In disposing of high-level radioactive waste, the drop in pH in the repository as the iron overpack corrodes must be considered. Plutonium migration behavior may be affected by the pH of pore water in compacted bentonite barriers in high-level waste repositories. To examine the effect of pH on migration behavior, H-bentonite was prepared by treating Japanese Na-bentonite, Kunipia-F, with hydrochloric acid. Diffusion experiments were performed with mixtures of Na- and H-bentonites. The pH value in the pore water of the water-saturated bentonite mixtures decreased from 8 to 3 as the mixing ratio of H-bentonite increased. Diffusion experiments were carried out by using 238Pu then apparent diffusion coefficients were determined from the plutonium distribution in the specimens. The apparent diffusion coefficients were on the order of 10-13 to 10-12 m2/s at pH values lower than 4, whereas they were less than 10-14 m2/s at pH values higher than 6.5. These results indicate that plutonium diffused faster as Pu3+ or PuO22+ due to disproportionation at lower pH while plutonium could be retarded as Pu(OH)40 by sorption on bentonite at higher pH.

The swelling stress of bentonite which is one of the engineered barriers and backfill materials for radioactive waste disposal is strongly dependent on water chemistry such as saline water. The authors have developed a thermodynamic model for calculating the swelling stress (pressure) of bentonite, based on the thermodynamic data of interlayer water in Na-montmorillonite obtained in earlier studies. In this work, the swelling stress of water-saturated Na-bentonite was calculated for various bentonite dry densities and solution conditions such as sodium chloride concentration and nitrate concentration and compared to the measured data.

Swelling stress versus montmorillonite partial density was estimated for solutions containing sodium chloride ([NaCl] = 0−3.4 m, m: molality) and nitrate concentrations ([NaNO3] = 0−6 m) and compared to data measured for bentonites with various montmorillonite contents and silica sand contents. The calculated swelling stresses commonly decreased with increasing [NaCl] and [NaNO3] for the same montmorillonite partial density. The trend of swelling stress versus [NaCl] was in a good agreement with the measured results. The calculated swelling stress versus [NaCl] was also quantitatively in a good agreement within the scattering of the measured data. The trend versus [NaNO3] was also similar to that versus [NaCl]. However, the calculated results were quantitatively different from the measured data ([NaNO3] = 3, 5 m, montmorillonite partial density = 0.76−0.87 Mg/m3). Even though those measurements were conducted under the condition of high ionic strength, the measured data of swelling stresses were almost the same as in the condition of distilled water. Since the measurement periods were quite rather short, it is conceivable that the measurements were not done at equilibrium.

Bentonite is used as one of the materials for engineered barrier systems in a radioactive waste repository. Since the major clay mineral constituent of bentonite is montmorillonite, its physico-chemical properties are important. Basal spacing of water-saturated Na-montmorillonite is reported to decrease with increasing Na-montmorillonite density. This paper presents a thermodynamic model to calculate change in the interlayer space of Na-montmorillonite based on the relative partial molar Gibbs free energy (dG) of interlayer water as contacted with a solution of an arbitrary salinity (NaCl concentration). Directly change in montmorillonite density (ρdm) against salinity was calculated by the thermodynamic model. The dG of interlayer water as contacted with a solution of an arbitrary salinity can be calculated by dG = dGH2O+ dGS (dGH2O: relative partial molar Gibbs free energy of interlayer water, dGS: that of water in a solution of an arbitrary salinity). The author previously reported an empirical correlation of dGH2O vs. water content for Na-montmorillonite. The dependence of ρdm on salinity was calculated by replacing dGH2O in the empirical correlation with dG. ρdm increased with salinity. Concretely, initially the ρdm-values of 0.5 and 1.0 Mg/m3 increased to 1.05 and 1.16 Mg/m3 under 0.5 m-NaCl, respectively. Interlayer space vs. salinity was estimated based on the measured results of basal spacing vs. ρdm by XRD and the average density of montmorillonite vs. salinity calculated by this model.

Titration of bentonite, or clay materials in general, can be carried out in two ways, namely by semi-continuous method and batch method. This study compares these two methods, and, in addition, presents the results of two different evaluation and regression procedures used for fitting of experimental titration curves and for calculation of corresponding parameters. Bentonites B75 and S65, and montmorillonite SAz-1, were chosen for this study. Regarding the semi-continuous method, the technique developed by Wanner et al. was used. As for the batch method, two-week contact time for selected solid/liquid ratio 1 g per 250 ml was used on the base of preliminary experiments. The evaluation of result obtained using the batch method with back titration correction gave more relevant data, on the other hand, with the use of this method is connected a risk of solid phase dissolution. The chemical equilibrium model (i.e. non-electrostatic), incorporated into codes (I) PHREEQC + UCODE_2005, and (II) PDNLRG.fm (code package STAMB) + software FAMULUS, was used to the evaluation of the titration curves. The codes mentioned give only slightly different resulting values in the case of batch method.

Spent nuclear fuel from Finnish power plants is planned to be deposited deep in the crystalline bedrock in Olkiluoto, Finland. The bedrock needs to be well characterized to assess the risks inherent to the long term safety of the site. In the bedrock the possibly released radionuclides are mainly transported by water conducting fractures and their transport is retarded by diffusion and sorption. In porous materials these properties are typically linked to microscopic pore structure (pore size distribution, tortuosity and constrictivity) and chemical nature of the minerals and groundwater.

In this work transport properties of veined gneiss (VGN) and pegmatitic granite (PGR) samples from Olkiluoto were studied using various through diffusion experiments and the C-14-PMMA autoradiography. Through diffusion experiments were performed on rock cores using HTO and 36Cl in water phase and He in gas phase as tracers. The effective diffusion coefficients (De) determined for the VGN were found to be dependent on the tracer molecule (De(HTO) < De(He) < De(Cl)) whereas for the PGR such a dependence was not found. The porosity distributions determined by the C-14-PMMA autoradiography revealed the difference in the pore structure between the samples. The porosity of VGN consists mostly of grain boundary pores and pores between biotite lamellae. Due to a high content of nanometer scale pores anion exclusion affected the results of 36Cl and Knudsen diffusion the ones of He for VGN. Furthermore, in the PGR micrometer scale intra- and intergranular fissures form a connected network for diffusive transport and thus all tracers diffuse at the same rate.

Spent nuclear fuel from Finnish power plants is planned to be deposited deep in the crystalline bedrock in Olkiluoto, Finland. The bedrock and more specifically the elemental composition of ground water, which is composed of the fracture water and the matrix pore water, needs to be well characterized to assess the risks inherent to the long term safety of the site. To this end, it is valuable to investigate elemental composition of the matrix pore water since it tends to conserve hydrogeological signals for longer time spans compared to open fracture waters.

In this study, the chloride concentration of matrix pore water in veined gneiss (VGN) and pegmatitic granite (PGR) samples were investigated. Chloride was out-diffused from the naturally saturated rock cores into deionized water. Chloride pore diffusion coefficients were derived by modelling the chloride breakthrough curves obtained from the out-diffusion experiments. Two component modelling gave best fit to the experimental results. There two diffusion coefficients were (9±2)×10-11 m2/s and (0.5±0.1)×10-11 m2/s for PGR and (2.5±0.5)×10-11 m2/s and (0.4±0.1)×10-11 m2/s for VGN. Porosity distribution and total porosities of the rock samples were studied with the C-14-PMMA autoradiography. Porosity for PGR was found to be 0.6 % with large mineral transecting fissures, and porosity for VGN was found to be 0.7 % with highly porous mineral clusters connected to each other via grain boundaries and intragranular pores. The findings here show that heterogeneity has to be taken into account in modelling to find better agreement with the experimental results. C-14-PMMA autoradiography results indicate dual-component behavior for diffusion in PGR and VGN which were used in the modelling.

The migration behavior of the long-lived actinides was studied under the conditions of the deep disposal of the acidic liquid nuclear waste (LNW). Composition of LNW varies significantly including acidic technological wastes (pH ∼2.4), which consist of sodium nitrate, acetic acid, corrosion products (Fe, Cr, Mn, Ni, Al), fission products and actinides. Corrosion products tend to precipitate under the LNW disposal conditions that favor forming of the phases with high sorption capacity towards actinides. Sands of reservoir bed have their own initial sorbent surfaces besides new secondary phases that have formed as a result of interaction with acidic LNW. The nearest to the injection well conditions are gradually changing from pH ∼2.4 till neutral values due to the dilution by groundwater with formation of new precipitated phases of corrosion products. The solid phases characterization is a necessary step on the path of knowledge of migration behavior of actinides. The secondary phases of both corrosion products and sands of reservoir bed under LNW disposal conditions were characterized using XRD, SEM and Mössbauer spectroscopy. The recent results of the analyses of the behavior of actinides (Pu. U, Np, Am) under the conditions of the injection of the acid LNW are presented in the paper.

Study of sorption behavior of radionuclides toward five rock samples drilled in the deep-well R12 in the exocontact zone of Nizhnekansky granitoid massif (“Eniseysky” area) in the range of depth 166 m – 476 m was carried out. The sorption kinetics and Kd values of the long-lived radionuclides of different chemical behavior 137Cs, 226Ra, 79Se, 237+239Np, 239,240Pu, 241,243Am were determined. Experiments were performed under the conditions that are relevant for the future high level waste disposal (atmosphere; composition, pH and Eh of the solutions). Digital radiography demonstrated heterogeneous sorption of all the investigated radionuclides and revealed phases with higher sorption ability. It was established that sorption of metal cations is fast and reaches steady state in a few hours while Se in the form of selenate-ion sorbs insignificantly.

Bentonite clay is proposed to be a buffer material in the KBS-3 method of deep geological disposal of spent nuclear fuel in Sweden and Finland. It is essential to know the long term behaviour of the material when evaluating the safety performance of the barrier system. Therefore, various experiments and modelling have been carried out with bentonite in different conditions during the last decades. However, most of the laboratory experiments in controlled conditions last less than few years only.

In this study 15 years old compacted MX-80 bentonite samples were studied. They stayed closed in a copper canister in contact with low salinity groundwater both in aerobic and anaerobic conditions. After dismantling the experiment a thorough characterization was carried out, including analysis of chemical composition of bentonite and external solution, clay microstructure, mineralogy and microbial activity.

In this presentation the focus is on the microstructural features of the samples. Small-angle X-ray scattering (SAXS), nuclear magnetic resonance (NMR) and anion exclusion measurement were used to investigate the nano and micro porosity features of the material. The results are compared with similar analysis of MX-80 samples prepared in the laboratory and with short time equilibration with external solution. Calculations based on SAXS data give similar results but there are differences for NMR based results. This could be caused by dissimilarities of between the samples at the mesoscale level. No differences were observed between the oxic and anoxic conditions suggesting no influence of redox conditions on the bentonite microstructure.

Since 1988 the Swedish Nuclear Fuel and Waste Management Co. operates a repository for low- and intermediate-level short-lived radioactive waste, SFR, in Forsmark, Sweden. Due to decommissioning of the nuclear power plants additional storage capacity is needed. In December 2014, an application to extend the repository was therefore submitted. One key component of this application was an assessment of post-closure safety of the extended SFR. For this safety assessment, a methodology based on that developed by SKB for the spent nuclear fuel repository was used and the impact of the degradation of repository components, the evolution of the surface system and changes of future climate on the radiological safety of the repository was assessed over a period of 100,000 years. The central conclusion of the SR-PSU safety assessment is that the extended SFR repository meets requirements on protection of human health and of the environment that have been established by the Swedish radiation safety authority for the final disposal of radioactive waste. Furthermore, the design of the repository was shown suitable for the waste selected and the applied methodology suitable for the safety assessment.

There are some regions where it is difficult to exclude the possibility of appearance of new volcanic activity on the site even if the geological disposal site is located far from existing volcanos in Japan. In order to identify the influence of volcanic eruption at the site to the public if it occurs, public exposure doses were evaluated based on the two scenarios considering types of eruption at new volcanic activity in Japan. One is the exposure by volcanic ash widespread by Strombolian eruption and deposited on the ground surface, including radionuclides from vitrified waste forms after a volcanic conduit penetrated the repository. The other is that by waste forms appeared at the ground surface by Merapi type pyroclastic flow. Exposure doses of the residents living on the volcanic ash do not exceed 1mSv/y even when the eruption occurs at 1,000 years after closure of disposal site. Exposure dose rate for the volcanic researchers temporarily approaching naked waste forms without realizing radioactive wastes is 0.8mSv/h when the eruption occurs 100,000 years after. It indicated that attention should be paid to the influence by Merapi type pyroclastic flow on researchers approaching waste forms appeared rather than that by Strombolian eruption on residents living on the volcanic ash widespread.