Transcript Nuclear Fusion Technology

India’s Strategy for Fusion
Energy
R. Srinivasan
Institute for Plasma Research,
Bhat, Gandhinagar – 382 428, India.
Energy scenario in India


Judicious mix of Non-fission & Fission to supply the immediate
needs
Fission Projection: 20 GWe by 2020, 60 GWe by 2030. Aim for 25 %
share by 2050
• Total Power Generated: 180.4 GW
• 65.0 % Total Thermal
• Coal 54.6%
• Gas 9.8%
• Oil 0.7%
• 21.0 % Hydro
• 2.7 % Nuclear
• 11.2 % Renewable
Ministry of Power, India, 31-07-2011
We need to build and exploit
Fusion reactors for
generating power for the
future
Installed capacity
• 1947  1363 MWe
• 1980-81  30,214 MWe
• 1990-91  66,086 MWe#
• 2003
 138,730 MWe
• Growth rates : 9.54,8.14
and 6.26%/yr
• Beyond 2022, intensity fall
by 1.2 %/yr
R. B. Grover et al., Energy Policy (2006) 2834
#
Shah RKD, Indian National
Academy of Engineering (1998)
Installed capacity : Beyond 2050
Without fusion
Shows 890 GWe (34 %) by Nuclear
in 2100
R. Srinivasan and the Indian
DEMO Team, JPFRS (2010)
With 10 % fusion
Fall of contribution from coal near
2100. 2 GWe by 2060 and 250 GWe
(10%) by fusion in 2100
Indian Fusion Program
Power Plant
2050
Fusion Power Reactor
DEMO 2037
• Qualification of Technologies
• Qualification of reactor
components & Process
• Qualification of materials
2 x 1GWe Power
plant by 2060
Indigenous Fusion Experiment
SST-2 2027
ITER Participation 2005
scientific and technological
feasibility of fusion energy
SST-1 1996
Steady State Physics and
related technologies
ADITYA 1984
First Tokamak
Note: Years represent start of project
Technologies to realize DEMO
Technologies needed for DEMO
• Tritium breeding blankets
• Divertor components capable of taking high heat flux
• Fuel Cycle
• Materials which can withstand high heat flux and neutron
irradiation and their joining technologies
• High Power Heating and Current Drive Systems
• Large sized superconducting magnets
Kick-start activities for
SST-2 & DEMO
Programs Initiated at present in 5-Year Plans
•
•
•
•
•
•
•
•
Materials development & qualification program
Blanket technology development program
Divertor technology development program
Fuel Cycle technology development program
Magnet technology development program
NBI system development program
RF system development program
Remote Handling Technology development program
Color
2007-2017
2012-2022
Critical areas
• Fusion grade materials
– Development of structural and functional materials
– Irradiation & test facilities for qualification
– Capacity building for large scale production
• Tritium fuel cycle
– Tritium startup inventory
– Tritium extraction and fuel processing
– Storage
• Large size reactor components and their fabrication
issues
– Blanket, Divertor, Magnets and VV
– Remote handling, fabrication techniques like hipping, EB
wielding
Other Areas
• Technologies related to auxiliary systems
–
–
–
–
–
–
RF sources : gyrotron, klystron, and tetrode tubes
Ion sources, high heat transfer elements, RHVPS
HTC leads
Cryosorption pumps, extruders for pellet injectors
Heat extraction system for Pb-Li loop and He loop
Plant control
Some of these areas can
be addressed with
international collaborations
Specific issues to be addressed before
DEMO
Tritium fuel cycle
• Uncertainty in tritium loss from reactor
• What is the acceptable level of Tritium Breeding Ratio
(TBR)?
• TBR > 1.1 or 1.2 , decides the design of breeding
blanket concept, thickness of breeding zones
• ITER TBM program may not be able to answer this
• Needs an integrated testing of breeding blanket
• Medium size tokamak with D-T operation to produce
tritium may answer this
Reactor Availability Issues
• Indian DEMO is expected to have 30 % availability at the
start and has to be maximized by gaining experience in
operation
• Quantifying reactor availability before DEMO is crucial
• ITER operation may give estimate about availability but
ITER is without breeding blankets
• The maintenance/ repairing requirements of breeding
blankets may not be realized in ITER device
• Remote Handling of such module needs to be developed
by experience
There seems to be a need of a an interim device with all
breeding blankets
Divertor
• ITER will establish the capability of handling heat load
about 5-8 MW/m2
• For DEMO, this will be higher by a factor of 2 (15 -20
MW/m2)
• Presently available materials are W and W-alloys
• Develop new materials to take such high heat flux
• Qualification for high dpa
• Innovative divertor concepts like X-div., liquid div., also
need to explored during design.
Experiments in SST-1 will demonstrate Double null Vs Single
null operations. Innovative concepts will also be tried out
Ignited plasma issues
• Alpha particle will provide the dominant heating
mechanism in DEMO
• Alpha particle heating has to be supported with external
heating
• Identifying the state of plasma operation and control the
power accordingly
• ITER may tell about the alpha heating in presence of
dominant external heating (Q~10)
• ITER experiments may reveal the future direction in this
aspect
Needs DEMO like
machine
Future devices to address these issues
SST-2
• Should act as first step for verifying the choices being
made for DEMO
• A medium size tokamak with pulsed D-T operation
• With breeding blanket at the outboard side
• Should provide the first integrated test of some systems
being developed for DEMO
• Should address the tritium breeding, possible losses
and recovery
• Will be able to address alpha particle issues
• Remote handling of components and maintenance
• Address availability of machine with breeding blankets
SST-2
– Build with existing
technologies
– Pulsed D-T machine
– Low Q machine and less
fusion power output
– Experience in tritium
handling
– Achieving steady Q &
Fusion power output
– Tritium breeding will not be
self-sufficient (should test
the breeding performance)
– Should be the test bed for
all developmental activities
Plasma
parameters
SST-2
R0
4.4
a
1.5
A
3.0
Bt (T)
5.4
Ip(MA)
11
fbs(%)
11.5
Ploss(MW)
40
Pfusion (MW)
100
Paux(MW)
20
Q
5
n/nGW
0.93
<T> keV
4.5
N
1.31
DEMO
•
•
•
•
•
•
DEMO should have most features of the power plant
Thermal efficiency should be maximized
Should couple electricity to the grid
Should address the integrated machine performance
Tritium self-sufficiency should be achieved
Machine availability should be enhanced for realizing a
power plant
• All the issues expected in an ignited plasma scenario
should be addressed in this device
Indian DEMO
– Production of more than 1
GW of net electricity
– with 30 % availability
– Less aggressive (any
improvement will be a
boost)
– Try to improve the
availability
– Performance of reactor and
its optimization
Plasma
parameters
Indian
DEMO
R0
7.7
a
2.6
A
3.0
Bt (T)
6.0
Ip(MA)
17.8
fbs(%)
50
Ploss(MW)
720
Pfusion (MW)
3300
Paux(MW)
110
Q
30
n/nGW
0.93
<T> keV
21.5
N
3.3
Choices for Indian DEMO
•
•
•
•
•
•
•
•
•
TF with Nb3Sn
Plasma facing components with W and W-alloys
Blanket concept : LLCB with Pb-Li and LiTi2O3
Structural : IN-RAFMS
VV : SS316LN
Shielding : borated steel
Allowable dpa on structural material < 50 (?)
Double null or Single null
Thermal efficiency with 30 %
Design will have many variants with mid-term, longterm projections. Few choices on dream materials or
concepts have to be made and pursued
Indian R & D efforts
RAFMS –
Structural Material Development
2 nos. of 3000 Kg commercial melts completed
 Chemical Composition under control
Forging and rolling into Plates completed
IN-RAFMS INGOT
Forging of IN-RAFMS
Characterisation under progress
IN-RAFMS Plates
22
Photomicrograph of Li2TiO3 after sintering at
1250oC, 4 hours, by SOL-GEL Process:
Other materials :
• SS316LN
• borated steel
Samples of various tungsten materials
produced using powder metallurgical route
Large scale production of materials
has to be established through
Indian industries/national labs
Test facilities to qualify materials
• High- heat flux facility
– 200 kW of electron beam testing facility to test the helium and
water cooled components
– To simulate HHF during ELMs, a test facility is being planned
• Neutron Irradiation facility
–
–
–
–
fission reactors available to test up to a fraction of dpa
Effect of 14 MeV neutrons will be important to qualify materials
SST2 will be used as the test facility for 14 MeV neutrons
Interested to participate in other international irradiation facilities
Preliminary results for CICC
Cross section of
20x20mm CICC
containing 336
wires of 0.8mm
dia out of which
48 nos. are SC
wires
0.8mm dia SC
wire having 492
Nb-Ti Filaments
• CICC developed at IPR and BARC have shown that 11000 A of current
could be passed at 6 K against designed value of 10000 A at 4.5 K
supercritical helium.
• This hybrid conductor has about 25% less superconductor compared to
that of SST-I conductor.
Magnet testing
Large Experimental Cryostat (6 m high, 5 m diameter
VPI Facility developed
Internal Tin Nb3Sn strands being characterized
Lead-Lithium Loop at IPR
Experiment: Corrosion Studies
Loop Parameters:
Hot leg temperature : 550 C
Temperature difference between
hot and cold legs ~95 C
Flow velocity - 5 cm/sec.
Corrosion Sample - RAFMS
Neutral Beams : Negative ion beams
RF based Source
Integrated source in
operation (source
under IPP agreement)
Experimental program of production of RF based Negative ion



Experience in coupling of RF power to produce plasma in the source;
Characterization of plasma
Study various filter field configurations for optimal solution
Beam extraction , acceleration & characterization
28
Human Resource Development
Program
• Initiated to bring various labs, universities and
industries to participate in the R&D program of
fusion reactor
• Provided engineering services to many ITER
tasks and this is available for our own program
• These activities will nucleate various working
groups required for the fusion reactor
• Future
human resources for fusion will be
developed through this program
• Need of innovative ideas to attract young
minds to sustain this long term program
Conclusions
• Indian DEMO roadmap is driven by the energy requirement
• The commercial power plant is expected by 2060. If this
can be accelerated, this will have major impact on Indian
energy scenario
• Materials and other fusion technologies pursued with
definite goals
• Network research and a strong interaction between R & D
labs and the Indian industries is being pursued
• Through BRFST, a reasonable achievement in network
research has been achieved and is expected to grow
rapidly in the coming decades
• Fusion program has a strong momentum now which is
going to become more intense and focused in the coming
decade
• Interaction with like minded groups around the world is
going to play a crucial peer group role in these
developments.
Thank you