 3; Tomes, Matej1, 3; Varju, Josef1; Vondracek, Petr1, 3

1Institute of Plasma Physics of the Czech Academy of Sciences, Prague, Czech Republic2Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, Prague, Czech Republic3Faculty of Mathematics and Physics, Charles University in Prague, Prague, Czech Republic

The Institute of Plasma Physics of the CAS in Prague has recently started construction of new COMPASS-U tokamak . It will be a compact, medium-size (R = 0,85 m, a = 0,3 m), high-mag-netic-field (5 T) device . COMPASS-U will be equipped by a flexible set of poloidal field coils and capable to operate with plasma current up to 2 MA and, therefore, high plasma density (~ 10^20 m^-3) . The device is designed to generate and test various DEMO relevant magnetic configura-tions, such as conventional single null, double null, single and double snow-flake . The plasma will be heated using 4 MW Neutral Beam Injection (NBI) heating system with future extension by at least 4 MW Electron Cyclotron Resonant Heating (ECRH) system .

COMPASS-U will be equipped with lower and upper closed, high neutral density divertors . Due to high PB/R ratio COMPASS-U will represent a device which will be able to perform ITER and DEMO relevant studies in important areas, such as the plasma exhaust or development of new confinement regimes . The divertors will use conventional materials in the first stage, however, in the later stage, the liquid metal technology, which represents a promising solution for the power exhaust in DEMO, will be installed into the lower COMPASS-U divertor . The metallic first wall will be operated at high temperature (approx . 300 °C) during plasma discharge, which will enable to explore the edge plasma regimes relevant to ITER and DEMO operation . The first plasma is scheduled for 2022 .

In this contribution, we will present the conceptual design of the COMPASS-U tokamak as well as the main tokamak components .


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O1A .6

Development of HINEG and its experimental campaignsWu, Yican; Wang, Yongfeng; Liu, Chao; Wang, Zhigang; Li, Taosheng; Jiang, Jieqiong; Long, Pengcheng; Hu, Liqin; Wang, Jianye; Team, FDS

Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, China

Fusion energy becomes essential to solve the problem of increasing energy demands . A high inten-sity D-T fusion neutron generator is keenly needed for the research and development (R&D) of fusion technology, especially for fusion materials research .

The Institute of Nuclear Energy Safety Technology (INEST), Chinese Academy of Sciences (CAS) has launched the High Intensity D-T Fusion Neutron Generator (HINEG) project . The R&D of HINEG includes three phases: HINEG-I has been constructed and successfully produced a D-T fu-sion neutron yield of up to 6 .4E12 n/s . The mechanism research of irradiation damage for materi-als can be carried out . HINEG-II aims at a high neutron yield of 1E15~1E16 n/s neutrons via high speed rotating tritium target system and high intensity ion source, which could be used to conduct material irradiation damage testing . The preliminary design and research on key technologies are on-going . HINEG-III is a volumetric fusion neutron source with yield of more than 1E18 n/s . The integration testing of nuclear system engineering could be performed .

As an important platform for fusion technology and safety research, HINEG can be used to car-ry out the neutron activation and irradiation testing not only for structural but also functional materials to assess the performance and reliability, such as structural materials for the blanket, neutron multipliers and ceramic breeders for tritium fuel production, suitable radiation resistant thermosets for the electrical insulation of the superconducting magnets, in-vessel conductor coils, liquid-metal coolants, etc . The fusion neutron irradiation testing is being conducted on China Low Activation Martensitic (CLAM) steel, which has been developed by INEST and selected as the primary candidate structural material for Chinese Helium Cooled Ceramic Breeder ITER Test Blanket Module (CN HCCB TBM) . Moreover, the performance of components under neutron irradiation can also be assessed on HINEG platform, such as the tritium breeding blanket and shielding blanket .


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O1A .7

Neutron spectrum determination at the ITER material irradiation stations at JETPacker, Lee1; Batistoni, Paola2; Bradnam, Steven1; Colling, Bethany1; Conroy, Sean3; Ghani, Zamir1; Gilbert, Mark1; Jedonorg, Slawomir4; Laszynska, Ewa4; Leichtle, Dieter5; Lengar, Igor6; Mietelski, Jerzy7; Misiak, Ryszard7; Pillon, Mario2; Popovichev, Sergei1; Radulovic, Vladimir6; Stamatelatos, Ion8; Vasilopoulou, Theodora8; Wójcik-Gargula, Anna7

1Nuclear Technology, UKAEA, Abingdon, United Kingdom2Department of Fusion and Technology for Nuclear Safety and Security, ENEA, Frascati, Italy3Department of Physics and Astronomy, Uppsala University, Uppsala, Sweden4Nuclear dept, Institute of Plasma Physics and Laser Microfusion, Warsaw, Poland5Nuclear safety, Fusion for Energy, Barcelona, Spain6Reactor Physics Department, Josef Stefan Institute, Ljubljana, Slovenia7Institute of Nuclear Physics, Polish Academy of Sciences, Krakow, Poland8Technology, Energy and Safety, Institute of Nuclear and Radiological Sciences, Athens, Greece

The experiments that are planned over the next few years at the Joint European Torus (JET), no-tably including a deuterium-tritium (DT) experimental phase, are expected to produce large neu-tron yields, up to 1 .7E21 neutrons . The scientific objectives of the experiments are linked with a technology programme, WPJET3, to deliver the maximum scientific and technological return from those operations, with particular emphasis on technology exploitation via the high neutron fluxes predicted in and around the JET machine . Importantly, the programme aims to extract experi-mental data relevant to the international effort to design, construct and operate ITER . The data expected to be retrieved under the JET experimental program will support, develop and improve the radiation transport and activation simulation capabilities via benchmarking and validation in relevant operational conditions . Such capabilities are important and are applied extensively to predict a wide range of nuclear phenomena and impacts associated with components and materials that will be used in ITER operations .

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


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O1B .1

Myth of Initial Loading Tritium-2 : Practical commissioning strategy of DEMO-Japan without external sourceKonishi, Satoshi1; Hiwatari, Ryoji2

1Institute of Advanced Energy, Kyoto University, Uji, Japan2Department of Fusion Reactor Systems Research, National Institute for Quantum Science and Technology Research, Aomori, Japan

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


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O1B .2

Manufacturing of the first ITER Torus CryopumpDremel, Matthias1; Canadell, Francina2; Pearce, Robert1; Quinn, Eamonn1; Chitu, Florin1

1PED, ITER Organization, 13067 St Paul Lez Durance Cedex, France2F4E, EU Domestic Agency, 08019 Barcelona, Spain

The first of nine ITER Torus and Cryostat Cryopumps has been successfully manufactured and delivered to ITER in summer 2017 . This Pre- Production Cryopump is the first of the ITER cry-opumps and may be used for the first pump down of the vacuum vessel or the cryostat .

The pump has a 1 .8 m diameter and a length of about 3 m and contains cryogenic pressure equip-ment with a charcoal coated adsorption stage integrated in a casing with a vacuum vessel plug combined with the largest all-metal vacuum valve ever built resulting in an overall weight of 8 tons . The design of the cryopump is the result of more than ten years of research and development finalized to the ITER built to print design to comply with the demanding requirements to be ful-filled during its operation starting from first plasma .

The paper will outline the experience gained with the cryopumps manufacture and assembly . We discuss the components with confinement function as ITER style vacuum flanges, double bellows for the valve assembly and electrical feedthroughs . Many different requirements had to be ad-dressed for the manufacture of these components and their integration in the cryopump .

ITER is a Nuclear Facility, INB-174, and requirements for the operation in the primary vacuum and the nuclear confinement function demand a high level of quality control and inspection needs during all manufacturing stages . The Pre-Production Cryopump has been built in close and suc-cessful cooperation with F4E reflected in the adequate surveillance of the 27 suppliers required to fabricate the cryopump . The successfully built and delivered Pre-Production Cryopump will give a reliable basis for the F4E procurement of the eight Torus and Cryostat Cryopumps which are required for first plasma .

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


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O1B .3

Neutronics of the IFMIF-DONES irradiation facilityFischer, Ulrich1; Bienkowska, Barbara2; Drozdowicz, Krzysztof3; Frisoni, Manuela4; Mota, Fernando5; Ogando, Francisco6; Qiu, Yuefeng1; Stankunas, Gediminas7; Tracz, Grzegorz3

1Karlsruhe Institute of Technology, Eggenstein-Leopoldshafen, Germany2Institute for Plasma Physics and Laser Microfusion (IPPLM), Warsaw, Poland3Institute of Nuclear Physics (IFJ PAN), Cracow, Poland4ENEA-Bologna, Bologna, Italy5Fusion National Laboratory, CIEMAT, Madrid, Spain6Universidad Nacional de Educación a Distancia (UNED), Madrid, Spain7Lithuanian Energy Institute (LEI), Kaunas, Lithuania

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

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

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

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


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O1B .4

Innovative Technology for 6Li Enrichment using Electrodialysis with Lithium Ionic ConductorHoshino, Tsuyoshi

Fusion Energy Research and Development Directorate, National Institutes for Quantum and Radiological Science and Technology (QST), Rokkasho-mura, Kamikita-gun, Japan

Tritium needed as a fuel for fusion reactors is produced via neutron capture by lithium-6 (6Li) . However, natural Li contains only about 7 .8% 6Li, and enrichment of 6Li up to 90% is required for adequate tritium breeding in fusion reactors . In Japan, lithium isotope enrichment methods have been developed to avoid the environmental hazards of using mercury . However, the isotope separa-tion coefficient and efficiency is too low to meet the practical need of large mass production of 6Li .

Therefore, new Li isotope separation technique using a Li ionic superconductor functioning as a Li isotope separation membrane (LISM) have been developed . First of all, I investigated the ionic mobility of lithium isotopes in ionic superconductor . Combing the first principle and the kinetics Monte Calro simulation, I calculate the diffusion constant of 6Li and 7Li .

Furthermore, examinations of Li isotope separation using LISM with electrodialysis ware per-formed . Because the mobility of 6Li ions is higher than that of 7Li ions, 6Li can be enriched on the cathode side of a cell . Using Li0 .29La0 .57TiO3 (LLTO) as the Li ionic superconductor was prepared . After electrodialysis, I obtained a maximum of 1 .05 for the 6Li isotope separation coeffi-cient . This result showed that the 6Li isotope separation coefficient of this method is the same as that of the amalgamation process using mercury (1 .06) .


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O1B .5

Status of the EU DEMO breeding blanket manufacturing R&D activitiesForest, Laurent1; Aktaa, Jarir2; Virgilio Boccaccini, Lorenzo2; Emmerich, Thomas2; Eugen-Ghidersa, Bradut2; Froio, Antonio3; Namburi, Hygreeva4; Neuberger, Heiko2; Li Puma, Antonella1; Rey, Jörg2; Savoldi, Laura3; Sornin, Denis5; Vala, Ladislav4

1DEN-Service d‘études mécaniques et thermiques (SEMT), CEA, Université Paris-Saclay, Gif sur yvette, France2KIT, Karlsruhe Institute of Technology, Karlsruhe, Germany3Dipartimento Energia, Politecnico di Torino, NEMO group, Torino, Italy4Centrum výzkumu Řež, Husinec, Czech Republic5DEN- Service de Recherches Métallurgiques Appliquées (SRMA), CEA, Université Paris-Saclay, Gif sur yvette, France

The realization of a DEMOnstration Fusion Power Reactor (DEMO) to follow ITER, with the ca-pability of generating several hundred MW of net electricity and operating with a closed fuel-cycle by 2050, is viewed by Europe as the remaining crucial step towards the exploitation of fusion pow-er . The EUROfusion Consortium, in the frame of the European Horizon 2020 Program, is assessing four different breeding blanket concepts in view of selecting the reference one for DEMO . This paper describes technologies and manufacturing scenarios developed and envisaged for the four blanket concepts, including nuclear “conventional” assembly processes as TIG, electron beam and laser welding, Hot Isostatic Pressing (HIP), and also more advanced (from the nuclear standpoint) technologies as additive manufacturing techniques .

With regard to welding processes, topics as the metallurgical weldability of EUROFER steel and the associated risks or the development of appropriate filler wire are discussed .

The development of protective W-coating layers on First Wall, with Functionally Graded (FG) interlayer as compliance layer between W and EUROFER substrate, realized by Vacuum Plasma Spraying method, is also propounded . First layer systems showed promising layer adhesion, ther-mal fatigue and thermal shock properties . He-cooled mock-ups, representative of the First Wall with FG W/EUROFER coating will be fabricated for test campaigns in the HELOKA facility under relevant heat fluxes .

First elements of Double Walled Tubes (DWT) manufacturing and tube/plate assembly for the water cooled concept are given, comprising test campaign aiming at assessing their behaviour un-der corrosion .

Developments described in the paper are performed in conformity with international standards and/or design/manufacturing codes .

Eventually, further development strategies are suggested .


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O1B .6

Experimental refutation of the deuterium permeability in vanadium, niobium and tantalumMalo, Marta; Garcinuño, Belit; Rapisarda, David

National Fusion Laboratory, CIEMAT, Madrid, Spain

Unique gas retention and transport characteristics of group V elements (V, Nb, Ta) have long attracted a significant interest, in particular among the nuclear fusion community . The nominally high hydrogen isotope permeability and diffusion at the expected operational temperatures, to-gether with the negative activation energy for the solubility present these materials as a promising choice for the fabrication of tritium recycling structures .

However, before seriously considering these materials, one should question the accuracy of the available data, given the remarkable lack of direct experimental measurements in support of the traditionally accepted transport properties of these materials . Furthermore, it must be considered that data have been mostly obtained by combining results obtained by different authors and methods .

The extensive literature review presented in this paper shows that existing experimental results not only contradict the semi-empirical values assumed for these materials but also present a broad dispersion .

In order to clarify this, deuterium permeability data for the three materials was obtained at the THERMOPERM facility at Ciemat (Madrid, Spain) in a relevant range of pressures and temper-atures . Experimental difficulties together with the role of surface oxidation which may become a major issue for practical uses are also assessed .


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O1B .7

Multifunctional nanoceramic coatings for future generation nuclear systemsDi Fonzo, Fabio1; Vanazzi, Matteo1; Iadicicco, Daniele1; Utili, Marco2; Bassini, Serena2; Tarantino, Mariano2; Hernandez, Teresa3; Morono, Alejandro3; Munoz, Patricia3

1Center for Nano Science and Technology, Istituto Italiano di Tecnologia, Milano, Italy2ENEA FSN-ING C.R. Brasimone, ENEA, Camugnano (BO), Italy3Fusion Technology Division, CIEMAT, Madrid, Spain

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


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O1C .1

Deep Learning: Towards Autonomous Remote MaintenanceSkilton, Robert

Remote Applications in Challenging Environments (RACE), UK Atomic Energy Authority, Abingdon, Uni-ted Kingdom

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

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

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


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O1C .2

Reconstructing JET using LiDAR-vision fusionJonasson, Emil1; Kyberd, Stephen2; Boeuf, Jonathan2; Skilton, Robert1; Burroughes, Guy1; Collins, Steve1; Amayo, Paul2

1RACE - Remote Applications in Challenging Environments, UK Atomic Energy Authority, Abingdon, United Kingdom2Oxford Robotics Institute, University of Oxford, Oxford, United Kingdom

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

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

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


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O1C .3

Engineering and integration design risks arising from advanced magnetic divertor configurationsKembleton, Richard1; Federici, Gianfranco2; Ambrosino, Roberto3; Maviglia, Francesco2; Siccinio, Mattia2; Reimerdes, Holger4; Ha, Samuel5; Merriman, Samuel5

1EUROfusion, Abingdon, United Kingdom2PPPT, EUROfusion PMU, Garching, Germany3Consorzio CREATE, Università Federico II di Napoli, Napoli, Italy4EPFL, SPC, Lausanne, Switzerland5CCFE, Culham Science Centre, Abingdon, United Kingdom

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

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

In this paper we describe the work being undertaken, and main results so far, in assessing the im-pact of incorporating these alternative configurations into DEMO whilst respecting requirements on remote ha

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