guidelines for the design of a tokamak device

Guidelines for the design of a tokamak

device

R. Albanese, CREATE – Univ. Napoli Federico II

WPDTT2 Project Leader

22nd Int. Conf. on Plasma Surface Interactions in Controlled

Fusion Devices (22nd PSI), May-June 2016, Rome, Italy

http://www.psi2016.enea.it/index.php/tutorials

Tutorials – PSI 2016 | Guidelines for the design of a tokamak device | R. Albanese | Rome | 29 May 2016 | PAGE 2

Outline

• Introduction

• Definition of role and objectives

• Physics basis

• Physical and technological requirements

• Choice of main machine parameters

• Specifications

• Main subsystems

• Cost, site, licensing, organization, risks, schedule


Introduction

Tokamaks are among the most complex machines ever conceived by the

mankind:

• Coexistence of temperatures close to highest and lowest values in the

universe

• Nuclear environment, high magnetic fields, vacuum requirements, large

heat fluxes

• All fields of science and engineering involved: large teams needed


Definition of role and objectives

Just a few examples:

JET (in operation since 1983, with various upgrades later): designed to study

plasma behaviour in conditions and dimensions approaching those required in a

fusion reactor (including D-T operation)

TORE SUPRA (in operation since 1988, now being upgraded to WEST):

devoted to the study of physics and technologies for long pulse plasma scenarios

ITER (under construction): reactors-scale international experiment designed to

deliver ten times more power than it consumes (burning plasma with Q 10)

DTT (proposal): a facility addressed to develop and test integrated, controllable

power exhaust solutions for DEMO including plasma, PFCs, control

diagnostics/actuators

DEMO (predesign activities): expected to be the first fusion plant to provide

electricity to the grid

...


Physics basis

The physics guidelines are used as a basis for the definition of a tokamak

concept as well as its performance expectations:

• possibly supported by current knowledge in tokamak physics (or limited

extrapolation from ongoing experimental data)

• necessarily consistent with role and objectives of the tokamak

Example in DTT:

• Power exhaust

• Alternative configurations

• Liquid metal targets

• ...

Example in ITER:

Nuclear Fusion, Volume 39, Number 12

Plasma confinement and transport - MHD stability, operational limits and

disruptions- Power and particle control - Physics of energetic ions - Plasma auxiliary

heating and current drive- Measurement of plasma parameters- Plasma operation

and control - Opportunities for reactor scale experimental physics

...

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

http://iopscience.iop.org/journal/0029-5515



http://iopscience.iop.org/0029-5515/39



http://iopscience.iop.org/0029-5515/39/12







Physical and technological requirements

Strictly related to the objectives

Example in DTT (accent on SOL & power exhaust):

Physical requirements

• Preservation of 4 DEMO relevant parameters: Te , *=Ld/λei, Δd/λ0 , β

• Relaxation on normalized Larmor radius: (ρi/Δd)*R value similar to DEMO

• Integrated scenarios: solutions compatible with plasma performance of DEMO

Technological requirements

• 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 compatible with technological constraints of DEMO

• Budget constraint: within 500 M€

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


Choice of main machine parameters (1/3)

How to select the main parameters so as to guarantee that the requirements

are fulfilled?

1. Proper definition of the constraints is needed

2. A suitable figure of merit should be chosen

3. Physics and engineering parameters must be consistent with one another

The first two points (constraints and figure of merit) are within the terms of reference

of the project and should be specified by the design team.

The third task is not simple, as the large number of parameters involved are highly

dependent on one another. Therefore this issue is usually addressed with the

support of computer programs known as systems codes.

These programs are based on 0-D models of all parts of a tokamak, from the basic

plasma physics to the buildings. In their simplest form they can be excel spread

sheets, however there are more comprehensive codes, e.g. PROCESS used by

CCFE (http://dx.doi.org/10.1016/j.fusengdes.2014.09.018), a system code used by

KAERI (DOI: 10.1016/j.fusengdes.2008.07.037), GA-system code (DOI:

10.1016/S0920-3796(00)00300-8), …



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In case of multiple objectives, Pareto optimization

can be used, e.g. in DTT reduce (relevance of SOL

physics) and increase R (flexibility and relevance)

Main DTT proposal assumptions: • Preservation of 4 DEMO relevant parameters: Te ,

*=Ld/λei, Δd/λ0 , β

• Relaxation on normalized Larmor radius: (ρi/Δd) R value similar to DEMO

• Preservation of EU DEMO parameters: Psep/R15 MW/m, R/a3, 1.7, 0.3

• Cost scaling with magnetic energy: Cost B 2R 3

Psep/R vs R Cost (excluding Paux) vs R • 0.75 is the largest

value compatible with

the constraint

Psep/R15 MW/m

• R 2.2 m is the largest

value compatible with

a cost of 500 M€ (150

of which for Paux)



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

MAIN DTT PARAMETERS FOR THE

REFERENCE SINGLE NULL SCENARIO

Comparison with ITER and DEMO


Specifications

• Once the main machine parameters have been fixed, detailed technical

specifications have to be provided for each subsystem

• The concept ideally starts from the plasma reference scenario

• Then a vacuum vessel able to contain the plasma and the in-vessel components

(first wall, blanket, divertor, diagnostics, etc.) has to be designed; thickness and

material should be chosen so as to take into account the electromagnetic

interactions and the mechanical stress

• Then a set of TF coils is designed taking account of geometrical (vessel +

shield), mechanical (forces and torques), and electromagnetic constraints

(current density limits, ripple, etc.)

• Then the CS and the PF coil system is designed so as to guarantee the flux

consumption during flat top and the equilibria for full bore plasmas

• Then the design continues with a cryostat that acts as an external containment

structure

• Then the specification for other fundamental subsystems (auxiliary heating,

power supplies, remote maintenance system, pumps, cooling, etc.) can be given

• Once the various subsystem are designed, then one or more iterations are

needed to take account of the inevitable interactions between the various

subsystems


• The definition of the subsystems may vary a lot from tokamak to tokamak

• To fix the ideas, in the following I refer to the DTT proposal

• This DTT has most of the features of a next generation tokamaks but:

• no tritium no fusion reactions, no blanket, no double-walled VV (even

if the expected neutron flux is significant, namely 91011 n cm-2 s-1)

• no significant current drive contribution

Main subsystems


July 2015

and

R. Albanese, A. Pizzuto et al. "The DTT proposal: introduction and executive

summary“, submitted to Fusion Engineering and Design, Special Issue for DTT





Reference plasma scenario

• 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

Main subsystems: reference plasma 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

Main subsystems: alternative configurations


1

2

4

3

5

11

22

44

33

5

Main subsystems: vacuum vessel (1/2)


• Plasma disruptions

• TF discharges

L/R time constants of DTT VV

VS: 2070 s-1, ms 0.40.8

the VV also acts an EM shield

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

Main subsystems: vacuum vessel (2/2)


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

Main subsystems: plasma facing components


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

Main subsystems: divertor

A LM box divertor


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

Main subsystems: shield and cryostat


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

Main subsystems: magnets

DTT


Each of the 18 D-shaped TF coils has 78 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 (6 .0T @ 2.15 m)

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

Main subsystems: toroidal field coils

Bending moment free "D"-shaped toroidal field magnets are placed around the vacuum vessel

produce a magnetic field whose primary function is to confine the plasma particles and guarantee a

sufficiently high safety factor.


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

Main subsystems: central solenoid

Bz (T)

13.2

0 0.497 0.840 R (m)

Jmax 29 MA/m2 ; Rmin = Rmax Bmax /(0Jmax)

= Bmax(Rmin2 +RminRmax+Rmax

2)/3 17.5 Vs

Rmax


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)

Main subsystems: poloidal field coils

in-vessel coils C1-C8 used for

plasma control or local field

modifications in the divertor region


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

Main subsystems: additional heating


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.

Main subsystems: power supplies

Power supplies and electrical distribution system


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

Main subsystems: diagnostics and control


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)

Main subsystems: other subsystems and upgrades


DTT investment costs

Cost, site, licensing, organization, risks, schedule (1/7)

Main Components Cost (M€)

Load Assembly 224.10

Auxiliary Heating Systems 96.00

Principal diagnostic systems 8.00

Controls and Data Acquisition System 4.50

Cooling System 27.40

Power Supply 78.00

Remote Handling 14.00

New buildings 11.00

Assembly 11.00

Contingency 25.00

Total 499.00



The candidate site for DTT is Frascati. The ENEA FRC has the possibility

to realize the DTT facility, given its capability to meet the various technical

requirements. The presence of FTU Tokamak facility would make much

easier the authorization and licensing procedures of the new machine.

aerial view on of the present FTU

buildings, with the necessary

upgrades for DTT highlighted in yellow

design of the new hall and

the present FTU hall



DTT licensing scheme



DTT proposal



DTT organization



WPDTT2 risk register



schedule

Load Assembly

Magnets

Vacuum Vessel

First Wall

Divertor

Criostato

Lay out

Additional Power

ICRH:15 MW

ECRH: 10 MW

Controls and data Acquisition

Cooling

Helium cooling

Water cooling

Electric Power Supply

Sub-station

power Supply

Remote Maintenance

in vessel

ex vessel

Buildings: FTU Hall modification and Service …

Assembly

Licensing

Commissioning

guidelines for the design of a tokamak device

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