The immobilization of soluble Cs from spent fuel elements by ion exchange and direct chemical reaction with clay minerals or shales was investigated under hydrothermal conditions. Various clay minerals or shales were reacted with likely Cs sources and water at 300 bars pressure and 100°, 200°, and 300°C for 4, 2, and 1 months, respectively. Pollucite was the principal product, but CsAlSiO4 was also observed, along with unreacted or hydrothermally altered aluminosilicates. From Cs concentrations of the product solutions partition of Cs between liquid and solids was found to vary depending on the Cs source, the clay or shale phase, temperature, and run duration. For example, illite-Cs2MoO4 interactions resulted in 19, 32, and 95% fixation of added Cs at 100°, 200°, and 300°C respectively. Fixation of as much as 97% of the Cs in some solids was observed. In addition to Cs-aluminosilicates, Cs was fixed on cation-exchange sites by interlayer collapse in montmorillonite. Reactions with Cs2MoO4 also produced powellite because Ca was available in the reaction mixture. The U6+ from β-Cs2U2O7 was reduced to form uraninite by sulfide- and/or organic-rich shales. (Cs,Na)2(UO2)(Si2O5)3·4H2O, an analog of weeksite, was produced in reactions with β-Cs2U2O7. The reaction products pollucite and uraninite can immobilize much of the Cs and U from spent fuel elements because Cs in pollucite is extremely difficult to exchange and U in uraninite is insoluble.

Reactions of Sr, La, and Nd (the latter two elements simulating Am and Cm) with potential backfill minerals for radioactive waste storage and with repository wall rocks, such as shale, were investigated under simulated repository conditions of 200° and 300°C for 12 weeks under a confining pressure of 30 MPa. The solid and solution reaction products were characterized to determine the nature and extent of reaction. Chlorite, illite, kaolinite, montmorillonite, mordenite, and clinoptilolite and four shales removed as much as 61.2 and 98.5% of the added SrCl2 and Sr(OH)2 from solution, respectively, by ion exchange and/or by forming new strontium compounds such as SrAl2Si2O8, Sr2MgSi2O7, SrCO3 (strontianite), and SrAl2Si4O12·2H2O (Sr-wairakite). The formation of these sparingly soluble Sr phases by the reaction of the soluble Sr compounds with such backfill materials indicates that the backfill may serve as a barrier during the thermal period of the waste in the life of a repository. These same minerals and shales removed as much as 99.99% of the added La or Nd from solution at 300°C by forming new phases such as LaOHCO3, NdOHCO3, and possibly La or Nd oxides and hydroxides. Zeolites reacted with La and Nd to form smectite. Thus, if La and Nd truly simulate the reactivity of Am and Cm, properly designed backfills can serve as a barrier to the migration of transuranic elements of nuclear wastes.

Secondary minerals formed during simulated weathering of nuclear waste glasses have been identified by analytical electron microscopy. A complete description of the reacted glass, from the outermost surface in direct contact with the leachant solution to the reacting front that migrates into the bulk glass, was obtained. Manganese and iron oxyhydroxide phases and saponite were found to have precipitated onto the residual glass surface from the leachant solution. Iron-bearing smectite, serpentine, and manganese and uranium-titanium oxyhydroxides formed in situ in the glass in several distinct bands at different depths beneath the original surface. This sequential development of secondary phases displays a clear trend toward more order and crystallinity in the phases farthest from the reaction front and indicates that complete restructuring of the glass into crystalline phases did not occur at the interface with fresh glass. Additionally, the formation of a discrete uranium-bearing phase, as opposed to uranium uptake by precipitated phases, suggests that stable actinide phase formation rather than ion exchange may be a possible mechanism for retarding radionuclide release to the environment.

Two industrial bentonites, IBECO SEAL-80 and TIXOTON TE, have been proposed as potential backfill material in the German Asse salt dome, a test field for the disposal of low- to medium-grade active nuclear waste. Considering the unlikely but possible case of a barrier breakdown with infiltration of a highly concentrated salt brine, the physicochemical stability and material behavior of these bentonites in a saturated salt brine (predominantly MgCl2) at 25°C were studied over the time period of 150 days. The results show that no mineral transformations occurred throughout the duration of the experiments and minor dissolution was only active during the first days. Some chemical properties, namely sorption capability and swelling, were reduced during contact with the salt brine, but could be reversed by removing the salt after treatment. Despite restriction of the CEC in the presence of salt solution, interlayer cation exchange reactions are still active in this environment. The long-term chemical stability of smectite in salt brine is predicted under these low-temperature conditions, but the increased permeability during aggregate formation could lead to physical breakdown of the backfill component.

Geological disposal is the preferred option for the final storage of high-level nuclear waste and spent nuclear fuel in most countries. The selected host rock may be different in individual national programs for radioactive-waste management and the engineered barrier systems that protect and isolate the waste may also differ, but almost all programs are considering an engineered barrier. Clay is used as a buffer that surrounds and protects the individual waste packages and/or as tunnel seal that seals off the disposal galleries from the shafts leading to the surface.

Bentonite and bentonite/sand mixtures are selected primarily because of their low hydraulic permeability in a saturated state. This ensures that diffusion will be the dominant transport mechanism in the barrier. Another key advantage is the swelling pressure, which ensures a self-sealing ability and closes gaps in the installed barrier and the excavation-damaged zone around the emplacement tunnels. Bentonite is a natural geological material that has been stable over timescales of millions of years and this is important as the barriers need to retain their properties for up to 106 y.

In order to be able to license a final repository for high-level radioactive waste, a solid understanding of how the barriers evolve with time is needed. This understanding is based on scientific knowledge about the processes and boundary conditions acting on the barriers in the repository. These are often divided into thermal, hydraulic, mechanical, and (bio)chemical processes. Examples of areas that need to be evaluated are the evolution of temperature in the repository during the early stage due to the decay heat in the waste, re-saturation of the bentonite blocks installed, build-up of swelling pressure on the containers and the surrounding rock, and degradation of the montmorillonite component in the bentonite. Another important area of development is the engineering aspects: how can the barriers be manufactured, subjected to quality control, and installed?

Geological disposal programs for radioactive waste have generated a large body of information on the safety-relevant properties of clays used as engineered barriers. The major relevant findings of the past 35 y are reviewed here.

In deep geological repositories for high-level nuclear wastes, interactions between steel canisters and clay-rich materials may lead to mineralogical transformations with a loss of the confining properties of the clays. Experiments simulating the conversion of smectite to Fe-rich clay phases in contact with Fe metal have been carried out to evaluate such a possibility by taking into account the effects of a series of critical parameters, including temperature, pH, and Fe/clay (Fe/C) and liquid/clay (L/C) ratios. The mineralogical and chemical transformations observed in these experiments have been compared with data from the literature, and subsequently used to propose a conceptual model for the main mineralogical transformations which can be expected in clay formations surrounding high-level nuclear waste repositories. In the presence of Fe metal and under low oxygen fugacity (<-40) the main mineralogical sequences are as follows:

1. (1) up to 150°C, under neutral pH, and L/C > 5: dioctahedral smectite (di-sm) → 7 Å Fe-rich phase (berthierine, odinite-cronstedtite) for large Fe/C ratios (>0.5), or di-sm → Fe-rich di-sm + Fe-rich trioctahedral smectite (tri-sm) for small Fe/C ratios (0.1)

2. (2) up to 150°C, under alkaline pH (10–12), and L/C > 5: di-sm → Fe di-sm (±palygorskite) for a small Fe/C ratio (0.1)

3. (3) at 300°C, Fe/C = 0.1, and L/C > 5: di-sm → Fe-rich saponite → trioctahedral chlorite + feldspar + zeolite (near-neutral pH); di-sm → Fe-rich vermiculite + mordenite (pH 10–12).

Low temperatures (<150°C) and large L/C and Fe/C ratios seem to favor the crystallization of the serpentine group minerals instead of Fe-rich trioctahedral smectites or chlorites, the latter being favored by higher temperatures. The role of L/C and Fe/C ratios and the competition between them at different temperatures is a crucial point in understanding the transformation of smectite in contact with Fe metal.

Hydrothermal alteration of Proterozoic granitic rocks in the Äspö underground laboratory, southestern Sweden, resulted in the formation of chlorite with large variations in textural and chemical characteristics, which reflect differences in formation temperatures, fluid composition, and reaction mechanisms. The mineral assemblage associated with chlorite, including Ca-Al silicates (prehnite, pumpellyite, epidote, and titanite), Fe-oxides, calcite, albite and K-feldspar, suggests that chloritization occurred at temperatures of between 200–350°C during various hydrothermal events primarily linked to magmatism and rock deformation. Petrographic and electron microprobe analyses revealed that chlorite replaced biotite, amphibole and magnetite, and hydrothermal chlorite phases filled fractures and vugs in the granitic rocks. While fracture-filling chlorite reduces fracture permeability, chloritization reactions in the host granite resulted in the formation of new localized microporosity that should thus be taken into consideration when evaluating the safety of the granitic basement rocks as a repository for nuclear waste. It is also important to take into account that similar alteration reactions may occur at the site of stored nuclear waste where temperatures in excess of 100°C might be encountered.

Hectorite was synthesized from a Eu(III)-bearing brucite precursor in a multistep procedure. In a separate experiment, Eu(III) ions were adsorbed onto hectorite in suspension. Colloids were extracted from both samples. The size distributions in the colloidal fractions were characterized by application of the asymmetrical flow field-flow fractionation (AsFlFFF) method and the corresponding elemental compositions were obtained by ICP-MS. Extended X-ray absorption fine structure (EXAFS) spectroscopy was used to characterize the local chemical environment surrounding Eu in the bulk samples and in the colloidal fractions.

The EXAFS results show that Eu is associated with hectorite upon co-precipitation or adsorption. Results from AsFlFFF suggest that Eu is structurally associated with the colloidal fraction extracted from bulk Eu-bearing co-precipitated hectorite. The AsFlFFF data are different for the colloidal fraction containing Eu(III) adsorbed on hectorite; in this sample they are consistent with a surface retention mechanism. These small but significant differences enable surface sorbed Eu to be distinguished from co-precipitated Eu. Eu is very likely located in a clay-like environment in the co-precipitation experiment, and it forms inner-sphere surface complexes in the adsorption experiment. The results obtained using the different experimental techniques agree, and show the benefits of using multiple methods of analysis.

Trivalent europium was used as non-radioactive chemical homologue for trivalent actinides. Similar retention mechanisms are expected for the trivalent actinides if they are co-precipitating with or adsorbing onto sheet silicates. The present study provides information which can be usefully added to the safety assessments required for deeply buried nuclear waste disposal sites.

Pyrochlore has been considered as a waste form to immobilize high-level nuclear waste such as excess weapons-grade plutonium. In order to study the aqueous stability of pyrochlore, we have carried out hydrothermal experiments with a natural microlite (Nb + Ta > 2Ti; Ta ⩾ Nb) at 175°C in a neutral and acidic solution for 14 days. The starting material and the experimental products were studied by electron microprobe, backscattered electron (BSE) imaging, powder infrared (IR) and micro-Raman spectroscopy, and powder X-ray diffractrometry (XRD). The microlite has small U (∼200 ppm) and Th (∼800 ppm) contents and is crystalline. The hydrothermal treatment in a solution containing l mol/l HCl and l mol/l CaCl2 causes the partial replacement of the microlite (up to ∼10 μm) by a new pyrochlore phase. This new phase is characterized by a larger unit-cell volume and contains a large number of vacancies at the A site (A = Ca, Na) as well as anion vacancies, molecular water, and possibly OH groups. Analyses of the experimental fluid further revealed that U was also selectively lost to the solution. Treatment in pure water did not produce reaction zones detectable by BSE imaging or powder XRD. However, significant spectral changes in powder IR spectra of the reaction product and the detection of Na and Ca in the solution indicate that the microlite has also reacted in pure water. The experimental alteration features bear a remarkable resemblance to those seen in naturally altered microlite samples, suggesting that short-term experimental results can be applied to natural systems and vice versa in order to evaluate the long-term stability of a pyrochlore waste form.

During the past 70 years, more than 2000 metric tonnes of Pu, and substantial quantities of the ‘minor’ actinides such as Np, Am and Cm, have been generated in nuclear reactors. Some of these transuranium elements can be a source of energy in fission reactions (e.g. 239Pu), a source of fissile material for nuclear weapons (e.g. 239Pu and 2Np), and of environmental concern because of their long half-lives and radiotoxicity (e.g. 239Pu and 237Np). There are two basic strategies for the disposition of these transuranium elements: (1) to ‘burn’ or fission the actinides using nuclear reactors or accelerators; (2) to dispose of the actinides directly as spent nuclear fuel or to ‘sequester’ the actinides in chemically durable, radiation-resistant materials that are also suitable for geological disposal. For the latter strategy, there has been substantial interest in the use of actinide-bearing minerals, especially isometric pyrochlore, A2B2Oi (A = rare earths; B = Ti, Zr, Sn, Hf), for the immobilization of actinides, particularly plutonium, both as inert matrix fuels and nuclear waste forms. Systematic studies of rare-earth pyrochlores have led to the discovery that certain compositions (B = Zr, Hf) are stable to very high doses of α-decay event damage. Recent developments in the understanding of the properties of actinide-bearing solids have opened up new possibilities for the design of advanced nuclear materials that can be used as fuels and waste forms. As an example, the amount of radiation damage that accumulates over time can be controlled by the selection of an appropriate composition for the pyrochlore and a consideration of the thermal environment of disposal. In the case of deep borehole disposal (3—5 km), the natural geothermal gradient may provide enough heat to reduce the amount of accumulated radiation damage by thermal annealing.

Crystals of the compound Na4(UO2)(CO3)3 have been synthesized and the structure has been solved. It is trigonal with a= 9.3417(6), c = 12.824(1) Å, V = 969.2(1) Å3, space group Pc1 and Z = 4. The structure was refined on the basis of F2 (wR2 = 4.2%) for all unique data collected using Mo-Kα X-radiation and a CCD-based detector. The final R1 was 2.0%, calculated for 534 unique observed (Fo ≥ 4σF) reflections, and the goodness-of-fit (S) was 0.91. The structure contains a uranyl tricarbonate cluster composed of a uranyl hexagonal bipyramid that shares three equatorial edges with CO3 triangles. The uranyl tricarbonate clusters are connected through NaO6 and NaO5 polyhedra, forming a heteropolyhedral framework structure. This compound may be related to a uranyl carbonate phase with the same composition which has been reported as an alteration phase on the surface of Chernobyl ‘lava’, and as a mineral in the Jachymov ore district, Czech Republic.

Materials designed for nuclear waste disposal include a range of ceramics, glass ceramics and glass waste forms. Those with crystalline phases have provided the momentum for studies of minerals as a means to understand aspects of waste-form crystal chemistry, behaviour in aqueous systems and radiation damage over geological periods of time. Although the utility of natural analogue studies varies, depending upon the degree of analogy to the proposed geological repository and other factors such as chemical composition, the available data suggest that Th-U host phases such as brannerite, monazite, pyrochlore, zircon and zirconolite are resistant generally to dissolution in aqueous fluids at low temperatures. Geochemical durability may or may not extend to hydrothermal systems depending on the specifics of fluid composition, temperature and pressure. At elevated temperatures, for example, davidite may break down to new phase assemblages including titanite, ilmenite and rutile. Perovskite is generally less resistant to dissolution at low temperatures and breaks down to TiO2, releasing A-site cations to the aqueous fluid. Studies of radiation damage indicate that the oxide and silicate phases become amorphous as a result of the gradual accumulation of alpha-recoil collision cascades. Monazite tends to remain crystalline on geological time scales, a very attractive property that potentially eliminates major changes in physical properties such as density and volume, thereby reducing the potential for cracking, which is a major concern for zircon. In spite of recent success in describing the behaviour of Th-U minerals in geological systems, considerable work remains in order to understand the P-T-X conditions during alteration and T-t history of the host rocks.

We revise the data fitting in our original paper [The dissolution rates of simulated UK Magnox - ThORP blend nuclear waste glass as a function of pH, temperature and waste loading, Miner. Mag.79 (2015) 1529–1542]. The intrinsic rate constant data were calculated incorrectly, the corrected data are presented herein. To support the corrected analysis we have also taken the opportunity to report some additional 90°C data. The conclusions of the original paper remain sound.

It is UK Government policy to dispose of higher activity radioactive waste through geological disposal into an engineered deep underground geological disposal facility (GDF; DECC, 2014). Those wastes include low-level (LLW) and intermediate-level (ILW) radioactive wastes that are very heterogeneous, containing a range of inorganic and organic materials, the latter including cellulosic items. After closure of the GDF, eventual resaturation with groundwater is expected, resulting in the development of a hyperalkaline environment due to the proposed use of a cementitious backfill. Under these high-pH conditions, cellulose is unstable and will be degraded chemically, forming a range of water-soluble, low molecular weight compounds, of which the most abundant is isosaccharinic acid (ISA). As ISA is known to form stable soluble complexes with a range of radionuclides, thereby increasing the chance of radionuclide transport, the impact of microbial metabolism on this organic substrate was investigated to help determine the role of microorganisms in moderating the transport of radionuclides from a cementitious GDF. Anaerobic biodegradation of ISA has been studied recently in high-pH cementitious ILW systems, but less work has been done under anaerobic conditions at circumneutral conditions, more representative of the geosphere surrounding a GDF. Here we report the fate of ISA in circumneutral microcosms poised under aerobic and anaerobic conditions; the latter with nitrate, Fe(III) or sulfate added as electron acceptors. Data are presented confirming the metabolism of ISA under these conditions, including the direct oxidation of ISA under aerobic and nitrate-reducing conditions and the fermentation of ISA to acetate, propionate and butyrate prior to utilization of these acids during Fe(III) and sulfate reduction. The microbial communities associated with these processes were characterized using 16S rRNA gene pyrosequencing. Methane production was also quantified in these experiments, and the added electron acceptors were shown to play a significant role in minimizing methanogenesis from ISA and its breakdown products.

Two 20-cm long columns of MX-80 bentonite compacted at a nominal dry density of 1.7 g/cm3 with a water content of 17% were tested in thermo-hydraulic (TH) cells with the aim of simulating the conditions of a sealing material in a nuclear waste repository. On top of the columns a hydration surface simulated the host rock supplying groundwater and at the bottom a heater simulated the waste canister. The tests comprised two phases: a heating phase and a ‘heating + hydration’ phase. The temperatures at the ends of the columns were set during the last phase to 30°C at the top and 140°C at the bottom, respectively. The thermo-hydraulic treatment resulted in major changes along the bentonite columns. These changes led to significant gradients along the column with respect to the physical state (water content, dry density) and geochemistry of the bentonite. Smectite dissolution processes occurred. As a result, colloids were probably produced, particularly in the more hydrated areas. In the warmest part of the columns precipitation of carbonates took place, caused by their solubility decrease with temperature and the evaporation. The increase in water content reduced the ionic strength of the pore water in the more hydrated areas where species such as gypsum were dissolved. The solubilized ions were transported towards the bottom of the columns; Na+, Ca+, Mg2+ and SO42− moved at a similar rate and K+ and Cl− moved farther. These solubilized ions precipitated in the form of salts farther away along the columns as the test was longer. The TH treatment implied the loss of exchangeable positions in the smectite, particularly towards the heater. The cation exchange complex was also modified.

Laser-driven relativistic electrons can be focused onto a high-Z convertor for generating high-brightness γ-rays, which in turn can be used to induce photonuclear reactions. In this work, photo-transmutation of long-lived radionuclide 135Cs induced by laser–plasma–interaction-driven electron source is demonstrated using Geant4 simulation (Agostinelli et al., 2003 Nucl. Instrum. Meth. A506, 250). High-energy electron generation, bremsstrahlung, as well as photonuclear reaction are observed at four different laser intensities: 1020, 5 × 1020, 1021, and 5 × 1021 W/cm2. The transmutation efficiency depends on the laser intensity and target size. An optimum laser intensity, namely 1021 W/cm2, was found, with the corresponding photonuclear reaction yield reaching 108 J−1 of the laser energy. Laser-generated electrons can therefore be a promising tool for transmutation reactions. Potential application in nuclear waste management is suggested.

The analytical evaluation of the capability of Bremsstrahlung highly directional energetic γ-beam to induce photo transmutation of 90Sr (γ,n) 89Sr is presented. Photo transmutation of hazardous nuclear waste of 90Sr, one of the two main sources of heat and radioactivity in spent fuel into valuable nuclear medicine radioisotope of 89Sr is explained. Based on the calculations, a fairly decent fraction of gamma rays in this range are used in transmuting of 90Sr into 89Sr where according to the available experimental data it is shown that by irradiating a 1-cm thick 90Sr sample with lasers of intensity of 1021 W/cm2 and repletion rate of 100 Hz for an hour, the reaction activity would be 1.45 kBq. It is shown that there is not a linear relationship between the growth of the activity and increasing the laser intensity, but there is a dramatic increase in the growth rate especially between 1020and 1021 W/cm2. In this work, the advantage of photonuclear transmutation over the neutron capture transmutation for 90Sr isotope is also discussed.