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Overview of the ARIES-CSCompact Stellarator Power Plant Study
Farrokh Najmabadi and the ARIES TeamUC San Diego
Japan-US Workshop on Fusion Power Plants and Related Advanced Technologies with EU participationFebruary 5-7, 2007Kyoto, Japan
Electronic copy: http://aries.ucsd.edu/najmabadi/TALKSARIES Web Site: http://aries.ucsd.edu/aries/
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For ARIES Publications, see: http://aries.ucsd.edu/For ARIES Publications, see: http://aries.ucsd.edu/
GIT
Boeing GA
INEL
MIT ORNL
PPPL RPI
U.W.
CollaborationsFZK
UC San Diego
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Goals of the ARIES-CS Study
Can compact stellarator power plants be similar in size to advanced tokamak power plants?
Reduce aspect ratio while maintaining “good” stellarator properties.
Include relevant power plants issues (α particle loss, Divertor, Practical coils).
Identify key areas for R&D (what areas make a big difference)
Impact of complex shape and geometry
Configuration, assembly, and maintenance drives the design
Complexity-driven constraints (e.g., superconducting magnets)
Complex 3-D analysis (e.g., CAD/MCNP interface for 3-D neutronics)
Manufacturability (feasibility and Cost)
First design of a compact stellarator power plant
Design is pushed in many areas to uncover difficulties
Can compact stellarator power plants be similar in size to advanced tokamak power plants?
Reduce aspect ratio while maintaining “good” stellarator properties.
Include relevant power plants issues (α particle loss, Divertor, Practical coils).
Identify key areas for R&D (what areas make a big difference)
Impact of complex shape and geometry
Configuration, assembly, and maintenance drives the design
Complexity-driven constraints (e.g., superconducting magnets)
Complex 3-D analysis (e.g., CAD/MCNP interface for 3-D neutronics)
Manufacturability (feasibility and Cost)
First design of a compact stellarator power plant
Design is pushed in many areas to uncover difficulties
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Goal: Stellarator Power Plants Similar in Size to Tokamak Power Plants
Approach:
Physics: Reduce aspect ratio while maintaining “good” stellarator properties.
Engineering: Reduce the required minimum coil-plasma distance.
Approach:
Physics: Reduce aspect ratio while maintaining “good” stellarator properties.
Engineering: Reduce the required minimum coil-plasma distance.
0
2
4
6
8
10
12
14
0 4 8 12 16 20 24
Pla
sma
Asp
ect
Rat
io <
R>
/<a>
Average Major Radius <R> (m)
Stellarator Reactors
HSR-5
HSR-4SPPS
CompactStellaratorReactorsARIES
AT ARIESRS
FFHR-1
MHR-S
Circle area ~ plasma areaTokamak Reactors
Need a factor of 2-3 reductionNeed a factor of 2-3 reductionMultipolar external field -> coils close to the plasma
First wall/blanket/shield set a minimum plasma/coil distance (~2m)
A minimum minor radius
Large aspect ratio leads to large size.
Multipolar external field -> coils close to the plasma
First wall/blanket/shield set a minimum plasma/coil distance (~2m)
A minimum minor radius
Large aspect ratio leads to large size.
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Physics Optimization Approach
NCSX scale-up
Coils1) Increase plasma-coil separation2) Simpler coils
High leverage in sizing.
Physics1) Confinement of α particle2) Integrity of equilibrium flux surfaces
Critical to first wall & divertor.
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Optimization of NCSX-Like Configurations: Increasing Plasma-Coil Separation
LI383
A series of coil design with Ac=<R>/∆min ranging 6.8 to 5.7 produced. Large increases in Bmax only for Ac < 6. α energy loss is large ~18% .
A series of coil design with Ac=<R>/∆min ranging 6.8 to 5.7 produced. Large increases in Bmax only for Ac < 6. α energy loss is large ~18% .
Ac=5.9
For <R> = 8.25m: ∆min(c-p)=1.4 m ∆min(c-c)=0.83 m Imax=16.4 MA @6.5T
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A bias is introduced in the magnetic spectrum in favor of B(0,1) and B(1,1)A substantial reduction in α loss (to ~ 3.4%) is achieved.
The external kinks and infinite-n ballooning modes are marginally stable at 4% β with no nearby conducting wall.Rotational transform is similar to NCSX, so the same quality of equilibrium flux surface is expected.
A bias is introduced in the magnetic spectrum in favor of B(0,1) and B(1,1)A substantial reduction in α loss (to ~ 3.4%) is achieved.
The external kinks and infinite-n ballooning modes are marginally stable at 4% β with no nearby conducting wall.Rotational transform is similar to NCSX, so the same quality of equilibrium flux surface is expected.
Optimization of NCSX-Like Configurations: Improving α Confinement & Flux Surface Quality
N3ARE
Fre
qu
ency
*40
96
N3ARELI383
Energy (keV) Energy (keV)
Baseline Configuration
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Physics Optimization Approach
NCSX scale-up
Coils1) Increase plasma-coil separation2) Simpler coils
High leverage in sizing.
Physics1) Confinement of α particle2) Integrity of equilibrium flux surfaces
Critical to first wall & divertor.
New classes of QA configurations
Reduce consideration of MHD stability in light of W7AS and LHD results
MHH21) Develop very low aspect ratio geometry2) Detailed coil design optimization
“Simpler” coils and geometry?
SNS1) Nearly flat rotational transforms 2) Excellent flux surface quality
How good and robust the flux surfaces one can “design”?
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Two New Classes of QA Configurations
II. MHH2Low plasma aspect ratio (Ap ~ 2.5) in 2 field period.Excellent QA, low effective ripple (<0.8%), low α energy loss (≤ 5%) .
II. MHH2Low plasma aspect ratio (Ap ~ 2.5) in 2 field period.Excellent QA, low effective ripple (<0.8%), low α energy loss (≤ 5%) .
III. SNSAp ~ 6.0 in 3 field period. Good QA, low ε-eff (< 0.4%), α loss ≤8% .Low shear rotational transform at high β, avoiding low order resonances.
III. SNSAp ~ 6.0 in 3 field period. Good QA, low ε-eff (< 0.4%), α loss ≤8% .Low shear rotational transform at high β, avoiding low order resonances.
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Minimum Coil-plasma Stand-off Can Be Reduced By Using Tapered-Blanket Zones
Thickness(cm)
| |Thickness
(cm)
Non-uniformBlanket
&Shield
@ ∆min
FullBlanket
& Shield
Replaceable FW/Blkt/BW
∆ ≥ 179 cm
SO
L
Vac
uu
m V
esse
l
FS
Sh
ield
(per
man
ent)
Ga p
3.8
cm F
W
Gap
+ T
h. I
nsu
lato
r
Win
din
g P
ack
Pla
sma
5 >2 ≥232 19.42.228
Coi
l Cas
e &
In
sula
tor
5 c m
Bac
k W
a ll
28
25 cmBreeding
Zone-I
25 cmBreeding Zone-II
0.5 cmSiC Insert
1.5 cm FS/He1.
5 c m
FS
/He
63| | 35
He
& L
iPb
Man
ifo l
ds
Vac
uu
m V
esse
l
Gap
Gap
+ T
h. I
nsu
lato
r
Coi
l Cas
e &
In
sula
tor
Win
din
g P
ack
Pla
sma
2 2 19.42.228
| |∆min = 130.7 cm
Str
ong
Bac
k
28
SO
L
Bac
k W
all
5 14 5
FW
3.8F
S S
hie
ld-I
(re p
lace
a ble
)34
WC
Sh
ield
(per
man
ent)
Str
ong
Bac
k
25
Bla
nk
et
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Resulting power plants have similar size as Advanced Tokamak designs
Trade-off between good stellarator properties (steady-state, no disruption , no feedback stabilization) and complexity of components.
Complex interaction of Physics/Engineering constraints.
Trade-off between good stellarator properties (steady-state, no disruption , no feedback stabilization) and complexity of components.
Complex interaction of Physics/Engineering constraints.
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Resulting power plants have similar size as Advanced Tokamak designs
Major radius can be increased to ease engineering difficulties with a small cost penalty.
Major radius can be increased to ease engineering difficulties with a small cost penalty.
5
5.5
6
6.5
7
7.5
8
8.5
9
4.2 4.4 4.6 4.8 5 5.2
pn,wall,max
(MW/m2)
Bmax
/2 (T)
Baxis
(T)
<R> (m)
0.1 COE (1992 mills/kWhe)
<R>min
(m)
2,1891,7572,633Total Direct Cost (M$)
1,3869001,642Reactor Plant Equip. (M$)
12,6795,22610,96221,430FPC Mass, tonnes
5.0%9.2%5.0%5.0%<β>
8.05.95.75.0<Bo>, T
5.55.27.7514.0<R>, m
ARIES-RSARIES-ATARIES-CSSPPS
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Complex plasma shape and plasma-coil relative position drives many engineering
systems
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First ever 3-D modeling of complex stellarator geometry for nuclear assessment using CAD/MCNP coupling
Detailed and complex 3-D analysis is required for the design
Example: Complex plasma shape leads to a large non-uniformity in the loads (e.g., peak to average neutron wall load of 2).
Detailed and complex 3-D analysis is required for the design
Example: Complex plasma shape leads to a large non-uniformity in the loads (e.g., peak to average neutron wall load of 2).
Poloidal AngleIB IB
Tor
oida
l Ang
le
Distribution of Neutron wall load
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Option 1: Inorganic insulation, assembled with magnet prior to winding and capable to withstand the heat treatment process.
Option 1: Inorganic insulation, assembled with magnet prior to winding and capable to withstand the heat treatment process.
Coil Complexity Impacts the Choice of Superconducting Material
Strains required during winding process is too large.
NbTi-like (at 4K) ⇒ B < ~7-8 T
NbTi-like (at 2K) ⇒ B < 9 T, problem with temperature margin
Nb3Sn ⇒ B < 16 T, Conventional technique does not work because of inorganic insulators
Strains required during winding process is too large.
NbTi-like (at 4K) ⇒ B < ~7-8 T
NbTi-like (at 2K) ⇒ B < 9 T, problem with temperature margin
Nb3Sn ⇒ B < 16 T, Conventional technique does not work because of inorganic insulators
Option 3: HTS (YBCO), Superconductor directly deposited on structure.Option 3: HTS (YBCO), Superconductor directly deposited on structure.
Option 2: conductor with thin cross section to get low strain during winding. (Low conductor current, internal dump).
Option 2: conductor with thin cross section to get low strain during winding. (Low conductor current, internal dump).
SC strands
High RRR Support plateHe coolant
InsulationStructure
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Coil Complexity Dictates Choice of Magnet Support Structure
It appears that a continuous structure is best option for supporting magnetic forces.
Net force balance between field periods (Can be in three pieces)
Absence of disruptions reduces demand on coil structure.
Superconductor coils wound into grooves inside the structure.
It appears that a continuous structure is best option for supporting magnetic forces.
Net force balance between field periods (Can be in three pieces)
Absence of disruptions reduces demand on coil structure.
Superconductor coils wound into grooves inside the structure.
28 cm
Nominally20 cm
Strongback
Inter-coil Structure
Coil dimensions19.4 cm x 74.3 cmFilled with cables
28 cm
Strongback
Inter-coil Structure
Coil dimensions
Filled with cables
Cover plate 2 cm thick
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Port Assembly: Components are replaced Through Ports
Modules removed through three ports using an articulated boom.
Modules removed through three ports using an articulated boom.
Drawbacks:
Coolant manifolds increases plasma-coil distance.
Very complex manifolds and joints
Large number of connect/disconnects
Drawbacks:
Coolant manifolds increases plasma-coil distance.
Very complex manifolds and joints
Large number of connect/disconnects
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Dual coolant with a self-cooled PbLi zone and He-cooled RAFS structure Originally developed for ARIES-ST, further developed by EU (FZK), now is considered as ITER test module
SiC insulator lining PbLi channel for thermal and electrical insulation allows a LiPb outlet temperature higher than RAFS maximum temperature
Self-cooled PbLi with SiC composite structure (a al ARIES-AT)
Higher-risk high-payoff option
Dual coolant with a self-cooled PbLi zone and He-cooled RAFS structure Originally developed for ARIES-ST, further developed by EU (FZK), now is considered as ITER test module
SiC insulator lining PbLi channel for thermal and electrical insulation allows a LiPb outlet temperature higher than RAFS maximum temperature
Self-cooled PbLi with SiC composite structure (a al ARIES-AT)
Higher-risk high-payoff option
Blanket Concepts are Optimized for Stellarator Geometry
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Heat/particle flux on divertor was computed by following field lines outside LCMS.
Because of 3-D nature of magnetic topology, location & shaping of divertor plates require considerable iterative analysis.
Heat/particle flux on divertor was computed by following field lines outside LCMS.
Because of 3-D nature of magnetic topology, location & shaping of divertor plates require considerable iterative analysis.
A highly radiative core is needed for divertor operation
W alloy outer tube
W alloy inner cartridge
W armor
Divertor module is based on W Cap design (FZK) extended to mid-size (~ 10 cm) with a capability of 10 MW/m2
Divertor module is based on W Cap design (FZK) extended to mid-size (~ 10 cm) with a capability of 10 MW/m2
Top and bottom plate location with toroidal coverage from -25° to
25°.
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Summary of the ARIES-CS Study
Goal 1: Can compact stellarator power plants similar in size to advancedtokamak power plants?
Reduce aspect ratio while maintaining “good” stellarator properties.
Include relevant power plants issues (α particle loss, divertor, practical coils).
Identify key areas for R&D (what areas make a big difference)
Results:
Compact stellarator power plants can be similar in size to advanced tokamaks (The best “size” parameter is the mass not the major radius).
α particle loss can be reduced substantially (how low is low enough?)
A large number of QA configurations, more desirable configurations are possible. In particular, mechanism for β limit is not known. Relaxing criteria for linear MHD stability may lead to configurations with a less complex geometry or coils.
Goal 1: Can compact stellarator power plants similar in size to advancedtokamak power plants?
Reduce aspect ratio while maintaining “good” stellarator properties.
Include relevant power plants issues (α particle loss, divertor, practical coils).
Identify key areas for R&D (what areas make a big difference)
Results:
Compact stellarator power plants can be similar in size to advanced tokamaks (The best “size” parameter is the mass not the major radius).
α particle loss can be reduced substantially (how low is low enough?)
A large number of QA configurations, more desirable configurations are possible. In particular, mechanism for β limit is not known. Relaxing criteria for linear MHD stability may lead to configurations with a less complex geometry or coils.
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Summary of the ARIES-CS Study
Goal 2: Understand the impact of complex shape and geometry
A. Configuration, assembly, and maintenance drives the design
A high degree of integration is required
Component replacement through ports appears to be the only viable method.
Leads to modules that can be fitted through the port and supported by articulated booms.
Large coolant manifold (increase radial build), large number of connects and disconnects, complicated component design for assembly disassembly.
B. Complexity-driven constraints (e.g., superconducting magnets)
Options were identified. (e.g., base case for superconducting magnets requires development of inorganic insulators.)
Goal 2: Understand the impact of complex shape and geometry
A. Configuration, assembly, and maintenance drives the design
A high degree of integration is required
Component replacement through ports appears to be the only viable method.
Leads to modules that can be fitted through the port and supported by articulated booms.
Large coolant manifold (increase radial build), large number of connects and disconnects, complicated component design for assembly disassembly.
B. Complexity-driven constraints (e.g., superconducting magnets)
Options were identified. (e.g., base case for superconducting magnets requires development of inorganic insulators.)
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Summary of the ARIES-CS Study
Goal 2: Understand the impact of complex shape and geometry
C. Complex 3-D analysis
3-D analysis is required for almost all cases (not performed in each case).
CAD/MCNP interface for 3-D neutronics, 3-D solid model for magnet support, …
D. Manufacturability (feasibility and Cost)
Feasibility of manufacturing of component has been included in the design as much as possible.
In a large number of cases, manufacturing is challenging and/or very expensive.
Goal 2: Understand the impact of complex shape and geometry
C. Complex 3-D analysis
3-D analysis is required for almost all cases (not performed in each case).
CAD/MCNP interface for 3-D neutronics, 3-D solid model for magnet support, …
D. Manufacturability (feasibility and Cost)
Feasibility of manufacturing of component has been included in the design as much as possible.
In a large number of cases, manufacturing is challenging and/or very expensive.
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From ITER to an attractive final product
Plans for the Next ARIES Study
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ITERIntegration of fusion plasma with fusion
technologies
A 1st of the kind Power Plant!
“Fusion Power: Research and Development Requirements.”Division of Controlled Thermonuclear Research (AEC).
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World-wide Development Scenarios use similar names for devices with different missions!
Com
mer
cial
ITE
R +
IF
MIF
+ B
ase
Demo Proto
EU or Japan
CTF Demo
US
Demo-ProtoDemo (R&D)
EU or Japan (Fast Track)
∗ Combine Demo (R&D) and Proto in one device
Proto Demo
US (1973 AEC)
An R&D Device A Power Plant
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In the ITER area,we need to develop
a 5,000 ft view
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A holistic or integrated optimization approachshould drive the development path.
Traditional Approach: Ask each scientific area (i.e., plasma, blanket, …)
What are the remaining major R&D areas?
Which of the remaining major R&D areas can be explored in existing devices or simulation facilities (e.g., fission reactors)? What other major facilities are needed?
Traditional Approach: Ask each scientific area (i.e., plasma, blanket, …)
What are the remaining major R&D areas?
Which of the remaining major R&D areas can be explored in existing devices or simulation facilities (e.g., fission reactors)? What other major facilities are needed?
Holistic Approach: Fusion energy development should be guided by therequirements for an attractive fusion energy source
What are the remaining major R&D areas?
What it the impact of this R&D on the attractiveness of the final product.
Which of the remaining major R&D areas can be explored in existing devices or simulation facilities (i.e., fission reactors)? What other major facilities are needed?
Should we attempt to replicate power plant conditions in a scaled device or Optimize facility performance relative to scaled objectives
Holistic Approach: Fusion energy development should be guided by therequirements for an attractive fusion energy source
What are the remaining major R&D areas?
What it the impact of this R&D on the attractiveness of the final product.
Which of the remaining major R&D areas can be explored in existing devices or simulation facilities (i.e., fission reactors)? What other major facilities are needed?
Should we attempt to replicate power plant conditions in a scaled device or Optimize facility performance relative to scaled objectives
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Elements of the Case for Fusion Power Were Developed through Interaction with Representatives of U.S. Electric Utilities and Energy Industry
Have an economically competitive life-cycle cost of electricity
Gain Public acceptance by having excellent safety and environmental characteristics
No disturbance of public’s day-to-day activities
No local or global atmospheric impact
No need for evacuation plan
No high-level waste
Ease of licensing
Reliable, available, and stable as an electrical power sourceHave operational reliability and high availability
Closed, on-site fuel cycle
High fuel availability
Capable of partial load operation
Available in a range of unit sizes
Fu
sion
ph
ysic
s &
tec
hn
olog
y
Low-activation material
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Existing facilities fail to address essential features of a fusion energy source
Metricwaste 3 need to deal with it, but wrong materials, little fluence
reliability 3 some machine operation, little fluence
maintenance 5 unprototypic construction, modules replaced
fuel 3 tritium handling, but no breeding, no fuel cycle
safety 6 hazards are lower, operations different
partial power 4 experience with operating modes
thermal efficiency 0 no power production, low temperature, wrong materials
power density 5 low average power density, local regions of HHF
cost 5 1st of a kind reactor costs, cost reduction needed
ITER
Metricwaste 0 little relevance
reliability 1 some machine operation, no fluence
maintenance 1 experience moving tokamak equipment
fuel 1 Some tritium handling, no breeding, no fuel cycle
safety 2 hazards much lower, operations much different
partial power 2 experience with operating modes
thermal efficiency 0 no power conversion
power density 1 low power handling required, some divertor heating
cost 1 not relevant to a power plant
D3/JET
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ITER is a major step forward but there is a long road ahead.
Present Experiments
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A holistic optimization approachshould drive the development path.
Traditional Approach: Ask each scientific area (i.e., plasma, blanket, …)
What are the remaining major R&D areas?
Which of the remaining major R&D areas can be explored in existing devices or simulation facilities (e.g., fission reactors)? What other major facilities are needed?
Traditional Approach: Ask each scientific area (i.e., plasma, blanket, …)
What are the remaining major R&D areas?
Which of the remaining major R&D areas can be explored in existing devices or simulation facilities (e.g., fission reactors)? What other major facilities are needed?
Holistic Approach: Fusion energy development should be guided by therequirements for an attractive fusion energy source
What are the remaining major R&D areas?
What it the impact of this R&D on the attractiveness of the final product.
Which of the remaining major R&D areas can be explored in existing devices or simulation facilities (i.e., fission reactors)? What other major facilities are needed?
Should we attempt to replicate power plant conditions in a scaled device or Optimize facility performance relative to scaled objectives
Holistic Approach: Fusion energy development should be guided by therequirements for an attractive fusion energy source
What are the remaining major R&D areas?
What it the impact of this R&D on the attractiveness of the final product.
Which of the remaining major R&D areas can be explored in existing devices or simulation facilities (i.e., fission reactors)? What other major facilities are needed?
Should we attempt to replicate power plant conditions in a scaled device or Optimize facility performance relative to scaled objectives
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ARIES studies emphasize holistic R&D needs and
their design implications
Plasma
Blankets
Divertors
Magnets
Vacuum vessel
Traditional approach
Power control
Power and particle management
Fuel management
Maintenance
Safety
Waste
Cost
Concurrent engineering/physics
This approach has many benefits (see below) This approach has many benefits (see below)
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Examples of holistic issues for Fusion Power
• Power & Particle management: Demonstrate extraction of power core high-grade heat, divertor power and particle handling, nuclear performance of ancillary equipment.
• Fuel management: Demonstrate “birth to death” tritium management in a closed loop with self-sufficient breeding and full accountability of tritium inventory.
• Safety: Demonstrate public and worker safety of the integral facility, capturing system to system interactions.
• Plant operations: Establish the operability of a fusion energy facility, including plasma control, reliability of components, inspectability and maintainability of a power plant relevant tokamak.
• Power & Particle management: Demonstrate extraction of power core high-grade heat, divertor power and particle handling, nuclear performance of ancillary equipment.
• Fuel management: Demonstrate “birth to death” tritium management in a closed loop with self-sufficient breeding and full accountability of tritium inventory.
• Safety: Demonstrate public and worker safety of the integral facility, capturing system to system interactions.
• Plant operations: Establish the operability of a fusion energy facility, including plasma control, reliability of components, inspectability and maintainability of a power plant relevant tokamak.
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Power & particle management: Demonstrate extraction of power core high-grade heat, divertor power and particle handling, nuclear performance of ancillary equipment.
Power & particle management: Demonstrate extraction of power core high-grade heat, divertor power and particle handling, nuclear performance of ancillary equipment.
Fission:Fission:
Divertor
First wall
Prad
Pcond
In-vessel
PFC’s
Pα
Pinjected
Pfusion
corepower
rf antennas, magnets, diagnostics, etc.
edgepower
Prad
Pcond
FusionFusion: Pneutron
Blanket
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A holistic approach to Power and Particle Management
Does not allow problem cannot be solved by transferring to another system:A 100% radiating plasma transfers the problem from divertor to the first wall.
Allows Prioritization of R&D:Systems code can be used to find power plant cost (or any other metric) as a function of divertor power handling. This leads to a “benefit” metric that can be compared to other R&D areas, for example increasing plasma β. We can then answer: should we focus on power flow or improving plasma β.Most technologies for current experiments & ITER are NOT transferable to a power plant. Is this an optimum approach?
Solution may come from other areas:Low recirculating powerA higher blanket thermal efficiency reducing input fusion power
This particular area may have a profound impact on next-step facilities.
Does not allow problem cannot be solved by transferring to another system:A 100% radiating plasma transfers the problem from divertor to the first wall.
Allows Prioritization of R&D:Systems code can be used to find power plant cost (or any other metric) as a function of divertor power handling. This leads to a “benefit” metric that can be compared to other R&D areas, for example increasing plasma β. We can then answer: should we focus on power flow or improving plasma β.Most technologies for current experiments & ITER are NOT transferable to a power plant. Is this an optimum approach?
Solution may come from other areas:Low recirculating powerA higher blanket thermal efficiency reducing input fusion power
This particular area may have a profound impact on next-step facilities.
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Fuel management: Demonstrate “birth to death” tritium management in a closed loop with self-sufficient breeding and full accountability of tritium inventory.
Fuel management: Demonstrate “birth to death” tritium management in a closed loop with self-sufficient breeding and full accountability of tritium inventory.
inventory
pumps
breeder
coolant
breeder processing
coolant processing
vacuum processing
fueling
D+T
D+T+α
n
TFuelprocessing
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Fuel Management divides naturally along physical boundaries
Can be done in a fission facility.
Demonstrate in-situ control of breeding rate (too much breeding is bad).
Demonstrate T can be extracted from breeder in a timely manner (minimum inventory).
Can be done in a fission facility.
Demonstrate in-situ control of breeding rate (too much breeding is bad).
Demonstrate T can be extracted from breeder in a timely manner (minimum inventory).
ITER provides most of the required data.
Issues include minimizing T inventory and T accountability
ITER provides most of the required data.
Issues include minimizing T inventory and T accountability
(the rest)
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For the next three years, ARIES program will examine fusion development scenarios
What are the remaining major R&D areas? What are the data base needed to field a commercial power plant (e.g., licensing, operation, reliability, etc.)?
What it the impact of this R&D on the attractiveness of the final product.
Which of the remaining major R&D areas can be explored in existing devices or simulation facilities (i.e., fission reactors)? What other major facilities are needed?
Break the systems along physics boundaries (instead of scientific disciplines).
What other major facilities are needed (CTF, Fast track, etc.) What are the possible embodiments for CTF and what are the theircost/performance attributes.
We should consider the needs of next-step facilities in the R&D in current facilities as well as initiating R&D needed to ensure maximum utilization of those facilities.
What are the remaining major R&D areas? What are the data base needed to field a commercial power plant (e.g., licensing, operation, reliability, etc.)?
What it the impact of this R&D on the attractiveness of the final product.
Which of the remaining major R&D areas can be explored in existing devices or simulation facilities (i.e., fission reactors)? What other major facilities are needed?
Break the systems along physics boundaries (instead of scientific disciplines).
What other major facilities are needed (CTF, Fast track, etc.) What are the possible embodiments for CTF and what are the theircost/performance attributes.
We should consider the needs of next-step facilities in the R&D in current facilities as well as initiating R&D needed to ensure maximum utilization of those facilities.
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Any Questions?