Systems-level comments on liquid PFMs from a materials science perspective

1. Systems-level comments on liquid PFMs from a materials science perspective Steven J Zinkle University of Tennessee, Knoxville, TN USA Oak Ridge National Laboratory, Oak Ridge, TN USA Fusion materials planning meeting University of Tennessee, Knoxville, TN July 29, 2016

2. General comments • Li vs. SnLi vs. Ga vs. molten salt vs. ??? • What is the appropriate prioritization to identify most promising reference liquid concept? (including both plasma confinement and reactor system perspectives) • Optimized choice depends on fusion issue: • Heat removal; • material sputtering/redeposition; • plasma stability/performance; • safety and operations (T2 sequestration) • What may be good for plasma performance may be bad for overall fusion energy systems performance • Tritium sequestration/inventory and permeation in structure • High partial pressure vs. low partial pressure tritium transport from divertor to tritium extraction apparatus (high vs low T2 solubility) • Can have a large impact on required tritium breeding ratio for device (particularly if part/all of first wall also uses liquid walls) • importance dependent on liquid PFC design: Low recycling –vs- High recycling • Can we afford to throw away the heat deposited in the divertor? (COE vs ops)

3. Tritium removal from liquid PFCs • Potential advantage of liquid lithium divertor concept is the ability to accumulate most of exhausted tritium in lithium and remove tritium continuously • Development of detritiation system is key for liquid lithium divertor • Tritium inventory limit (e.g., 1 kg=356 PBq for ITER) • Detritiation performance (detritiation factor, purification limit concentration, and purification time) determines tritium and lithium inventory. • Several detritiation methods have been proposed • Molten salt extraction process [5] • Yttrium hot trap for IFMIF/EVEDA [6,7] • Molten salt liquid‐liquid extraction and RbCl containing salts (LLNL LDRD) • Direct LiT electrolysis (SRNL LDRD) • These methods are (in principal) technically feasible for tritium removal, but (significant) further R&D is required to demonstrate economical feasibility and meet required detritiation performance for liquid lithium divertor concept. • Typically energy-intensive processes and slow purification time • Often requires use of hazardous materials (e.g. halides, hydrofluoric acid) [5] V.A.Maroniet.al., Nucl. Technol. 25 (1974) 83 [6] S. Fukadaet.al., Fus. Sci. Technol. 54 (2008) 117. [7] Y. Edaoet.al., Fus. Eng. Des. 85 (2010) 53. M. Shimada (INL) | Fusion Material Workshop | July 27, 2016 |

4. Li-LiH equilibrium phase diagram High solubility of T2 in liquid Li is a potential advantage or disadvantage Formation of solid LiH for H concentrations >1-5 at.% at 200-500oC Potential concerns: enhanced erosion of piping in flowing Li solidification/plugging in cold regions of circuit

5. Common assumptions/assertions (some of which need further consideration) Technology basis: Molten salt extraction method (cf. V.A. Maroni et al., Nucl. Tech. 25 (1975) 83 Gas sparging, distillation, evaporation, permeable membrane approaches unlikely to be viable due to anticipated requirement to reduce tritium concentration to <10 ppm Cold trapping (solid H gettering) techniques might achieve <10 ppm T2, but difficult to use in continuous mode, regeneration concerns, etc. • H removal from Li is a relatively easy technology Schematic of centrifigal contactor (never constructed/ operated) Scoping test used ~1 cm3 each of Li and salt in a Nb-1Zr capsule, exposed to 2 mCi tritium in 1 liter Ar gas to evaluate partitioning between Li and molten salt 1. Li and salt are mixed to enable transfer of tritium to salt; 2. Li and salt are demixed (<0.1% residual salt) 3. tritium is extracted from the salt (requires extraction electrode yet to be designed, so evolved tritium is efficiently collected; 20% T2 removal efficiency assumed) Mixing and separating equipment must have good compatibility with Li and molten salt at ~500oC

6. Alternative approaches for tritium removal from Li Use Y, Ti, etc. to getter H isotopes; periodically extract tritium from getter by high T anneals, chemical dissolution, etc. -difficulties with continuous operation (e.g., surface fouling), etc. • Cold- or hot-trapping (proposed for IFMIF Li loop, etc.) • Distillation column High power consumption; very high operating temperature (separation factor becomes >1 only for T>1100oC for equilib. distillation) Separation factor Nonequilibrium distillation: appropriate for dilute T2 concentrations LiT + Li2T Molecular T2 V.N. Tebus et al., Atom. Energy 84, 3 (1998) 202

7. Only refractory alloys are compatible with Li at T>1000oC: numerous fabrication and operational issues (UHV-scale impurity control, etc.) Tmelt embrittlement Corrosion ~200°C window

8. Tritium recovery from Pb-Li and FliBe (low H solubility) coolant • Example of one systems-level analysis for a liquid breeder • S. Fukada et al., Fus. Sci. Technol. 61, 1T (2012) 58 From abstract: “Simultaneous transfer of heat and T needs new configuration in order to satisfy the necessary conditions of low T leak and high heat transfer” T2 concentration in PbLi needs to be reduced to ~10-10 before power conversion to achieve acceptably low tritium release rates Flibe has unacceptably high T2 leakage rates and requires development of tritium permeation barriers

9. Li-Sn equilibrium phase diagram Sn-20at%Li H solubility in Sn is low L.A. Sedano, Fus. Sci Tech. 48 (2005) 605 TM~320oC

10. Very few materials are compatible with Sn at elevated temperatures T melt Window almost closed

11. Irradiation damage (cavity formation) will result in significantly higher tritium retention compared to H solubility in unirradiated materials Example of H sequestration in fission neutron irradiated Type 316 stainless steel 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) F.A. Garner et al., J. Nucl. Mater. 356 (2006) 122 Fusion may need to avoid operation at conditions that produce fine-scale cavities in structural materials Zinkle et al., Nucl. Fus. 53 (2013) 104024; based on F.A. Garner et al., J. Nucl. Mater. 356 (2006) 122

12. 764°C, 0.16 dpa Irradiated Tungsten Microstructure 800°C, 0.98 dpa 742°C, 2.2 dpa L Loop Selected for TEM (Still in Hot Cells) L Loop + Void Literature (Hasegawa et al). P Void + Precipitate 900 Voids P Temperature range for enhanced T2 sequestration in W L 800 P L L Precipitates 700 2.2 dpa Precipitate 724°C, 0.639 dpa Loop & Void 2.2 dpa 600 6.4 dpa 3.8 dpa 742°C, 2.2 dpa & Void 500 Irradiation Temperature (°C) L 3.8 dpa 400 L Dislocation loops 300 Voids 2.2 dpa 6.4 dpa Loop 200 397°C, 0.032 dpa 90°C, 0.639 dpa 100 L L 6.4 dpa 0 1 2 1.5 0.5 Irradiation dose (dpa) Dislocation loops Dislocation loops Snead : ICFRM-17 Aachen

13. Several materials-tritium issues require additional investigation • Identification of a robust, efficient and economic method for extraction of tritium from high temperature coolants • No current options above TRL~2 • Tritium sequestration/inventory may be an important parameter • Fission power reactors (typical annual T2 discharges of 100-800 Ci/GWe; ~10% of production) are drawing increasing scrutiny • A 1 GWe fusion plant will produce ~109Ci/yr; typical assumed releases are ~0.3 to 1x105Ci/yr (<0.01% of production) • Will 1-10 Ci/day release from a fusion power plant be acceptable? • Neutron-induced cavity formation may lead to significant trapping of hydrogen isotopes in the PFC and blanket structure (depends on Tirr, dose) • Tritium trapping efficacy of precipitates and nanoscale solute clusters (blanket & piping) is poorly understood from a fundamental perspective