Long-lived Radioactive Waste Research Articles

Experimental methods for determining the effective neutron multiplication factor of the “Yalina-Thermal” subcritical assembly

To study the kinetics of subcritical systems and determine the optimal conditions for the transmutation of longlived radioactive waste in the neutron spectrum of ADS-systems the “Yalina” research nuclear facility was created at Joint Institute for Power and Nuclear Research – Sosny (Minsk, Belarus). The main safety indicator of a subcritical system (active zone reactivity) was measured for a “Yalina-Thermal” assembly via three independent methods: inverse multiplication, probabilistic and impulse ones. For the inverse multiplication method, the neutron flux density was monitored during assembly loading. For a fuel load of 285 EK-10 rods the neutron multiplication was M = 22.3±0.6, and the effective neutron multiplication coefficient was keff = 0.9551± 0.0016. The probabilistic method (Feynman-alpha method), based on measuring fluctuations in the neutron density level within a system with a fission chain reaction, gave the ratio of the variance to the average counting rate value D/n = 1.779±0.005, which corresponds to keff = 0.9597 ±0.0003. The pulse method is aimed at studying the neutron flux behavior of after the neutron pulse injection into the breeding system. Measurements were held with the same setup, used in the Feynman-alpha method. The measured decay constant of instantaneous neutrons is α = –670±0.7 1/s, which corresponds to keff = 0.9560±0.0001. The effective multiplication factor keff of the subcritical assembly “Yalina-Thermal”, obtained via three different independent methods, is around average value of keff = 0.9569 ± 0.0018. The methods considered can be used for subcritical level monitoring for ADS-systems and research nuclear facilities.

Tungsten–chromium–yttrium alloys as first wall armor material: Yttrium concentration, oxygen content and transmutation elements

Tungsten (W) is a prime candidate as first wall armor material of future fusion power plants as W withstands extreme particle, heat, and radiation loads without forming long-lived radioactive waste. The release of radioactive material from the reactor to the environment should be suppressed in case of an accident such as a loss of coolant (LOCA) with simultaneous air ingress into the vacuum vessel. W oxidizes and sublimates in case of a LOCA. Therefore, oxidation resistant tungsten, chromium, yttrium (W–Cr–Y) alloys are developed to provide intrinsic safety in case of such an accident. In this paper, the optimization of the yttrium (Y) concentration is presented on bulk samples compacted by field assisted sintering technology (FAST). W with 11.4 weight (wt)% Cr and 0.6 wt% Y appears to be an optimum regarding the oxidation resistance. Further, first preparations for industrial upscaling, which may increase the impurity level, are addressed. The oxygen (O) content is varied systematically. It is shown that a good oxidation resistance requires a low O level. The exposure of the material to fusion neutrons is another issue addressed on W–Cr–Y alloys. In a non-activated environment it is shown that 1 wt% rhenium (Re) dramatically changes the oxidation kinetics: at 1273 K the mass gain of W–Cr–Y–Re follows a cubic rate law while W–Cr–Y follows a linear rate law for two days. Further, the influence of the alloying elements on the neutron transport and transmutation of W is studied by simulating the exposure of spatially heterogeneous high-resolution models of the W–Cr–Y alloys to 14 MeV fusion neutrons.

Neptunium extraction by N,N-dialkylamides

Abstract Separation of neptunium by solvent extraction has been based on tributylphosphate (TBP) for decades, but TBP is not fully incinerable, which adds to the burden of long-lived radioactive waste. Alternatives to TBP for uranium and plutonium extraction, such as the N,N-diakylamides, previously have been explored in the hopes of transitioning to an extractant that is incinerable. Four N,N-diakylamides, N,N-dihexylhexanamide (DHHA), N,N-dihexyloctanamide (DHOA), N,N-di(2-ethylhexyl)butanamide (DEHBA), and N,N-di(2-ethylhexyl)-iso-butanamide (DEHiBA) were considered in this work for their potential to extract millimolar concentrations of Np(IV), Np(V), and Np(VI) from nitric acid solutions into organic solutions containing 1 M extractant in Exxsol D60. Under these conditions the branching of the alkyl substituents affects the extractability of Np(VI) and Np(IV), causing three of the dialkylamides, DHHA, DHOA and DEHBA, to extract neptunium in the expected order Np(VI) > Np(IV) > > Np(V). In contrast, branched DEHiBA is so poor an extractant for Np(IV) that the extraction order becomes Np(VI) > > Np(V) > Np(IV) between 0.1 and 5.6 M HNO3 due to partial oxidation of the Np(V) in nitric acid.

Scientific and technical problems of CNFC in two-component NPE and their solution in “Proryv” project

Modern Nuclear Energetics (NE) based on Thermal Reactors (TR) with Uranium fuel in Open Nuclear Fuel Cycle (ONFC) has systemic problems that limit its further development: low utilization efficiency of extracted U, absence of ecologically suitable solution for long-lived high-level radioactive waste treatment and nonproliferation. Besides that one of the most serious barriers for modern NE development is the problem of competitiveness that is closely related to safety problem. Attempts to solve the safety problem by development of additional active means of safety protection led to the decrease of competitive ability of NE compared to organic power industry. Fast Reactors with inherent safety as the basis for the New Technological Platform (NTP) are to overcome the current development barriers. The near future transfer towards the two-component NE structure with Fast and Thermal Reactors and CNFC is the key direction of the nuclear energy development strategy. Reprocessing of spent fuel (SF) and recycling of accumulated Pu and unburned U in Fast Reactors fuels cycle allows to cut the need in natural U in 100 times and in 10 times the mass of heavy nuclei in long-lived high-level radioactive waste, which is one of the key means of ensuring ecological safety of NTP. This article presents the goals and means of achieving technological and ecological safety, political neutrality, resource stability and competitive ability of the New Technological Platform. Introduction

Water-soluble bitumen degradation products can fuel nitrate reduction from non-radioactive bituminized waste

An important fraction of the currently stored volume of long-lived intermediate level radioactive waste in Belgium is immobilized as Eurobitum. This type of waste typically contains large amounts of NaNO3 homogeneously dispersed in a hard bituminous matrix. Geological disposal of Eurobitum in a water-saturated sedimentary formation such as Boom Clay will result in the leaching of high concentrations of NaNO3 in the Boom Clay formation. This could cause a geochemical perturbation of the surrounding clay, possibly affecting some of the favorable characteristics of the host formation such as its hydraulic conductivity, sorption potential or the redox conditions. The latter might result in a decrease of its reducing capacity and an increase in the mobility of redox-sensitive radionuclides. Microbial nitrate reduction is a well-known process, which could be stimulated by the chemical and radiolytical water-soluble organic bitumen degradation products. The present study carried out different series of long-term anoxic batch experiments to investigate the ability of the microbial community of Boom Clay borehole water to reduce nitrate, leaching from thermally aged non-radioactive Eurobitum in the presence or absence of acetate, formate and oxalate, being part of the most important organic bitumen degradation products. Obtained results indicate that all three organic bitumen degradation products can be used as electron donor to fuel microbial nitrate reduction, albeit with a different efficiency. The highest nitrate reduction rate was observed in the presence of acetate, while oxalate was the least efficient electron donor for nitrate reduction. Moreover, organic compounds that leached from the Eurobitum during the course of the experiment were used as electron donor for microbial nitrate reduction in all conditions. Furthermore, calcium oxalate crystals were observed, indicating that if oxalate is present, it will probably be less bioavailable compared to other organic compounds.

Сравнительная оценка безопасности глубинных захоронений радиоактивных отходов открытого и замкнутого топливного циклов: радиологическая миграционная эквивалентность

The article presents analysis of safety of long-lived high level radioactive waste (RW) disposed for permanent storage in a deep geological repository. According to the study protocol it is necessary to estimate the human and environmental health safety of RW, generated in different nuclear fuel cy-cles, during their long-term stay in the repository. To estimate biological effects, RW composition from thermal reactors in open nuclear fuel cycle (“TR”) and from simultaneously operated thermal and fast reactors in closed nuclear cycle (“TR+FR”) in different time after their disposal have been evaluated. It is suggested that RW to be safe to humans and the environment if committed effective doses from RW components including minor actinides and fission products and from natural urani-um are equivalent (radiation equivalence) or lifetime attributable risks of cancer mortality from RW and natural uranium are equivalent (radiological equivalence). For this purpose it is important to evaluate the time of achievement of radiation or radiological equivalence. To evaluate health effect of RW containing radio-nuclides on the public and the environment their migration activity has been evaluated with the use of distribution coefficient and retardation factor. Probable annual effective doses to the public from exposure to the radionuclides released from the repository to the biosphere at different time after the RW disposal have been estimated. Radiation doses to the public following the consumption of water from the well installed on the repository, and lifetime attributable risk (LAR) of cancer mortality associated with the exposure to radionuclides contained in “TR” and ”TR+FR” generated waste have been evaluated. It turned out that LARs of cancer mortality from “TR+FR” radionuclides is lower than LAR from natural uranium, it means that radiological migration equivalence after its achievement will exists until the repository conditions allow, about 1 billion years. We have found that LARs from “TR+FR” waste is about 100 times lower than the negligible level, 10-6, established in the Russian National Radiation Safety Standards (NRB-99/2009). Estimat-ed LAR of cancer mortality from “TR”-generated waste is 360 times higher than LAR from “TR+FR”-generated waste. As evaluated for the waste from TR radiological equivalence will be achieved not earlier than 1 million years after permanent disposal. If RW is generated in the closed cycle estimat-ed LAR value from consumption of well water does not exceed 10-5 year-1, if RW is generated in open fuel cycle, the LAR value is 10-3 year-1, this is socially unacceptable risk. In the first 10 thousand years after the disposal the major dose- and risk-forming radionuclides in the well water will be 241Am, 239Pu и 240Pu. Estimates of radiation detriment from ”TR” generated waste put to the perma-nent deep geological disposal are given in the article.

Americium Transmutation in the SVBR-100 Reactor

One of the postponed problems of nuclear power (NP) is the problem of the management of long-lived radioactive waste (RAW), and, first of all, with minor actinides (MA), of which americium-241 is the most difficult. The aim of this work is to study the efficiency of americium transmutation in a fast reactor with a heavy liquid metal coolant lead-bismuth eutectic alloy. The article presents the results of calculations of the transmutation of americium in the SVBR-100 reactor using standard uranium oxide fuel with the addition of americium-241. The obtained values of the rate of transmutation of americium are compared with similar values for the SVBR-100 reactors on MOX-fuel and in the BN-800 reactor.

Capabilities of JSC “SSC RIAR” in Producing 227ThCl4

227Th is considered to be one of the promising radionuclides for target radionuclide therapy of cancer. Its advantages are chemical properties which are good for the synthesis of the complexes with chelating agents and emission of five alpha particles with complete decay of one 227Th nucleus. The apparent disadvantage is the presence of 223Ra among daughter products the chemical properties of which are fundamentally different from thorium that can cause its redistribution in the human body. A rather big half-life of 227Th (18.7 days), on the one hand, is good for the preparation production and shipment, and on the other hand, it specifies high-level requirements to the preparation stability in the human body. In addition to using 227Th for radiopharmaceutical synthesis, it can be applied in manufacturing 223Ra generators. While such generators have a rather small period of use, they have a number of advantages over 227Ac/223Ra generators. In particular, there is no need to check every batch for the 227Ac content, and there are no long-lived radioactive wastes. JSC “SSC RIAR” performs experiments to produce trial samples of 227Th by generating from long-lived parent radionuclide 227Ac extracted from the 226Ra targets irradiated during 20-25 days in the SM reactor neutron trap. To separate 227Th and 227, thorium is sorbed on BioRad AG 18 (NO3-) strongly basic anion-exchange resin from 8M HNO3 with further elution using 0.1-0.5 HNO3 or HCl. To produce the preparation of desired radionuclide purity, the purification process must be carried out at least twice. The key challenge in producing 227Th is determination of the 227Ac content. It is not possible to directly determine 227Ac by its own alpha radiation at its activity of less than 1% of the 227Th activity. The yield of alpha radiation in 227Ac decay is only 1.38 %, and its peak in the spectrum is on the low-energy tail of 227Th and its daughter products. 227Th conversion electrons and beta radiation of the daughter products, namely 211Pb and 211Bi, are obstacles in measuring the 227Ac activity by beta radiation. The presentation discusses possible methods to check the content of 227Ac with its preliminary chemical extraction from a preparation aliquot and gives the characteristics of 227Th experimental samples.

High Temperature and Ion Implantation-Induced Phase Transformations in Novel Reduced Activation Si-Fe-V-Cr (-Mo) High Entropy Alloys

For fusion to be realized as a safe, sustainable source of power, new structural materials need to be developed which can withstand high temperatures and the unique fusion radiation environment. An attractive aspect of fusion is that no long-lived radioactive wastes will be produced, but to achieve this structural materials must comprise reduced activation elements. Compositionally complex alloys (CCAs) (also called high entropy alloys, HEAs) are promising candidates for use in extreme environments, including fusion, but few reported to date have low activation. To address these material challenges, we have produced novel, reduced activation, HEAs by arc-melting, and investigated their thermal stability, and radiation damage resistance using 5 MeV Au2+ ion implantation. Whilst the alloys were designed to form single phase BCC, using room temperature and non-ambient in situ X-ray diffraction we have revealed the thermodynamically stable structure of these alloys is in fact a sigma phase. We propose that a BCC phase is formed in these alloys, but at high temperatures (>1000°C). A BCC phase was also formed during heavy ion implantation, which we propose to be due to the rapid heating and cooling that occurs during the thermal spike, effectively freezing in the BCC phase produced by an implantation induced phase transformation. The BCC phase was found to have high hardness and a degree of ductility, making these new alloys attractive in the development of reduced activation HEAs for nuclear applications.

How rock samples can be representative of in situ condition: A case study of Callovo-Oxfordian claystones

Within the framework of feasibility studies for a reversible, deep geological repository of high- and intermediate-level long-lived radioactive waste (HLW, IL-LLW), the French National Radioactive Waste Management Agency (Andra) is investigating the Callovo-Oxfordian (COx) formation near Bure (northeast part of France) as a potential host rock for the repository. The hydro-mechanical (HM) behaviour is an important issue to design and optimise different components of the disposal such as shaft, ramp, drift, and waste package disposal facilities. Over the past 20 years, a large number of laboratory experiments have been carried out to characterise and understand the HM behaviours of COx claystones. At the beginning, samples came from deep boreholes drilled at the ground surface with oil base mud. From 2000 onwards, with the launch of the construction of the Meuse/Haute-Marne Underground Research Laboratory (MHM URL), most samples have been extracted from a large number of air drilled boreholes in the URL. In parallel, various constitutive models have been developed for modelling. The thermo-hydro-mechanical (THM) behaviours of the COx claystones were investigated under different repository conditions. Core samples are subjected to a complex HM loading path before testing, due to drilling, conditioning and preparation. Various kinds of effects on the characteristics of the claystones are highlighted and discussed, and the procedures for core extraction and packaging as well as a systematic sample preparation protocol are proposed in order to minimise the uncertainties on test results. The representativeness of the test results is also addressed with regard to the in situ rock mass.

Sorption of radium onto early cretaceous clays (Gault and Plicatules Fm). Implications for a repository of low-level, long-lived radioactive waste

The repository of low-level and long-lived radioactive waste from REE purification and treatment is presently under investigation by the French National Radioactive Waste Management Agency (Andra). This waste consists of Ra-bearing solids, mainly Ra-barite (Ba,Ra)(SO4), along with other salts such as NH4NO3 and monazite. The current repository concept under study is based on a shallow (∼30 m deep) geological disposal in a clay formation. This work deals with the experimental study of the sorption of Ra onto two clay rocks from Lower Cretaceous geological formations in the Department of Aube, in France, the Gault (tégulines) Formation and Plicatules Formation.Clay samples were extracted from different boreholes from the both geological clay formations, and consisted mainly of illite, illite-smectite mixed layers (35–65% and 25–70%). Sorption experiments were carried out on raw materials as well as on samples exchanged with different cations (homoionic forms of Na, NH4, Ba and Ca, respectively). The results indicate that Ra sorption in raw materials is linear within the range of concentration investigated (10−8 - 10−11 M). In most of the samples, Kd values from 170 to 300 g·mL−1 were measured for Ra and from 30 to 70 g·mL−1 for Ba. The sorption experiments (from 1 h to 92 days) of Ra showed small kinetic effects as Kd values tend to increase with time from approximately logKd = 2.15 to 2.65. In all exchanged samples, a dependence of sorption on the ionic strength was observed -the higher the ionic strength, the lower Kd values-; except for Na, this dependence is consistent with an ionic exchange process and selectivity coefficients (Ra-Ca; Ra-Ba and Ba-Ca) have been determined. A non-negligible effect of pH on sorption was also observed that could be explained considering surface complexation on the amphoteric groups present at the solid surface. In these materials, the selectivity of Ra was higher than that of Ba (half an order of magnitude).The experimental results indicate that sorption of Ra in the studied clay formations is high enough to limit the radium migration in the near field of a low-level, long lived disposal site.

France's State of the Art Distributed Optical Fibre Sensors Qualified for the Monitoring of the French Underground Repository for High Level and Intermediate Level Long Lived Radioactive Wastes.

This paper presents the state of the art distributed sensing systems, based on optical fibres, developed and qualified for the French Cigéo project, the underground repository for high level and intermediate level long-lived radioactive wastes. Four main parameters, namely strain, temperature, radiation and hydrogen concentration are currently investigated by optical fibre sensors, as well as the tolerances of selected technologies to the unique constraints of the Cigéo’s severe environment. Using fluorine-doped silica optical fibre surrounded by a carbon layer and polyimide coating, it is possible to exploit its Raman, Brillouin and Rayleigh scattering signatures to achieve the distributed sensing of the temperature and the strain inside the repository cells of radioactive wastes. Regarding the dose measurement, promising solutions are proposed based on Radiation Induced Attenuation (RIA) responses of sensitive fibres such as the P-doped ones. While for hydrogen measurements, the potential of specialty optical fibres with Pd particles embedded in their silica matrix is currently studied for this gas monitoring through its impact on the fibre Brillouin signature evolution.

Vertical distribution of helium and 40 Ar/ 36 Ar in porewaters of the Eastern Paris Basin (Bure/Haute-Marne): constraints on transport processes through the sedimentary sequence

Abstract As part of its ongoing project on repositories for high-activity, long-lived radioactive waste, a 2000 m deep borehole was drilled by the French Nuclear Waste Agency (ANDRA) in the layered structure of alternating aquifers and aquitards of the Eastern Paris Basin. Among the information retrieved from this borehole, the vertical distribution of chloride in porewaters showed that, in addition to vertical diffusion, lateral advection in the aquifers plays a major part in transporting chlorine away from the study area. Helium concentrations were also measured in porewaters along the borehole. Because the helium input function is different from that of chlorine, it represents an excellent alternative tracer to further constrain transport characteristics. We applied an advection–diffusion model to the helium profiles with the appropriate source term for 4 He based on U–Th measured concentrations of uranium and thorium. 40 Ar/ 36 Ar data, which were available along the whole sequence, were also simulated. The modelled and measured 4 He profiles were in good agreement, indicating that the transport parameters used for the chlorine simulations were robust. 40 Ar/ 36 Ar simulations also gave coherent results and confirmed that most of the radiogenic 40 Ar remained trapped in the rocks (primarily in clays and feldspars).

An Advanced Energy System Using a Small Fast Reactor as an Energy Source

An advanced energy system has been proposed that involves a supercritical carbon dioxide gas turbine fast reactor (S-CO2 FR) as a dispersed energy source, a new waste-heat recovery system from the FR, and a bioconversion system using the recovered waste heat. The FR with S-CO2 gas turbine achieves higher cycle efficiency than conventional sodium-cooled FRs with steam turbines, eliminating problems of conventional FRs related to safety, plant maintenance, and construction costs. The S-CO2 FR consumes minor actinide elements produced in light water reactors as fuel, thereby reducing long-lived radioactive wastes and environmental loads imparted by long-term geological disposal. The recovered waste heat is used for methane and methanol production through fermentation of human and animal wastes in cities and farms. The methane and methanol are easily transported and can accommodate demand changes; they are useful as fuel of fuel cells, automobiles, and gas turbine power plants. The total energy utilization efficiency in electricity and heat is estimated to be higher than 85%, contributing to saving of energy resources and reduction of greenhouse-gas emissions. Consumption of the waste products of cities and farms for production of methane and methanol fosters a recycling society. Compact and high-performance microchannel heat exchangers are used in the S-CO2 FR, the new waste-heat recovery system and the fermentation system.

Feasibility study of fissile mass quantification by photofission delayed gamma rays in radioactive waste packages using MCNPX

The feasibility of fissile mass quantification in large, long-lived medium activity radioactive waste packages using photofission delayed gamma rays has been assessed with MCNPX. The detection limit achievable is lower than the expected uranium mass in these waste packages, but the important sensibility to the waste matrix density and sample localization imposes to get an accurate measurement of these parameters. An isotope discrimination method based on gamma-ray ratios has been evaluated showing that photofission delayed gamma rays can be used to measure the fissile mass as well as the total uranium mass.

Electrochemical Measurement of Li2O in Molten LiCl Salt

Nuclear fuel reprocessing is an important technology designed to increase the sustainability of nuclear power generation, as well as minimize long-lived radioactive waste production. There are multiple reprocessing schemes, most notably PUREX and pyroprocessing. While PUREX and related aqueous separation processes technically have superior maturity due to large scale implementation in France, Japan, Great Britain, and Russia, pyroprocessing has attracted substantial interest in recent years largely based on the perception that it represents a much lower proliferation threat and is ideally suited for metallic fast reactor fuel recycle. The Republic of Korea in particular has invested a tremendous amount of research funding over the last decade to develop pyroprocessing for spent fuel from commercial pressurized water reactors. Pyroprocessing utilizes a high temperature electrochemical cell called an electrorefiner to separate actinides from fission products. To apply this technology to oxide fuel, as in the case of the Korean spent fuel, an oxide reduction step must precede electrorefining. The current leading technology for oxide reduction is a high temperature electrolytic process carried out in molten lithium chloride salt. Electrolytic reduction relies on a lithium cycle, which consists of oxidation of lithium metal to lithium oxide at the cathode and reduction of the lithium oxide back to lithium metal at the anode. The net result is that uranium oxide is reduced to uranium metal, and oxygen gas is evolved. Lithium oxide concentration in the molten salt electrolyte is an important process variable. If it decreases, it can be a sign of lithium oxide accumulation in the fuel basket and will result in increased anode potential that ultimately leads to anode degradation. Currently, the only proven way of measuring lithium oxide concentration in lithium chloride involves sampling followed by acid-base titration. Measuring lithium oxide concentration in real time is an important objective to maintain optimal process control. The technique being developed for this purpose involves measuring the change in the open circuit potential (OCP) between a molten pool of a lead-lithium alloy and a Ni/NiO reference electrode. Because OCP measures equilibrium potential, this method can be combined with the Nernst Equation to continuously measure lithium oxide concentration. Mixtures of lithium chloride and lithium oxide were prepared with lithium oxide concentrations ranging from 0.00 to 2.75 wt%. A lead-lithium pool containing 2.0 wt% lithium was placed below the molten salt pool. After the salt melted it was allowed an hour to equilibrate, then an OCP measurement was obtained. As expected, the OCP increased with increasing concentration of Li2O. The measured sensitivity was about 200 mV for a Li2O concentration change from 0 to 2.5 wt%. Achieving repeatable and stable reference electrode potential is key to making accurate measurements from this method. The reference electrodes were made of high-density MgO tubes with low density, porous MgO plugs. Nickel wire and NiO powder were inserted into these tubes. The reference potential varied significantly between different sets of reference electrodes, which is hypothesized to be due to variations in the thickness of the porous MgO plugs. Methods for standardizing the reference electrodes have been attempted, and the results will be discussed. Another factor that may effect this measurement is the solubility of lithium metal in the zero oxidation state in the molten LiCl. Lithium solubility measurements were made in LiCl-Li2O with varying concentrations of Li2O. The results varied from 0.7 to 1.5 mole%. This low level of solubility is expected to cause minimal loss of lithium from the lead-lithium sensor. And experiments with different volumes of salt over a fixed volume of lead-lithium confirmed that the measured OCP is relatively insensitive to any decrease in the lithium concentration in the lithium-lead pool.

Microbial sulphate-reducing activity over load pressure and density in water saturated Boom Clay

The Boom Clay formation in Mol, Belgium, is studied as a reference host rock for the future Belgian repository for high-level and long-lived radioactive wastes. An apparently dormant sulphate-reducing bacteria (SRB) population in Boom Clay can be activated during repository construction and reduce sulphate to sulphide which may enhance the corrosion of metallic components of the engineered barriers. Thirteen test cells constructed of titanium were installed with saturated Boom Clay cores at three different wet densities, 1800, 1900 and 2000kgm−3. For the purpose of analysing microbial sulphate-reduction to sulphide, a previously developed method utilizing the radiotracer 35SO42− was applied. Copper discs were installed towards which produced sulphide diffused and reacted to form CuxS. The amounts of radioactive sulphide on the copper disc surfaces were analysed and the sulphide production rates in the clay were modelled using a diffusion coefficient for sulphate that was determined to 2.2×10−12m2s−1 for a fully saturated Boom Clay at a wet density of 2000kgm−3. The diffusion coefficient for sulphide was set to 4.4×10−12m2s−1. Heat treated clay (120°C, 48h) was included as negative controls in 4 test cells. The analysis for SRB reported from 107 up to 109cellsL−1 pore water for 34 sample positions and 5 positions in the negative control clay cores were below detection limit. These numbers were within the range of cultivable bacteria observed previously in Boom Clay. There was no clear cut-off in density with respect to presence of cultivable SRB and sulphide production, but it appeared as if sulphide production was increasingly possible at or below 1800kgm−3. At higher densities, numbers were lower, but the SRB were still cultivable and active which suggests that SRB can be active and produce sulphide in a Boom Clay repository for as long as sulphate is available.

The Status of Water in Swelling Shales: An Insight from the Water Retention Properties of the Callovo-Oxfordian Claystone

The Callovo-Oxfordian (COx) claystone is considered in France as a possible host rock for the disposal of high-level long-lived radioactive waste at great depth. During the operational phase, the walls of the galleries and of the disposal cells will be successively subjected to desaturation induced by ventilation followed by resaturation once the galleries are closed. To better understand this phenomenon, a sound understanding of the water retention properties of the COx claystone is necessary. Following a previous study by the same group, this paper presents an investigation of microstructure changes in COx claystone under suction changes. Microstructure was investigated by means of mercury intrusion porosimetry tests on freeze-dried specimens previously submitted to various suctions. Along the drying path, the initial microstructure, characterised by a well-classified unimodal pore population around a mean diameter value of 32 nm, slightly changed with the same shape of the PSD curve and slightly moved towards smaller diameters (27–28 nm) at suctions of 150 and 331 MPa, respectively. The infra-porosity too small to be intruded by mercury (diameter smaller than 5.5 nm) reduced from 4.3 to 3.3 %. Oven drying reduced the mean diameter to 20 nm and the infra-porosity to 1 %. Wetting up to 9 MPa suction leads to saturation with no significant change in the PSD curve, whereas wetting at zero suction gave rise to the appearance of a large pore population resulting from the development of cracks with width of several micrometres, together with an enlargement of the initial pore population above the mean diameter. The concepts describing the step hydration of smectites (by the successive placement within the clay platelets along the smectite faces of 1, 2, 3 and 4 layers of water molecules with respect to the suction applied) appeared relevant to better understand the changes in microstructure of the COx claystone under suction changes. This also allowed to better define the status of water in claystones and shales containing smectite, with a distinction made between the water adsorbed within the clay platelets, and the free inter-platelet water involved in hydromechanical couplings through changes in pore pressure and water transfers.

Fundamental aspects of the hydromechanical behaviour of Callovo-Oxfordian claystone: From experimental studies to model calibration and validation

Within the framework of feasibility studies for a reversible, deep geological repository of high- and intermediate-level long-lived radioactive waste (HLW, IL-LLW), the French National Radioactive Waste Management Agency (Andra) is investigating the Callovo-Oxfordian (COx) claystone formation near Bure (northeast France) as a potential host formation for the repository. In 2000, Andra initiated construction of the Meuse Haute-Marne Underground Research Laboratory (MHM URL). Concurrently with on-site construction work, an extensive research program has been conducted, including theoretical and experimental studies and constitutive modelling. In-situ experiments provide an extensive data set for performance evaluation of models describing the thermo-hydro-mechanical behaviour of COx claystone under different conditions (excavation, heat, etc.). In 2012, a benchmark exercise was initiated to evaluate proposed models, including basic assumptions, mathematical descriptions, input variables and parameters, and to determine how these assumptions influence model outcomes with respect to in-situ experimental observations. The present paper outlines the main features of the hydromechanical behaviour of COx claystone and presents a series of laboratory test results (triaxial compression and creep tests) providing model development teams with a coherent database for model calibration and validation.

Overview of the clay-mineralogy studies presented at the ‘Clays in natural and engineered barriers for radioactive waste confinement’ meeting, Brussels, March 2015

This special issue of Clay Minerals contains 13 full papers presented at the 6th conference on ‘Clays in natural and engineered barriers for radioactive waste confinement’ held in Brussels during 2015. The conference was organized by the Belgian Agency for Radioactive Waste and Enriched Fissile Materials (ONDRAF/NIRAS) together with partner organizations ANDRA (France), COVRA (The Netherlands), NAGRA (Switzerland), NWMO (Canada), POSIVA (Finland) and SKB (Sweden) and a broad international scientific committee. Since 2002, this conference has developed into the most important event for scientists from the world over, dealing with the disposal of highly and long-lived radioactive waste. After each of these conferences, a special issue of a journal was published which focused on clay radioactive waste (radwaste), research which has contributed in a significant way to the outstanding scientific level of the research in this field ( e.g . Landais & Aranyossy, 2011; Landais et al ., 2013; Norris et al ., 2014). Therefore, Clay Minerals - Journal of Fine Particle Science , is happy to be able to provide the present compilation of recent HLW-disposal research in clay mineralogy in an open access issue. Most of the 13 papers published in this special 2016 issue were taken from the session on alteration processes. Papers from this conference which focus on large-scale geological characterization, general strategies for clay-based disposal systems, geomechanics, or mass and gas transfer will be published in a ‘Special Publication’ of the Geological Society of London. Many countries have chosen to dispose of all or some of their radioactive waste in facilities constructed in stable geological formations. Geological disposal as a safe solution for the long-term management of radioactive waste is in line with international recommendations and practices. The development of a geological disposal facility at a specific site requires a systematic and integrated approach taking …