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Research performed at UNLV on the chemistry of Technetium in the nuclear fuel cycle

Research performed at UNLV on the chemistry of Technetium in the nuclear fuel cycle. 1. Separation U/Tc and synthesis of solids form 2. Synthesis and characterization of Tc-Zr alloys. Background. In the US: Spent fuel inventory in 2014: 65 000 MT of spent fuel ~ 50 MT of 99 Tc

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Research performed at UNLV on the chemistry of Technetium in the nuclear fuel cycle

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  1. Research performed at UNLV on the chemistry of Technetium in the nuclear fuel cycle 1. Separation U/Tc and synthesis of solids form 2. Synthesis and characterization of Tc-Zr alloys

  2. Background In the US: Spent fuel inventory in 2014: 65 000 MT of spent fuel ~ 50 MT of 99Tc DOE: Various options for nuclear waste management 1. Direct disposal of spent fuel: Deep bore hole

  3. 2. Reprocessing and development of waste storage forms. • Development of experimental separation process : • UREX process: U recovered and Tc placed in a waste form for storage •  No PUREX because of proliferation concerns Study at UNLV focused on Tc separation for UREX process and development of metallic technetium waste form

  4. UREX process: Suite of solvent extractions. UREX segment

  5. UREX segment 1. Acetohydroxamic acid :AHA  Reduction Np, Pu  Prevent extraction by TBP 2. TBP in dodecane  Extraction: U&Tc 1.5 M H+, 4 M NO3- Pu, Tc, U, Np [U]= 50-100 g/L [Tc] = 60-130 mg/L 0.01 M HNO3 3. 0.01 M HNO3 Tc & U back extracted U& Tc TBP/dodecane

  6. 1. U/Tc Separation for UREX processsynthesis of solids forms

  7. Separation U/Tc already been studied at ANL. Labscale -Demonstration of UREX process using spent fuel  Tc separated from U using Anionic exchange resin No waste Tc form synthesized Separation U/Tc at ANL: anionic exchange resin Spent fuel Goal : Separation U/Tc & Synthesis Tc waste form

  8. A- Lab scale demonstration B- Synthesis and characterization of solid forms C- Conclusion

  9. A- Lab scale demonstration Solution: [U]= 100 g/L, [Tc]= 130 mg/L in 1L 0.01M HNO3 Separation of Tc from U using anionic exchange resin Experimental condition: Elution column: 7 g of treated Reillex Guard column: 1 g of Reillex Elution: -350 ml of 1M NH4OH (flow rate = 4 ml/min) Set-up for lab scale demonstration

  10. Results Elution profile Total Sorption yield : 97.7%. Elution yield of treated Resin : 93.7 % Products obtained after separation Uranium: Technetium: 1 liter of UO2(NO3)2 in 0.01M HNO3 350 ml of TcO4- in 1 M NH4OH

  11. B- Synthesis of solid forms 1.Uranium 1.Synthesis of Uranyl hydroxide 1.Precipitation NH4OH Uranyl nitrate Uranyl hydroxide 2. Filtration 2.Conversion to uranium ammonium oxide 250 ºC 3 hours 119.44 g), [Tc] < DL Ammonium Uranyl hydroxide(172.60 g) Tc < DL

  12. 2.Technetium 1.Synthesis of (n-Bu4N)TcO4 Dissolution Evaporation Precipitation (n-Bu4N)HSO4 Centrifugation NH4TcO4 (15 %) NH4NO3 (85%) Need to separate! 2.Conversion to Tc metal (Steam reforming) Tc Reduction at 800 °C under wet Ar: (n-Bu4N)TcO4 + 2H2 → Tc metal + 2 H2O  H2 /CO produced by reaction between Carbone and H2O at 800 °C T= 800 °C Wet Ar, 5 hours Arc melting (n-Bu4N)TcO4: 520 mg Tc metal : 68.2 mg

  13. Characterization Tc metal XRD XRD • Tc hexagonal • No other phase EXAFS EXAFS 13(2) Tc @2.72 Å • Tc hexagonal

  14. Recovering of the Technetium on the guard column by pyrolysis (Steam reforming) Before pyrolysis After pyrolysis 900 ° C Wet Ar Resin in “Tea bag” Set up used for pyrolysis Optical and SEM microscopy Before pyrolysis After pyrolysis 900 ° C Wet Ar Tc metal : SEM x 300 R- TcO4 Tc metal: x 40

  15. C- Conclusion • 1. Optimization Uranium/Technetium separation •  Tc Elution yield of 93 % on Reillex HP resin • 2. Synthesis of U and Tc solid form •  U product is free of Tc and was recovered in a yield of 99.4%. • Tc metal is free of U and was obtained in a yield of 52.5%.

  16. 2. Synthesis and Characterization of Tc Waste Form

  17. Reprocessing activity of spent fuel will produce technetium stream DOE: Technetium plan to be incorporated into a metallic waste form • Two metallic waste forms considered: • 1. Tc metal • Possibility to transmute into stable Ru • 2. Tc-Zr alloys • Make a combined waste form with the Zr from the cladding • Permit to decrease the melting point of waste form  Determination Tc-Zr phase diagram  Stability of Tc-Zr alloys

  18. Four composition analyzed: Tc6.1Zr, Tc2.1Zr, TcZr1.1, TcZr5.7 Arc melted Annealed at 1400 °C Mixed Pressed Tc + Zr Four different phases observed • Tc6.2Zr and Tc4.6Zr (a-Mn, cubic) • Tc2Zr (Zn2Mg, hexagonal) • b-Zr(Tc) (solid solutions of Tc in Zr) Poineau, F., et al. Inorg. Chem. (2010) 49, 1433.

  19. Experimental Tc-Zr phase diagram at 1400 °C

  20. Behavior of Tc-Zr in oxidizing conditions Tc6Zr, Tc2Z and TcZr treated 3 days at 1500 °C under Ar Low presence of O2 in the system (release from alumina tube) • Zr complete oxidation to ZrO2 • Tc remain as the metal Dark: ZrO2 Light: Tc TcZr sample After treatment SEM: Phase separation XRD: Tc metal and ZrO2 • Tc metal more stable than Zr toward oxidation • Oxygen free atmosphere required to develop Tc-Zr waste form • Tc metal might be a more stable waste form than Tc-Zr

  21. Questions

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