Spent nuclear fuel

Spent nuclear fuel

Mats Jonsson

KTH Chemical Science and Engineering, Royal Institute of Technology, Stockholm, Sweden

II Letnia Szkoła Energetyki i Chemii Jądrowej

Spent nuclear fuel

Mats Jonsson, KTH Chemical Science and Engineering, Royal Institute of Technology, Stockholm, Sweden

E-mail: matsj@kth.se

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

A reminder

> 100 000 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 100 000 years???

Safety assessment

• Extreme extrapolations

• Impossible to perform experiments on real materials (a typical scenario is that a canister fails after 1000- 3000 years)

• Very difficult to predict changes in climate and in human activities

• How do we communicate with future generations?

Purpose of a geological repository

• To isolate the radioactive waste from the biosphere until the radioactivity is comparable to a uranium ore.

Proposed repository concepts

• Deep boreholes

• Clay formations

• Brines (salt)

• Granitic bedrock

The KBS-3 concept

The repository site in Sweden

(Forsmark)

Site investigations

• Oskarshamn vs Forsmark

• Both municipalities wanted to host the repository!

Fuel, canister and bentonite

Barriers

0. The fuel

1. The canister

2. The bentonite clay 3. The bedrock

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 30-40 years

• Encapsulation

• Final/long-term storage in deep repository

Spent nuclear fuel in Sweden

CLAB in Oskarshamn

The fuel

• UO₂ has very low solubility in reducing groundwater

The canister

• Mechanical support

• Corrosion resistance

The bentonite clay

• Mechanical support

• Self sealing

• Diffusion barrier

The bedrock

• Low flow

• Retention of radionuclides

Processes that can influence the safety of a repository

• The fuel

• The canister

• The bentonite

Dissolution of spent nuclear fuel

Dissolution of spent nuclear fuel

• Instant release

• UO₂ matrix dissolution

• The cladding is assumed to be gone (in a typical scenario)

Instant release

• Readily soluble fission products at the surface are rapidly released to the groundwater.

The Fuel Matrix (Spent Nuclear Fuel)

• ~95 % UO₂

• BUT ALSO

• Highly radioactive (depending on age)

• Insoluble noble metal inclusions (fission products)

• Rare earth oxides (fission products)

• Heavy actinides (activation)

Dissolution of UO

/SNF in groundwater

• UO₂ 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 UO₂ increases the matrix solubility by several orders of magnitude.

• Oxidants will be produced upon radiolysis of groundwater.

• Ox + UO₂  UO₂²⁺  UO₂²⁺(aq)

System description (somewhat simplified)

• Mixed radiation field (a, b and g) (gradient)

• Water radiolysis (including groundwater components)

• Surface reactions:

• 1. Oxidation

• 2. Reduction

• 3. Dissolution

• 4. Precipitation

• 5. Catalysis

• Diffusion

HETEROGENEOUS!

We have to start with something simpler

• Radiation induced dissolution of pure UO₂

• When we understand this system, the complexity can be increased:

• 1. Effect of H₂

• 2. Effect of noble metal inclusions

• 3. Effect of rare earth oxide doping

Radiation induced dissolution of UO

• Reactivity of aqueous radiolysis products (oxidants) towards UO₂

OH

, HO

, H

O

, O

, CO

UO

oxidation kinetics (powder)

• The rate constant for oxidation is a function of reduction potential.

-25 -20 -15 -10 -5 0

-0,5 0,5 1,5 2,5

E⁰ (V)

ln k

1 2

8

4 3 6 5

7

J. Nucl. Mater. 2003, 322, 242-248 Diffusion limit

Oxidation (by H

O

) vs Dissolution (Effect of HCO

)

0,00E+00 5,00E-07 1,00E-06 1,50E-06 2,00E-06 2,50E-06 3,00E-06 3,50E-06 4,00E-06 4,50E-06 5,00E-06

0 20 40 60 80 100 120

[HCO₃^-]/mM

k/m min-1

0,00E+00 5,00E-07 1,00E-06 1,50E-06 2,00E-06 2,50E-06 3,00E-06 3,50E-06 4,00E-06 4,50E-06 5,00E-06

0 0,2 0,4 0,6 0,8 1 1,2

[HCO3 -]/mM

k/m min-1

J. Nucl. Mater. 2006, 358, 202-208

Oxidation is the rate limiting step > 1 mM HCO

How to use the rate constants for radiation induced UO

dissolution

• Total rate of oxidation:

  



 

 ⁿ

ox

e ox

UO VI

U n

Ox k

dt A rate dn

1 )

(

2

2

J. Nucl. Mater. 2006, 355, 38-46

g-radiolysis of UO

-suspensions

0,0 1,0 2,0 3,0 4,0 5,0 6,0 7,0 8,0

0 50 100 150 200 250 300

Irradiation time (min) nU(VI) (μmol)

Calculated Experimental

J. Nucl. Mater. 2006, 355, 38-46

g-radiolysis

Good agreement!

Relative impact of (a-) radiolysis products

H₂O₂ O₂ O₂^•- HO₂^• CO₃^•- OH^• No additives 100.0 % 0.01 % 0 % 0.03 % 0 % 0 %

H₂ (40 bar) 99.9 % 0 % 0 % 0.02 % 0 % 0.03 %

H₂ (40 bar)

HCO₃^- (10 mM) 100.0 % 0 % 0 % 0 % 0.02 % 0 %

HCO₃^- (10 mM) 99.9 % 0.09 % 0 % 0 % 0 % 0 %

 

rate ^e

1 ox UO

U(VI)

2







 n

k dt A

dn ⁿ

ox

H

O

is the major oxidant!

J. Nucl. Mater. 2006, 355, 38-46

Next step

dx O

H G x

D r

x

x O

H











2 2

0

2

2

)

( )

( 



Environ Sci. Technol. 2007, 41, 7087-7093

Can this be simplified?

Rate of H

O

production

Radiolysis: Geometrical dose

distribution (based on RN inventory)

0 0,05 0,1 0,15 0,2 0,25 0,3 0,35 0,4 0,45 0,5

0 5 10 15 20 25 30 35 40

Distance from fuel surface (m)

Dose rate (Gy/s)

alpha beta total

J. Nucl. Mater. 2006, 359, 1-7

First test

• Simulation including:

1. Dose profile (H₂O₂ production)

2. H₂O₂ consumption at the UO₂ surface 3. Diffusion in one dimension (x)

Concentration profile as a function of time

0,00E+00 1,00E-10 2,00E-10 3,00E-10 4,00E-10 5,00E-10 6,00E-10 7,00E-10 8,00E-10 9,00E-10

0 20 40 60 80 100 120 140 160 180 200

Distance/µm [H2O2]/M

1 s

3 s

10 s

41 s

Steady-state

J. Nucl. Mater. 2008, 372, 32-35 and J. Nucl. Mater. 2008, 374, 286-289

Steady-state

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

Does the steady-state approach work?

Material (Dose rate)

p(H₂) [HCO₃^-] (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) 10 % U-

233

(99 Gy/h)

(Ar) 1.6810^-3 47 7.0510^-8 1.5010^-9 6.4010^-8 10 % U-

233

(99 Gy/h)

(1,2 % O2) 1.0710^-3 126 1.9610^-6 1.5610^-8 5.6910^-7 SF

(a = 828 Gy/h b = 31 Gy/h)

(Ar) 1010^-3 40 1.4210^-4 3.5410^-6 5,9610^-5

SF

(a = 828 Gy/h b = 31 Gy/h)

5 bar 1010^-3 376 0 0 1.70 x 10^-10

We are ready to increase the

complexity!

The effect of H

• H₂ inhibits dissolution of spent nuclear fuel.

• Why?

• Reduces the rate of H₂O₂ production (not sufficient)

J. Nucl. Mater. 2010, 396,163-169

The mechanism

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) 7087–7093.

g

e^-_aq H^ H₂ OH^ H₂O₂ H₂O

Spent Nuclear Fuel Fission products

Actinides

2 e^- 2 H⁺

H₂

a b

Oxidizing species Reduced species U(IV)

UO₂²⁺ (aq)

Fission products Actinides

HCO₃^- U(VI)

e

g

e^-_aq H^ H₂ OH^ H₂O₂ H₂O

Spent Nuclear Fuel Fission products

Actinides

2 e^- 2 H⁺

H₂

a b

Oxidizing species Reduced species U(IV)

UO₂²⁺ (aq)

Fission products Actinides

HCO₃^- U(VI)

e

But how efficient is the noble metal

catalyzed reduction by H

?

Experiment

• Uranium release from pellets containing Pd particles

• Exposed to H₂O₂

• 0 – 40 bar H₂

Noble metal catalyzed inhibition of H

O

induced dissolution

(Pd-doped UO

pellets)

-0,1 0,1 0,3 0,5 0,7 0,9 1,1 1,3

-10 10 30 50 70 90 110 130

rdiss/rox

ex pH₂ (bar)

0%

0.1 % 1%

3%

k = 10

m s

(diff. limited)

Martin Trummer, Sara Nilsson and Mats Jonsson, J. Nucl. Mater. 2008, 378, 55-59

Noble metal inclusions

• Also catalyze oxidation of pellets

Rare earth oxides

H

O

induced oxidative dissolution (doped UO

pellets)

0 5E-10 1E-09 1,5E-09 2E-09 2,5E-09 3E-09

Y2O3 Y2O3 / Pd Pd UO2

r_diss(U(VI)) / mol L-1 s-1

1 % Pd, 0.3 % Y2O3

N2

Radiation (g) induced dissolution

0 10 20 30 40 50 60 70 80

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000

c(Uranyl) / µM

time / min

N2 experiments

UO2 Y2O3 Y2O3/Pd Pd

H

O

consumption SIMFUEL/UO

0 0,5 1 1,5 2 2,5

0 50 100 150 200 250 300 350

H2O2 concentration (mM)

reaction time (min)

Rate as expected from k(H

O

)*

*

H

O

induced U(VI) dissolution SIMFUEL/UO

-10 0 10 20 30 40 50 60 70 80 90 100

0 100 200 300 400 500 600

concentration U(VI) μM

reaction time (min)

UO

SIMFUEL

J. Nucl. Mater. 2011, 410, 89-93

Radiation (g) induced dissolution of SIMFUEL/UO

-1 0 1 2 3 4 5 6 7 8

0 1000 2000 3000 4000 5000 6000 7000

U(VI) concentration μM

irraditation time (min)

UO

SIMFUEL

J. Nucl. Mater. 2011, 410, 89-93

What happens to H

O

?

H

O

+ UO

UO

+ 2 OH

(Dissolution)

H

O + ½ O

+ UO

Catalytic decomposition of H

O

H₂O₂  2HO^• (surface catalyzed)

HO^• + H₂O₂  H₂O + HO₂^• HO₂^• + HO₂^•  O₂ + H₂O₂ S 2H₂O₂  O₂ + 2H₂O

(ZrO₂) Cláudio Lousada and Mats Jonsson, J. Phys. Chem. C, 2010, 114 (25)

Detection of OH

OH^• + TRIS   Formaldehyde

H

O

+ ZrO

0 20000 40000

0 2 4 6

H₂O₂ HO^.

time (s)

[H 2 O 2 ] ( m M )

0,0 0,2 0,4

[HO .] ( m M )

Cláudio Lousada and Mats Jonsson, J. Phys. Chem. C, 2010, 114 (25)

Why do metal oxides catalyze decomposition of H

O

?

• Adsorption of H₂O₂ and OH^•

Metal oxide H

O

HO

HO

Hydroxyl radical affinity

0 0,02 0,04 0,06 0,08 0,1 0,12 0,14 0,16

1600 2600 3600 4600 5600

scavenged [HO·] mM

Irradiation time (s)

No Oxide Present ZrO2 TiO2 Y2O3

Lousada et al. to be published

TiO

<ZrO

<Y

O

0 20000

0.0 0.2 0.4 0.6 0.8 1.0

Y₂O₃ TiO₂

[H 2O 2]/[H 2O 2] 0

time (s)

UO

powder

0 0,05 0,1 0,15 0,2 0,25 0,3 0,35 0,4 0,45 0,5

0,00 1,00 2,00 3,00 4,00 5,00 6,00

0 1000 2000 3000 4000 5000

[OH] mM

[H2O2] mM

time (s)

UO

pellet

0 2 4 6 8 10 12 14 16 18 20

0 0,2 0,4 0,6 0,8 1 1,2 1,4 1,6 1,8 2

0 20000 40000 60000 80000 100000 120000

[U] microM

[HO] mM

time (s)

SIMFUEL pellet

0 2 4 6 8 10 12 14 16 18 20

0 0,2 0,4 0,6 0,8 1 1,2 1,4 1,6 1,8 2

0 20000 40000 60000 80000 100000 120000

[U] microM

[HO] mM

time (s)

Dissolution yields: [U(VI)]/[H

O

]

• UO

powder: 83 % (80 %)

• UO

pellet: 2 % (14 %)

• SIMFUEL pellet: 0 % (0.2 %)

• Catalytic decomposition is the main reaction path for H

O

on pellets!

• Why is UO

NOT oxidized by catalytically

produced OH?

What is the origin of the effect of dopants?

• Effects on k_cat

• And/or

• Effects on k_ox

OH production (pellets)

Uranium dissolution (pellets)

0 2 4 6 8 10 12 14 16 18 20

0 20000 40000 60000 80000 100000 120000

[ U(VI) ]normalized to westinghouse microM

time (s)

Pd Y UO2 YPd

Westinghouse SIMFUEL

Redox reactivity

• Oxidants: MnO₄^- and IrCl₆^2-

Oxidation of UO

by MnO

0 0,2 0,4 0,6 0,8 1 1,2

0 100 200 300 400

Norm. [MnO 4- ]

Time (min)

UO2

SIMFUEL

Activation energies (MnO

)

UO₂ (Westinghouse): 7.4 kJ mol^-1 SIMFUEL: 12.9 kJ mol^-1

Rate constants for pellet oxidation

Pellet k(H2O2)/min^-1 k(MnO4-)/min^-1 k(IrCl62-)/min^-1 UO2 (Westinghouse) 1.5 x 10^-4 3.3 x 10^-3 2.0 x 10^-2

SIMFUEL 1.4 x 10^-6 6.0 x 10^-4 1.7 x 10^-2

UO2 4.3 x 10^-5 6.2 x 10^-3 2.0 x 10^-2

UO2/Y2O3 1.3 x 10^-5 5.3 x 10^-3 2.1 x 10^-2

UO2/Y2O3/Pd 4.3 x 10^-6 4.6 x 10^-3 2.3 x 10^-2

UO2/Pd 6.6 x 10^-5 7.8 x 10^-3 2.2 x 10^-2

Relative rate constant as a function of standard potential

-5 -4,5 -4 -3,5 -3 -2,5 -2 -1,5 -1 -0,5 0

0 0,2 0,4 0,6 0,8 1

ln(k(SIMFUEL)/k(UO 2))

E^o (Oxidant) (V vs NHE)

SIMFUEL vs UO

Oxidants k_ox (UO₂) m s^-1 k_ox (SIM) m s^-1

O₂ 2.7 x 10^-11 1.3 x 10^-15

H₂O₂ 1.0 x 10^-8 9.5 x 10^-11

CO₃^•- 1.7 x 10^-5 1.7 x 10^-5

OH^• 1.7 x 10^-5 1.7 x 10^-5

-1 0 1 2 3 4 5 6 7 8

0 2000 4000 6000 8000

U(VI) concentration μM

irraditation time (min)

This explains the g- radiolysis results:

Lower impact of

molecular oxidants!

Corrosion of copper in groundwater

Corrosion of copper in groundwater

• Heavily debated but not possible according to thermodynamics

• Radiation could have an effect

Radiation induced corrosion of Copper

Å. Björkbacka, S. Hosseinpour, M. Johnson, C. Leygraf, M. Jonsson, Rad. Phys. Chem. 2013, 92, 80-86

Unirradiated reference

g-irradiated: 43 kGy

Radiation induced corrosion of copper

The Cu-concentration in solution increases with increasing g-dose

Copper in solution as a function of absorbed dose

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.

Bentonite

• Radionuclide diffusion

• Bentonite erosion

• Colloid formation

• Colloid diffusion in compacted bentonite

• Effects of ionizing radiation

Radionuclide diffusion

• Cationic radionuclides are adsorbed to the bentonite.

Strong retention

• No or very little retention of anionic radionuclides

Bentonite erosion

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

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

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

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

Microbiology

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

Is KBS-3 a safe concept?

End of part 2