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

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

4. 4 Stellarators Offer Unique Engineering Features and Challenges Minimum radial standoff at  min controls 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. 5 ARIES-CS Requirements Guide In-vessel Components Design Overall TBR 1.1 (for T self-sufficiency) Damage to Structure 200 dpa - advanced FS (for structural integrity) 3% burn up - SiC Helium Production @ Manifolds and VV 1 appm (for reweldability of FS) S/C Magnet (@ 4 K): Fast n fluence to Nb 3 Sn (E n > 0.1 MeV)10 19 n/cm 2 Nuclear heating 2mW/cm 3 dpa to Cu stabilizer6x10 -3 dpa Dose to electric insulator > 10 11 rads Machine Lifetime 40 FPY Availability 85%

6. 6 Reference Dual-cooled LiPb/FS Blanket Selected with Advanced LiPb/SiC as Backup BreederMultiplierStructureFW/BlanketShieldVV CoolantCoolantCoolant Internal VV: FlibeBeFSFlibeFlibeH 2 O LiPb (backup) –SiCLiPbLiPbH 2 O LiPb (reference)–FSHe/LiPbHe H 2 O Li 4 SiO 4 BeFSHeHe H 2 O External VV: LiPb–FSHe/LiPbHe or H 2 O He Li–FSHe/LiHeHe

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

8. 8 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. CAD geometry engine Monte Carlo method Ray object intersection CAD based Monte Carlo Method CAD geometry file Neutronics input file

9. 9 3-D Neutron Wall Loading Profile Using CAD/MCNP Coupling Approach IBOB –R = 8.25 m design. –Neutrons tallied in discrete bins: – Toroidal angle divided every 7.5 o. – Vertical height divided into 0.5 m segments. –Peak  ~3 MW/m 2 at OB midplane of  = 0 o –Peak to average  = 1.52  = 0  = 60 Peak 

10. 10 Novel Shielding Approach Helps Achieve Compactness Benefits: –Compact radial build at  min –Small R and low B max –Low COE. 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. WC Shield-only Zone (6.5% of FW area) Transition Zone (13.5% of FW area)

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

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

13. 13 Blanket Design Meets Breeding Requirement Local TBR approaches 1.3 3-D analysis 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. 14 Preliminary 3-D Results Using CAD/MCNP Coupling Approach Model * 1-D3-D Local TBR1.2851.316± 0.61% Energy multiplication (M n )1.14 1.143± 0.49% Peak dpa rate (dpa/FPY)40 39.4± 4.58% FW/B lifetime (FPY)5 5.08± 4.58% Nuclear heating (MW): FW156 145.03±1.33% Blanket1572 1585.03±1.52% Back wall13 9.75± 6.45% Shield71 62.94± 2.73% Manifolds18 19.16± 5.49% Total1830 1821.9± 0.49% ____________________________ * Homogenized components. Uniform blanket everywhere. No divertor. No penetrations. No gaps. Future 3-D analysis will include blanket variation, divertor system and penetrations to confirm 1.1 overall TBR and M n.

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

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

17. 17 Key Design Parameters for Economic Analysis Integral system analysis will assess impact of  min, M n, and  th on COE LiPb/FS/He LiPb/SiC (reference)(backup)  min 1.19 1.14 Overall TBR1.1 1.1 Energy Multiplication (M n )1.14 1.1 Thermal Efficiency (  th )40-45% * 55-63% * FW Lifetime (FPY)5 6 System Availability~85%~85% __________ * Depending on peak .

18. 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. 19 Decay Heat WC decay heat dominates at 3 h after shutdown

20. 20 Thermal Response during LOCA/LOFA Event 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 o C limit. WC shield modules may need to be replaced. More realistic accident scenario will be assessed. Nominal Blanket/Shield T max ~ 711 o C Shield-only Zone T max ~ 1060 o C 740 C temp limit 1 s1 m1 h 1 d 1 mo 740 C temp limit 1 s1 m1 h 1 d 1 mo

21. 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: – Recycling should 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. 22 ARIES-CS Generates Only Low-Level Waste WDR Replaceable Components: FW/Blanket-I0.4 WC-Shield-I0.9 Permanent Components: WC-Shield-II 0.3 FS Shield0.7 Vacuum Vessel0.05 Magnet< 1 Confinement building<< 0.1 LLW (WDR < 1) qualifies for near-surface disposal or, preferably, recycling

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

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

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

26. 26 Well Optimized Radial Build Contributed to Compactness of ARIES-CS 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. UWTOR-M 24 m 0 5 10 152025 2 4 6 8 m Average Major Radius (m) ASRA-6C 20 m HSR-G 18 m SPPS 14 m FFHR-J 10 m ARIES-CS ARIES-ST Spherical Torus 3.2 m ARIES-AT Tokamak 5.2 m Stellarators | | 7 m 8.25 m 1982 19872000 1996 20002005

27. 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. 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).