Actinide Content Research Articles

Thin extractive membrane for monitoring actinides in aqueous streams

Alpha spectrometry and solid state nuclear track detectors (SSNTDs) are used for monitoring ultra-trace amount of alpha emitting actinides in different aqueous streams. However, these techniques have limitations i.e. alpha spectrometry requires a preconcentration step and SSNTDs are not chemically selective. Therefore, a thin polymer inclusion membrane (PIM) supported on silanized glass was developed for preconcentraion and determination of ultra-trace concentration of actinides by α-spectrometry and SSNTDs. PIMs were formed by spin coating on hydrophobic glass slide or solvent casting to form thin and self-supported membranes, respectively. Sorption experiments indicated that uptakes of actinides in the PIM were highly dependent on acidity of solution i.e. Am(III) sorbed up to 0.1molL−1 HNO3, U(VI) up to 0.5molL−1 HNO3 and Pu(IV) from HNO3 concentration as high as 4molL−1 . A scheme was developed for selective sorption of target actinide in the PIM by adjusting acidity and oxidation state of actinide. The actinides sorbed in PIMs were quantified by alpha spectrometry and SSNTDs. For SSNTDs, neutron induced fission-fragment tracks and α-particle tracks were registered in Garware polyester and CR-39 for quantifications of natural uranium and α-emitting actinides (241Am/239Pu/233U), respectively. Finally, the membranes were tested to quantify Pu in 4molL−1 HNO3 solutions and synthetic urine samples.

Versatile Monazite: Resolving geological records and solving challenges in materials science: Monazite as a promising long-term radioactive waste matrix: Benefits of high-structural flexibility and chemical durability

Monazite (Ln 3+ PO 4 ) and related solid solutions are a well-known source of rare earth elements on earth. They may also accommodate large amounts of thorium and uranium without sustaining damage to the structure by self-irradiation. Such observations led to monazite-type structures being proposed as a potential host matrix for sequestering long-lived radionuclides produced during the nuclear fuel cycle and/or plutonium and americium from dismantled nuclear weapons. Monazite has two main advantages as a matrix for the containment of radioactive waste (or “radwaste”). The first is a highly flexible structure that permits accommodation of high concentrations of actinides. The incorporation of trivalent elements may be achieved by direct synthesis of An 3+ PO 4 (An 3+ = plutonium, Pu to einsteinium, Es), while tetravalent cation incorporation requires coupled substitutions, either on the anionic site (leading to monazite-huttonite solid solution) or on the cationic site (monazite-cheralite solid solution). Various methods developed for the preparation of such compounds are summarized here, as well as the experimental conditions required for the production of sintered pellets, with a particular focus on plutonium-bearing compositions. The second highly favorable property of monazite is its high chemical durability. Several experimental procedures developed to determine normalized leaching rates are reviewed, as well as results obtained from natural and synthetic monazite. Potential phases formed during dissolution were considered because they also partially control the concentration of actinides in the media. A preliminary list for such phases of interest, as well as corresponding thermodynamic data, is presented.

Determination of the 237Np(n,f) reaction cross section for En = 4.5-5.3 MeV, using a MicroMegas detector assembly

In the present work, the measurement of the 237Np(n,f) cross section with reference to the standard 238U(n,f) reaction was performed for the first time with a MicroMegas detector, especially developed at CERN, for these measurements, within the framework of the n_TOF Collaboration. The incident monoenergetic neutron beam with energies in the range 4.5-5.3MeV was produced via the 2H(d,n) reaction at the neutron facility of the 5.5MV Tandem accelerator laboratory at NCSR “Demokritos”. The mass of the actinide content of the targets used and of their impurities was quantitatively determined via alpha spectroscopy. Furthermore, their thickness and homogeneity have been examined via the RBS (Rutherford Backscattering Spectrometry) technique. Monte Carlo simulations were carried out using the code MCNP5 implementing the neutron beam setup and the MicroMegas assembly in order to determine the neutron flux for each target, as well as the existence of possible low-energy neutrons due to scattering. Additional simulations with FLUKA were performed, studying the energy deposition of the fission fragments in the active area of the detector, in order to accurately estimate the detection efficiency. Fair discrimination of the heavy and light mass peaks of the fission fragments at the energy spectra was achieved. The present work is compared to existing evaluations and previous data which present discrepancies of up to 8% within the same energy range.

Irradiation of a thorium–plutonium rodlet: Experiment and benchmark calculations

A benchmark exercise for thorium–plutonium fuel, based on experimental data, has been carried out. A thorium–plutonium oxide fuel rodlet was irradiated in a PWR for four consecutive cycles, to a burnup of about 37 MWd/kgHM. During the irradiation, the rodlet was inserted into a guide tube of a standard MOX fuel assembly. After the irradiation, the rod was subjected to several PIE measurements, including radiochemical analysis. Element concentrations and radial distributions in the rodlet, multiplication factors and distributions within the carrier assembly of burnup and power were calculated. Four participants in the study simulated the irradiation of the MOX fuel assemblies including the thorium–plutonium rodlet using their respective code systems; MCBurn, HELIOS, CASMO-5 and ECCO/ERANOS combined with TRAIN. The results of the simulations and the measured results of the radiochemical analysis were compared and found to be in fairly good agreement when the calculated results were calibrated to give the same burnup of the thorium–plutonium rodlet as that experimentally measured. Average concentrations of several minor actinides and fission products were well reproduced by all codes, to the extent that can be expected based on known uncertainties in the experimental setup and the cross section libraries. Calculated results which could not be confirmed by experimental measurement were compared and only two significant anomalies were found, which can probably be addressed by limited modifications of the codes.

Isotopic Fingerprinting of the World’s First Nuclear Device Using Post-Detonation Materials

In the event of a rogue nuclear attack or interception of illicit nuclear materials, timely forensic investigations are critical for accurate source attribution. Uranium (U) and plutonium (Pu) isotopic ratios of intercepted materials or postdetonation samples are, perhaps, the most valuable evidence in modern nuclear forensics. These ratios simultaneously provide information regarding the material's ''age'' (i.e., time elapsed since last purification), actinide concentrations, and relevant isotopic ratios/enrichment values. Consequently, these isotope signatures are invaluable in determining the origin, processing history, and intended purpose of any nuclear material. Here we show, for the first time, that it is feasible to determine the U and Pu isotopic compositions of historic nuclear devices from their postdetonation materials utilizing in situ U isotopic measurements. The U isotopic compositions of trinitite glass, produced subsequent to the world's first atomic explosion, indicate two sources: the device's tamper, composed of natural U that underwent fission during detonation, and natural U from the geological background. Enrichments in (234,235,236)U reflect the in situ decay of (238,239,240)Pu, the fuel used in the device. Time-integrated U isotopic modeling yields "supergrade" compositions, where (240)Pu/(239)Pu ≈ 0.01-0.03 and (238)Pu/(239)Pu ≈ 0.00011-0.00017, which are consistent with the Pu originating from the Hanford reactor. Spatially resolved U isotopic data of postdetonation debris reveal important details of the device in a relatively short time frame (hours). This capacity serves as an important deterrent to future nuclear threats and/or terrorist activities and is critical for source attribution and international security.

Transmutation of Nuclear Waste and the future MYRRHA Demonstrator

While a considerable and world-wide growth of the nuclear share in the global energy mix is desirable for many reasons, there are also, in particular in the "old world" major objections. These are both concerns about safety, in particular in the wake of the Fukushima nuclear accident and concerns about the long-term burden that is constituted by the radiotoxic waste from the spent fuel. With regard to the second topic, the present contribution will outline the concept of Partitioning & Transmutation (P&T), as scientific and technological answer. Deployment of P&T may use dedicated "Transmuter" or "Burner" reactors, using a fast neutron spectrum. For the transmutation of waste with a large content (up to 50%) of (very long-lived) Minor Actinides, a sub-critical reactor, using an external neutron source is a most attractive solution. It is constituted by coupling a proton accelerator, a spallation target and a subcritical core. This promising new technology is named ADS, for accelerator-driven system. The present paper aims at a short introduction into the field that has been characterized by a high collaborative activity during the last decade in Europe, in order to focus, in its later part, on the MYRRHA project as the European ADS technology demonstrator.

Characterization of actinide targets for fission cross section measurements

In this work the characterization of thin actinide targets ( 237Np, 238U, 235U), man- ufactured for measurements of fission cross section, will be presented. The determi- nation of the mass of the actinide content of each target has been performed via alpha spectrometry, utilizing two SSB detectors, with 50mm2 and 3000mm2 active surface, in order to combine good resolution and good statistics respectively and obtain a smaller error in the final mass result. The impurities of the targets were quantitatively estimated with the same technique. The homogeneity of the samples was examined with two independent methods. Firstly, RBS measurements were per- formed using a proton beam of 2 MeV, provided by the 5.5 MV Tandem Van de Graaff accelerator of the N.C.S.R. ”Demokritos”. The thickness of each target was quantitatively estimated at various points. Secondly, CR-39 plastic track detectors were used to measure the alpha tracks from the surface of the targets and provide an image of the activity of the surface.

Accommodation of multivalent cations in fluorite-type solid solutions: Case of Am-bearing UO2

The radiotoxicity of spent nuclear fuel is mainly due to the content of minor actinides, which could be substantially reduced by Partitioning and Transmutation. A possible transmutation method would be to employ americium-bearing uranium oxide materials as blanket fuels in fast neutron reactors. In order to maintain fuel performance and reactor safety, it is mandatory to control the structural homogeneity and oxygen stoichiometry during the sintering process. In this work, U0.85Am0.15O2−x materials, fabricated by a solid state chemistry process, were sintered at 2023K under three oxygen potentials, i.e. −375, −350 and −325kJmol−1, thereby significantly extending the range of a previous study. By coupling X-ray diffraction and X-ray absorption spectroscopy measurements, it was shown that fluorite solid solutions are obtained whatever the sintering conditions. The presence of U(+V), pointed out in a previous work for oxygen potentials equal to −520 and −450kJmol−1, was confirmed. This result constitutes an experimental proof of the existence of U(+V) in An-doped UO2 fluorite-type structure materials. These experimental data are major results in view of the developing thermodynamical model of the U–Am–O system. Considering the now available extended range, the effect of the oxygen potential is discussed in terms of charge distribution and local structure.

Radiometric detector options to aid in DOE high activity waste tank in situ characterization efforts

Radioactive waste cleanup and subsequent closure of waste storage tanks is currently underway at the Savannah river site, prompting the need to characterize the residual contents (heels) of the tanks. Occasionally, results from laboratory analyses indicate alternative sub-sampling strategies are needed, resulting in repetitive efforts to sample and analyze tank bottoms. The development of a system for in situ tank analyses using a radiometric probe, which could be lowered into a waste tank, could aid in identifying waste structures on tank bottoms requiring further sampling and characterization. Ideally, the probe would provide information for determining which structures were higher in concentrations of actinides and fission products characteristic of DOE high level waste (HLW) heels. Although only a limited set of isotopes can be measured directly without extensive radiochemical separations, the low-energy photon spectra of HLW do offer some intriguing possibilities for characterization using a radiometric probe. One possibility for obtaining a low-energy photon spectrum in the presence of high levels of interfering radiation would be to design a probe primarily based upon recently developed technology from Amptek Inc. Such a detector would be relatively insensitive to the high photon background, which would paralyze conventional gamma probes (i.e. sodium iodide) subjected to the same radiological conditions. The prototype detector is capable of successfully obtaining high resolution measurements at very high count rates (in excess of 500,000 counts per second). An overview of measurements obtained from various HLW samples using the prototype Amptek detector, as well as some additional detector technologies, which could enhance this prototype, will be discussed.

Plutonium isotopes and 241Am in the atmosphere of Lithuania: A comparison of different source terms

137Cs, 241Am and Pu isotopes collected in aerosol samples during 1994–2011 were analyzed with special emphasis on better understanding of Pu and Am behavior in the atmosphere. The results from long-term measurements of 240Pu/239Pu atom ratios showed a bimodal frequency distribution with median values of 0.195 and 0.253, indicating two main sources contributing to the Pu activities at the Vilnius sampling station. The low Pu atom ratio of 0.141 could be attributed to the weapon-grade plutonium derived from the nuclear weapon test sites. The frequency of air masses arriving from the North-West and North-East correlated with the Pu atom ratio indicating the input from the sources located in these regions (the Novaya Zemlya test site, Siberian nuclear plants), while no correlation with the Chernobyl region was observed. Measurements carried out during the Fukushima accident showed a negligible impact of this source with Pu activities by four orders of magnitude lower as compared to the Chernobyl accident. The activity concentration of actinides measured in the integrated sample collected in March–April, 2011 showed a small contribution of Pu with unusual activity and atom ratios indicating the presence of the spent fuel of different origin than that of the Chernobyl accident.

Determination of the radionuclide composition and burn-up of high-burn-up WWER-1000 fuel by destructive methods

A sample of WWER-1000 spent nuclear fuel (SNF) with the burn-up of approximately 50 GW day (t U)−1 was studied by methods of destructive analysis. Data on the content of isotopes of U, Pu, Am, Cm, Nd, and Cs are given. The methods used are briefly described. These include anion exchange and extraction chromatography, thermal ionization mass spectrometry, luminescence analysis, and γ- and α-ray spectrometry. The main method of the study is isotope dilution followed by mass-spectrometric analysis. The data obtained will be used for developing of a low-waste and low-cost technology for preprocessing of highburn-up SNF and also for checking the accuracy of calculation of heavy actinide content in SNF and for improving computation methods and programs intended for calculation of the SNF nuclide composition in relation to the fuel burn-up in WWER-type reactors. The Nd and Cs isotopes were used as burn-up monitors.

Optimization of parameters of alpha spectrometry with silicon detector for low level measurements of actinides in environmental samples

Determination of actinides in environmental and biological samples is an important activity of radiation protection program at nuclear energy facilities. High resolution alpha spectrometry with passivated ion implanted Silicon detectors is widely used for the determination of actinides concentration. Low levels of activity concentrations in these samples often require long counting duration of a few days to obtain accurate and statistically significant data for further impact assessment. In alpha spectrometry, the chamber in which Si detector operated is a critical component and maintained at a desired vacuum for minimizing the alpha particle attenuation. Experimental evaluation of variations in energy resolution and tailing of alpha spectra was investigated under different chamber air pressures from about 6.7Pa to more than 2700Pa under the chamber hold mode and pump electrically switched off conditions. As part of validation, data collected on an IAEA inter-comparison exercise sample are presented under short and long counting durations with pump operating and switched off conditions respectively. It has been observed that the FWHM values do not significantly degrade, to impact the low and medium level concentration alpha spectra, for variations in vacuum chamber pressures from about 6.7Pa to 2700Pa.

Comparison of MCB and FISPACT burn-up performances using the HELIOS experiment technical specifications

Abstract The objective of the HELIOS experiment was to test innovative, uranium-free nuclear fuel containing americium and to study helium release and fuel swelling behaviour during irradiation. The experiment consisted of irradiation of five fuel samples, each containing various initial concentrations of Pu and Am in different inert matrices. The irradiation was performed in the High Flux Reactor (HFR) at Petten during nine cycles (i.e. about nine months) for a total duration of 240 equivalent full-power days. The neutronic calculations play an important role in preparation and analysis of the irradiation experiments. Therefore, this paper presents an explicit burn-up analysis of the HELIOS samples using two numerical tools, the MCB and FISPACT codes. The focus of this calculation is on the concentrations of Pu, minor actinides (MAs) and He during irradiation time and on total sample power due to nuclear reactions and decay. Particular attention was paid to Am242m, the fission cross-section of which is significant in the thermal neutron spectrum. The MCB code incorporates an improved method of producing this, which was subsequently tested and qualified. The sensitivity of the nuclide evolution to the choice of JEF2.2 and JEFF3.1 transport cross-section libraries was also investigated.

Measurement of the Neutron Induced Fission of 245Cm in the Resolved Resonance Region and Its Resonance Analysis

In order to improve the prediction of heavy actinide concentration in reactor fuel elements and to study the reduction of the long-term nuclear waste radiotoxicity by using transmutation, cross sections for neutron induced reactions are highly required by nuclear industry for many minor actinides. In this context, a new measurement of the Cm-245(n,f) cross section has been performed at the GELINA neutron facility of the Institute for Reference Materials and Measurements (IRMM) in Gee], Belgium. This measurement was motivated by the fact that in the thermal region a strong dispersion between measurements is observed and in the resolved resonance region, only a few old measurements exist with in some cases a poor energy resolution. The energy of the neutrons is determined applying the time of flight method using a flight path length of about 9 m. In the present work, the incident neutron energy covers the resolved and unresolved resonance regions up to a few keV. A highly enriched Cm-245 sample (98.48%) was mounted back-to-back with a B-10 sample in the centre of a vacuum chamber together with two surface barrier detectors positioned outside the neutron beam. One detector measured the B-10(n,alpha)Li-7 reaction products for the neutron flux determination, while the second one registered the Cm-245(n,f) fragments. In this way, the neutron flux can be determined simultaneously with the fission fragments. A control measurement has been performed replacing the Cm-245 sample with a U-235 sample in order to check that the well-known U-235(n,f) cross section can be reproduced. In comparison with previous Cm-245 fission cross section measurements, our results show a nice improvement of the energy resolution, in particular below 20 eV. A new set of resonance parameters is determined using the evaluation code CONRAD which was recently developed by CEA-Cadarache.

The reactor physics characteristics of a transuranic mixed oxide fuel in a heavy water moderated reactor

The reprocessing actinide materials extracted from spent fuel for use in mixed oxide fuels is a key component in maximizing the spent fuel repository utility. While fast spectrum reactor technologies are being considered in order to close the fuel cycle, and transmute these actinides, there is potential to utilize existing pressurized heavy water reactors such as the CANDU ® 1 1 CANDU is a trademark of Atomic Energy of Canada Limited. design to meet these goals. The use of current thermal reactors as an intermediary step which can burn actinide based fuels can significantly reduce the fast reactor infrastructure needed. This paper examines the features of actinide mixed oxide fuel, TRUMOX, in a typical CANDU nuclear reactor. The actinide concentrations used were based on extraction from 30 year cooled spent fuel and mixed with natural uranium in 4.75% actinide MOX fuel. The WIMS-AECL model of the fuel lattice was created and the two neutron group properties were transferred to RFSP in order to create a 3 dimensional time average full core model. The model was created with typical CANDU limits on bundle and channel powers and a burnup target of 45 MWd/kgHE. The TRUMOX fuel design achieved its goals and performed well under normal operations simulations. This effort demonstrated the feasibility of using the current fleet of CANDU reactors as an intermediary step in burning reprocessed spent fuel and reducing actinide burdens within the end repository. The recycling, reprocessing and reuse of spent fuels produces a much more sustainable and efficient nuclear fuel cycle using existing and proven reactor technologies.

Thermodynamic Approach for Predicting Actinide and Rare Earth Concentrations in Leachates from Radioactive Waste Glasses

Studies aimed primarily at determining leach rates of different elements from doped glasses have resulted in computerized models for predicting leachate concentrations. However, leach rate related data should be limited to predicting the stability behavior of the glass matrix; the radionuclide release data based on these studies are empirical and are highly dependant on many variables and processes which have not been systematically evaluated and thus do not provide a reliable method of predicting leachate concentrations. A better approach is available for those elements that can readily form relatively insoluble solids during preparation of glass or glass/water interactions. This alternate approach relies on the experimental solubilities of pulverized doped glasses, in a wide range of well-controlled important variables such as pH and pe, and their comparisons at the given aqueous composition to predicted solubilities of known solid phases from the thermodynamic data. These comparisons are used to indirectly identify specific solubility-controlling solids in doped glass/water systems to determine scientifically defensible aqueous concentrations of different elements for any given groundwater composition, independent of glass dissolution kinetics and independent of time. This paper summarizes data available for the application of this alternate approach to reliably predict concentrations of thorium, uranium, neptunium, plutonium, and trivalent actinides and rare earth elements leachable from the doped glasses. The thermodynamic data, in addition to that reported in recent critical reviews, includes new data that were developed for the solubility products of Th3(PO4)4(s) and the solid solutions of trivalent actinides and rare earth hydroxides. Thermodynamic interpretations of the doped glass solubility data show specifically that tetravalent actinide hydrous/crystalline oxides and solid solutions of trivalent actinides and rare earths hydroxides in non-phosphate glasses and Th3(PO4)4(s) and MPO4(s), where M denotes trivalent actinides or rare earths, in phosphate-containing glasses are the dominant solubility-controlling solids. Needed future research in this area is briefly outlined.

Murataite matrices for actinide wastes

Matrices for actinide wastes, consisting of complex murataite-type oxides, were studied. The samples were prepared by sintering of the oxide charge at 1100–1400°C or by fusion at 1450–1600°C followed by crystallization. Along with the structural analog of the natural mineral (murataite 3C), appreciable role in the samples is played by the phases (hereinafter murataites 5C, 7C, and 8C) consisting of pyrochlore and murataite 3C units. The fraction of wastes in the samples is about 10 wt %, which is close to the values for the pyrochlore matrix for Pu immobilization. In the ceramics prepared from the melt, the murataite grains have a zonal structure. They are built of murataite 3C at edges and murataite 5C (rarely 7C) in the center where the actinide content is maximal. This fact accounts for their high capability to isolate radionuclides. Amorphization of the structure only slightly affects the stability of murataite in solution. The optimal procedure for the industrial production of the matrices is their melting by induction heating in cold crucible (IMCC) and crystallization.

Radionuclide Geomicrobiology of the Deep Biosphere

This review summarizes research into interactions between microorganisms and radionuclides under conditions typical of a repository for high-level radioactive waste in deep hard rock environments at a depth of approximately 500 m. The cell–radionuclide interactions of strains of two bacterial species (i.e., Shewanella putrefaciens and Desulfovibrio aespoeensis) with Cm, Pm, and Pu were investigated in vitro and the results were found to agree with literature data. Siderophores are capable of binding actinides strongly and need to be considered in terms of radionuclide mobility in the subsurface. Siderophores and other bioligands were found to have a generally very strong mobilizing effect on Am, Cm, Fe, Np, Pm, Pu, Th, and U. Where reduced groundwater enters an aerobic environment, such as a large open fracture or fracture zone (e.g., in tunnels), there is the possibility of rapid aerobic bacterial metabolism, microbial proliferation, biofilm development, and iron oxide formation. In these environments, the stalk-forming bacterium Gallionella may act as a scaffold for iron oxide precipitation on biological material. In situ work in the Äspö Hard Rock Laboratory tunnel indicated that the concentrations of biological iron oxides, lanthanides, and actinides correlated positively with Gallionella biomass, a finding that compares well to literature data. In deep oligotrophic subsurface granitic rock environments, fracture biofilms reach a threshold of approximately 2–5 × 106 cells cm−2. The cells in these biofilms are spatially distinct and are surrounded by an extracellular polysaccharide matrix that constitutes up to 60% of the total organic carbon. Calcium-rich amorphous masses are associated with this base layer of cells and organic exudates. In situ, these biofilms have been found to influence the adsorption and immobilization of Am, Np, Pm, Th, and U. This review demonstrates that microorganisms can influence, and sometimes even control, the migration behavior of radionuclides in deep geological environments typical of future sites for radioactive waste repositories.

Hypothetical integration of a fast reactor to the fuel cycle in Slovakia

The inspiration for dealing with the topic of fuel cycle back-end was attendance at a European project called RED-IMPACT – Impact of Partitioning Transmutation and Waste Reduction Technologies. This paper includes an image how to re-use energetic potential of stored spent fuel and at the same time how to effectively reduce spent fuel and radioactive waste volumes aimed for deep repositories. The first part is based on the analysis of Pu and minor actinides (MA) content in actual VVER-440 spent fuel stored in Slovakia. The next parts present the hypothetical possibilities of reprocessing and Pu re-use in a fast reactor under Slovak conditions. For the hypothetical transmutation of heavy nuclides (Pu and MA) contained in Slovak spent fuel a SUPERPHENIX (SPX) fast reactor with increased power was chosen because a fast nuclear reactor cooled by sodium belongs to the group of Generation IV reactor systems. This article deals with the analysis of power production and fuel cycle indicators. The indicators of the SPX calculation model were compared with the results of the VVER-440 spent fuel with the initial fuel enrichment of 4.25% U-235 + 3.35% Gd 2O 3. The created SPX model in the spectral computer code HELIOS 1.10 consists of a fissile (fuel) and a fertile part (blanket). All kinds of calculations were performed by the computer code HELIOS 1.10. This study also exposes the HELIOS modelling and simulating borders.

Thorium partitioning in Greek industrial bauxite investigated by synchrotron radiation and laser-ablation techniques

Typical red–brown (Fe-rich) and high-quality white–grey (Fe-depleted) bauxite samples from active mines of the Parnassos-Ghiona area, central Greece, were investigated. According to XRF and ICP-MS analyses their actinide content, and particularly of Th, is relatively increased. Fe-depleted samples contain up to 62.75 ppm Th corresponding to 220 Bq/kg due to 228Ac ( 232Th-series), whereas Fe-rich samples are less Th-radioactive (up to 58.25 ppm Th, 180 Bq/kg due to 228Ac). Powder-XRD patterns showed that Th-enriched (Fe-depleted) bauxite consists mostly of diaspore (AlOOH polymorph), anatase and rutile (TiO 2 polymorphs). SEM-EDS indicated the presence of Ti–Fe–containing phases (e.g. ilmenite, FeTiO 3), chromite (Cr-spinel) and besides LREE-minerals (mostly bastnäsite/parisite-group) and zircon (ZrSiO 4) hosting a part of the bulk Th. The presence of Th in diaspore and in Ti-containing phases (not detected by SEM-EDS as in the case of REE-minerals and zircon) was investigated, into distinct pisoliths of Fe-depleted bauxite, using μ-XRF and μ-XAFS in the SUL-X beamline of the ANKA Synchrotron facility (KIT, Germany). XAFS spectra of Th salts and Th-containing reference materials were obtained as well. Accordingly it was revealed, for the first time in the literature, that Ti-phases, and particularly anatase, host significant amounts of Th. This novel conclusion was complementary supported by LA-ICP-MS analyses indicated an average of 73 ppm Th in anatase grains together with abundant Nb (3356 ppm), Ta (247 ppm) and U (33 ppm). The Th L III-edge XAFS spectra as compared to reference materials, give also evidence that Th 4+ may not replace Ti 4+ in distorted [TiO 6] fundamental octahedral units of anatase and ilmenite lattice (CN = 6). The occupation of either extraframework sites of higher coordination (CN = 6.9 or even CN = 7.4), according to EXAFS signals evaluation, or of defected/vacant (**) sites is more probable. This is likely explained by the difficulty of Th 4+ to replace directly Ti 4+ in [6]-coordinated (octahedral) sites due to the large difference in the relevant ionic radii (0.940 and 0.605 Å respectively).