MARTIN P. et al.

1

OVERVIEW OF THE DIVERTOR TOKAMAK TEST FACILITY PROJECT

P. MARTIN Università degli Studi di Padova and Consorzio RFX Padova, Italy Email: martin@igi.cnr.it

R. ALBANESE Università degli Studi di Napoli Federico II and Consorzio CREATE Napoli, Italy Email: raffaele.albanese@unina.it F. CRISANTI ENEA, Dipartimento Fusione e Sicurezza Nucleare, Frascati,Italy Email: flavio.crisanti@enea.it A. PIZZUTO ENEA, Dipartimento Fusione e Sicurezza Nucleare, Frascati,Italy Email: aldo.pizzuto@enea.it THE DTT TEAM

Abstract

The Divertor Tokamak Test Facility (DTT) is a new tokamak whose construction has recently been approved by the Italian government. 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=2.10 m, minor radius a=0.65 m and average triangularity 0.3. The main role of DTT is to contribute to the development of a reliable solution for the power and particle exhaust in a reactor, a challenge commonly recognised as one of the major issues in the road map towards the realisation of a nuclear fusion power plant. Following the project approval, since June 2017 the design review of DTT has started. This paper will present the device by summarizing its main physics goals and the present status of the design.

.

1. INTRODUCTION

The Divertor Tokamak Test Facility (DTT) is a new tokamak whose construction has recently been approved by the Italian government. DTT will be a high magnetic field superconducting toroidal device (6 T) carrying plasma current up to 5.5 MA in pulses with length up to 100s, with an up-down symmetrical D-shape defined by major radius R=2.10 m, minor radius a=0.65 m and average triangularity 0.3. The auxiliary heating power will be 45 MW.

The main role of DTT is to contribute to the development of a reliable solution for the power and particle exhaust in a reactor, a challenge commonly recognised as one of the major issues in the road map towards the realisation of a nuclear fusion power plant. The divertor physics and technology issues will be tackled integrated with state of the art plasma scenarios, which will allow DTT to be a key player for the development of core physics, too.

After an extensive activity, which involved Italian labs and scientists from other European labs, the DTT preliminary design report was released in June 2015 [1] and later documented in a series of journal papers [2]. In the first semester of 2017 the Italian government identified the funding strategy of the experiment and authorized its start. In October 2017 EUROfusion noted that the DTT proposal may provide important elements for finding solutions to the plasma exhaust problem delivering information in operational ranges or

IAEA-CN- FIP/P8-31

configurations that are not accessible for the present devices or JT-60SA and approved its involvement in the DTT facility at a date around 2022-23. In April 2018 the DTT site has been selected to be in Frascati.

Figure 1: global view of the DTT device This paper will present the device by summarizing its main physics goals and the present status of the

design.

2. BACKGROUND AND SCOPE

The controlled exhaust of energy and particle from a fusion reactor is a difficult issue that has to be solved before starting the design of DEMO. According to experimental and theoretical work (see for example [3]) one of the major risks comes from the size of the scrape off layer (SOL) power flow decay length lq, which – from data taken in existing experiments - scales like [4]

𝜆" 𝑚𝑚 = 1,35𝑃+,-././1𝑅/./3𝐵5./.63𝜀/.31 1

where PSOL is the power in the scrape-off-layer, R the major radius of the machine, Bp the poloidal magnetic field and ε the aspect ratio.

If this scaling holds true for ITER, it means that in that device lq is expected to be of the order of 1 mm. Considering the Q=10 scenario, with 500 MW of net fusion power (400 MW brought by neutrons and 100 MW lost through radiation and particles) the expected power exhausted on the divertor in a low radiation case is about 90 MW. This means a heat flux on the divertor ~ 50 MWm-2, a value far above the limit of present target materials. The solution to this issue is based on three lines of science and technology activities: — Development of plasma facing components to cope with very large power fluxes. At the moment, the

technological limit is ~ 10-20 MWm-2; — Development of the physics understanding of the exhaust process – in particular through first principle

models – and of the divertor design. Aims of divertor studies are the study of modifications of its magnetic topology (e.g. double null, snowflakes, X-divertor, Super-X divertor configurations) to decrease the divertor plates power flux by increasing the divertor “wetted” surface, and the exploitation of liquid metals as plasma facing components;

— Development of plasma scenarios where plasma energy is removed by means of radiation before it reaches the plasma facing components while keeping optimized plasma core performance.

To cope with these challenges and following the recommendations of the EUROfusion roadmap [5] the Italian fusion community has proposed in 2015 the DTT experiment. Its main physics and technology goals and requirements are the following:

Physics: — Matching 4 DEMO relevant parameters: the electron temperature Te the normalized collisionality ν*, the

ratio between the SOL thickness and the neutral mean free path Δd/λ0 and the plasma pressure normalized to the magnetic one β;

MARTIN P. et al.

3

— Relaxation of the normalized Larmor radius: ρ*reactor=ρ*

scaledRg, where R is the major radius and g is a controlled scaling parameter, which has been chosen as g = 0.75;

— Availability of integrated scenarios, so that the power exhaust options to be extrapolated to DEMO will be studies in core/edge plasma conditions relevant to DEMO.

Technology: — Matching the PSEP/R values, where Psep is the power flowing through the last closed magnetic surface, with

ITER and DEMO. This means that the machine has to be designed with PSEP/R ≥ 15 MW/m; — Flexibility for divertor choices, which calls for a design that guarantees the possibility to allocate in the

vessel and to test several divertor options, that keeps high flexibility of the poloidal coils to test alternative magnetic configurations

— Possibility to test liquid metals and to close the loop, i.e. to collect liquid metal on the first wall. The surface temperature of the first wall should be between 200-300 °C to guarantee the flow of the liquid metal on it.

— the power exhaust alternatives should be compatible with technological constraints of DEMO (e.g. plasma bulk performances, poloidal field coil system, materials, space for the blanket and neutron shielding).

In addition to the aforementioned requirements, the DTT device design has evolved since its 2015

version in order to provide a complete up-down symmetry, to allow the study of double null (DN) divertor configurations and to improve its portfolio of auxiliary heating systems so that the maximum power coupled to the plasma is 45 MW, distributed among ECRH, ICRH and NBI. The up-down symmetry will be implemented for the superconducting toroidal field and poloidal field coils, in-vessel coils, first wall, divertor structures, vacuum vessel, portholes (Fig. 2).

Figure 2: left, the 2015 DTT vessel design. Right, the 2018 design, fully up-down symmetric

3. SURVEY OF MAIN ENGINEERING CHOICES

3.1. Radial build

With respect to the original 2015 design [1,2], the device is being redesigned to make it fully up-down symmetric. In order to keep the cost constant, some minor changes in the machine size and in the plasma current where necessary. The main technical and target physics quantities for the present DTT design are summarized in Table I, where for comparison the 2015 design values are also shown. No major impact of this changes on the plasma performance is expected.

The inboard radial build takes into account the inboard radius of the plasma (1.45 m), a minimum clearance of 30 mm between plasma and first wall, and 62.5 mm for the first wall structure and the in-vessel magnetic diagnostics. On the inboard side, the vacuum vessel is made by a two shell SS AISI 316 LN structure. Each shell is 15 mm thick and they are separated by a neutron shield 95 mm thick with 70 mm of borated water. A 25 mm thick thermal shield operating at 80 K is located behind the vacuum vessel with a gap of 20 mm. The structure of the inner legs of the TF (toroidal field) coils spans from 820 mm to 1175 mm. The central solenoid is made by six identical modules wound in three layers, spanning from 429 to 762 mm.

IAEA-CN- FIP/P8-31

Plasma parameters 2015 2018 Major radius R (m) 2.15 2.10 Minor radius a (m) 0.70 0.65 Aspect ratio A 3.0 3.2

Volume V (m3) 33 28 Plasma current Ip (MA) 6.0 5.5

Magnetic field on axis Btor (T) 6.0 6.0 Confinement enhancement H98 1.0 1.0

Table 1. Main plasma parameters of DTT original proposal (2015) and design review (2018).

On the outboard side of the device the radial build is guided by the requirement on the magnetic field

ripple, such that (Bmax-Bmin)/2B0 < 0.5%. This requires that the outboard TF coil leg in the equatorial plane spans from 1622 to 3953 mm. This leaves significant room, since the first wall is located at a radius of 2.86 m, i.e. with a gap of 110 mm from the nominal plasma. The advantage of this is an ample flexibility for alternative plasma shapes. The vacuum vessel on the outboard has two 15 mm thick shells, with an inner radius of 3.17 m, leaving space for the in-vessel coils and the magnetic diagnostics. The cavity between the shells where the borated water flows is larger than the inboard (16-20 cm).

3.2. Magnet system

To guarantee 100 s pulses, the toroidal magnet and the central solenoid will be realized by superconducting Nb3Sn and the rest of the poloidal coils by Nb3Sn. The magnet system [6] includes 18 Nb3Sn CICC (Cable-In-Conduit Conductor) TF coils operating with 26.9 kA in 130 turns at a maximum field of 11.7 T. The layer wound option selected in 2015 has been replaced by Double-Pancake winding, which simplifies coil manufacturing. The 6 NbTi PF (poloidal field) coils are designed so to have up-down symmetry. They can work at a maximum field of 5.5 T, with a current of 30 kA. The number of turns is 324 in coils PF1 and PF6, 160 in coils PF2 and PF5, and 196 in coils PF3 and PF4. The central solenoid consists of 6 identical modules wound in three layers and is designed to allow for the insertion of an additional high temperature superconductor (HTS) coil. This HTS addendum will allow for 10% flux increase and will work as test bed for the development of the next generation magnets for fusion applications.

3.3. Vacuum vessel and cryostat

The machine core is surrounded by the Cryostat Vessel, a 40 mm thick vacuum tight container, which provides the vacuum for the superconducting magnets and forms part of the secondary confinement barrier. The double shell stainless steel SS AISI 316 LN vacuum vessel is located inside the magnet system and is made by 18 welded sectors. Each shell of the vessel is 15 mm thick, and they are separated by a neutron shield where borated water flows at 50-80 °C [7,8]. The thickness of the neutron shield is 95 mm inboard (with 25 mm of solid components) and 160-200 mm outboard. This shield is designed for a DD neutron yield rate of 1×1017 n/s - plus 1×1015 n/s of 14 MeV neutron production due to triton burn-up - during the high-performance phase. It will allow to keep the nuclear heating of the first toroidal field coil layer below 1 mW/cm3 [7,9]. The vessel ports are up-down symmetric both in the 6 sectors dedicated to upper and lower divertor remote handling and in the remaining 12 sectors reserved for heating and diagnostics.

Exposed to the plasma the first wall is designed with coaxial pipes [10]. Two alternative concepts are considered, one using CuCrZr and one in stainless steel, both coated with a 3 mm thick tungsten layer deposited by plasma spray technique. The first wall will be compatible with liquid metal divertors, which require heating at a temperature above 200 °C to avoid liquid metal condensation.

3.4. Divertor

The divertor system will be a key asset of the device and the central component to perform the DTT scientific mission. The DTT divertor will be realized in collaboration with the EUROfusion “Plasma Exhaust (PEX)” initiative. The final design will be influenced by a series of dedicated experiments on Asdex Upgrade,

MARTIN P. et al.

5

TCV and MAST-Upgrade and it is expected to be finalized around 2022-23. For this reason, the device load assembly is presently being designed with enough flexibility to be able to accommodate various types of divertors. Presently different types of materials for the divertor plates are under consideration: the ITER like tungsten mono-blocks and liquid metals (Li or Sn).

As a preliminary example, Fig. 3 reports the results of magnetic equilibrium simulations, showing that DTT is capable to produce Double (a) and Single Null (b) configurations @5.5 MA, Snowflake divertor (c) @4.5MA, Single Null with negative triangularity (d) @5.5MA and super X divertor @3MA €.

Figure 3: divertor magnetic configurations feasible in DTT

3.5. In-vessel coils

Inside the vacuum vessel several coils will be installed. In particular: — 2 independent n=0 copper coils and 2 stabilizing plates for vertical stabilization, fast radial control during

breakdown and H-L transitions, and double null wobbling; — 4 independent n=0 divertor coils for magnetic configuration control in the divertor region and strike point

sweeping; — n>0 ELM/RWM coils, whose design is still ongoing

3.6. Additional heating.

The availability of additional heating will follow a staged approach, with a final target of 45 MW coupled to the plasmas [11]. The heating mix will be provided by ECRH, ICRH and NBI. The precise sharing among the three sources is still under study, together with the possibility of double frequency for the ECRH (140-170 GHz). The ranges that are presently considered for the full power operation are 3-9 MW of ICRH, 20-30 MW ECRH (@170 GHz) and 7.5-15 MW of negative NBI @400 keV. In the initial phase 25 MW of additional power will be coupled to the plasma, with the following heating mix: 15 MW of ECRH with 16 gyrotrons, 3 MW of ICRH, and 7.5 MW of negative neutral beam injection provided by a single tangential injector.

3.7. Disruption mitigation

Electromagnetic loads on the vacuum vessel due to the current quench in a plasma disruption have been studied, and on the vessel the magnitude of the force is predicted to be close to 106 N. Given the outmost importance of the disruption issue for ITER, the option to equip DTT with a disruption mitigation system which could be used as test-bed for future devices is under investigation.

4. PRELIMINARY PLASMA PARAMETERS

Following the revision of the machine parameters, the updated expeted plasma parameters are shown in Table 2. For comparison the first two columns of this table report the design ITER and DEMO relevant parameters. Note that the key Psep/R has not changed.

IAEA-CN- FIP/P8-31

ITER DEMO 2015 proposal

Revised proposal

Tped(KeV) 4.3 5.5 3.1 3 nped(1020m-3) 0.8 0.6 1.4 1.4 ν*

ped (10-2) 2.3 1.4 2.4 2.5 ELMs En. (MJ) 24 85 1.2 1.0 L-H Pow (MW) 60-100 120-200 16-22 15-19 Psep (MW) 87 150 32 32 lint 2.2 2.2 1.7 1.7 PDIV(MW/m2) (no Rad) 55 84 54 79 PDIV(MW/m2) (70% Rad)

27 42 27 24

q// »PTOTB/R (MWT/m) 100 290 125 128 Pulse length (s) 400 7600 100 95

Table 2. Main edge parameters of DTT (2015 and 2018 versions), compared with ITER and DEMO.

5. CONCLUSIONS

The design of the divertor tokamak test facility has started, following the approval of the project by the Italian government. The superconducting tokamak will be realized in Frascati. The design allows ample flexibility in terms of divertor systems. The device will represent a unique facility to study in an integrated way the plasma exhaust issues in high performance tokamak scenarios and it will be a key actor to support ITER operation and DEM design.

MARTIN P. et al.

7

1 http://fsn-fusphy.frascati.enea.it/DTT/downloads/Report/DTT_ProjectProposal_July2015.pdf 2 Special Section on "DTT. Divertor Tokamak Test facility", Fusion Eng. Des., Volume 122, Pages 253-394, e1-e26, Nov. 2017, http://www.sciencedirect.com/science/journal/09203796/122 3 LOARTE A., NEU G. Fusion Engineering and Design, Volume 122, November 2017, Pages 256-273 4 EICH T. et al. Nuclear Fusion 53 (2013) 093031 5 https://www.euro-fusion.org/eurofusion/roadmap/ 6 DI ZENOBIO A. et al., “Updated Conceptual Design of the DTT magnet system”, presented at SOFT 2018 Conference, Giardini Naxos, Italy, Sept. 2018. 7 MAZZITELLI G. et al., “Role of the Italian DTT in the Power Exhaust implementation strategy”, presented at SOFT 2018, Giardini Naxos, Italy, Sept. 2018. 8 DI ZENOBIO A. et al., “Updated Conceptual Design of the DTT magnet system”, presented at SOFT 2018, Giardini Naxos, Italy, Sept. 2018. 9 COLANGELI A. et al., “Neutronics study for DTT tokamak building”, presented at SOFT 2018, Giardini Naxos, Italy, Sept. 2018. 10 MAVIGLIA F. et al., “Thermal-hydraulic analysis for first wall and vacuum vessel thermal shield of Divertor Tokamak Test facility, presented at SOFT 2018, Giardini Naxos, Italy, Sept. 2018 11 GRANUCCI G. et al., “The heating systems capability of DTT”, presented at SOFT 2018, Giardini Naxos, Italy, Sept. 2018