Principal Research Results

The Spent Fuel Integrity during Dry Storage−An Evaluation of Hydrogen Migration in Cladding Existing Axila Temperature Gradient−
BackgroundAt present, spent nuclear fuels coming from nuclear plant are stored at reactor site areas before shipment to reprocessing
plants. Japanese government and electric utilities have a plan to operate interim dry storage facilities away from reactor sites till 2010.
An experiment using the real spent fuel cladding and following evaluation of fuel integrity focused to dry storage is necessary prior tothe start of operation of the interim dry storage facilities away from reactor sites.
One of the phenomena that will affect fuel integrity during the dry storage is the migration of hydrogen in cladding in theaxial direction. If there is a hydrogen pickup in cladding at irradiation, hydrogen will move from hotter to colder cladding region in theaxial direction under fuel temperature gradient during storage and will be precipitated as hydride, and consequently hydride brittlenessmay take place in the cladding (Fig.1).

Objectives
To carry out the experiment focused to hydrogen migration relating to hydride brittleness and obtain the data to evaluate thehydrogen migration by using real twenty years dry (air) stored spent PWR-UO2 fuel rods. To evaluate a hydrogen distribution in thecladding after dry storage by calculation.

Principal Results
The hydrogen diffusion coefficient, the solubility limit of cladding and the heat of transport of hydrogen that relates to migration caused by temperature gradient are necessary to evaluate a hydrogen migration in the cladding during storage. To obtain these
data, the hydrogen migration experiments were carried out on the real twenty years stored PWR-UO2 fuels (burn-up: 31MWd/kgHM
and 58MWd/kgHM) and hydrogen distribution in the cladding after dry storage was predicted by the calculation.
(1) The hydrogen concentration measurements were carried out on cladding specimens after hydrogen migration experiment (Fig.2).
In this study, the heat of transport, diffusion coefficient and solubility limit of real stored spent fuel cladding, whose data werenone or few, were obtained by fitting the hydrogen flux equation based on Fick’s law to hydrogen concentration distribution obtained by measurement. The heat of transport of irradiated cladding was significantly larger than that of unirradiated cladding. Thismeans that hydrogen in irradiated cladding can move easier than that in unirradiated cladding (Table 1).
(2) There are few differences in the diffusion coefficients and solubility limits between the irradiated cladding and unirradiatedcladding (Fig.3). The conservative safety evaluation on hydrogen migration can be possible using the diffusion coefficient and solubility limit of unirradiated cladding.
(3) The hydrogen distribution in the cladding after dry storage for forty years was evaluated by one-dimensional diffusion calculation
using the measured values. In calculation, the maximum values as the heat of transport, diffusion coefficient and solubility limit
were selected to evaluate conservatively. The axial hydrogen migration was not significant after dry storage for forty years in helium atmosphere, except at fuel rod ends (Fig.4).
It is concluded that hydrogen migration has few effect on the fuel integrity during dry storage.
Main Researcher: Akihiro Sasahara,
Research Scientist, Nuclear Power Generation Technology Sector, Nuclear Technology Research Laboratory
Reference
A.Sasahara, T.Matsumura, “Post irradiation experiments of twenty years stored spent fuels”, Storage of Spent Fuel from Power
Reactors, IAEA, 2003, June 2-6, Vienna.

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C. Harmonization of energy and environment

Fig.1 Effect of hydrogen on spent fuel cladding

Fig.2 Hydrogen concentration in the irradiatedFig.3 Diffusion coefficient of irradiated andcladding after migration experiment

unirradiated cladding
Table 1 The heat of transport obtained by this study
SampleFig.4 Hydrogen distribution by calculationin the cladding after forty years storage
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Heat of transport (kcal/mol)

Unirradiatedcladding

4.5〜 9.7

Irradiated andstored cladding

10.6〜17.7