disruption mitigation with high-pressure gas...
TRANSCRIPT
Disruption Mitigation with
High-Pressure Gas Jets
D. Whyte1, R. Granetz2, V. Izzo2, T. Biewer2, M. Reinke2, J. Terry2,
A. Bader2, M. Bakhtiari1, T. Jernigan3, G. Wurden4
Workshop on Active MHD Control 2005
Madison, Nov. 2, 2005
1 University of Wisconsin2 MIT Plasma Science and Fusion Center3 Oak Ridge National Laboratory4 Los Alamos National Laboratory
Disruption mitigation with
high-pressure noble gas jet
High-pressure noble gas jets can mitigate 3 problems arising from
disruptions, without contaminating subsequent discharges.
1) Divertor thermal loading: sudden heat load ablates/melts divertor material
Solution: Deliver large quantities of impurity into core plasma to dissipate
high fraction of plasma energy by relatively benign, isotropic radiation
2) Halo currents: large mechanical J B forces on vessel/first wall components
Solution: Rapid thermal quench, resulting in a plasma that remains centered
in vessel during current quench, substantially reducing vessel halo currents
3) Runaway electrons: Relativistic MeV electrons from avalanche amplification
during current quench in large-scale tokamaks
Solution: Suppression by large density of bound electrons in plasma volume.
These issues are particularly severe for ITER
Outline
• New experimental results from Alcator C-Mod
• Modeling: Gas delivery and radiation
• ITER:
• Using disruptions for Tritium recovery
• Beryllium wall melting and MHD stability
Why do high-pressure noble gas jet
experiments on C-Mod?
Gas jet mitigation has been studied in DIII-D (D. Whyte et al, PRL 89, 55001)
It was postulated that the impurities penetrated as a neutral gas jet, since
Pjet (20-30 kPa) > Pplasma (8 kPa vol. avg, 30 kPa on axis)
Alcator C-Mod has:
• O(10x) higher pressure
• O(10x) higher Wth density
• O(10x) higher Wmag density
• Metallic wall
• Faster disruption timescale
– Challenging test of ability to convert
plasma energy to radiation on a fast
enough timescale (~1 GW)
Cooling front propagation into DIII-D
Specific goals of initial C-Mod experiments
• Study penetration of gas jet/impurities:
–Fast camera, Te and X-ray profiles
–NIMROD modeling, KPRAD modeling
• Disruption mitigation:
–Halo currents (current quench time, vertical displacement)
–Thermal deposition to divertor (IR camera, radiated fraction)
• Engineering/operational issues:
–Optimization of gas jet system (quantity & speed at LCFS)
–Reliability, reproducibility, post-disruption recovery
Not addressed: gas jet mitigation of actual disrupting plasma
–Gas delivery speed; realtime disruption sensing
C-Mod gas jet system optimized
based on DIII-D experience
To
vacuum
vessel
Plenum (70 bar) filled with
He, Ne, Ar, or Kr
Fast valve (ORNL)
Tokamak valve at port flange
Valve outside TF/ neutrons +
Gas delivery through ~m tube
desirable for application of
gas jets on ITER
C-Mod gas jet system optimized
based on DIII-D experience
Outlet nozzle is
extremely close to
plasma edge (2-3 cm)
to maximize gas
injected into plasma.
Nozzle is pointed at
plasma center
Injects 0.5-1.0 1023
atoms in a few ms
(plasma inventory is1.5-3.0 1020 D+, e–)
Example gas jet shot (Helium)
Gas jet valve fires at t=0.8 s
Gas jet valve fires at t=0.8 s
Example gas jet shot (Helium)
Thermal quench occurs a
few milliseconds later
n > 2x1021 m-3
Example gas jet shot (Helium)
Gas jet valve fires at t=0.8 s
Followed by current spike
and current quench
– Loss of vertical stability
– Halo currents
Thermal quench occurs a
few milliseconds later
n > 2x1021 m-3
Effect of gas jet on current quench
Typical (no gas jet) Argon gas jet
Fastercurrentquench
Less verticaldisplacement
Less halocurrent indivertor
Halo current reduction improves with Z
Fraction of energy radiated increases with Z
IR imaging of divertor surfaces
Inb
oard
Outboarddivertor
wall
Floor
Thermal deposition is not toroidally uniform,
but rather concentrated at leading edges
Temperature image from IR camera
Gas jet reduces energy deposition on
divertor surface
Temperature
differences
evident
during
cooldown
High-speed camera images indicate
shallow penetration of gas jet as neutrals
200 _s frame
Analysis of images of neutral gas
jet shows predominantly toroidal
flow, not deep radial penetration
divertor
High-Z vs. Helium penetration characteristics.
Argon HeliumSXR(au)
tTQ
0
-2
High-Z vs. Helium penetration characteristics.
Mitigation effectiveness does not seem linked to
“strong” particle penetration as found in He case.
Argon Helium
~ Uniform Tedrop beforeCurrentquench
“Weak”particlepenetrationArgon@ r/a < 0.6
Cold frontpropagates
Correlated“Strong”He particlePenetration
Central impurity density (r/a<0.9)
nAr < 1019 m-3 nHe ~ 5x1020 m-3
NIMROD simulations (V. Izzo):
MHD plays a major role for high-Z cases
A 2/1 instability destroys outer flux surfaces, 1/1 mode flattens core temperature
Fast cooling of edge region triggers MHD modes:
Therefore only shallow (r/a>0.85) impurity penetration is
required to collapse core temperature on a fast timescale.
Effect of MHD mixing? Helium case
stands out again.
RMS amplitude of MHD
fluctuations from midplane
pickup coils shown relative to
the time of the final thermal
quench
Magnitude and history very
similar for Ne, Ar, Kr.
Helium gas jet leads to earlier
MHD rise…effect of forced
pressure gradient by particle
(ion) penetration?
He
Ar Kr
Ne
Summary: Operational results
• Helium, neon, argon, and krypton gas jets used successfully
• Very reproducible effects and timing (± 0.3 ms)
• Proved to be benign – no problem with following discharge
• No runaways generated (unlike with high-Z killer pellets)
Summary: Gas jet/impurity penetration
• Impurities do not penetrate far into plasma as neutral gas
• For higher Z gas jets (Ne, Ar, Kr) NIMROD modeling shows
triggered MHD playing dominant role
• Different picture for He gas jet with good particle penetration
and different MHD characteristics
Conclusion: Since deep gas jet neutral penetration is not required,
gas jet mitigation seems plausible in ITER and reactors.
Summary: Disruption mitigation
• Halo currents are reduced by as much at 50%
–Faster current quench --> less vertical displacement
– Improves with Z of gas (higher resistivity)
• More plasma energy (~100%) is converted to benign radiation
–Less heating of divertor surfaces
–Radiated energy fraction improves with Z of gas
Conclusion: Radiated power levels are high enough to affect
energy balance on the short C-Mod disruption timescale.
Higher Z impurities, which are better radiators, are more
effective.
Future work
• Extend to higher performance plasmas
–Higher Wth: 0.11 MJ --> 0.25+ MJ (ITER energy densities)
–Higher Ip: 1 MA --> 2+ MA
• Gas jet mitigation of disrupting plasmas
–Fire into programmed VDE’s
–Gas flow rate through system may matter: possible tradeoff in
Z of gas (higher speed vs better radiator; mixed gases)
–Realtime disruption sensing and gas jet firing (ultimate goal)
• Further analysis of energy accounting (particularly for He);
Address toroidal symmetry questions; NIMROD modeling using
KPRAD, NIMROD modeling of halo current reduction; etc.
Outline
• New experimental results from Alcator C-Mod
• Modeling: Gas delivery and radiation
• ITER:
• Using disruptions for Tritium recovery
• Beryllium wall melting and MHD stability
Frictional dissipation of the gas “shock” is important for gas
delivery to plasma on disruption timescale, and will set
requirements for disruption trigger time
Realistic Volumetric Impurity
Deposition intoRadiation Code
Friction-freeshock with uniformgas delivery
Gas delivery with friction
0-D KPRAD coupled with pipe code
We have have considered acurrent decay equation inKPRAD.
dIpdt
=2 R0L
Et
Input from pipe code
Assume uniformdeposition to plasmavolume
No other free parametersfor ionization/thermal balanceKPRAD calculation.
Coupled calculations of time-dependent gas
delivery and energy / radiation balance
compared to C-Mod data
0
1
2
3
4
5
6
7
8
0 5 10 15 20 25 30 35 40
Shot No.
Tim
e (
ms)
Helium
Neon
Argon
Krypton
KPRAD
Time delay between Triggering valveAnd Beginning of Current Quench
Compared to KPRAD + Gas Flow Model
Current decay rate is reproducible &
set by gas species injected
• “Dip” in -dIp/dt at
t=0.5 - 1.5 ms is loss of
closed flux surfaces.
• Current quench rate clearly
controlled by resistivity of
impurity-dominated
plasma….
• Particle “penetration” is very
good in ~ zero Beta CQ
plasma as expected from gas
delivery.
Kr
Ar
Ne
He
0.0
0.2
0.4
0.6
0.8
1.0
1.2
1.4
1.6
0 5 10 15 20
Gas atomic number
L/
R d
ecay t
ime
for c
ore p
lasm
a (
ms)
KPRAD + Idealized shock
KPRAD + Frictional shock
Data
Coupled calculations of time-dependent gas delivery and
energy / radiation balance compared to C-Mod data
Physics in hand:
Gas delivery trigger needs,
Global energy balance,
CQ equilibrium Halo mitigation
/w runaway electron suppression
He Ne Ar
Current Quench L/R time Core cooling time
Physics missing:
Coupled roles of edge neutral
penetration, radiation, MHD, and
heat conduction
NIMROD + KPRAD
Outline
• New experimental results from Alcator C-Mod
• Modeling: Gas delivery and radiation
• ITER:
• Using disruptions for Tritium recovery
• Beryllium wall melting and MHD stability
Exploit rapid dissipation of plasma energy to wall
surface in order to desorb H/D/T without damaging
material surfaces.
-20
-10
0
10
20
30
40
50
-20 0 20 40 60 80 100
Stored energy at disruption (kJ)
Gas r
eco
vered
-
Gas i
np
ut
(To
rr-L
)
Plasma shots
No plasma
Energy density difference:Unmitigated disruption in C-Mod
compared to radiativeterminations in ITER.
Disruptions with closed pumpvalves for H/D particle balance
• C-Mod goal:
Control H/D ratio after vent and surface
cleaning for ICRH minority heating
– Surface analysis shows H/D ~ 5-10 in
near surface, due to H20 absorption.
– Large starting H+D inventory:
• H/(B+Mo) ~ 10-40% in first
micron(s)
• H inventory > 1000 Torr-L H2
• ITER goal:
Routine in-situ Tritium recovery
(PSI 2004)
• Expect threshold in local energy
density for removal: strong Tsurf
dependence on H diffusion in Mo/B.
– Estimate T th> 500-1000 C << Tmelt
H/D recovery by planned disruptions was
successful on C-Mod
Favorable scaling for ITER
0
2
4
6
8
10
12
14
16
18
20
H2
rem
oval
eff
icie
ncy
(To
rr-L
/ 1
5 m
in.
)
ECDC Fiducial
shotsq=2
terminations
VDE
terminations
Recovered ~30% H2 in single operation day.H/D reduced ~35%Wall H/D depleted
5-10x more efficient forH2 recovery thantypical techniques
Beryllium Melt Layer MHD
stability with disruptions
• Ideal rapid, quasi-uniform radiation flash desired for disruption mitigation can lead
to melting of the ~500 m2 Tm beryllium wall.
• Q=10 plasma termination produces a thin melt layer (~10’s microns) that remains
molten during thermal quench ~ 0.2 - 2 ms
• Because the layer melt simultaneously with the dissipation of plasma Wdia, the layer
always sees a radial inward, mobilizing JxB force resulting from induced eddy
currents by the expulsion of the diamagnetic toroidal flux.
• Surface tension acts as a stabilizing force, but for nominal tile size is insufficient to
balance JxB (~ 1mm surface “dimples” may work)
• As a result, 10’s of kg of Be will likely “splash” onto other main-wall surfaces
and divertor.
Disruption mitigation: sequence determined
by impurity “radiation instability”
• Impurities injected (gas, pellets) to dissipate
Wth by uniform radiation
• But Erad is predicted to arrive as a “burst”,
rather than uniformly in time
– Ionization instability when T < 200 eV.
– Not very dependent on mixing mechanism.
• Verified by fast bolometry on DIII-D.
Example: ITER Q=10, neon injection,
Prad spatial peaking factor =1.5 --> 50 µm thick Be melt layer over 300 m2
Energy &IonizationBalance
Be surfaceheating fromPrad
ImpurityInjectionRequired forRE control
Thermal quench
Sudden expulsion of toroidal diamagnetic flux leads to
inductive poloidal eddy currents in surrounding Be tiles that
produce radially inward JxB forces
• Relevant parameters for ITER Q=10
with Be first wall tiles.
Jdia
B
plasma
p
Inductivee.m.f.
plasma
Jtile x B
2 cm
B
2.5
cm
2.5 cm
1.5 V sToroidal Flux swing(DINA / Sugihara)
= (dtile )2 µo 4 ms >> melt
Low Be resistivity--> long skin time
= 4 10 8 m
Wall segments: ~ 106 Ltile,eddy < 10-8 H
Itile Nseg L> 100 A a
Jtile x Btor > 200 g
Prad
meltlayer
Stabilizing forces against jxB= p
Surface tension & Toroidal eddy current
• Surface tension resists change in shape of film.
• Approximate stabilizing body force by pinning film
near eddy current “corners”
• For d~50 microns, aS.T. ~ 30 g
• Stabilization can be improved by intentional
“dimpling” of original surface (like a golf ball) to
enhance capillary effect.
Eddycurrent
aJxBJtile x Btor
aS.T . dtile film
Surface tension
• The sudden inward radial movement of plasma at thermal
quench can induce toroidal eddy currents that produce radial
forces from IT x Bp
– Forces tend to be weaker since Bp ~ 10% BT
– Biggest stabilizing effect at inner midplane, de-stable at outer.
– More modeling required on poloidal dependence.
Bp
JT
Timing of Be melting and solidification critical to melt-
layer stability:
Scope Prad uniformity & Beta-Loss
1.E-06
1.E-05
1.E-04
1.E-03
1.E-02
0 0.0002 0.0004 0.0006 0.0008
Layer d
isp
lacem
en
t in
tau
_m
elt
(m
)
0.0E+00
5.0E+03
1.0E+04
1.5E+04
2.0E+04
2.5E+04
3.0E+04
0 0.0002 0.0004 0.0006 0.0008
Layer a
ccele
rati
on
(m
/s^
2)
0.0E+00
5.0E-04
1.0E-03
1.5E-03
2.0E-03
2.5E-03
0 0.0002 0.0004 0.0006 0.0008
t of final Beta collapse (s)
Melt
du
rati
on
(s)
Prad peaking
1.25 1.5
2.0 2.5
Neontermination
Melt duration increaseswith local Erad
Acceleration ismaximized whensurface “just melts”(max. surface current)
Radial displacement istypically larger than filmdepth in all conditionsUNSTABLE
Timing of Be melting and solidification critical to melt-
layer stability:
Scope Prad uniformity & Beta-Loss
Prad peaking
1.25 1.5
2.0 2.5
Neontermination
Melt thickness increasesAs Beta-drop becomesfaster
10’s of kg of molten Beare mobilized
Rdimple ~ mm stabilizesjxB movement bycapillary effect.
0.E+00
1.E-03
0 0.0002 0.0004 0.0006 0.0008
Dim
ple
siz
e
(m
)
0
10
20
30
40
50
60
0 0.0002 0.0004 0.0006 0.0008
Mo
lten
mass (
kg
)
0.E+00
5.E-05
1.E-04
0 0.0002 0.0004 0.0006 0.0008
Layer t
hic
kn
ess (
m)
ITER Summary
• Disruptions are an inevitable consequence of an experimental
tokamak.
• Controlling the timing and consequences of disruptions will be
an integral part of the operational availability and success of
ITER
– Controlled energy dissipation may help control Tritium inventory.
– Wall component viability.
– Reliable plasma breakdown and current ramp.
– Scientific aggressiveness in exploring burning plasmas.
“Shock-capturing” code developed for gas
flow down long pipes
Solves Euler’s equations for a compressible fluid.
Normalized Length along tube
Friction-free benchmark to exact solution
Friction factor dependent onReynold’s # & pipe roughness…but uncertain for transient flow
w =1
2fu u
DFriction force /wf = friction factor
Code
Data
Benchmarked “shock-capturing” code used for gas
flow down long pipes: ORNL data
f=0.07 provides good fitto measured exit pressure vs. time (inverted pressure signal)
Total gas flow rate~ 4x105 Torr-L/s ~ 1025/s
Data
Code
Gas valve pulse time (ms)S.K. Combs, et al .