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3.2 Spent fuel
The release of radionuclides from spent fuel is classically described by two terms:
The Instant Release Fraction (IRF) is assumed to be immediately released under repository
conditions upon contact with groundwater. This concerns mainly the fraction of the radionuclide
inventory located in the gap between fuel and cladding, and in the grain boundaries.
The radionuclides that are embedded in the spent fuel matrix will dissolve slowly as a result
of the matrix alteration.
The main results related to these two terms are as follows:
Instant release fraction and long term fuel evolution
The programme focussed on obtaining new experimental IRF data and on updating the IRF model
developed in the SFS project (FP5, [3]). This model considered both the initial IRF after irradiation
and its potential increase with time. Significant results have been obtained on both aspects.
First, regarding the characterisation of the IRF of “young” irradiated fuels, new IRF values have
been obtained both for high burn-up UOX and for MOX fuel, even though the measured values may
be overestimated due to inclusion of fractions from oxidized fuel surfaces. In particular, these results
seem to evidence that in case of water contact to the centre of the fractured fuel, the rim zone
of high burnup UOX fuel contributes less to the IRF, compared to the fuel in the centre of the fuel
rod, although the opposite was expected. Release of 36 Cl form UOX spent fuel may be much faster
than previously anticipated.
Second, one of the critical questions was whether this IRF will remain constant or grow with time
due to decay enhanced diffusion and fracturing by helium ingrowth. It was demonstrated that this
diffusion process is so slow that it will not significantly increase the IRF. Furthermore, a new micromechanical
model was developed to assess the impact of He accumulation due to decay. Results
show that He accumulation will lead to bubble formation, but for UOX fuel, the critical bubble
pressure will most probably not be sufficiently high to fracture the fuel. Hence the present day surface
area and IRF values of UOX fuel are expected to remain roughly constant with disposal time,
at least in absence of water. The situation for MOX fuels is more complex and still needs to be assessed.
So, NF-PRO has significantly reduced the uncertainties regarding the IRF quantification and a new
pessimistic and realistic set of instant release values has been provided for a number of potentially
mobile radionuclides.
Spent fuel matrix dissolution in the presence of a corroding container
Both studies of the dissolution of UO2 doped with emitters to simulate the radiation field of spent
fuel of a few thousand years, and of real spent fuel were carried out under different conditions
which simulate European deep repository conditions. The following results were obtained:
Tests with doped material have shown very low dissolution rates, calculated by the isotope
dilution method, while the measured 238 U concentrations are very low. For performance
assessment this may imply that under reducing conditions, solubility controlled dissolution
models cannot be used to describe long-term corrosion, and the radiation field of 3000-
10000 y old fuel does not promote the oxidative dissolution of fuel.
Tests with doped UO2 in the presence of bentonite confirm the counteracting effect of dissolved
hydrogen on the UO2 dissolution, but also suggest important sorption of U on the
clay. Hydrogen is assumed to slow down U(VI) dissolution, but not U(IV) dissolution. The
latter is assumed to continue until saturation of the sorption sites of the bentonite.
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