Spent nuclear fuel. II Letnia Szkoła Energetyki i Chemii Jądrowej

Transcription

1 Spent nuclear fuel Mats Jonsson KTH Chemical Science and Engineering, Royal Institute of Technology, Stockholm, Sweden II Letnia Szkoła Energetyki i Chemii Jądrowej

2 Spent nuclear fuel Mats Jonsson, KTH Chemical Science and Engineering, Royal Institute of Technology, Stockholm, Sweden

3 Part 2: Geological repositories for spent nuclear fuel Purpose of a geological repository Proposed repository concepts The KBS-3 concept Processes that can influence the safety of a repository

4 A reminder

5 > years! That is for how long the repository must remain safe That is also how many years ago Homo sapiens first appeared What could happen in the next years???

6 Safety assessment Extreme extrapolations Impossible to perform experiments on real materials (a typical scenario is that a canister fails after years) Very difficult to predict changes in climate and in human activities How do we communicate with future generations?

7 Purpose of a geological repository To isolate the radioactive waste from the biosphere until the radioactivity is comparable to a uranium ore.

8 Proposed repository concepts Deep boreholes Clay formations Brines (salt) Granitic bedrock

9 The KBS-3 concept

10 The repository site in Sweden (Forsmark)

11 Site investigations Oskarshamn vs Forsmark Both municipalities wanted to host the repository!

12 Fuel, canister and bentonite

13 Barriers 0. The fuel 1. The canister 2. The bentonite clay 3. The bedrock

14 Handling the spent nuclear fuel Storage at nuclear power plant (in pools) for 1-2 years Transportation to interim storage (by ship) Interim storage (in pools) for years Encapsulation Final/long-term storage in deep repository

15 Spent nuclear fuel in Sweden CLAB in Oskarshamn

16 The fuel UO 2 has very low solubility in reducing groundwater

17 The canister Mechanical support Corrosion resistance

18 The bentonite clay Mechanical support Self sealing Diffusion barrier

19 The bedrock Low flow Retention of radionuclides

20 Processes that can influence the safety of a repository The fuel The canister The bentonite

21 Dissolution of spent nuclear fuel

22 Dissolution of spent nuclear fuel Instant release UO 2 matrix dissolution The cladding is assumed to be gone (in a typical scenario)

23 Instant release Readily soluble fission products at the surface are rapidly released to the groundwater.

24 The Fuel Matrix (Spent Nuclear Fuel) ~95 % UO 2 BUT ALSO Highly radioactive (depending on age) Insoluble noble metal inclusions (fission products) Rare earth oxides (fission products) Heavy actinides (activation)

25 Dissolution of UO 2 /SNF in groundwater UO 2 has very low solubility under the expected groundwater conditions (reducing). Soluble radionuclides present at the fuel surface are readily released upon contact with water (Instant release). Oxidation of UO 2 increases the matrix solubility by several orders of magnitude. Oxidants will be produced upon radiolysis of groundwater. Ox + UO 2 UO 2 2+ UO 2 2+ (aq)

26 System description (somewhat simplified) Mixed radiation field (a, b and g) (gradient) Water radiolysis (including groundwater components) HETEROGENEOUS! Surface reactions: 1. Oxidation 2. Reduction 3. Dissolution 4. Precipitation 5. Catalysis Diffusion

27 We have to start with something simpler Radiation induced dissolution of pure UO 2 When we understand this system, the complexity can be increased: 1. Effect of H 2 2. Effect of noble metal inclusions 3. Effect of rare earth oxide doping

28 Radiation induced dissolution of UO 2 Reactivity of aqueous radiolysis products (oxidants) towards UO 2 OH, HO 2, H 2 O 2, O 2, CO 3 -

29 ln k UO 2 oxidation kinetics (powder) The rate constant for oxidation is a function of reduction potential. E 0 (V) 0-0,5 0,5 1,5 1 2,5 Diffusion limit J. Nucl. Mater. 2003, 322,

30 k /m min -1 Oxidation (by H 2 O 2 ) vs Dissolution (Effect of HCO 3- ) 5,00E-06 4,50E-06 4,00E-06 3,50E-06 k /m min -1 5,00E-06 3,00E-06 Oxidation is the rate limiting 4,50E-06 step > 1 mm HCO - 3 2,50E-06 4,00E-06 3,50E-06 2,00E-06 1,50E-06 1,00E-06 5,00E-07 0,00E+00 3,00E-06 2,50E-06 2,00E-06 1,50E-06 1,00E-06 5,00E-07 0,00E ,2 0,4 0,6 0,8 1 1,2 [HCO - 3 ]/mm [HCO - 3 ]/mm J. Nucl. Mater. 2006, 358,

31 How to use the rate constants for radiation induced UO 2 dissolution Total rate of oxidation: rate dn U ( VI ) dt A UO 2 n ox 1 k ox Ox n e 2 J. Nucl. Mater. 2006, 355, 38-46

32 n U(VI) (μmol) g-radiolysis of UO 2 -suspensions 8,0 7,0 g-radiolysis 6,0 5,0 4,0 Good agreement! 3,0 Calculated Experimental 2,0 1,0 0, Irradiation time (min) J. Nucl. Mater. 2006, 355, 38-46

33 Relative impact of (a-) radiolysis products rate dn U(VI) dt A UO 2 n ox 1 k ox n e Ox 2 H 2 O 2 O 2 O 2 - HO 2 CO 3 - OH No additives % 0.01 % 0 % 0.03 % 0 % 0 % H 2 (40 bar) 99.9 % 0 % 0 % 0.02 % 0 % 0.03 % H 2 (40 bar) HCO 3 - (10 mm) % 0 % 0 % 0 % 0.02 % 0 % HCO 3 - (10 mm) 99.9 % 0.09 % 0 % 0 % 0 % 0 % H 2 O 2 is the major oxidant! J. Nucl. Mater. 2006, 355, 38-46

34 Next step Rate of H 2 O 2 production r H 2 O 2 x max x 0 D ( x) G( H 2 O2 ) dx Can this be simplified? Environ Sci. Technol. 2007, 41,

35 Dose rate (Gy/s) Radiolysis: Geometrical dose distribution (based on RN inventory) 0,5 0,45 0,4 0,35 alpha beta total 0,3 0,25 0,2 0,15 0,1 0, Distance from fuel surface ( m) J. Nucl. Mater. 2006, 359, 1-7

36 First test Simulation including: 1. Dose profile (H 2 O 2 production) 2. H 2 O 2 consumption at the UO 2 surface 3. Diffusion in one dimension (x)

37 [H 2 O 2 ]/M Concentration profile as a function of time 9,00E-10 8,00E-10 7,00E-10 6,00E-10 Steady-state 5,00E s 4,00E-10 3,00E-10 2,00E-10 1 s 3 s 10 s 1,00E-10 0,00E Distance/µm J. Nucl. Mater. 2008, 372, and J. Nucl. Mater. 2008, 374,

38 Steady-state 90% of the steady-state (surface) concentration is reached in a very short time (Seconds-Minutes)

39 Does the steady-state approach work? Material (Dose rate) 10 % U- 233 (99 Gy/h) 10 % U- 233 (99 Gy/h) SF (a = 828 Gy/h b = 31 Gy/h) SF (a = 828 Gy/h b = 31 Gy/h) p(h 2 ) [HCO 3 - ] (mol dm -3 ) Time (days) Calc. final conc (mol dm -3 ) Calc. diss rate (mol dm -3 d -1 ) Experimental final conc. (mol dm -3 ) (Ar) (1,2 % O2) (Ar) , bar x 10-10

40 We are ready to increase the complexity!

41 The effect of H 2 H 2 inhibits dissolution of spent nuclear fuel. Why? Reduces the rate of H 2 O 2 production (not sufficient) J. Nucl. Mater. 2010, 396,

42 The mechanism 1,2 1 M.E. Broczkowski, J.J. Noël, D.W. Shoesmith, J. Nucl. Mater. 346 (2005) 16 2 M. Jonsson, F. Nielsen, O. Roth, E. Ekeroth, S. Nilsson, M.M. Hossain, Environ. Sci. Technol. 41 (2007)

43 Spent Nuclear Fuel Fission products Actinides a b g H 2 O e - aq H H 2 OH U(IV) Oxidizing species Reduced species H 2 O 2 H 2 e 2 e - 2 H + HCO 3 - U(VI) UO 2 2+ (aq) Fission products Actinides

44 Spent Nuclear Fuel Fission products Actinides a b g H 2 O e - aq H H 2 OH U(IV) Oxidizing species Reduced species H 2 O 2 H 2 e 2 e - 2 H + HCO 3 - U(VI) UO 2 2+ (aq) Fission products Actinides

45 But how efficient is the noble metal catalyzed reduction by H 2?

46 Experiment Uranium release from pellets containing Pd particles Exposed to H 2 O bar H 2

47 Noble metal catalyzed inhibition of H 2 O 2 induced dissolution (Pd-doped UO 2 pellets) 1,3 1,1 0% 0.1 % 1% 3% 0,9 r diss /r ox 0,7 0,5 0,3 0,1-0, e x ph 2 (bar) k = 10-6 m s -1 (diff. limited) Martin Trummer, Sara Nilsson and Mats Jonsson, J. Nucl. Mater. 2008, 378, 55-59

48 Noble metal inclusions Also catalyze oxidation of pellets

49 Rare earth oxides

50 r_diss(u(vi)) / mol L-1 s-1 H 2 O 2 induced oxidative dissolution (doped UO 2 pellets) 3E-09 1 % Pd, 0.3 % Y2O3 2,5E-09 2E-09 1,5E-09 1E-09 N2 5E-10 0 Y2O3 Y2O3 / Pd Pd UO2

51 c(uranyl) / µm Radiation (g) induced dissolution 80 N2 experiments UO2 Y2O3 50 Y2O3/Pd 40 Pd time / min

52 H2O2 concentration (mm) H 2 O 2 consumption SIMFUEL/UO 2 2,5 2 1,5 Rate as expected from k(h 2 O 2 )* 1 0, reaction time (min) *M. M. Hossain, E. Ekeroth, M. Jonsson. J. Nucl. Mater. 2006, 358,

53 concentration U(VI) μm H 2 O 2 induced U(VI) dissolution SIMFUEL/UO UO SIMFUEL reaction time (min) J. Nucl. Mater. 2011, 410, 89-93

54 U(VI) concentration μm Radiation (g) induced dissolution of SIMFUEL/UO UO SIMFUEL irraditation time (min) J. Nucl. Mater. 2011, 410, 89-93

55 What happens to H 2 O 2? UO OH - (Dissolution) H 2 O 2 + UO 2 H 2 O + ½ O 2 + UO 2

56 Catalytic decomposition of H 2 O 2 H 2 O 2 2HO (surface catalyzed) HO + H 2 O 2 H 2 O + HO 2 HO 2 + HO 2 O 2 + H 2 O 2 S 2H 2 O 2 O 2 + 2H 2 O (ZrO 2 ) Cláudio Lousada and Mats Jonsson, J. Phys. Chem. C, 2010, 114 (25)

57 Detection of OH OH + TRIS Formaldehyde

58 [H2O2] (mm) [HO. ] (mm) H 2 O 2 + ZrO 2 6 0,4 4 2 H 2 O 2 HO. 0,2 0 0, time (s) Cláudio Lousada and Mats Jonsson, J. Phys. Chem. C, 2010, 114 (25)

59 Why do metal oxides catalyze decomposition of H 2 O 2? Adsorption of H 2 O 2 and OH

60 H 2 O 2 HO HO Metal oxide

61 scavenged [HO ] mm Hydroxyl radical affinity 0,16 0,14 0,12 0,1 0,08 0,06 0,04 0, Irradiation time (s) No Oxide Present ZrO2 TiO2 Y2O3 [H2O2]/[H 2 O 2 ] 0 TiO 2 <ZrO 2 <Y 2 O Y2O3 TiO2 Lousada et al. to be published time (s)

62 [H2O2] mm UO 2 powder 6,00 5,00 4,00 3,00 2,00 1,00 0,00 0,5 0,45 0,4 0,35 0,3 0,25 0,2 0,15 0,1 0, time (s) [OH] mm

63 [HO] mm UO 2 pellet ,8 18 1,6 16 1,4 1, [U] microm 0,8 8 0,6 6 0,4 4 0, time (s)

64 [HO] mm SIMFUEL pellet ,8 18 1,6 16 1,4 14 1,2 1 0, [U] microm 0,6 6 0,4 4 0, time (s)

65 Dissolution yields: [U(VI)]/[H 2 O 2 ] UO 2 powder: 83 % (80 %) UO 2 pellet: 2 % (14 %) SIMFUEL pellet: 0 % (0.2 %) Catalytic decomposition is the main reaction path for H 2 O 2 on pellets! Why is UO 2 NOT oxidized by catalytically produced OH?

66 What is the origin of the effect of dopants? Effects on k cat And/or Effects on k ox

67 OH production (pellets)

68 [ U(VI) ]normalized to westinghouse microm Uranium dissolution (pellets) Pd Y UO2 YPd Westinghouse SIMFUEL time (s)

69 Redox reactivity Oxidants: MnO 4 - and IrCl 6 2-

70 Norm. [MnO 4 - ] Oxidation of UO 2 by MnO 4-1,2 1 0,8 0,6 0,4 UO2 SIMFUEL 0, Time (min)

71 Activation energies (MnO 4- ) UO 2 (Westinghouse): 7.4 kj mol -1 SIMFUEL: 12.9 kj mol -1

72 Rate constants for pellet oxidation Pellet k(h 2 O 2 )/min -1 k(mno - 4 )/min -1 k(ircl 2-6 )/min -1 UO 2 (Westinghouse) 1.5 x x x 10-2 SIMFUEL 1.4 x x x 10-2 UO x x x 10-2 UO 2 /Y 2 O x x x 10-2 UO 2 /Y 2 O 3 /Pd 4.3 x x x 10-2 UO 2 /Pd 6.6 x x x 10-2

73 ln(k(simfuel)/k(uo 2 )) Relative rate constant as a function of standard potential 0-0,5-1 -1,5-2 -2,5-3 -3,5-4 -4, ,2 0,4 0,6 0,8 1 E o (Oxidant) (V vs NHE)

74 U(VI) concentration μm SIMFUEL vs UO 2 Oxidants k ox (UO 2 ) m s -1 k ox (SIM) m s -1 O x x H 2 O x x CO x x 10-5 OH 1.7 x x irraditation time (min) This explains the g- radiolysis results: Lower impact of molecular oxidants!

75 Corrosion of copper in groundwater

76 Corrosion of copper in groundwater Heavily debated but not possible according to thermodynamics Radiation could have an effect

77 Radiation induced corrosion of Copper Å. Björkbacka, S. Hosseinpour, M. Johnson, C. Leygraf, M. Jonsson, Rad. Phys. Chem. 2013, 92, 80-86

78 Unirradiated reference

79 g-irradiated: 43 kgy

80 Radiation induced corrosion of copper The Cu-concentration in solution increases with increasing g-dose

81 Copper in solution as a function of absorbed dose

82 Confusing observation More copper is released than can be accounted for by aqueous radiolysis. The copper concentrations are higher than the solubility of the oxides that are formed.

83

84 Bentonite Radionuclide diffusion Bentonite erosion Colloid formation Colloid diffusion in compacted bentonite Effects of ionizing radiation

85 Radionuclide diffusion Cationic radionuclides are adsorbed to the bentonite. Strong retention No or very little retention of anionic radionuclides

86 Bentonite erosion Flowing water and low salinity could result in erosion of the barrier (has been shown experimentally)

87 Colloid formation At low ionic strength (e.g. glacial melting water) bentonite colloids can be formed Bentonite colloids can act as carriers of radionuclides and influence migration

88 Colloid diffusion in compacted bentonite Compacted bentonite acts as an efficient filter for colloids. However Humic colloids are flexible and can change conformation (not filtered) Small enough colloids can pass through compacted bentonite

89 Effects of ionizing radiation on the integrity of the bentonite barrier 1. Increases bentonite colloid stability 2. Lowers the cation sorption capacity 3. Changes the Fe(II)/Fe(III) ratio

90 Microbiology Has an impact on numerous processes in a deep repository (not covered in this lecture)

91 Is KBS-3 a safe concept?

92 End of part 2