705 research outputs found
Institute of Energy and Climate Research IEK-6 : nuclear waste management & reactor safety report 2009/2010 ; material science for nuclear waste management
No full textThis is the first issue of a new series of bi-annual reports intended to provide an overview of
research activities for the safe management of nuclear waste in the Institute of Energy and
Climate Research (IEK-6), Nuclear Waste Management and Reactor Safety devision in
Jülich. The report gives a thematic overview of the research in 2009 and 2010 by short
papers of five to eight pages. Some papers are discussing the work within different projects
with intensive overlap, such as the work on irradiated (reactor) graphite. Other projects are
represented by several papers, such as the work on ceramic waste forms. Furthermore,
some background information about the IEK-6 completes the picture of the activities of the
Institute. In summary, a hopefully comprehensive overview of the current research at IEK-6
will be provided.
During the years 2009 and 2010 a number of substantial changes took place within IEK-6. In
April 2009 Dirk Bosbach succeeded Reinhard Odoj as head of the nuclear waste
management part of the Institute. Bruno Thomauske joined the Institute in August 2009 as
head of the nuclear fuel cycle part. Therefore, the continuation of the nuclear waste
management research activities in Jülich is ensured. ..
Keramiken des Monazit-Typs zur Immobilisierung von minoren Actinoiden und Plutonium
No full textThe safe disposal of radioactive waste in deep geological formations is a challenging task of present and future generations. Innovative strategies as the conditioning of radionuclides in ceramic matrices can make a contribution here. This work points out monazite-type ceramics as potential waste forms for minor actinides and Pu. Several aspects concerning nuclear disposal as well as fundamental structural information were investigated. Lanthanide phosphate endmembers (LnPO) within the stability field of monazite (Ln = La-Gd) were synthesised within the scope of this work. To extend the knowledge of monazite phases, monoclinic TbPO- and DyPO-phases were prepared and characterised. Tb- and Dy-phosphates are situated in the xenotime stability field close to that of monazite. They can exist as metastable monazite phases. Structural characterisations of long- and short-range order were performed by X-ray diffraction, infrared (IR) and Raman spectroscopy. Structural data could be complemented, enhanced and gaps of knowledge could be filled by the first systematic consideration of the complete Ln-monazite-series (Ln = La-Dy). Furthermore, this work focuses on Sm-monazite phases. Samarium with an atomic number of 62 is located in the middle part of the lanthanides showing the monazite structure. Accordingly, it has a mean cationic radius within the Ln-monazite-series and hence shows a relative high flexibility regarding the incorporation of radionuclides with different radii. Sintering densities of SmPO ceramics were optimised by varying process parameters like pressure and number of pressing steps. An irregular texture as well as densities of 94% of the theoretical value could be achieved. The resistance of Sm-monazite against ionising radiation were examined. Radiation damages caused by the -decay of radionuclides incorporated in a ceramic matrix were simulated by computer calculations and experimentally by heavy ion bombardment of SmPO. Thin layers of the samples bombarded with Au ions show initiating effects of radiation damages enduring doses of D = 0.02 and 0.06, respectively, whereas a dose of D = 1.65 produced an amorphisation to a great extent in a lamella sample. However, crystalline areas remained due to recrystallisation processes in the material. The flexibility of the Sm-monazite’s crystal structure, necessary for the incorporation of radionuclides, is required for the use of SmPO ceramics as a suitable waste form. Therefore, Sm was substituted by other lanthanide cations, that served as surrogates for minor actinides such as Np, Am, and Cm or for Pu. Actinides reveal a similar [...
SGSreco – Radiologische Charakterisierung von Abfallfässern durch Segmentierte -Scan Messungen
No full textStarting from 2021, low and intermediate level radioactive waste produced in the Federal Republic of Germany will be finally disposed at a depth from 800m to 1300m in the Konrad Repository, close to the city Salzgitter. A prerequisite for the final disposal of radioactive waste packages is their conformance with national acceptance criteria. These acceptance criteria include among others radiological requirements for waste packages. To ensure a conformance of waste packages with these radiological requirements, experimental techniques are applied to characterize their radionuclide inventories. For this purpose, segmented -scanning is used worldwide as the standard non-destructive assay for the radiological characterization of waste drums. Segmented -scanning investigates predefined parts of a waste drum independently of each other using -spectrometry with a collimated detection system. Radionuclides are identified by their characteristic -linesin each recorded -spectrum, and two-dimensional count rate distributions are determined depending on the positions of the investigated predefined parts. The reconstruction of radionuclide specific activities by conventional methods requires a homogeneous matrix and radionuclide distribution within the whole drum. Thus, radionuclide specific activities are estimated using an analytical model based on the average count rate of a characteristic -line over all investigated parts of the waste drum. However, only 25% of all waste drums meet these requirements. It is therefore expected that the radionuclide specific activities for the majority of waste drums are miscalculated by several orders of magnitude. In this work, an analysis framework known as SGSreco is presented. SGSreco aims to ensure an accurate anda reliable reconstruction of radionuclide specific activities for homogeneous and spatially concentrated (point sources) radionuclide inventories. SGSreco uses an inverse approach. Within a first-guess reconstruction, point sources are identified by peaks in the associated count rate distributions. Using this information, the measuredcount rate distributions are fitted with a physically motivated geometric model by a likelihood minimization. Finally, the positions and activities of point sources as well as the activity of a homogeneous distribution of a radionuclide are reconstructed. For the first time, a priori unknown parameters, such as the density of the waste matrix or the thickness of absorber boxes, can be determined by SGSreco. Furthermore, SGSreco is capable of estimating reasonable statistical and systematic uncertainties. Reconstructions with SGSreco are benchmarked by simulation studies with waste matrices ranging from densities of 0.5 g cm to 2.3 g cm and by measurements with a reference drum and a test drum segment using the key nuclides 60Co and 137Cs. The performance of SGSreco is tested in terms of homogenous radionuclide inventories, ensembles of up to five point sources, single point sources in unknown waste matrices or absorber boxes, and spatially extended sources. It is shown, for example, that the activity reconstruction of homogenous radionuclide inventories is improved from deviations of around 30% for the conventional method to deviations of no more than 6%. The total activity of ensembles with up to five sources can be reconstructed with deviations between around −7% and 14%. Moreover, it is shown that the reconstructions with SGSreco are very accurate in all investigated cases. Point sources are localized with a precision of only a few mm. Due to computation times of less than one minute, SGSreco is applicable for the routine characterization of waste drums without limitations. Furthermore, SGSreco is compatible with existing -scanning hardware, making its implementation quick and flexible. In summary, SGSreco significantly improves the accuracy and reliability of the radionuclide specific activity reconstruction, allows a reasonable determination of statistical and systematic uncertainties, and enables the reconstruction of previously inaccessible parameters of matrix and radionuclide distribution, which may have a direct impact on the acceptance of waste for final disposal
Monazite type ceramics for conditioning of minor actinides : structural characterization and properties
No full textThe minor actinides (MA) neptunium, americium, and curium are mainly responsible for the long-term radiotoxicity of the High Active Waste (HAW) generated during the nuclear power operation. If these long-lived radionuclides are removed from the HAW by partitioning and converted by neutron fission (transmutation) into shorter-lived or stable elements, the remaining waste loses most of its long-term radiotoxicity. Thus, partitioning and transmutation (P&T) are considered as attractive options for reducing the burden on geological disposals. As an alternative, these separated MA can also be conditioned (P&C strategy) in specifically adapted ceramics to ensure their safe final disposal over long periods. At the moment, spent fuel elements are foreseen either for direct disposal in deep geological repositories or for reprocessing. The highly active liquid waste that is produced during reprocessing is conditioned industrially using a vitrification process before final disposal. Although the widely used borosilicate glasses meet most of the specifications needed, ceramic host matrices appear to be even more suitable in terms of resistance to corrosion. The development of new materials based on tailor-made highly specific ceramics with extremely stable behavior would make it possible to improve the final storage of long-lived high-level radiotoxic waste. In the framework of this PhD research project, monazite-type ceramics were chosen as promising host matrices for the conditioning of trivalent actinides. The focus on the monazite-type ceramics is justified by their properties such as high chemical durability. REPO4 ceramics are named monazite for RE = La - Gd (monoclinic symmetry) and xenotime for RE = Tb - Lu and Y (tetragonal symmetry). The objective of this study is to contribute to the understanding of the alteration behavior of such ceramics under the repository conditions. REPO4 (with RE = La, Eu) is prepared by hydrothermal synthesis at 200°C. Structural and morphological characteristics (using X-ray diffraction (XRD) and scanning electron microscope (SEM)) combined with physical and thermal properties of samples (using thermogravimetry, differential scanning calorimetry (TG-DSC) and dilatometry) are realized in order to study the behavior of monazite-type powder and pellets. The access to short-range-order spectroscopy (time resolved laser fluorescence spectroscopy (TRLFS) and extended X-ray absorption fine structure (EXAFS)) permits to understand the structure of ceramic waste forms at the molecular level. La-monazite matrices being doped with Eu (III) (as a non-radioactive chemical homologue for Am (III)) and Cm (III), TRLFS is used to explore the local structural environment of Eu and Cm within the monazite crystal structure. Eu (III) and Cm (III) are substituted on the La site of LaPO4. The single site of Cm (III) is found in four slightly different environments which is assumed to be due to a difference in the four La sites within a LaPO4 unit cell. Structural parameters of the Eu (III) species were also analyzed by EXAFS. The nearest neighbors of Eu (III) are modeled as 9.5 oxygen atoms. An essential parameter that describes the stability of the host phases is their dissolution rate obtained under conditions of relevance for final repositories. In this context, a set-up is developed and tested on crushed pellets. Normalized weight losses of lanthanum-phosphates and europium-doped lanthanum-phosphates, measured in acidic media at 90°C, are interpreted and compared against the previous findings from the literature. The normalized dissolution rate for La and Eu within (La, Eu)PO4 is between 1•10-5 and 1•104 g•m2•d-1, whereas the rate of Na, Cs and Sr in phosphate glass at room temperature in deionized water is about 1•10-2 g•m-2•d-1. Another essential parameter is their resistance to radiation. The damage created by the recoils accompanying alpha-decay can be simulated on ceramic matrices. Preliminary experiments are realized by means of ion bombardment. Kr2+ ions are implanted in La-monazite-type pellets, and the effects on the LaPO4 structure resulting from the Raman spectroscopy are poor. On a laboratory scale, the promising characteristics of the monazite mineral are found again in the synthetic phosphate. In particular, the doping of actinide surrogates is successful and the corrosion tests under repository conditions show a good resistance of the samples. The results achieved in this work confirm that among other favorable ceramics the “monazite-route” has to be further pursued regarding the research on the conditioning of MA
Production and Characterization of Monodisperse Uranium Particles for Nuclear Safeguards Applications
No full textThe International Atomic Energy Agency (IAEA) is the official body to apply nuclear safeguards toverify compliance with existing legal bilateral or multilateral safeguards agreements [a]. Environmental sampling is a very effective measure to detect undeclared nuclear activities. Generally, samples are taken as swipe samples on cotton. These swipes contain minute quantities of particulates which have an inherent signature of their production and release scenario. These inspection samples are assessed for their morphology, elemental composition and their isotopic vectors. Mass spectrometry plays a crucial role in determining the isotopic ratios of uranium. Method validation and instrument calibration with well-characterized quality control (QC)-materials, reference materials (RMs) and certified reference materials (CRMs) ensures reliable data output. Currently, the availability of suitable well defined microparticles containing uranium and plutonium reference materials is very limited. Primarily, metals,oxides and various uranium and plutonium containing solutions are commercially available. Therefore, the IAEA’s Safeguards Analytical Services (SGAS) cooperates with the Institute of Nuclear Waste Management and Reactor Safety (IEK-6) at the Forschungszentrum Jülich GmbH in a joint task entitled “Production of Particle Reference Materials”. The work presented in this thesis has been partially funded by the IAEA, Forschungszentrum Jülich GmbH and the Federal Ministry of Economic Affairs and Energy(BMWi) through the “Joint Program on the Technical Development and Further Improvement of IAEA Safeguards between the Government of the Federal Republic of Germany and the IAEA” (in brief: German Support Program, GER SP). In order to strengthen the IAEA’s analytical capabilities, a broad range of tailor-made uranium and plutonium containing particles with consistent characteristics are needed: (1) mono disperse particles with a certified value on the number of atoms per particle (2) mixed particles sizes and (3) artificial QC samples by embedding various monodisperse particle populations with different particle sizes onto swipe samples (these swipes could additionally contain a “dirt” matrix to simulate real-life samples). In the long run, these particles are targeted to be used for quality assurance, method validation and interlaboratory performance evaluations and finally as reference materials or even certified reference materials. The first step towards monodisperse microparticles was the development of pure uraniumoxide particles made from certified reference materials. This work in this thesis represents the efforts and results made during the last three years. A comprehensive outlook will be given later on. The focus of the dissertation is (1) the implementation of a working setup to produce monodisperse uranium oxide particles and (2) the characterization of these particles towards the application as QC-material. A successful working setup was implemented at IEK-6. Monodisperse uranium oxide particles were produced by spray pyrolysis. Spray pyrolysis is the production of aerosols and the subsequent thermal conversion to its corresponding oxides: A dilute hydro-alcoholic solution made from certified uranyl nitrate solutions was used to produce monodisperse aerosol droplets. Monodisperse aerosol droplets were generated using a vibrating orifice aerosol generator (VOAG). Particles were dried and thermally converted to uranium oxide within a preheating system and a four zone oven and after cooling they are removed from the system by inertial collection. All in all, the entire setup was designed to be a closed system that can even be operated inside a glove box. All components were designed to be easily replaceable. As cost-effective connections and tubes, Swagelok and KF connectors and flanges were used, which ensure a gas tight connection
Dissolution Behaviour of Innovative Inert Matrix Fuels for Recycling of Minor Actinides
No full textDuring the peaceful use of nuclear energy high level wastes, which contain long-lived radionuclides (plutonium, minor actinides) with high radiotoxicity, are generated worldwide. About 10,000 t of spent fuel are unloaded from commercial reactors each year. Most countries including Germany favour the direct disposal of spent fuel in deep geological formations. Other countries prefer the reprocessing of the spent fuel to recycle uranium and plutonium. In Europe commercial reprocessing is currently performed in La Hague and Sellafield. A future alternative would be provided by closing the fuel cycle in the context of the partitioning and transmutation strategy (P&T). This method separates and converts the long-lived radionuclides into stable or short-lived nuclides via neutron induced reactions in dedicated facilities. The P&T strategy has potential to significantly reduce the radiotoxicity and the volume of the radioactive waste; however it cannot obviate the need of a final repository. The transmutation of minor actinides can be performed in different reactor types, including accelerator driven systems, which consist of a subcritical reactor core and an external accelerator. The accelerator driven system (ADS) fuel consists of the fissile material (AnO) which is spread in an inert matrix to improve the thermal properties of the fuel. Within this work two different fuels containing actinide oxides as fissile material and ceramic magnesium oxide (CerCer) or metallic molybdenum (CerMet) as matrix material are under investigation. The dissolution and separation issues for inert matrix fuels (IMF) have not yet been investigated coherently. It is of crucial importance to take into account the behaviour of the matrix elements in the dissolution and separation processes and to check their compatibility with future waste management requirements. The dissolution and the subsequent hydro or pyrometallurgical treatment of the fuel are crucial steps in the reprocessing process. A complete dissolution of the actinide oxide and the matrix material, or a selective dissolution, where one of the components remains undissolved, can be considered. To investigate there processability of molybdenum and magnesia based inert matrix fuels reference samples containing variable amounts of CeO, which serves as surrogate for plutonium dioxide, have been prepared based on a comprehensive compactibility and sinterability investigation. The pellets were thoroughly characterized by means of density measurements, micro hardness measurements, scanning electron microscopy (SEM) investigation, and X-ray diffraction(XRD). The dissolution rate was studied in macroscopic experiments as a function of acid concentration and temperature. Magnesium oxide is soluble even under mild conditions. The dissolution rates of MgO at different acid concentrations are rather similar, whereas the dissolution rate is strongly dependent on the temperature. Additionally, the MgO dissolution process was investigated following a microscopic approach. Detailed SEM investigations show a heterogeneous reactivity of the MgO pellet’s surface. A model was developed to describe the evolution of the pellet surface area and a surface normalized dissolution rate was calculated. The activation energies of MgO dissolution in nitric acid have been calculated from the Arrhenius plot for different acid concentrations and indicate a surface controlled dissolution mechanism. During the dissolution of MgO/CeO pellets the MgO dissolves completely, while the bulk of CeO remains undissolved, allowing a separation of the actinides and the matrix during the dissolution process
Production and characterization of monodisperse uranium particles for nuclear safeguards applications
No full textThe International Atomic Energy Agency (IAEA) is the official body to apply nuclear safeguards to verify compliance with existing legal bilateral or multilateral safeguards agreements [a]. Environmental sampling is a very effective measure to detect undeclared nuclear activities. Generally, samples are taken as swipe samples on cotton. These swipes contain minute quantities of particulates which have an inherent signature of their production and release scenario. These inspection samples are assessed for their morphology, elemental composition and their isotopic vectors. Mass spectrometry plays a crucial role in determining the isotopic ratios of uranium. Method validation and instrument calibration with well-characterized quality control (QC)-materials, reference materials (RMs) and certified reference materials (CRMs) ensures reliable data output. Currently, the availability of suitable well defined microparticles containing uranium and plutonium reference materials is very limited. Primarily, metals, oxides and various uranium and plutonium containing solutions are commercially available. Therefore, the IAEA’s Safeguards Analytical Services (SGAS) cooperates with the Institute of Nuclear Waste Management and Reactor Safety (IEK-6) at the Forschungszentrum Jülich GmbH in a joint task entitled“Production of Particle Reference Materials”. The work presented in this thesis has been partially funded by the IAEA, Forschungszentrum Jülich GmbH and the Federal Ministry of Economic Affairs and Energy(BMWi) through the “Joint Program on the Technical Development and Further Improvement of IAEA Safeguards between the Government of the Federal Republic of Germany and the IAEA” (in brief: German Support Program, GER SP). In order to strengthen the IAEA’s analytical capabilities, a broad range of tailor-made uranium and plutonium containing particles with consistent characteristics are needed: (1) monodisperse particles with a certified value on the number of atoms per particle (2) mixed particles sizes and (3) artificial QCsamples by embedding various monodisperse particle populations with different particle sizes onto swipe samples (these swipes could additionally contain a “dirt” matrix to simulate real-life samples). In the long run, these particles are targeted to be used for quality assurance, method validation and interlaboratory performance evaluations and finally as reference materials or even certified reference materials. The first step towards monodisperse microparticles was the development of pure uraniumoxide particles made from certified reference materials. This work in this thesis represents the efforts and results made during the last three years. A comprehensive outlook will be given later on. The focus of the dissertation is (1) the implementation of a working setup to produce monodisperse uranium oxide particles and (2) the characterization of these particles towards the application as QC-material. A successful working setup was implemented at IEK-6. Monodisperse uranium oxide particles were produced by spray pyrolysis. Spray pyrolysis is the production of aerosols and the subsequent thermal conversion to its corresponding oxides: A dilute hydro-alcoholic solution made from certified urany lnitrate solutions was used to produce monodisperse aerosol droplets. Monodisperse aerosol droplets were generated using a vibrating orifice aerosol generator (VOAG). Particles were dried and thermally converted to uranium oxide within a preheating system and a four zone oven and after cooling they are removed from the system by inertial collection. All in all, the entire setup was designed to be a closed system that can even be operated inside a glove box. All components were designed to be easily replaceable. As cost-effective connections and tubes, Swagelok and KF connectors and flanges were used, which ensure a gas tight connection. It was demonstrated that the particle size can be controlled primarily by the aerosol precursor solution and the production parameters during the aerosol generation - in particular the liquid feed rate and the frequency of the orifice. The final particle morphology is controlled by the precipitation conditions during the conversion from aerosol droplets to solid entities. Small changes to these parameters have a significant influence on the final geometry, size and morphology. The second part of this thesis deals with the characterization of microparticles. A selection of particles was chosen to present the developments over a period of 12 months. Scanning Electron Microscopy coupled with Energy Dispersive X-Ray Spectroscopy (SEM-EDX) was used for various applications, e.g. to verify the elemental content and to assess the size and geometry of the particles. Furthermore, automated particle assessments over large areas were performed. It was demonstrated that the particle batches show an almost monodisperse size distribution. Combined Focused Ion Beam(FIB-SEM) studies revealed the presence of a porous inner structure for all solid particles. Hence, there sulting overall density was less than expected. Time of Flight Secondary Ionization Mass Spectrometry (TOF-SIMS) studies evaluated the elemental content and demonstrated the need for cleanliness since minute quantities of contaminations could be found in single particles. Micro Raman investigations were used to determine the crystallinity, crystal orientation and uranium species. The measurements showed that particles primarily consist of U3O8. Parts consist of Meta-schoepite and U(IV)-hydroxide which indicates residual water inside the crystal lattice. Micro Raman investigations were performed at CEA(Ile de France) and at the TU-Vienna. SIMS measurements were performed at Safeguards Analytical Services – Environmental Sample Laboratory (SGAS-ESL) on the Large Geometry-SIMS (LG-SIMS) with the scope to assess their performance as a QC material. Particles produced at Jülich were also compared directly against existing QC- and reference materials. Investigations and characterization assays on monodisperse microparticles indicate reproducible results and LG-SIMS investigations indicate equal or even better performance than existing reference materials. It can be concluded that SIMS-experiments indicate a consistent uranium mass per particle. Furthermore, SIMS analysis implies consistent and predictable performance regarding the isotopic content, hydride formation rate, total evaporation profiles and a better performance than existing certified reference materials (CRMs)
Secondary Uranium Phases of Spent Nuclear Fuel– Coffinite, USiO, and Studtite, UO . 4HO – Synthesis, Characterization, and Investigations Regarding Phase Stability
No full textThe miscibility behavior of the USiO – ThSiO system was investigated. The end members and ten solid solutions UxTh()SiO with x = 0.12 – 0.92 were successfully synthesized, without formation of other secondary uranium or thorium phases. Lattice parameters of the solid solutions evidently follow Vegard’s Law. Investigation of the local structure with EXAFS reveals small differences between U and Th environment attributed to different atomic radii of the metal atoms but no implications for a miscibility gap. The data provided confirms complete miscibility for the system USiO – ThSiO. The structure of the end members was studied in detail with XRD and discussed with special regard to the oxygen positions and the often neglected Si-Obond length. USiO could be obtained without UO impurities and the lattice parameters derived from Rietveld refinement as c = 6.2606(3) and a = 6.9841(3) . The Si-O distance in USiO appears to be 1.64 , which is more reasonable than earlier reported values. Synchrotron X-ray powder diffraction pattern and Raman spectra of synthetic coffinite, USiO, were obtained for pressures up to 35 GPa and 18 GPa, respectively. From the changes in the diffraction pattern it can be concluded that USiO undergoes a first order phase transition from zircon-type (space group I 4/amd) to scheelite-type structure (space group I 4/a) at 15 GPa and room-temperature. Contrary to earlier reports, the data indicates that this transition is completely reversible upon pressure release. Pressure dependencies of the Raman modes for the zircon structured phase are larger than those reported for hafnon, HfSiO, and zircon, ZrSiO, indicating that coffinite, USiO, is more compressible than these orthosilicates. Bulk moduli fitted from the p-V data for the zircon-type and scheelite-type USiO phase are compared to those known to literature for other MSiO (M = U, Hf, Zr) compounds. The bulk modulus for zircon-type USiO is 180(7) GPa and hence lower than those of ZrSiO (205 GPa) as expected from the larger unit cell. The pressure dependence of the Raman modes of USiO was studied upto 18 GPa, yet no abrupt changes of peaks or in the peak shifts appear. Furthermore it could be established, that the B- and the A-modes of the SiO -tetrahedron in the Raman spectrumare very close and overlap at ambient conditions. [...
Fundamental Insights into the Radium Uptake into Barite by Atom Probe Tomography and Electron Microscopy
No full textRecently, the BaRaSO solid solution has been investigated with regard to its applicability to the long-term safety of spent nuclear fuel (SNF) disposal. As Ra originates from the U decay chain, its concentration in SNF builds up with time. In some scenarios for the direct disposal of SNF taken from the Swedish license applicationfor a final SNF repository, Ra dominates the dose after 100,000 years. Currently,the solubility of Ra is considered to be controlled by the formation of Ra-SO in the Swedish license application as the BaRaSO solid solution characteristics were not sufficiently investigated at the point of submission. The BaRaSO solid solution could be considered as solubility controlling phase for Ra if the uptake mechanism of Ra into barite was understood in more detail. Barite can occur as a primary phase in the surrounding of the future repository or as a secondary phase within nuclear waste due to the different positions of Ba and Ra within SNF. In the case of SNF corrosion, Ba would come in contact with water first. Sulfate-containing water would lead to barite precipitation. Therefore, a system is most likely where pre-existing barite is in equilibrium with an aqueous solution into which Ra then enters. Recent studies comprising long-term batch recrystallization experiments propose a kinetically influenced uptake of Ra into barite that equilibrates into a thermodynamically controlled situation within 800 days. This thesis provides the first detailed four -dimensional characterization of the Ra uptake into barite by combining three-dimensional sample characterization with the temporal evolution. To understand the mechanism of Ra uptake into barite, two types of barites (SL and AL barite) obtained from batch recrystallization experiments of previous studies were characterized prior to, during and after the Ra uptake. Acombination of different state-of-the-art high-resolution microscopy techniques was used to answer the questions regarding (1) the internal microstructure of the initial barite (2) the role of this internal microstructure during the Ra uptake and (3) the changes in the Ra distribution within the barite. [...
Cyclotron irradiation on tungsten & co-relation of thermo-mechanical properties to displacement and transmutation damage
No full textNeutron damage is a major deciding factor in the commercialisation of a fusion power plant. Neutron damage inflicted on the walls of the reactor during operation, leads to changes in the behaviour of materials and ultimately decides the life time of the component. Consequently, it is essential that fusion relevant materials are tested under fusion irradiation conditions in order to qualify them, prior to use. Tungsten is a key material for the plasma facing component in a fusion reactor, and is located directly in the path of high energy fusion neutrons. Currently, it is not possible to test the change in material behaviour under high energy neutrons as there exits no high flux fusion neutron source. Moreover, high flux fission reactors are unable to re produce the high energy neutron damage. However, this work demonstrates the use of 30 MeV protons to induce fusion relevant neutron damage on tungsten. This work involves the first irradiation of tungsten using high energy protons (30 MeV). A complete irradiation cycle, including irradiation planning, sample design and manufacturing, polishing, irradiation, the setting up of post irradiation devices and post irradiation investigation was carried out within the scope of this work. Optimal sample geometry for accelerator irradiations, which is also directly comparable and compatible with fission reactor irradiations, was manufactured. The sample holder was designed such that in-situ temperature measurements were possible for the first time. Additionally, hot cell and remote handling conforming, punch and indentation testing have been developed and demonstrated through the use of irradiated active samples. In order to understand proton damage, pure tungsten was irradiated using three different proton energies 3, 16 & 30 MeV. The 3 MeV proton irradiation produces pure displacement damage, while the 16 & 30 MeV induce a combination of displacement and transmutation damage. Moreover, instrumented indentation was performed on the irradiated samples in a radiation environment (controlled areas). For all proton energies, an irradiation hardening of ~0.6 GPa was observed at low doses of 0.003 dpa for 30 MeV protons and 0.005 dpa for 3 MeV protons. Further experiments with 3 MeV protons displayed an initial further increase followed by saturation at 0.4 dpa. A similar behaviour has been reported with self ion irradiations on pure tungsten. The TEM observations of 3 MeV proton irradiated tungsten shows the development of dislocation loops, which grow in size but also achieve a saturation in loop density. This correlates well with the saturation in irradiation hardening. Irradiation modelling was performed using MCNP6.1 and FISPACT-II on both the sample and the sample holder to estimate the damage capability of 30 MeV protons. Post irradiation, gamma analysis showed good agreement with the modelling. Additionally, dose rate measurements were in-line to estimates from the simulations. This, by extension validates the transmutation capability of 30 MeV protons and their ability to simulate fusion neutron damage in W
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