1. Low Aspect Ratio FNSF Mission, Features, Readiness Issues – a discussion for feedback
Martin Peng, ORNL
FNST Meeting
August 18-20, 2009
UCLA, Los Angeles, U.S.A.

2. What does CTF do? [Abdou et al.]
“Scientific Exploration” to discover and investigate unexpected physical properties, and improve.
“Engineering Verification” to compare designs and select winners.
“Reliability Growth” to establish fusion nuclear component database for fusion DEMO (NRC required).
Physical properties important to FNST can have time constants up to 106 s

3. D-D Tokamak Confinement
Where should CTF be located in fusion neutron fluence rate and plasma burn duration? PP
U.S.
Demo
 IFMIF (20-55 dpa/yr) 
MTS (20 dpa/yr) 
SNS (5 dpa/yr) 
E.U.
Demo
1.0
0.1
0.01
CTF-III
CTF-II
Fusion Neutron Fluence Rate (MW-yr/m2-yr)
ITER: Burning Plasma
CTF-I
105
104
103
102
106 10
107
108
Plasma Burn Duration (s)
D-D Tokamak Confinement

4. Tungsten erosion redeposition lifetime D-D Tokamak Confinement
Where should physical properties of interest to Plasma-Material Interface and Fusion Power in a full fusion nuclear environment be located in the same parameter space? Examples - updated. PP
U.S.
Demo
 IFMIF (20-55 dpa/yr) 
MTS (20 dpa/yr) 
SNS (5 dpa/yr) 
E.U.
Demo
1.0
0.1
0.01
CTF-III
CTF-II
Fusion Neutron Fluence Rate (MW-yr/m2-yr)
T build-up hot FS
and tungsten
T build-up in
solid breeders
ITER: Burning Plasma
(TBM)
T build-up in
PbLi Breeder
Tungsten erosion redeposition lifetime
CTF-I
T Recycling
with wall bulk;
T measurement
105
104
103
102
106 10
107
108
Fusion Burn Duration (s)
D-D Tokamak Confinement

5. Fusion Burn Duration (s)
FNSF tests physical properties of interest to Plasma-Material Interface and Fusion Power, to establish the understanding needed to begin engineering development on a FNSF-CTF PP
U.S.
Demo
 IFMIF (20-55 dpa/yr) 
MTS (20 dpa/yr) 
SNS (5 dpa/yr) 
E.U.
Demo
1.0
0.1
0.01
CTF-III
CTF-II
Fusion Neutron Fluence Rate (MW-yr/m2-yr)
T permeation
thru hot RAFS
with barriers
FNSF
Fusion Nuclear Science
R&D using increasing burn
durations & fluence rate
T extraction in
solid breeders
ITER: Burning Plasma
(TBM)
T build-up in
PbLi Breeder
CTF-I
T Recycling
with wall bulk;
T measurement
105
104
103
102
106 10
107
108
Fusion Burn Duration (s)
D-D Tokamak Confinement

6. bN  0.75x no-wall limit; HH  1.25; q95  2xlimit; JavgTF  4 kA/cm2
Conservative parameters are available for R0 = 1.3m, A = 1.7 plasmas to address the FNSF mission
bN  0.75x no-wall limit; HH  1.25;
q95  2xlimit; JavgTF  4 kA/cm2
FNSF-CTF
Performance level
JET-DD
JET-DT
2xJET
3xJET
Tests enabled
divertor
FNS
FNST
WL (MW/m2)
0.005
0.25
1.0
2.0
Current, Ip (MA)
4.2
6.7
8.4
Field, BT (T)
2.7
2.9
3.6
Safety factor, q95
12.7
8.6
Toroidal beta, bT (%)
4.4
10.1
10.8
Normal beta, bN
2.1
3.3
3.5
Avg density, ne (1020/m3)
0.54
1.1
1.5
Avg ion Ti (keV)
7.7
7.6
10.2
11.8
Avg electron Te (keV)
4.3
5.7
7.2
BS current fraction
0.45
0.47
0.50
0.53
NBI H&CD power (MW) 26 22 44 61
Fusion power (MW)
0.4 19 76
152
NBI energy (kV)
120
235
330
Aspect ratio and plasma aggressiveness have high leverage on tradeoffs in readiness, performance, cost, and risk.

7. D-D Tokamak Confinement
FNSF-CTF R&D can be staged: JET-level DD-only for divertor-PMI verification (requiring full remote handling); JET-level DT for FNS; 2xJET for CTF-II; 3xJET for CTF-III PP
U.S.
Demo
 IFMIF (20-55 dpa/yr) 
MTS (20 dpa/yr) 
SNS (5 dpa/yr) 
E.U.
Demo
1.0
0.1
0.01
CTF-III
CTF-II
FNS R&D
Fusion Neutron Fluence Rate (MW-yr/m2-yr)
3xJET-DT
T permeation
thru hot RAFS
with barriers
T extraction in
solid breeders
2xJET-DT
ITER: Burning Plasma
(TBM)
T build-up in
PbLi Breeder
CTF-I
JET-DT
T Recycling
with wall bulk;
T measurement
105
104
103
102
106 10
107
108
Fusion Burn Duration (s)
JET-DD
D-D Tokamak Confinement

8. Divertor testing goals at JET-level DD operation
Required capabilities
Peak heat fluxes  10 MW/m2
Pulse lengths progressively up to  106 s
Remote handling, efficient replacement
Extremely rare plasma-induced disruptions
Disruption mitigation, e.g. “killer pellets”
Plasma material interface diagnostics
Tritium diagnostics
Testing goals – measure, understand, improve, and verify
Life time limiting mechanisms – divertors, plasma facing surfaces
Tritium transport mechanisms – permeation, distribution at very low levels
Can ITER tungsten divertor be extended up to 106 s?
FNSF-CTF divertors

9. PMIF: Can a Linear Test Stand Provide High Plasma Heat Fluxes, Ti and Te for up to 106 s?
Magnetic Mirror with RF Heating
Test Target Transfer Cask
Helicon
PMI Test Station
PMI Analysis & Preparation Stations
US-J PSI July 2009

10. Feasibility for high heat flux in test stand, required to enable FNSF-CTF, is being answered
High recycle target (Tt < 10 eV): Tu is limited to eV, requiring q|| ~ MW/m2 and nu ~4-8x1019/m3
Detachment (Tt < 2-3 eV): Tu is limited to 20 eV, requiring q|| ~ MW/m2 and same nu range
Can Helicon plasma at ~5 eV be heated to ~35 eV?
Can Helicon reach such high nu (data so far  2-3x1019/m3)?
High Recycle
Detachment
Target Plasma T(eV)
High Recycle
Detachment
Upstream Plasma T(eV)

11. Divertor testing goals at JET-level DT operation with WL = 0.25 MW/m2
Additional testing goals – measure, understand, improve, and verify
Degradation due to neutron damage – decreased thermal conductivity, material strength
Enhancements to life time limiting mechanisms – divertors, plasma facing surfaces
Enhancements in tritium transport mechanisms – permeation, distribution
Can ITER tungsten divertor still be extended up to 106 s?
Allowable stress to limit radiation induced creep
FNSF-CTF divertors
Note that RAFS is not expected to suffer He-embrittlement for up to dpa [ref: Zinkle et al, TOFE 2002 review paper]
Divertor and plasma facing surfaces improved to handle WL = 1-2 MW/m2 would enable the start of CTF-II and CTF-III R&D

12. Upper and lower breeding blankets
Breeding blanket testing goals at JET-level DT operation with WL = 0.25 MW/m2
Upper and lower breeding blankets
Required capabilities
Peak plasma heat fluxes  1 MW/m2
Pulse lengths progressively up to  106 s
Remote handling, efficient replacement
Extremely rare plasma-induced disruptions
Disruption mitigation, e.g. “killer pellets”
Blanket tritium diagnostics
Blanket performance instrumentation
Testing goals – measure, understand, improve, and verify
Tritium breeding and power extraction
Blanket tritium transport mechanisms – permeation, distribution at power relevant levels
Can ITER TBM designs be extended up to 106 s?

13. Mid-plane Test Blanket Modules
Test Blanket Module testing goals at JET-level DT operation with WL = 0.25 MW/m2
Required capabilities
Peak plasma heat fluxes  2 MW/m2
Pulse lengths progressively  106 s
Remote handling, efficient replacement
Extremely rare plasma-induced disruptions
Disruption mitigation, e.g. “killer pellets”
Blanket performance instrumentation
Tritium diagnostics
Testing goals – measure, understand, improve, and verify
Power extraction and tritium breeding
Test blanket tritium transport mechanisms – permeation, distribution at power relevant levels
Can ITER TBM designs be extended up to 106 s?
Mid-plane Test Blanket Modules

14. Single-turn toroidal field coil center post
Single-turn TF center post testing goals at JET-level DT operation with WL = 0.25 MW/m2
Required capabilities
JavgTF  3 kA/cm2
ITFC = 17.7 MA
Lifetime  1 dpa and up to 10 dpa
Remote handling, efficient replacement
Testing goals – measure, understand, improve, and verify
Allowable stresses of hardened and embrittled GlidCop to limit crack growth
Proper combination of load bearing design (compression, locking against twisting force)
Low-A power plant study estimated  3.7 kA/cm2 [ref: Song et al, FED 2005]
Single-turn toroidal field coil center post
GlidCop is highly resistant to He-embrittlement for > 100 dpa of fission neutron damage [ref: Cu alloy review, DOE/ER-0313/16, 1994]

15. Additional FNSF chamber components will require FNS R&D at 0.25 MW/m2
Positive-ion NBI (120 kV)
Pulse lengths progressively up to  106 s
Remote handling, efficient replacement
Tritium diagnostics
Testing goals – measure, understand, improve, and verify
Life time limiting mechanisms – source, cryo, power handling, etc.
Can ITER NBI technologies be applied and extended to very long pulses
Plasma diagnostic port assemblies
Can ITER diagnostic systems be extended progressively up to  106 s?
FNSF-CTF

16. FNS R&D can be enabled using JET-level DT plasmas, providing WL = 0
FNS R&D can be enabled using JET-level DT plasmas, providing WL = 0.25 MW/m2 with Q~0.8
FNSF is the scientific R&D stage of CTF, to improve component designs before engineering and technology testing for Demo
Very conservative plasmas are possible for FNSF-CTF eventually to deliver WL = 2 MW/m2, without plasma-induced disruptions for  106 s
Remaining new R&D: startup, electron transport, high heat flux, TFC
JET-level DD plasma adequate for divertor verification and improvements
A very long pulse linear high plasma heat flux test stand being assessed
JET-level DT plasma would enable full FNS R&D for PMI and Harnessing Fusion Power
Divertors, plasma facing surfaces
Blankets – breeding and mid-plane test modules
TFC center post
Conclusions applicable to normal aspect ratio designs – as a first stage of FDF
US-J PSI July 2009

17. We will continue to work closely with the FNST stakeholders to
Describe the scientific mission
Develop major components of a National FNS Program
Example:
What are the scientific questions? – tritium transport through divertor
Why are these important? – at low burn fraction, can only lose <1% for fuel self-sufficiency
Basic physical models and uncertainties? – tritium implantation on plate, permeation through RAFS, barriers, GlidCop, crossing into coolant
Experimental conditions needed – test stands, subsystems  test components in full fusion nuclear environment
Measurements needed and techniques
Improvements in understanding, or reductions in uncertainty required to determine component design for CTF-II and CTF-III testing
US-J PSI July 2009