ARIES Engineering Activities

1. ARIES Engineering Activities UC San Diego UW Madison GIT ARIES PPPL INL M. S. Tillack for the ARIES Team Boeing GA System Studies Peer Review 29 August 2013

2. Outline • Part I. Solutions for high heat flux and transients • He-cooled W divertor design and analysis • Experimental validation of heat transfer in impinging jet designs • Thermomechanics evaluation, including plasticity, creep and fracture • Transient (ELM and disruption) responses • Part II. System integration and design innovations • ACT1 power core concept • Nuclear analysis: Neutronics, tritium breeding, shielding, activation • Liquid metal blanket analysis and manifolding improvements • Thermal conversion cycle • Vacuum vessel with enhanced safety and damage attributes • Response to off-normal events (LOCA/LOFA) • ACT2 power core concept

3. The ARIES Team proposed and led the development of He-cooled W-alloy divertors • ARIES-ST was the first power plant study to adopt this concept. • The motivation was to eliminate duplex structures in the HHF zone. • Exploit attractive properties of W: high coolant temperature (800 ˚C) for high hth, high heat flux (up to 15 MW/m2) to meet requirements. • Now internationally regarded as a top candidate for power plants. • Alloy development is expected to improve properties (e.g. ductility). plate T-tube finger plate/finger M. S. Tillack, X. R. Wang, J. Pulsifer, I. Sviatoslavsky and the ARIES Team, "ARIES-ST Plasma-Facing Component Design and Analysis, Fusion Eng Design 49-50 (2000) 363. M. S. Tillack, A. R. Raffray, X. R. Wang, S. Malang, S. Abdel-Khalik, M. Yoda and D. Youchison "Recent US Activities on Advanced He-Cooled W-Alloy Divertor Concepts for Fusion Power Plants," Fusion Engineering and Design, 86 (2011) pp. 71-98.

4. Impinging jet designs have been examined in detail, and shown to provide very high heat flux capability • Characteristic number / dimension:Plate: 104 / 100 cm T-tube: 105 / 10 cm Finger: 106 / 1.5 cm • ACT1 adopted a combined plate/finger design, using fingers only where needed • Total pumping power optimized by tailoring jets X. R. Wang, S. Malang, M. S. Tillack, "High Performance Divertor Target Concept for a Power Plant: a Combination of Plate and Finger Concepts,” Fusion Sci Tech 60 (1) 218-222 (2011). X. R. Wang, S. Malang, M. S. Tillack, J. Burke, "Recent Improvements of the Helium-Cooled W-based Divertor for Fusion Power Plants,” Fusion Eng Design 87 (5-6) 732-736 (2012).

5. ARIES supported a series of gas-cooled divertor experiments to validate our analysis • Objectives: • Evaluate thermal performance (max. heat flux, pumping power) of leading helium-cooled divertor designs under prototypical conditions • Provide design correlations for system codes • Approach: • Extrapolate dynamically similar experiments with different coolants to He at prototypical conditions • Provide parametric design curvesto estimate how changes in operating conditions affect divertor thermal performance • Georgia Tech He loop (new) • Supplies up to 10 g/s He at 10 MPa and up to 400 ˚C • Accommodates test sections with pressure drops > 0.7 MPa

6. Results demonstrate that multi-jet designs can meet design requirements for heat flux >10 MW/m2 Multi-jet test section • Prototypical conditions: 7.3 g/s mass flowrate, max. heat flux = 12 MW/m2 on pressure (structural) boundary (11 MW/m2 on hexagonal tile) for avg. boundary temp. Ts = 1200°C. • Pumping power (as fraction of incident thermal power based on max. heat flux)  = 12%. • HEMJ has the best thermal performance among designs studied here. W-1.0% La2O3 M. Yoda, S. Abdel-Khalik, D. Sadowski, B. Mills and J. Rader, “Experimental Evaluation of the Thermal-Hydraulics of Helium-Cooled Divertors,” Fusion Sci Tech, to appear. B. Mills, J. Rader, D. L. Sadowski, M. Yoda, and S. I. Abdel-Khalik, "An Experimental Study of the Effects of Thermal Conductivity Ratio in Helium-Cooled Divertor Modules," Fusion Sci Tech, 64(3) 670-674 (2013). J. Rader, B. Mills, D. L. Sadowski, M. Yoda, and S. I. Abdel-Khalik, "Verification of Thermal Performance Predictions of Prototypical Multi-Jet Impingement Helium-Cooled Divertor Module,” Fusion Sci Tech 64(3) 282-287 (2013).

7. 3D elastic-plastic analysis is now routinely employed, demonstrating higher performance in plastic regime • Stress relaxation allows higher performance vs. 3Sm criterion. We typically use cumulative plastic strain < 50% uniform elongation. • Birth-to-death analysis considers fabrication steps, cycling (warm and cold shutdown). • “Souped-up” (100 GB) PC can model large structures; analysis is performed by students (undergrad and grad) as well as staff. First wall fabrication cycle braze temper Elastic analysis, 15 MW/m2 Elastic-plastic analysis, 15 MW/m2 First wall operating cycle with warm shutdown Safety factor <<1 Allowable plastic strain for pure W is 0.8% @270ºC and 1.0% @1200ºC. Peak in simulations is only 0.13%. emax=0.13%

8. An external transition joint accommodates differential expansion between the divertor and coolant pipes • Place transition outside of high heat flux zone • Ta-2.5W interlayer with intermediate CTE • Careful design is needed to avoid ratchetting Original design suffered ratchetting in the Ta interlayer subjected to cold shutdown cycles Final design exhibits shakedown followed by no ratchetting D. Navaei, X. R. Wang, M. S. Tillack, S. Malang, "Elastic-plastic analysis of the steel-to-tungsten transition joint for a high performance divertor," Fusion Eng Design 88 (5), 2013, pp. 361–367. D. Navaei, X. R. Wang, M. S. Tillack, S. Malang, "Elastic-plastic analysis of transition joints for high performance divertor target plate," Fusion Sci Tech 60 (1) July 2011, 233-237.

9. Fracture mechanics analysis of the finger divertor shows crack growth may be less serious than feared 1. Base of grooves 2. Coolant channel surface Crack-Free Stress State J. P. Blanchard, C. J. Martin, M. S. Tillack, and X. R. Wang, "Ratchetting Models for Fusion Component Design," Fusion Sci. Tech 60 (1) July 2011, 313-317. J. P. Blanchard and C. J. Martin, "Fracture in All-Tungsten Divertors for ARIES,” Fusion Sci Tech 64(3) 435-439 (2013).

10. Results show that the maximum stress intensity is below the (unirradiated) fracture toughness Crack on coolant surface (hot) 2c a Crack in notch (at shutdown) B. Gludovatz, S. Wurster, A. Hoffmann, R. Pippan, “Fracture Toughness of Polycrystalline Tungsten Alloys.” 17th Plansee Seminar 2009, Vol. 1.

11. Transient loads and responses were evaluated; similar to ITER, worst-case scenarios are severe • First comprehensive description of power plant transients. • Steady state loads estimated using JET/ITER scaling laws. • ELM energy (10-20% of pedestal energy), energy partition (IB, OB, divertor and FW) and time scale (a few particle parallel transit times) are based on experimental observations, scaled to our plasma. • Disruptions (thermal quench, current quench, runaways) also scaled from experiments. Time-dependent temperatures near divertor surface from a single Type-I ELM (1128 MW/m2) Melting threshold for tungsten C. E. Kessel, M. S. Tillack and J. P. Blanchard "The Evaluation of the Heat Loading from Steady, Transient, and Off-Normal Conditions in ARIES Power Plants, Fusion Sci Tech64(3) 440-448 (2013). James P. Blanchard and Carl Martin, “Thermomechanical Analysis for an All-Tungsten ARIES Divertor,” Fusion Sci Tech, to appear (2014).

12. Outline • Part I. Solutions for high heat flux and transients • He-cooled W divertor design and analysis • Experimental validation of heat transfer in impinging jet designs • Thermomechanics evaluation, including plasticity, creep and fracture • Transient (ELM and disruption) responses • Part II. System integration and design innovations • ACT1 power core concept • Nuclear analysis: Neutronics, tritium breeding, shielding, activation • Liquid metal blanket analysis and manifolding improvements • Thermal conversion cycle • Vacuum vessel with enhanced safety and damage attributes • Response to off-normal events (LOCA/LOFA) • ACT2 power core concept

13. The ACT1 power core evolved from ARIES-AT Similarities Differences • SiC composite breeding blanket with PbLi at To~1000 C • Brayton power cycle, h~58% • He-cooled W divertor • Steel structural ring (vs. SiC) • Simplified blanket coolant paths • Simplified vacuum vessel with external LT shield X. R. Wang, M. S. Tillack, S. Malang, F. Najmabadi, "Overall Power Core Configuration and System Integration for the ARIES-ACT Fusion Power Plant,” Fusion Sci Tech 64 (3) 455-459 (2013). X.R. Wang, M.S. Tillack, C. Koehly, S. Malang, F. Najmabadi and the ARIES Team, “ARIES-ACT1 System Configuration, Assembly and Maintenance, Fusion Sci Tech, to appear.

14. Nuclear Analysis Covers Three Closely-Related Tasks • Neutronics: • Poloidal distribution of neutron wall loading and surface heat flux • Tritium breeding ratio (TBR) for T-self sufficiency • Radial and poloidal nuclear heating distribution (for thermal hydraulic analysis) • Nuclear energy multiplication (for power balance) • Radiation damage to structural materials (dpa, He production, etc.) • Service lifetimes of all components based on neutron dose. • Shielding: • Shielding of permanent components • Radial and vertical build definition (for physics code, CAD drawings, and systems code analysis) • Neutron streaming through penetrations and assembly gaps • Bioshield specifications. • Activation: • Radioactive product inventory (for safety, environmental, and licensing assessments). • Environmental impact of fusion: radwaste classification, recycling, clearance • Decay heat (for thermal response of components during accidents) • Biological dose (for maintenance crew, workers and public).

15. State-of-the-Art 3-D nuclear analysis was used • In 2000’s, UW-FTI neutronics team developed a new approach to allow fully accurate 3-D nuclear analysis of complex devices by coupling CAD geometry directly with 3-D MCNP neutronics code (DAGMC). • Main reasons for developing sophisticated 3-D nuclear analysis: • Model details of blanket reduce uncertainty in 3-D modeling • High fidelity in TBR results assure T self-sufficiency • Lower the breeding requirement thinner blanket and lower T inventory. • New approach applied to several ARIES designs: • ARIES-CS (first application; very complex geometry) • ARIES-ACT with DCLL Blanket (detailed TBR analysis) • ARIES-ACT with SiC Blanket (detailed TBR analysis).

16. Our new approach to tritium breeding analysis helps us better understand the sensitivities Add walls to blankets Add cooling channels Add Stabilizing Shells Vary blanket thickness Add SiC FCI Vary Li Enrichment 1. 1D infinite cylinder 2. 3D toroidal model 3. assembly gaps 4. FW added 5. steel box 6. grid plates added 7. SiC FCI’s 8. W shells 9. OB penetrations 10. Li6 enrichment L. El-Guebaly, A. Jaber, L. Mynsberge, “State-of-the-Art 3-D Assessment of Elements Degrading TBR of ARIES ACT-SiC Blanket,” Fusion Sci Tech 64 (3) 582-586 (2013). L. El-Guebaly, A. Jaber, and S. Malang, "State-of-the-Art 3-D Assessment of Elements Degrading the TBR of the ARIES DCLL Blanket,” Fusion Sci Tech 61 (4) (May 2012).

17. Radial and vertical builds for ARIES-ACT1 Radiation limits and requirements

18. 3D MHD was considered for the first time in design; simplified coolant flow paths were developed 180˚ bends ARIES-AT Manifolding and distribution C. Koehly, M. S. Tillack, X. R. Wang, Farrokh Najmabadi, S. Malang, “Flow distribution systems for liquid metal cooled blankets,” Proc. 25th Symp on Fusion Eng, San Francisco, 2013. Expansion/ contraction

19. Liquid metal heat transfer was analyzed to establish blanket operating conditions • Stagnation caused by curved ducts was analyzed and included in design • 2D MHD heat transfer in central and annular ducts M. S. Tillack, X. R. Wang, S. Malang, F. Najmabadi, "ARIES-ACT1 Power Core Engineering,” Fusion Science and Technology 64 (3) 427-434 (2013).

20. Primary stresses were shown to meet requirements First wall • Bonding the inner and outer tubes substantially reduces pressure stress.

21. A high-efficiency (58%) power cycle is key to system performance • Like DCLL, need to match all of the coolant temperatures. • Result depends on inlet temperature as well as outlet; >57% could be achieved with 550˚C inlet. • Results are shown for hrecuperator==96%, hturbine=92%

22. A new vacuum vessel concept was developed • Elimination of water, 350˚C operating temperature minimizes T inventory. • Low-activation 3Cr-3WV bainitic steel • Lower activation than 316SS • No post-weld heat treatment • Ample volume to accommodate a fullHe LOCA with <1 atm overpressure. • No need to support other components. • 10 cm total thickness, including embedded He cooling channels. • Low disruption forces. H. H. Toudeshki, F. Najmabadi, X. R. Wang, "Vacuum Vessel Analysis and Design for the ARIES-ACT Fusion Power Plant,” Fusion Science and Technology 64 (3) 675-679 (2013). "Design Challenges and Activation Concerns for ARIES Vacuum Vessel,” L. El-Guebaly, T. Huhn, A. Rowcliffe, S. Malang, Fusion Science and Technology 64 (3) 449-454 (2013).

23. The passive safety features of ARIES-ACT1 have been verified in a LOFA and multiple LOCA scenarios • A new coupling method (coupled iterative single-fluid runs) was developed to accurately model two-fluid problems (PbLi and water), a first for MELCOR. Natural convection in the water-cooled shield removes decay heat during a long-term station blackout. Temperatures of outer structures peak at 2-3 days and do not threaten the structural integrity of components. Outboard temperatures during LOFA (˚C) P. W. Humrickhouse and B. J. Merrill, “MELCOR accident analysis for ARIES-ACT,” Fusion Science and Technology 64 (2) 340-344 (2013). MELCOR representation of ARIES-ACT1 power core

24. Activation is assessed with the ultimate goal of recycling and clearance • Candidate W Alloys: • W-W composites (100% W) • W-Re (74 % W, 26 % Re) • W-La2O3 (99% W, 1% La2O3) • W-TiC (98.9 % W, 1.1% TiC) • W-VM (W doped with 70 ppm K) • W-K (W doped with 40 ppm K) • W-Ta (95 % W, 5% Ta). 0.7 MW/m2 3.8 FPY 3 MWy/m2 All materials could be handled and recycled within days after replacement using advanced remote handling equipment (capable of handling fusion components with dose > 10,000 Sv/h Avoid using W-Re for generating high-level waste Divertordoes not qualify for clearance A. Robinson, L. El-Guebaly and D. Henderson, "Activation and Radiation Damage Characteristics of W-Based Divertor of ARIES Power Plants,” Fusion Science and Technology 60 (1) 715-719 (2011). M. Zucchetti, L. Di Pace, L. El-Guebaly, B.N. Kolbasov, V. Massaut, R. Pampin, and P. Wilson, "The Back-End of Fusion Materials Cycle: Recycling and Clearance, Avoiding Disposal,” Fusion Science and Technology 56 (2) 781-788 (2009).

25. ACT2 is our first fully integrated study of the DCLL blanket in a tokamak (work in progress) • ARIES-ST was the first integrated study to adopt a dual-cooled PbLi blanket. Also chosen for ARIES-CS. • PbLi/SiC/He/steel provides a pathway from near-term to advanced concepts • It has become the de facto US reference concept 500˚C allowable PbLi/steel interface 550˚C allowable structure Low surface heat flux and low MHD Dp give ACT2 a comfortable design margin.

26. ARIES-ACT issues are well known, and the subject of ongoing R&D • Characterization of steady and transient surface heat loads. • MHD effects on flow and heat transfer. • Fabrication, assembly and joining of complex structures made of SiC composites, tungsten alloys, and low activation ferritic steels. • Failure modes and rates: Mechanical behavior of steel, W and SiC structures, including fracture mechanics, creep/fatigue, and irradiation effects. • Upper and lower temperature limits of W alloys and advanced ferritic steels. • Fluence lifetime of components under anticipated loading conditions. • Erosion of plasma-facing components. • Tritium containment and control.