RWM Control to Enhance ITER Performance

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RWM Control to Enhance ITER Performance and Test
Advanced Modes for DEMO
Dale Meade work being carried out by G.
Navratil and J. Bialek Columbia University, M.A
Ulrickson, Sandia National Laboratory and C.
Kessel PPPL. December 8, 2005
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Control is the Central Issue for a Self-Heated
Burning Plasma (CTR)

• The goal is to create, control and sustain a
  burning plasma
• (Wright Flyer, Chicago Pile 1 gt Hanford Pu
  Production)
• Control power will be limited in high Q plasmas
  ( Pcontrol lt 16 for Q 25)
• Global Burn control thermal stability
  including fuel mix, ash and impurities
• Profile Burn control J(r) ECCD, NBI, LHCD
  instability-NTM, RWM
• Power Handling Control fast position,
• edge ergodic/gas,
• disruptions-gas jets, fast B (VDE)

• Particle Handling Control fuel mix, ash
  pumping, edge radiation
• This is how we control combustion
  now

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What has Happened in RWM Control Since Snowmass
2002?
Tokamak experiments have demonstrated AT
regimes with bN 4 approaching those needed for
AT regimes in burning plasmas. Garofalo et al
APS_DPP 2005 Experiments on present tokamaks
have shown the possibility of increasing the b -
limit (power density) using feedback coils to
stabilize n 1 resistive wall mode (RWM).
Major US focus. References Navratil-APS_DPP_2004,
Menard this meeting. Tokamak Simulation Code
Studies for FIRE demonstrated the possibility of
high-b steady-state burning plasma with
approaching DEMO parameters 100 non-inductive
(fbs 78), bN 4, 5 MWm-3, GN 2 MWm-2 for
4 tCR. VALEN studies for RWM coils mounted on
the shielding port plugs on FIRE have shown that
n 1 RWM could be stabilized up to bN 4.2.
These results and those expected in the next 5
- 10 years justify a serious look at implications
of implementing AT modes and RWM control on ITER.
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How would RWM Control Benefit ITER?
Studies in the US, Japan and Europe of
concepts for Fusion Power Plants have shown the
need for advances beyond the Design Physics Basis
for ITER. Having the flexibility in ITER to
explore and exploit a DEMO relevant power density
would significantly increase the value of ITER
for nuclear component testing and for providing a
design basis for DEMO. The physics
performance of ITER would be significantly
enhanced if the b - limit could be raised from bN
2 to 4. This would allow ITER to test
advanced modes with lower plasma current at
present design power densities. If successful,
this could help provide the physics basis for an
upgrade to increase the neutron wall loading in
ITER. This would also require an upgrade in the
power handling capability of the ITER first
wall/blanket and divertor. Kessel IEA BP
W60-Tarragano 2005
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Achieving AT Goals will require control of RWMs
Modification of APS-DPP VG shown by A. Garofalo
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What can the US Do?
In FY 2005 the FIRE/NSO activity evolved into
a study to assess the benefits and feasibility of
RWM coils to increase the operational b - limit
in ITER. Coils closer to the plasma have a
larger benefit to plasma performance but are more
difficult from the engineering point of view.
A number of locations have been considered for
the feedback coils including - coils
outside the TF coils - outside the vacuum
vessel but inside the TF coils - inside the
vacuum vessel - mounted behind the Blanket
Shield Module on the port plugs The feedback
coils might also be operated to generate edge
ergodic field layers for ELM control. Also some
potential for fast plasma position control.
The FIRE/NSO activity is focused on RWM coils
mounted on the mid-plane port plugs. This
configuration allows the coils to be
maintained/replaced by removal or exchange of
port plug assemblies.
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Steps in RWM Mode Study for ITER
The present study is for RWM coils mounted on
mid-plane port plugs. This configuration allows
the coils to be maintained/replaced by
removal/exchange of port plug assemblies.
1. Assessment of RWM requirements using VALEN -
High Priority Task assigned to US by ITER
International Team. (Navratil et al- Columbia)
2. Assessment of Eddy Current Penetration
through blanket shield modules with 3-D effects.
(Ulrickson et al - Sandia) 3. Update of VALEN
assessment using improved eddy current
penetration model. Be prepared to participate in
anticipated ITER Design Review next year. 4.
Define physical properties of the coils and
sensors, and overall system. 5. Integration of
internal RWM coils with Internal Ergodic
Coil (EU group -P. Thomas, E. Becoulet)
Vacuum Vessel and first wall engineering
group Diagnostic Interface Group 3 D
neutronics analysis
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Schematic Layout of RWM Coils on ITER
The study analyses the effect of a single turn
coil in the 10 cm gap between the blanket shield
module (BSM) and the port extension of a
mid-plane port plug.
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Applying Internal RWM Feedback Coils to the Port
Plugs in ITER Increases b-limit for n 1 from
bN 2.5 to 4
VALEN Analysis Columbia University
J. Bialek, G. Navratil, Columbia University
RWM Coil Concept for ITER
Baseline RWM coils located outside TF coils
7 RWM Coils mounted behind the BSM in every other
port except NBI ports. (assumes 9 ms time
constant for each BSM)
IEA BP W60 Tarragono http//fire.pppl.gov/iea_bp_2
005.html
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Eddy Current Modeling of Blanket Shield Module
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Tasks for the US Related to RWM in BP
1. Continue active RWM studies on existing
devices to understand physics basis of RWM
stabilization as a key element in the US advanced
confinement program aimed at an advanced DEMO.
2. Continue development of high-beta
steady-state Advanced DEMO regime for ITER - (bN
4, 100 non-inductive, 80 fbs) - expts and
modelling 3. Extend the ITER RWM analyses to
the next level of detail and address practical
aspects of implementing RWM control on
ITER Refine system specifications -
Preconceptual design of system (integration
with diag, RF in port plug) Neutron damage
of coil insulation - streaming through gap,
materials Coordinate with related ITER
efforts Internal magnetic diagnostic
coils Ergodic coils (EU Thomas,
Becoulet) Fast Internal control coils (VDE
control?) Begin Development of a simulation
of Integrated Plasma Control