ARIESCS Overview and Engineering Approach

1
ARIES-CS Overview and Engineering Approach

• Presented by A. R. Raffray
• University of California, San Diego
• L. El-Guebaly
• F. Najmabadi
• S. Malang
• T. K. Mau
• X. Wang
• and the ARIES Team
• FZK Visit
• Karlsruhe, Germany
• June 22-25, 2004

2
Outline

• Objectives of ARIES-CS study
• Physics summary
• Engineering plan of action
• Maintenance approaches
• Blanket designs
• Divertor considerations
• Conclusions and future work

3
Background
Assessment of Compact Stellarator option as
a power plant to help - Advance
physics and technology of CS concept and
address concept attractiveness
issues in the context of power plant
studies - Identify optimum CS configuration for
power plant - NCSX plasma/coil configuration as
starting point - But optimum plasma/coil
configuration for a power plant may be
different
4
ARIES-CS Program is a Three-Phase 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.
• Explore attractive coil topologies.

5
Overall Design Process Requires Close Interaction
among Physics, System and Engineering Studies
Calculate Consistent Machine Parameters Based on
System Studies Physics parameters (B, b, a
loss, etc..) and reactor size (aspect ratio,
major radius) for ignition for given
- Space
between coils
- Minimum plasma edge to coil distance
- Fusion power
Proposed CS Configuration(s) from Physics
Analysis (plasma shape coil configuration)

Start with NCSX based configuration
Explore alternate
concepts that might better extrapolate to power
plant
2-field period configuration

Others
Develop Engineering Design Configuration
Blanket, Shield, Vacuum Vessel,
Divertor Maintenance Scheme
Analysis of Design Requirements and Performance
Parameters Tritium breeding Shielding
requirements - Magnet configuration and heating
limits - Reweldability Thermal
efficiency Maintenance requirements - Size and
weight of blanket unit - Access Safety
requirements
6

• Physics Summary

7
Compact Stellarator Reactor Vision Physics
Parameters

• A steady-state toroidal reactor with
• - No disruptions
• - No conducting structures or active feedback
  control of instabilities
• - No current drive (? minimal recirculating
  power)
• - High power density (3 MW/m2)
• Likely configuration features
• - Rotational transform from a combination of
  bootstrap and externally- generated
  sources. (how much of each?)
• - 3D plasma shaping to stabilize limiting
  instabilities. (how strong?)
• - Quasi-symmetric to reduce helical ripple
  transport, alpha losses, flow damping. (how
  low must ripple be?)
• - Power and particle exhaust via a divertor.
  (what magnetic topology?)
• - R/?a?4.4 (how low?) and ?4 (how high?)
• ARIES aims to develop the physics basis and its
  reactor implications to determine an optimum CS
  configuration for an attractive power plant.
• Optimum design involves tradeoffs between
  physics requirements and engineering constraints.

8
Recent Stellarator Experiment Results are
Promising

• Wendelstein 7-AS
• (Germany)
• - b gt 3
• - enhanced confinement.
• - density control enhanced performance
  w/island divertor.

• Large Helical Device (Japan)
• - b  gt 3.
• - Te0 10 kev, Ti0 7 keV.
• - enhanced confinement t 0.36s
• 150 s pulses.

Helically Symmetric Experiment
(U. Wisc.) - Successful test of quasi-symmetry.
9
ARIES Program Plan to Assess Compact Stellarators

• Optimizing the compact stellarator reactor
  configuration.
• - Minimize alpha particle losses
• - Minimize ripple or maximize quasi-axisymmetry
  (sufficient to confine alphas)
• - Provide sufficient space between coil and
  plasma for blanket and shield
• - Identify simplest coil geometry with sufficient
  space between coils for simple maintenance
• - Preserve the overall size compactness
• - Explore configuration alternatives to best
  achieve above
• - Plasma configurations of periods, beta,
  aspect ratio, shaping.
• - Coil configurations simple coils
  alternatives to modular coils.
• Identify high-leverage issues for further
  physics research
• - What are the cost sensitivities?
• Develop an attractive design around the optimum
  configuration.
• - Requires close interaction between
  physicists and engineers

10
Initial Configurations for ARIES-CS Optimization
Studies

• We have developed two initial candidate CS
  configurations, with 2 and 3 field periods, for
  self-consistent evaluation and comparison.
• Parameter 3-field period (NCSX) 2-field
  period
• Coil-plasma distance, D (m) 1.4
  1.4
• ltRgt (m) 9.7 7.5
• ltagt (m) 2.15 2.0
• Aspect ratio 4.5 3.75
• b () 4.15 4.0
• Number of coils 18 16
• Bo (T) 5.65 5.0
• Bmax (T) 14.4 14.4
• Avg. wall load (MW/m2) 1.9 2.7
• Cases of 12 and 8 coils
• are also being considered
• for the 2-field period
• configuration.

11
Example of a Two-Field Period Stellarator with 8
Coils and A2.5 Designed by Running the NSTAB
Equilibrium Code (recently proposed by P.
Garabedian)

• Lots of space for maintenance.
  
• An engineers dream!
• But needs further study to assess feasibility.

12
Initial System Runs Indicate Possibility of More
Compact Reactor

• Example parameters for a case based on NCSX
  (3-field period) with - Bmax16 T Minimum
  shield thickness 0.86 m (WC case)

• When including peaking factor (1.5-2), wall
  load is too high to be accommodated by
  blanket/first wall design
• New system runs being done for more moderate
  wall load
• System code still under development and cost
  is being implemented to get a better indication
  of the trade-offs
• However, initial results encouraging in term
  of putting the C in CS!

13
Configurations Optimized for a Energy Loss Were
Found

• The losses are localized in narrow helical
  bends just below the OB midplane, and the peak
  heat load may be high (several MW/m2).
  
• May lead to challenging local first wall design
  
• Can divertor be placed at this location?
• Can alpha loss be minimized (case with 10 loss
  shown here but can be up to 25)

Footprints of escaping as on LCMS for N3B5D.
B5.5 T, Vol1000 m3.
a loss 10
Heat load may be localized and high
14
A New Class of QA Configurations with Good Flux
Surface Quality at High b

• Negative magnetic shear tailored to match
  magnitude of the bootstrap current such that the
  presence of low order resonance is avoided at the
  target beta (6 in the present study).
• Good QA with low residue non-axisymmetric fields
  (1-2) and low effective ripple (lt1).
• Good a-particle confinement with energy loss
  fraction in 1000 m3 reactors at 6.5 T lt 10
  (confinement proportional to B2).
• Deep magnetic well in vacuum. Configurations with
  4 -9 a-particle loss found.
• Toroidally averaged elongation (gt1.8) and
  triangularity (gt 0.7) match those in advanced,
  high beta tokamaks and other classes of QA
  stellarators with good MHD stability properties.
• Reasonably simple shape
• Being further studied.

15

• Engineering Plan of Action

16
Plan for Engineering Activities
Machine Parameters and Coil Configurations
Maintenance Scheme 2
Maintenance Scheme 1
Maintenance Scheme 3
Evolve in conjunction with scoping study of
maintenance scheme and blkt/shld/div.
configurations
Blkt/shld/div. 1
Blkt/shld/div. 1
Blkt/shld/div. 2
Blkt/shld/div. 3
Blkt/shld/div. 2
Blkt/shld/div. 3
Blkt/shld/div. 1
Blkt/shld/div. 2
Blkt/shld/div. 3
Phase I
Optimization in conjunction with maintenance
scheme design optimization
Optimize configuration and maintenance scheme
Optimize configuration and maintenance scheme
Optimize configuration and maintenance scheme
Phase II
Overall Assessment and Selection
Phase III
Detailed Design Study and Final Optimization
17
Engineering Activities Phase I
Perform Scoping Assessment of Different
Maintenance Schemes and Consistent Design
Configurations (coil support, vacuum vessel,
shield, blanket) - Three Possible
Maintenance Schemes 1. Field-period based
replacement including disassembly of modular coil
system (e.g. SPPS, ASRA-6C) 2. Replace
ment of blanket modules through small number of
designated maintenance ports (using
articulated boom) 3. Replacement of blanket
modules through maintenance ports arranged
between each pair of adjacent modular
coils (e.g. HSR) - Each maintenance
scheme imposes specific requirements on machine
and coil geometry
18
Engineering Activities Phase I
Scoping analyses of the following
blanket/shield/VV configurations have been
completed 1. Self-cooled liquid metal blanket
(will probably need He-cooled divertor
depending on heat flux) a) with
SiCf/SiC (Pb-17Li) b) with insulated
ferritic steel and He-cooled structure (Pb-17Li
or Li) 2. He-cooled blanket (liquid breeder
or solid breeder) with ferritic steel and
He-cooled divertor 3. Flibe-cooled ferritic
steel blanket (might need He-cooled divertor
depending on heat flux) Information on these
different concepts from presentations at
ARIES project meetings (see
http//aries.ucsd.edu/ARIES/ ) Here, as
illustration, the following are
described - one example concept with
field-period based maintenance - one modular
concept with port-based maintenance
19

• Example field period-based maintenance approach
  and blanket design

20
Example Scoping Analysis of Maintenance Scheme
and Engineering System Design for NCSX-Based
Coils and Plasma Shape with 3-Field Period
21
Maintenance Approaches for NCSX-Like Power Plant

• For blanket module-based maintenance scheme
• - Desirable to accommodate 2
  m x 2 m x 0.5 m module
• - Modular maintenance scheme with limited
  number of ports and articulated boom seems
  possible
• - Not suited for maintenance scheme with ports
  between each pair of adjacent coils unless
  reactor size is increased and/or 2- field
  period is considered (resulting in larger
  individual port sizes)
• Seems best suited for field period-based
  maintenance

22
Example Design Approach for Field-Period Based
Maintenance Scheme
Need to Design Coil Support Structure to
Accommodate the Three Kinds of Forces Acting on
Coils while Being Compatible with Maintenance
Scheme - Large centering forces need to be
reacted by strong bucking cylinder. - No net
forces between coils from one field period to the
other. - Out-of plane forces acting between
neighbouring coils inside a field period require
strong inter-coil structure. - Weight of the
cold coil system has to be transferred to the
warm foundation without excessive heat
ingress. - The radial movement of a field
period as required for blanket replacement should
be possible without disassembling coils,
inter-coil structure and thermal insulation in
order to avoid unacceptably long down time.
- Transfer of large forces within a field period
and between coils and bucking cylinder is not
possible between cold and warm elements. This
means that the entire support structure has
to be operated at cryogenic temperature. - To
facilitate opening of the coil system for
maintenance, separate cryostats for the
bucking cylinder in the centre of the torus
and for every field period are envisaged.
23
Proposed Solution Enclose the Individual
Cryostats in a Common External Vacuum Vessel
Minimize fabrication costs and mass of waste by
separating regions requiring replacement
(replacement units) from the life-time
components (permanent zone). Divide neutron
shielding into high temp. and low temp. regions
for two reasons - to enable passive low
temperature cooling in the outer shield region
for safety reasons, - to minimize low
temperature heat (waste heat) to lt1 of total
heat
24
Schematic of Arrangement of all Coils of a Field
Period on a Supporting Tube
Winding of Coils into Grooves in the Supporting
Tube
25
Arrangement of Long Legs to Support the Weight of
the Cold Structure

• Conflicting requirements
• Sufficiently large cross section of legs required
  to minimize mechanical stresses
• Small ratio between cross section and length of
  the legs required to minimize heat ingress ? long
  legs
• e.g. for a total cold structure weight of 1,800
  tons and 3x3 legs
• - required cross section of one cylindrical
  tube leg 100mm diameter and 3 mm wall
  thickness
• - corresponding heat flow between 300 K and 4 K
  through 0.5 m long legs 500 watt for all legs
• (Much longer legs shown in sketch)

26
Arrangement of Legs Supporting the Weight of
FWBlanketShields and of Coolant Routing
All access tubes are coming from the bottom.
Location of the access tubes close to points
of mechanical support. Concentric pipes with
inlet flow in the annulus and outlet flow in the
central pipe allow for sliding seals in the
centre tube.
Much larger weight to be supported than for
cold structure but legs are between warm
regions and there is nearly no limit on cross
section. - e.g. 4 legs in each field period,
500 mm diameter, 35 mm wall thickness
27
Steps to be Performed for an Exchange of Blankets
(at end of life or because of failure)
1. Warm up the coil system including the
mechanical support structure. 2. Flood the
plasma chamber and the cryostats with inert
gas. 3. Cut all coolant connections to the field
period to be moved, using in-bore tools operating
from the bottom. 4. Open the outer side of the
external VV in the range of the field period to
be moved. 5. Slide the entire field period,
composed of blankets, shields, cryostats, coils
with inter coil structure, and all thermal
insulation, outward in radial direction an the
flat surface at the bottom. It should be possible
to use an air-cushion system for this
movement. 6. Cut the coolant access tubes to the
blanket to be replaced. 7. Slide out from the
openings at both ends the replacement units,
composed of FW, breeding zone, and a structural
ring (serving at the same time as a first high
temperature shield). The replacement unit is
moved outwards in the toroidal direction on a
rail attached to the replacement zone, and
sliding on the low temperature shield (possibly
with an air-cushion system).
28
Extrapolation of the Maintenance Scheme to a CS
For simplicity, the design and maintenance
scheme have been illustrated for a circular
plasma with uniform cross-section. They have
to be adapted to the particular geometry of a
CS In particular, can a replacement unit be
extracted toroidally in a CS geometry?
Replacement unit in pink (blanket) Permanent
components hot shield in blue and cold shield in
green
60
0
30
29
Removal Blanket Replacement Unit under Proposed
Maintenance Scheme is Possible for CS
Replacement unit can be removed toroidally by
- using off-axis track system - locally
decreasing shield and correspondingly increasing
blanket to provide a minimum of 100 cm shield
e.g.
Removal would be easier for geometries with
smaller variation of the plasma cross
section, and coils with smaller deviations from
circular shape.
30
Example of a Blanket Concept Suitable for Field
Period-Based Maintenance
This maintenance scheme allows for very large
blanket units with virtually no weight limit.
Self-cooled liquid breeder (Li, Pb-17Li )or
dual-coolant concept would be good
candidates Dual coolant Li/He concept with
ferritic steel as structural material used as
example - no need for additional neutron
multipliers as in solid breeder or FLiBe
blankets. - no need for large internal heat
transfer surfaces as in solid breeder blankets.
- volumetric heating of the breeder/coolant
provides the possibility to set the coolant
outlet temperatures beyond the maximum structural
temperature limits. - FW and the entire steel
structure cooled with helium. - Li flowing
slowly toroidally (parallel to major component of
magnetic field) to minimize MHD pressure drop
used as breeder/coolant in the breeding zone.
- electrically insulating coating between Li
and FS not required but thermal insulating
layer might be needed to maintain Li/FS temp.
within its limit (600C)
31
Schematic of Blanket Concept Showing Cooling
Arrangement
Cross section of toroidal cooling channels
32
Example Blanket Parameters for 2 GW Fusion Power
Plant

• For one replacement unit (1/6 of entire
  machine)
• - Total thermal power to be removed
  400 MW
• - Heat to be removed with Li/He
  300/100 MW
• Helium cooling of FW
• - Pressure
  8MPa
• - Inlet/outlet temperature
  400/500C
• - Velocity
  70 m/s
• - Heat transfer coefficient 4,200 W/(m2-K)
• - Pressure drop 0.1
  MPa
• Lithium cooling of breeding zone
• - Inlet/outlet temperature
  500/800C
• - Velocity
  0.12 m/s
• Heat transfer coefficient 450 W/(m2-K)
• Pressure drop (assuming perpendicular B1T) 0.1
  MPa
• Tritium self-sufficiency has been estimated
  with breeding zones 47-62 cm

33

• Example modular maintenance approach and blanket
  design

34
Modular Maintenance Approach
ITER-like rail system articulated boom
extremely challenging in CS geometry due to
roller coaster effect and to non-uniform plasma
shape and space Preferable to design
maintenance based on articulated boom only

• From EDITH-system, boom built with
• - a total length of 10m
• - a reach of /- 90 in NET
• - pay load of 1 ton
• - maximum height of 2 m
• 3 field-period configuration could be
  compatible with maintenance through limited
  number of ports

Experimental -In-Torus Maintenance System for
Fusion Reactors, FZKA-5830, Nov. 1966.
35
2-Field Period Configuration (from P. Garabedian)
Results in More Space for Maintenance Ports that
Could Allow for Maintenance Scheme with Ports
Between each Pair of Adjacent Coils
Initial Assumed Parameters R 8 m ltagt 2.3 m A
3.5 16 coils ( 8 per period ) Thickness 57
cm. Aspect ratio 2 Coil-plasma min. distance
1.5 m Example of Parameters from System Study R
6.62 m B (axis) 6.55 T ltbgt 4.47 16 coils
( 8 per period ) Coil pack dimension 40 cm x 40
cm Fusion power 2 GW Plasma aspect ratio 3.75
Even more port space with newer
configurations featuring fewer coils (12 and
8)
36
Comparison of Horizontal Port Access Area Between
Adjacent Coils for Different Configurations
Horizontal space available between
coils, toroidal dimension x poloidal dimension (m
x m)
Assuming a coil cross-section of 0.57 m x 1.15 m
37
Considerations on Choice of Module Design and
Power Cycle for a Ceramic Breeder Concept

• The blanket module design pressure impacts the
  amount of structure required, and, thus, the
  module weight size, the design complexity and
  the TBR.
• For a He-cooled CB blanket, the high-pressure He
  will be routed through tubes in the module
  designed to accommodate the coolant pressure. The
  module itself under normal operation will only
  need to accommodate the low purge gas pressure
  ( 1 bar).
• The key question is whether there are accident
  scenarios that would require the module to
  accommodate higher loads.
• If coupled to a Rankine Cycle, the answer is yes
  (EU study)
• - Failure of blanket cooling tube subsequent
  failure of steam generator tube can lead to
  Be/steam interaction and safety-impacting
  consequences.
• - Not clear whether it is a design basis (lt10-6)
  or beyond design basis accident (passive means
  ok).
• To avoid this and still provide possibility of
  simpler module and better breeding, we
  investigated the possibility of coupling the
  blanket to a Brayton Cycle.

38
Low-Pressure Requirement on Module Leads to
Simpler Design

• Simple modular box design with coolant flowing
  through the FW and then through the blanket
• - 4 m (poloidally) x 1 m (toroidally) module
• - Be and CB packed bed regions aligned
  parallel to FW
• - Li4SiO4 or Li2TiO3 as possible CB
• In general modular design well suited for CS
  application
• - accommodation of irregular first wall
  geometry
• - module size can differ for different port
  location to accommodate port size

39
Arrangement of the Breeder and Beryllium Pebble
Beds
Inside the breeding zone, each breeder bed is
enclosed by two cooling plates. This assembly
is filled outside the blanket box with ceramic
pebbles, and closed. All the cooling plates
are welded to larger manifold plates before
inserting the breeding zone into the blanket box.
Beryllium pebbles are filled into any empty
space inside the box, and compacted by vibrating
the module.
- To weld the cooling plate(ODS steel) to the
manifold(ODS steel) is a major issue because of
difficulty of producing high strength welds
on ODS steel.
40
Access Tube Shielding Plug for Cutting Tube
Prior to Removing Blanket Module
Cut the assembly weld in the front disk at the
FW first. Pull out the shielding plug with
inner tube. Cut the outer tube weld located
behind the permanent shield. Open/Remove the
attachment bolts. Pull out the blanket module.
41
Steps to be Performed for an Exchange of Ceramic
Breeder Blankets
Pull out first the Closing Plugs from access
port Open and remove the first and second
doors. Cut the coolant access tubes from
back. Pull out the closing plug and insert the
articulated boom into the plasma region. The
boom has to be equipped with two classes of
tools Tools for opening attachment bolts,
inserted from the plasma region through radial
gaps between the modules. Tools for
cutting/re-welding the front disk at the FW as
well as the coolant access tubes at the back of
blanket module. Remove other blanket
modules Cut the weld in the front disk at the
module FW and remove module shielding
plug. Cut the weld of the coolant access tubes
at the back of blanket. Remove the attachment
bolts.

42
Ceramic Breeder Blanket Module Configuration
Initial number and thicknesses of Be and CB
regions optimized for TBR1.1 based
on - Tmax,Be lt 750C - Tmax,CB lt
950C - kBe8 W/m-K - kCB1.2 W/m-K - dCB
region gt 0.8 cm 6 Be regions 10 CB regions
for a total module radial thickness of 0.65 m
He flows through the FW cooling tubes in
alternating direction and then through
3-passes in the blanket
43
Example Brayton Cycle350 MWE Nuclear CCGT
(GT-MHR) Designed by U.S./Russia
44
He-Cooled Blanket Brayton Cycle Blanket
Coolant to Drive Cycle or Separate Coolant HX?

• Brayton cycle efficiency increases with
  increasing cycle He max. temp. and decreases with
  increasing cycle He fractional pressure drop,
  DP/P
• - Using blanket coolant to drive power cycle
  will increase cycle He DP/P
• - With a separate cycle He, DP/P can be reduced
  by increasing system pressure (15 MPa)
• - From past studies on CB blanket with He as
  coolant, a max. pressure of 8 MPa has been
  considered
• - Utilizing a separate blanket coolant and a HX
  reduces the maximum He temperature in the cycle
  depending on DTHX
• Tradeoff study required on case by case basis
  to help decide whether or not to use separate He
  for cycle
• For this scoping study, a separate He cycle and
  a HX are assumed to enable separate optimization
  of blanket and cycle He conditions (in particular
  pressure) and to provide independent flexibility
  with increasing cycle He fractional pressure drop

45
Two Example Brayton Cycle Configurations
Considered

• Brayton II
• A higher performance - configuration with
  4-stage compression 3 inter- coolers and
  4-stage expansion 3 re-heaters
• (shown on next slide)
• - Considered by P. Peterson for flibe blanket

Brayton I - A more conventional
configuration with 3-stage compression 2
inter- coolers and a single stage expansion
46
Brayton II
47
Comparison of T-S Diagrams of Brayton I and
Brayton II
Brayton I 3-stage compression 2
inter- coolers and a single stage expansion
Brayton II 4-stage compression 3
inter-coolers and 4-stage expansion 3
re-heaters More severe constraint on
temperature rise of blanket coolant
48
Procedure for Blanket Thermal Hydraulic and
Brayton Cycle Analysis
Start with number of breeder and multiplier
regions from initial neutronics optimization
calculations to provide the required breeding,
TBR1.1 - 6 Be regions 10 CB regions for a
total module radial thickness of 0.65 m
Perform calculations for different wall loading
by scaling the qvol in the different regions
and setting the individual region thicknesses and
coolant conditions to accommodate the maximum
material temperature limits and pressure drop
constraints. - Tmax,Be lt 750C Tmax,CB lt
950C Tmax,FS lt 550C or 700C (ODS) - kBe8
W/m-K kCB1.2 W/m-K - dCB region gt 0.8 cm dBe
region gt 2 cm - P pump/P thermal lt 0.05 (unless
specified otherwise) Coolant conditions set in
combination with Brayton cycle under following
assumptions - Blanket outlet He is mixed
with divertor outlet He (assumed at 750C and
carrying 15of total thermal power) and then
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
49
Radial Build from Original Neutronics
Optimization Calculations
qvol based on wall load of 4.5 MW/m2 and
scaled for different wall
loads qplasma unspecified and assumed as
0.5 MW/m2 in all calculations (unless
otherwise specified)
50
Example Optimization Study of CB Blanket and
Brayton Cycle
Cycle Efficiency (h) as a function of neutron
wall load (G) under given constraints For a
fixed blanket thickness (Dblkt,radial) of 0.65 m
(required for breeding), a maximum G of 5 MW/m2
can be accommodated with Tmax,FSlt550C h
35 Tmax,FSlt700C h 42 The max. h
corresponds to G 3 MW/m2 Tmax,FSlt550C h
36.5 Tmax,FSlt700C h 44 The max. h 47
for G 3 MW/m2 for Brayton II. However, as
will be shown, Ppump/Pthermal is unacceptably
high in this Brayton II case.
51
Corresponding He Coolant Inlet and Outlet
Temperatures
Difference in blanket He inlet and outlet
temperatures much smaller for Brayton II
because of reheat HX constraint - Major
constraint on accommodating temperature
and pressure drop limits
52
Corresponding Maximum FS Temperature
For lower G (lt3 MW/m2), Tmax,FS limits the
combination of blanket outlet and inlet
He coolant temperatures For higher
G(gt3MW/m2), Tmax,CB and Tmax,Be limit the
combination of blanket outlet and inlet He
coolant temperatures
53
Corresponding Ratio of Pumping to Thermal Power
for Blanket He Coolant
An assumed limit of Ppump/Pthermal lt 0.05 can
be accommodated with Brayton I. With Brayton
II the smaller coolant temperature rise requires
higher flow rate (also for better convection) and
Ppump/Pthermal is much higher particularly for
higher wall loads On this basis, Brayton II
does not seem suited for this type of blanket as
the economic penalty associated with pumping
power is too large
54
Effect of Changing Blanket Thickness on Brayton
Cycle Efficiency
Decreasing the total blanket thickness to from
0.65 m to 0.6 m allows for accommodation of
slightly higher wall load, 5.5 MW/m2 and allows
for a gain of 1-2 points in cycle efficiency at a
given neutron wall load But is it acceptable
based on tritium breeding?
55
Effect of Changing the Plasma Surface Heat Flux
on Brayton I Cycle Efficiency
The efficiency decreases significantly with
increasing plasma surface heat flux. This is
directly linked with the decrease in He coolant
temperatures to accommodate max. FS temp. limit
in the FW (700C). (see next slide)
56
Effect of Changing the Plasma Surface Heat Flux
on Blanket He Temperatures
The key constraint is the max. FS temp. limit
in the FW (700C). To accommodate the increase
in plasma heat flux, the He coolant temperatures
must be decreased, in particular the outlet He
temperature Challenging to accommodate this
design with a Brayton cycle for plasma heat flux
much higher than 0.5 MW/m2
57

• Divertor Considerations

58
Designing a Divertor for ARIES-CS is a
Challenging Task

• Heat load and location highly uncertain
• Need complete tools for accurate analysis
• Focus of future effort
• Preliminary scoping analysis based on q 10
  MW/m2
• He-cooled concept best suited with different
  blanket concepts considered
• - Simple channel with enhancement
  (micro-channel?)
• - Porous media (high pressure drop)
• - Fins
• Liquid divertor?

59
Example Results for He Flow in a Regular Tube of
Length 0.5 m Illustrates the Limitation of this
Simple Configuration
Must be above DP 0.4 MPa line to
maintain DP/Pin0.05 for example
Pin8 MPa For tungsten,
Twall lt 1000-1100C, there is no
solution.
- Thus, simple channel configuration very limited
in max. q that can be accommodated (ltlt10
MW/m2) - Perhaps very short microchannels would
help, but very challenging
60
Example Calculations Using the MERLOT Code for a
Porous Flow Divertor Concept under 10 MW/m2

• Set rin and rout based on pressure drop
  optimization for given flow rate
• Set inlet temperature based on a Tungsten
  design.

61
Temperature Pressure Drop as a Function of
Velocity for 10 MW/m2 Example
Temperature limit based on 2 mm W wall (assume
100 0C per mm surface 1,300 0C)
1,100 0C
60 m/s
114 m/s
60 m/s
DP 3.7 MPa
114 m/s
DP 2.9 MPa
Very high pressure drop to provide required
heat transfer performance Some help by reducing
Tinlet (is it ok with W?) and optimizing porous
microstructure, but concept still marginal
in particular if q is even higher Need to
look at other concepts (e.g. FZK pin concept)
62
Future Engineering Activities
On the engineering front, we are making good
progress in many areas - Blanket/vessel/coil
configuration - Maintenance scheme - Ready to
start converging on a couple of concepts for more
detailed study within the next few months
(probably one concept with field period-based
maintenance and one concept with port
maintenance) - Start detailed point design study
within a year However, one key area where we
are lagging is the divertor - Need better
definition of location and heat loads - Need a
concept with good heat transfer performance
(accommodating qgt10 MW/m2) and acceptable
pumping power - Focus of our immediate
effort - Tools for calculating heat flux on
physics side
63
Much Work Done During Phase I - Results will be
Presented at 16th TOFE

• Invited Oral Papers for ARIES Special Session
• 1. F. Najmabadi and the ARIES Team, Overview of
  ARIES-CS Compact Stellarator Study
• 2. P. Garabedian, L. P. Ku, and the ARIES Team,
  Reactors with Stellarator Stability and Tokamak
  Transport
• 3. J.F. Lyon, L. P. Ku, P. Garabedian and the
  ARIES Team, Optimization of Stellarator Reactor
  Parameters
• 4. A. R. Raffray, L. El-Guebaly, S. Malang, X.
  Wang and the ARIES Team, Attractive Design
  Approaches for Compact Stellarator
• 5. L. El-Guebaly, R. Raffray, S. Malang, J. Lyon,
  L.P. Ku and the ARIES Team, "Benefits of Radial
  Build Minimization and Requirements Imposed on
  ARIES-CS Stellarator Design"
• Additional Papers Accepted for Presentation
• 6. L. El-Guebaly, P. Wilson, D. Paige and the
  ARIES Team, "Initial Activation Assessment for
  ARIES-CS Stellarator Power Plant"
• 7. L. El-Guebaly, P. Wilson, D. Paige and the
  ARIES Team "Views on Clearance Issues Facing
  Radwaste Management of Fusion Power Plants"
• 8. L. Bromberg, J. Schultz, L. El-Guebaly and the
  ARIES Team, Advanced Options for Modular
  Stellerator Magnets
• 9. S. Abdel-Khalik, S. Shin, M. Yoda, and the
  ARIES Team, "Design Constraints for
  Liquid-Protected Divertors"
• 10. X. Wang, S. Malang, A. R. Raffray and the
  ARIES Team, Maintenance Approaches for ARIES-CS
  Power
• T. K. Mau and the ARIES Team, Alpha Particle
  Loss and Heat Load Assessments for a Compact
  Stellarator Power Plant
• 12. A. R. Raffray, S. Malang, L. El-Guebaly, X.
  Wang and the ARIES Team, Ceramic Breeder Blanket
  for ARIES-CS