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

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Karel Lemmens, Christelle Cachoir, Karine Ferrand, Thierry Mennecart, ... The waste form contributes to the safety of the disposal system because it ... – PowerPoint PPT presentation

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


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
Temperature
16 C
lt100 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
00015 000 years)
pH
Temperature
8.3-8.6
water
Radionuclides
13.5
Overpack corrosion products (magnetite ?)
Altered concrete (?)
7
Stage 2 Vitrified waste after overpack failure
situation at first water contact fast
dissolution expected
(altered) concrete
magnetite
Overpack
H2 gas Water vapour
Cement water pH 13.5
glass
crack
8
Stage 2 vitrified waste after overpack
failurepH decreasing processes pH 13.5 ? pH 9
-10 (?)
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 gt one order
    magnitude lower
  • Dissolution rate at pH 9-10 gt 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
16C
lt 25C
Radionuclide release
13.5
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 (lt 2500
years)No spent fuel dissolution
Temperature
pH
16 C
8.3-8.6
lt100 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
1000015000 years)
pH
Temperature
16C
8.3 8.6
lt25C
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)
concrete
diffusionsorption
magnetite
Overpack ( iron insert cladding)
H2 gas Water vapour
Gap
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
U(IV)
H2
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
1638C
lt25C
Reducing conditions
Fe(II)
13.5
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
altered concrete
threshold activity for oxidative dissolution ?
magnetite
U sorption
a
H2O2
Fe(II) effect ?
U(IV)
U(VI)
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 ! Im finished !
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