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Spent Fuel Characterisation

Objectif

Specific Objectives 8.1.2.1. Fuel Corrosion and Radiolysis Studies Leaching experiments on alpha-doped UO2 (carbonated water, RT, anoxic/reducing conditions) with H2, O2 measurement and in presence of Fe) with H2, O2 measurement Contrasting effects of H2O2 and other ionic species on the dissolution behaviour of UO2, SIMFUEL Leaching of unirradiated MOX under reducing conditions (carbonated water, RT, co-precipitated MOX) Leaching of actinide-containing pyrochlores and/or ZrO2 and of Th-MOX 8.1.2.2. Radiation Damage and Gas Behaviour in Spent Fuels Irradiation with fission products (Xe) of transmutation (ZrN) and/or storage matrices (Nd2Zr2O7) at IRRSUD beamline, GANIL and TEM investigation on ion- irradiated Nd2Zr2O7 and ZrN Investigation of the alpha-damage annealing in MOX by thermal diffusivity measurements.

Comparative study of the helium vs. fission gas kinetics in nuclear fuels by annealing in a Knudsen Cell (experimental and evaluation) Complete the characterization of Cm-doped pyrochlore He release behaviour, effusion tests on actinide-bearing IM (pyrochlores) Thermodynamic solubility of helium in UO2 and IMF's (YSZ) 8.1.2.3. Spent Fuel Corrosion Chemistry and Dynamic Leaching Testing of spent and unirradiated UO2 and MOX-fuel: electrochemical corrosion tests SUPERFACT fuel: - examination of corroded fuel rods; optical & SEM/EPMA analysis (XRD) Autoclave leaching of spent fuel under hydrogen pressure. Dynamic leaching of spent UO2 and MOX-fuel: Cutting & preparation and instant release of UO2 spent fuel 65 MWd/kgU and MOX spent fuel 44 MWd/kgU Anticipated milestones and schedule.
Planned Deliverables 8.1.2.1. Fuel Corrosion and Radiolysis Studies U, Pu concentration vs. time and H2, O2 concentration vs time; redox evolution U, H2, O2, relevant fission products (SIMFUEL) concentration vs. time; redox evolution (A03) U, Pu concentration vs. time for MOX fuel Actinide and Th, Pu concentration vs. time; comparison with other waste forms 8.1.2.2.

Radiation Damage and Gas Behaviour in Spent Fuels Result report on TEM investigation on ion irradiated Nd2Zr2O7 and ZrN Thermal diffusivity vs. T vs. time for alpha-doped UO2 and old MOX samples Comparison of gas release curves for irradiated MOX, alpha-doped UO2, HBRP Hardness; oxygen potential; XRD,; specific heat; annealing behaviour; effusion of pyrochlore He release curves; comparison of effusion between actinide loaded-unloaded samples Preliminary results: qualitative effusion tests 8.1.2.3. Spent Fuel Corrosion Chemistry and Dynamic Leaching Characterisation of alpha-doped UO2, MOX, HBU UO2 Data on SUPERFACT fuel analysis Ac and FP concentration in leachate of alpha-doped UO2, MOX, HBU UO2 vs hydrogen Dissolution rates of the matrix and dissolution mechanisms of the different radionuclides.
Summary of the Action:

The work aims at better understanding 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 (such as radiolysis, inter-surface state, groundwater composition and redox conditions) 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. For in-situ measurements, electrochemical corrosion studies and thin film simulation experiments will be used. Rationale The management of spent nuclear fuel and highly radioactive waste is of considerable importance as about 45000 t of irradiated fuel is in interim storage in the EU. Decisions are needed on the options to be explored and to be pursued (example, law 1991 of French Parliament).

The two approaches for spent nuclear fuel management favoured by the Member States of the European Union can be schematised as follows: intermediate storage with subsequent conditioning for final disposal, and intermediate storage with subsequent reprocessing (partitioning) for lowering the radiotoxicity (transmutation) before final disposal in geological formations. Detailed design proposals for final repositories for spent fuel are now under regulatory review in some Member States as a part of a stepwise procedure. In the mean time, there are adequate facilities for safe short-term storage of spent fuel and high level waste. Development of a common understanding of the frequently used concept of "reasonable assurance of safety" for the long-term, post-closure phase of final repositories is highly desirable to ensure reasonably equivalent legal interpretations among EU countries in the environmental impact assessment and licensing procedures.

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Coordinateur

Institute for Transuranium Elements
Contribution de l’UE
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Adresse
PO Box 2340
D-76125 Karlsruhe
Allemagne

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