An in situ Gas Generation Experiment was initiated at Olkiluoto, Finland, to simulate gas generation from low-level maintenance waste under underground repository conditions. Although the primary focus was on gas generation, the 26-year experiment also provided an opportunity to investigate carbon steel corrosion and concrete degradation upon its termination in 2023. Early heterogeneous conditions enabled the formation of microbial niches and the initiation of gas generation, despite the influence of initially high-pH cementitious water. Approximately 7% of the theoretical gas generation potential was realized over the 26-year duration of the experiment, and at the time of termination, gas production was continuing at a stabilized rate. Microbiological analyses confirmed the presence of microbial groups essential for the complete biodegradation of waste, including bacteria capable of decomposing and fermenting organic matter, as well as acetogens and methanogens. Methane and carbon dioxide were the main gases produced, with CO2 precipitating as CaCO3. Simulated disturbances, including sulphate addition and an increase in pH, had no long-term impact on gas generation. At termination, the exterior paint on the steel drums remained in surprisingly good condition, although it had become brittle and exhibited blistering. The drum interiors showed both uniform and pitting corrosion. The concrete retained its strength within the design limits, with carbonation attributed to pre-experiment storage conditions.

The chlorination of oxides of major concern in cassiterite concentrate with various chlorinating agents is investigated in light of their thermodynamic feasibilities to extract and recover their valuable metal components. Mechanisms responsible for the processes and their Gibbs free energy changes as a function of temperature to selectively separate and/or recover the metal(s) of interest and unwanted ones as their metallic chlorides are identified. Attention is given to gaseous (Cl2 and Cl2 + CO mixture) and solid (CaCl2 and MgCl2) chlorine sources, from which Cl2 + CO shows no reaction selectivity for any of the oxides but a feasible metal chloride formation for all. Chlorine gas (Cl2), on the other hand, could selectively form chlorides with metals of +2 oxidation state in their oxides, leaving those of high oxidation state unreacted. MgCl2, unlike CaCl2, is found capable of producing calcium, ferrous, and stannic chloride from their metallic oxides with enhanced reaction tendencies in the presence of silicon dioxide (SiO2). An overall study of the thermodynamic feasibility of all chlorine sources looked at alongside operational and environmental viabilities suitably suggests MgCl2 for a selective extraction of the valuable metal components in a cassiterite concentrate, in which case, moderate temperatures seem promising.

Enrichment of impurities occurs in steam generators of nuclear plants due to intense boiling conditions pre-vailing within them. In this work, deposition and release of chloride and sulfate impurities were studied using a dedicated water circulation loop with Alloy 690 as steam generator tubing. Electrochemical impedance spec-troscopy was employed to follow in-situ the corrosion behavior in non-boiling, boiling and after-boiling con-ditions. Deposition/release of impurities is quasi-first-order vs. concentration. Interpretation of electrochemical impedance data using an updated version of the Mixed-Conduction Model allows to obtain reaction orders of oxide growth and corrosion release with respect to chloride and sulfate additives.

This study deals with an investigation of how substitution of an alcohol group by a thiol group in mixtures of choline chloride with a series of bio-sourceable molecules affects the physico-chemical properties of the mixtures and their ability to dissolve metal oxides. All of the thiol mixtures studied showed a higher affinity and selectivity for late transition metals and the physical properties of the mixtures were improved compared to their alcohol analogues (i.e. lower viscosity, wider liquid range). The metal solubility was assessed via determinination of the final concentration of the metal oxides dissolved in thiol mixtures via inductively coupled plasma optical emission spectroscopy (ICP-OES). The thiol function selectively improved the solubilities of the late transition metal oxides (i.e. copper and zinc), which are valuable metals often present as residue in industrial waste. The solubility of iron oxides was much lower than that of the valuable metals, which is a significant benefit in industrial applications. The different solubilization behaviour of metal oxides in the thiol and alcohol mixtures was further investigated via UV-vis absorption and infrared spectroscopy. This study allowed to assess the potential of these low-temperature-transition mixtures for the selective recovery of metals.

This study deals with an investigation of how substitution of an alcohol group by a thiol group in mixtures of choline chloride with a series of bio-sourceable molecules affects the physico-chemical properties of the mixtures and their ability to dissolve metal oxides. All of the thiol mixtures studied showed a higher affinity and selectivity for late transition metals and the physical properties of the mixtures were improved compared to their alcohol analogues (i.e.lower viscosity, wider liquid range). The metal solubility was assessedviadetermination of the final concentration of the metal oxides dissolved in thiol mixturesviainductively coupled plasma optical emission spectroscopy (ICP-OES). The thiol function selectively improved the solubilities of the late transition metal oxides (i.e.copper and zinc), which are valuable metals often present as residue in industrial waste. The solubility of iron oxides was much lower than that of the valuable metals, which is a significant benefit in industrial applications. The different solubilization behaviour of metal oxides in the thiol and alcohol mixtures was further investigatedviaUV-vis absorption and infrared spectroscopy. This study allowed the potential of these deep-eutectic solvents for the selective recovery of metals to be assessed.

The objective of this article is to study the activities and special characteristics of graphite and Fluental neutron moderator material in FiR1 TRIGA Mark II type research reactor in Finland. Quantifying the nuclide vector and total activities of decommissioning waste is a legal requirement and provides input to many other aspect of decommissioning project. Because earlier calculations of decommissioning waste nuclide inventories used partially assumed input data e.g. material compositions, measurement data is needed to validate the calculated results. Graphite and Fluental are especially important in long term final disposal of the waste, because graphite contains volatile radionuclides (C-14, H-3, Cl-36) that spread rapidly in final disposal conditions and the corrosion of aluminium in Fluental will result in hydrogen pressure generation, which can accelerate the diffusion of radionuclides in the bedrock. Moreover, these types of materials are not used in Finnish power reactors and their behaviour in final disposal conditions have not been studied so thoroughly earlier. This study reports composition and activity measurements from graphite, activation and leakage of tritium in Fluental and reviews main phenomena that could occur in final disposal environment. Measured activities were compared to estimates calculated with a point-depletion code modelling the reactor irradiation history. Results are used to validate the calculated estimates of total activities and the scaling matrix method that will be used to classify the packed decommissioning waste. Due to limited number of samples, numerical results still contain variations. However, developed methods are still valuable in future analyses and measured data also rules out unexpected activities.

Phosphorus and uranium are both vital elements for society. In recent decades, fears have arisen about the future availability of low-cost phosphorus and uranium. This has resulted in pressure to de-centralize production of both elements by utilizing lower-grade or complex deposits. The research presented here focused on phosphorus-containing apatite ores with uranium impurities; in order to separate uranium by selective and sequential bioleaching before phosphorus leaching. This would create an alternative process route for solvent-extraction, used to remove/recover uranium from the phosphorus acid product of apatite H2SO4 wet process. In this work, it was seen that the used fluorapatite ore required 24 h leaching at pH 1 by H2SO4 to result in 100% leaching yield for phosphorus. As this ore did not contain much uranium, an artificial fluorapatite-uranium ore was prepared by mixing standard uranium ore and fluorapatite. The research with this ore showed that 89% of uranium dissolved in 3 days at pH > 2 and leaching was improved by applying Fe3+ oxidant. In these conditions only 4% of phosphorus was leached. By prolonged (28 days) leaching 95% uranium yield was reached. According to the experiments, the iron in the uranium leach solution would be mainly Fe3+, which allows the use of H2O2 for uranium recovery and then direct use of spent leachate for another uranium leaching cycle. After the dissolution of uranium, 90% of phosphorus was dissolved by decreasing the pH to 1.3. This was done by bioleaching, by utilizing biogenic sulfur oxidation to sulfuric acid.

The objective of the study has been to verify the calculated residual activity in the decommissioning waste of the TRIGA Mark II-type research reactor FiR 1 in Finland. Knowledge of the radioactive inventory of irradiated materials is important in the planning of decommissioning activities and is essential for predicting the radiological impact to personnel and the environment. Measurements are performed for low active material samples from outer parts of the reactor. Methods include gamma spectrometric measurements, composition measurements with mass spectrometry, oxidation measurements of especially C-14 in graphite, and full combustion measurements of lithium-enriched shielding materials. Results are compared to estimates calculated with a combined Monte Carlo model of the reactor and a point-depletion code modeling the irradiation history. Decommissioning waste consists mainly of ordinary concrete, aluminum, steel, and graphite parts. Only preliminary measurements of low active samples are reported so far, but the same methods will be used later for characterizing and classifying dismantling waste. Some discussion of characterization requirements and future sampling is also included.

The next generation application of nuclear energy has shown an interest towards ThO2 or mixed oxides fuels. However, more knowledge is needed, for example, on the microstructural and matrix dissolution behaviour of ThO2. The objective of this study was to examine surface properties of thick ThO2 pellets by non-destructive alpha analysis method accompanied with Monte Carlo simulations before and after a leaching experiment. This study has been conducted to investigate the contents, and surface layer thickness of crystalline (ThO2)-Th-232 pellet leached in aqueous solution with excess of Th-229 tracer in the beginning of the experiment. The simulation results have suggested a maximum 0.1 mu m concentrated layer of Th-229 on the pellet surface. The presence of this layer suppresses further dissolution of the pellet. (C) 2017 Elsevier B.V. All rights reserved.

The sparingly soluble ThO2 dissolution behaviour was studied under conditions that are expected to prevail in bedrock after the closure of the spent nuclear fuel disposal facility. The objective was to investigate the characteristics of initial dissolution of crystalline 232ThO2 by adding 229Th tracer to the aqueous phase in the beginning of the 534 days experiment.The evolution of 232Th concentration and 229Th/232Th ratio were followed by sector field ICP-MS (inductively coupled plasma mass spectrometer). Selected samples were measured also by alpha spectrometry to compare the results from the used analysis techniques and they gave comparable results. In the early stage of dissolution there was no significant control by chemical equilibrium and the dissolution process seemed to be controlled by the stability of surfaces.

The objective of this study was to investigate the dissolution rate of ThO2 which was synthesised to approximate, as closely as possible, the microstructure of UO2 in a nuclear fuel matrix. The optimal sintering temperature for ThO2 pellets was found to be 1750 degrees C, which produced pellets with a microstructure similar to UO2 nuclear fuel pellets, with randomly oriented grains ranging in size from 10 to 30 mu m. Dissolution was conducted using ThO2 particles of different size fractions (80 to 160 mu m and 2 to 4 mm) in the presence and absence of carbonate, in solutions with pH from 2 to 8 and at 80 degrees C. Dissolution rates were calculated from Th released from the solid phase to solution. Particles of ThO2 were also leached with 1 M HNO3 at 80 degrees C in order to investigate the morphological changes at the particle surfaces. The concentration of Th was found to be >= 10(-9) mol/L at pH <= 4, lower than the theoretical solubility of crystalline ThO2. At higher pH values, from 4 to 8, the measured concentrations (10(-10) to 10(-12) mol/L) were between the theoretical solubility of ThO2 and Th(OH)(4). Grain boundaries were shown to exert an influence on the dissolution of ThO2 particles. Using high resolution aqueous solution analysis, these data presented here extend the current understanding of Th solubility in solution.

The objective of this work is to determine whether the presence of trace elements in natural groundwaters affects the dissolution rate of uranium dioxide in the presence of alpha radiation that causes radiolysis of water. The study is a part of the project Reducing Uncertainty in Performance Prediction (REDUPP) under the Seventh Framework Programme of the European Atomic Energy Community (EURATOM). The project aims to reduce uncertainties related to the extrapolation of the results of laboratory experiments to the conditions expected under geologic disposal. Thus far, synthetic groundwater has been normally used in the experiments. The synthetic groundwaters used do not contain all of the chemical elements that occur in natural groundwaters. Three natural groundwaters were chosen for the dissolution experiments with 0%, 5%, and 10% 233U-doped UO2 samples. These include a brackish groundwater, a saline groundwater and a low ionic strength groundwater. At the time of writing this paper, the dissolution experiments have been finished in the first groundwater, which was a moderately saline, brackish groundwater. The groundwater samples for the experiments were taken from a borehole in the Olkiluoto site in Finland. The measurements for dissolution rates were conducted under reducing conditions established using metallic iron in solution and an argon atmosphere in the glove box. The isotope dilution method was used to decrease uncertainties due to precipitation and sorption effects. The resulting dissolution rates in OL-KR6 natural groundwater were generally somewhat higher than the rates measured previously in synthetic groundwaters under similar redox conditions. No clear effect of alpha radiolysis could be seen for tests with lower SA/V, while those for higher SA/V indicated that the dissolution rate was higher for the 10% 233U-doped UO2, suggesting the effect of alpha radiolysis under these conditions.