Institute of plasma physics › Structure of IPP › Fusion Plasma Division › COMPASS Upgrade Project › Tokamak overview

Tokamak overview

Main design parameters

Toroidal magnetic field	Bt = 5 T
Plasma current	I_p = 2 MA
Major radius	R = 0.894 m
Minor radius	a = 0.27 m
Aspect ratio	A = 3.3
Triangularity	δ = 0.3-0.6
Elongation	κ = 1.8
Flat top duration	1-3 s (up to 11 s in low-parameter lower single null plasmas)

Available plasma shapes

Lower single null, negative triangularity with limited plasma parameters (Phase 1-2)
Double null (Phase 2-3)
Snowflake, negative triangularity (Phase 3-4)
Replaceable divertor design allows various divertors to be mounted

Heating power

NBI (Phase 1): 3-4 MW plus two diagnostic beams at 0.5 MW
ECRH (Phase 1): 1 MW
P_heatingB/R ratio (Phase 1): ~25
Phase 2: P_NBI up to 8 MW, P_ECRH up to 10 MW
PB/R ~ 100

3D view of the COMPASS-U tokamak

Visualisation of the COMPASS-U tokamak inside the experimental hall

Plasma scenarios: lower single null, double null, negative triangularity, snowflake divertor

Plasma facing components

Fully metallic (Inconel, W-coated Inconel and W)
First wall and vacuum vessel operated at high temperatures (nominally 300 °C, up to 500 °C), heated by gaseous He or CO₂

Inner wall limiters

8 guard limiters made of tungsten tiles
Support the plasma column during the plasma start-up and termination (~0.4 s)
Protect the remaining inner wall tiles (Inconel 718)

Divertor

32 tungsten cassettes, bolted to a toroidally continuous outer ring held by 16 flexible supports
PFC tiles bolted from the cassette back side
Divertor baffled may be made of with W-coated Inconel
Heat loads up to ~100 MW/m² -> heat dissipation is required (detachment, strike point sweeping) to lower the heat fluxes
Designed to routinely endure 20 MW/m² for 2-3 s

Tokamak plasma facing components (tungsten - magenta, tungsten-coated Inconel - cyan, Inconel - grey)

Tokamak plasma facing components

(tungsten - magenta, tungsten-coated Inconel - cyan, Inconel - grey)

View inside the COMPASS-U vacuum vessel

Vacuum vessel and ports

Made of Inconel 625 stainless steel, up to 35 mm thick
Total weight: ~9 t (including PSP)
Supported from the bottom by 8 flexible Inconel 625 beams

Designs of various tokamak ports

Vacuum vessel

Various designs of tokamak ports Vacuum vessel

Vacuum vessel heating and cooling

First wall and vacuum vessel will operate nominally at 300 °C, up to 500 °C
Vacuum vessel is thermally insulated by 20 mm of multilayer insulation (MLI)
Piping: Inconel 625, welded on the inner side of the vacuum vessel
Heating/cooling medium: He or CO₂ (gaseous)
Heating power ~40 kW heats the vacuum vessel and internal components to working temperature in ~24 hours
Cooling power ~33 kW removes the energy deposited by the plasma during the discharge (max. 40 MJ) in 20 min

Heating/cooling pipes

Temperature distribution during tokamak operation

Heating/cooling pipes Temperature distribution during tokamak operation

Toroidal field coils (TF coils)

16 coils made of OFHC (Oxygen Free High Conductivity) copper, CuAg0.1 or CuZr0.1, 7 turns each
Central solenoid and poloidal field coils inside the toroidal field coils -> the TF coils are dismountable with a sliding joint at the crown and a bolted joint at the outer midplane
Cooled to 80 K by gaseous He
Can carry 200 kA of current, producing B_t = 5 T at the magnetic axis (R = 0.894 m)
Toroidal field ripple at separatrix δ < 0.5 %
Assuming CuAg0.1 as the TF coil material and sliding joint resistance 0.2 μΩ, during a 3s flat top at B_t = 5 T, the TF coils will heat up by several tens of K -> must be cooled after the discharge

Toroidal field coil system

Poloidal field coils and central solenoid

4+4 poloidal field coils and 8 identical central solenoid segments
Hollow conductors made of OFHC (Oxygen Free High Conductivity) copper, CuAg0.1 (C10700), half or full hard
Cooled to 80 K by a gaseous coolant (He or H₂)
Located inside the toroidal field coils
1 power supply per pair of CS coils -> 14 power supplies in total
Conductor insulation: 1 mm S2 glass tape + kapton
Inter-layer insulation: 0.6 mm S2 glass tape
Ground insulation: 3 mm S2 glass tape
Vacuum pressure impregnation using epoxy resin

Poloidal field coils and central solenoid placement

Central solenoid structure and mechanical loads

Poloidal field coils and central solenoid placement Central solenoid structure and mechanical loads

Cryostat and support structure

Cryostat

Stainless steel AISI 304
Volume ~100 m³, weight ~50 t
The tokamak is placed on top of the cryostat base
8 massive steel supports attached to the 0.8 m thick steel-reinforced concrete slab of the experimental hall
Multilayer thermal insulation (MLI) on the inner surface

Support structure

Material SS 304(L)N or 316(L)N
Height ~4.4 m, diameter ~4.4 m, total weight ~190 t
16 C-frames + flexible supports
Cooled to 80 K with gaseous He led through stainless steel pipes welded to machined grooves
Cool-down in ~1 week time, accompanied by vertical contraction of ~14 mm
Can be vertically disassembled

Cryostat

Tokamak support structure

Cryostat Tokamak support structure

3D view of the support structure

Cryogenics

After a discharge, tokamak systems (plasma facing components, TF coils, PF coils, central solenoid...) must be cooled down; this takes <60 min
TF coils ~250 MJ, PF coils ~50 MJ -> required cooling power ~100 kW
Cooling performed by multiple closed gaseous He loops
CS - high pressure (p_base = 60 bar, Δp = 4 bar, ṁ = 80 g/s)
PF - medium pressure (p_base = 20 bar, Δp = 1 bar, ṁ = 160 g/s)
TF - low pressure (p_base = 20 bar, Δp = 0.1 bar, ṁ = 800 g/s)
Main cold source – liquid nitrogen heat exchanger
Cycle cooler (Brayton, J-T, G-M, ...) for subcooling under 80 K

Simulations of tokamak cooling systems

Simulations of tokamak systems cooling

Neutral beam injectors (NBI)

3-4 MW of NBI power at neutral energy 80 keV, organized in 2 x 2 MW units
Ions supplied by two ion RF sources above each other, inclined by ~7° from horizontal plane
The first unit is about to be delivered by BINP Novosibirsk (will be installed on COMPASS)
Injectors aim between the magnetic axis and the HFS wall - tangency radius R < 0.65 m
The two existing COMPASS NBIs (each 0.3 MW at 40 keV) will be upgraded and used for diagnostic purposes

Setup of neutral beam injectors

Electron cyclotron resonance heating (ECRH)

ECRH system enlarges the COMPASS-U operating space and plasma performance
Deposition on-axis is achieved for B_t = 1-2.5 and 5 T, toroidal steering needed for B_t = 3-4 T
Simulations in TORBEAM and ASTRA are ongoing

ECRH components

Gyrotrons: dual frequency 105-140 GHz, 1 MW, pulse length 3-5 s
Waveguides: 63.5 mm in diameter, total length < ~30 m
Launchers: large equatorial port, steering mirrors

ECRH operation modes

Power supply system

has to provide approx. 270 MW of power and 500 MJ of energy
uses the existing flywheel generators (50 MVA, 50 MJ each) and two new flywheel generators (106 MVA, 195 MJ each)

PF coils power supply

Power and energy: 85 MW - 90 MJ
Based on IGBT H-bridges

TF coils power supply

Power and energy 140 MW - 340 MJ
Based on thyristor converters

Auxiliary heating + reserve: 38 MW - 60 MJ

Status

FDR completed in February 2019
Contract signed in February 2020
December 2021 - first components manufactured (HV distribution station, flywheel starting motor, transformators, etc.)

Tokamak power supply system schematic

Tokamak overview

Institute of Plasma Physics of the CAS