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Tritium Retention in Graphite and Carbon Composites

Tritium Retention in Graphite and Carbon Composites. Rion Causey Sandia National Laboratories Livermore, CA 94550. Sandia National Laboratories. Carbon. Graphite occurs naturally. Nuclear graphites are made by Acheson process Crush, mill, size Add coal tar pitch and anneal at 1200 K

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Tritium Retention in Graphite and Carbon Composites

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  1. Tritium Retention in Graphite and Carbon Composites Rion Causey Sandia National Laboratories Livermore, CA 94550 Sandia National Laboratories

  2. Carbon • Graphite occurs naturally. • Nuclear graphites are made by Acheson process • Crush, mill, size • Add coal tar pitch and anneal at 1200 K • Graphitize at 2900 K to 3300 K • Product has density of 1.8 to 1.9 g/cm3 (very porous) • Grain size is ~ 10 µm • Grain composed of microcrystallites (~5 nm) • Carbon composites are very similar to graphite • High strength-to-weight ratio • Can be tailored to have directional properties

  3. Carbon • Tritium Retention in Carbon • Saturated layer • Absorption and diffusion along porosity • Intergranular diffusion and trapping • Codeposition Sandia National Laboratories

  4. Absorption and Diffusion Along Graphite Porosity Tritium Profile in POCO Graphite • Examination of tiles removed from JET and TFTR have shown tritium profiles extending completely through the tiles. It is the diffusion along porosity that produces these profiles. Removal may be difficult by heating alone. Heating may just drive the tritium into the grains where much higher temperatures will be required. Heating to moderate temperatures (300 to 500 C) in the presence of atomic hydrogen or oxygen should effectively remove this tritium. The inventory associated with the process is small. Sandia National Laboratories

  5. Codeposition is Expected to be a Major Source of In-Vessel Tritium Inventory • Carbon erosion from high flux areas results in redeposition of carbon along with tritium. • Tritium concentrations of 0.4 T/C are expected in a typical DT device. • (JET had 1.0 (D+T)/C on the louvers) • The thickness of the codeposited layer increases monotonically with discharge time. Codeposited film on TFTR bumper limiter (B. Mills, SNL)

  6. Other Useful Codeposition Information • As the thermal decomposition data to the right shows, codeposition can not occur on hot (>1000 K) surfaces. • Most codeposited layers are really coimplanted layers (energetic hydrogen neutrals strike the redeposited carbon layer). • The codeposited layer found on the JET louvers had a (D+T)/C ratio of approximately 1.0. • ITER-FEAT is predicted to have 1 to 2 grams of tritium codeposit with carbon per pulse. (Federici et al. J. Nucl.Mater. 290-293 (2001) 260) • Glow discharge (and He/O glow discharge) cleaning is too slow to be effective in removing the codeposited layer. Thermal Stability of the Codeposited Layer Sandia National Laboratories

  7. Removal of the Codeposited Layer • One proposed technique for the removal of the codeposited layer is to heat the entire vessel to a temperature of about 550 K in the presence of air (see figure to the right). • Concerns with the heating in air technique include damage to the vacuum vessel as well as the reconditioning of the vessel after the layer removal. • Similar techniques include the use of UV and ozone at lower temperatures. The same concerns listed above apply to these techniques. Stability of the Codeposited Layer in Air Sandia National Laboratories

  8. Novel laser detritiation technique shows promise • Scanning Nd laser heats surface to ≈1500 C and thermally desorbs tritium • Up to 87% of tritium has been removed from TFTR and JET carbon tile samples • Advantages for tokamak application: • fiber optic coupling to in-vessel scanner • fast - potential overnight cleanup in a next-step machine. • no oxygen to decondition PFC’s • no HTO to process Laser spot TFTR tile with codeposit Pyrometer • Heating by scanning laser mimics heat loads in slow transient off-normal events in tokamaks. • Opens new technique for studying high heat flux interactions, and “brittle destruction”. • Preprints available on PPPL website“http://www.pppl.gov/”PPPL-3603, PPPL-3630, PPPL-3604, PPPL-3662 Tritium release Charles Skinner, PPPL.

  9. Intergranular Diffusion with Trapping in Graphite • At temperatures of 1000 K and above, tritium begins to diffuse into the graphite grains where it it trapped at high energy trap sites (4.3 eV). • Each of the ~10 µm grains to the right-> are composed on smaller microcrystallites (5 nm). We think that the hydrogen diffusion occurs along the edges of the smaller crystallites. We also think that the high energy traps only occur along the prism plane (not on the chemically inert basal plane). Graphites and composites with large crystals (low surface to volume ratio) or large basal plane to prism plane ratios exhibit lower trapping densities. Sandia National Laboratories

  10. Intergranular Diffusion with Trapping • Unirradiated graphites typically have a tritium trap density of 10 to 20 appm. Neutron irradiation can increase this trap density to values above 1000 appm. • It is almost certain that irradiation at higher temperatures will limit the production rate of these traps. The higher temperatures simply allow some recovery of the radiation damage. • Experimental determination of the tritium trap production as a function of temperature is needed. Tritium Trapping in N3M Graphite Causey et al., Fusion Technol. 19 (1991) 1585 Sandia National Laboratories

  11. Radiation Induced Trapping in Graphite • The trap density appears to saturate at rather low dose levels • The saturation damage fluence of 0.1 dpa is equivalent to that expected for the old ITER Physics Phase. Causey et al., Fus. Technol. 19 (1991) 1587 Sandia National Laboratories

  12. Intergranular Diffusion with Trapping in Graphite • Experiments on two pitch based fiber composites demonstrated resistance to tritium trap generation. • These composites were known to have very limited fractions of prism planes. • Unfortunately, these fiber composites are extremely expensive. Irradiated at Room Temperature Sandia National Laboratories

  13. Inventory Predicted for Neutron Irradiated Graphites or Composites (DIFFUSE Code) Inventory After 3 Years of Continuous Exposure to Tritium Gas (10 µm grain size) Temperature (K) Graphite with Graphite with 100 appm traps 1000 appm traps* 4 gm/m3 17 gm/m3 23 gm/m3 22 gm/m3 21 gm/m3 20 gm/m3 19 gm/m3 14 gm/m3 56 gm/m3 164 gm/m3 225 gm/m3 220 gm/m3 209 gm/m3 191 gm/m3 1000 1100 1200 1300 1400 1500 1600 * It may not be possible to have 1000 appm trapping at these elevated temperatures Sandia National Laboratories

  14. Conclusion Carbon used in fusion reactors can retain large quantities of tritium through either codeposition or trapping at 4.3 eV traps Codeposition is a low temperature process Trapping at 4.3 eV traps is a high temperature process Experiments are needed on tritium trapping in graphites and composites irradiated at higher temperatures Limited experiments on silicon carbide suggest this material to present less trapping at elevated temperatures (appears to be very radiation resistant) Sandia National Laboratories

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