Implementation of a long leg X-point target divertor in the ARC fusion pilot plant
A.Q. Kuang, N.M. Cao, A.J. Creely, C.A. Dennett, J. Hecla, H. Hoffman, M. Major, J. Ruiz Ruiz, R.A. Tinguely, E.A. Tolman D. Brunner, B. LaBombard, B.N. Sorbom, D.G. Whyte Massachusetts Institute of Technology, Cambridge, MA P. Grover, C. Laughman Mitsubishi Electric Research Laboratories, Cambridge, MA
See Dennett – JP11.00083 Tuesday, 2 pm ARC – A compact high-field tokamak
• ITER power levels (Pfusion = 525 MW) in a JET sized (R0 = 3.3 m) tokamak.
Key Design Parmeter Value Fusion Power 525 MW Major Radius 3.3 m Toroidal Magnetic Field 9.2 T
Sorbom. B.N., et al. (2015). Fusion Engineering and Plant Lifetime 9 years Design. Vol. 100, p378-405. 1 of 12 ARC – A compact high-field tokamak
• ITER power levels (Pfusion = 525 MW) in a JET sized (R0 = 3.3 m) tokamak. • High Temperature Superconductors (HTS) enable on-axis field of 9.2 Tesla.
Key Design Parmeter Value Fusion Power 525 MW Major Radius 3.3 m Toroidal Magnetic Field 9.2 T
Sorbom. B.N., et al. (2015). Fusion Engineering and Plant Lifetime 9 years Design. Vol. 100, p378-405. 1 of 12 ARC – A compact high-field tokamak
• ITER power levels (Pfusion = 525 MW) in a JET sized (R0 = 3.3 m) tokamak. • High Temperature Superconductors (HTS) enable on-axis field of 9.2 Tesla. • Toroidal field magnets designed with demountable joints.
Key Design Parmeter Value Fusion Power 525 MW Major Radius 3.3 m Toroidal Magnetic Field 9.2 T
Sorbom. B.N., et al. (2015). Fusion Engineering and Plant Lifetime 9 years Design. Vol. 100, p378-405. 1 of 12 ARC – A compact high-field tokamak
• ITER power levels (Pfusion = 525 MW) in a JET sized (R0 = 3.3 m) tokamak. • High Temperature Superconductors (HTS) enable on-axis field of 9.2 Tesla. • Toroidal field magnets designed with demountable joints. • Vacuum vessel immersed in a molten FLiBe blanket that acts as both the neutron shield and the coolant.
Key Design Parmeter Value Fusion Power 525 MW Major Radius 3.3 m Toroidal Magnetic Field 9.2 T
Sorbom. B.N., et al. (2015). Fusion Engineering and Plant Lifetime 9 years Design. Vol. 100, p378-405. 1 of 12 ARC – A compact high-field tokamak
• ITER power levels (Pfusion = 525 MW) in a JET sized (R0 = 3.3 m) tokamak. • High Temperature Superconductors (HTS) enable on-axis field of 9.2 Tesla. • Toroidal field magnets designed with demountable joints. • Vacuum vessel immersed in a molten FLiBe blanket that acts as both the neutron shield and the coolant. • Vacuum vessel designed to be replaced every 1-2 years during Key Design Parmeter Value the 9 full power years of the Fusion Power 525 MW plant lifetime. Major Radius 3.3 m Toroidal Magnetic Field 9.2 T
Sorbom. B.N., et al. (2015). Fusion Engineering and Plant Lifetime 9 years Design. Vol. 100, p378-405. 1 of 12 ARC – A compact high-field tokamak
• ITER power levels (Pfusion = 525 MW) in a JET sized (R0 = 3.3 m) tokamak. • High Temperature Superconductors (HTS) enable on-axis field of 9.2 Tesla. • Toroidal field magnetsHowever, designed the with initial demountable design forjoints. • Vacuum vessel immersedARC in hada molten an intentionally FLiBe blanket that acts as both the neutronsimplified shield anddivertorthe coolant. geometry. • Vacuum vessel designed to be replaced every 1-2 years during Key Design Parmeter Value the 9 full power years of the Fusion Power 525 MW plant lifetime. Major Radius 3.3 m Toroidal Magnetic Field 9.2 T
Sorbom, B.N., et al. (2015). Fusion Engineering and Plant Lifetime 9 years Design. Vol. 100, p378-405. 1 of 12 Outline
MCNP simulations were performed for • Demountable TF coils and the FLiBe immersion blanket the full 3D vacuum vessel geometry enable:
• Internal PF coils • Implementation of advanced divertor geometries • Maintaining core plasma volume • Shielded PF coils • Keeping tritium breeding ratio greater than unity
2 of 12 Outline
MCNP simulations were performed for • Demountable TF coils and the FLiBe immersion blanket the full 3D vacuum vessel geometry enable:
• Internal PF coils • Implementation of advanced divertor geometries • Maintaining core plasma volume • Shielded PF coils • Keeping tritium breeding ratio greater than unity
• Double-null magnetic topology with secondary X-point target divertor configuration was selected for maximum power handling capabilities.
2 of 12 Outline
MCNP simulations were performed for • Demountable TF coils and the FLiBe immersion blanket the full 3D vacuum vessel geometry enable:
• Internal PF coils • Implementation of advanced divertor geometries • Maintaining core plasma volume • Shielded PF coils • Keeping tritium breeding ratio greater than unity
• Double-null magnetic topology with secondary X-point target divertor configuration was selected for maximum power handling capabilities.
• Long leg passively stable robust divertor systems provides a means to handle and actively control the high heat exhaust in a fusion reactor.
2 of 12 Original ARC magnetic topology with simplified divertor
3 of 12 Original ARC magnetic topology with simplified divertor
• Original ARC divertor geometry was intentionally over simplified.
3 of 12 Need to select a magnetic topology that can cope with reactor relevant divertor heat fluxes
• Fusion power plants all face the same problem of having extreme heat flux levels to the divertor1.
10 LH ARC LH ACT2 8 C-Mod 8T ADX 8T
ADX 6.5T 6 LH ACT1
]
T LH ITER
[ C-Mod* 5.4T
B 4 KSTAR AUG Maximum possible EAST P B/R from device JET sol SST-1 Psol B/R at L-H 2 DIII-D JT-60SA LH power threshold TCV Original ITER QDT=10 NSTX-U world tokamaks * operation point 0 MAST 0 550 10010 15015 40040 60060 q// ~ Psol B/R [MW-T/m] 1LaBombard, B., et al. (2015) Nuclear Fusion. Vol. 55, No. 5. ≈ q [GW/m2] ∥
4 of 12 Need to select a magnetic topology that can cope with reactor relevant divertor heat fluxes
• Fusion power plants all face the same problem of having extreme heat flux levels to the divertor1.
10 LH ARC LH ACT2 8 C-Mod 8T ADX 8T
ADX 6.5T 6 LH ACT1
]
T LH ITER
[ C-Mod* 5.4T
B 4 KSTAR AUG Maximum possible EAST P B/R from device JET sol SST-1 Psol B/R at L-H 2 DIII-D JT-60SA LH power threshold TCV Original ITER QDT=10 NSTX-U world tokamaks * operation point 0 MAST 0 550 10010 15015 40040 60060 q// ~ Psol B/R [MW-T/m] 1LaBombard, B., et al. (2015) Nuclear Fusion. Vol. 55, No. 5. ≈ q [GW/m2] ∥
4 of 12 Need to select a magnetic topology that can cope with reactor relevant divertor heat fluxes
Data from Alcator C-Mod, • Fusion power plants all face the same problem of H-Mode, 0.8 MA 1 having extreme heat flux levels to the divertor . 휆푞~ 1 푚푚
• Double null geometry allows for the power sharing between outer strike points and reduces heat flux to the inner strike point.1
1LaBombard, B., et al. (2015) Nuclear Fusion. Vol. 55, No. 5. 2Brunner, D., et al. (in progress) Nuclear Fusion. Brunner, D. et al. (2016) APS DPP, San Jose. 4 of 12 Need to select a magnetic topology that can cope with reactor relevant divertor heat fluxes
• Fusion power plants all face the same problem of having extreme heat flux levels to the divertor1.
• Double null geometry allows for the power sharing between outer strike points and reduces heat flux to the inner strike point.2
• X-point target divertor geometry has been shown in simulation to have the highest detachment threshold and largest stable detachment power window3.
1LaBombard, B., et al. (2015) Nuclear Fusion. Vol. 55, No. 5. 2Brunner, D., et al. (in progress) Nuclear Fusion. Brunner, D. et al. (2016) APS DPP, San Jose. 3Umansky, M., et al. (2017), Physics of Plasmas. Vol. 24. 4 of 12 A long legged X-point divertor magnetic geometry
• Significant proportion of the magnetic volume is not being utilized due to the need for neutron shielding.
5 of 12 A long legged X-point divertor magnetic geometry
• Significant proportion of the magnetic volume is not being utilized due to the need for neutron shielding. • Double-null magnetic topology with secondary ‘X-point target’ divertor1 • May allow for stable, detached operation2.
1LaBombard, B., et al. (2015), Nuclear Fusion. Vol. 55, No. 5. 2Umansky, M., et al. (2017), Physics of Plasmas. Vol. 24. 5 of 12 A long legged X-point divertor magnetic geometry
• Significant proportion of the magnetic volume is not being utilized due to the need for neutron shielding. • Double-null magnetic topology with secondary ‘X-point target’ divertor1 • May allow for stable, detached operation2. • Internal PF coils made possible by demountable TF coil design3.
1LaBombard, B., et al. (2015), Nuclear Fusion. Vol. 55, No. 5. 2Umansky, M., et al. (2017), Physics of Plasmas. Vol. 24. 3Mangiarotti, F.J., et al. (2015), IEEE Transactions on Applied Superconductivity. Vol. 25, Issue 3. 5 of 12 Reduced coil current-turns and size
• Simple coil set involving 3 divertor coils.
6 of 12 Reduced coil current-turns and size
• Simple coil set involving 3 divertor coils.
• Reduced PF coil current-turns due to proximity to the plasma (25% of current- turns in ITER PF).
6 of 12 Reduced coil current-turns and size
• Simple coil set involving 3 divertor coils.
• Reduced PF coil current-turns due to proximity to the plasma (25% of current- turns in ITER PF).
• Coils sized to critical current densities of 350 A/mm2 (performance based of 2015 REBCO HTS data operated at 20 K and a background magnetic field of 17 T).
• HTS cable has yet to be developed, but preliminary design work suggests that 20% superconductors and 80% structure to be a conservative estimate.
• PF coils shown in figure are to scale. 6 of 12 Reduced coil current-turns and size
• Simple coil set involving 3 pull coils.
• Reduced PF coil current-turns due to proximity to the plasma (25%All while of current maintaining:- turns in ITER PF). • TF coil geometry • Coils sized to critical current densities of 350 A/mm2 (performance• basedTritium of 2015 breeding ratio (TBR) REBCO HTS data operated• atTF 20 andK and PF a coil lifetimes background magnetic field of 17 T).
• HTS cable has yet to be developed, but preliminary design work suggests that 20% superconductors and 80% structure to be a conservative estimate.
• PF coils shown in figure are to scale. 6 of 12 PF and TF coil lifetimes greater than 9 FPY
Energetic • ARC has a plant lifetime of 9 full power years Neutron flux (FPY) set by neutrons at the inner leg of the (>100keV) TF remains unchanged. 1.6E15
9.2E12
5.3E10
3.1E8
1.8E6
n/cm^2*s
7 of 12 PF and TF coil lifetimes greater than 9 FPY
12.5 FPY 11.3 FPY Energetic • ARC has a plant lifetime of 9 full power years Neutron flux (FPY) set by neutrons at the inner leg of the (>100keV) TF remains unchanged. 1.6E15 • The PF coils achieved similar coil lifetime requirements with the addition of a solid 9.2E12 neutron shield layer at the edge of the FLiBe 5.3E10 tank. 3.1E8
1.8E6
n/cm^2*s
7 of 12 PF and TF coil lifetimes greater than 9 FPY
12.5 FPY 11.3 FPY Energetic • ARC has a plant lifetime of 9 full power years Neutron flux (FPY) set by neutrons at the inner leg of the (>100keV) TF remains unchanged. 1.6E15 • The PF coils achieved similar coil lifetime requirements with the addition of a solid 9.2E12 neutron shield layer at the edge of the FLiBe 5.3E10 tank. 3.1E8
• Lifetime estimate based on data established 1.8E6 8 2 for NB3Sn (3 × 10 n/cm , for neutron energies > 100 keV). This is conservative as n/cm^2*s HTS is expected to have higher thresholds1.
1Bromberg, L., et al. (2001). Fusion Engineering and Design, Vol. 54, p167 7 of 12 Tritium breeding ratio maintained greater than unity
5
D-T External Plasma FLiBe tank • No loss of TBR due to the large volume 4 of breeding material that is now taken up by the divertor. Tritium First wall 3 produced per source FLiBe in neutron coolant × 10−8 channels 2
1
0 8 of 12 Tritium breeding ratio maintained greater than unity
5
D-T External Plasma FLiBe tank • No loss of TBR due to the large volume 4 of breeding material that is now taken up by the divertor. Tritium produced • With FLiBe flowing in the cooling First wall 3 per source channels of the vacuum vessel where FLiBe in neutron fast neutron dominate the spectrum. coolant × 10−8 channels 2
1
0 8 of 12 Tritium breeding ratio maintained greater than unity
5
D-T External Plasma FLiBe tank • No loss of TBR due to the large volume 4 of breeding material that is now taken up by the divertor. Tritium produced • With FLiBe flowing in the cooling First wall 3 per source channels of the vacuum vessel where FLiBe in neutron fast neutron dominate the spectrum. coolant × 10−8 channels 2 • It is optimally located for tritium generation. Thus maintaining a TBR of 1.08. 1
0 8 of 12 Reduced neutron damage in divertor due to leg geometry
• Neutron damage in the divertor region is significantly reduced due to extended leg.
Divertor Region 4.5 – 9.0 dpa/yr He/dpa ~ 5.5 – 6.6
Midplane 16.4 – 27.7 dpa/yr He/dpa ~ 6.8 – 7.5 9 of 12 Reduced neutron damage in divertor due to leg geometry
• Neutron damage in the divertor region is significantly reduced due to extended leg. Reduced fast neutron population • Softening of the neutron spectrum for divertor components of the vacuum vessel further reduce He production. ~102 reduction in the magnitude of the neutron spectrum
Divertor Region 4.5 – 9.0 dpa/yr He/dpa ~ 5.5 – 6.6
Midplane 16.4 – 27.7 dpa/yr He/dpa ~ 6.8 – 7.5 9 of 12 The ARC divertor seperates and resolves key challenges
• Plasma erosion and high heat flux • Stable detachment across a wide power window1 reduces plasma temperature at plasma facing components and minimizes sputtering without affecting core plasma performance. • Long leg geometry spreads heat flux over a larger area. Initial simulations2 have peak heat fluxes of ~6 MW/m2. • Harsh neutron environment • Reduced neutron damage levels implies a possible separation of function between high heat flux handling and neutron damage resistant components. • Components only have to last 1-2 year before vacuum vessel is replaced3.
10 of 12 The ARC divertor seperates and resolves key challenges
• Plasma erosion and high heat flux • Stable detachment across a wide power window1 reduces plasma temperature at plasma facing components and minimizes sputtering without affecting core plasma performance. Psol = 88 MW 0.5% Neon • Long leg geometry spreads heat flux over a Super-X case 2 larger area. Initial simulations have peak 휆푞~0.6 mm heat fluxes of ~6 MW/m2. • Harsh neutron environment • Reduced neutron damage levels implies a possible separation of function between high heat flux handling and neutron damage resistant components. • Components only have to last 1-2 year before vacuum vessel is replaced3.
1Umansky, M., et al. (2017), Physics of Plasmas. Vol. 24. 2Wigram, M., et al. (2017), Plasma Edge Theory Conference, Marseilli, France. 10 of 12 The ARC divertor seperates and resolves key challenges
• Plasma erosion and high heat flux • Stable detachment across a wide power window1 reduces plasma temperature at plasma facing components and minimizes sputtering without affecting core plasma performance. Psol = 88 MW 0.5% Neon • Long leg geometry spreads heat flux over a Super-X case 2 larger area. Initial simulations have peak 휆푞~0.6 mm heat fluxes of ~6 MW/m2. • Harsh neutron environment • Reduced neutron damage levels implies a possible separation of function between high heat flux handling and neutron damage resistant components. • Components only have to last 1-2 year before vacuum vessel is replaced3.
1Umansky, M., et al. (2017), Physics of Plasmas. Vol. 24. 2Wigram, M., et al. (2017), Plasma Edge Theory Conference, Marseilli, France. 3Sorbom, B.N., et al. (2015). Fusion Engineering and Design. Vol. 100, p378-405. 10 of 12 Long leg divertors provide a means to handle and actively control high divertor heat exhaust • Present experiments use active feedback systems to control divertor detachment due to the narrow power window. • But this cannot scale to a reactor1: • Changes in heat flux can occur on < 10 ms time scales while feedback systems respond at ~1 s timescales. • Sensors used today likely would not survive in a reactor.
1Brunner, D., et al. (2017) Nuclear Fusion. Vol. 57, No.8. 2Umansky, M., et al. (2017), Physics of Plasmas. Vol. 24. 11 of 12 Long leg divertors provide a means to handle and actively control high divertor heat exhaust • Present experiments use active feedback systems to control divertor detachment due to the narrow
power window. 푃푖푛 • But this cannot scale to a reactor1: Ionization front location • Changes in heat flux can occur on < 10 ms Increasing power time scales while feedback systems respond to the divertor at ~1 s timescales. X-point • Sensors used today likely would not survive in Target a reactor. • A robust passively stable detached divertor is the ↑ 푃 ↑↑ 푃 key. 푖푛 푖푛
1Brunner, D., et al. (2017) Nuclear Fusion. Vol. 57, No.8. 2Umansky, M., et al. (2017), Physics of Plasmas. Vol. 24. 11 of 12 Long leg divertors provide a means to handle and actively control high divertor heat exhaust • Present experiments use active feedback systems to control divertor detachment due to the narrow
power window. 푃푖푛 • But this cannot scale to a reactor1: Ionization front location • Changes in heat flux can occur on < 10 ms Increasing power time scales while feedback systems respond to the divertor at ~1 s timescales. X-point • Sensors used today likely would not survive in Target a reactor. • A robust passively stable detached divertor is the ↑ 푃 ↑↑ 푃 key. 푖푛 푖푛 • Focus on adjusting detachment front location over manageable timescales (~1 sec). • Reliant only on neutron-tolerant diagnostics such as microwave reflectometry/interferometry system.
1Brunner, D., et al. (2017) Nuclear Fusion. Vol. 57, No.8. 2Umansky, M., et al. (2017), Physics of Plasmas. Vol. 24. 11 of 12 Conclusion
MCNP simulations were performed for • Demountable TF coils and the FLiBe immersion blanket the full 3D vacuum vessel geometry enable:
• Internal PF coils • Implementation of advanced divertor geometries • Maintaining core plasma volume • Shielded PF coils • Keeping tritium breeding ratio greater than unity
• Double-null magnetic topology with secondary X-point target divertor configuration was selected for maximum power handling capabilities.
• Long leg passively stable robust divertor systems provides a means to handle and actively control the high heat exhaust in a fusion reactor. See Dennett – JP11.00083 Tuesday, 2 pm 12 of 12