Slide 1 Mario Merola – Nuclear Fusion Engineering Masters, Torino 24 th January 2011 Mario Merola - ITER International Organization Internal Components Division Head ITER In-Vessel Components: Blanket, 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
Slide 2Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Nuclear Fusion and the ITER Project
ITER In-Vessel Components
ITER Divertor
ITER Blanket
Design Criteria
Conclusions
Outline
Slide 3Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Outline
Nuclear Fusion and the ITER Project
ITER In-Vessel Components
ITER Divertor
ITER Blanket
Design Criteria
Conclusions
Slide 4Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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:
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
Slide 6Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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
Slide 7Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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”.
Slide 8Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
ITER Agreement
The agreement was signed on the
21st November 2006
at the Elysée Palace in Paris
International Organization started.
Slide 9Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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
Slide 10Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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
Slide 11Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Tokamak – 29 m high x 28 m dia. & ~23000 t
Slide 12Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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
Slide 13Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
ITER Magnetic Field
ITER Field
~10 Tesla or 200,000 x HigherEarths Magnetic Field
~ 0.5 gauss or 0.5x10-4 Tesla
Slide 14Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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
Slide 15Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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
Slide 16Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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)
Slide 17Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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
Slide 18Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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
Slide 19Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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
Slide 20Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Outline
Nuclear Fusion and the ITER Project
ITER In-Vessel Components
ITER Divertor
ITER Blanket
Design Criteria
Conclusions
Slide 21Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Outline
Nuclear Fusion and the ITER Project
ITER In-Vessel Components
ITER Divertor
ITER Blanket
Design Criteria
Conclusions
Slide 22Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
ITER In-Vessel Components
• 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
Slide 23Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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.
Slide 24Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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
Slide 25Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
• 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.
Design inputs for PFCs
Slide 26Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Design inputs for PFCs
Inner Vertical Target
Volumetric heat deposition
Resulting temperature
Slide 27Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
• 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.
Design inputs for PFCs
Red: He production in appm
Black: dpa
1 year full-time irradiation
ITER is ~0.5 year
Slide 28Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Design inputs for PFCs
• 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
Slide 29Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Plasma current in MA
Plasma radial position in m
Plasma vertical
position in m
Slide 30Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Design inputs for PFCs
• 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.
Slide 31Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Armour
Materials
Be: port limiter, first wall
W: upper VT,
dome, liner
CFC: lower VT
Slide 32Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Beryllium
• Low atomic number
• Oxygen gettering capability
• Absence of chemical sputtering
• High thermal conductivity
Slide 33Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
CFC
• Longest lifetime
• Absence of melting
• Excellent thermal shock resistance
• Very high thermal conductivity
• Low atomic number
Slide 34Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Tungsten
• Lowest sputtering
• Highest melting point
• High thermal conductivity
• No concerns over tritium inventory
• Reference grade: sintered and rolled pure tungsten
Slide 35Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Thermal expansion at 300 °C
Slide 36Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Armour to heat sink joints
Beryllium CFC or tungsten
Pure copper interlayer
CuCrZr CuCrZr
Slide 37Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Terminology, Flat Tile and Monoblock
Armour
Heat sink
Cooling tu be
Coolant
Support ing
st ruct ure
Slide 38Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
0
0.5
1
1.5
2
%
Manuf. Strain Strain range
Flat tile
Monoblock
M
M
FT
FT
Slide 39Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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)
Slide 40Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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
Slide 41Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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
Slide 42Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Outline
Nuclear Fusion and the ITER Project
ITER In-Vessel Components
ITER Divertor
ITER Blanket
Design Criteria
Conclusions
Slide 43Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Outline
Nuclear Fusion and the ITER Project
ITER In-Vessel Components
ITER Divertor
ITER Blanket
Design Criteria
Conclusions
Slide 44Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Divertor system main functions :
• Minimize the helium and impurities
content in the plasma
• Exhaust part of the plasma thermal
power
Divertor
Slide 45Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
54 Cassettes
in a circular array
held in position by
two concentric
radial rails .
Divertor Cassette Layout
Slide 46Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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
Slide 47Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Divertor System
54 divertor cassettes
Mass: 470 tons
Slide 48Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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
Slide 49Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Divertor System
Slide 50Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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
Slide 51Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Divertor System
Slide 52Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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.
Slide 53Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Summary of TerminologyDivertor System
Slide 54Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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
Slide 55Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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
Slide 56Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Vertical TargetsPlasma-Facing Components
Slide 57Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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
Slide 58Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
XM-19 316L pipes
W Flat Tiles
5 MW / m2
W Tile
Copper Interlayer
CuCrZr Heat Sink
316L(N)
DomePlasma-Facing Components
Slide 59Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
– 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
Slide 60Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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
Slide 61Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Critical Heat Flux
Slide 62Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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
Slide 63Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
• 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
Slide 64Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
1995
2002
1998
Divertor Technology Evolution
Slide 65Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Neutron-Irradiation Experiments PARIDE 1 - 4
PARIDE 1:
· temperature: 350°C
· target fluence: 0.5 dpa
PARIDE 2:
· temperature: 700°C
· target fluence: 0.5 dpa
High Flux Reactor
Petten, Netherlands
PARIDE 3:
· temperature: 200°C
· target fluence: 0.2 dpa
PARIDE 4:
· temperature: 200°C
· target fluence: 1 dpa
HHF Technologies Irradiation
Slide 66Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
HHF Technologies Irradiation
Forschungszentrum Jülich
in der Helmholzgemeinschaft
EURATOM-Association
Slide 67Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Main features:
- Three divertor sets Two divertor replacements
- Off-line refurbishment, 2 sets of cassette bodies
Divertor Remote Handling Maintenance Strategy
Slide 68Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Outline
Nuclear Fusion and the ITER Project
ITER In-Vessel Components
ITER Divertor
ITER Blanket
Design Criteria
Conclusions
Slide 69Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Outline
Nuclear Fusion and the ITER Project
ITER In-Vessel Components
ITER Divertor
ITER Blanket
Design Criteria
Conclusions
Slide 70Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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
Slide 71Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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)
Slide 72Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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
Slide 73Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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
Slide 74Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Why shaping is needed ?
First wall
Shield block
Plasma
Toroidal direction
Toroidal gap : 16 mm on the inboard
Inboard wallHorizontal view
Slide 75Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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
Slide 76Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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
Slide 77Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Plasma
RH access
Exaggerated
shaping
Allow good access for RH
Shadow leading edges
Shaping of the First Wall panel
Slide 78Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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%
Slide 79Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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
Slide 80Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
• 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)
Slide 81Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
“At the back” of the Blanket Modules…
Slide 82Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Flexible attachments
Electrical Straps
Branch Pipes
“At the back” of the Blanket Modules
Slide 83Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Manifolds and Cooling Pipe Connections
Water Manifolds Branch Pipes
Shield Block Flow-separator-type
Connector
Slide 84Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Blanket Shield Module Attachments
Each Blanket Module is
mechanically attached to the
Vacuum Vessel by means of
4 “flexible cartridges”
Slide 85Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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
Slide 86Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
What else behind the Blanket Module ?
In-Vessel Coils
Blanket Manifold
Slide 87Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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.
Slide 88Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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
Slide 89Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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
Slide 90Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Blanket Manifold Routing in the Upper Port
Slide 91Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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
Slide 92Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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
Slide 93Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Outline
Nuclear Fusion and the ITER Project
ITER In-Vessel Components
ITER Divertor
ITER Blanket
Design Criteria
Conclusions
Slide 94Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Outline
Nuclear Fusion and the ITER Project
ITER In-Vessel Components
ITER Divertor
ITER Blanket
Design Criteria
Conclusions
Slide 95Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Slide 96Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Slide 97Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Load Category
1
1
2
3
No damage
No damage
Negligible damage
May need to
inspect and
repair/replace
Slide 98Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Stress Definition and Classification: Primary Stress
stress
strain
Generated to withstand imposed mechanical loads
Slide 99Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Stress Definition and Classification: Secondary Stress
stress
strain
Generated to withstand imposed deformation
(incl. thermal stress)
Slide 100Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Stress Definition and Classification: Breakdown of Primary
Slide 101Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
Stress Definition and Classification: Classification
Slide 102Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
OutlineOutline
Nuclear Fusion and the ITER Project
ITER In-Vessel Components
ITER Divertor
ITER Blanket
Design Criteria
Conclusions
Slide 103Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
OutlineOutline
Nuclear Fusion and the ITER Project
ITER In-Vessel Components
ITER Divertor
ITER Blanket
Design Criteria
Conclusions
Slide 104Mario Merola – Nuclear Fusion Engineering Masters, Torino 24th January 2011
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