ARIES-CS Radial Build Definition and Nuclear System Characteristics

1. ARIES-CS Radial Build Definition and Nuclear System Characteristics L. El-Guebaly, P. Wilson, D. Henderson, T. Tautges, M. Wang, C. Martin, J. Blanchard and the ARIES Team Fusion Technology Institute UW - Madison US / Japan Workshop on Power Plant Studies January 24 - 25, 2006 UCSD

2. Nuclear Areas of Research Neutron Wall Loading Profile: – Toroidal & poloidal distribution – Peak & average values Radial Build Definition: – Dimension of all components – Optimal composition Blanket Parameters: – Dimension – TBR, enrichment, Mn – Nuclear heat load – Damage to FW – Service lifetime High-performance shielding module at min Blanket Shield Manifolds • Radiation Protection: • – Shield dimension & optimal composition • – Damage profile at shield, manifolds, VV, and magnets • – Streaming issues VV • Activation Issues: • – Activity and decay heat • – Thermal response to LOCA/LOFA events • – Radwaste management Magnet

3. Nuclear Task Involves Active Interaction with many Disciplines Design Requirements Prelim. Physics (R, a, Pf, ∆min, plasma contour, magnet CL) Blanket Concept Init. Magnet Parameters NWL Profile ( peak, average, ratio) 1-D Nuclear Analysis (∆min, TBR, Mn, damage, lifetime) Init. Divertor Parameters Radial Build Definition @ ∆min and elsewhere (Optimal dimension and composition, blanket coverage, thermal loads ) no ∆min match or insufficient breeding 3-D Neutronics (Overall TBR, Mn) Activation Assessment (Activity, decay heat, LOCA/LOFA, Radwaste classification) CAD Drawings Systems Code (R, a, Pf) Blanket Design Safety Analysis

4. Stellarators Offer Unique Engineering Features and Challenges • Minimum radial standoff at mincontrols machine size and cost. •  Well optimized radial build particularly at min • Sizable components with low shielding performance (such as blanket and He manifolds) should be avoided at min. • Could design tolerate shield-only module (no blanket) at min? Impact on TBR, overall size, and economics? • Compactness mandates all components should provide shielding function: • Blanket protects shield • Blanket and shield protect manifolds and VV • Blanket, shield, and VV protect magnets • Highly complex geometry mandates developing new approach to directly couple CAD drawings with 3-D neutronics codes. • Economics and safety constraints control design of all components from beginning.

5. ARIES-CS Requirements Guide In-vessel Components Design Overall TBR1.1 (for T self-sufficiency) Damage to Structure200 dpa - advanced FS (for structural integrity) 3% burn up - SiC Helium Production @ Manifolds and VV1 appm (for reweldability of FS) S/C Magnet (@ 4 K): Fast n fluence to Nb3Sn (En > 0.1 MeV) 1019 n/cm2 Nuclear heating 2 mW/cm3 dpa to Cu stabilizer 6x10-3 dpa Dose to electric insulator > 1011 rads Machine Lifetime40 FPY Availability85%

6. Reference Dual-cooled LiPb/FS Blanket Selected with Advanced LiPb/SiC as Backup BreederMultiplierStructureFW/Blanket ShieldVVCoolantCoolantCoolant Internal VV: Flibe Be FS Flibe Flibe H2O LiPb (backup) – SiC LiPb LiPb H2O LiPb(reference) – FS He/LiPb He H2O Li4SiO4Be FS He He H2O External VV: LiPb – FS He/LiPb He or H2O He Li – FS He/Li He He

7. min  = 0  = 60 FW Shape Varies Toroidally and Poloidally - a Challenging 3-D Modeling Problem Beginning of Field Period 3 FP R = 8.25 m Middle of Field Period

8. CAD geometry file Neutronics input file CAD based Monte Carlo Method CAD geometry engine Monte Carlo method Ray object intersection UW Developed CAD/MCNP Coupling Approach to Model ARIES-CS for Nuclear Assessment • Only viable approach for ARIES-CS 3-D neutronics modeling. • Geometry and ray tracing in CAD; radiation transport physics in MCNPX. • This unique, superior approach gained international support. • Ongoing effort to benchmark it against other approaches developed abroad (in Germany, China, and Japan). • DOE funded UW to apply it to ITER - relatively simple problem.

9. IB OB 3-D Neutron Wall Loading ProfileUsing CAD/MCNP Coupling Approach  = 0  = 60 Peak  • – R = 8.25 m design. • – Neutrons tallied in discrete bins: • – Toroidal angle divided every 7.5o. • – Vertical height divided into 0.5 m segments. • – Peak ~3 MW/m2 at OB midplane of  = 0o • – Peak to average = 1.52

10. Novel Shielding Approach Helps Achieve Compactness • Benefits: • Compact radial build at min • Small R and low Bmax • Low COE. WC Shield-only Zone (6.5% of FW area) Transition Zone (13.5% of FW area) • Challenges: • Integration of shield-only and transition zones with surrounding blanket. • Routing of coolants to shield-only zones. • Higher WC decay heat compared to FS.

11. Toroidal / Radial Cross Section(R = 8.25 m ) Vacuum Vessel 28 Magnet Gap LiPb & He Manifolds 35 28 FS-Shield 28 Back Wall 5 min = 119 cm WC-Shield-II 38 38 Blanket 54 Non-uniform Blanket 13 WC-Shield-I 17 Divertor System (10% of FW area) 8 FW 4 5 cm SOL Plasma | | Nominal Blanket/shield Zone (~80%) WC-Shield only Zone (~6.5%) Transition Region (~13.5%)

12. Blanket Shield Manifolds WC Shield-only or Transition Region VV Magnet Radial Build Satisfies Design Requirements(3 MW/m2 peak ) Permanent Components Replaceable | | | Thickness (cm) 63 28 35 10 2.2 28 31 >2 5 ≥2 Blanket/ Shield Zone SOL FS Shield Plasma 25 cm Breeding Zone-I 25 cm Breeding Zone-II 3.8 cm FW Vacuum Vessel Winding Pack Manifolds Gap + Th. Insulator Coil Case & Insulator External Structure 1.5 cm FS/He Gap 5 cm BW 1.5 cm FS/He 0.5 cm SiC Insert | | >181 cm Thickness (cm) 5 38 28 2 17 2 5 Shield Only Zone @ min Back Wall WC Shield-I (replaceable) WC Shield-II (permanent) Plasma SOL FW Vacuum Vessel Coil Case & Insulator Winding Pack Gap Gap + Th. Insulator External Structure 1.5 cm FS/He min = 119 cm |

13. Blanket Design Meets Breeding Requirement • Local TBR approaches 1.3 • 3-Danalysis confirmed 1-D local TBR estimate for full blanket coverage. • Uniform and non-uniform blankets sized to provide 1.1 overall TBR based on 1-D results combined with blanket coverage. To be confirmed with detailed 3-D model.

14. Preliminary 3-D Results Using CAD/MCNP Coupling Approach Model*1-D3-D Local TBR 1.285 1.316 ± 0.61% Energy multiplication (Mn) 1.14 1.143 ± 0.49% Peak dpa rate (dpa/FPY) 4039.4 ± 4.58% FW/B lifetime (FPY) 55.08 ± 4.58% Nuclear heating (MW): FW 156 145.03 ±1.33% Blanket 1572 1585.03 ±1.52% Back wall 13 9.75 ± 6.45% Shield 71 62.94 ± 2.73% Manifolds 1819.16 ± 5.49% Total 1830 1821.9 ± 0.49% Future 3-D analysis will include blanket variation, divertor system and penetrations to confirm 1.1 overall TBR and Mn. ____________________________ * Homogenized components. Uniform blanket everywhere. No divertor. No penetrations. No gaps.

15. He Tube (32 cm ID) Blanket Shield Manifolds WC Shield-only or Transition Region VV Magnet He Access Tubes Raise Streaming Concern | | Thickness (cm) 63 28 35 10 2.2 28 30 5 ≥2 ? FS Shield 5 cm BW Manifolds Blanket/ Shield Zone SOL Plasma 25 cm Breeding Zone-I 25 cm Breeding Zone-II 3.8 cm FW Vacuum Vessel Winding Pack Gap + Th. Insulator Coil Case & Insulator External Structure 1.5 cm FS/He Local Shield 1.5 cm FS/He 0.5 cm SiC Insert | | >181 cm Thickness (cm) 5 38 28 2 17 5 Shield Only Zone @ min Back Wall WC Shield-I (replaceable) WC Shield-II (permanent) Plasma SOL FW Vacuum Vessel Coil Case & Insulator Winding Pack Gap Gap + Th. Insulator External Structure 1.5 cm FS/He min = 119 cm |

16. Shield Manifolds Back Wall Plasma FW/Blanket Magnet VV Local Shield He Access Tube Local Shield Helps Solve Neutron Streaming Problem Hot spots at VV and magnet No Local Shield Ongoing 3-D analysis will optimize dimension of local shield with Tube

17. Key Design Parameters for Economic Analysis LiPb/FS/He LiPb/SiC (reference) (backup) min 1.19 1.14 Overall TBR 1.1 1.1 Energy Multiplication (Mn) 1.14 1.1 Thermal Efficiency (th) 40-45%* 55-63%* FW Lifetime (FPY) 5 6 System Availability ~85% ~85% __________ * Depending on peak . Integral system analysis will assess impact of min, Mn, and th on COE

18. Activation Assessment • Main concerns: • – Decay heat of FS and WC components. • – Thermal response of blanket/shield during LOCA/LOFA. • – Waste classification: • - Low or high level waste? • Any cleared metal? • – Radwaste stream.

19. Decay Heat WC decay heat dominates at 3 h after shutdown

20. 740 C temp limit 740 C temp limit 1 mo 1 d 1 s 1 m 1 h 1 mo 1 d 1 s 1 m 1 h Thermal Response duringLOCA/LOFA Event Nominal Blanket/Shield Tmax ~ 711 oC Shield-only Zone Tmax ~ 1060 oC • 3-coolant system rare LOCA/LOFA event (< 10-8/y) • Severe accident scenario: He and LiPb LOCA in all blanket/shield modules & water LOFA in VV. • For blanket, FW temperature remains below 740 oC limit. • WC shield modules may need to be replaced. More realistic accident scenario will be assessed.

21. Waste Management Approach • Options: • Disposal in LLW or HLW repositories • Recycling – reuse within nuclear facilities • Clearing – release to commercial market, if CI < 1. • Repository capacity is limited: • – Recyclingshould be top-level requirement for fusion power plants • – Transmute long-lived radioisotopes in special module • – Clear majority of activated materials to minimize waste volume.

22. ARIES-CS Generates Only Low-Level Waste WDR Replaceable Components: FW/Blanket-I 0.4 WC-Shield-I 0.9 Permanent Components: WC-Shield-II 0.3 FSShield 0.7 Vacuum Vessel 0.05 Magnet < 1 Confinement building << 0.1 LLW (WDR < 1) qualifies for near-surface disposal or, preferably, recycling

23. Majority of Waste (74%) can be Cleared from Regulatory Control In-vessel components cannot be cleared Dispose or Recycle Clear

24. Building Constituents can be Cleared 1-4 y after Plant Decommissioning Building composition: 15% Mild Steel, 85% Concrete

25. Stellarators Generate Large Radwaste Compared to Tokamaks not compacted, no replacements Means to reduce ARIES-CS radwaste are being pursued (more compact machine with less coil support and bucking structures)

26. Well Optimized Radial Build Contributed to Compactness of ARIES-CS m ARIES-ST Spherical Torus 3.2 m 8 Stellarators | | 6 2005 2000 2000 1987 1996 1982 4 ARIES-AT Tokamak 5.2 m ARIES-CS 7 m 8.25 m HSR-G 18 m ASRA-6C 20 m UWTOR-M 24 m 2 SPPS 14 m FFHR-J 10 m 10 15 20 25 5 0 Average Major Radius (m) Over past 25 y, stellarator major radius more than halved by advanced physics and technology, dropping from 24 m for UWTOR-M to 7-8 m for ARIES-CS, approaching R of advanced tokamaks.

27. Concluding Remarks • Novel shielding approach developed for ARIES-CS. Ongoing study is assessing benefits and addressing challenges. • CAD/MCNP coupling approach developed specifically for ARIES-CS 3-D neutronics modeling. • Combination of shield-only zones and non-uniform blanket presents best option for ARIES-CS. • Proposed radial build satisfies design requirements. • No major activation problems identified for ARIES-CS. • At present, ongoing ARIES-CS study is examining more compact design (R < 8 m) that needs further assessment.

28. ARIES-CS Publications • L. El-Guebaly, R. Raffray, S. Malang, J. Lyon, and L.P. Ku, “Benefits of Radial Build Minimization and Requirements Imposed on ARIES-CS Stellarator Design,” Fusion Science & Technology, 47, No. 3, 432 (2005). • L. El-Guebaly, P. Wilson, and D. Paige, “Initial Activation Assessment of ARIES Compact Stellarator Power Plant,” Fusion Science & Technology, 47, No. 3, 440 (2005). • M. Wang, D. Henderson, T. Tautges, L. El-Guebaly, and X. Wang, “Three -Dimensional Modeling of Complex Fusion Devices Using CAD-MCNP Interface,” Fusion Science & Technology, 47, No. 4, 1079 (2005). • L. El-Guebaly, P. Wilson, and D. Paige, “Status of US, EU, and IAEA Clearance Standards and Estimates of Fusion Radwaste Classifications,” University of Wisconsin Fusion Technology Institute Report, UWFDM-1231 (December 2004). • L. El-Guebaly, P. Wilson, and D. Paige, “Evolution of Clearance Standards and Implications for Radwaste Management of Fusion Power Plants,” Journal of Fusion Science & Technology, 49, 62-73 (2006). • M. Zucchetti, L. El-Guebaly, R. Forrest, T. Marshall, N. Taylor, K. Tobita, “The Feasibility of Recycling and Clearance of Active Materials from Fusion Power Plants,” Submitted to ICFRM-12 conference at Santa Barbara (Dec 4-9, 2005). • L. El-Guebaly, R. Forrest, T. Marshall, N. Taylor, K. Tobita, M. Zucchetti, “Current Challenges Facing Recycling and Clearance of Fusion Radioactive Materials,” University of Wisconsin Fusion Technology Institute Report, UWFDM-1285. Available at: http://fti.neep.wisc.edu/pdf/fdm1285.pdf • L. El-Guebaly, “Managing Fusion High Level Waste – a Strategy for Burning the Long-Lived Products in Fusion Devices,” Fusion Engineering and Design. Also, University of Wisconsin Fusion Technology Institute Report, UWFDM-1270. Available at: http://fti.neep.wisc.edu/pdf/fdm1270.pdf • R. Raffray, L. El-Guebaly, S. Malang, and X. Wang, “Attractive Design Approaches for a Compact Stellarator Power Plant” Fusion Science & Technology, 47, No. 3, 422 (2005). • J. Lyon, L.P. Ku, P. Garabedian, L. El-Guebaly, and L. Bromberg, “Optimization of Stellarator Reactor Parameters,” Fusion Science & Technology, 47, No. 3, 4414 (2005). • R. Raffray, S. Malang, L. El-Guebaly, and X. Wang, “Ceramic Breeder Blanket for ARIES-CS,” Fusion Science & Technology, 48, No. 3, 1068 (2005).