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Technological readiness comparison for Helical and Tokamak DEMO
A. Sagara1, R. Wolf2 and H. Neilson3
1) National Institute for Fusion Science, Japan 2) Max-Planck-Institut fur Plasmaphysik, Germany
3) Princeton Plasma Physics Laboratory, USA
3rd IAEA DEMO Programme Workshop
11-14, May 2015 , Hefei
@University of Science and Technology of China
Special Session 2
Topics 1 : Contribution of integrated devices to closing the gaps (1)
3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 1 / 34
(1) This opportunity is a very good chance to know an example
of young researcher’s opinions in Japan on Technical Readiness Level (TRL) for Helical/Stellarater systems
in comparison with Tokamak systems.
(2) TRL is one of practical measures for discussing and having common
senses in the research communities. However, it should be noted that
TRL assessments must be made against established criteria*,
taking care not to inflate assessments and being as objective as possible.
There are no clear timelines pointed out in the TRL tables.
(3) In this presentation, therefore, some comments or new information
are added as far as possible for enhancing workshop discussions.
*USDOE example: DOE G 413.3-4A, “Technology Readiness Assessment Guide,” https://www.directives.doe.gov/directives/0413.3-EGuide-04a. (But
currently not applied to fusion by DOE)
3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 2 / 34
Ryuta Kasada*1, Takuya Goto*2, Shinsuke Fujioka*3, Ryoji Hiwatari*4, Naoyuki Oyama*5, Hiroyasu Tanigawa*5, Junichi Miyazawa*6
Young Scientists Special Interest Group on Fusion Reactor Realization
1Institute of Advanced Energy, Kyoto University 2NIFS
3Instituteo of Laser Engineering, Osaka University 4CRIEPI
5JAEA
US-Japan Workshop on
Fusion Power Plants and Related Advanced Technologies
8-9 March 2012
UC San Diego, Center for Magnetic Recording Research
3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 3 / 34
The Young Scientists Special Interest Group on Fusion Reactor Realization is an active volunteer group which consists of the spirited young researchers beyond the frame of organizations or specialties who has a burning ambition to realize fusion reactors in their lifetime.
This work has been carried out by this voluntary group including many young researchers.
3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 4 / 34
Visualization of Current States of the Key
Technologies
About TRL
Prerequisites for the TRL definition
Results of TRL evaluation
Tokamak
Helical
General issues
Perspective of R&D on the Key Technologies
3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 5 / 34
After the FUKUSHIMA accident (3.11), R&Ds on energy technologies in Japan is being judged distinctly by the stakeholders whether they can contribute to the energy demand in near-future.
R&D on fusion reactors will be also judged from a viewpoint of not plasma science but energy development.
So some of young researchers (DEMO generation) decided to prepare the explanation to state the current and future of the key technologies for realization of fusion reactors.
A technology readiness levels (TRL) assessment is a powerful tool to visualize the maturity of the technologies under development.
3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 6 / 34
A Technology Readiness Level (TRL) describes the maturity of a given technology
relative to its development cycle. TRL assessment were originally developed by NASA is to provide a common language among the
technology developers, engineers who will adopt/use the technology, and other stakeholders.
Dr. Tillack and the ARIES Team have already reported the TRL evaluation
for the fusion reactor.(2008)
While many types of reactor concepts have been studied in Japan, there were no activity to evaluate the technology maturity based on TRL.
TRL 1 2 3 4 5 6 7 8 9
Tech. A
Tech. B
Tech. C
Tech. D
completed In progress 3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 7 / 34
Ref.
UCSD-CER-08-01
3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 8 / 34
To identify Japanese customer (or young researchers) needs: Our target is to realize fusion reactor in 2050 when most of us will
alive.
For Magnetic confinement fusion: We set TRL5-6 for ITER and TRL7-9 for DEMO reactor.
Our TRL evaluation assumes two different scales of DEMO reactors allowing the development of components; such as 1GW fusion power (0.3GWe class) and a 3GW fusion power (1GWe class).
For Inertial fusion
Electrical power supply for solo reactor can be 0.2GWe because 5 sets of the reactor can generate 1GWe.
We set TRL6-7 for DEMO and TRL8-9 for prototype power reactor.
3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 9 / 34
TRL Definition
State for magnetic
confinement fusion
(incl. R&D in DEMO)
State for inertial fusion
TRL1 Basic principles observed and reported
Proof of concept Proof of concept TRL2
Technology concepts and/or applications formulated
TRL3 Analytical and experimental demonstration of critical function and/or proof of concept
Component validation Component validation
(including validation of
reactor-core physics and fusion
burning)
TRL4 Component and/or bench-scale validation in a laboratory environment
TRL5 Component and/or breadboard validation in a relevant environment
ITER (Exp. Reactor) TRL6
System/subsystem model or prototype demonstration in relevant environment
Demonstration reactor TRL7
System prototype demonstration in prototypic environment
DEMO reactor TRL8 Actual system completed and qualified through test and demonstration
Prototype reactor TRL9
Actual system proven through successful operations
Remarks After GNEP’s definition Main machines: ITER &
DEMO
Main machines: (NIF, LMJ, FIREX etc),
DEMO, Prototype
GNEP: Global Nuclear Energy Partnership 3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 10 / 34
Mt. Takao
599m
Mt. Fuji
3,776m
Everest 8,848m
Attention: Higher mountain is more different to climb; TRL is not same quantity.
Top
Top
Top
Pictures from Wikipedia and Google
half
half
half
Because the expected outcome and the degree of difficulty for commercialization are
different. TRL is the evaluation of individual technical achievement. It is important to
clarify that clear criteria and specific measures rather than the numerical value of TRL.
Yama-girl
Pro. Climber
3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 11 / 34
Summarized by Ryoji HIWATARI & Naoyuki OYAMA
3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 12 / 34
Currently it has become possible to build a technology roadmap by the accumulation of knowledge based on the past reactor designs.
Dr. Okano team build a technology map based on the WBS (work breakdown structure) which consists of 18 area including >1000 issues.
We can evaluate the TRL on the WBS.
3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 13 / 34
TRLs are evaluated for the different electrical power plants of 0.3GWe and 1GWe.
Even supposing the electrical output, fusion power output and the size is not determined.
We show typical examples of the combination of size, plasma performance and engineering conditions.
And we evaluate the TRL on these examples.
3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 14 / 34
Major radius: 6.5m Major radius: 8.5m
0.2GWe for
thermal
efficiency of
30% (0.3GWe for thermal
efficiency of
40%)
• bN>3.0⇒ITER high-performance plasma
• Thermal efficiency > 30%⇒to be tested in ITER TBM
• Higher strength magnetic field is
needed for TF coil.
• Neutron wall load: ~1.5MW/m2
• Full-sector removal maintenance
• bN=2.0以上⇒ITER normal plasma • Thermal efficiency > 30%⇒to be
tested in ITER TBM
• Higher strength magnetic field is
needed for bigger TF coil.
• Neutron wall load: ~0.75MW/m2
• Module maintenance
1GWe for
thermal
efficiency of
40% (0.6GWe for thermal
efficiency of
30%)
• bN>4.0以上⇒Advanced plasma ( over ITER) with conducting walls in blanket
• Thermal efficiency > 40%⇒advanced blanket
• Higher strength magnetic field is
needed for TF coil
• Neutron wall load: ~3.5MW/m2
• Full-sector removal maintenance
• NBI system efficiency: >30%
• bN>3.0⇒ITER high-perfomance plasma
• Thermal efficiency > 40%⇒advanced blanket
• Higher strength magnetic field is
needed for bigger TF coil.
• Neutron wall load: ~2.0MW/m2
• Module maintenance
3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 15 / 34
Scale Lab Large Device ITER DEMO
TRL 1 2 3 4 5 6 7 8 9
System design
Core plasma βN~4.0
βN~2.0
TF coil 16T
13T
Blanket T.E. 40%
T.E. 30%
Under discussion on DEMO mission.
ITER normal operation
JT-60SA
ITER scale
Higher magnetic field coil needs R&D on materials.
WCSB-WLK:ITER-TBM
Analysis for the selection and specification required to determine the primary design is underway.
・When the DEMO design is based on the findings of ITER, there is no Tokamak-specific bottleneck technology.
・TF coil technology can be a bottleneck, which depends on DEMO mission.
・Blanket technologies are not specific for Tokamak.
3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 16 / 34
Construction of a new 600 W He refrigerator/liquefier A 600 W He refrigerator/liquefier with an additional function
is replacing.
・ Its completion will be on March 31, 2015.
・ He coolant of variable temperature can be supplied to
test facilities for superconductors and magnets.
Specifications
・Refrigeration / liquefaction capacities
600W at 4.5K / 250L/h
・Supply capacities of He coolant
350 W at 4.55 K (50 g/s SHe)
1.0 kW at 20 K – 30 K (18 g/s GHe)
1.5 kW at 40 K – 50 K (20 g/s GHe)
A unique test facility of superconductors in the world by
operating with a 13 T magnet and a 75 kA power supply
13 T & 70 cm boar 75 kA power supply
Corroboration researches of wide scope are acceptable.
・HTC and new superconductor tests with the 4.5 K – 50 K
temperature range
・CIC conductors and coils cooled by SHe
・Conventional superconductors such as NbTi and Nb3Sn
Unloading of the cold box
Unloading of
a He compressor
S. Hamaguchi, et al., P6-11, 24th ITC
New
function
3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 17 / 34
ブランケット交換で性能向上可能、主・副半径をどうする 主要課題
冷却材条件(冷媒、圧力、温度) 変更 、発電システム総交換 必要性
原型炉概念Demo-CREST(主半径7.5m)で提案中
Major radius: 6.5m Major radius: 8.5m
0.2GWe for
thermal
efficiency of
30% (0.3GWe for thermal
efficiency of
40%)
• bN>3.0⇒ITER high-performance plasma
• Thermal efficiency > 30%⇒to be tested in ITER TBM
• Higher strength magnetic field is
needed for TF coil.
• Neutron wall load: ~1.5MW/m2
• Full-sector removal maintenance
• bN=2.0以上⇒ITER normal plasma • Thermal efficiency > 30%⇒to be
tested in ITER TBM
• Higher strength magnetic field is
needed for bigger TF coil.
• Neutron wall load: ~0.75MW/m2
• Module maintenance
1GWe for
thermal
efficiency of
40% (0.6GWe for thermal
efficiency of
30%)
• bN>4.0以上⇒Advanced plasma ( over ITER) with conducting walls in blanket
• Thermal efficiency > 40%⇒advanced blanket
• Higher strength magnetic field is
needed for TF coil
• Neutron wall load: ~3.5MW/m2
• Full-sector removal maintenance
• NBI system efficiency: >30%
• bN>3.0⇒ITER high-perfomance plasma
• Thermal efficiency > 40%⇒advanced blanket
• Higher strength magnetic field is
needed for bigger TF coil.
• Neutron wall load: ~2.0MW/m2
• Module maintenance
Upgrade by replacement of blanket 3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 18 / 34
U.S. ARIES team has compared “advanced” vs.
“conservative” assumptions for 1 GWe tokamak power plant designs
Conservative ARIES-ACT2
Advanced ARIES-ACT1
Ph
ys
ics
A
ss
um
pti
on
s
βNtotal 2.6 (no wall) 5.8 (with wall)
H98 1.25 1.65
n/nGr minimize minimize
Te
ch
no
log
y
As
su
mp
tio
ns
Blanket: DCLL (RAFM)
(ηth~0.44)
SCLL (SiC-comp)
(ηth~0.58)
qdivpeak <10 MW/m2 <15 MW/m2
Re
su
lts
R 9.75 m 6.25 m
BTaxis 8.75 6.0
Pfus 2.6 GW 1.8 GW
QDT 25.0 42.5
QENG 3.1 6.5
Special Issue (12 papers): Fusion Science and Technology 67, Vol. 1 (Jan. 2015)
3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 19 / 34
Summarized by Takuya GOTO & Junichi MIYAZAWA
3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 20 / 34
The NIFS fusion engineering research project (FERP):
has initiated the conceptual design activity of LHD-type helical reactor “FFHR-d1” by 13 task groups.
Taking the knowledge of past commercial type design concept FFHR-2m , the project aims to demonstrate the next generation reactors as soon as possible.
TRL evaluation:
Identification of major R & D items by the Task activities.
Because of no ITER-scale device for helical system, DD experiment phase, large scale engineering component tests, and numerical simulation is defined as TRL=5, 6.
Specific-issues for helical type reactor
Original fabrication and maintenance method should be developed for 3D
structure containing winding helical coil and blanket.
Due to the big size, building and fabrication process is probably special.
A big advantage is that constant heating input is not needed.
3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 21 / 34
Scale Lab LHD/ITER-EDA LHD-DD/ITER
& DEMO-EDA
DEMO
TRL 1 2 3 4 5 6 7 8 9
System integration
Core plasma
Supeconducting magnet coil with low Temp. system
Blanket technologies
Internal components
Plasma heating
Plasma measurements
Tritium & safety
Power supply system
Building
To determine key parameters of the reactor FFHR-d1.
By extrapolation of the LHD experiments and numerical analysis.
Various conductors are examined. It is necessary to develop an
efficient method of winding.
Materials evaluation and neutronics design have
been progressed mainly for molten salt blanket.
Studies on 3D layout design and remote maintenance is needed.
Long pulse, high-efficiency
Not helical-specific issue.
Needs for active control of magnetic axis.
Scale is larger than the tokamak building. The schedule should be
consistent with the main component construction process.
Under validation for LHD-DD experiments
(using LHD and ITER tech.)
3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 22 / 34
Progress of 100-kA-Class HTS Conductor and Coil Design
HTS Conductor
Current 100 kA ×
1 hour
FFHR-d1 Helical Coil
(Helium gas cooling 20 K)
Mechanical
Lap Joint
4.2 K
100 kA-class HTS
conductor design (2014)
100-kA-class High-Temperature Superconducting
(HTS) conductor renovates the world record 100 kA current sustained for 1 hour [1, 2]
Tensile test for a single-tape joint was conducted Contact pressure of 50 MPa is needed to withstand the
shear strength evaluated by 3D-FEM [3]
(1) N. Yanagi et al., FIP/P8-21, 25th IAEA Fusion Energy
Conference (2014), St. Petersburg, Russia.
(2) S. Ito et al., Plasma and Fusion Research 9 (2014)
3405086.
(3) S. Ito et al., IEEE Transactions on Applied
Superconductivity 25 (2015) 4201205.
3D-FEM analysis of EM stress
on the helical coil and conductor
Tensile and shear strength test
for single-tape joint
Collaboration with
Tohoku Univ.
Maximum shear
strength obtained by 3D-FEM
analysis 50 MPa
3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 23 / 34
(comments from Helias) C. Beidler, F. Warmer, R. Wolf et al.
HELIAS as an advancement of the optimized stellarator W7-X
Central points which are also important for TRL
• Plasma performance: W7-X designed is based on an self-consistent optimization
w.r.t. neoclassical transport, plasma equilibrium, exhaust concept and fast ion
confinement
Good confinement properties and high b require high density
Development of scenarios with optimized fuelling, density and impurity control
Issue: Conditions to excite fast ion driven instabilities difficult to achieve in W7-X
• Plasma heating based on steady-state ECRH
Extrapolation to high density needs to be demonstrated
• Comparatively low magnetic field ITER superconductor and coil technology is sufficient
• Blanket integration under investigation ( Karlsruhe Institute of Technology)
• EU Roadmap assumes intermediate step between W7-X and commercial
Stellarator-PP, in particular for demonstrating burning plasma and fast ion confinement (further improvement of configuration)
Step to commercial Stellarator-PP based on this and technology development
from tokamak DEMO
Decision point how to proceed when W7-X high performance steady-state
plasmas have been demonstrated (~2025) 3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 24 / 34
Summarized by Ryuta KASADA & Hiroyasu TANIGAWA
3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 25 / 34
Scale Lab Large device ITER DEMO
TRL 1 2 3 4 5 6 7 8 9
Blanket: RAF-Solid-Water (BA etc) (ITER-TBM
etc)
(ITER-TBM
etc)
(DEMO-
TBM etc)
Blanket : advanced
Diverter (for ITER) Should be discussed
Diverter (for DEMO)
S.C. coil (for ITER) Nb3Sn+JJ1+Insulator
S.C. coil (advanced) Nb3Al+?St. Mat. +?Insulator
Tritium fuel cycling system
Vacuum tech. (for ITER) Should be discussed
Vacuum tech. (for DEMO)
Tritium recovery system
Tritium safety system
ITER-TBM like component tests is needed.
Effect of neutron irradiation should be examined.
No candidates for structural material and insulator,
Significant advance from ITER is needed for steady state operation of DEMO.
3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 26 / 34
Flinak Loop
SC magnet (3T
perp. to flow)
Li-Pb Loop
Oroshhi-2 in NIFS(2015~) A. Sagara et al., FS&T, in press
3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 27 / 34
LiPb loop
FLiNaK loop
3 T superconducting
magnet
(EM: electro- magnetic)
Phased experiments on MHD effects on LiPb Corrosion and mass transfer under no-equilibrium and DT
Heat transfer on FLiNaK under high B Hydrogen charging and recovery
Metal powder mixed system Operation of S-CO2 system for Flinak
A. Sagara, FED 89 (2014) 2114–2120
Future plan
10MPa S-CO2 compressor
and turbine system
~100kW
A. Sagara et al., FS&T, in press
nano-fluid
3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 28 / 34
Tritium cycle
to reduce the total inventory
R. Sakamoto, NIFS Annual repo. 2010-2011. A. Sagara et al., Fusion Eng. Design 87 (2012) 594.
3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 29 / 34
Scale Lab Large device ITER DEMO
TRL 1 2 3 4 5 6 7 8 9
Recovery of Li
from seawater
Enrichment of
Li-isotope
Back-end tech.
Only basic research in Japan.The Hg amalgam is not suitable for environments.)
Cost can be drastically reduced due to the demand of Li-ion buttery.
• Although limited to the magnetic confinement fusion systems, with respect to the
divertor technology, the existing technology intended for ITER is a Dead-End
Technology. It is necessary to expedite the technology development for DEMO.
• For handling tritium in large facilities, the challenge is particularly tritium recovery
system. The first entire operation of the fuel circulation system is examined in ITER.
Extraction and proposals for improvements of the problem is expected. Towards
the DEMO reactor, development of tritium measurement-control technology,
specifically for the technology to predict and measure the inventory is needed.
• Lithium isotope enrichment technology development is very important.
These common issues mostly need facilities such
as IFMIF for the evaluation of blanket, divertor
evaluation facility, and Tritium pilot plant, in addition
to ITER.
3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 30 / 34
As conclusion of this talk
3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 31 / 34
As results of the TRL evaluation for the realization of three types of fusion reactor concept, it is seen that the specialty of each concept has high TRL but some of engineering issues has extremely low TRL. For example, TRL on measurement and control of burning plasma scenario is
very low.
For blanket technologies including structural materials, smooth progress
has been confirmed by TRL evaluation. Now jump to the R&D on "mass manufacturing technology" is required.
TRL related to the tritium fuel system in all concepts is relatively low. R&D with large-scale handling is needed.
It is essential to establish a method for lithium procurement. As for the magnetic confinement concepts, it is necessary to harmonize
research and development of the divertor component based on their integrated design because the current technology is probably dead-end one for the DEMO divertor
There are many difficulties in engineering issues from ITER to DEMO. It is necessary to set the test phase in DEMO (like a DEMO-TBR) and to make engineering machines to test diverter, blanket with tritium handling.
3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 32 / 34
The TRL assessment summarized the current state of the technical challenges of each element, clarified the criteria and lead to the "visualization" of technological bottleneck for the realization of various fusion reactor concepts.
The young researchers have successfully obtained the point of view of reactor design aiming to realize fusion reactors and mutual understanding between young researchers of different fields in fusion engineering and science has been promoted. 3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 33 / 34
Summary of comments 1. TRL assessments must be made against
established criteria*, taking care not to inflate
assessments and being as objective as possible.
2. There are no clear timelines pointed out in the TRL
tables. Feedforward to a road map is expected.
3. To fulfil all the required developments in the orange
bars, a huge effort is required. Compatibility with
the current research efforts should be assessed.
4. While ITER will be the first test-bed for a fusion
blanket, other developments towards DEMO are
not so clear.
5. For helical system, without burning devices,
extrapolation to the core plasma is the key issue.
3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 34 / 34
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