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Impact of the cementitious engineered barriers on the stability of the high-level waste matrices

Impact of the cementitious engineered barriers on the stability of the high-level waste matrices. Karel Lemmens, Christelle Cachoir, Karine Ferrand, Thierry Mennecart, Ben Gielen, Regina Vercauter Euridice Exchange meeting Mol, January 29th 2009.

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Impact of the cementitious engineered barriers on the stability of the high-level waste matrices

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  1. Impact of the cementitious engineered barriers on the stability of the high-level waste matrices Karel Lemmens, Christelle Cachoir, Karine Ferrand, Thierry Mennecart, Ben Gielen, Regina Vercauter Euridice Exchange meeting Mol, January 29th 2009

  2. Safety functions attributed to the waste forms • The waste form contributes to the safety of the disposal system because it spreads the radionuclide release in time  determine the release rate in supercontainer conditions (programme 2004-2008) • The interaction between waste form and concrete may lead to lower dissolved radionuclide concentrations by inclusion of the leached radionuclides in secondary minerals (not studied so far) Euridice Exchange meeting, January 29th 2008, Mol 2/26

  3. Outline of the presentation • Expected evolution of vitrified waste in supercontainer conditions • Expected evolution of spent fuel in supercontainer conditions Euridice Exchange meeting, January 29th 2008, Mol 3/26

  4. Part I Expected evolution of vitrified waste in supercontainer conditions Stage 1: thermal stage (containment by overpack) Stage 2 : perforated, corroding overpack Stage 3: post-overpack

  5. Reference disposal design for vitrified wasteStage 1: at time of gallery closure and during thermal stage pH 8.3-8.6 13.5 Temperature 16 °C <100 °C Waste glass containers cracks Concrete overpack Host rock (Boom Clay)

  6. Stage 2: Vitrified waste after overpack failure, until complete overpack corrosion (800 to 10 000…15 000 years) 16°C < 25°C pH Temperature 8.3-8.6 water 13.5 Radionuclides Overpack corrosion products (magnetite ?) Altered concrete (?)

  7. Stage 2: Vitrified waste after overpack failure: situation at first water contact : fast dissolution expected Si Al B RN (altered) concrete magnetite Overpack H2 gas Water vapour Cement water pH 13.5 glass crack

  8. Stage 2 : vitrified waste after overpack failure:pH decreasing processes: pH 13.5  pH 9 -10 (?) Ca Si Al Ca Si Al Al RN RN RN Si, Al, Ca precipitation (+ sorption ?) pH 13.5 Secondary phases Secondary RN phases pH 9- 10 (?)

  9. R7T7 glass : preliminary quantitative results Dissolution rate of Cogema R7T7 glass at pH 13.5 : 9 – 30 gram (glass) m-² year -1 (expert range) • Estimation of effective surface area of glass block = outer surface x cracking factor (internal surface): • Cracking factor 5 - 40 (literature) • Time for complete dissolution of a glass block 400 – 10 000 years • pessimistic because of assumed constant pH 13.5 Dissolution rate at pH 12.5 – 11.7 : > one order magnitude lower Dissolution rate at pH 9-10: > two orders magnitude lower

  10. Stage 3: Vitrified waste after complete overpack corrosion thick alteration layer on the outside pH Temperature 8.3-8.6 16°C < 25°C 13.5 Radionuclide release 9 -13.5 (?) Glass replaced by thick layer of secondary phases

  11. Vitrified waste : priorities for the future • Current dissolution rate estimations based on small number of experiments •  Decrease of uncertainty on the proposed values is necessary • The current dissolution rate estimations are very pessimistic  Improved understanding of the processes to estimate degree of pessimism, and – if possible – to propose more realistic, probably lower dissolution rates • e.g. due to pH decrease • Evolution of effective glass surface area at high pH • (less prioritary topic ?): effect of secondary phases on radionuclide concentrations at the glass/cement interface

  12. Part II Expected evolution of spent fuel in supercontainer conditions Stage 1: thermal stage (containment by overpack) Stage 2: perforated, corroding overpack Stage 3: post overpack

  13. Simplified disposal design for spent fuelat time of gallery closure Spent fuel concrete overpack Host rock (Boom Clay)

  14. Details of a charged spent fuel container Cast iron insert Overpack Fuel assembly with cladded fuel (and filling material)

  15. Stage 1: Spent fuel during thermal stage (< 2500 years)No spent fuel dissolution Temperature pH 16 °C 8.3-8.6 <100 °C Reducing conditions Fe(II) 13.5 H2 Decay of b,g isotopes Overpack corrosion products (magnetite) (altered) concrete

  16. Stage 2: Spent fuel after overpack failure, until complete overpack corrosion (2500 – 10000…15000 years) pH Temperature 16°C 8.3 – 8.6 <25°C Reducing conditions 13.5 Fe(II) H2 a emitters Overpack corrosion products (magnetite) (altered) concrete

  17. Stage 2: Spent fuel after overpack failure(1) fast release of instant release fraction (IRF)  f(pH) He IRF concrete diffusion+sorption magnetite Overpack (+ iron insert + cladding) Gap H2 gas Water vapour Crack UO2.x P P P P P cracking P P P

  18. Stage 2: Spent fuel after overpack failure (2) matrix dissolution =f(pH) concrete magnetite H2 Water vapour a H2 U(IV) H2O2 H2 H2 OH° U(VI) RN RN RN RN UO2 RN RN RN RN RN RN RN RN RN

  19. Stage 3: Spent fuel after complete overpack corrosion Temperature pH 8.3 – 8.6 16–38°C <25°C Reducing conditions 13.5 Fe(II) a emitters Overpack replaced by layer of magnetite and/or layer of altered concrete (?)

  20. Stage 3: Spent fuel after complete overpack corrosion slow matrix dissolution threshold activity for oxidative dissolution ? altered concrete magnetite U sorption a H2O2 Fe(II) effect ? U(VI) U(IV) OH°

  21. Some typical results of the tests with depleted UO2 at high pH (programme 2004-2008) Red: pH 13.5 Green: pH 12.5 Blue: pH 11.7 [U] for pH 13.5 in the range 10-8 - 10-6 mol.L-1 Uranium conc. (M) Dissolution rate initially high, then close to zero  use pessimistic average rate Days

  22. Spent Fuel : preliminary results UO2 dissolution rate at high pH in the range 40 – 8000 µg m-2 year -1 • Estimation of effective surface area of irradiated UO2 pellets (literature): • As removed from reactor: cracking factor 15 • Increase of surface area during (geological) disposal : cracking factor 45 • Time for complete dissolution of a UO2 pellet (crack. factor 45) 105 years – 24.106 years • same range as for neutral pH, but still large uncertainty Euridice Exchange meeting, January 29th 2008, Mol 22/26

  23. Spent Fuel : preliminary results • U concentrations at pH 13.5 from 10-8 tot 10-6 M •  10-6 M high compared to neutral pH ([U] = 10-8 – 10-9 M) • (but confirmation is necessary) Euridice Exchange meeting, January 29th 2008, Mol 23/26

  24. Spent Fuel : priorities for the future • Reduce the uncertainties on the dissolution rate estimation • distinguish the high initial dissolution rate from the low long term rate • effect of cement phases in the system (a.o. sorption) • effect of alpha activity (threshold for oxidative dissolution ?) • Reduce the uncertainties on the fuel surface evolution • Reduce undertainties on uranium solubility at high pH in contact with spent fuel • Other topics: temperature, sorption on magnetite, sand.., effect of hydrogen, effect of burn-up, data for MOX… Euridice Exchange meeting, January 29th 2008, Mol 24/26

  25. General conclusions • Vitrified waste • The supercontainer conditions probably increase the radionuclide release rate from vitrified waste, compared to the previous, bentonite based engineered barrier system. • Current rate estimations are pessimistic, larger database and better understanding necessary • High pH may decrease RN solubility at glass/cement interface by secondary phase formation (not yet studied) Euridice Exchange meeting, January 29th 2008, Mol 25/26

  26. General conclusions • Spent fuel • The supercontainer conditions seem to have no clear impact on the radionuclide release rate from spent fuel, compared to the previous, bentonite based engineered barrier system. • Larger database and better understanding necessary to improve the dissolution rate estimations • Reduce uncertainties on uranium solubility at fuel/concrete interface Euridice Exchange meeting, January 29th 2008, Mol 26/26

  27. Wake up ! I’m finished !

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