ARIES: Fusion Power Core and Power Cycle Engineering

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(No Transcript)
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An Overview of Fusion Technology and Advanced
Design Activities at UCSD

• René Raffray
• University of California, San Diego
• La Jolla, CA, USA
• Presented at the IKET-Kolloquium
• Forschungszentrum Karlsruhe
• July 4, 2006

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Outline

• Center for Energy Research at UCSD
• Summary of Fusion-Related Activities
• - PISCES
• - Laser/material laboratory
• - HAPL
• - ARIES (focus)

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Center for Energy Research

• The mission of the Center for Energy Research
  (CER) at the University of California, San Diego
  (UCSD) is to foster research and educational
  activities devoted to critical energy needs. The
  CER provides an academic research unit for
  interdisciplinary interactions among UCSD
  faculty, research staff and students aimed at
  promoting and coordinating energy research and
  education. The CER also provides a vehicle for
  developing other dimensions of energy research,
  including energy policy, economics and ecology.

Web site http//cer.ucsd.edu
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The Center for Energy Research (CER) is
Affiliated with UCSD Jacobs School of Engineering
Enrolled Application Freshmen 771
6,627 New graduate student 319 4,629
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Two Major Research Areas under CER Fusion
Energy Applied Plasma Physics Combustion

• CER Director Prof. Forman Williams
• Deputy Director Prof. Farrokh Najmabadi

• Fusion Researchers at JSOE
• Faculty and Senior Researchers 10
• Junior Researchers Post-docs 17
• Students 23
• Engineering Staff 6

Combustion Researchers at JSOE Faculty and
Senior Researchers 5 Junior Researchers
Post-docs 1 Students 12
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Fusion Research at JSOE Has Grown by 60 in the
Last Five Years
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Fusion Energy Applied Plasma Physics Research

• ARIES Program
• - Performing advanced integrated design studies
  of long-term fusion energy concepts to identify
  key RD directions and to provide visions for
  the fusion program
• PISCES Program
• - Exploring plasma boundary science and
  plasma-material interactions as they relate to
  magnetic fusion confinement devices and related
  applications.
• Magnetic Confinement of Plasma
• - Collaboration with General Atomics (DIII-D),
  Princeton Plasma Physics Laboratory (NSTX and
  NCSX) and KFA Juelich (Textor).
• HAPL (High Average Power Laser) Program
• - Developing the science and technology for
  Laser Inertial Fusion Energy UCSDs
  contribution involves the reaction chamber,
  final optics, and target survival and tracking.
• Laser-Matter Interactions
• - Includes thermomechanical response of surfaces
  exposed to short-pulse high-energy laser
  irradiation, laser ablation plume dynamics, laser
  plasma light and particle emissions for
  next-generation, extreme UV lithography.
• High Energy Density Physics Research
• - Includes Fast Ignition for Inertial
  Confinement Fusion, wire array Z-pinches,
  compact x-ray and neutron sources for
  different applications.

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• PISCES Program

Web site http//cerfe.ucsd.edu/
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PISCES Program

• Objective
• - Perform Basic Plasma-Materials Interaction
  Boundary Plasma Research Needed for ITER PFC
  Design Validation and Performance Predictions
• Approach- Perform Controlled Be-C, Be-W, Be-C-W
  PMI Experiments
• - Steady-state Transient Thermal Loads
• - Make Related Edge/SOL Plasma Transport Studies
• - Develop and Validate Edge Plasma/PFC Models
  Via Collaborations
• Focus of Several Major Collaborations
• - US-EU EFDA Collaboration on Mixed Materials
• - UCSD Hosts On-Site Long-Term EU Visitors
• - EU Contributes to PISCES-B Facility
  Capabilities
• - US-Japan Collaborative Exchanges
• - Model Validation Studies
• - Argonne WBC
• - KFA-Juelich ERO
• - UCSD Theory - UEDGE

Slide courtesy of R. Doerner and G. Tynan
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Facilities at UCSD in Support of PISCES Program
PISCES-A PMI Supporting Research
Be-Compatible PISCES-B Facility
CSDX Supporting Research

• Surface Science Diagnostics
• In-situ XPS, Auger, SIMS
• SEM EDX
• Ex-situ SIMS, XPS, TDS

Slide courtesy of R. Doerner and G. Tynan
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A Range of Experiments are Conducted as Part of
PISCES e.g. Controlled Mixed Matl PMI Expts
in Be Seeded PISCES-B Plasmas
Slide courtesy of R. Doerner and G. Tynan
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Example Results from PISCES A small beryllium
impurity concentration in the plasma drastically
suppresses carbon erosion
-50 V bias, 200ºC, Te 8 eV, ne 3 x1012 cm-3
Chemical erosion
Physical sputtering
Slide courtesy of R. Doerner and G. Tynan
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• Laser-Matter Interactions

Web site http//aries.ucsd.edu/LMI/
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UCSD Research on Laser Plasma and Laser-Matter
Interactions Spans a Wide Range of Physics Regimes

• Laser plasmas
• EUV lithography
• HED studies (XUV, x-rays)

Relativistic laser plasma (Beg)
Thermal, mechanical and phase change behavior
Laser ablation plume dynamics, LIBS,
micromachining
IFE optics
Slide courtesy of M. Tillack
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Research Activities in the UCSD Laser Plasma and
Laser-Matter Interactions Laboratory

• IFE optics development
• Laser plasma physics
• EUV lithography
• HED studies (EUV/XUV)
• Atmospheric plasmas LIBS, micromachining

• Laser ablation plume dynamics in gases and
  magnetic fields
• Thermal, mechanical and phase change behavior of
  surfaces exposed to short-pulse laser irradiation

Slide courtesy of M. Tillack
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UCSD is Studying Laser-Produced Plasma Light
Sources for Next-Generation Semiconductor
Lithography

• Conventional lithography uses 193-248 nm lasers,
  transmissive mask and optics
• EUV lithography uses discharge or laser-produced
  plasmas emitting 13.5 nm light

Intels EUV MicroExposure Tool (MET)
mask
EUV source and condenser lens
laser
High energy ion damage and contamination are
critical issues for a commercial system Ion
energies in the keV range are most damaging
wafer
EUV lithography tool light path
Slide courtesy of M. Tillack
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UCSD Developed a Double-Pulse Technique to
Virtually Eliminate Debris Ion Damage to
Condenser Optics

• A 1-mJ pre-pulse forms a target plasma
• A 1-J main pulse generates light from pre-pulse
  plasma

1 mm

• Expansion dynamics from pre-plasma are entirely
  different than solid targets
• A factor of 30 reduction in energy was observed
  using the optimum delay
• Time-resolved interferometry verifies the plasma
  density profile evolution
• No loss of conversion efficiency!

Slide courtesy of M. Tillack
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• Contributions to
• High Average Power Laser (HAPL) Program (led by
  NRL)

Web site http//aries.ucsd.edu/HAPL/
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The HAPL Program Aims at Developing a New Energy
Source IFE Based on Lasers, Direct Drive Targets
and Solid Wall Chambers
(Major chamber interfacing systems/components)
System (including power cycle)
Blanket (make the most of MFE design and RD
info)
Dry wall chamber (armor must accommodate
ionphoton threat and provide required lifetime)

• Modular, separable parts lowers cost of
  development and improvements
• Conceptually simple spherical targets, passive
  chambers
• Currently toward the end of Phase I effort,
  which is focused on the basic science and
  technology of different components

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Chamber Physics Needs to be Understood
Ion and Photon Threat Spectra and Chamber
Conditions Prior to Each Shot Must Be Well
Characterized (UW) Attenuation of ion and
photon spectra seen by wall by using a
protective chamber gas
Chamber Gas Ion and Photon Attenuation for
Chamber with R6.5 m (from BUCKY)
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Must Ensure Successful Injection Tracking and
Survival of Target (GA/UCSD/LLE/LANL)
Target injection tracking - Developing
systems for tracking direct-drive inertial
fusion targets and steering multiple laser
beams onto the target - Injection system
includes mechanical, gas gun or
electromagnetic (depending on
requirements such as injection velocity) -
5 mm target injection accuracy, 2 mm
target/laser accuracy
GA target injection facility
Target survival - Must accommodate heat
transfer from background gas and wall
radiation and meet target integrity
requirements based on target implosion
physics
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Long Term Survival Optical Fidelity Required of
Final Optics
We are developing damage-resistant final optics
based on grazing-incidence metal mirrors and
testing them (effort coordinated by UCSD LLNL)
Mirror requirements - 5 J/cm2 - 2 yrs, 3x108
shots - 1 spatial non-uniformity - 20 mm
aiming - 1 beam balance
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Credible Armor/First Wall Configuration to
Accommodate the Threat Spectra and Provide the
Required Lifetime
Separation of function armor for threat
accommodation FW for structural
function - Front runner configuration thin W
armor ( 1 mm) on FS
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Assessment of Potential Causes of Armor Failure
and Consideration Advanced Engineered Material
for Potentially Superior Performance
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The Blanket and Beyond Strategy for Blanket
Development and Integrated System Studies Study
(UCSD/UW/LLNL)
Blanket strategy aims at making the most of MFE
design and RD info in developing an attractive
IFE blanket concept - Self-cooled Li blanket
developed for large chamber configuration - Othe
r concepts considered for magnetic intervention
case

• System studies to help develop an integrated
  HAPL power plant conceptual design
• - Develop tools and constraints to evolve
  parameter space that accommodates material
  and design constraints.
• - Perform trade-off studies

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Example Integration Analysis of Chamber
Armor/FW/Blanket

• Start with spectra from NRL 154 MJ direct-drive
  target
• - Photon
• Fast ions
• Debris ions

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Calculate Energy Deposition in Armor Based on
Spectra and Time of Flight Effect
Use results of photon and ion energy deposition
analysis as input in RACLETTE-IFE code to
calculate cyclic armor thermal response
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Example Results Comparing W Temperature Histories
for Armor Thicknesses of 0.05 mm and 0.5 mm,
respectively
dW0.05mm
dW0.5mm
154 MJ yield No gas Rep Rate 10 Rchamber 6.5
m dFS 2.5mm Tcoolant 500C
Not much difference in maximum W temperature
and in number of cycles to ramp up to the maximum
temperature level
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Example Results Comparing FS Temperature
Histories for W Armor Thicknesses of 0.05 mm and
0.5 mm, Respectively
dW0.05mm
dW0.5mm
Substantial differences in max. TFS and cyclic
DTFS at FS/W interface depending on dW Can
adjust Tmax by varying Tcoolant and
hcoolant Design for separate function and
operating regime - armor function under cyclic
temperature conditions - structural material,
coolant and blanket operation designed for
quasi steady-state
154 MJ yield No gas Rep Rate 10 Rchamber 6.5
m dFS 2.5mm Tcoolant 500C
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Example Results of Armor Parametric Analysis
Illustrating Combination of Xe Chamber Pressure,
Yield and Chamber Size to Maintain W Armor Within
2400C for a Fusion Power of 1800 MW




Required P
as a Function of Yield to Maintain T
lt2400?C
W temperature limit of 2400C assumed for
illustration purposes Actual limit based
results of ongoing experimental and modeling
armor RD effort
W,max
Xe
for 1800 MW Fusion Power and Different R
chamber
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40
R
(m)
1 mm W
chamber
3.5 mm FS
5.7
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T
572?C
50
coolant
6.5
2
h67 kW/m
-K
30
25
7
Xe Pressure (_at_ST) (mtorr)
Repetition Rate
20
15
8
10
10
5
0
0
50
100
150
200
250
300
350
400
450
Yield (MJ)
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• ARIES Program

Web site http//aries.ucsd.edu/ARIES/
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ARIES Program
National multi-institution program led
by UCSD (Program leader Prof.
F.Najmabadi) - Perform advanced integrated
design studies of long-term fusion
energy concepts to identify key RD
directions and to provide visions for the
fusion program
Currently completing ARIES-CS, a study of a
Compact Stellarator option as a
power plant to help - Advance physics and
technology base of CS concept and address concept
attractiveness issues in the context of
power plant studies - Identify optimum CS
configuration for power plant
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The ARIES Team is Completing the Last Phase of
the ARIES-CS Study

• Phase I Development of Plasma/coil Configuration
  Optimization Tool
• Develop physics requirements and modules (power
  balance, stability, a confinement, divertor,
  etc.)
• Develop engineering requirements and constraints
  through scoping studies.
• Explore attractive coil topologies.

• Phase II Exploration of Configuration Design
  Space
• Physics b, aspect ratio, number of periods,
  rotational transform, shear, etc.
• Engineering configuration optimization through
  more detailed studies of selected concepts
• Trade-off studies (systems code)
• Choose one configuration for detailed design.

Phase III Detailed system design and optimization
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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. - Engi
neering Reduce the required minimum coil-plasma
distance.
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We Considered Different Configurations Including
NCSX-Like 3-Field Period and MHH2-Field Period
Configurations
NCSX-Like 3-Field Period
Example Parameters for NCSX-Like 3-Field Period
from Latest System Optimization Runs
MHH2 2-Field Period
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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.
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Blanket Concepts
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Five Blanket Concepts Were Evaluated During Phase
I
3 4) Dual-coolant blankets with He-cooled FS
structure and self-cooled Li or Pb-17Li breeder
(ARIES-ST type)
2) Self-cooled Pb-17Li with SiCf/SiC (ARIES-AT
type)
1) Self-cooled flibe with ODS FS
5) He-cooled ceramic breeder with FS structure
Flibe/FS/Be LiPb/SiC CB/FS/Be LiPb/FS Li/FS ?mi
n 1.11 1.14 1.29 1.18 1.16 TBR 1.1 1.1 1.1 1.1
1.1 Energy Multiplication (Mn) 1.2 1.1 1.3 1.15
1.13 Thermal Efficiency (?th) 42-45 55-60 42
42-45 42-45 FW Lifetime (FPY) 6.5 6 4.4 5
7
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Selection of Blanket Concepts for Detailed Study

• Dual Coolant concept with a self-cooled Pb-17Li
  zone and He-cooled RAFS structure.
• He cooling needed for ARIES-CS divertor
• Additional use of this coolant for the
  FW/structure of blankets facilitates
  pre-heating of blankets, serves as guard
  heating, and provides independent and redundant
  afterheat removal.
• Generally good combination of design
  simplicity and performance.
• Build on previous effort, further evolve and
  optimize for ARIES-CS configuration
• - Originally developed for ARIES-ST
• - Further developed by EU (FZK)
• - Now also considered as US ITER test module

• Self-cooled Pb-17Li blanket with SiCf/SiC
  composite as structural material.
• Desire to maintain a higher pay-off, higher
  risk option as alternate to assess the potential
  of a CS with an advanced blanket

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Dual Coolant Blanket Module Redesigned for
Simpler More Effective Coolant Routing
(applicable to both port and field-period based
maintenance schemes)
10 MPa He to cool FW toroidally and box Slow
flowing (lt10 cm/s) Pb-17Li in inner channels
RAFS used (Tmaxlt550C)
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Blanket Optimized Shield to Minimize
Coil-Plasma Stand-off (machine size) while
Providing Required Breeding (TBR gt 1.1) and
Shielding Performance (coil protection)
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Pb-17Li/He DC Blanket Coupled to a Brayton Cycle
Through a HX
Power core He flown through HX to transfer
power to the cycle He with DTHX 30C
- Minimum He temperature in cycle (heat
sink) 35C - hTurbine 0.93 -
hCompressor 0.89 - eRecuperator
0.95 - Total compression ratio lt 2.87
Example Brayton cycle with 3- stage compression
2 inter-coolers and a single stage expansion
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Coolant Routing Through HX Coupling Blanket and
Divertor to Brayton Cycle
Div He Tout Blkt Pb-17Li Tout Min. ?THX
30C PFriction ?pump x Ppump
Example Power Parameters
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Maintenance Scheme
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Port-Based Maintenance Chosen as Reference Scheme
(with Field-Period Maintenance as Back-Up)

• Vacuum vessel is internal to the coils.
• - for blanket maintenance, no disassembling
  and re-welding of VV required and modular
  coils kept at cryogenic temperatures.
• One dedicated port per field period
• - 4 m high by 1.8 m wide
• - Possibility of using smaller ECH port (1
  m2) per field period for inserting remote
  maintenance tools and fixtures.
• - Modular design of blanket (2m x 2m x
  0.63 m)and divertor plates ( 2m x 1.5m x
  0.2 m) compatible with maintenance scheme.
• Closing plug used in access port.
• Articulated boom utilized to remove and
  replace blanket modules (5000 kg).

Cross section of 3 field-period configuration at
0 showing port location for one field period.
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A Key Aim of the Design is to Minimize Thermal
Stresses
Hot core (including shield and manifold)
(450C) as part of strong skeleton ring
(continuous poloidally, divided toroidally in
sectors) separated from cooler vacuum vessel
(200C) to minimize thermal stresses.
Concentric coolant access pipes for both He
and Pb-17Li, with return He in annulus (at
450C) and inlet Pb-17Li in annulus (at 450C)
to maintain near uniform temperature in skeleton
ring.

Each skeleton ring sector rests on sliding
bearings at the bottom of the VV and can freely
expand relative to the VV. Blanket modules are
mechanically attached to this ring and can float
with it relatively to the VV. Bellows are used
between VV and the coolant access pipes at the
penetrations. These bellows provide a seal
between the VV and cryostat atmospheres, and only
see minimal pressure difference. Temperature
variations in blanket module minimized by cooling
the steel structure with He (with ?Tlt100C).
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Blanket Module Replacement for Port-Based
Maintenance Assumes Prior Removal of Adjacent
Module and Access from Plasma Side
Example of Pipe Cutting/Rewelding For He Supply
to Blanket Modules Following Removal of Port
Modules (1A and 1B)
Pipe cutting/rewelding from outside preferred.
Use of equipment similar to what is already
commercially-available. Shield pieces first
removed to access coolant piping. First cut
then performed and shielding ring (protecting
rewelding area from neutron streaming) removed
from inside piping Final coolant piping cut
performed at the back of the shield where He
production is small enough to allow re-welding (lt
1 appm He).
3
3
2
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Structural Design and Analysis of Coils
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Desirable Plasma Configuration should be Produced
by Practical Coils with Low Complexity
Complex 3-D geometry introduces severe
engineering constraints - Distance between
plasma and coil - Maximum coil bend radius
- Coil support - Assembly and
maintenance Superconducting material Nb3Sn ?
B lt 16 T wind react heat treatment to
relieve strains - Need to maintain structural
integrity during heat treatment (700o C for
100s hours) - Need inorganic insulator
Coil structure - JK2LB (Japanese austenitic
steel) preferred - Much less contraction than
316 at cryogenic temp. - Relieve stress
corrosion concern under high temp., stress
and presence of O2 (Incoloy 908) - Potentially
lower cost - YS/UTS _at_4K 1420/1690 MPa - More
fatigue and weld characterization data needed
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Coil Support Design Includes Winding of All Coils
of One Field-Period on a Supporting Tubular
Structure
Winding internal to structure. Entire coil
system enclosed in a common cryostat. Coil
structure designed to accommodate the forces
on the coil
Reacted by connecting coil structure together
(hoop stress) Reacted inside the field-period
of the supporting tube. Transferred to
foundation by 3 legs per field-period. Legs are
long enough to keep the heat ingress into the
cold system within a tolerable limit.

• Large centering forces pulling each coil
  towards the center of the torus.
• Out-of plane forces acting between
  neighboring coils inside a field period.
• Weight of the cold coil system.
• Absence of disruptions reduces demand on
  coil structure.

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Detailed EM and Stress Analysis Performed with
ANSYS
Shell model used for trade-off studies A
case with 3-D solid model done for comparison to
help better understand accuracy of shell
model A case with penetration will be done to
characterize required rib structure.

• As a first-order estimate, structure
  thickness scaled to stress deflection
  results to reduce required material and
  cost
• - Avg. thickness inter-coil structure 20 cm
• - Avg. thickness of coil strong-back 28 cm

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Divertor Study
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Conceptual Divertor Design Study for ARIES-CS
Evolve design to accommodate a max. q of at
least 10 MW/m2 (in anticipation of better
estimates from physics study) - Productive
collaboration with FZK - Absence of disruptions
reduces demand on armor (lifetime based on
sputtering) Previous He-cooled divertor
configurations include - W plate design (1 m)
- More recently, finger configuration with W
caps with aim of minimizing use of W as
structural material and of accommodating
higher q with smaller units (1-2 cm)
(FZK) Build on the W cap design and explore
possibility of a new mid-size configuration with
good q accommodation potential, reasonably
simple (and credible) manufacturing
and assembly procedures, and which could be well
integrated in the CS reactor design.
- "T-tube" configuration (10 cm)
- Cooling with discrete or continuous
jets - Effort underway at PPI to develop
fabrication method
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T-Tube Configuration Looks Promising as Divertor
Concept for ARIES-CS
Encouraging analysis results from ANSYS
(thermomechanics) and FLUENT (CFD) for q 10
MW/m2 - W alloy temperature within
600- 1300C (assumed ductility and
recrystallization limits) - Maximum thermal
stress 370 MPa Initial results from
experiments at Georgia Tech. seem to confirm
thermo-fluid modeling analysis.
sth,max 370 MPa
Good heat transfer from jet flow
Example Case Jet slot width 0.4
mm Jet-wall-spacing 1.2-1.6 mm Specific
mass flow 2.12 g/cm2 Mass flow per tube 48
g P 10 MPa, ?P 0.1 MPa ?T 90 K for q
10 MW/m2 THe 605 - 695C
Tmax 1240C
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Divertor Manifolding and Integration in Core

• T-tubes assembled in a manifold unit
• Typical target plate (1.5 m x 2 m) consists of
  a number of manifold units
• Target plate supported at the back of VV to
  avoid effect of hot core thermal expansion
  relative to VV
• Concentric tube used to route coolant and to
  provide support
• Possibility of in-situ alignment of divertor
  plate if needed

Details of T-tube manifolding to keep FS manifold
structure within its temperature limit
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Alpha Loss
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Alpha Particle Loss is a Concern
Example Spectrum of Lost Alpha Particles
Significant alpha loss in CS (5)
represents a not only loss of heating
power in the core, but adds to the heat load
on PFCs. Depending on the magnetic topology,
a fraction of these particles are promptly
lost from the plasma and hit the PFCs at
energies lt3.5 MeV. Thus, not only must the
PFC surface accommodate the heat load of the
alpha particle flux but it must also
accommodate these high-energy alpha
fluxes and provide the required lifetime.
Footprints of escaping ? on LCMS for N3ARE
Heat load and armor erosion maybe localized and
high
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Accommodating Alpha Particle Heat Flux
High heat flux could be accommodated by
designing special divertor-like modules (allowing
for q up to 10 MW/m2). e.g. for alpha
loss of 5-10 - Pfusion 2350 MW - Max.
neutron wall load 5 MW/m2 - FW Surface Area
572 m2 - Assumed alpha module coverage
0.05 - Ave. q on alpha modules 0.82-1.64
MW/m2 - Max q constrained to lt10
MW/m2 - Alpha q peaking factor lt 12-24 If
the alpha particles end up on the divertor, the
combined load on the divertor would have to be
within the 10 MW/m2 limit. Impact of alpha
particle flux on armor lifetime (erosion) is more
of a concern.
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Inventory of He in W Based on Example ?-Particle
Implantation Case
Simple effective diffusion analysis for
different characteristic diffusion dimensions for
an activation energy of 4.8 eV (vacancy
dissociation) Not clear what is the max. He
conc. limit in W to avoid exfolation (perhaps
0.15 at.) High W temperature needed in this
case Shorter diffusion dimensions help, perhaps
a nanostructured porous W (PPI) e.g. 50-100 nm
at 1800C or higher
An interesting question is whether at a high W
operating temperature (gt1400C), some
annealing of the defects might help the
tritium release. This is a key issue for a CS
which needs to be further studied to make
sure that a credible solution exists both in
terms of the alpha physics, the selection of
armor material, and better characterization of
the He behavior under prototypic conditions.
61
PPIs Progress in Manufacturing Porous W with
Nano-Microstructure

• Plasma technology can produce tungsten nanometer
  powders.
• - When tungsten precursors are injected into the
  plasma flame, the materials are heated, melted,
  vaporized and the chemical reaction is induced
  in the vapor phase. The vapor phase is quenched
  rapidly to solid phase yielding the ultra pure
  nanosized W powder
• - Nano tungsten powders have been successfully
  produced by plasma technique and the product
  is ultra pure with an average particle size of
  20-30nm. Production rates of gt 10 kg/hr are
  feasible.
• Process applicable to molybdenum, rhenium,
  tungsten carbide, molybdenum carbide and other
  materials.
• The next step is to utilize such a powder in the
  Vacuum Plasma Spray process to manufacture porous
  W (10-20 porosity) with characteristic
  microstructure dimension of 50 nm .

TEM images of tungsten nanopowder, p/n S05-15.
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Extra Slides
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ARIES-CS Summary (I)
New configurations have been developed, others
refined and improved, all aimed at low plasma
aspect ratios (A 6), hence compact size - Both
2 and 3 field periods possible. - Progress has
been made to reduce loss of a particles to 5
this may be still higher than desirable. - Resulti
ng power plants have similar size as Advanced
Tokamak designs. Modular coils were designed to
examine the geometric complexity and the
constraints of the maximum allowable field,
desirable coil-plasma spacing and coil-coil
spacing, and other coil parameters. Assembly
and maintenance is a key issue in configuration
optimization.
64
ARIES-CS Summary (II)
Engineering effort has yielded some interesting
and new evolutions in power core
design - Blanket/shield optimization to minimize
plasma to coil minimum distance and reduce
machine size - Separation of hot core components
from colder vacuum vessel (allowing for
differential expansion through slide
bearings) - Design of coil structure over one
field-period with variable thickness based on
local stress/displacement when combined with
rapid prototyping fabrication technique, this
can result in significant cost reduction. - Mid
-size divertor unit (T-tube) applicable to both
stellarator and tokamak (designed to
accommodate at least 10 MW/m2) - Possibility of
in-situ alignment of divertor if required.
65
DC Blanket Parameters for Reference Case
R 7.75 m Fusion power 2355 MW Avg. wall
load 2.6 MW/m2 Max. wall load 4
MW/m2 Avg. plasma q 0.6 MW/m2 Max. plasma
q0.8 MW/m2
3-mm ODS FS Tmax/Tmin/Tavg 643C/564C/604C
Plasma q
1-mm RAFS Tmax/Tmin/Tavg 564C/536C/550C
FW He Coolant
Tcool 426 C