Implementation of a long leg X-point target divertor in the ARC fusion pilot plant

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1 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 JP Tuesday, 2 pm

2 ARC A compact high-field tokamak ITER power levels (P fusion = 525 MW) in a JET sized (R 0 = 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 Design. Vol. 100, p Plant Lifetime 9 years 1 of 12

3 ARC A compact high-field tokamak ITER power levels (P fusion = 525 MW) in a JET sized (R 0 = 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 Design. Vol. 100, p Plant Lifetime 9 years 1 of 12

4 ARC A compact high-field tokamak ITER power levels (P fusion = 525 MW) in a JET sized (R 0 = 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 Design. Vol. 100, p Plant Lifetime 9 years 1 of 12

5 ARC A compact high-field tokamak ITER power levels (P fusion = 525 MW) in a JET sized (R 0 = 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 Design. Vol. 100, p Plant Lifetime 9 years 1 of 12

6 ARC A compact high-field tokamak ITER power levels (P fusion = 525 MW) in a JET sized (R 0 = 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 the 9 full power years of the plant lifetime. Key Design Parmeter Fusion Power Major Radius Toroidal Magnetic Field Value 525 MW 3.3 m 9.2 T Sorbom. B.N., et al. (2015). Fusion Engineering and Design. Vol. 100, p Plant Lifetime 9 years 1 of 12

7 ARC A compact high-field tokamak ITER power levels (P fusion = 525 MW) in a JET sized (R 0 = 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 simplified shield and divertor the coolant. geometry. Vacuum vessel designed to be replaced every 1-2 years during the 9 full power years of the plant lifetime. However, the initial design for ARC had an intentionally Key Design Parmeter Fusion Power Major Radius Toroidal Magnetic Field Value 525 MW 3.3 m 9.2 T Sorbom, B.N., et al. (2015). Fusion Engineering and Design. Vol. 100, p Plant Lifetime 9 years 1 of 12

8 Outline Demountable TF coils and the FLiBe immersion blanket enable: MCNP simulations were performed for the full 3D vacuum vessel geometry 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

9 Outline Demountable TF coils and the FLiBe immersion blanket enable: MCNP simulations were performed for the full 3D vacuum vessel geometry 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

10 Outline Demountable TF coils and the FLiBe immersion blanket enable: MCNP simulations were performed for the full 3D vacuum vessel geometry 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

11 Original ARC magnetic topology with simplified divertor 3 of 12

12 Original ARC magnetic topology with simplified divertor Original ARC divertor geometry was intentionally over simplified. 3 of 12

13 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 divertor LH LH ARC ACT2 C-Mod 8T ADX 8T 1 LaBombard, B., et al. (2015) Nuclear Fusion. Vol. 55, No. 5. B [T] 6 4 KSTAR EAST SST-1 DIII-D LH LH ACT1 ITER C-Mod * 5.4T JET AUG ADX 6.5T Maximum possible Psol B/R from device Psol B/R at L-H power threshold JT-60SA 2 TCV Original ITER Q DT =10 NSTX-U operation point world tokamaks * MAST q // ~ Psol B/R [MW-T/m] q [GW/m 2 ] LH 4 of 12

14 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 divertor LH LH ARC ACT2 C-Mod 8T ADX 8T 1 LaBombard, B., et al. (2015) Nuclear Fusion. Vol. 55, No. 5. B [T] 6 4 KSTAR EAST SST-1 DIII-D LH LH ACT1 ITER C-Mod * 5.4T JET AUG ADX 6.5T Maximum possible Psol B/R from device Psol B/R at L-H power threshold JT-60SA 2 TCV Original ITER Q DT =10 NSTX-U operation point world tokamaks * MAST q // ~ Psol B/R [MW-T/m] q [GW/m 2 ] LH 4 of 12

15 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 divertor 1. Data from Alcator C-Mod, H-Mode, 0.8 MA λ q ~ 1 mm Double null geometry allows for the power sharing between outer strike points and reduces heat flux to the inner strike point. 1 1 LaBombard, B., et al. (2015) Nuclear Fusion. Vol. 55, No Brunner, D., et al. (in progress) Nuclear Fusion. Brunner, D. et al. (2016) APS DPP, San Jose. 4 of 12

16 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 divertor 1. 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 window 3. 1 LaBombard, B., et al. (2015) Nuclear Fusion. Vol. 55, No Brunner, D., et al. (in progress) Nuclear Fusion. Brunner, D. et al. (2016) APS DPP, San Jose. 3 Umansky, M., et al. (2017), Physics of Plasmas. Vol of 12

17 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

18 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 divertor 1 May allow for stable, detached operation 2. 1 LaBombard, B., et al. (2015), Nuclear Fusion. Vol. 55, No Umansky, M., et al. (2017), Physics of Plasmas. Vol of 12

19 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 divertor 1 May allow for stable, detached operation 2. Internal PF coils made possible by demountable TF coil design 3. 1 LaBombard, B., et al. (2015), Nuclear Fusion. Vol. 55, No Umansky, M., et al. (2017), Physics of Plasmas. Vol Mangiarotti, F.J., et al. (2015), IEEE Transactions on Applied Superconductivity. Vol. 25, Issue 3. 5 of 12

20 Reduced coil current-turns and size Simple coil set involving 3 divertor coils. 6 of 12

21 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 currentturns in ITER PF). 6 of 12

22 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 currentturns in ITER PF). Coils sized to critical current densities of 350 A/mm 2 (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

23 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% of currentturns in ITER PF). Coils sized to critical current densities of 350 A/mm 2 (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. All while maintaining: TF coil geometry Tritium breeding ratio (TBR) TF and PF coil lifetimes 6 of 12

24 PF and TF coil lifetimes greater than 9 FPY ARC has a plant lifetime of 9 full power years (FPY) set by neutrons at the inner leg of the TF remains unchanged. Energetic Neutron flux (>100keV) 1.6E15 9.2E12 5.3E10 3.1E8 1.8E6 n/cm^2*s 7 of 12

25 PF and TF coil lifetimes greater than 9 FPY ARC has a plant lifetime of 9 full power years (FPY) set by neutrons at the inner leg of the TF remains unchanged. The PF coils achieved similar coil lifetime requirements with the addition of a solid neutron shield layer at the edge of the FLiBe tank FPY 11.3 FPY Energetic Neutron flux (>100keV) 1.6E15 9.2E12 5.3E10 3.1E8 1.8E6 n/cm^2*s 7 of 12

26 PF and TF coil lifetimes greater than 9 FPY ARC has a plant lifetime of 9 full power years (FPY) set by neutrons at the inner leg of the TF remains unchanged. The PF coils achieved similar coil lifetime requirements with the addition of a solid neutron shield layer at the edge of the FLiBe tank. Lifetime estimate based on data established for NB 3 Sn ( n/cm 2, for neutron energies > 100 kev). This is conservative as HTS is expected to have higher thresholds FPY 11.3 FPY Energetic Neutron flux (>100keV) 1.6E15 9.2E12 5.3E10 3.1E8 1.8E6 n/cm^2*s 1 Bromberg, L., et al. (2001). Fusion Engineering and Design, Vol. 54, p167 7 of 12

27 Tritium breeding ratio maintained greater than unity 5 No loss of TBR due to the large volume of breeding material that is now taken up by the divertor. D-T Plasma First wall FLiBe in coolant channels External FLiBe tank Tritium produced per source neutron of 12

28 Tritium breeding ratio maintained greater than unity 5 No loss of TBR due to the large volume of breeding material that is now taken up by the divertor. With FLiBe flowing in the cooling channels of the vacuum vessel where fast neutron dominate the spectrum. D-T Plasma First wall FLiBe in coolant channels External FLiBe tank Tritium produced per source neutron of 12

29 Tritium breeding ratio maintained greater than unity 5 No loss of TBR due to the large volume of breeding material that is now taken up by the divertor. With FLiBe flowing in the cooling channels of the vacuum vessel where fast neutron dominate the spectrum. It is optimally located for tritium generation. Thus maintaining a TBR of D-T Plasma First wall FLiBe in coolant channels External FLiBe tank Tritium produced per source neutron of 12

30 Reduced neutron damage in divertor due to leg geometry Neutron damage in the divertor region is significantly reduced due to extended leg. Divertor Region dpa/yr He/dpa ~ Midplane dpa/yr He/dpa ~ of 12

31 Reduced neutron damage in divertor due to leg geometry Neutron damage in the divertor region is significantly reduced due to extended leg. Softening of the neutron spectrum for divertor components of the vacuum vessel further reduce He production. ~10 2 reduction in the magnitude of the neutron spectrum Reduced fast neutron population Divertor Region dpa/yr He/dpa ~ Midplane dpa/yr He/dpa ~ of 12

32 The ARC divertor seperates and resolves key challenges Plasma erosion and high heat flux Stable detachment across a wide power window 1 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 simulations 2 have peak heat fluxes of ~6 MW/m 2. 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 replaced of 12

33 The ARC divertor seperates and resolves key challenges Plasma erosion and high heat flux Stable detachment across a wide power window 1 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 simulations 2 have peak heat fluxes of ~6 MW/m 2. 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 replaced 3. P sol = 88 MW 0.5% Neon Super-X case λ q ~0.6 mm 1 Umansky, M., et al. (2017), Physics of Plasmas. Vol Wigram, M., et al. (2017), Plasma Edge Theory Conference, Marseilli, France. 10 of 12

34 The ARC divertor seperates and resolves key challenges Plasma erosion and high heat flux Stable detachment across a wide power window 1 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 simulations 2 have peak heat fluxes of ~6 MW/m 2. 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 replaced 3. P sol = 88 MW 0.5% Neon Super-X case λ q ~0.6 mm 1 Umansky, M., et al. (2017), Physics of Plasmas. Vol Wigram, M., et al. (2017), Plasma Edge Theory Conference, Marseilli, France. 3 Sorbom, B.N., et al. (2015). Fusion Engineering and Design. Vol. 100, p of 12

35 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 reactor 1 : 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. 1 Brunner, D., et al. (2017) Nuclear Fusion. Vol. 57, No.8. 2 Umansky, M., et al. (2017), Physics of Plasmas. Vol of 12

36 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 reactor 1 : 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. X-point Target Ionization front location P in Increasing power to the divertor A robust passively stable detached divertor is the key. P in P in 1 Brunner, D., et al. (2017) Nuclear Fusion. Vol. 57, No.8. 2 Umansky, M., et al. (2017), Physics of Plasmas. Vol of 12

37 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 reactor 1 : 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. X-point Target Ionization front location P in Increasing power to the divertor A robust passively stable detached divertor is the key. Focus on adjusting detachment front location over manageable timescales (~1 sec). Reliant only on neutron-tolerant diagnostics such as microwave reflectometry/interferometry system. P in P in 1 Brunner, D., et al. (2017) Nuclear Fusion. Vol. 57, No.8. 2 Umansky, M., et al. (2017), Physics of Plasmas. Vol of 12

38 Conclusion Demountable TF coils and the FLiBe immersion blanket enable: MCNP simulations were performed for the full 3D vacuum vessel geometry 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 JP Tuesday, 2 pm 12 of 12

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