Cement-based materials, like concrete and mortar, are widely used in radioactive waste repositories. A deep characterization of these heterogeneous materials, and of their main phases, is necessary to evaluate their capability of retaining critical radionuclides (RN).

In this study, the ion beam technique micro- Particle Induced X- Ray Emission (μPIXE) is used to characterize the concrete and mortar used in the Spanish low level waste repository. Two calcium silicate hydrate (CSH) phases with different Ca/Si ratio are also studied, because they are known to be amongst the most relevant phases, formed upon cement hydration, that retain RN. The retention of thorium on the above mention materials, as relevant tetravalent actinide, is also analyzed. Results are compared with Scanning Electron Microscopy- Energy Dispersive X-Ray Spectroscopy (SEM-EDX) analyses.

The National Atomic Energy Commission of the Argentine Republic is developing a nuclear waste disposal management programme that contemplates the design and construction of a facility for the final disposal of intermediate-level radioactive wastes. The repository is based on the use of multiple, independent and redundant barriers. The major components are made in reinforced concrete so, the durability of these structures is an important aspect for the facility integrity. This work presents an investigation performed on a reinforced concrete specifically designed for this purpose, to predict the service life of the intermediate level radioactive waste disposal facility from data obtained with several techniques. Results obtained with corrosion sensors embedded in a concrete prototype are also included. The information obtained will be used for the final design of the facility in order to guarantee a service life more or equal than the foreseen durability for this type of facilities.

Cement-based materials used in the construction of the repository for high/low level radioactive wastes may produce a highly alkaline calcium-rich groundwater (plume). The Ca ions react with soluble silicic acid, depositing calcium-silicate-hydrate (CSH) gel on the surfaces of the groundwater flow-paths and decreasing the permeability of the bedrock. Such a decrement of permeability may play a role in retarding the migration of radionuclides. In this study, the deposition behavior in a fracture was experimentally examined by using a micro flow-cell consisting of silicon plate (including a slit (60 mm×5 mm, or 60 mm×2 mm)) and granite-chip. The initial equivalent-aperture based on the square law was estimated in the range of 26 μm to 45 μm from the flow test of pure water.

In the experiments, a Ca(OH)2 solution of 6.36 mM (pH: 12.2 to12.5, including NaOH) was continuously injected into the flow system at a constant flow rate of 1 or 2 ml/h. The solution flowed on the surface of the granite-chip. In this study, we prepared two kinds of chips that differed in the treatment of the surface. One chip was roughly ground with #2000 sandpaper (hereinafter referred to as rough surface) and another was polished to mirror-like surface. As a result, on the rough surface the deposits of CSH gel appeared along flow-channels across mineral grain-boundaries, while the deposits on the mirror-like surface were relatively uniform. Furthermore, the permeability in the case of rough surface became smaller than that in the case of mirror-like surface, showing the repeats of rapid decrement and increment due to the relatively large roughness of the surface. In order to estimate the decrement degrees of permeability, a simple, one-dimensional mathematical model is proposed in this study.

Research reactors spent fuels disposal is a problematic area that conduces to the quest of feasible solutions for ensure safe destination of the spent fuels. In this work, a new method for processing spent fuel from MTR reactors is presenting. The main objective of this process, is the immobilization of principal radioactive elements that are present in the spent fuel, in order to achieve a suitable material which could be temporally stored safely. The Vitrocerus method involves ordinary physical procedures performed in a hot cell. It differs from conventional vitrification because there is a reduction in volume of glass material added.

The process propose the ceramization of a mixture of spent fuel plates (wich were milled and calcined previously) with natural uranium oxide (U3O8) to obtain the desired isotopic dilution (to low the U235 enrichment). At the same time, a small fraction of VG98/12 glass [1] was added to enhance low temperature sintering. The treatment and conditioning tasks proposed in Vitrocerus were tested on MTR fuel miniplates that simulate a real U3Si2 dispersed in Al fuel, which were successfully transformed into ceramic sintered pellets [2,3] with low enrichment, structural integrity, outstanding mechanical properties and water corrosion resistance. Eventually these pellets could be stored safely in an interim dry storage facility.

The results of sequential large-scale molecular dynamics (MD) simulations of radiation damage cascades in Gd2Ti2O7 and Gd2Zr2O7 are presented. Twelve alpha recoil cascades, each due to a recoil atom with 40 keV of kinetic energy, are performed in both materials and a stark contrast in behaviour observed. Topological connectivity analysis is used to analyse the structural evolution of the two systems. Our results provide important insight into accumulation of disorder in pyrochlore-structured ceramics.

The corrosion behavior of oxygen-free copper in anoxic sulfide solutions under nuclear waste disposal conditions was studied using electrochemical impedance spectroscopy (EIS), scanning electron microscopy (SEM) equipped with a focused ion beam (FIB), X-ray photoelectron spectroscopy (XPS) and Micro X-ray diffraction (μXRD). The film growth process and mechanism were elucidated using an Au marker test, and the contribution of solution diffusion to corrosion was demonstrated in magnetically-stirred experiments. The effect of groundwater chemistry, particularly chloride content on copper corrosion and film properties was characterized using long-term corrosion experiments.

A series of uranium-containing gadolinium zirconate samples have been fabricated at 1723 K in air. X-ray diffraction and Raman spectroscopy have confirmed pyrochlore or defect fluorite structures, while diffuse reflectance, X-ray absorption near edge structure and X-ray photoelectron spectroscopies indicate a predominantly U6+ oxidation state, even when Ca2+ was added to charge balance for U4+. The results demonstrate the potential of gadolinium zirconates as host materials for actinides.

Gas phase-optimized structures are used to assemble an ab initio molecular dynamics simulation of plutonium (IV) solvated in water. Hydrolysis is observed, and results are compared to experimental EXAFS data.

While simulation time is insufficient to be conclusive, evidence suggests that the 7-coordinate singly-hydrolysed complex, [Pu (OH) (OH2)6]3+, is most stable in our simulated environment. Energetic differences between the gas-phase optimised structure and the prevalent dynamic simulation structure are shown to be relatively small.

An ancient iron sword (ca 2,000 year old) excavated from Ohtake-nishi archaeological site in Osaka Prefecture, is one of the oldest cast iron swords in Japan. Because of a good state of preservation, such as the sharply-defined “Shinogi” (the longitudinal ridge) on the surface, we studied relationship between the burial environment and corrosion behavior. As environmental conditions, we investigated groundwater chemistry and corrosion rate with iron probe monitor, etc. Regarding corrosion state, we analyzed corrosion depths with an X-ray computed tomography and corrosion products with a portable X-ray diffractometer equipped with X-ray fluorescence spectrometer. As results, we found that the redox potential and dissolved oxygen level as environmental conditions were very low, and that the corrosion rate (7.5×10-4 mm/y) evaluated from measured corrosion depths was smaller than the probe corrosion rate (3.2–5.2×10-2 mm/y) by two orders of magnitude. The results suggested that the corrosion layer of siderite formed on the iron sword surface inhibited corrosion reaction.

Several national programs for geological disposal of higher activity radioactive waste will initiate siting via a call for volunteers. This is probably furthest advanced in Japan (NUMO) and the UK (NDA RWMD). In both these cases, rather than focusing on a reference repository concept, a catalogue of options has been developed. This requires a process to tailor design options to sites and compare particular design/site combinations as part of the site selection process. As has been noted in the longer-running Japanese program, the assessment of options must be as realistic as possible, rather than conservative, as in past nominating approaches to siting. Past demonstration work has focused on feasibility, showing that it is possible to implement a design. Determining practicality, which incorporates many more complex factors, is more difficult. Extensive desk studies can form a basis for such analysis but these must be rigorously tested, which makes full-scale tests in Underground Rock Laboratories (URLs) an essential component of any program. This paper highlights the importance of such demonstration projects and illustrates how they are vital in both improving understanding of the engineered barrier system and stakeholder engagement.

In many deep repository concepts spent nuclear fuel (SNF) will be disposed in canisters containing large amounts of iron. Intrusion of groundwater in a failed canister may occur under the presence of hydrogen, expected to be produced by the anoxic corrosion of iron. Compelling evidence now exists that hydrogen inhibits oxidative dissolution of SNF, the mechanism is however not fully understood. Hydrogen generally requires a catalyst in order to operate as a reductant. The metallic inclusions (ε-particles) present in SNF are a likely catalyst for this process due to their noble metal content. There is also evidence that the SNF UO2 matrix or doping of the UO2 with fission products can activate hydrogen. In most spent fuel experiments carried out under hydrogen, a decrease in concentration of all redox sensitive nuclides originating from a pre-oxidized layer is observed. Given their low concentrations and abundance in the fuel, it has however been difficult to detect any reductive precipitation on the fuel surfaces.

In this study, Cr(VI) oxyanions were employed as a redox sensitive marker, as Cr(VI) is expected to precipitate as Cr(III) oxide on the catalyst that activates hydrogen.

In the experiments PWR spent fuel (43 MWd/kgU) was leached in simulated groundwater (10 mM NaCl, 2 mM NaHCO3) at 25 and 70 ◦C under 5 MPa of hydrogen and dissolved Cr(VI). Dark green, Cr(III)-oxide was found to precipitate; mapping by electron microscopy (SEM-WDS) evidenced a Cr rich layer covering the fuel, suggesting that the whole fuel surface is catalyzing the reduction of chromium.

Ion-exchange resins are used for purification of radioactive liquid waste from nuclear reactors. After exhaustion, resins become intermediate level radioactive waste to be managed. They have to be immobilized before being stored to improve the leach resistance of the waste matrix and to maintain mechanical stability for safety requirements. The immobilized resins are thus contained in steel drums that can undergo internal corrosion depending on the presence of certain contaminants. This work shows an study of the corrosion susceptibility of steel drums in contact with cemented ion-exchange resins with different types and contents of aggressive species. Results show that the corrosion depth of the steel drums after a period of 300 years (foreseen life-span of the radioactive waste disposal facility), in the most unfavorable case (high chloride contamination), will be considerably lower than the thickness of the wall of the drums.

Termination of the Yucca Mountain repository program implies that U.S. spent nuclear fuel (SNF) and high-level radioactive waste may be stored for very long periods, i.e., more than 100 years. Many of the storage systems for commercial SNF are large, cylindrical, welded, stainless-steel canisters containing 24-80 SNF assemblies in an inert gas (helium) environment. Each canister is contained in a massive reinforced concrete structure for shielding, safety, and security reasons. Airflow in the annular space between the canister and the concrete structure removes SNF decay heat by natural convection.

Undetected defects in the canister-lid welds and stress corrosion cracking (SCC) in the canister-lid weld area may combine to form pathways for slow escape of the inert fill gas from the canister. SCC may not occur until many decades after SNF is loaded into the canister because SNF decay heat initially would keep the canister surface temperatures above the temperature where liquid water, a prerequisite for SCC, could condense. The period required for inert gas to escape from the canister once pathways have formed and for air subsequently to enter the canister may also be measured in decades, depending on the size and shape of the pathways.

In this paper, the authors explore the limits of weld-flaw detectability; when necessary conditions for SCC could occur in the canister-lid weld area; the period for loss of inert fill gas if through-weld cracks occur; and the possible consequences of fill gas loss. The authors also develop a list of research and study needs to address the possibility of SCC resulting in the loss of fill gas.

During irradiation in the reactor, a fraction of the fission product inventory will have segregated either to the gap between the fuel and the cladding or to the grain boundaries in the fuel. Of these nuclides, the behavior of the fission gases is best known. The part of the inventory that is rapidly released upon contact with water is designated the instant release fraction (IRF). Previous studies have shown that IRF and fission gas release (FGR) seem to be correlated. Studies of the instant release fraction from high burnup fuel is of interest for the assessment of the safety of a geological repository.

The instant release fractions of 129I and 137Cs from five different light water reactor (LWR) fuel rods with a burnup range of 43 to 75 MWd/kgU and a fission gas release range from 0.9 to 5.0 % were studied. Four types of fuel samples (pellet, fragment, powder and fuel rodlet) have been used in the experiments. The results show that the fuel sample preparation method has a significant impact on the release from high burnup fuel samples over the time period covered by this study. Leaching of high burnup fuel samples with fuel detached from the cladding shows the highest release. The fractional 129I release from such fragment samples is similar to the FGR in the corresponding rod. On the other hand, corresponding fractional release of 137Cs is lower.

Deep boreholes have been proposed for many decades as an option for permanent disposal of high-level radioactive waste and spent nuclear fuel. Disposal concepts are straightforward, and generally call for drilling boreholes to a depth of four to five kilometers (or more) into crystalline basement rocks. Waste is placed in the lower portion of the hole, and the upper several kilometers of the hole are sealed to provide effective isolation from the biosphere. The potential for excellent long-term performance has been recognized in many previous studies. This paper reports updated results of what is believed to be the first quantitative analysis of releases from a hypothetical disposal borehole repository using the same performance assessment methodology applied to mined geologic repositories for high-level radioactive waste. Analyses begin with a preliminary consideration of a comprehensive list of potentially relevant features, events, and processes (FEPs) and the identification of those FEPs that appear to be most likely to affect long-term performance in deep boreholes. The release pathway selected for preliminary performance assessment modeling is thermally-driven flow and radionuclide transport upwards from the emplacement zone through the borehole seals or the surrounding annulus of disturbed rock. Estimated radionuclide releases from deep borehole disposal of spent nuclear fuel, and the annual radiation doses to hypothetical future humans associated with those releases, are extremely small, indicating that deep boreholes may be a viable alternative to mined repositories for disposal of both high-level radioactive waste and spent nuclear fuel.

The first step in the processing of low- and medium-level liquid radioactive waste is the reduction in the volume of liquid containing small concentrations of radionuclides. Various methods for concentration of radioactive waste have been studied and developed at the Institute of Nuclear Chemistry and Technology, including membrane processes. Reverse osmosis was implemented at the Radioactive Waste Management Plant. Other methods such as ultrafiltration, membrane distillation, adsorption and different integrated processes were studied in the scope of national and international projects.

This paper presents the results of research performed at Institute of Nuclear Chemistry and Technology in Warsaw concerning radioactive liquid waste treatment and the programs of implementation of these methods at nuclear centers producing such wastes, as well as the plans for the utilization of the knowledge and experience and designs of schemes for radioactive waste management in future nuclear power industry. The integrated system developed on the basis of research will be a prototype for further implementation.

The microstructure change of the uranium-plutonium mixed oxide fuels (MOX fuels) irradiated in a fast reactor occurs because of a radial temperature gradient. To make detailed observations and elemental analyses of fuel samples, a field emission scanning electron microscope (FE-SEM) equipped with a wavelength-dispersive X-ray spectrometer (WDX) was installed in a hot laboratory.

Because fuel samples have high radioactivities and emit α-particles, the instrument was modified as follows :

1. 1) The instrument was attached to a remote control air-tight sample transfer unit between a shielded hot cell and the FE-SEM.

2. 2) The FE-SEM was installed in a lead shield box and the control unit was separately located outside the box.

After the installation, the microscopy and elemental analyses were applied to low burnup fuel samples. High resolution images were obtained and characteristic X-rays (U, Pu, and so on) emitted from the sample surface were measured. The technique has the great advantage of being able to evaluate the irradiated fuels in detail. In future work, samples of even higher radioactive will be observed and analyzed.

Development of numerical tools for performance assessment studies of radioactive waste disposal facilities, must address the management of the wide-ranging uncertainties associated with the long-term behaviour of these complex systems. Different approaches and assumptions are made in order to identify and describe relationships between the disposal system and its environment. They take into account, among other factors, the uncertainties associated with temporal evolution of the system within a proposed scenario; the landscape changes arising from future human actions, climate and geological events and processes; the relationships between components of the disposal system and its immediate environment; the behaviour and characteristics of radionuclides within the system and their role in contributing to radiation exposure. In all cases, the different scenario-based models are typically used to determine the radiological significance of potential future discharges from waste disposal facilities. However, it is important to keep always in mind that in any specific case, the purpose of developing and/or applying a model may vary from a simple calculation (e.g. to support concept development) to detailed site-specific performance assessment in support of a disposal license application. The assumptions and modelling simplifications that are appropriate to one type of calculation may not be so easily justified in different circumstances. In order to develop the capability of modelling different long-term scenarios for a generic disposal site for low and intermediate level radioactive wastes, implementation of models of both the near-field/geosphere and biosphere were performed using general approaches for geosphere-biosphere interface, with sub-models for the whole system.

The electrochemical reduction of H2O2 on SIMFUEL was investigated over the pH range 1 to 4. The mechanism at pH 4 is known to occur on UV species incorporated into a surface layer of UIV1-2xUV2xO2+x. However, below pH 3, reduction occurs on an adsorbed UVO2(OH) state which is unstable and oxidizes to insulating UVI before dissolving as UO22+. Both schemes are observed at intermediate pH’s. The presence of both low and high acidic regions at the electrode surface is determined by the combination of peroxide concentration, bulk pH and the surface diffusion conditions.

The geometries, relative energies and spectroscopic properties of α-isosaccharinate and D-gluconate complexes of uranyl(VI) are studied computationally using density functional theory. The effect of pH is accommodated by varying the number of water and hydroxide ligands accompanying gluconate in the equatorial plane of the uranyl unit. Their relative energies are found to be pH dependent, although the energetic differences between them are not sufficient to exclude the possibility of multiple speciation. The calculated uranyl stretching frequency decreases as pH increases, in agreement with previous experimental data. Three different coordination modes are studied.