The final disposal of nuclear waste is at the interface between the technologies of the nuclear fuel cycle that produce the waste and the natural hydrologic and geochemical cycles of geologic repositories. Despite this broad interdisciplinary scope, nuclear waste management, as practiced, remains “balkanized” among the relevant disciplines. The individual subdisciplines continue to work in relative isolation from one another: materials science dealing with the immobilization of nuclear waste; engineering science dealing with the design, construction and operation of the repository; geoscience dealing with the long-term behavior of host rocks and the hydrology; health science dealing with the effects of radiation; social sciences dealing with the issues of trust, risk and ethics. Understanding how these very different disciplines interact is fundamental to creating and managing a nuclear waste organization. Based on a comprehensive review of the scholarly and scientific literature of waste management, we have analyzed the evolution and structure of research in nuclear waste management between 1979 and 2017. Focusing on materials science, we show that some research themes have been isolated from the most central themes of nuclear waste management. Moreover, we observed a relative decline of the fundamental research in materials science. This decline was evidenced by a drop in the number of articles published in the proceedings of the MRS symposia “Scientific Basis for Nuclear Waste Management” since 2000. We argue for the need to more precisely and inclusively define the field of nuclear waste management.

Pure and Nb-doped zirconium germanate materials of composition K2-xZr1-xNbxGe3O9.H2O where x = 0, 0.1, 0.2 and 0.25 with the structure of the mineral umbite have been successfully synthesised. The parent material displays negligible ion exchange of K+ for Cs+ but the doped materials shows much improved exchange. Synchrotron X-ray diffraction shows substantial peak splitting which varies with increasing niobium content. Preliminary Rietveld refinements suggest a two phase model with a caesium and potassium rich doped umbite phase.

Two glass series in the ternary systems K2O-B2O3-SiO2 (KBS) and Na2O-B2O3-SiO2 (NBS) were studied in order to identify the main factors influencing the solubility of iodine. We established that iodine incorporation is strongly linked to the bulk chemistry, i.e. the SiO2/(B2O3+SiO2) molar ratio, and to the physical properties of the glasses, and we assessed three different solubility limits. Iodine in Si-rich glasses has a low solubility (≤1 mol% I) regardless of the alkali ion present. On the contrary, in B-rich glasses, the solubility is five times higher than in Si-rich glasses for Na-glasses, and more than six times higher for K-glasses. The strong dependence of iodine solubility on the bulk chemistry is related to the adaptability of the glass network. Furthermore, our data suggest that iodine is stable with different redox states in the glasses here analyzed.

In the context of safety assessment, the fraction of inventory that is expected to rapidly dissolve when water contacts the spent fuel is called the Instant Release Fraction (IRF). Conceptually, this fraction consists of radionuclides outside of the uranium dioxide matrix and therefore the fraction can be further divided into the radionuclides in the fuel/cladding gap and radionuclides in the grain boundaries. The relative importance of these two fractions is investigated here for two Swedish high burnup fuels through simultaneous grinding and leaching fuel fragments in simplified groundwater for a short period of time. The hypothesis is that this will expose grain boundaries to leaching solution and provide an estimate of the release of the grain boundary inventory upon contact with water. The studied fragments were used in previous leaching experiments and thus pre-washed to remove any pre-oxidized phases. The results showed a significant release of iodine, cesium and rubidium and to a lower extent molybdenum and technetium. The fraction of inventory in the aqueous phase of actinides and lanthanides was 1-2 orders of magnitude lower than for the elements associated to the IRF. Both fuels displayed a very similar behavior and no correlation as a function of burnup or fission gas release was found.

The “safety case” approach has been developed to address the issue of evaluating the performance of a geologic repository in the face of the large uncertainty that results for evaluations that extend over hundreds of thousands of years. This paper reviews the concept of the safety case as it has been defined by the international community. We contrast the safety case approach with that presently used in the U.S. repository program. Especially, we focus on the role of uncertainty quantification. There are inconsistencies between the initial proposal to dealing with uncertainties in a safety case and current U.S. practice. The paper seeks to better define the safety case concept so that it can be usefully applied to the regulatory framework of the U.S. repository program.

The forward rate of dissolution of the International Simple Glass (ISG) was determined under alkaline conditions at 40 °C using the Single Pass Flow Through (SPFT) method. Forward rates were consistent with those obtained in the literature for this glass composition. The formation of altered gel layers and surface pits was observed on the surface of glass particles, especially at the very highest pH values, despite the application of high flow rates to prevent the build-up of solubility limiting phases. These features could be attributed to preferential localized dissolution at sites with a higher alkali concentration or from a separate, less durable, vitreous phase. These results may indicate that surface pit and altered gel formation occurs under the forward rate of dissolution as imposed by the SPFT method, particularly for simplified borosilicate glass materials.

The accident at the Fukushima Dai-ichi Nuclear Power Plant occurred following the Great East Japan Earthquake on March 11, 2011, and led to the release of volatile radionuclides, which were deposited on the environment in the Fukushima prefecture and the neighbouring areas. After the short-lived I-131, radiocaesium such as Cs-134 and Cs-137 have controlled radiation dose rate. The authors derived the apparent diffusion coefficients (Da) of some radionuclides such as Cs-134 and Cs-137 based on time variation of the depth distributions of respective radionuclides in soil obtained in field investigations in earlier studies. Almost all Da-values were of order 10-14 (m2·s−1) and well consistent with distribution coefficients (Kd) obtained from batch experiments. Whilst, field investigations for the relaxation mass depths of Cs-137 in soil by a scraper plate method were conducted at totally 85 locations over a period of nearly 6 years from December 2011 in the Fukushima prefecture and the neighbouring prefectures, and time variation of the effective relaxation mass depths was recently reported. Consequently, the effective relaxation mass depths of Cs-137 showed a tendency to linearly increase with increasing time. This indicates that radiocaesium gradually moves to the deeper part of soil with time. In this study, Da based on Fick’s diffusion equation was derived based on time variation of the effective relaxation mass depths of Cs-137 in soil. In order to calculate the Da based on Fick’s law, correlation between relaxation depth and square root of elapsed time was derived from the correlation between effective relaxation mass depth and elapsed time (where, relaxation depth is defined as the depth of 1/e of radionuclide concentration at the ground surface and can calculate by relaxation mass depth/soil density). The calculated Da of Cs-137 was of order 10-12 (m2·s−1) , which was about 2 orders of magnitude higher than Da-values that the authors previously reported as described above. Considering that almost all relaxation depths of Cs-137 were shallow within 2cm in depth from the ground surface and near the surface layer of soil is unsaturated, it is considered that Da estimated in this analysis includes the effect of dispersion by advection (by flow in the vertical direction of rainwater).

A glass-ceramic borosilicate waste form is being considered for immobilization of waste streams of alkali, alkaline-earth, lanthanide, and transition metals generated by transuranic extraction for reprocessing used nuclear fuel. Waste forms are created by partial crystallization on cooling, primarily of oxyapatite and powellite phases. In-situ neutron diffraction experiments were performed to obtain detailed information about crystallization upon cooling from 1200°C. The combination of high temperatures and reactivity of borosilicate glass with typical containers used in neutron experiments, such as vanadium and niobium, prevented their use here. Therefore, methods using sealed thick-walled silica ampoules were developed for the in-situ studies. Unexpectedly, high neutron absorption, low crystal fraction, and high silica container background made quantification difficult for these high temperature measurements. As a follow-up, proof-of-concept measurements were performed on different potential high-temperature container materials, emphasizing crystalline materials so that residual glass in the waste form sample could be more easily analyzed. Room temperature measurements were conducted with a pre-crystallized sample in ‘ideal’ containers stable at low temperatures (i.e., vanadium and thin-wall silica capillaries) and compared to the same measurements in containers stable at high temperatures (i.e, platinum, single crystal sapphire, and thick-walled silica ampoules). Results suggested that Pt is probably the best choice if suitably sealed to prevent contamination from the sample after neutron activation.

The apparent diffusion coefficients of strontium in compacted bentonites were investigated at various concentrations of NaHCO3. Purified sodium bentonite Kunipia-F® was compacted with a jig into cylindrical pellets 10 mm in diameter and 10 mm high with dry densities of 1.0 to 1.6 Mg/m3. Each bentonite pellet was inserted into an acrylic resin column and saturated with carbonated water containing 0.1 to 1.0 M NaHCO3 for more than 1 month. The face of the bentonite specimen was spiked with 5 μL of 1.0 M SrCl2 tracer solution. After a few weeks, the strontium diffusion profiles were measured by inductively coupled plasma-mass spectrometry. The apparent diffusion coefficients of strontium decreased slightly with increasing dry density. NaHCO3 concentrations of 0.5 M decreased the apparent diffusion coefficients of strontium by half at a dry density of 1.0 Mg/m3 and quarter at 1.6 Mg/m3. At a higher NaHCO3 concentration of 1.0 M, no strontium diffusion profile was observed, whereas white precipitate was observed on the face of the bentonite specimen where it was spiked with strontium. This white precipitate could be strontianite, which is strontium carbonate. Diffusion experiments using cesium were carried out for comparison, and the presence of carbonate had no effect on the apparent diffusion coefficient.

Nuclear wastes generated from reprocessing of used nuclear fuel tend to contain a large fraction of rare earth (RE, e.g., Nd3+), transition (TM, e.g., Mo6+, Zr4+), alkali (A, e.g., Cs+), and alkaline earth cations (AE, e.g., Ba2+, Sr2+). Various strategies have been considered for immobilizing such waste streams, varying from nominally crystal-free glass to glass-ceramic to multi-phase ceramic waste forms. For glass and glass-ceramic waste forms, the added glass-forming system is generally alkali-alkaline earth-aluminoborosilicate (i.e., Na-Ca-Al-B-Si oxide). In a US-UK collaborative project, summarized here, we investigated the glass structure and crystallization dependence on compositional changes in simulated nuclear waste glasses and glass-ceramics. Compositions ranged in complexity from five – to – eight oxides. Specifically, the roles of Mo and rare earths are investigated, since a proposed glass-ceramic waste form contains crystalline phases such as powellite [(AE,A,RE)MoO4] and oxyapatite [(RE,AE,A)10Si6O26], and the precipitation of molybdenum phases is known to be affected by the rare earth concentration in the glass. Additionally, the effects of other chemical additions have been systematically investigated, including Zr, Ru, P, and Ti. A series of studies were also undertaken to ascertain the effect of the RE size on glass structure and on partitioning to crystal phases, investigating similarities and differences in glasses containing single RE oxides of Sc, Y, La, Ce, Nd, Sm, Er, Yb, or Lu. Finally, the effect of charge compensation was investigated by considering not only the commonly assessed peralkaline glass but also metaluminous and peraluminous compositions. Glass structure and crystallization studies were conducted by spectroscopic methods (i.e., Raman, X-ray absorption, nuclear magnetic resonance (NMR), optical absorption, photoluminescence, photoluminescence excitation, X-ray photoelectron spectroscopy), microscopy (i.e., scanning electron microscopy, transmission electron microscopy, electron probe microanalysis), scattering (i.e., X-ray and neutron diffraction, small angle measurements), and physical characterization (i.e., differential thermal analysis, liquidus, viscosity, density). This paper will give an overview of the research program and some example unpublished results on glass-ceramic crystallization kinetics, microstructure, and Raman spectra, as well as some examples of the effects of rare earths on the absorption, luminescence, and NMR spectra of starting glasses. The formal collaboration described here has resulted in the generation of a large number of results, some of which are still in the process of being published as separate studies.

Reactions of alkali salts (nitrates, sulfates, carbonates, halides, borates) play a key role in the low temperature feed conversion occurring at the cold cap during processing of Hanford Low Activity Waste (LAW) glass melters. An alkali salt phase can sometimes form, and preferentially incorporate radionuclides of Cs, Cl, I, and Tc. During melting of the slurry feed, some of the feed components sequentially break down with increasing temperature to form gases (i.e., nitrates ➔ NOx, carbonates ➔ CO2, and boric acid ➔ H2O) or partially volatilize (halides). Sulfate, however, tends not to volatilize but has limited solubility in the final borosilicate glass waste form. To improve understanding of these low temperature processes and their composition dependencies, a scoping study was undertaken to synthesize salt systems that remain amorphous at room temperature, thus facilitating structural study. Melts of equimolar ratios of K2SO4-ZnSO4 (a known ionic glass-forming system) with added nitrates, halides, or carbonates, were melted and quenched. Some of the materials formed single phase glasses and some underwent crystallization upon quenching. Characterization of these quenched materials by thermal analysis, infrared absorption, and diffraction was performed. Addition of other anions to the sulfate base glass resulted in a distortion of the sulfate tetrahedron, as evidenced by infrared absorption. Carbonates strongly promoted crystallization, mostly of carbonate phases. Nitrates promoted crystallization of ZnO, and the nitrate volatilized with some incorporating into the glass. Halides tended to incorporate into the glass, but the small (F) and large (I) halogens promoted crystallization of sulfate-containing crystals, while moderate sized (Cl) halogens produced single-phase ionic glasses.