D.H. Wegen

Use of UO2 films for electrochemical studies

Abstract UO 2 films have been prepared by dc reactive sputtering of a uranium metal target in an Ar/O 2 atmosphere. We have used the films deposited on gold substrates as working electrodes for electrochemical investigations as simulating the surfaces of fuel pellets. Film composition was determined by photoelectron spectroscopy (XPS and UPS) and X-ray diffraction (XRD). The oxide stoichiometry as a function of deposition conditions was determined and the appropriate conditions for UO 2.0 formation established. AC impedance and cyclic voltammetry measurements were performed. A double RC electrical equivalent circuit was used to fit the data from impedance measurements, similar to those used in unirradiated UO 2 or spent fuel pellets. However due to the porosity or adhesion defects on the thin films that permitted a direct contact between the solution and the gold substrate, we were obliged to add a contribution simulating the water–gold system. Cyclic voltammetry measurements show the influence of pH on the dissolution mechanism. Alkaline solutions permit the formation of an oxidised layer (UO 2.33 ) which is not present in the acidic solutions. In both pH=2 and pH=6 solutions, a U VI species layer is formed.

Promises and pitfalls in the reliable determination of 233U using high resolution ICP-OES

This study highlights that the currently employed analytical approach for the ICP-OES determination of 233U at λ = 385.950 nm is neither specific for this U isotope nor reliable and provides potential solutions for a selective and reliable quantification of 233U, 235U, and 238U using high resolution ICP-OES. While being a sensitive U wavelength region, the small isotopic line splitting of 233U (λ = 385.950 nm), 235U (λ = 385.955 nm) and 238U (λ = 385.958 nm) cannot be resolved with any commercial high resolution optical spectrometer. In addition, the width of 5.67 pm of the 233U peak, defined as full width at half maximum, is particularly larger than the 238U peak width of ∼4.3 pm, thus indicating splitting of the 233U emission signal into multiple peaks. In contrast, superior U isotopic information can be obtained from isotope specific emission lines centred around λ = 411.6 nm and λ = 424.4 nm with a spectral distance between 233U and 238U of 25 pm and 39 pm, respectively. The 233U signal of both wavelength regions, however, suffers from a potential interference with thorium that requires peak deconvolution for accurate quantification of 233U.

Reducing uncertainties affecting the assessment of the long-term corrosion behavior of spent nuclear fuel.

Reducing the uncertainties associated with extrapolation to very long term of corrosion data obtainable from laboratory tests on a relatively young spent nuclear fuel is a formidable challenge. In a geologic repository, spent nuclear fuel may come in contact with water tens or hundreds of thousands of years after repository closure. The corrosion behavior will depend on the fuel properties and on the conditions characterizing the near field surrounding the spent fuel at the time of water contact. This paper summarizes the main conclusions drawn from multiyear experimental campaigns performed at JRC-ITU to study corrosion behavior and radionuclide release from spent light water reactor fuel. The radionuclide release from the central region of a fuel pellet is higher than that from the radial periphery, in spite of the higher burnup and the corresponding structural modifications occurring at the pellet rim during irradiation. Studies on the extent and time boundaries of the radiolytic enhancement of the spent fuel corrosion rate indicate that after tens or hundreds of thousands of years have elapsed, very small or no contribution to the enhanced corrosion rate has to be expected from α radiolysis. A beneficial effect inhibiting spent fuel corrosion due to the hydrogen overpressure generated in the near field by iron corrosion is confirmed. The results obtained so far point toward a benign picture describing spent fuel corrosion in a deep geologic repository. More work is ongoing to further reduce uncertainties and to obtain a full description of the expected corrosion behavior of spent fuel.

Spent Fuel as Waste Material

In general, this chapter deals with spent LWR fuel, mainly UO 2 and mixed oxide fuel, with respect to the change of its chemical and mechanical properties during long-time storage in a deep underground repository.

Leaching of 53 MW/d kg U spent nuclear fuel in a flow-through reactor

Abstract The dissolution behaviour of powdered commercial spent fuel (UO2 with burn-up of 53 MW/d kg U) has been studied in a carbonate-containing solution ([HCO3-] =0.001 mol dm-3) by using a flow-through reactor specially designed for the use in a hot cell. This method allows studying spent fuel dissolution while avoiding the parallel process of secondary solid phase formation. The dissolution behaviour of U, Np, Pu, Sr and Cs was studied. The main trend of the results obtained in this work is that only neptunium releases congruently with uranium (FIAPNp/FIAPU=1.21±0.01) because both strontium and caesium have higher FIAP values (FIAPSr/FIAPU=2.3±0.8; FIAPCs/FIAPU=5±1) and plutonium lower (FIAPPu/FIAPU=0.07±0.02). The FIAP value for uranium at the steady-state is 4(±2)×10-4.

Release of Radiotoxic Elements from High Burn-Up UO2 and MOX Fuel in a Repository

In a spent fuel repository the processes that govern the release of radionuclides are dissolution and transport in a possible groundwater flow. The cladding will be the last barrier before the water comes into contact with the fuel, namely with the outer rim of the pellet. Here the heterogeneity of the material due to the irradiation process is responsible for a complex release process. Fission products and minor actinides inventories are considerably higher at the pellet periphery as a result of increased epithermal neutron capture and of migration in the case of the volatile fission products. The present paper gives a review of experimental activities at the Institute for Transuranium Elements (ITU). Both single effects studies and integral tests are carried out to study the behavior of spent fuel under storage conditions. Leaching of irradiated UO 2 (up to 50 GWd/tU) and MOX (up to 25 GWd/tU) fuel rods with preset cladding defects at 100°C under anoxic or reducing conditions should simulate the realistic case of groundwater coming into contact with a spent nuclear fuel repository. For all main radionuclides the release process can be described considering a two-step dissolution mechanism that includes the initial dissolution of an oxidized layer present on the fuel surface followed by a long-term oxidative matrix dissolution. By means of α-doped ( 238 Pu) UO 2 it could be demonstrated, that radiolysis has a significant influence on this dissolution. Especially high initial release rates were found for the volatile cesium and iodine for the reasons mentioned above. Besides the conventional leaching experiments electrochemical techniques are used to investigate for instance the complex corrosion behavior of the heterogeneous MOX fuel materials or the influence of α-radiolysis on spent fuel dissolution. In the integral tests mentioned above with large S/V values, reprecipitation of U is likely to happen. Therefore special dynamic test are conducted where this reprecipitation is prohibited and true U solubility can be determined. Thin layer of UO 2 and (U,Pu)O 2 doped with various fission products and minor actinides are prepared to study the influence of these elements on the matrix dissolution. When Cs is for instance co-deposited, the U oxidation state changes from U 4+ to U 6+ for the same O 2 pressure possibly indicating a stable Cs uranate. This could be an indirect proof of the existence of such a species in irradiated fuel (e.g. at the grain boundaries).

RN Fractional Release of High Burn-Up Fuel: Effect of HBS and Estimation of Accessible Grain Boundary

The so-called Instant Release Fraction (IRF) is considered to govern the dose released from Spent Fuel repositories. Often, IRF calculations are based on estimations of fractions of inventory release based in fission gas release [1]. The IRF definition includes the inventory located within the Gap although a conservative approach also includes both the Grain Boundary (GB) and the pores of restructured HBS inventories. A correction factor to estimate the fraction of Grain Boundary accessible for leaching has been determined and applied to spent fuel static leaching experiments carried out in the ITU Hot Cell facilities [2]. Experimental work focuses especially on the different properties of both the external rim area (containing the High Burn-up Structure (HBS)) and the internal area, to which we will refer as Out and Core sample, respectively. Maximal release will correspond to an extrapolation to simulate that all grain boundaries or pores are open and in contact with solution. The correction factor has been determined from SEM studies taking into account the number of particles with HBS in Out sample, the porosity of HBS particles, and the amount of transgranular fractures during sample preparation.

Mechanisms governing the release of radionuclides from spent nuclear fuel in geological repository: major outcomes of the European Project SFS

European Commission supported a wide research project entitled “Spent Fuel Stability under repository conditions” (SFS) within the 5 th FWP, the aim of which was to develop a common understanding of the radionuclides release from spent nuclear fuel in geological disposal and build a RN release model in order to assess the fuel performance. This project achieved by the end of 2004 focuses both on the Instant Release Fraction (IRF) model and the Matrix Alteration Model (MAM). A new IRF model was developed based on the anticipated performances of the various fuel microstructures (gap, rim, grains boundaries) and the potential diffusion of RN before the canister breaching. However, this model lets the choice to the end-user about the degree of conservativeness to consider. In addition, fuel alteration has been demonstrated to be linked to the production of radiolytic oxidants by water radiolysis at the fuel interface, the oxidation of the fuel interface by radiolytic oxidants and the subsequent release of uranium under the influence of aqueous ligands. A large set of experimental data was therefore acquired in order (i) to upgrade the current radiolytic kinetic scheme, (ii) to experimentally correlate the fuel alteration rate and the fuel specific alpha activity by performing experiments on alpha doped samples, (iii) to experimentally test the potential inhibitor effect of hydrogen on fuel dissolution. Based on these results, a new MAM was developed, which was also calibrated using the experiments on inactive UO 2 samples. This model was finally applied to representative granitic, salt and clayey environment to predict spent fuel long-term fuel performance.

Summary of the euratom collaborative project FIRST-nuclides and conclusions for the safety case

Abstract A consortium of 10 partners from seven European countries and the European Commission collaborated in investigating the short-term release of radionuclides from disposed spent nuclear fuel upon canister failure. The Collaborative Project FIRST-Nuclides was implemented in the scope of the 7th Euratom Framework Programme in the period from 2012 to 2014. The objectives and organization of the project are presented, as well as the experiments with highly radioactive materials under investigation. The outcome of the project summarizes the measured instant release fraction (IRF) of safety-relevant isotopes from high burnup spent UO2 nuclear fuels (SNFs). Specifically discussed are the dependencies of the IRF on the sample properties, the gap and grain boundary releases, and the behavior and IRFs of elements such as cesium, iodine, and selenium. The IRFs of nonstandard SNFs were also investigated. The summary is complemented by the presentation of the modeling approaches within the project.