1. Leaching experiments on alpha-doped UO2 performed on low- activity samples (233U dopant) simulating 103-105 years old fuel and reproducing some relevant repository conditions;
2. Highly reducing repository conditions will be examined using the H2 overpressure autoclave;
3. Investigation of chemistry vs. radiolysis effects by comparing the leaching behaviour of alpha-doped (238Pu) UO2 to MOX (239Pu);
4. Examination of 10 years old SUPERFACT fuel to simulate behaviour of spent fuel during ~100 years of storage;
5. Electrochemical studies: comparison irradiated UO2-MOX vs. alpha-doped UO2, and development of the noise technique;
6. The dynamic leaching apparatus will be used on irradiated fuels;
7. Alterations induced by accumulation of alpha-damage and He in spent fuel during storage will be studied. Both short-term and long-term disposal will be considered;
8. He release behaviour from MOX will be studied by performing annealing tests using the Knudsen cell apparatus combined with micro-structural TEM examinations.
1. U, Pu concentrations vs. leaching time in carbonated and NaCl solutions;
2. U concentrations vs. leaching time under H2 overpressure;
3. Redox, pH evolution during leaching;
4. Dissolution rates of irradiated fuel measured using dynamic leaching set up;
5. Corrosion potential of irradiated fuel and electrochemical noise measurements.
Specific deliverables to DGs:
- Alert function on safety relevant data related to long-term storage;
- Reports on leaching and dissolution rates of spent fuel under realistic conditions;
- Consultancy for EU policies with regard to waste management.
As a result of the research:
Publications, patents, upgrade of safety relevant databases, training of staff from Member State authorities.
Summary of 2000 Deliverables: 31/12/2000
The behaviour of spent nuclear fuel subject to extended interim storage (up to ~500 years) or to final disposal in a geologic repository is to be investigated. Alterations occurring in the fuel during storage, caused mainly by accumulation of alpha-decay damage and of helium in the structure, and basic corrosion/dissolution processes of spent fuel exposed to groundwater are studied. Focal point for these studies is the age of the fuel either at the time of retrieval from interim storage, or when it will be in contact with groundwater in the final repository. This latter event is very unlikely to occur before storage times of the order of hundreds thousand years have elapsed, due to the multiple containment barriers which safely isolate spent fuel from the repository environment.
No "old" nuclear fuel is available today for studying, due to the young age of nuclear technology. The experiments aim at reproducing and studying realistic conditions governing the behaviour of spent fuel after relevant storage times. Alpha decay will play an important role, damaging the fuel structure and enhancing its dissolution through radiolysis of the water molecules. To study these effects, ITU fabricates and studies samples of uranium oxide containing different concentrations of short-lived alpha emitters (e.g. uranium-233 or plutonium-238). New experiments are now superimposing to these effects the opposite, dissolution inhibiting influence due to the high (~50 bar) hydrogen pressure, which also characterizes the repository. The (mostly inhibiting) effects of other relevant repository materials or conditions on the leaching behaviour of spent fuel are also investigated.
Highlight 2000: Alpha-radiolysis studies for spent fuel characterisation in view of long term storage.
Leaching experiments on alpha-doped UO2 under anoxic conditions with continuous monitoring of pH and redox potential allowed a direct measurement of effects associated with the alpha-radiolysis of the water. With increasing leaching time, increasingly higher redox potential values, e.g. more oxidising conditions, occur in the solutions for the alpha-doped samples, due to the radiolytic production of oxidising species. Correspondingly, higher fractions of U were dissolved in the case of alpha-doped UO2 compared to undoped UO2. A clear dependence of this process on the alpha-activity of the materials was observed by adopting relatively low values of the S/V (sample surface/solution volume) ratio. A relatively small alpha-activity, however, was seen to severely increase leach rates under experimental conditions characterised by a high S/V ratio, as could be expected for some real storage scenarios.
Electrochemical corrosion testing enable the dissolution rate-determining step to be clarified under repository-relevant conditions ranging from strongly reducing to oxidising conditions. The installation for the dynamic dissolution studies has been completed in 2000.
Output Indicators and Impact
- Participation to main spent fuel shared cost action in 5th FWP;
- Attendance of spent fuel workshops.
Global output : 3 publications in refereed journals and 5 conference participation.
Summary of the project
Most of the existing experimental data on the long-term behaviour of spent fuel were obtained using non-irradiated uranium oxide. This is the reason why Member State authorities in charge of intermediate or long-term fuel storage press for investigations of spent fuel under realistic storage conditions (granite, salt, clay).
The objective of the work is to better understand the basic mechanisms and kinetics of leaching of irradiated uranium oxide and mixed oxide fuel. In addition to real fuel samples, well defined synthetic fuel forms will be used and solid state physical techniques will be employed, for example, for single effect studies in order to assist the interpretation of the results obtained with spent fuel.
For the determination of the source term of spent fuel, instruments and methods will be further improved. Experimental results will be compared with theoretical predictions. An important new item in the project will be the study of the influence of radiolysis on spent fuel dissolution.
For the management of spent fuel discharged from power reactors, Member States of the European Union have adopted one (either one or a combination) of the following approaches:
- Intermediate dry or wet storage with subsequent direct conditioning for final disposal in geological formations;
- Intermediate storage followed by reprocessing, e.g., recycling of uranium and plutonium and subsequent conditioning of highly active waste for final disposal.
For the direct long-term storage of spent fuel, e.g., without reprocessing, the behaviour and interaction of the fuel with its surroundings need to be well known and understood, based on realistic conditions, in particular the oxidation corrosion and leaching behaviour under various conditions.
Complementary to basic work carried out in European universities and research centres, the Institute is using its facilities and equipment for experiments with real spent fuel or other highly active waste forms.