ITER In-Vessel Components: Blanket, Divertor - staff.polito.itstaff.polito.it/roberto.zanino/sub1/teach_files/current_topics/lect_merola.pdf · ITER In-Vessel Components •Divertor

Slide 1Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011

Mario Merola - ITER International Organization

Internal Components Division Head

ITER In-Vessel Components: Blanket, Divertor


Nuclear Fusion and the ITER Project

ITER In-Vessel Components

ITER Divertor

ITER Blanket

Design Criteria

Conclusions

Outline


Outline



ITER Divertor

ITER Blanket

Design Criteria

Conclusions


Fusion powers the sun and the stars

“…Prometheus stole the fire from the heaven”

• Essentially limitless fuel, available all

over the world

• No greenhouse gases

• Intrinsic safety

• No long-lived radioactive waste

• Large-scale energy production

On Earth,

fusion could provide:

http://it.wikipedia.org/wiki/File:Heinrich_fueger_1817_prometheus_brings_fire_to_mankind.jpg

Slide 5Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011J Jacquinot, Geneva FEC 2008

5JJ OCS Cannes 17 March 085

• The Sun has a radius of 0.7

Million kilometer

• A core temperature of 10

Million deg

• A power density of 0.01 W/m3

• The Tokamak chamber has a

radius of 2 meter

• A core temperature of 100 Million

deg

• A power density of 500,000 W/m3


Nuclear Fusion is an Energy Option…

• Take the lithium from the battery of a single laptop

computer, add half a bathtub of water, and it can give you

200,000 kilowatt hours of electricity

• That's enough to power one person in the EU for 30

years, including his share of industrial electricity.

+ = Energy


The Way to Fusion Power – The ITER Story

The idea for ITER originated from the Geneva Superpower Summit on

November 21,1985, when the Russian Premier Mikhail Gorbachev

and the US-President Ronald Reagan proposed that an international

Project be set up to develop fusion energy “as an essentially

inexhaustible source of energy for the benefit of mankind”.


ITER Agreement

The agreement was signed on the

21st November 2006

at the Elysée Palace in Paris

International Organization started.


JET

Tokamak’s

ITER

JET – Internals & Plasma

ITER will allow us to produce plasmas with

temperatures of 100 - 200 million ºC(10 times the temperature of the sun’s core)

500 Megawatts of fusion power


ITER - A Unique Scientific, Technological and

Industrial Project

•Objective - Demonstrate the scientific and

technological feasibility of fusion energy

•Goal - produce a significant fusion power amplification

(10x the energy input):output 500 MW

Seven Party Sharing


Tokamak – 29 m high x 28 m dia. & ~23000 t


ITER Size Comparison

ITER~28 m Tall x 29 m Dia.

Jefferson Memorial

(Washington DC)~29 m Tall (floor to top of dome)

~28 m Tall


ITER Magnetic Field

ITER Field

~10 Tesla or 200,000 x HigherEarths Magnetic Field

~ 0.5 gauss or 0.5x10-4 Tesla


Magnet Energy Comparison

Superconducting

Magnet Energy:~51 GJ

Charles de Gaulle Kinetic Energy: ~38000 t at ~14 km/hr

or

The kinetic energy of ~1900

Audi A5’s each at ~150 km/hr


TF Coil – Mass

Mass of 1 TF Coil: 16 m Tall x 9 m Wide, ~360 t

Boeing 747-300

(Maximum Takeoff Weight) ~377 t

http://image067.mylivepage.ru/chunk67/1488486/1086/Boeing 747-31 1.jpg


Vacuum Vessel Mass Comparison

VV & In-vessel components mass:

~8000 t

19.4 m outside diameter x 11.3 m tall

Eiffel Tower mass:

~7300 t

324 m tall

(Completed 1889)


ITER – Key facts

• Designed to produce 500 MW of fusion power for an extended period of time

• 10 years construction, 20 years operation

• Cost: 10 billion Euros for construction, and 5 billion for operation and decommissioning


Baseline Schedule for 2019 First PlasmaFirst Plasma

2010 2011 2012 2013 2014 2015 2016 2017 2018 2019 2020

ITER Construction

TF Coils (EU)

Tokamak Assembly

Tokamak Basic Machine Assembly

Ex Vessel Assembly

In Vessel Assembly

Start Install CS Start Cryostat Closure

Pump Down & Integrated Commissioning

Start Machine Assembly

2021 2022

ITER Operations

Assembly Phase 2

Assembly Phase 3

Plasma Operations

2023

Buildings & Site

CS Coil

Case Winding Mockups Complete TF10 TF15

VV Fabrication Contract Award VV 05 VV09 VV07

Vacuum Vessel (EU)

CS Final Design Approved CS3L CS3U CS Ready for Machine Assembly

Construction Contract Award Tokamak Bldg 11 RFE


Personnel

26 Different nationalities

Professional

staff

Support

staffTotal

CN 16 4 20

EU 182 125 308

IN 12 16 28

JA 25 7 32

KO 22 5 27

RU 19 3 22

US 24 8 32

Total 301 168 469


Outline



ITER Divertor

ITER Blanket

Design Criteria

Conclusions


Outline



ITER Divertor

ITER Blanket

Design Criteria

Conclusions



• Divertor and Blanket directly face the thermonuclear plasma and cover an area of about 210 + 620 m2, respectively.

• All these removable components are mechanically attached to the Vacuum Vessel or Vessel Ports.

• Max heat released in the PFCs during nominal pulsed operation: 847 MW

– 660 MW nuclear power

– 110 MW alpha heating

– 77 additional heating

• Removed by three independent water loops (~1200 ks/s each) for the blanket + port plugs and one loop for the divertor (~1000 kg/s), at 3 and 4.2 MPa water pressure, ~100 (inlet), ~150 (outlet) °C

Blanket

Divertor


Design inputs for PFCs

• Surface heat flux due to the radiative and particle flux from the plasma.

This is of particular concern for the next generation of fusion machines

where, due to the high number of operating cycles, a thermal fatigue

problem is anticipated. Particularly harmful are the off-normal heat

loads, which are associated to plasma instabilities (such as a plasma

disruption or vertical displacement). Up to some tens of MJ/m2 can be

deposited onto the PFCs in a fraction of a second resulting in melting

and evaporation of the plasma facing material. About 10% of the

discharges are anticipated to end with plasma instability in the next

generation of fusion machines, whereas this figure should decrease to

less than 1% in a commercial reactor.

• Neutron flux from the plasma. The neutron flux is referred to as “wall

loading” and measured in MW/m2. This is the power density transported

by the neutrons produced by the fusion reaction. The wall loading

multiplied by the total plasma burn time gives the neutron fluence, which

is measured in MW-year/m2. The two main effects of the neutron flux

are the volumetric heat deposition and the neutron damage.


Power Handling

HIGH HEAT FLUX

COMPONENTS

FOSSILE FIRED

BOILER WALL

(ABB)

FISSION

REACTOR

(PWR) CORE

ITER DIVERTOR

DESIGN

12/15 mm

ID/OD

HEAT FLUX

- average MW/m2

- maximum MW/m2

0.2

0.3

0.7

1.5

3 – 5

10 – 20

Max heat load MJ/m2

Lifetime years

Nr. of full load cycles

Neutron damage dpa

Materials

-

25

8000

-

Ferritic-Martens. steel

-

4

10

10

Zircaloy - 4

10

~ 5-8

3000 - 16000

0.2

CuCrZr & CFC/W

Coolant

- pressure MPa

- temperature °C

- velocity m/s

- leak rate g/s

Water-Steam

28

280-600

3

<50

Water

15

285-325

5

<50(SG)

Water

4

100 – 150

9 – 11

<10-7

Comparisons


• The volumetric heat deposition has a typical maximum value of a

few W/cm3 in the FW structures and then decreases radially in an

exponential way. It has mainly an impact on the design of the

supporting structures, which thus need to be actively cooled.




Inner Vertical Target

Volumetric heat deposition

Resulting temperature


• The neutron damage will be

the main lifetime limiting

phenomenon in a commercial

reactor. It is measured in

“displacements per atom” (dpa)

that is the number of times an

atom is displaced from its

position in the lattice due to the

action of an impinging particle.

The dpa is proportional to the

neutron fluence. As an example

1 MW-year/m2 causes about 3

and 10 dpa in beryllium and

copper or steel, respectively.

The dpa value is a measure of

the neutron damage. Typical

effects of this damage are

embrittlement and swelling.


Red: He production in appm

Black: dpa

1 year full-time irradiation

ITER is ~0.5 year



• Electromagnetic loads. During a plasma instabilities eddy currents are

induced in the PFCs. These currents interact with the toroidal magnetic

field thus resulting in extremely high forces applied to the PFCs. These

forces can generate mechanical stresses up to a few hundreds of MPa

with a consequent strong impact in the design of the supporting

structures.

Toroidal field

Plasma current

Eddy currents

In PFCs

Poloidal field


Plasma current in MA

Plasma radial position in m

Plasma vertical

position in m



• Surface erosion. The particle flux impinging onto the PFCs causes

surface erosion due to physical sputtering (and also chemical

sputtering in the case of carbon).

– One effect of this phenomenon is that the thickness of the plasma facing

material is progressively reduced.

– Furthermore the eroded particles can migrate into the plasma thus

increasing the radiative energy loss by bremsstrahlung and diluting the

deuterium and tritium concentration.

– Another consequence is that some eroded particle (like carbon or beryllium

oxide) may trap tritium atoms when they redeposit onto the surface of the

PFCs (the so-called “co-deposition”). This results in an increase of the

tritium inventory in the plasma chamber with the associated safety

concerns.


Armour

Materials

Be: port limiter, first wall

W: upper VT,

dome, liner

CFC: lower VT


Beryllium

• Low atomic number

• Oxygen gettering capability

• Absence of chemical sputtering

• High thermal conductivity


CFC

• Longest lifetime

• Absence of melting

• Excellent thermal shock resistance

• Very high thermal conductivity

• Low atomic number


Tungsten

• Lowest sputtering

• Highest melting point

• High thermal conductivity

• No concerns over tritium inventory

• Reference grade: sintered and rolled pure tungsten


Thermal expansion at 300 °C


Armour to heat sink joints

Beryllium CFC or tungsten

Pure copper interlayer

CuCrZr CuCrZr


Terminology, Flat Tile and Monoblock

Armour

Heat sink

Cooling tu be

Coolant

Support ing

st ruct ure


0

0.5

1

1.5

2

%

Manuf. Strain Strain range

Flat tile

Monoblock

M

M

FT

FT


Non-destructive testing:

Be/Cu alloy, W/Cu and Cu/Cu alloy joint

• Ultrasonic examination is the best technique.

• Defects of 2 mm can be detected reliably

• Inspection better performed from the rear side, prior to machining the

cooling channels, when:

– CFC armour

– Be or W armour and fine castellation (< 10x10mm)

• Main issue: differences in the attenuation of the ultrasonic waves (up to

16 dB)


Non-destructive testing: CFC/Cu joint

Ultrasonic examinations

• Ultrasounds can hardly propagate inside the CFC material, therefore the CFC/Cu joint

can only be inspected from the Cu side.

• The acoustic impedance of CFC and Cu is significantly different

• AMC: laser structured surface

CfC

CuCrZr

• Thermographic examination Cu


Transient

Thermography

Inspection

Element to be tested

Reference

Flow direction

0,0 20,0 40,0 60,0 80,0 100,0 120,0138,00,0

10,0

20,0

30,0

40,0

50,0

60,0

70,0

80,0

90,0

100,0

110,0

120,0

130,0

140,0

150,0

159,2

Xmin normiert

0,0 20,0 40,0 60,0 80,0 100,0 120,0138,0-0,3

1,0

2,0

3,0

4,0

5,0

6,0

7,0

8,0

9,3

deltaT norm

0,0 20,0 40,0 60,0 80,0 100,0 120,0138,00,0

10,0

20,0

30,0

40,0

50,0

60,0

70,0

80,0

90,0

100,0

110,0

120,0

130,0

140,0

150,0

159,2

Xmin normiert

- =Tmin,Ref (t)

temperature evolution of

the reference tile

Tmin (t)

temperature evolution of

the tile to be tested

DTmin,Ref (t)

maximum temperature

difference


Outline



ITER Divertor

ITER Blanket

Design Criteria

Conclusions


Outline



ITER Divertor

ITER Blanket

Design Criteria

Conclusions


Divertor system main functions :

• Minimize the helium and impurities

content in the plasma

• Exhaust part of the plasma thermal

power

Divertor


54 Cassettes

in a circular array

held in position by

two concentric

radial rails .

Divertor Cassette Layout


Divertor: Key Facts

Number of Cassette Assemblies: 54

Mass per Cassette Assemblies : ~9 tons

Total Mass: ~490 tons

Armour: CFC / Tungsten

Heat Sink: CuCrZr

Steel Structure: 316L(N)-IG / XM-19

Max total thermal load: 204 MW


Divertor System

54 divertor cassettes

Mass: 470 tons


The divertor CB is reusable to minimise activated

waste; it provides neutron shielding, routes the

water coolant and supports the different PFCs

Divertor System Cassette body


Divertor System


Inner VT

The inner and outer vertical targets (VTs), are the

PFCs that, in their lower part, intercept the

magnetic field lines, and therefore remove the

heat load coming from plasma via conduction,

convection and radiation.

Outer VT

Divertor System Vertical Targets


Divertor System


The “Umbrella”, which is located

below the separatrix, baffles

neutrals, particularly helium.

DomeDivertor SystemThe inner and outer neutral “Particle

Reflector Plates” protect the CB from

plasma radiation, allow transient

movements of the strike points and provide

improved operational flexibility of the

divertor in terms of magnetic configuration.


Summary of TerminologyDivertor System


CF

C

W

Divertor PFC materials choice

Non-active phase (H, He):CFC at the strike points, W on the baffles

All-W from the start of D operations

Rationale:

• Carbon easier to learn with

• No melting easier to test ELM and disruption mitigation strategies before nuclear phase

• T-retention expected to be too high in DT phase with CFC targets


Optical Diagnostic Box

Features to locate and cool diagnostic box integrated in all cassettes

– To allow implementation of mirror box , neutron flux monitor , pick-up coils, bolometers

– Pipes plugged when no diagnostic box is required


Vertical TargetsPlasma-Facing Components


W monoblock

5 MW / m2

10 MW / m2

W monoblock

Copper Interlayer

CuCrZr Heat SinkSmooth Tube

Plasma-Facing Components

316L(N)-IG CFC monoblock

20 MW / m2 10 secCFC monoblockFor first divertor set

Copper Interlayer

CuCrZr Heat SinkTwisted tape: To increase

the margins against the

Critical Heat Flux

Vertical Targets

XM-19

XM-19


XM-19 316L pipes

W Flat Tiles

5 MW / m2

W Tile

Copper Interlayer

CuCrZr Heat Sink

316L(N)

DomePlasma-Facing Components


– The PFCs of the first divertor set are designed to withstand 3000 equivalent

pulses of 400 s duration at nominal parameters, including 300 slow transients

– During normal operational conditions:

• vertical target has a design surface heat flux up to 10 MW/m2 (strike

point region) and 5 MW/m2 (baffle region)

– Under slow transient thermal loading conditions:

• lower divertor vertical target geometry has a design surface heat flux

up to 20 MW/m2 for sub-pulses of less than 10 s

– The dome shall sustain design heat fluxes of up to 5 MW/m2

– The umbrella and the particle reflector plates shall sustain local heat flux up to

10 MW/m2, which can be transiently swept across the surface (about 2 s) as

the plasma is returned to its correct position

Power Handling Quasi-Stationary Design Values


Fundamentals of Water Heat Transfer

Heat Transfer Coeff. (W/m2K)

Wall Temperature (K)

Turbolence flow

Subcooled boiling

Critical Heat Flux

xSaturation Temp.

Divertor cooling regime


Critical Heat Flux


Vertical Target

Medium-Scale Prototype

• W macrobrush:

15 MW/m2

x 1000 cycles

• CFC monoblock

20 MW/m2

x 2000 cycles

• CHF test > 30 MW/m2

Test results


• W monoblocks:

10 MW/m2

x 1000 cycles

• CFC monoblock

10 MW/m2

x 1000 cycles

20 MW/m2

x 1000 cycles

23 MW/m2

x 1000 cycles

Vertical Target

Full-Scale Prototype


1995

2002

1998

Divertor Technology Evolution


Neutron-Irradiation Experiments PARIDE 1 - 4

PARIDE 1:

· temperature: 350°C

· target fluence: 0.5 dpa

PARIDE 2:



High Flux Reactor

Petten, Netherlands

PARIDE 3:



PARIDE 4:


· target fluence: 1 dpa

HHF Technologies Irradiation


HHF Technologies Irradiation

Forschungszentrum Jülich

in der Helmholzgemeinschaft

EURATOM-Association


Main features:

- Three divertor sets Two divertor replacements

- Off-line refurbishment, 2 sets of cassette bodies

Divertor Remote Handling Maintenance Strategy


Outline



ITER Divertor

ITER Blanket

Design Criteria

Conclusions


Outline



ITER Divertor

ITER Blanket

Design Criteria

Conclusions


Blanket system main

functions :

• Exhaust the majority of the

plasma power

• Contribute in providing neutron

shielding to superconducting

coils

• Provide limiting surfaces that

define the plasma boundary

during startup and shutdown.

Blanket System


Module

s 1

-6

Modules 7-10

Module

s 1

1-1

8

~1240 – 2000 mm

~850 –

1240 m

m

Shield Block (semi-permanent)

FW Panel(separable)

http://fusionforenergy.europa.eu/

http://www.iterrf.ru/

http://www.iterkorea.org/


Blanket System: Key Facts

Number of Blanket Modules: 440

Max allowable mass per module: 4.5 tons

Total Mass: 1530 tons

Armour: Beryllium

Heat Sink: CuCrZr

Steel Structure: 316L(N)-IG

n-damage (Be / heat sink / steel): 1.6 / 5.3 / 3.4 (FW) 2.3 (SB) dpa

Max total thermal load: 736 MW


Design of FW Panel

Deep

Central Bolt

Keys reacting

radial torque

Pads reacting

poloidal torque

I shape beam to poloidal torque mitigation

Slitting to reduce

Eddy current


Why shaping is needed ?

First wall

Shield block

Plasma

Toroidal direction

Toroidal gap : 16 mm on the inboard

Inboard wallHorizontal view


But the two sides have to be considered

First wall

Toroidal direction

5 mm

The two situations are equally probable

So chamfering on both sides is necessary

First wall First wall

First wall

5 mm

First wall

Toroidal direction


Shaping of the First Wall panel

First wall

Toroidal direction

First wall

5 mm

16 mm

q

Plasma

RH access

Exaggerated

shaping

Allow good access for RH

Shadow leading edges


Plasma

RH access

Exaggerated

shaping

Allow good access for RH

Shadow leading edges

Shaping of the First Wall panel


First Wall Panels: Design Heat Flux0 150 <= 1.3 MW/m² 34%

1.7 MW/m² <= 68 <= 2.0 MW/m² 15%

3.0 MW/m² <= 162 <= 3.9 MW/m² 37%

4.0 MW/m² <= 60 14%

Total 440 100%


Normal Heat Flux Finger:

Concept with Steel Cooling Pipes

Enhanced Heat Flux Finger:

Concept with Rectangular Channels

SS Back Plate

CuCrZr Alloy

SS Pipes

Be tiles

Be tiles

FW Finger Design


• Slits to reduce EM loads, additional slits on the back were

introduced to minimize the thermal expansion and bowing

(mitigation of EM load)

• Cooling holes are optimized for Water/SS ratio (Improving nuclear

shielding performance)

• Poloidal coolant arrangement make large cutouts feasible

Shielding Module Design

260

280

300

320

340

360

380

0.7 0.75 0.8 0.85 0.9 0.95 1

Volume fraction of SS in blanket shield block

Inboard

TF c

oil n

ucle

ar

heat

(#1-#14)

W/le

g

New Mix, Water30%-0.95g/cc (fendl2.1)

NAR,Water16%-0.9g/cc(fen1)

Water16%-0.9g/cc(fen2)

Water16%-0.9g/cc(fen2.1)


“At the back” of the Blanket Modules…


Flexible attachments

Electrical Straps

Branch Pipes

“At the back” of the Blanket Modules


Manifolds and Cooling Pipe Connections

Water Manifolds Branch Pipes

Shield Block Flow-separator-type

Connector


Blanket Shield Module Attachments

Each Blanket Module is

mechanically attached to the

Vacuum Vessel by means of

4 “flexible cartridges”


The flexible support is designed to take the radial force while provide lateral

flexibility. Radial load, lateral displacement and flange rotation are main

parameters for flexible support.

Blanket Shield Module Flexible Attachments

Vacuum Vessel Blanket Module

Access hole


What else behind the Blanket Module ?

In-Vessel Coils

Blanket Manifold


In-Vessel Coil System

The vacuum vessel (VV) has an In-Vessel (IV) coil system that mitigates

potentially high power loads on the divertor as well as other IV components

while controlling and stabilizing the plasma position.

There are two sets of coil systems installed between the blanket modules

and the outboard inner wall of the vacuum vessel.

The first is the Edge-Localized Mode (ELM) coil system and the second is

the Vertical Stabilization (VS) coil system. The two systems have distinctly

different functions.

The ELM coils generate resonant magnetic perturbations in order to

minimize high power deposition in the divertor induced by ELM heating and

control moderately unstable Resistive Wall Modes (RWM).

The VS coils provide fast vertical stabilization of the plasma.


VS & ELM coils

Upper VS : 240 kAt

Lower VS: 240 kAt

Upper ELM: 90 kAt

Mid ELM : 90 kAt

Lower ELM : 90kAt

Implementation of 27

segregated ELM and 2 VS

coils with feeders


Blanket Manifold Design

- 12 pipes feeding the outboard blanket modules & 8 pipes feeding inboard blanket moduled

- Lines free to expand

- Fixations independent of the ELM coil fixations


Blanket Manifold Routing in the Upper Port


Integrate ELM Coils and Blanket Manifolds

Coils and feeders

integration :

- optimize cross-section of

coils & feeders

- shift stub-keys

- reorganize branch pipes

and electrical straps


Blanket Main Milestones

February 2010 Conceptual Design Review

End 2011 Preliminary Design Review

End 2012 Final Design Review

Mid-2013 Start procurement

End-2019 Last Delivery on Site


Outline



ITER Divertor

ITER Blanket

Design Criteria

Conclusions


Outline



ITER Divertor

ITER Blanket

Design Criteria

Conclusions




Load Category

1

1

2

3

No damage

No damage

Negligible damage

May need to

inspect and

repair/replace


Stress Definition and Classification: Primary Stress

stress

strain

Generated to withstand imposed mechanical loads


Stress Definition and Classification: Secondary Stress

stress

strain

Generated to withstand imposed deformation

(incl. thermal stress)


Stress Definition and Classification: Breakdown of Primary


Stress Definition and Classification: Classification


OutlineOutline



ITER Divertor

ITER Blanket

Design Criteria

Conclusions


OutlineOutline



ITER Divertor

ITER Blanket

Design Criteria

Conclusions


When will fusion be ready?

Lev Artsimovitch’s celebrated reply:

The need is clear. We must aspire and work to deliver fusion as

fast as we can.

"Термоядерная энергия будет получена тогда, когда она станет

необходима человечеству"

“Fusion will be ready when society needs it”

CHECK ITER WEB PAGE http ://www.iter.org

Documents

ITER In-Vessel Components: Blanket, Divertor - staff.polito.itstaff.polito.it/roberto.zanino/sub1/teach_files/current_topics/lect_merola.pdf · ITER In-Vessel Components •Divertor