Guidelines for the design of a tokamak

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

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

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: T

, 

=L

/λ

, Δ

/λ

, β

• Relaxation on normalized Larmor radius: (ρ

/Δ

)*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), …

Choice of main machine parameters (2/3)

[www.noesissolutions.com]

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: T

, 

=L

/λ

, Δ

/λ

, β

• Relaxation on normalized Larmor radius: (ρ

/Δ

)  R

value similar to DEMO

• Preservation of EU DEMO parameters: Psep/R15 MW/m, R/a3,  1.7, 0.3

• Cost scaling with magnetic energy: Cost  B

R

Psep/R vs R Cost (excluding Paux) vs R

•   0.75 is the largest value compatible with the constraint

Psep/R15 MW/m

• R  2.2 m is the largest

value compatible with

a cost of 500 M€ (150

of which for Paux)

Choice of main machine parameters (3/3)

R (m) 2.15 b_N 1.5

a (m) 0.7 t_Res (sec) 8 I_P (MA) 6 V_Loop (V) 0.17

B_T (T) 6 Z_eff 1.7

V (m³) 33.0 P_Rad (MW) 13 P_ADD (MW) 45 P_Sep (MW) 32 H₉₈ 1 T_Ped (KeV) 3.1

<n_e> (10²⁰ m^-3) 1.7 n_Ped (10²⁰ m^-3) 1.4

n_e/n_eG 0.45 b_p 0.5

<T_e> (KeV) 6.2 P_Div (MW/m²)

(No Rad) ~ 55 t (sec) 0.47 P_Sep/R

(MW/m) 15 n_e(0) (10²⁰ m^-3) 2.2 P_TotB/R (MW

T/m) 125 T_e(0) (KeV) 10.2 λ_q (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  910

n cm

s

)

• no significant current drive contribution

Main subsystems

http://fsn-fusphy.frascati.enea.it/DTT 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 1 1

2 2

4 4 3 3

5

Main subsystems: vacuum vessel (1/2)

• Plasma disruptions

• TF discharges

L/R time constants of DTT VV VS:   2070 s^-1, m_s  0.40.8

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

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

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/cm^3. 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: B

: 12.0 T, B

: 6.0 T, 65 MAt;

6 CS coils: B

: 12.5 T, 

|N

I

| =51 MAt; available poloidal flux: 17.6 Vs;

6 PF coils: B

: 4.0 T, 

|N

I

| =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, T_marg= 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, T

= 1.5 K

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

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 Nb

Sn 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 J_max  29 MA/m² ; R_min = R_max B_max /(₀J_max)

 = B_max(R_min² +R_minR_max+R_max²)/3  17.5 Vs

R_max

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

Cost, site, licensing, organization, risks, schedule (2/7)

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

Cost, site, licensing, organization, risks, schedule (3/7)

DTT licensing scheme

Cost, site, licensing, organization, risks, schedule (4/7)

DTT proposal

Cost, site, licensing, organization, risks, schedule (5/7)

DTT organization

Cost, site, licensing, organization, risks, schedule (6/7)

WPDTT2 risk register

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