Research activities on radioactive waste management and on the back-end of the nuclear fuel cycle performed by the Joint Research Centre of the European Commission

The Euratom Research and Training Programme contributes, within its portfolio of activities, to establish and improve the scientific basis of knowledge for the safe management of spent nuclear fuel and radioactive waste. This includes research and innovation activities undertaken by the Joint Research Centre (or JRC, the European Commission’s science and knowledge service) in its laboratories. This paper provides an overview and some highlights of the Joint Research Centre (JRC) activities which are dedicated to the safety of spent fuel and high level radioactive waste forms. The fields of experimental and modelling research address various stages of spent fuel management after discharge from the reactor core: cooling in the spent fuel pool; handling, transport, extended interim storage and retrieval thereafter; disposal in a deep geological repository and long term behaviour of the spent fuel/waste form after disposal. The safety of the “back-end” of nuclear fuel cycles which include U-Pu recycling and/or a “fully closed” cycle with minor actinides separation and transmutation is also a major area of research. Both normal operation and accident scenarios, which cause fuel degradation/melting, are investigated. Possible applications for legacy waste management, decommissioning, and safeguards are considered. The relevance of the research is linked to the possibility of investigating “real” spent fuel and highly radioactive compounds using JRC’s research infrastructure, which includes hot cells and shielded facilities, and state of the art experimental methods that are (in some cases) rare or even unique. The activities are performed in collaboration with partners and/or in the context of international initiatives. Opportunities and perspectives for enhanced cooperation, including access and sharing of infrastructure are being developed

Experimental system to displace radioisotopes from upper to deeper soil layers: chemical research

BACKGROUND: Radioisotopes are introduced into the environment following nuclear power plant accidents or nuclear weapons tests. The immobility of these radioactive elements in uppermost soil layers represents a problem for human health, since they can easily be incorporated in the food chain. Preventing their assimilation by plants may be a first step towards the total recovery of contaminated areas. METHODS: The possibility of displacing radionuclides from the most superficial soil layers and their subsequent stabilisation at lower levels were investigated in laboratory trials. An experimental system reproducing the environmental conditions of contaminated areas was designed in plastic columns. A radiopolluted soil sample was treated with solutions containing ions normally used in fertilisation (NO(3)(-), NH(4)(+), PO(4)(--- )and K(+)). RESULTS: Contaminated soils treated with an acid solution of ions NO(3)(-), PO(4)(--- )and K(+), undergo a reduction of radioactivity up to 35%, after a series of washes which simulate one year's rainfall. The capacity of the deepest soil layers to immobilize the radionuclides percolated from the superficial layers was also confirmed. CONCLUSION: The migration of radionuclides towards deeper soil layers, following chemical treatments, and their subsequent stabilization reduces bioavailability in the uppermost soil horizon, preventing at the same time their transfer into the water-bearing stratum

Hydrogen suppresses UO2 corrosion

Release of long-lived radionuclides such as plutonium and caesium from spent nuclear fuel in deep geological repositories will depend mainly on the dissolution rate of the UO2 fuel matrix. This dissolution rate will, in turn, depend on the redox conditions at the fuel surface. Under oxidative conditions UO2 will be oxidised to the 1000 times more soluble UO2.67. This may occur in a repository as the reducing deep groundwater becomes locally oxidative at the fuel surface under the effect of alpha-radiolysis, the process by which alpha-particles emitted from the fuel split water molecules. On the other hand, the groundwater corrodes canister iron generating large amounts of hydrogen. The role of molecular hydrogen as reductant in a deep bedrock repository is questioned. Here we show evidence of a surface-catalysed reaction, taking place in the H-2-UO2-H2O system where molecular hydrogen is able to reduce oxidants originating from alpha-radiolysis. In our experiment the UO2 surface remained stoichiometric proving that the expected oxidation of UO2.00 to UO2.67 due to radiolytic oxidants was absent. As a consequence, the dissolution of UO2 stopped when equilibrium was reached between the solid phase and U4+ species in the aqueous phase. The steady-state concentration of uranium in solution was determined to be 9 x 10(-12) M, about 30 times lower than previously reported for reducing conditions. Our findings show that fuel dissolution is suppressed by H-2. Consequently, radiotoxic nuclides in spent nuclear fuel will remain immobilised in the UO2 matrix. A mechanism for the surface-catalysed reaction between molecular hydrogen and radiolytic oxidants is proposed. (C) 2009 Elsevier Ltd. All rights reserved

Assessment of a Compton-event Suppression gamma-Spectrometer for the Detection of Fission Products at Trace Levels.

Abstract not availableJRC.E-Institute for Transuranium Elements (Karlsruhe

Corrosion of high burn-up structured UO2 fuel in presence of dissolved H-2

The influence of high burn-up structured material on UO2 corrosion has been studied in an autoclave experiment. The experiment was conducted on spent fuel fragments with an average burn-up of 67 GWd/tHM. They were corroded in a simplified groundwater containing 33 mM dissolved H-2 for 502 days. All redox sensitive elements were reduced. The reduction continued until a steady-state concentration was reached in the leachate for U at 1.5 x 10(-10) M and for Pu at 7 x 10(-11) M. The instant release of Cs during the first 7 days was determined to 3.4% of the total inventory. However, the Cs release stopped after release of 3.5%. It was shown that the high burn-up structure did not enhance fuel corrosion. (C) 2009 Elsevier B.V. All rights reserved

Corrosion of irradiated MOX fuel in presence of dissolved H-2

The corrosion behaviour of irradiated MOX fuel (47 GWd/tHM) has been studied in an autoclave experiment simulating repository conditions. Fuel fragments were corroded at room temperature in a 10 mM NaCl/2 mM NaHCO3 solution in presence of dissolved H-2 for 2100 days. The results show that dissolved H-2 in concentration 1 mM and higher inhibits oxidation and dissolution of the fragments. Stable U and Pu concentrations were measured at 7 x 10(-10) and 5 x 10(-11) M, respectively. Caesium was only released during the first two years of the experiment. The results indicate that the UO2 matrix of a spent MOX fuel is the main contributor to the measured dissolution, while the corrosion of the high burn-up Pu-rich islands appears negligible. (C) 2009 Elsevier B.V. All rights reserved

Corrosion of high burn-up structured UO2 fuel in presence of dissolved H-2

The influence of high burn-up structured material on UO2 corrosion has been studied in an autoclave experiment. The experiment was conducted on spent fuel fragments with an average burn-up of 67 GWd/tHM. They were corroded in a simplified groundwater containing 33 mM dissolved H-2 for 502 days. All redox sensitive elements were reduced. The reduction continued until a steady-state concentration was reached in the leachate for U at 1.5 x 10(-10) M and for Pu at 7 x 10(-11) M. The instant release of Cs during the first 7 days was determined to 3.4% of the total inventory. However, the Cs release stopped after release of 3.5%. It was shown that the high burn-up structure did not enhance fuel corrosion. (C) 2009 Elsevier B.V. All rights reserved