Plasma Facing Components (PFC)

S. W. Lisgo ¹

with the PFC part based heavily on presentations by

F. Escourbiac ¹ , T. Hirai ¹

and, in general, helped by many conversations with

G. De Temmerman

and additional material provided by

S. Carpentier-Chouchana ¹ , D. Coster ² , A. Dvornova ¹ , J. Gunn ³ , J. Harrison ⁵ , M. Kocan ¹ , S. Panayotis ¹ , R.A. Pitts ¹ , J. Terry ⁶ ,

and the ITER Divertor Design Team ^1,4

Plasma-Wall Interaction (PWI) and Plasma Facing Components (PFC) Plasma-Wall Interaction (PWI) and Plasma Facing Components (PFC)

The views and opinions expressed herein most certainly do not reflect those of the ITER Organization, at all

1

ITER Central Team, France

2

IPP Garching, Germany

3

IRFM/CEA, France

4

ITER Domestic Agencies, Earth

5

CCFE, UK

6

PSFC, MIT, USA

 Boundary plasma transport

 Plasma-material interaction

 Plasma facing components

The (ITER) tokamak (the “model tokamak” for this presentation)

The (ITER) tokamak

cryostat

The (ITER) tokamak

vacuum vessel port extensions:

diagnostics, heating, fueling, pumps, in-vessel component feeds, remote access

The (ITER) tokamak

The (ITER) tokamak

toroidal field coils

The (ITER) tokamak

vacuum vessel

blanket

divertor

central solenoid

The (ITER) tokamak

Magnetic field structure: toroidal field

Magnetic field structure: poloidal field

B _poloidal  0.1 B _toroidal

plasma current

Magnetic field structure: magnetic field line trajectory (q = 3)

B _poloidal  0.1 B _toroidal

Magnetic field structure: magnetic field line trajectory (q = 3)

MACHINE GEOMETRY MAGNETIC GEOMETRY

B _poloidal  0.1 B _toroidal

Magnetic field structure: x-point geometry directs plasma toward divertor

Magnetic field structure: “open” field lines intersect solid surfaces

Magnetic field structure: strong radial magnetic shear near the x-point

Plasma transport from the core to the boundary: turbulence

gravity

FLUID DYNAMICS: RALEIGH-TAYLOR INSTABILITY ( link)

Plasma transport from the core to the boundary: turbulence

C-MOD

gravity

EB

HIGH SPEED D _a GAS PUFF IMAGING PLASMA DYNAMICS:

“INTERCHANGE” INSTABILITY

[J. Terry]

DEUTERIUM

GAS PUFF

Plasma transport from the core to the boundary: turbulence

C-MOD

1 cm

gravity

EB

HIGH SPEED D _a GAS PUFF IMAGING PLASMA DYNAMICS:

“INTERCHANGE” INSTABILITY

[J. Terry]

fluctuations can also confine

plasma DEUTERIUM

GAS PUFF

Transport from the core to the boundary: ballooning transport

MAX = 4.310

ph m

s

INVERTED He ⁺¹ @ 468 nm IMAGE

5 mm mesh, 500500 image, shot 11849

MAST LIGHT FROM SINGLY CHARGED HELIUM

[S. Lisgo]

Main chamber recycling (MCR): plasma contact with main chamber surfaces

[A. Loarte 2007]

FAR-SOL FAR-SOL NEAR-SOL

NEAR-SOL

separat rix

RADIAL PLASMA DENSITY PROFILE

Main chamber recycling (MCR): plasma contact with main chamber surfaces

[A. Loarte 2007]

FAR-SOL FAR-SOL NEAR-SOL

NEAR-SOL

separat rix

RADIAL PLASMA DENSITY PROFILE

ESTIMATED RADIAL PLASMA VELOCITY

Turbulence suppression: H-mode MAST

15622, unfiltered

TURBULENT BOUNDARY ELMing H-MODE

Turbulence suppression: H-mode MAST

15622, unfiltered

TURBULENT BOUNDARY ELMing H-MODE

[Eich 2013]

Power exhaust: peak steady-state heat-flux to the divertor targets

l _q,|| for ITER currently predicted to be 1 mm, significantly smaller than the design value of 5 mm  reduced operational

space? [Kukushkin 2015]

S scaling? [Scarabosio 2015]

l _int

l _int  radial heat-flux decay length at target

l _q,||

l _q,||  radial decay length of heat-flux into divertor

S

S  heat-flux spreading due to divertor transport

l _int  l _q,|| + 1.64 S

Power exhaust: cross-field turbulent transport in the divertor (related to S)

LOCALIZED DIVERTOR TURBULENCE

X-POINT RADIAL MAGNETIC SHEAR

Power exhaust: ELM transient heat loads to PFCs

 Electron beam thermal cycling of ITER mono-block components indicate failure at

~0.5 MJ m ^-2 for both CFC and W [J. Linke et al JNM (2007)]

 Natural ELMs in ITER predicted to be 10-20 MJ m ^-2  ELM control required

Power exhaust: ELM control with Resonant Magnetic Perturbation (RMP) EMC3 SIMULATION OF DIII-D

MAST

He ⁺ , NO RMP

WITH RMP RMP

COILS

Tungsten influx during an ELM and core-boundary impurity coupling SOLPS-DIVIMP ELM SIMULATION

W DENSITY, SUM OVER ALL STATES

TARGET ELECTRON TEMPERATURE

[D. Coster 2015, S. Lisgo]

Impurities can be a very good thing: control actuator for dissipation

INFRARED, VIEW OF DIVERTOR FROM ABOVE [A. Kallenbach]

ASDEX UPDGADE

 Boundary plasma transport

 Plasma-material interaction

 Plasma facing components

Plasma-surface interaction in the divertor

ION FROM “UPSTREAM” ACCELERATING TOWARD THE TARGET

Angle between the magnetic field, B, and the surface is typically a few degrees

At 20 eV and 5 Tesla, the gryo-radius for a deuterium ion is

0.1 mm

The sheath

ELECTROSTATIC DEBYE SHEATH NEAR THE SURFACE

Electrostatic potential drop (acceleration) experienced by an ion passing through the sheath:

V _sh  2 T _i + 3 Z _ion T _e At T _e = 20 eV and n _e = 10 ¹⁹ m ^-3 , the sheath thickness is

0.01 mm

Plasma recycling

PLASMA RECYCLING, FLUX AMPLIFICATION

Plasma recycling

PLASMA RECYCLING, FLUX AMPLIFICATION

ion flux  n e,upstream

when ionisation near the surface (local to the divertor)

becomes important

2

Charge exchange

CHARGE EXCHANGE (CX)

For hydrogen, the CX rate always larger than the ionisation rate

Important process for the transport of

energy across the

magnetic field, i.e. CX

Physical sputtering

PHYSCIAL SPUTTERING

4 𝑀 ₁ 𝑀 ₂ (𝑀 ₁ + 𝑀 ₂ ) ² Maximum energy

fraction transferred in a head-on collision:

= 100% for M ₁ = M ₂

and 4% for D  W

Physical sputtering

SPUTTERING YIELD PER INCIDENT DEUTERIUM ION [Stangeby pg 119]

D threshold T

C W



W self-sputtering yield > 1

(normal incidence)

implantation

Prompt re-deposition

GROSS AND NET EROSION, PROMPT RE-DEPOSITION

Net erosion can be much less than gross erosion

For example, 4 orders of magnitude lower for W sputtering during ELMs 

prompt re-deposition

[Chankin 2014]

Temperature gradients in the boundary

PARALLEL TEMPERATURE GRADIENTS, HIGH RECYCLING DIVERTOR

Energy dissipation in the boundary due to recycling and

impurity radiation can lead to strong temperature

gradients between

the core and divertor

Plasma “detachment”

“GAS TARGET” DIVERTOR, STRONG DISSPATION

For T _e < 5 eV, plasma- surface interaction can be strongly reduced 

“detachment”

“Deep” detachment

can affect core plasma

performance

BALMER GAMMA EMISSION  T _e < 1.5 eV

[C. Boswell 2001]

An example of a strongly dissipating divertor

C-MOD (SCALE 10)

ITER

Many open questions related to plasma-material interaction MANY ACTIVE AREAS OF RESEARCH

• mixed materials

• weakly bound atoms

• chemical erosion

• surface roughness

• incident angle

• plasma chemistry

• heavy ions

• plasma transients

• magnetic pre-sheath

ITER issue: tritium retention by co-deposition

 A 400 s Q=10 pulse will require ~50 g of T fuel, but the maximum mobilisable in- vessel T inventory is limited to 640 g (+180 g in pumps, +180 g uncertainty)

− nuclear safety issue (expensive too); tritium burn-up is only ~0.3% in ITER

T:Be trapped fraction

~1:20 @ 200 C

[G. De Temmerman, R. P. Doerner, et al., NF (2009)]

T:Be trapped fraction

~1:20 @ 200 C

[G. De Temmerman, R. P. Doerner, et al., NF (2009)]

T:Be <1:100 @ 300+ C T:Be <1:100 @ 300+ C

 T:Be depends sensitively on deposition rate, incoming particle energy, and surface

temperature  complex problem  on-going efforts to predict the level of T-

ITER issue: fluence

 Long-pulse, large size, and high density operation combine to give a significant increase in the ion fluence to the wall

No.

Pulses

Time in Diverted Phase (hours)

Outer Divertor Ion Fluence

JET 13466 40.5 ~5x10 ²⁷ [1]

ITER (Q=10)

No.

Pulses

Time in Diverted Phase (hours)

Outer Divertor Ion Fluence

JET 13466 40.5 ~5x10 ²⁷ [1]

ITER (Q=10) 1 0.15 ~1.5x10 ²⁷ [2]

DIVERTOR ION FLUX COMPARISON WITH JET (2000-2008, campaigns C1-C19)

[1] M.F. Stamp, CCFE, private communication

[2] SOLPS4.3 code result: run #1514, A. S. Kukushkin

 9 years of JET operation ≡ 3 ITER pulses @ Q=10 (~1.5 hours of real time)

− 3 decades of JET, or ~half the time humanity has spent on controlled fusion, in a morning

 The high resulting material turnover affects tritium retention, material mixing, layer

growth, and dust production

ITER issue: surface morphology: helium and tungsten [G. De Temmerman]

 Want to commission ITER systems during the non-nuclear phase (H and He plasmas), but H-mode may only be accessible with He plasmas

Nano-bubbles ‘Fuzz’

~ 600 - 700 K ~ 900 – 1900 K

500nm [M. Miyamoto et al, JNM, 415 (2011)]

> 2000 K

Large voids

1mm

 Helium insoluble in tungsten but can self-trap

 Clustering of helium to form

bubbles  material ‘swell’

 Boundary plasma transport

 Plasma-material interaction

 Plasma facing components

How can PFCs survive a steady-state plasma environment?

 Most tokamaks have been “short pulse” (~seconds)  PFCs heat up during a discharge and cool down before the next one  “inertial cooling”

− main thermo-physical property is the specific heat capacity of the material, C

(J Kg

K

)

− armour tiles held in place mechanically with bolts (C / Mo reliable up to ~2500 C)

 Long pulse (superconducting)  active cooling  complex PFCs

− thermal conductivity, k (W m

K

), is the most important material property (a function of temperature and neutron fluence)

− bonded components that are welded to the structure (bolts OK below 1-2 MW m

)

HEAT

COOLANT (H ₂ O) MAIN STRUCTURAL MATERIAL (316L(N)-IG SS)

PLASMA FACING MATERIAL, PFM (Be, W)

HIGH l STRUCTURAL MATERIAL (CuCrZr)

INTERLAYER (Cu)

The power of active cooling

 Boiling water in a paper cup…

− (warning: enthalpy of vapourization important here – not directly applicable to ITER PFCs)

[L. Jedral and his high school physics class, Ontario Public School Board]

Time-averaged heat flux design limit at the target: 10 MW m ^-2

 What does 10 MW m ^-2 look like?

 If there is no energy dissipation in the ITER boundary plasma (very unlikely), then for the ITER baseline assumption of l _q,|| = 5 mm, there could be 40 MW m ^-2 at the

H EA T FL UX DE NSI TY (M W m -2 )

DISTANCE FROM CENTER OF ARC (mm)

[A. B. Murphy CSIRO]

HEAT FLUX TO THE WORKPIECE FOR AN ARC WELDER

[http://www.youtube.com/watch?v=Ng-urnWhBR8&hd=1]

40 MW m ^-2

10 MW m ^-2

Divertor cassettes

 54 cassettes, 8.7 tonnes each, “easily” removable

− 6 months to replace

Divertor cassettes

Divertor cassettes

Divertor cassettes

Vertical target and plasma facing unit (PFU) configuration

PLASMA FACING UNIT (PFU)

water flowing in a tube

CuCrZr

 A water cooled tube made from a high thermal

conductivity metal can remove heat very efficiently

Vertical target and plasma facing unit (PFU) configuration

PLASMA FACING UNIT (PFU)

 Increase tube thickness for greater strength and to maximize lifetime (tokamak plasma contact will erode the tube)

CuCrZr

Vertical target and plasma facing unit (PFU) configuration

PLASMA FACING UNIT (PFU)

W (6-8 mm thick) Cu

 A good structural metal may not be compatible with plasma interaction, so armour the tube with an

erosion resistant and high melting point material such as tungsten (note: tungsten is a poor structural material at tube temperatures appropriate for water cooling)

CuCrZr

 A layer of copper (soft) compensates for the thermal

expansion mismatch between the armour and tube

Vertical target and plasma facing unit (PFU) configuration

PLASMA FACING UNIT (PFU)

Cu

 Castellate to reduce thermomechanical stress in the armour and on the bond with the tube, as well as electromagnetic forces during plasma disruptions

W (6-8 mm thick)

CuCrZr

Vertical target and plasma facing unit (PFU) configuration

“VAPOUR MATTRESS”

• heat transfer dramatically reduced

• “critical heat flux” (CHF)  imminent failure of component

“NUCLEATE BOILING”

• onset at T _sat

• heat transfer coefficient goes up by a factor 3-8

[SOLPS, A. Kukushkin]

[RACLETTE, A. Sashala Naik]

Vertical target and plasma facing unit (PFU) configuration

T _wall “VAPOUR

MATTRESS”

• heat transfer dramatically reduced

• “critical heat flux” (CHF) 

Vertical target and plasma facing unit (PFU) configuration

T _wall “VAPOUR

MATTRESS”

• heat transfer dramatically reduced

• “critical heat flux” (CHF)  imminent failure of component

 The CHF is a function of the overall system design and can be

raised by increasing the coolant velocity + pressure, the PFC

complexity, and by sub-cooling  more money

Vertical target and plasma facing unit (PFU) configuration

 The CHF is a function of the overall system design and can be raised by increasing the coolant velocity + pressure, the PFC complexity, and by sub-cooling  more money

“SWIRL TUBE”

 The CHF is a function of the overall system design and can be raised by increasing the coolant velocity + pressure, the PFC complexity, and by sub-cooling  more money

Vertical target and plasma facing unit (PFU) configuration

ICHF VERSUS COOLANT FLOW VELOCITY

[A.R. Raffray, 1999]

 The CHF is a function of the overall system design and can be raised by increasing the coolant velocity + pressure, the PFC complexity, and by sub-cooling  more money

Vertical target and plasma facing unit (PFU) configuration

ICHF VERSUS COOLANT FLOW VELOCITY

[A.R. Raffray, 1999]

SCREW TUBE SCREW TUBE SCREW TUBE

Screw tube not chosen despite higher CHF  concerns that the grooves will seed crack formation in the tube

[A.R. Raffray, Fus. Eng. and Design 45 (1999) 377–407]

Vertical target and plasma facing unit (PFU) configuration

SMOOTH TUBE

SWIRL TUBE WITH TWISTED TAPE

HIGHER CHF

CuCrZr Cu

W

The dome

W FLAT TILE

ARMOUR

HYPER-

VAPOTRON

Hypervapotron

NUCLEATE BOILING IMAGES (low T)

[P. Chen, PhD thesis, 2007]

 Good CHF (to ~20 MW m ^-2 so far) with little loss in pressure  turbulence mixing of coolant across the square channels

− limitation perhaps from structural integrity at present (thermal gradients), rather than CHF

[J. Milnes, PhD thesis, 2011]

EDDY SIMULATIONS

[O. Aybay, Fluid Mechanics and Multidisciplinary Design, 2010]

FLOW

Exposed “leading edges” can receive the full parallel heat-flux

 Magnetic field lines can penetrate gaps between

monoblocks, allowing the plasma to heat the “leading edges”

of the PFCs  power is concentrated onto a smaller area, resulting in strong local heating

a

2D 2L

Exposed “leading edges” can receive the full parallel heat-flux

 Magnetic field lines can penetrate gaps between

monoblocks, allowing the plasma to heat the “leading edges”

of the PFCs  power is concentrated onto a smaller area, resulting in strong local heating

a

2D 2L

Exposed “leading edges” can receive the full parallel heat-flux

 Magnetic field lines can penetrate gaps between

monoblocks, allowing the plasma to heat the “leading edges”

of the PFCs  power is concentrated onto a smaller area, resulting in strong local heating

a

2D 2L

 The shallow angle between the magnetic field lines and the

surface, due to the large toroidal component of the field and

the vertical target geometry, means that small mono-block

Misalignments accounted for in the manufacturing specifications

 Geometrical tolerances are specified considering operational requirements at global and local scales

− all monoblocks aligned within 0.5 mm range

− steps of neighbouring blocks smaller than 0.3 mm

 To achieve the required tolerance during manufacture:

− identify poor tolerance operation: joining (welding) and assembly

− recover by higher tolerance operation (machining) and custom

machining of tolerance compensators

 Monoblock edges must be protected in an all-W divertor (C is more “forgiving”  “plasma polishing”)

− straight chamfer at 0.45º angle protects 0.3 mm with factor 2 margin 

~23% loss of wetted area for a = 3º incident angle

Protecting leading edges with monoblock shaping

 Extensive and detailed field line tracing studies have been performed that take into account variation of magnetic

a

2D 2L

h

x

b

ANSALDO W THICKNESS 6.0 mm PLANSEE W THICKNESS 7.5 mm

ITER Divertor Test Facility (IDTF) for high heat-flux testing of PFUs

 Comprehensive small-scale prototype testing

− “self-castellation” observed with some samples  does not affect qualification process

Advanced high-heat flux PFC technologies: helium

 He cooling  single phase coolant  10 MW m ^-2 system demonstrated [KIT]

− high temperature targets possible  increased efficiency of electricity production

− a lot of power may be required to move the helium…

KIT He COOLED “THIMBLE” (HEMJ)

Advanced high-heat flux PFC technologies: liquid lithium targets

 Can’t melt what’s already melted!  but early days…

− lots of potential issues, other metals being considered

NSTX LIQUID LITHIUM DIVERTOR (LLD)

Summary

 There are many active areas of PWI research, i.e. open questions

 The plasma boundary is complex and dynamic

 Plasma physics is an integral part of the PWI solution for high high-performance, high duty-cycle fusion devices

 A lot of recent focus has been provided by acute ITER needs, i.e. design guidance and nuclear regulation

 High-performance and robust active cooling technologies are required for the next- step in fusion energy development

 ITER PFC designs are based on conventional technologies and are well advanced