Developing and Qualifying Materials and Components for Next Fusion Energy Step Steve Zinkle Governor’s Chair, University of Tennessee and Oak Ridge National Laboratory With contributions from Nasr Ghoniem (UCLA) and Lance Snead (SUNY-StonyBrook) Town Hall Meeting Symposium on Fusion Engineering Ponte Vedra, FL June 4, 2019

Materials R&D and deployment is historically a long process Phases in commercialization process Commercialization of New Materials for a Global Economy, Nacional Academies Press, 1993, p. 15 M. Boren et al., McKinsey & Co. Thomas Eager, Technology Review, no. 2 (1995) 42

Fusion materials challenges and opportunities Increasing opportunities for leveraging broader mater. sci. community Challenges Plasma facing components Will tungsten work? Tritium containment and online extraction/fuel reprocessing Nonstructural materials lifetime in a DT fusion environment Plasma diagnostics (optical fibers, electrical insulators, etc.) Plasma heating feedthrough insulators Next generation magnet systems (insulation, HTC superconductors) Ceramic breeders Structural materials T2 sequestration in radiation-induced cavities Is there a viable option beyond 5 MW-yr/m2? (50 dpa)

3 High-Priority Materials R&D Challenges Is there a viable divertor & first wall PFC solution for DEMO/FNSF? Is tungsten armor at high wall temperatures viable? Do innovative divertor approaches (e.g., Snowflake, Super-X, or liquid walls) need to be developed and demonstrated? Can a suitable structural material be developed for DEMO? What is the impact of fusion-relevant transmutant H and He on neutron fluence and operating temperature limits for fusion structural materials? Is the current mainstream approach for designing radiation resistance in materials (high density of nanoscale precipitates) incompatible with fusion tritium safety objectives due to tritium trapping considerations? Can recent advanced manufacturing methods such as 3D templating and additive manufacturing be utilized to fabricate high performance blanket structures at moderate cost that still retain sufficient radiation damage resistance? What range of tritium partial pressures are viable in fusion coolants, considering tritium permeation and trapping in piping and structures? What level of tritium can be tolerated in the heat exchanger primary coolant, and how efficiently can tritium be removed from continuously processed hot coolants? PbLi, water coolant might not be viable options based on tritium considerations. S.J. Zinkle, A. Möslang, T. Muroga and H. Tanigawa, Nucl. Fusion 53 (2013) 104024

Initially ductile W-Cu laminates rapidly embrittle during irradiation at 400-800oC Vertical Target Dome KIT/ORNL collaboration

Creep rupture behavior for TMT vs. conventional 9Cr steels Thermo-mechanical treatment (TMT) 9Cr steels designed using computational thermodynamics 50-100% improvement in creep rupture strength for newly designed reduced activation steels TMT RAFMs are competitive with 4th gen FM steel creep strengths Predicted significant improvement in radiation resistance as well due to high precipitate density S.J. Zinkle et al., Nucl. Fusion 57 (2017) 092005 6

Next-generation (TMT, ODS) steels Effect of Initial Sink Strength on the Radiation Hardening of Ferritic/martensitic Steels Current steels Next-generation (TMT, ODS) steels Dramatic reduction in radiation hardening occurs when average spacing between defect cluster nuclei (dislocation loops, etc.) is much greater than average spacing between defect sinks Nloop-1/3 >> Stot-1/2 or equivalently, Stot >> Srad defects Needs further confirmation; generally would not be expected to be valid for FCC or high Z BCC metals due to in-cascade formation of large sessile defect clusters Note: SiC vacancy effective sink strength is >1e17/m2 S.J. Zinkle and L.L. Snead, Ann Rev. Mat. Res., 44 (2014) 241; S.J. Zinkle et al., Nucl. Fusion 57 (2017) 092005

Materials-tritium issues require additional investigation Identification of a robust, efficient and economic method for extraction of tritium from high temperature coolants Large number of potential tritium blanket systems is both advantageous and a hindrance Current materials science strategies to develop radiation-resistant materials may (or may not) lead to dramatically enhanced tritium retention in the fusion blanket Fission power reactors (typical annual T2 discharges of 100-800 Ci/GWe; ~10% of production) are drawing increasing scrutiny >70% of US reactor sites (>50% in last 10 years) have reported T2 groundwater contamination levels exceeding EPA safe drinking water limits* A 1 GWe fusion plant will produce ~109 Ci/yr; typical assumed releases are ~0.3 to 1x105Ci/yr (<0.01% of production) Nanoscale cavity formation may lead to significant trapping of hydrogen isotopes in the blanket (and FW/divertor) structure Note for tritium 1 g=9650 Ci *https://www.nrc.gov/docs/ML1723/ML17236A511.pdf (Sept. 2017)

H retention increases dramatically in the presence of cavity formation => Fusion may need to avoid operation at conditions that produce fine-scale cavities in structural materials 3 to 5x increase in retained hydrogen when cavities are present, even with 2-3x reduction in neutron dose 500-700 appm H (few cavities) 1700-3700 appm H (rad.-induced cavities present) Retained H level is ~100x higher than expected from Sievert’s law solubilities Measured H conc. (~10dpa irrad. 316SS with cavities) Sievert’s Law (defect-free 316SS) Note: 1000 appm corresponds to 5 kg Tritium per 100 tonnes of steel; ITER site administrative limit for tritium is 700 grams (Roth 2009) Note FFTF samples in contact with Na (H solubility equipartition coefficient of 170-320 for Ni and Fe at 500C, respectively) still had retained H contents of 320-570 appm in 316 SS samples with ~0.3 to 8% cavity swelling; also note that Na H content was maintained at 1ppm H by cold trapping. I presume Sieverts law curves for Fig. 6 in Garner paper were generated by assuming 1 atm H pressure & then multiplying bulk H solubility by volume fraction of cavities (CH~d^3) Garner paper states that doses for the baffle bolt have been revised to be 19.5, 12.2 and 7.5 dpa based on retrospective dosimetry analysis. Bolt head 1 mm 320oC, 19.5 dpa Bolt shank 25 mm 343oC, 12.2 dpa Near threads 55 mm 333oC, 7.5 dpa Baffle-former bolt removed from Tihange-1 (Belgium) pressurized water reactor Type 316 austenitic stainless steel F.A. Garner et al., J. Nucl. Mater. 356 (2006) 122 Zinkle, Möslang, Muroga, Tanigawa, Nucl. Fus. 53 (2013) 104024

Notional operating temperature windows for ferritic martensitic steels in fusion reactors T2 sequestration at cavities Cavity swelling Radiation embrittlement Thermal creep Potential impact of T2 sequestration in blanket structure Traditional operating window Thermal creep Cavity swelling Radiation embrittlement Lower temperature bound for T2 sequestration assumed to be equal to the onset temperature for cavity formation T2 sequestration issues may eliminate high T2 pressure breeding blanket concepts from consideration Zinkle & Ghoniem Fus. Eng. Des. 51-52 (2000) 55; A. Hishinuma et al. J. Nucl. Mat. 258-263 (1998) 193 after Zinkle, Möslang, Muroga, Tanigawa, Nucl. Fus. 53 (2013) 104024

Material and Component Qualification Structural materials are not universally “qualified” Relevant code cases (ASME, etc.) provide format for data compilations that summarize test temperatures where robust data have been generated and provide guidance on allowable stresses High temperature design methodology needs to be developed for irradiated materials Component testing facilities are valuable to validate predicted behavior from single-effects tests and to check for potential synergistic effects Although numerous such facilities exist for nonirradiated materials, irradiated component test facilities are lacking Halden (fission fuel rod testing) is closed

Multiphysics-Multiscale Framework Global Multiphysics Global-Local Structural Analysis CFD P Elasticity Fracture mechanics (MS-FAD) P, v T Heat transfer Elasto-plasticity Crystal plasticity N.M. Ghoniem, UCLA

Steady State Plasticity Limits Depend on Model Fidelity Neuber method Post-processing step based on elastic analysis results Energy conservation principle with stress-strain curve Global-local Elasto-plasticity More accurate than Neuber method Can be extended to transient analysis Max stress ≈ 425 MPa Stress-strain curve with Neuber Hyperbola Von-Mises stress, global-local analysis [MPa] Max stress ≈ 600 MPa Max stress ≈ 450 MPa Equivalent Neuber stress [MPa] Von-Mises stress, global elastic analysis [MPa] N.M. Ghoniem, UCLA

Advanced Materials Developmnt & Synthesis Example of potential fusion materials R&D changes: Green text indicates current base program. Red indicates suggested new task areas.   Current Base Program  Funding Increment (Assumes Redirection to Design-­guided Program)  +4 M$ Increment  +8 M$ Increment    Design &  Mechanics  Conceptual Design/Systems Studies  SMI-­‐1, SMI-­‐2, SM-­‐3  Linked Materials Database Thermomechanics module Design Integration Framework  SMI-­‐1, SMI-­‐2, SMI-­‐3  High Temp. Design Methodology Coupled Fundamental PMI   Coupled Fundamental Radiation Eff. Thermomechanics module   Plasma Materials  Interactions  Limited multi-­‐variable PSI Limited PSI and Edge diagnostics  PMI-­‐1, PMI-­‐2  Modeling SOL and Divertor  Multi-­‐scale modeling Edge and PSI Expanded multi-­‐variable PSI Expanded PSI and Edge diagnostics  Single and Coupled PSI Effects Lab Dedicated Edge Diagnostics Lab Multi-­‐scale modeling Edge and PSI Expanded multi-­‐variable PSI   Structure  & Properties Evolution  Limited  W  Development Limited HHF Test Stand Develop Tungsten Irradiat. Effects Steel Development and Radiation Eff  Modeling He and Cascade Effects  MPE-­‐2  Bulk Isotopic Tailoring Experiments Limited W Development  Limited HHF Test Stand Develop Tungsten Irradiation Effects MPE-­‐1, MPE-­‐2  Functional PFM Irradiation Effects Re-‐engage IFMIF Internat. Commun. Limited HHF Test Stand Develop Tungsten Irradiation Effects Expanded He and Cascade Effects   Advanced Materials Developmnt & Synthesis  AMDS-­‐3  Framework for Materials Database   AMDS1, AMDS-­‐2, AMDS-­‐3  Fusion Materials Database Prototype Testing  Non-­‐nuclear Environmental Effect Advanced Manufacturing Technique  Advanced Materials Development  L.L. Snead et al., Initiatives white paper submitted to 2014 FESAC Strategic Planning Panel

Fusion n source Concluding Comments Multiple options are available for high performance structural materials for nuclear environments High confidence of suitability for fission neutron environments Uncertain suitability of FM steels for fusion beyond ~5 MW-yr/m2 Potential impact of tritium retention in cavities needs to be assessed (requires systems-level analysis for specific blanket concepts) Many of the critical path items for DEMO are associated with fusion materials and technology issues (PMI, etc.) Low-TRL issues can often be resolved at low-cost Alternative energy options are continuously improving Passively safe fission power plants with accident tolerant fuel that would not require public evacuation for any design-basis accident Lower-cost solar, wind (coupled with lower-cost energy storage) => Fusion energy concepts must continue to evolve to remain competitive Fusion n source

Partial list: possible APS DPP CPP fusion materials white paper topics Fusion materials research and development needs for next-step device Scope: Materials development; operating environment effects; qualification path Fusion neutron spectrum effects in materials Scope: Experimental & modeling of fusion neutron effects; fusion neutron source Component testing needs/ capability Scope: suite of experimental test/validation of components (incl. nuclear aspects) Plasma-Materials Interactions science Scope: PMI materials science, PFM/C science and engineering, research initiatives Advanced manufacturing for fusion Scope and approach: how to leverage emerging advanced manufacturing capability Tritium and fuel cycle scope: blanket technology program/tritium materials science High Temperature Superconductor evaluation, development Scope: radiation effects, manufacturing, materials science base, industrial capability Radiation-resistant diagnostics/sensors Scope: leverage fission research initiatives Chemical compatibility of material systems Scope: compatibility of proposed material systems; design corrosion resistance Joining & repair R&D Scope: Leverage recent joining and component repair technology advances 10 topics listed