Intermediate Level Radioactive Waste Research Articles

Bimodal structured chromium-tungsten composite as plasma-facing materials: Sinterability, mechanical properties, and deuterium retention assessment

Even though tungsten (W) is considered to be the most favored plasma-facing material (PFM) for nuclear fusion reactors thanks to its beneficial physical properties, its advantage for those plasma-facing components to be subjected to medium heat flux and high neutron irradiation dose is compromised by unfavorable nuclear behavior such as intermediate-level radioactive waste activation and embrittlement caused by lattice damage and transmutation. Chromium (Cr) or Cr-based materials may have the potential to mitigate these shortcomings while maintaining acceptable thermal and physical performances required for PFM. They could offer attractive design options, particularly for those parts where the particle fluxes are relatively low and the heat flux is not too high (≤ 5 MW/m²) while nuclear loads are high (≥ 5 dpa). In this article, we report the first results of a preliminary technology feasibility study for an engineered Cr-based material fabricated based on a powder-metallurgical route. To explore an optimal metallurgical condition ensuring reasonable mechanical properties and low deuterium retention, a novel two-phase Cr-W composite was developed via current-assisted rapid sintering at low sintering temperature. The composite showed a cell-like bimodal microstructure consisting of a coarse-grained Cr phase surrounded by a 3D interconnected network of an ultrafine-grained W phase. The composite exhibited substantial improvements in ductility as well as high-temperature strength compared to pure W and Cr. After undergoing W heavy ion irradiation, the damaged Cr-W composite with a two-phase bimodal microstructure showed a comparable deuterium retention capacity to pure W. Finally, the potential applicability as PFM is discussed.

An experimental study on the measurement of gas diffusivities in un-saturated clay based materials

Gas migration in porous media in the context of nuclear waste disposal is dominated by diffusion. The availability of abundant data on diffusion coefficients of different gases in many clay based materials such as Boom Clay, Opalinus Clay and Callovo-Oxfordian Clays provides a good overview of the rates of gas transport through these different clay rocks [1]. Such formations are being considered as potential host rocks for the disposal of high and intermediate level radioactive wastes across Europe. Diffusion coefficients have been studied with particular interest due to the unavoidable gas generation phase in the nuclear waste repository [2]. However, all the diffusion measurements performed so far have been under fully saturated conditions, with little experimental data on gas diffusivities in partially saturated conditions, which have been hypothesized to exist during the early part of the life cycle of the repository [3]. The objective of the current study is to provide an overview of the effect of desaturation on the rate of diffusive transport of gases in partially saturated clay based materials.
 The experimental methodology is based on the double through diffusion setup used in previous studies to measure gas diffusivities in saturated clays [1]. However, the setup is modified so as to keep a fixed level of unsaturation in the clay sample. This is done by employing the vapour equilibrium technique, using oversaturated salt solutions in both upstream and downstream reservoirs to maintain a constant relative humidity in the entire setup [4].
 Preliminary pore network modelling has already shown that the gas diffusivity increases with a decrease in water saturation of the clay material. However, the predicted trend also suggests that this increase is only marginal and should not normally exceed an order of magnitude. This is shown in the figure below.
 All the diffusion measurements are carried out on synthetic clay samples of dimensions 3.8 cm (diameter) and 2 cm (height). In Table 1, the Sw stands for water saturation. The compositions of the synthetic clay samples (60% clay, 20% sand, 20% silt) have been adjusted so as to most closely mimic the mineralogical compostion of Boom Clay. Synthetic samples are used instead of natural samples like Boom Clay because of their better physical response to drying than Boom Clay.
 The selected control variable to lower the saturation of the material is suction, which is imposed by changing the relative humidity of the environment using oversaturated salt solutions [4]. Suction can be kept constant given constant ambient temperature and pressure. Thus, the clay sample is conditioned at a certain suction, which in turn leads to desaturation in the range of 70-100% of total saturation depending on the level of suction. For this study, the saturation range has been fixed in between 70-100% of total saturation. This range has been chosen because of what is expected in a real life nuclear waste repository. The first measured diffusion coefficients in two partially saturated samples are shown in Table 1. The initial measurements of gas diffusion coefficients in unsaturated synthetic clays show that the increase in diffusivity of helium is higher than that of argon. Helium is the second lightest known element and has an extremely small molecular size and hence the impact of desaturation on helium diffusivity could be more pronounced than the same for heavier gases.

Two-phase reactive transport modeling of heterogeneous gas production in a low- and intermediate-level waste repository

Abstract. The widely proposed approach to dispose of low- and intermediate-level radioactive waste is to store it in a deep underground repository with multiple barriers. A typical gallery is filled with concrete containers accommodating cemented steel drums of waste. Inside a container, different gases may be produced by (bio)chemical reactions, which include pH-dependent anoxic corrosion of metals and the degradation of organic matter. Both reactions consume water and may lead to pressure buildup and transport of gas, both within and around the repository. In order to investigate the controlling factors of this gas production process, a coupled reactive transport model of component-based two-phase flow in the OpenGeoSys framework is adopted here. The numerical study of Huang et al. (2021) has shown that a realistic internal structure of a waste package, including the heterogeneous distribution of materials with different chemical and hydrological properties, and the exchange of mass at the boundaries are key factors that determine the evolution of the waste package. Based on the study of Huang et al. (2021), the geometric configuration of the model has been further extended to reflect the various conditions of a multi-container disposal in a gallery. In a two-dimensional setup several model scenarios have been designed and simulated to study the change in gas production rate over time in relation to water availability in various geological and waste storage setups. In this presentation, we show simulation results covering the geochemical evolution of a waste package over 500 years. It is found that the initial water content in the waste compartment only controls the gas production rate for the first 40 to 60 years. The early pressure buildup and gas release rates are largely controlled by several critical parameters, including the permeability of cement material and water availability at the boundary. The sensitivity of these parameters is currently being investigated in detail.

Radiolytic degradation of cellulosic materials in nuclear waste: Effect of oxygen and absorbed dose

Lignocellulosic materials can be found in a significant fraction of the current low- and intermediate-level radioactive waste. During storage and disposal, radiolytic degradation of such materials can be expected, under oxic or anoxic conditions. This degradation may lead to a significant gas production and changes in the physico-chemical properties of the lignocellulosic materials, which can affect the formation of the known radionuclide-complexing agent isosaccharinic acid (ISA) as well as other (possibly complexing) degradation products during disposal. Hence, in the present work the radiolytic degradation of cellulosic tissues – realistically found in radioactive waste – was investigated under various storage and disposal conditions. For this, cellulosic tissues were exposed to γ-irradiation in gas-tight containers under oxic or anoxic conditions, at an absorbed dose ranging up to 1.4 MGy and at two different dose rates. Our results show that mainly H2, CO and CO2 are produced during irradiation of tissues, though also small amounts of CH4 are formed. The presence of oxygen does not affect the generation of H2, but results in a significant increase in the yields of CO, CO2 and CH4. Furthermore, radiation-induced chain scission is observed, causing a decreasing polymerization degree with increasing absorbed dose. Amorphization of the cellulose microstructure occurs significantly at high doses of gamma rays (≥ 0.8 MGy). An increase in the concentration of reducing functional groups is observed with increasing absorbed doses as well. For irradiation under anoxic conditions, this increase is correlated with the observed chain scission. In contrast, additional oxidation processes occur when irradiating cellulosic tissues in the presence of oxygen, resulting in a partially oxidized polymer backbone without causing considerably more chain scission or amorphization. These radiolytic changes to the cellulose structure, both under anoxic and oxic conditions, may enhance its hydrolytic degradation under the hyper-alkaline conditions of long-term final disposal, resulting in a faster production of radionuclide-complexing agents.

A novel method for gamma spectrum analysis of low-level and intermediate-level radioactive waste

A novel method for gamma spectrum analysis of low-level and intermediate-level radioactive waste

Contribution of HADES URL to the development of the Cigéo project, the French industrial centre for geological disposal of high-level and long-lived intermediate-level radioactive waste in a deep clay formation

Abstract In the 1980s, HADES (High-Activity Disposal Experimental Site) was the first underground research laboratory (URL) dedicated to the study of the geological disposal of radioactive waste in a deep clay formation, the Boom Clay. It was not until the early 2000s, after a siting process, that ANDRA implemented the Meuse/Haute-Marne URL, in the Callovo-Oxfordian formation at a depth of about 500 m in order to develop the Cigéo project (French industrial centre for geological disposal). ANDRA therefore relied heavily on the work carried out in HADES, through numerous co-operation projects (participation in in situ experiments) both between ANDRA and ONDRAF/NIRAS and SCK CEN (EURIDICE) and/or with Mont Terri consortium, and within European projects (CLIPEX, RESEAL, etc.). This was driven by a dual objective: (1) to prepare its own experimental programmes in the Meuse/Haute-Marne underground laboratory (methodology, experimental devices and protocols, etc.); and (2) to acquire general knowledge on the behaviour of argillaceous rocks, in particular in terms of similarity and differences between the various argillaceous rocks. This paper illustrates the contribution of HADES to the ANDRA programme. This concerns the characterization of the claystone behaviour, host rock and swelling clay-based seals (hydromechanical, thermo-hydromechanical, excavation damaged zone, etc.), and the design and the behaviour of underground structures and seals in deep clay formation (constructability, lining/support, etc.).

Microstructural investigation into the damage behaviour of MX80 bentonite pellet under wetting/drying path using Mercury Intrusion Porosimetry (MIP) and Micro-Computed Tomography (μCT)

MX80 bentonite pellet has been considered as a promising sealing/backfilling material in geological disposal facilities for high-level and intermediate-level radioactive wastes. As the sealing/backfilling material which is subjected to hydration by pore water or vapour from host rock can be damaged by wetting-induced cracking, it is essential to clarify the suction effect on the damage behaviour of the material. In this study, a combined Mercury Intrusion Porosimetry/Micro-Computed Tomography (MIP/μCT) method was applied to investigate the microstructural damage of the MX80 bentonite cylinder-shaped 32-mm pellets equilibrated under suctions s from 113 to 1.5 MPa for wetting and from 113 to 262 MPa for drying. The enlarged pore size distribution (PSD) curves by MIP/μCT revealed an insignificant drying-induced closure of cracks, as opposed to massively developed wetting-induced cracks, with the size of the largest cracks reaching several millimetres. This is confirmed by the μCT observation. Indeed, the μCT images showed a negligible effect of drying on cracking, contrary to the effect of wetting characterised by significant cracking from the marginal area to the central part of pellet, indicating significant damage of pellet by wetting. Additionally, only horizontal cracks were identified at s ≥ 82 MPa. Vertical cracks emerged and propagated only at s ≤ 57 MPa, together with further expanding of horizontal cracks, due to the preferentially horizontal orientation of clay particles resulting from the one-dimensional compaction in the pellet fabrication. A damage coefficient D was defined as the ratio of the volume of macro-pores to the total volume of pellet. D increased slightly upon drying from suction 113 to 262 MPa and significantly upon wetting from suction 113 to 3.2 MPa, but kept almost unchanged when wetting furthered from suction 3.2 to 1.5 MPa because the clogging of macro-pores by the exfoliation effect slowed down the cracking rate.

Strontium interaction with unblended and Low-pH cementitious wasteforms: pH-dependent leaching experiments and geochemical modelling

Cementitious materials are widely-applied for the solidification and stabilization of low and intermediate level radioactive waste streams, that often contain strontium-90. Low-pH cement pastes, based on ordinary Portland cement (OPC) supplemented with pozzolanic additives, were found to have better Sr retention capabilities than unblended OPC pastes. However, precise definition of the phases and processes that control Sr retention in unblended and low-pH OPC pastes is still lacking. The present study combines experimental work with thermodynamic modelling to investigate the retention of Sr in unblended pastes compared to low-pH pastes. The unblended and low-pH pastes were prepared each in two formulations: one based on reagent grade alite (tri-calcium silicate) and the other based on OPC. The chemical and mineralogical behavior of the paste were studied by pH-dependent leaching test. The pastes characteristics were compared in order to assess the impact of: (1) the addition of amorphous silica to the pastes and (2) the presence of minor OPC phases (which are absent in the alite-based pastes) on Sr binding. The low-pH systems showed one order of magnitude lower Sr leachability than the unblended systems at the natural pH. Moreover, the leaching curves indicated that the low-pH systems bind Sr effectively at pH > 7 while for the unblended systems binding is effective only at pH > 10. These results are attributed to the difference in Sr binding to low Ca/Si C–S–H, dominant in low-pH systems, versus the high Ca/Si C–S–H, dominant in unblended systems. Additionally, it was shown that Sr binds mainly to the C–S–H phase rather than to minor phases. The leaching results were thermodynamically modelled and correlated successfully with the experimental findings. A novel modelling approach is offered for the consideration of partial pozzolanic reaction and SiO2-gel formation. The effect of these processes on strontium retention has been successfully demonstrated.

Research on the γ spectrum-unfolding method of low- and intermediate-level radioactive waste based on LaBr3(Ce) detector

To apply the LaBr3(Ce) detector with a higher full-energy peak detection efficiency and larger activity detection range to the measurement of low- and intermediate-level radioactive waste (LILW), we examined the problem of poor energy resolution. We analyzed the characteristics of key nuclide energy spectra and established an adaptive filter function to form a deconvolution energy spectrum analysis method for intelligent nuclide identification and quantitative calculation. Compared with the traditional deconvolution method, the average overlapping peak resolution was improved by 74.5%, achieving the qualitative analysis and quantitative calculation of all nuclides. By using the novel method to analyze the experimental spectrum of Co-60 and Cs-137, the results showed that the nuclide was accurately identified with a mean relative error of 0.85% for the peak area calculations. The analysis of the spectra of the compositions of the five LILW nuclides was obtained from Monte Carlo simulations using different methods. The results showed that, for the overlapping peaks with peak area ratios of 1:1:1:1:1, compared with the traditional deconvolution method, the peak area calculation error of this method was reduced from 73.09% to 2.12%. For the overlapping peaks with the peak area ratios of 5:1:2:3:1 and 5:1:20:10:1, the traditional spectrum-unfolding method could not resolve the weak peaks, and the maximum energy deviation was 8 keV. This method achieved accurate identification of all energies, and the error of the weak peak was 7.19%, and the error of the other peaks was less than 2.23%. The results of this study provided key technical support for the application of LaBr3(Ce) detectors to the nuclide measurement of LILW.

Toward Deep Decontamination of Intermediate-Level-Activity Spent Ion-Exchange Resins Containing Poorly Soluble Inorganic Deposits

Spent ion-exchange resins (SIERs) generated yearly in large volumes in nuclear power plants (NPPs) require particular predisposal handling and treatment with the primary objectives of waste volume reduction and lowering the disposal class. Deep decontamination of the SIERs using solution chemistry is a promising approach to reduce the amount of intermediate-level radioactive waste (ILW) and, thus, SIER disposal costs. However, the entrapment of nonexchangeable radionuclides in poorly soluble inorganic deposits on SIERs significantly complicates the implementation of this approach. In this work, the elemental and radiochemical compositions of inorganic deposits in an intermediate-level-activity SIER sample with an activity of 310 kBq/g have been analyzed, and a feasibility study of SIER decontamination using solution chemistry has been conducted. The suggested approach included the magnetic separation of crud, removal of cesium radionuclides using alkaline solutions in the presence of magnetic resorcinol-formaldehyde resin, removal of cobalt radionuclides using acidic EDTA-containing solutions, and hydrothermal oxidation of EDTA-containing liquid wastes with immobilization of radionuclides in poorly soluble oxides. The decontamination factors for 137Cs, 60Co, and 94Nb radionuclides were 3.9 × 103, 7.6 × 102, and 1.3 × 102, respectively, whereas the activity of the decontaminated SIER was 17 Bq/g, which allows us to classify it as very low-level waste.

Design of an IoT-based indoor tracking and condition monitoring system for the safe and transparent management of drums storing low- and intermediate-level radioactive waste

ABSTRACT Regulatory agencies and civil society organizations concerned with radioactive waste management are demanding maximized safety management of radioactive-waste-containing drums using advanced technology for more efficient and safe management. In the present paper, we propose a sensor system that can track and monitor drums containing low- and intermediate-level radioactive waste using advanced technology based on the Internet of Things (IoT). The proposed system consists of drum nodes, pallet nodes, and gateways. A drum node and magnetic sensor installed on a drum can be used to remotely check whether the drum lid is open or closed, and the location of the drum can be tracked using a near-field communication reader and an infrared module installed on the pallet node. Considering that radioactive waste is stored for a long time period, the proposed sensors are designed for low power consumption. Moreover, a pilot test involving 48 drums and 12 pallets was conducted to verify the system prototype. Accordingly, from the results of the pilot test, drawbacks were noted and solutions were proposed to improve the system in future work. Implications: We designed a test bed by fabricating a sensor-system prototype and used it in a simulation experiment. The results of this study will be used as basic data for establishing safety measures for radioactive waste management in the future through computer simulation of radioactive waste anomalies in a digital-twin system.

Evaluation of Fillers for Magnesium Potassium Phosphate Cement (MKPC) for the Encapsulation of Low and Intermediate Level Metallic Radioactive Wastes.

This study investigates the effect of coal fly ash (FA), wollastonite (WO), pumice (PM), and metakaolin (MK) as filler materials in the rheological, mechanical, chemical, and mineralogical properties of a magnesium potassium phosphate cement (MKPC), designed for the encapsulation of low and intermediate level radioactive wastes containing reactive metals. Workability, compression strength, dimensional stability, pH, chemical composition, and mineralogical properties were studied in different pastes and mortars of MKPC with a fixed molar ratio of MgO/KH2PO4 = 1. No new mineral phases were found with the addition of the fillers, denoting their low chemical impact on the MKPC system. Moreover, all formulations with a water/cement mass ratio of <0.65 presented compressive strengths higher than 30 MPa after 90 days, and pH values lower than 8.5, corresponding to the passivation zone of aluminum corrosion.

Non-destructive verification of materials in waste packages using QUANTOM®

The nuclear and non-nuclear industry has produced a considerable amount of low and intermediate-level radioactive wastes during the last decades. The material characterization of waste packages recently became more and more important in order to dispose of these waste packages in a final underground repository. Material characterization remains an indispensable criterion to prevent pollution of the groundwater with toxic materials and is usually required by the national licensing and supervisory authorities. Information on the nature of waste materials can be obtained based on existing documentation or, if the documentation is insufficient, on further destructive or non-destructive analysis. Non-destructive methods are to be preferred to minimize radiation exposures of operating personnel as well as costs. Existing non-destructive techniques (Gamma scanning, X-ray, active/passive neutron counting, muon tomography) do not allow the identification of non-radioactive hazardous substances. An innovative non-destructive measurement system called QUANTOM® (QUantitative ANalysis of TOxic and non-toxic Materials) has been developed. It is based on the prompt and delayed gamma neutron activation analysis (P&DGNAA). This technology is able to identify and quantify the elemental composition (Cd, Cu, B, Pb, Hg, Fe, Al, …) in radioactive packages such as 200-l radioactive drums. This information helps waste producers verify the content of their radioactive wastes, especially regarding the presence of hazardous substances. Different reference materials have been analysed by means of the same technology (P&DGNAA) at the research reactor of BUDAPEST. A comparison of those results for five reference materials is presented. The results show a very good agreement between QUANTOM® and standardized reference analyses.

New Method for the Solidification of High-Concentration Radioactive Borate Solution by Cement-Based Materials

Cement is widely used for the solidification of low- and intermediate-level radioactive waste materials. Radioactive borate solution with a high concentration of boron is one of the main radioactive wastes produced in nuclear stations. It is difficult to solidify this solution by using cement because borate has a great inhibitory effect on the cement hydration process. In this study, the hydration kinetics, strength, durability, phase assemblage, and transportation and transformation of the silicon of the paste that blended Portland cement with 5 M borate solution were investigated. After the addition of sodium hydroxide and sodium metasilicate to the paste, the cement hydration process was restarted, and the 28-days strength of samples met the requirements of the Chinese standard. The mechanism of overcoming the retardation of cement hydration by the borate solution was attributed to the formation of calcium metaborate, ettringite, portlandite, and calcium silicate hydrate with the restarting of cement hydration, without the formation of ulexite.

Cellulose degradation products and high ionic strength enhance the retardation of strontium in hydrated cement paste

The present concept for the disposal of some low- and all intermediate-level radioactive waste in the United Kingdom involves grouting with cement in steel drums and placement in a geological facility. The vaults would then be backfilled with low strength, high porosity cement-based materials designed to promote a pervasive, alkaline environment in which many of the radioactive species present are sparingly soluble. This work investigates the interaction of strontium with such a backfill under both diffusive and advective conditions given the potential significance of the fission product 90Sr. An important characteristic of the United Kingdom waste inventory is the abundance of organic compounds, including cellulosic materials; consequently, the experiments were repeated with products of alkaline cellulose degradation. Additional experiments were performed at high ionic strength simulating anticipated changes in the salinity of the groundwater. The effective diffusivity (De) of strontium in the absence of the organic compounds ranged between 5.5 × 10−11 and 8.5 × 10−11 m2 s−1, with a distribution coefficient (Rd) of between 2.8 × 10−3 and 3.1 × 10−3 m3 kg−1. The presence of organic compounds and/or an increase of ionic strength enhanced the retardation of strontium. The results indicate that the retention of strontium by hydrated cement paste occurs via a reversible ion exchange mechanism rather than mineralisation and is promoted by decalcification of the calcium silicate hydrate phases.

Monitoring in a reinforced concrete structure for storing low and intermediate level radioactive waste. Lessons learnt after 25 years

Where concrete structures are designed to have a service life of over 100 years, their performance must be monitored, for the prediction models available are fraught with uncertainties that need to be eliminated. The present study was conducted to meet that need by monitoring a pilot structure for low and intermediate radioactive waste storage. Long-term operation of the sensors was observed to be adequate to determine the value of the parameters that characterise structural durability, such as corrosion current density. The parameters analysed were correlated to calculate their reciprocal impact: where applied in conjunction with artificial intelligence tools, temperature, for instance, was found suitable for finding activation energy and expansion coefficients and detecting outliers. The results showed the pilot structure to perform satisfactorily.

Characterization of bentonite, vermiculite and soil as potential materials to compose the natural barriers in a near-surface repository

The use of nuclear technology is growing in Brazil, where it is possible to carry out the entire nuclear fuel cycle. This technology generates radioactive waste that must be treated before being released into the environment or stored in suitable facilities. Then, the first national repository - CENTENA - will be implemented, where all low and intermediate level radioactive waste generated in Brazilian territory will be safely stored. One of the barriers that will compose the repository will be composed of natural materials such as soils and clays, the latter for having a high adsorption and cation exchange capacities and swelling potential. The purpose of the article is to show the research carried out in Centro de Desenvolvimento da Tecnologia Nuclear (CDTN) with natural materials aiming their use in natural barriers. Four materials were characterized: CDTN soil, this soil with bentonite, expanded vermiculite and bentonite, using the clay characterization protocol. Primary mineralogical, physical and chemical analyzes were performed, such as X-ray diffraction, scanning electron microscopy, particle size distribution, specific surface area and cation exchange capacity. The results showed that these clays are potentially candidates for composing CENTENA's natural barriers.

In Situ Synthesis of Zeolites in Solidified Fly Ash/Cement Matrices with High Na Content: Products, Pore Structure, and Leaching Behavior

Due to its high retention capacity and well durability, fly ash-based geopolymers rich of zeolites are considered to be potential solidified materials for low- and intermediate-level radioactive waste stream. This study mainly focused on the phase transformation, pore structure, and leaching characteristic of Sr2+ and Cs+ in solidified fly ash/cement matrices under different sodium hydroxide contents or sodium nitrate concentrations after thermal curing of 28 d. The results obtained from the experiments indicated that the geopolymers prepared from pure fly ash had the potential to transform into ordered structures at high temperatures of 90°C instead of room temperature, such as Na-P1 zeolite and chabazite. With the increase in sodium hydroxide contents, more sodium nitrate participated in the reaction to form a zeolite phase in solidified fly ash/cement matrices, while the coexistence phenomenon of chabazite and Na-P1 zeolite appeared under the condition of high sodium hydroxide contents of more than 5%. Just a proper number of zeolites can effectively inhibit the leaching of radionuclides Sr2+, as too many zeolites would lead to the increase in porosity, which was harmful to the retention of radionuclides. Solidified fly ash/cement matrices at a sodium hydroxide content of 5% or at a sodium nitrate concentration of 300 g/L obtained the lowest cumulative leaching fraction of Sr2+. Moreover, sodium hydroxide increased the cumulative leaching fraction of Cs+ which reversely decreased significantly with the increase in sodium nitrate concentrations.

PH and microbial community determine the denitrifying activity in the presence of nitrate-containing radioactive waste.

An important fraction of the currently stored volume of long-lived intermediate-level radioactive waste in Belgium contains large amounts of NaNO3 homogeneously dispersed in a hard bituminous matrix. Geological disposal of this waste form in a water-saturated sedimentary formation such as Boom Clay will result in the leaching of high concentrations of NaNO3, which could cause a geochemical perturbation of the surrounding clay, possibly affecting some of the favorable characteristics of the host formation. In addition, hyper-alkaline conditions are expected for thousands of years, imposed by the cementitious materials used as backfill material. Microbial nitrate reduction is a well-known process and can result in the accumulation of nitrite or nitrogenous gases. This could lead to the oxidation of redox-active Boom Clay components, which could (locally) decrease the reducing capacity of the clay formation. Here, we compared nitrate reduction processes between two microbial communities at different pH related to a geological repository environment and in the presence of a nitrate-containing waste simulate during 1 year in batch experiments. We showed that the microbial community from in Boom Clay borehole water was able to carry out nitrate reduction in the presence of acetate at pH 10.5, although the maximum rate of 1.3 ± 0.2 mM NO3 -/day was much lower compared to that observed at pH 9 (2.9 mM NO3 -/day). However, microbial activity at pH 10.5 was likely limited by a phosphate shortage. This study further confirmed that the Harpur Hill sediment harbors a microbial community adapted to high pH conditions. It reduced twice as much nitrate at pH 10.5 compared to pH 9 and the maximum nitrate reduction rate was higher at pH 10.5 compared to that at pH 9, i.e., 3.4 ± 0.8 mM NO3 -/day versus 2.2 ± 0.4 mM NO3 -/day. Both communities were able to form biofilms on non-radioactive Eurobitum. However, for both microbial communities, pH 12.5 seems to be a limiting condition for microbial activity as no nitrate reduction nor biofilm was observed. Nevertheless, pH alone is not sufficient to eliminate microbial presence, but it can induce a significant shift in the microbial community composition and reduce its nitrate reducing activity. Furthermore, at the interface between the cementitious disposal gallery and the clay host rock, the pH will not be sufficiently high to inhibit microbial nitrate reduction.

Cs Extraction from Chloride Media by Calixarene Crown-Ethers

Asse II salt mine, in Germany, contains low and intermediate-level radioactive waste that must be retrieved in the upcoming years. Potentially contaminated salts and brines will require treatment, with 137Cs being the main contaminant. Cs+ is problematic to selectively recover due to its chemical similarity with Na+ and K+ which are present in high quantities in a salt mine. This paper offers a novel solution for Cs+ separation from concentrated chloride salt media by solvent extraction with calixarene-crown-ether extractants in an alcoholic diluent. The proposed solvent extracts Cs+ at elevated chloride concentrations (3–4 M) while back-extraction is achieved by contacting the solvent with dilute (0.01 M) hydrochloric acid.