Kastriot Spahiu

The fate of radiolytic oxidants during spent fuel leaching in the presence of dissolved near field hydrogen

Summary In most European disposal concepts, one expects large amounts of dissolved hydrogen produced by the anoxic corrosion of iron inside a damaged container. At repository temperatures (<100 °C), dissolved hydrogen is quite inert and is not expected to contribute to the redox capacity of the deep groundwaters. In a previous work from this laboratory we observed a large impact of dissolved hydrogen on the dissolution of the spent fuel in stainless steel autoclaves. In this work we report data on spent fuel dissolution obtained using quartz lined autoclaves to ensure that there is no contact between the solution and the metallic parts; we also made a number of other experimental improvements. The autoclave was filled with a solution 10 mM NaCl, 2 mM NaHCO3 and with H2+0.03% CO2 at a pressure of 0.5 MPa in the remaining free volume. The leaching of PWR spent fuel powder, placed in a gold basket, was studied during more than one year by analysing solution samples taken at regular time intervals. Special care was devoted to the study of the initial phase of the dissolution that was not investigated previously. In spite of the ten times lower hydrogen concentration as compared to the study in stainless steel autoclaves, extremely low concentrations of uranium (less than 10-9 M) were again measured in the solution samples; this was the case also for most of the redox sensitive fuel components. The uranium levels in solution remained practically constant during the whole leaching period, indicating the absence of any oxidative dissolution of the spent fuel matrix. In order to follow the fate of radiolytic oxidants, gas phase analyses were also carried out. The radiolytic oxygen levels in the autoclave measured after one year leaching were below the detection limit. A discussion of the fate of the radiolytic oxidants in these experiments and the mechanism of the hydrogen activation is also presented. The main conclusion is that for concentrations of dissolved hydrogen above 4 mM, no measurable oxidative dissolution of the UO2 matrix seems to occur in the studied systems.

On Mo-Ru-Tc-Pd-Rh-Te alloy particles extracted from spent fuel and their leaching behavior under Ar and H2 atmospheres

Summary In the present work, we extracted Mo-Tc-Ru-Pd-Rh-Te alloy particles from spent fuel by a non-oxidative selective leaching of the fuel matrix, characterized them and studied their leaching behavior under different redox conditions. After selecting the optimal experimental conditions by using a synthetic alloy, the radioactive alloy particles were extracted from spent fuel and characterized by using SEM (scanning electron microscopy)-WDS (wavelength dispersive spectra). Micrometer sized alloy particles display very similar compositions, on average (wt.%): 32.7% Mo, 40.5% Ru, 7% Tc, 4.4% Rh, 12.2% Pd and 3.8% Te. The alloy particles were leached in 10 mM NaCl and 2 mM NaHCO3 solutions, purged with two kinds of gas mixtures. It was observed that, under purging with Ar + 0.03% CO2, the leaching rates of 99Tc and 100Mo were similar, about 1.5 ppb/day. The results obtained under purging with 89.97% Ar + 0.03% CO2 + 10% H2 indicate that the 4d alloy particles are very effective catalysts for reduction caused by hydrogen: the concentrations of 99Tc and 100Mo dropped to levels below 1 ppb. No significant leaching of Ru, Rh, Pd and Te was observed during purging with both gas mixtures.

The reduction of U(VI) by near field hydrogen in the presence of UO2(s)

Summary In several recent works a large impact of dissolved hydrogen on spent fuel dissolution has been observed: the solution concentrations of all redox sensitive components of the spent fuel decrease with time, while gas phase analyses indicate levels of radiolytic oxygen below the detection limit. This indicates that at the relatively low temperatures of these studies (≤70 °C) hydrogen is activated. The aim of this study is to test one of the proposed mechanisms for hydrogen activation, namely by UO2(s) surfaces, which may act as hydrogen catalysts. Dissolved U(VI) carbonate species were used as oxidized species in order to test the reducing ability of hydrogen in the presence and absence of UO2(s). In most cases parallel experiments with the same conditions, but under Ar(g) atmosphere were carried out as blanks. Quartz lined autoclaves avoiding any contact of the solution with metallic parts were used. The experiments were carried out in two stages: first the stability of the (10-7-10-5) M U(VI) in 10 mM NaCl, 2 mM NaHCO3 solutions in the presence of dissolved hydrogen was tested. Then a gold net basket containing UO2(s) fragments was introduced in the autoclave and the concentration of U(VI) was followed at different time intervals by ICP-MS. Various dissolved hydrogen concentrations, temperatures and U(VI) concentrations were used, always spanning the range of these parameters in a proposed deep rock repository. Especially at the lowest U(VI) concentrations tested (∼100 ppb) the activation of dissolved hydrogen by UO2(s) surfaces is clearly visible. As a conclusion of this work it may be stated that in the temperature range (20-70 °C) investigated a) dissolved hydrogen does not reduce U(VI) carbonate species in the absence of a catalyst and b) in the presence of UO2(s) surfaces dissolved hydrogen reduces U(VI) in carbonate solutions, very probably to UO2(s).

Assessment of the relevance of Coffinite formation within the near-field environment of spent nuclear fuel geological disposals

Experiments were performed in anoxic gloves box in an attempt to synthesise Coffinite both in representative near-field conditions, and in conditions which were expected to favour its precipitation according to thermodynamic calculations. The experimental results did not confirm the predictions. However, a new mineral was observed instead of Coffinite. In addition, accurate characterization of various natural samples demonstrate the permanent presence of U(VI) within Coffinite contradictory to its theoretical composition. Our observations raise the question on the validity and applicability of available –actually estimatedthermodynamic data of Coffinite. Based on kinetic hindrance of Coffinite formation, coffinitization of spent nuclear fuel in geological disposal is not anticipated to be a dominant short term process.

Extraction behavior of the synergistic system 2,6-bis-(benzoxazolyl)-4-dodecyloxylpyridine and 2-bromodecanoic acid using Am and Eu as radioactive tracers

Abstract In this study, the extraction properties of a synergistic system consisting of 2,6‐bis‐(benzoxazolyl)‐4‐dodecyloxylpyridine (BODO) and 2‐bromodecanoic acid (HA) in tert‐butyl benzene (TBB) have been investigated as a function of ionic strength by varying the nitrate ion and perchlorate ion concentrations. The influence of the hydrogen ion concentration has also been investigated. Distribution ratios between 0.03–12 and 0.003–0.8 have been found for Am(III) and Eu(III), respectively, but there were no attempts to maximize these values. It has been shown that the distribution ratios decrease with increasing amounts of ClO4 −, NO3 −, and H+. The mechanisms, however, by which the decrease occurs, are different. In the case of increasing perchlorate ion concentration, the decrease in extraction is linear in a log–log plot of the distribution ratio vs. the ionic strength, while in the nitrate case the complexation between nitrate and Am or Eu increases at high nitrate ion concentrations and thereby decreases the distribution ratio in a non‐linear way. The decrease in extraction could be caused by changes in activity coefficients that can be explained with specific ion interaction theory (SIT); shielding of the metal ions, and by nitrate complexation with Am and Eu as competing mechanism at high ionic strengths. The separation factor between Am and Eu reaches a maximum at ∼1 M nitrate ion concentration. Thereafter the values decrease with increasing nitrate ion concentrations.

Surface mediated processes in the interaction of spent fuel α-doped UO2 with H2

In most deep disposal concepts, large amounts of hydrogen are expected to be produced by the anoxic corrosion of massive iron containers. At repository temperatures, hydrogen is quite inert and is not expected to contribute to the redox capacity of the deep groundwaters. In several recent works, a large impact of dissolved hydrogen on the dissolution of the LWR or MOX fuel and UO 2(s) doped with 233U or 238Pu has been observed. For hydrogen concentrations above a certain limit, the dissolution rates of these highly radioactive materials drop to very low values. A discussion of the results obtained with spent fuel or ?-doped UO 2 in the presence of a range of hydrogen concentrations is presented. Typical for all measurements under such conditions are the very low long term concentrations of uranium and other redox-sensitive radionuclides, such as Tc and the minor actinides. The concentrations of U are systematically lower than the values measured during UO2(s) solubility measurements carried out in the presence of strong reducing agents. Measurements of the radiolytic oxygen after long leaching periods result in values below detection limit. The investigation of the surface of spent fuel or UO2(s) pellets doped with 233U by XPS after long periods of testing shows absence of oxidation. The kinetics of the release of non-redox sensitive elements such as Sr and Cs, used to estimate fuel matrix dissolution rates, is also discussed. An attempt is made to propose potential mechanisms responsible for the observed behaviour, based mainly on data from studies on the interaction of water adsorbed on the surfaces of metal oxides or actinide oxides with radiation. Another important effect observed in recent studies is the existence of a threshold for the specific alpha activity below which no measurable influence of the alpha radiolysis on the uranium release from UO2 is observed. The importance of such a threshold for the behaviour of spent fuel under repository conditions encompassing very long time scales will be discussed, as well as the necessity to better investigate the mechanisms of recombination reactions in a thin water layer on the surface of actinide oxides affected by ?- radiolysis.

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.

Redox Reactions of Iron and Uranium Dioxide in Simulated Cement Pore Water Under Anoxic Conditions

Results on the chemical behavior of Fe(0) and UO2(s), as well as the interaction between fresh and corroded iron with U(VI) in simulated cement contacting alkaline solution are reported. Batch experiments were conducted under anoxic conditions at different alkalinities and salt concentrations to investigate: (a) the corrosion of iron foils (b) the U(VI) removal by fresh, aged and pre-treated (with FeS or Fe3O4 layers) iron surfaces in a simulated cement pore fluid, (c) the dissolution rates of newly reduced UO2.00 slices in simulated cement pore fluid and KOH solutions (pH 12.7) and (d) the isotope exchange reactions between dissolved 235U(VI) and 238UO2(s). The reacted iron and UO2(s) surfaces were analyzed by X-ray diffraction (XRD), scanning electron microscopy-energy dispersive spectra (SEM-EDS), laser Raman spectroscopy and X-ray photoelectron, spectroscopy (XPS).

Immobilization of Radionuclides on Iron Canister Material at Simulated Near-field Conditions

This work is a continuation of a long-term spent fuel leaching and radionuclides immobilization (by iron canister) experiment under simulated near-field conditions, in deoxygenated 2 mM NaHCO3 solu ...

A Kinetic Model for the Stability of the Spent Fuel Matrix Under Oxic Conditions: Model Development Against Experimental Evidence

A kinetic model recently developed [1] for the radiolytically induced oxidative dissolution of the spent fuel matrix is presented. This is based on experimental studies on the generation and evolution of radiolytic products in a closed system containing fragments of PWR-fuel [2]. The outcome of this model is currently being integrated in the present PA exercise being prepared by SKB. The calibration of the model against various experimental information and its predictive capabilities for the long term performance of the spent fuel matrix are presented.