1. 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

2. 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

3. 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)

4. 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)

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

6. 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) 6

7. 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)

8. 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 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
* (Sept. 2017)

9. 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
appm H
(few cavities)
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 for Ni and Fe at 500C, respectively) still had retained H contents of 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)

10. 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 (2000) 55;
A. Hishinuma et al. J. Nucl. Mat (1998) 193
after Zinkle, Möslang, Muroga, Tanigawa, Nucl. Fus. 53 (2013)

11. 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

12. 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

13. 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

14. 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

15. 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

16. 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