1. Magnetic Fusion Power Plants Farrokh Najmabadi, Director, Center for Energy Research Prof. of Electrical & Computer Engineering University of California, San Diego EPRI, July 19, 2011

2. Conceptual designs studies of fusion power plants are performed by the ARIES national team  National ARIES Team comprises key members from major fusion centers (universities, national laboratories, and industry).  Many studies of evolving confinement concepts and different technologies.  National ARIES Team comprises key members from major fusion centers (universities, national laboratories, and industry).  Many studies of evolving confinement concepts and different technologies.

3. Framework: Assessment Based on Attractiveness & Feasibility Periodic Input from Energy Industry Goals and Requirements Scientific & Technical Achievements Evaluation Based on Customer Attributes Attractiveness Characterization of Critical Issues Feasibility Projections and Design Options Balanced Assessment of Attractiveness & Feasibility No: Redesign R&D Needs and Development Plan Yes

4. Utility/industrial advisory committees have defined customer requirements  Fusion Power Plant Studies Utility Advisory Committee and EPRI Fusion Working Group*  Chaired by Steve Rosen & Jack Kaslow respectively,  Met biannually in 1993-1995.  Helped define goals and top-level requirements.  Had a major impact on safety/licensing as well as configuration/maintenance approach.  Input as members of review committee for individual designs.  Fusion Power Plant Studies Utility Advisory Committee and EPRI Fusion Working Group*  Chaired by Steve Rosen & Jack Kaslow respectively,  Met biannually in 1993-1995.  Helped define goals and top-level requirements.  Had a major impact on safety/licensing as well as configuration/maintenance approach.  Input as members of review committee for individual designs. * See http::/aries.ucsd.edu/ARIES/DOCS/UAC/ for membership and meeting minutes.

5. Goals & Top-Level Requirements for Fusion Power Plants Were Developed in Consultation with US Industry  Have an economically competitive life-cycle cost of electricity  Gain Public acceptance by having excellent safety and environmental characteristics  No disturbance of public’s day-to-day activities  No local or global atmospheric impact  No need for evacuation plan  No high-level waste  Ease of licensing  Reliable, available, and stable as an electrical power source  Have operational reliability and high availability  Closed, on-site fuel cycle  High fuel availability  Available in a range of unit sizes Fusion physics & technology Low-activation material Fusion Fuel Cycle

6. Framework: Assessment Based on Attractiveness & Feasibility Periodic Input from Energy Industry Goals and Requirements Scientific & Technical Achievements Evaluation Based on Customer Attributes Attractiveness Characterization of Critical Issues Feasibility Projections and Design Options Balanced Assessment of Attractiveness & Feasibility No: Redesign R&D Needs and Development Plan Yes

7. Detailed analyses are necessary to understand trade-offs Plasma analysis Engineering Design Self-consistent point design for a fusion power plant System Analysis and Trade-offs

8. Detailed analyses are necessary to understand trade-offs Plasma analysis Engineering Design Self-consistent point design for a fusion power plant System Analysis and Trade-offs  High accuracy equilibria;  Large ideal MHD database over profiles, shape and aspect ratio;  RWM stable with wall/rotation or wall/feedback control;  NTM stable with LHCD;  Bootstrap current consistency using advanced bootstrap models;  External current drive;  Vertically stable and controllable with modest power (reactive);  Rough kinetic profile consistency with RS /ITB experiments, as well GLF23 transport code;  Modest core radiation with radiative SOL/divertor;  Accessible fueling;  No ripple losses;  0-D consistent startup;  High accuracy equilibria;  Large ideal MHD database over profiles, shape and aspect ratio;  RWM stable with wall/rotation or wall/feedback control;  NTM stable with LHCD;  Bootstrap current consistency using advanced bootstrap models;  External current drive;  Vertically stable and controllable with modest power (reactive);  Rough kinetic profile consistency with RS /ITB experiments, as well GLF23 transport code;  Modest core radiation with radiative SOL/divertor;  Accessible fueling;  No ripple losses;  0-D consistent startup;  Superconducting magnet design  First wall/blanket, and shield, Divertor; Current-drive systems (Launchers, transmission lines, sources),…  Configuration  Neutronics & Shielding  Thermo-fluid & thermo mechanical design  MHD effects  Tritium Breeding & management  Erosion  Off-normal events  Inventory  Waste Disposal  Safety Analysis  Maintenance  Superconducting magnet design  First wall/blanket, and shield, Divertor; Current-drive systems (Launchers, transmission lines, sources),…  Configuration  Neutronics & Shielding  Thermo-fluid & thermo mechanical design  MHD effects  Tritium Breeding & management  Erosion  Off-normal events  Inventory  Waste Disposal  Safety Analysis  Maintenance

9. DT Fusion requires a tritium-breeding blanket Plasma should be surrounded by a blanket containing Li  Through care in design, only a small fraction of neutrons are absorbed in structure and induce radioactivity*  Rad-waste depends on the choice of material: Low- activation material  For liquid coolant/breeders (e.g., Li, LiPb), most of fusion energy (carried by neutrons and n-Li reaction) is directly deposited in the coolant simplifying energy recovery  Issue: Large flux of high-energy neutrons through the first wall and blanket: Plasma should be surrounded by a blanket containing Li  Through care in design, only a small fraction of neutrons are absorbed in structure and induce radioactivity*  Rad-waste depends on the choice of material: Low- activation material  For liquid coolant/breeders (e.g., Li, LiPb), most of fusion energy (carried by neutrons and n-Li reaction) is directly deposited in the coolant simplifying energy recovery  Issue: Large flux of high-energy neutrons through the first wall and blanket: D + T  4 He (3.5 MeV) + n (14 MeV) n + 6 Li  4 He (2 MeV) + T (2.7 MeV) n T * A neutron multiplier, e.g., 7 Li, Pb, or Be, is needed to achieve tritium self-sufficiency.

10. Irradiation leads to a operating temperature window for material  Additional considerations such as He embrittlement and chemical compatibility may impose further restrictions on operating window Radiation embrittlement Thermal creep Zinkle and Ghoniem, Fusion Engr. Des. 49-50 (2000) 709  Carnot =1-T reject /T high Structural Material Operating Temperature Windows: 10-50 dpa

11. New structural material should be developed for fusion application Candidate “low-activation” structural material:  Fe-9Cr steels: builds upon 9Cr-1Mo industrial experience and materials database  9-12 Cr ODS steel is a higher-temperature option.  SiC/SiC: High risk, high performance option (early in its development path)  W alloys: High performance option for PFCs (early in its development path) Candidate “low-activation” structural material:  Fe-9Cr steels: builds upon 9Cr-1Mo industrial experience and materials database  9-12 Cr ODS steel is a higher-temperature option.  SiC/SiC: High risk, high performance option (early in its development path)  W alloys: High performance option for PFCs (early in its development path)

12. 1) Ceramic Solid Breeder Concepts (using He coolant and ferritic steel structure) Adopted from fission pebble-bed designs. Complex internal design of coolant routing to keep solid breeder within its design window. High structural content, low Li content, requires lots of Be multiplier. Low outlet temperature and low efficiency Large tritium inventory 1) Ceramic Solid Breeder Concepts (using He coolant and ferritic steel structure) Adopted from fission pebble-bed designs. Complex internal design of coolant routing to keep solid breeder within its design window. High structural content, low Li content, requires lots of Be multiplier. Low outlet temperature and low efficiency Large tritium inventory Many Blanket Concepts have been considered 2) Li (breeder and coolant) with vanadium structure Needs insulating coating for MFE (MHD effects). Special requirements to minimize threat of Li fires. Large tritium inventory in Li which can be released during an accident. 2) Li (breeder and coolant) with vanadium structure Needs insulating coating for MFE (MHD effects). Special requirements to minimize threat of Li fires. Large tritium inventory in Li which can be released during an accident.

13. 3. Dual coolant with a self-cooled PbLi zone, He-cooled RAFS structure and SiC insert  Steel First wall and partitioning walls are cooled with He.  Most of fusion neutron energy is deposited in PbLi coolant/breeder.  SiC insert separates PbLi from the walls: They reduce a) MHD effects and b) heating of the walls by LiPb  Outlet coolant temperature of ~700 o C (Max. steel temperature of ~550 o C)  Steel First wall and partitioning walls are cooled with He.  Most of fusion neutron energy is deposited in PbLi coolant/breeder.  SiC insert separates PbLi from the walls: They reduce a) MHD effects and b) heating of the walls by LiPb  Outlet coolant temperature of ~700 o C (Max. steel temperature of ~550 o C)

14. Outboard blanket & first wall 4. High-performance blanket with SiC Composite Structure and LiPb coolant  Simple, low pressure design with SiC structure and LiPb coolant and breeder.  Innovative design leads to high LiPb outlet temperature (~1,100 o C) while keeping SiC structure temperature below 1,000 o C leading to a gross thermal efficiency of ~ 59% (52% net)  Simple manufacturing technique.  Very low afterheat.  Class C waste by a wide margin.  Simple, low pressure design with SiC structure and LiPb coolant and breeder.  Innovative design leads to high LiPb outlet temperature (~1,100 o C) while keeping SiC structure temperature below 1,000 o C leading to a gross thermal efficiency of ~ 59% (52% net)  Simple manufacturing technique.  Very low afterheat.  Class C waste by a wide margin.

15. Framework: Assessment Based on Attractiveness & Feasibility Periodic Input from Energy Industry Goals and Requirements Scientific & Technical Achievements Evaluation Based on Customer Attributes Attractiveness Characterization of Critical Issues Feasibility Projections and Design Options Balanced Assessment of Attractiveness & Feasibility No: Redesign R&D Needs and Development Plan Yes

16. Configuration & Maintenance

17. ARIES-AT (tokamak) Fusion Core

18. The ARIES-AT utilizes an efficient superconducting magnet design  On-axis toroidal field:6 T  Peak field at TF coil:11.4 T  TF Structure: Caps and straps support loads without inter-coil structure;  On-axis toroidal field:6 T  Peak field at TF coil:11.4 T  TF Structure: Caps and straps support loads without inter-coil structure; Superconducting Material  Either LTC superconductor (Nb 3 Sn and NbTi) or HTC  Structural Plates with grooves for winding only the conductor. Superconducting Material  Either LTC superconductor (Nb 3 Sn and NbTi) or HTC  Structural Plates with grooves for winding only the conductor.

19. Configuration & Maintenance are important aspects of the design 1.Install 4 TF coils at a times 2.Insert ¼ of inner VV and weld 3.Complete the torus 4.Insert maintenance ports and weld to inner part of VV and each other 5.Install outer walls and dome of the cryostat 1.Install 4 TF coils at a times 2.Insert ¼ of inner VV and weld 3.Complete the torus 4.Insert maintenance ports and weld to inner part of VV and each other 5.Install outer walls and dome of the cryostat vacuum vessel Inner part: 4 pieces, Complete vessel, outer part is welded during assemblyentirely made of maintenance ports

20. Modular sector maintenance enables high availability  Full sectors removed horizontally on rails  Transport through maintenance corridors to hot cells  Estimated maintenance time < 4 weeks  Full sectors removed horizontally on rails  Transport through maintenance corridors to hot cells  Estimated maintenance time < 4 weeks ARIES-AT elevation view

21. ARIES-AT Fusion core is segmented to minimize rad-waste and optimize functions Shield Inboard FW/blanket 1 st Out-board FW/blanket 2 nd Out-board FW/blanket Stabilizing shells Divertor Blanket-2 and shield are life-time components

22. Safety, Licensing and Waste Disposal

23. After 100 years, only 10,000 Curies of radioactivity remain in the 585 tonne ARIES-RS fusion core. After 100 years, only 10,000 Curies of radioactivity remain in the 585 tonne ARIES-RS fusion core.  SiC composites lead to a very low activation and afterheat.  All components of ARIES-AT qualify for Class-C disposal under NRC and Fetter Limits. 90% of components qualify for Class-A waste.  SiC composites lead to a very low activation and afterheat.  All components of ARIES-AT qualify for Class-C disposal under NRC and Fetter Limits. 90% of components qualify for Class-A waste. Ferritic Steel Vanadium Radioactivity levels in fusion power plants are very low and decay rapidly after shutdown Level in Coal Ash

24. Safety analysis of off-normal events and accident scenarios indicate no evacuation plan is needed  Detailed accident analysis (e.g., loss of coolant, loss of flow, double break in a major coolant line) are performed:  Limited temperature excursion due to the use of low- activation material.  No evacuation plan is needed. Most of the off-site dose after an accident is due to tritium release from fusion core. Fusion core tritium inventory is ~ 1kg.  Components are designed to handle off-normal events:  Pressurization of blanket modules due internal break of He channels (Dual-cooled blanket)  Disruption forces and thermal loads  Quench of TF coils  Detailed accident analysis (e.g., loss of coolant, loss of flow, double break in a major coolant line) are performed:  Limited temperature excursion due to the use of low- activation material.  No evacuation plan is needed. Most of the off-site dose after an accident is due to tritium release from fusion core. Fusion core tritium inventory is ~ 1kg.  Components are designed to handle off-normal events:  Pressurization of blanket modules due internal break of He channels (Dual-cooled blanket)  Disruption forces and thermal loads  Quench of TF coils

25. Waste volume is modest (ARIES-AT)  1,270 m 3 of Waste is generated after 40 full-power year of operation. Coolant is reused in other power plants 29 m 3 every 4 years (component replacement), 993 m 3 at end of service  Equivalent to ~ 30 m 3 of waste per full-power operation. Effective annual waste can be reduced by increasing plant service life.  1,270 m 3 of Waste is generated after 40 full-power year of operation. Coolant is reused in other power plants 29 m 3 every 4 years (component replacement), 993 m 3 at end of service  Equivalent to ~ 30 m 3 of waste per full-power operation. Effective annual waste can be reduced by increasing plant service life.  90% of waste qualifies for Class A disposal

26. Costing

27. A cost break-down structure is used.  Costing is performed through a comprehensive cost break-down structure to component level.  Direct vendor quotes are used when available.  In the absence of vendor quotes, comparable technologies are used to cost a component.  Costing assumptions where calibrated against advanced fission and fossil plant economics.*  Costing is performed through a comprehensive cost break-down structure to component level.  Direct vendor quotes are used when available.  In the absence of vendor quotes, comparable technologies are used to cost a component.  Costing assumptions where calibrated against advanced fission and fossil plant economics.* No. Account 20Land and Land Rights 21Structures and Site Facilities 22Power Core Plant Equipment 22.01 Fusion Energy Capture and Conversion 22.01.01 First Wall and Blanket 22.01.02 Second Blanket 22.01.03 Divertor Assembly 22.01.04 High Temperature Shielding 22.01.05 Low Temperature Shielding 22.01.06 Penetration Shielding 22.02 Plasma Confinement 22.02.01 Toroidal Field Coils 22.02.02 Poloidal Field Coils 22.02.03 Feedback Coils 22.03 Plasma Formation and Sustainment 22.04 …. 22.14 23Turbine Plant Equipment 24Electric Plant Equipment 25Miscellaneous Plant Equipment 26Heat Rejection Equipment 27Special Materials 90.Direct Cost 91-98 Indirect Costs 99. Total Cost *J. Delene, Fusion Technology, 26 (1994) 1105.

28. Magnetic Fusion Power Systems are projected to be cost-competitive. Estimated Cost of Electricity (2009 c/kWh) Major radius (m)  Total Capital Cost ranges from $4B to $8B.  We are in the process of implementing Gen IV fission cost data base. This data base would lead:  Similar total Capital Cost  30% lower COE because of a lower fixed-cost rate (5.8% for Gen-IV vs 9.65% for Delene).  Total Capital Cost ranges from $4B to $8B.  We are in the process of implementing Gen IV fission cost data base. This data base would lead:  Similar total Capital Cost  30% lower COE because of a lower fixed-cost rate (5.8% for Gen-IV vs 9.65% for Delene).

29. Framework: Assessment Based on Attractiveness & Feasibility Periodic Input from Energy Industry Goals and Requirements Scientific & Technical Achievements Evaluation Based on Customer Attributes Attractiveness Characterization of Critical Issues Feasibility Projections and Design Options Balanced Assessment of Attractiveness & Feasibility No: Redesign R&D Needs and Development Plan Yes

30. Level Generic Description 1 Basic principles observed and formulated. 2 Technology concepts and/or applications formulated. 3 Analytical and experimental demonstration of critical function and/or proof of concept. 4 Component and/or bench-scale validation in a laboratory environment. 5 Component and/or breadboard validation in a relevant environment. 6 System/subsystem model or prototype demonstration in relevant environment. 7 System prototype demonstration in an operational environment. 8 Actual system completed and qualified through test and demonstration. 9 Actual system proven through successful mission operations. Technical Readiness Levels provides a basis for assessing the development strategy Increased integration Increased Fidelity of environment Basic & Applied Science Phase Validation Phase  See ARIES Web site: http://aries.ucsd.edu/aries/ (TRL Report) for detailed application of TRL to fusion systems

31. Fusion Nuclear technologies are in an early development stage  Fusion research has focused on developing a burning plasma.  Technology development has been based on the need of experiments as opposed to what is needed for a power plant.  Plasma support technologies (e..g, superconducting magnets) are at a high-level of technology readiness level.  Fusion Nuclear technologies, however, are at a low level of technology readiness level.  Material development has only focused on irradiation response of structural material due to the low level of funding.  A focused development program could raise the TRL levels of fusion nuclear technologies rapidly.  Fusion research has focused on developing a burning plasma.  Technology development has been based on the need of experiments as opposed to what is needed for a power plant.  Plasma support technologies (e..g, superconducting magnets) are at a high-level of technology readiness level.  Fusion Nuclear technologies, however, are at a low level of technology readiness level.  Material development has only focused on irradiation response of structural material due to the low level of funding.  A focused development program could raise the TRL levels of fusion nuclear technologies rapidly.

32. Example: TRLs for Plasma Facing Components Issue-Specific DescriptionFacilities 1 System studies to define tradeoffs and requirements on heat flux level, particle flux level, effects on PFC's (temperature, mass transfer). Design studies, basic research 2 PFC concepts including armor and cooling configuration explored. Critical parameters characterized. Code development, applied research 3 Data from coupon-scale heat and particle flux experiments; modeling of governing heat and mass transfer processes as demonstration of function of PFC concept. Small-scale facilities: e.g., e-beam and plasma simulators 4 Bench-scale validation of PFC concept through submodule testing in lab environment simulating heat fluxes or particle fluxes at prototypical levels over long times. Larger-scale facilities for submodule testing, High-temperature + all expected range of conditions 5 Integrated module testing of the PFC concept in an environment simulating the integration of heat fluxes and particle fluxes at prototypical levels over long times. Integrated large facility: Prototypical plasma particle flux+heat flux (e.g. an upgraded DIII-D/JET?) 6 Integrated testing of the PFC concept subsystem in an environment simulating the integration of heat fluxes and particle fluxes at prototypical levels over long times. Integrated large facility: Prototypical plasma particle flux+heat flux 7 Prototypic PFC system demonstration in a fusion machine. Fusion machine ITER (w/ prototypic divertor), CTF 8 Actual PFC system demonstration qualification in a fusion machine over long operating times. CTF 9 Actual PFC system operation to end-of-life in fusion reactor with prototypical conditions and all interfacing subsystems. DEMO

33. Example: TRLs for Plasma Facing Components Issue-Specific DescriptionFacilities 1 System studies to define tradeoffs and requirements on heat flux level, particle flux level, effects on PFC's (temperature, mass transfer). Design studies, basic research 2 PFC concepts including armor and cooling configuration explored. Critical parameters characterized. Code development, applied research 3 Data from coupon-scale heat and particle flux experiments; modeling of governing heat and mass transfer processes as demonstration of function of PFC concept. Small-scale facilities: e.g., e-beam and plasma simulators 4 Bench-scale validation of PFC concept through submodule testing in lab environment simulating heat fluxes or particle fluxes at prototypical levels over long times. Larger-scale facilities for submodule testing, High-temperature + all expected range of conditions 5 Integrated module testing of the PFC concept in an environment simulating the integration of heat fluxes and particle fluxes at prototypical levels over long times. Integrated large facility: Prototypical plasma particle flux+heat flux (e.g. an upgraded DIII-D/JET?) 6 Integrated testing of the PFC concept subsystem in an environment simulating the integration of heat fluxes and particle fluxes at prototypical levels over long times. Integrated large facility: Prototypical plasma particle flux+heat flux 7 Prototypic PFC system demonstration in a fusion machine. Fusion machine ITER (w/ prototypic divertor), CTF 8 Actual PFC system demonstration qualification in a fusion machine over long operating times. CTF 9 Actual PFC system operation to end-of-life in fusion reactor with prototypical conditions and all interfacing subsystems. DEMO Power-plant relevant high-temperature gas-cooled PFC Low-temperature water-cooled PFC

34. Application of TRL to Power Plant Systems

35. Application to power plant systems highlights early stage of fusion nuclear technology development TRL 123456789 Power management Plasma power distribution Heat and particle flux handling High temperature and power conversion Power core fabrication Power core lifetime Safety and environment Tritium control and confinement Activation product control Radioactive waste management Reliable/stable plant operations Plasma control Plant integrated control Fuel cycle control Maintenance Completed In Progress For Details See ARIES Web site: http://aries.ucsd.edu/aries/ (TRL Report) Basic & Applied Science Phase System demonstration and validation in operational environment (FNF) Demo/ 1 st power plant

36. ITER will provide substantial progress in some areas (e.g., plasma, safety) TRL 123456789 Power management Plasma power distribution Heat and particle flux handling High temperature and power conversion Power core fabrication Power core lifetime Safety and environment Tritium control and confinement Activation product control Radioactive waste management Reliable/stable plant operations Plasma control Plant integrated control Fuel cycle control Maintenance Completed In Progress ITER Absence of power-plant relevant fusion nuclear technologies severely limits ITER’s contributions in many areas. System demonstration and validation in operational environment (FNF) Demo/ 1 st power plant

37. In summary:  ITER will demonstrate “technical feasibility” of fusion power by generating copious amount of fusion power (500MW for 300s) with fusion power > 10 input power.  Tremendous progress in understanding plasmas has helped optimize plasma performance considerably.  Vision of attractive magnetic fusion power plants exists which satisy customer requirements.  Transformation of fusion into a power plant requires considerable R&D in material and fusion nuclear technologies (largely ignored or under-funded to date). This step, however, can be done in parallel with ITER  ITER will demonstrate “technical feasibility” of fusion power by generating copious amount of fusion power (500MW for 300s) with fusion power > 10 input power.  Tremendous progress in understanding plasmas has helped optimize plasma performance considerably.  Vision of attractive magnetic fusion power plants exists which satisy customer requirements.  Transformation of fusion into a power plant requires considerable R&D in material and fusion nuclear technologies (largely ignored or under-funded to date). This step, however, can be done in parallel with ITER

38. Thank You!

39. There has been substantial changes in our predications of edge plasma properties  Current expectation of much higher peak heat and particle flux on divertors:  Scrape-off layer energy e-folding length is substantially smaller.  Elms and intermittent transport  Gad-cooled W divertor designs with capability of 10-12MW/m 2 has been produced.  More work is needed to quantify the impact of the new physics predictions on power plant concepts.  Current expectation of much higher peak heat and particle flux on divertors:  Scrape-off layer energy e-folding length is substantially smaller.  Elms and intermittent transport  Gad-cooled W divertor designs with capability of 10-12MW/m 2 has been produced.  More work is needed to quantify the impact of the new physics predictions on power plant concepts. ARIES-CS T-Tube concept

40. Predicted Tritium Inventories in ARIES-CS Layout of ARIES-CS power core ARIES-CS coolant circuit schematic

41. ARIES-CS DCLL PbLi Heat Transport System (HTS) Schematic He inletHe outlet Vacuum pump Vacuum permeator Blanket Concentric pipes Heat Exchanger T2 outlet Pressure boundary (90 C) Closed Brayton Cycle PbLi (460 C) PbLi (700 C) PbLi pump Inter-coolerPre-coolerRecuperator Turbo-compressor Power turbine

42. TMAP ARIES-CS Model Schematic Concentric pipes PbLi Permeator PbLi core PbLi/He HX Non-Hartmann Gaps Hartmann Gaps First wall Second wall Rib walls Back plate Tritium cleanup system Helium pipes Shield Inter-cooler Pressure boundary Manifolds Brayton Cycle Closed Brayton CycleDCLL Blanket ARIES-CS HTS Inventories and Permeation Rates * TMAP Predictions per Sector (multiply by 6 for reactor totals) An additional 1 to 2 kg will also exist in T2 fueling and processing plant ARIES-CS relied on a high efficiency PbLi tritium extraction unit (vacuum permeator ~ 70%) and an actively cooled SS strong barrier enveloping the secondary Brayton cycle to meet safety goals Structure No ImplantationFW Implantation Inventory (g-T) Permeation into building (g/a) Inventory (g-T) Permeation into building (g/a) Blanket1.23E-01 5.05E+00 High temperature shield2.63E-03 5.31E-03 Manifold9.63E+01 2.47E+02 PbLi outlet pipe7.41E-031.68E+011.25E-023.52E+01 Pbli HTX tubes5.12E-02 1.06E+00 PbLi inlet pipe3.93E-01 7.44E+00 Helium outlet pipe3.60E-022.81E-017.55E-018.30E+00 Helium HTX tubes1.59E-035.36E-042.45E-022.30E-03 Helium inlet pipe1.01E-01 3.17E+00 Brayton cycle wall3.62E-01 3.70E-01 Permeator3.03E-02 5.75E-01 Sector total9.73E+011.71E+012.65E+024.35E+01 Release after 99% efficient cleanup1.71E-01 4.35E-01 575 Reactor Total 1590 ~1 ~2.6 Non-flow BC for conservatism