dtt: a divertor tokamak test facility for the study …...von mises stress ok for a 2.9 mm 316 ln...

DTT: a Divertor Tokamak Test facility

for the study of the power exhaust

issues in view of DEMO

R. Albanese, ENEA-CREATE (Italy)

on behalf of the WPDTT2 Team & the DTT report contributors

(work carried out in tight cooperation with WPDTT1)

R. Albanese | 1st IAEA Technical Meeting on Divertor Concepts | 29 SEPT. – 2 OCT. 2015 | PAGE 2

Outline

• WPDTT2: objectives, organization, activities

• Project status

− DTT requirements

− Choice of parameters

− DTT proposal

• Conclusions


WPDTT2: objectives, structure, schedule

http://users.jet.efda.org/iterphysicswiki

www.create.unina.it/dtt2

http://fsn-fusphy.frascati.enea.it/DTT

Objective: design an experiment addressed to the solution

of the power exhaust issues in view of DEMO. This derives

from the need to develop integrated, controllable exhaust

solutions for DEMO including plasma, PFCs, control

diagnostics and actuators, using experiments, theory and

modelling, so as to mitigate the risk that conventional

divertor might not be suitable for DEMO.

About 60 MEUR not yet allocated (mostly for HW)

8 RUs:

See also I-11: H. Reimerdes

http://www.create.unina.it/dtt2







Steps of WPDTT2 Project

Phase I

FIRST TWO STEPS

MAINLY IN SUPPORT

TO PHYSICS ACTIVITIES

OF WPDTT1

Phase II

FURTHER STEPS

MORE FOCUSED

ON DESIGN ACTIVITY

WITH THE SUPPORT

OF WPDTT1

START OF STEP 4

ANTICIPATED TO

APRIL 2015

TIGHT COOPERATION,

SAME PB,

COORDINATED

TIME PLAN,

SMOOTH HANDOVER

WPDTT1/WPDTT2



WPDTT2: objectives, structure, schedule O

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Project status : DTT requirements

• Preservation of 4 dimensional or dimensionless parameters: Te , *=Ld/λei,

Δd/λ0 , β + relaxation on normalized Larmor radius (ρi/Δd)*R 0.75

• Psep/R 15 MW/m

• Flexibility in the divertor region so as to possibly test several divertors

• Possibility to test alternative magnetic configurations

• Possibility to test liquid metals

• Integrated scenarios (solutions to be compatible with plasma

performance and technological constraints of DEMO)

• Budget constraint (within a reasonable cost)

R. Albanese, F. Crisanti, B. P. Duval, G. Giruzzi, H. Reimerdes, D. van Houtte, R. Zagorski,

“DTT - An experiment to study the power exhaust in view of DEMO”,

Presented at the3rd IAEA DEMO Programme Workshop (DPW-3) , Hefei, China, 11-15 May 2015


Project status : Choice of parameters

MAIN DTT PARAMETERS FOR THE REFERENCE SINGLE NULL SCENARIO

R (m) 2.15 bN 1.5

a (m) 0,7 tRes (sec) 8

IP (MA) 6 VLoop (V) 0.17

BT (T) 6 Zeff 1.7

V (m3) 33.0 PRad (MW) 13

PADD (MW) 45 PSep (MW) 32

H98 1 TPed (KeV) 3.1

<ne> (1020 m-3) 1.7 nPed (1020 m-3) 1.4

ne/neG 0.45 bp 0.5

<Te> (KeV) 6.2 PDiv (MW/m2) (No Rad) ~ 55

t (sec) 0.47 PSep/R (MW/m) 15

ne(0) (1020 m-3) 2.2 PTotB/R (MW T/m) 125

Te(0) (KeV) 10.2 λq (mm) ~ 2.0


Project status : Comparison with other devices

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Project status : DTT Program


Project status: DTT proposal

TD03-I: WP DTT2 Progress Report on the

preconceptual “baseline “design of DTT

EFDA_D_ 2D3KX2 (Apr. 2015)

https://idm.euro-fusion.org/?uid=2D3KX2

Italian EFSI proposal for DTT submitted to

the Italian Government (July 2015)


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Plasma scenarios

• Plasma-wall gaps 40 mm (power decay length at 6 MA is 2 mm at the outboard midplane);

• plasma shape parameters similar to the present design of DEMO: R/a≈3.1, k≈1.76, <δ>≈0.35;

• pulse length of more than 100 s (total available flux ≈ 45 Vs, Central Solenoid swing ≈ 35 Vs).

6 MA

SN

scenario



Conventional and alternative

magnetic configurations that

can be obtained using the DTT

PF system.

CS, PF and TF coils are

superconducting: plasma pulse

duration ~ 100 s without current

drive



Internal copper coils

can be used for

plasma control or

local modifications of

the magnetic

configuration in the

divertor region


See also P-2: F. Crisanti


Magnet system: CS, PF coils and TF Coils

18 TF coils: Bpeak: 12.0 T, Bplasma: 6.0 T, 65 MAt;

6 CS coils: Bpeak: 12.5 T, k |N kI k| =51 MAt; available poloidal flux: 17.6 Vs;

6 PF coils: Bpeak: 4.0 T, k |N kI k| =21 MAt.

CS, PF coils and TF Coils

in-vessel coils


DTT


Each of the 18 D-shaped TF coils has78 turns of Nb3Sn/Cu CIC conductor, carrying 46.3kA He

cooled (inlet T of 4.5K): max field 11.4 T, max ripple on the plasma 0.8%

Graded solution: Cable-In-Conduit (CIC) conductor layouts: 48 LF

turns with thicker 316 LN jacket and lower SC strand number, 30

HF turns. section wound in pancakes to reduce the He path

NI=65 MAt, Wm=1.96 GJ, Tmarg= 1.2 K

von Mises stress OK (<650 Mpa in 3D analyses)

Thotspot also OK (104 K all materials, 268 K Cu & SC only)

Based on ITER-like strands with slightly optimized performances, only

20% higher, which should be achievable

Jmax ~1.8 higher than ITER: possible SULTAN or EDIPO test facility

for both HF & LF grade and the test of full-size joints

If needed, a small reduction of Bmax by 5% would increase current

density limit by 20% in the HF grade and 10% in the LF grade



NI=51 MAt, Flux swing of 35 Vs, Tmarg= 1.5 K

von Mises stress OK for a 2.9 mm 316 LN jacket*: 346 Mpa

Thotspot also OK (86K all materials, 229K cable only)

DTT CS coil assembly

The CS operates at 12.5 T (13.2 T peak on the SC) and consists of 6 independent

modules based on Nb3Sn CICCs: 23 kA, 2220 turns (2x270+4x420).

ITER CS DTT CS

Operating current (kA) 45.0 23.0

Peak magnetic field (T) 13 13.2

Cumulative operating load 585 kN/m 288 kN/m

Conductor outer dimensions 49.0 mm x 49.0 mm 31.6 mm x 19.8 mm

Jacket Thickness 8.2 mm

(minimum value) 2.9 mm

Cable area (mm2) 771

(excluding central channel) 353

Steel section per turn (jacket) 1566 mm2 242.4 mm2

*900 MPa yield stress

ITER vs DTT CS



The 6 NbTi PF coils are in not-challenging conditions: separately fed, double-pancakes,

placed into clamps fixed to the TF coil structure,3mm thick epoxy-resin layer for ground

insulation around windings.

Vertical force limits (12.5 MN for CS coils,

19 MN for PF coils) scaled from DEMO.

PF1 PF2 PF3 PF4 PF5 PF6

Bmax (T) 3.70 3.00 2.35 3.36 3.85 4.02

Imax (MAt) 3.277 2.446 2.371 3.454 3.337 6.046

Name Isat

(kA)

Vsat (V) turns

CS3U 23 800 270

CS2U 23 800 420

CS1U 23 800 420

CS1L 23 800 420

CS2L 23 800 420

CS3L 23 800 270

PF1 25.2 800 130

PF2 22.6 800 108

PF3 21.2 1000 112

PF4 24.7 1000 140

PF5 23 800 152

PF6 23.3 800 260

C1 60 50 1

C2 60 50 1

C3 60 50 1

C4 60 50 1

C5 25 200 4

C6 25 200 4

C7 60 50 1

C8 60 50 1

Field and current limits

Current and voltage limits (4 quadrants)



-100 -50 0 50 100 150 2000

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

Time (s)

Pow

er fa

ctor

Imax, Imin, Break-down and PS voltages, SNU res.

-100 -50 0 50 100 150 200-40

-30

-20

-10

0

10

20

30

40

50

60

Time (s)

Pow

er

(MW

, M

var,

MV

A)

P

Q

S

Total active, reactive and apparent power for poloidal coils

CS3U CS2U CS1U CS1L CS2L CS3L PF1 PF2 PF3 PF4 PF5 PF6 IC5 IC60

100

200

300

400

500

600

700

800

900

Poloidal coil name

PS

curr

ent

and v

oltage,

SN

U R

and v

oltage

Ipos

(kA)

Ineg

(kA)

VPS

(V)

RSNU

(m)

VSNU

/10

-100 -50 0 50 100 150 20050

100

150

200

250

300

350

400

450

500

550

Time (s)

Pow

er

(MW

, M

var,

MV

A)

P

Q

S

Power supplies and electrical distribution system

SN scenario

-100 s 200 s 0 -100 s 200 s 0

0

900

0

1


-40

60

550

50 Time (s) Time (s)


Poloidal Toroidal Additional Auxiliary DTT Total +20%

P (MW) 20 (positive) 2.2 130 90 270

Q (Mvar) 60 2.7 150 80 350

S (MVA) 60 2.7 200 120 440

Power factor - - 0.65 0.75 0.67 (average)

Duty cycle 100s/3600s CW 100s/3600s CW -

Most power supplies have output DC current ±25 kA and output DC voltage ±800V

(except PF3, PF4, IC5 and IC6 PSs that have an output DC voltage ±1 kV). These AC/DC

converters are four quadrants, thyristor based 12 pulses with current circulating and

sequential control to reduce the reactive power, except IC5 and IC6 PSs that are IGCT

based to be fast enough to control the vertical position of plasma

The ENEA Research Centre of Frascati is a candidate site for DTT. It has been foreseen an

high voltage connection at 400 kV by an intermediate electric substation 400kV/150kV

(whose location is not still defined) and two underground electric cables up to the electric

substation 150kV/36kV of ENEA Research Centre of Frascati. The electric characteristics of

the power grid are not still available because it is ongoing a contract with TERNA for the

definition of connection characteristics and costs.


Power supplies and electrical distribution system


1

2

4

3

5

11

22

44

33

5



• Plasma disruptions

• TF discharges

L/R time constants of DTT VV

VS: 2070 s-1, ms 0.40.8

The maximum Von Mises Stress is lower than INCONEL

625 admissible stress limit (Sm =265Mpa) in VV

E

42 ms

Br

22 ms

Bv

16 ms

B

22 ms



Neutronics calculations show that without any additional shield

(considering only VV, FW and front casing) the TF coil nuclear

heating density on the first inboard turn is 3.77 mW/cm3. With

proper shielding design (5 cm inboard), the total nuclear loads

on the TF coil would be 5-10 kW. By increasing the shielding

thickness and improving VV design and/or by slighting reducing

the operational density, this figure could be reduced to 2-3 kW. Total neutron flux (n cm-2 s-1)

@ inboard midplane 9.1x1011



Plasma facing components

The FW consists of a bundle of tubes armored with plasma-sprayed tungsten (W). The

plasma facing tungsten is about 5 mm thick (except for the equatorial and upper inboard

segments where the tungsten layer is about 10 mm thick), the bundle of stainless steel

tubes (coaxial pipes in charge of cooling operation) is 30 mm thick, and the backplate

supporting the tubes is 30 mm thick of SS316L(N)

Poloidal profile 3D view FW support structure

FW layers

RH mandatory for the

non-negligible neutron flux



The main objective of the DTT project is to test several divertor design and configurations, so the

concept of the machine could change from the standard single null (SN) plasmas to alternative

configurations like X Divertor (XD) Snow Flake Divertor (SFD). Furthermore the design of VV,

ports and RH devices should take into account application and testing of a Liquid Metal Divertor.

A possible divertor compatible with SN & SF

RH

Liquid Li limiter

tested in FTU



Additional heating

A mix of different heating systems will provide the required 45MW power:

≈15MW ECRH at 170 GHz; ≈15MW ICRH at 60-90 MHz; ≈15MW NBI at 300 keV.

During the initial plasma operations 15 MW of ICRH and 10 MW of ECRH will be available.

4 antennas

16 RF generator units

2 auxiliary PS & 1 HVPS (with 8 units)

TLs + tuning and matching (16 units)

Cooling, control, data acquisition, test bed facility

15 MW ICRH system

gyrotrons

MHVPS

TL

Rem. part (cryom., BHVPS, PS filam., collector coils,

launcher, CODAS)

10 MW ECRH system

NBI



Data acquisition, diagnostics and control

Diagnostics

Parameters to be measured: Te Plasma Core, Ne Core, Ti, Ion Flow

Plasma Core, Plasma Current, Magnetic Field, Plasma position and

shape, Plasma Energy, q profile, MHD, Radiation, Zeff, Impurities Core,

Impurities SOL/Divertor, ni, Ti, flow, Divertor Te, ne, Divertor

Detachment, Neutrals (pressure), Wall Hot Spots, Escaping Fast ion,

Wall temperature, q, Runaway electrons, Halo/Hiro Currents, Vessel

deformation/displacement, Redeposition layers

Real time control (main components)

Overview of interferometer-

polarimeter 6+5 viewing chords

Diagnostic Actuator

Plasma Current Rogowsky Coils Magnetic Flux

Axisymmetric equilibrium Magnetic sensors PF coils

Electron Density Interferometer Gas valves/ Cryopumps

MHD /NTM Pick-up coils/ECE/SXR ECE/Control coils

ELM control Da, Stored energy

Control Coils, Plasma Shape

Control, Vertical kicks, Pellets ,

RMP’s

Power exhaust IR Cameras/thermocouples/ CCD

cameras/spectroscopy

Divertor and main plasma Gas

valves /impurity gas valves



Other systems and possible future upgrading

Other systems*:

• cooling systems (cryogenics & conventional)

• pumping & fuelling systems

• auxiliary systems

Possible future upgrading*:

• DTT upgrade with a liquid metal divertor

• First wall (FW) and alternative divertors

• Double Null (DN)


* details in the proposal


• This DTT facility would be fully dedicated to the power exhaust problem.

• The different schedules of EUROfusion and Italian Government made the

management of this WP complicated and sometimes awkward: the positive

attitude of various EUROfusion bodies was very helpful.

• The EFSI decision on Italian proposal for DTT funding is expected by Dec.

2015 (delays due to the Greek debt issues and the immigration problem):

technical and administrative staff of the Italian Government started contacts

with EIB to set up a financial plan in case of approval.

• The DTT assessment is expected in early 2016, based on the guidelines

provided by the EU PEX AHG Phase II and the activities carried out by

WPDTT1.

• Revision of the DTT project proposal is planned to take into account the

indications of the DTT assessment and the latest experimental results.

Conclusions

dtt: a divertor tokamak test facility for the study …...von mises stress ok for a 2.9 mm 316 ln...

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