In this study, the cation exchange capacity (CEC) and leached exchangeable cations (LC) of montmorillonite purified from bentonite produced in the Tsukinuno bentonite mine, Yamagata, Japan, were measured, and the exchangeable cations in the interlayer of the montmorillonite were discussed. A montmorillonite, in which the soluble minerals were completely removed, was prepared. Kunipia-F and Kunipia-P, for which both bentonites originally contain approximately 100 wt.% montmorillonite, were used as the initial material. All of the measurements were carried out in a N2 atmosphere-controlled glove-box.

The CEC values of montmorillonites for both bentonites (100-110 meq/100g) were similar to data conventionally reported, and the sum of LC was also approximately in good agreement with the CEC values. The share of Na+ in the interlayer of montmorillonite calculated from the LC was about 3/4 of the sum of the LC (≍ CEC), and Mg2+ and Ca2+ occupied about 7 and 19 %, respectively. Although montmorillonite in bentonite produced in the Tsukinuno bentonite mine is known as a Na type, the sum of Ca2+ and Mg2+ occupied about 26 % of all exchangeable cations in the interlayer. Based on these data, the ion exchange reaction constant between Na+ and H+ in the interlayer of montmorillonite was calculated to be -0.07. This is nearly 2 orders of magnitude lower than data that are usually adopted.

Colloids generated from the engineered barriers of a high level radioactive waste repository (HLWR) emplaced in crystalline rock could play a significant role in radionuclide transport and they are of concern for the safety assessment of these repositories.

The main objectives of this study are: a) to analyze the transport properties of colloids in a crystalline fractured rock under hydrodynamic conditions as similar as possible to those expected in a repository (i.e. low flow rates) and b) to discuss the effects of their presence on the transport of radionuclides.

Transport experiments with bentonite and latex colloids in a fractured granite column from the Grimsel Test Site (Switzerland) were carried out, under geochemical conditions ensuring colloid stability (alkaline and low ionic strength water). Transport experiments were also carried out with 85Sr and 233U and the results with and without the presence of bentonite colloids were compared.

Colloid filtration in the fracture was always observed, even when colloids presented high stability and the conditions were unfavorable to colloid attachment to rock surfaces, being both the colloids and the rock negatively charged and the fracture surface smooth. The retention in the fracture depended on the water flow rate, increasing the retention as the water flow decreased.

This work illustrates as both the mobile and retained fraction of colloids, which strongly depend on the hydrodynamic conditions, are of importance in the overall radionuclide mobility.

Diffusion and sorption of cesium (Cs) and iodine (I) were investigated in a purified and moderately compacted sodium montmorillonite (dry density of 800 kg m-3) saturated with 0.01, 0.1 and 0.5M NaCl solutions. The effective diffusivity (De) and capacity factor (α) for Cs and I were measured by through-diffusion experiments, coupled with multiple curve analyses, including tracer depletion, breakthrough and depth concentration curves, which could be fitted with a conventional diffusion model using only one set of parameters. The De values obtained for Cs were of the order of 10-9-10-10 m2 s-1 and decreased as salinity increased, and those for I were of the order of 10-11-10-12 m2 s-1 and showed the opposite dependency. The distribution coefficient (Kd) of Cs decreased from the order of 100 to 10-2 m3 kg-1 as salinity increased. Diffusion and sorption parameters for Cs were also obtained by in-diffusion and batch sorption experiments and showed good agreement with those obtained by the through-diffusion experiments. The diffusion model, based on homogeneous pore structure and electrical double layer (EDL) theory, predicted the salinity dependence of De reasonably well, showing the effect of cation excess and anion exclusion as a function of salinity. The apparent diffusivity (Da), which includes sorption effects, was also interpreted by a coupled sorption model.

In this investigation the swelling pressure of confined samples of a real waste form, consisting of bitumen containing washing water concentrate, was measured. The theoretically reachable swelling pressure was estimated as high as ∼ 273 bar (27 MPa).

A first simple experimental set-up was a sample of bituminized radwaste embedded in a cylinder of sintered bronze with a piezo-resistive pressure transducer. This was put into a leaching medium. A maximum pressure of 2 bar (0.2 MPa) was measured. The leaching rates of these confined samples were similar to those of unconfined ones.

The second set-up was a sample of bituminized radwaste embedded in a cylinder of stainless steel, also equipped with a pressure transducer. The open bottom of the cylinder was closed with a disc of porous cement with water/cement ratio of 1.3. In two similar experiments with this set-up the swelling pressure reached in one case a value of 9 bar = 0.9 MPa and in the second improved one about 15 bar = 1.5 MPa.

The FEBEX in-situ experiment, installed in 1997 at the Grimsel Test Site (GTS, Switzerland) 400 m depth under the Swiss Alps, simulates a high level radioactive waste repository (HLWR) emplaced in granite. Its initial aim was to study the performance of a bentonite engineered barrier but recently, two new boreholes were drilled in the granite to study the possible bentonite colloid formation and their migration in the granite.

This study presents the characterization performed, at the micrometer scale, of the threemain water conductive fractures that were identified on the granite cores extracted from the newboreholes. These fractures are possible pathways for bentonite colloid transport (or retention),may be source of natural colloids and may condition colloid stability. The nuclear ion beamtechniques µ-Particle X-Ray Emission (µPIXE) and Rutherford Backscattering Spectrometry(RBS) were applied for visualizing and quantifying the elemental composition of the fracturessurface and of the surrounding micro-fractures, as support of the bentonite colloid analyses.

Alloy 22 is highly resistant to all forms of corrosion; however, it may suffer crevice corrosion in presence of chloride ions especially at temperatures higher than ambient and at anodic potentials. The susceptibility of Alloy 22 to suffer crevice corrosion is highly dependent on the type of electrolyte solution that is in contact with the alloy including variety of species in solution and their relative concentration. Laboratory research has shown that at a constant chloride concentration, the susceptibility of Alloy 22 to crevice corrosion is not influenced by the nature of the cations present in solution. On the other hand, that nature of the anions is highly influential. Of the anions present in ground water, only chloride is detrimental and the others are inhibitors or innocuous for crevice corrosion susceptibility. That is, the presence of the other anions counter balances the negative effect of chloride. The inhibition effect is explained and the likelihood that Alloy 22 would suffer crevice corrosion in contact with ground water is discussed.

The major part of groundwater contaminants strongly interact with soils and aquifer rocks. Therefore sorption processes on porous matrix are of utmost importance in the frame of the nuclear waste disposal. The objectives of this study were to evaluate sorption uptake by silica sand of some safety-relevant metal ions such as Cs+, Sr2+, Cu2+, Ni2+ and to investigate the existence of competitive sorption processes between these ions. To this aim, kinetic and equilibrium, mono-component and multi-component, batch experiments were carried out in order to study: i) the influence of metal concentration, pH and contact time, on sorption onto silica sand of the above ions in aqueous solution, and ii) the presence of competition phenomena. Sorption data were well fitted by Langmuir and Freundlich models. Multi-component tests show that the uptake of each ion is reduced in presence of other ions in solution with respect to mono-component batch tests results and that competition between species appears influenced by the equilibrium times of the single species in solution and by pH.

From uranium-ore treatment to spent fuel recycling, waste treatment and conditioning and to final storage of waste packages, radionuclides are involved in numerous chemical and physical reactions. The understanding of their chemical forms (speciation) and behavior as well (retention, complexation,...), as a function of the environment conditions (T, P, solid/liquid/gas interfaces), are key issues for the development of the industrial nuclear activities. Dedicated analytical tools are needed to determine the radionuclide concentrations and speciation in the liquids, solids and gas, over a wide range of concentrations and matrixes. The obtained experimental data on radionuclide speciation are integrated in dedicated data bases, supporting various models used to simulate the system behavior (i.e. RN migration under geological disposal, RN contamination in the primary fluids of nuclear power plants, RN behavior in the PUREX process, etc.). There are several needs in the following domains of the fuel cycle :

♦ The development of innovative methods to enhance analytical performances of isotopic composition of elements in irradiated fuels or waste streams arising from processed spent fuels. Isobaric interferences may be suppressed by specific ion-molecules reactions in collision/cell coupled with Mass Spectrometer, instead of preliminary chromatographic separations.

♦ The thermo chemistry at high temperature and pressure of the coolant fluids of the nuclear power plants, to model the solid/liquids interactions controlling its contamination by the activated products and hideout processes.

♦ The development of scientific and operational models of the radiolysis of organic molecules and materials, under extreme conditions (γ and α radiolysis), to understand the controlling long-term degradation phenomena ( i.e. H2 degassing in the waste packages).

♦ The fundamental understanding of sorption processes of redox sensitive elements such as U on specific mineral surfaces, in the presence of organic molecules, to develop dedicated tools for radionuclide monitoring and measurement in the environment.

This study is focused on the alteration behavior of spent nuclear fuel when exposed to highly alkaline groundwater. Contact of highly alkaline solution with the waste product is considered in the Belgian concept for disposal in the Boom Clay formation. According to the “supercontainer design” the fuel will be encapsulated in carbon steel canisters, surrounded by a concrete over-pack. After saturation of the engineered barriers by porewater, interactions with the concrete will result in solutions rich in NaOH, KOH and Ca(OH)2. Using this type of solution at pH 12.5, spent nuclear fuel corrosion experiments were conducted over 378 days. Under anoxic conditions, parallel experiments were performed (a) in the absence of Fe phases and (b) in the presence of solid Fe phases representing container (corrosion) products. Both types of experiments resulted in relatively low matrix dissolution rates, around 10-7 per day, according to the fractional release of Sr. Solution concentrations of actinides are close to or below the detection limit, indicating an effective retention of these radioelements in the system. The observed precipitation of a Ca rich phase onto the surfaces of the corroded fuel samples may be related to the inhibited re-lease of actinides, Sr and other matrix bound radioelements.

Since 2006, French research on spent fuel has focused on the main issues related to transport and extended in-pool storage of spent fuel assembly. Studies on creep behaviour of irradiated cladding have resulted in a new creep model which is valid over a wide domain of temperature, internal pressure and time. Under nominal conditions, no evolution of the spent fuel rod is expected during in-pool storage. In case of defective fuel rods in the storage pool, the consequences of fuel alteration on the initial defect of the cladding depend on the matrix alteration rate and nature of the secondary phases formed. Considering the optional scenario of direct disposal, the long-term behaviour of the spent fuel is investigated focusing on helium consequences before water contact on the one hand and on the influence of repository conditions on matrix alteration on the other hand. The aim of the on-going studies is to improve the safety margins initially introduced in the radionuclide source term models.

Two important issues in Performance Assessment exercises regarding the alteration of Spent Nuclear Fuel (SNF) are the contribution of Instant or better called Fast release Fraction (FRF) and the effect of High Burn-Up Structure (HBS). Therefore this paper focuses on the effect of HBS in FRF of a PWR irradiated in a commercial reactor with a mean Burn-Up (BU) of 48 GWd/tU. Additionally, determined FRF are compared with previous experiments performed with PWR with a mean BU of 60 GWd/tU, in order to evaluate the effect of BU on FRF.

In order to study the HBS contribution, static leaching experiments are performed with two samples prepared from different radial positions, labeled Core, central region and Out, enriched with HBS. Two synthetic leaching solutions, bicarbonate and bentonitic granitic groundwater are used under oxic conditions.

The estimated FRF are calculated from the determined Fraction of Inventory in Aqueous Phase (FIAP) taking into account the inventory determined experimentally for each fraction.

Higher release is observed for Core sample, which can be an indicative that HBS does not increase the RN release. With the exception of Rb and Cs, which release is higher at lower BU, no clear effect of BU is observed within the studied range.

The formation of uranyl secondary solid phases onto the spent nuclear fuel surface might influence the radionuclide concentration in solution via, among others, sorption processes. In this work, the incorporation of some radionuclides onto the uranium peroxide studtite, UO2O2·4H2O, has been tested.

The study was carried out in batch experiments where a known amount of studtite (0.05 g) was put in contact with 20 cm3 of radionuclide solution. Once equilibrium was reached, radionuclide concentrations in solution were determined by ICP-MS. The radionuclide amount attached to the solid was calculated from the mass balance. The S/V values of the experiments were also determined from BET specific solid surface area measurements.

In this work, data on sorption of caesium, strontium, and selenium as a function of pH are presented. The behaviour of caesium and strontium are similar: a relatively high amount of radionuclide is sorbed at neutral to alkaline pH while there is almost no sorption at acidic pH. On the other hand, in the case of selenium, the sorption maximum occurs at acidic pH and there is almost no sorption at alkaline pH. The different behavior of the radionuclides is related to the element speciation in solution and the surface charge of the solid. Strontium and caesium are sorbed at alkaline pH because they are positively charged in solution and the surface of the studtite is negatively charged (>O- groups) while selenium(VI) sorbs at acidic pH because the surface of the studtite is positively charged, and the predominant selenium(VI) species in solution is anionic.

These preliminary data indicate that the sorption capacity of uranyl secondary solid phases such as studtite is an important process to be considered when establishing the migration of different radionuclides released from spent nuclear fuel.

To assess the long-term behavior of spent fuel in a nuclear waste repository, the chemical reactions between the fuel and possible intruding water must be understood and the resulting radionuclide release must be quantified. The instant release fraction (IRF) source term assumed to be instantaneously accessible to water after the failure of the waste container. Some IRF values for different kinds of spent fuel are available in the literature. However, the possible contribution the rim restructured zone for high-burnup UOX fuels, was not necessary taken into account. A specific study of the leaching behavior of the rim zone has been carried out on a UOX fuel sample with a burnup of 60 GWd·t-1 and 2.8% FGR. The 134/137Cs and 90Sr rim IRF are effectively higher than the gap and grain boundary inventories (respectively 4.4 wt% and 0.3 wt% of the total Cs and Sr inventories). Nevertheless because of the relative small volume of this zone in the pellet, the impact of the rim inventory appears to be limited and the complete Cs and Sr IRF (gap, grain boundaries and rim), were estimated at 1.85 and 0.3 wt%, respectively

The year 2009 marks 20 years since the Environmental Management program was first established in the Department of Energy. At that time, nearly 50 years of nuclear activity had left a legacy that included nuclear waste and environmental contamination at more than 100 sites across the United States. The extent of the risk to our citizens and communities was unknown, and certainly many of the processes and technologies to reduce that risk had not yet been invented. Since then, the Department of Energy has closed 86 of 108 sites originally assigned to the program nationwide. The Department of Energy has packaged and safely stored the nation’s entire excess plutonium inventory. The Department has pioneered new technologies that have allowed progress in retrieving millions of liters of tank waste and safely disposing of tens of thousands of cubic meters of transuranic waste. In Fiscal years 2006 and 2007 alone, the Department of Energy demolished approximately 500 buildings (nuclear, radioactive, and industrial) as part of our decontamination and decommissioning projects. Finally, there have been great strides in restoring groundwater contaminated with radionuclides using innovative treatment systems. In August 2005, a rigorous project management system was instituted. This Department of Energy program was built on the principle of prioritizing risk reduction supported by our four guiding tenets of safety, performance, clean-up, and closure. The mission activities at our clean-up sites are targeted at our highest risk activities. In planning its environmental clean-up efforts and developing the budget for those activities, the Department seeks to focus on work that will produce the greatest environmental benefit and the largest amount of risk reduction.

Low- and intermediate-level operational waste from the Swedish nuclear power plants is stored in the SFR repository. The facility is located in the bedrock at a depth of approximately 50 m and has been in operation for 20 years. According to present plans, all Swedish nuclear power plants will have been decommissioned by the end of the 2040s. The decommissioning phase will start by initially decommission the two Barseback reactors that have been non-operational since 1999 and 2005. In order to host the decommissioning waste, the storage volume of the SFR repository must be extended with approximately 140,000 m3. To do this, a project is ongoing with the objective of submitting an application to extend SFR by 2013 and the superior aim of having the facility in operation year 2020.

Cementitious materials are widely used in waste management systems with different aims and requirements for long term performance. Both conventional and novel cementitious materials are used to create reliable immobilising elements for safe storage and disposal of wastes. The barrier elements as well as interactions envisaged between various components are important to ultimately ensure the overall safety of a storage/disposal system. The behaviour and performance of cementitious materials including waste package components, wasteform and backfilling were analysed within the IAEA Coordinated Research Project which involved 26 research organizations from 21 Member States MS). The paper presents briefly the main research outcomes for conventional cementitious systems; novel materials and technologies; testing and waste acceptance criteria; and modelling long term behaviour.

In preparation for actual repository implementation, designs that could be tailored to specific sites need to be considered. This requires a paradigm shift in design philosophy, moving away from a single reference design towards a flexible and systematic “catalogue” of engineered barrier components and associated repository infrastructure. To support this development, novel methodology for both developing and comparing repository design options for Japanese vitrified high-level radioactive waste (HLW) has been examined. This paper will outline the background of the project and, in particular, some novel approaches examined for facilitating optimisation of practical repository designs. It is complemented by a companion paper - Matsumoto et al: Application of Formal Knowledge Engineering Approaches to Develop A Design Catalogue for A Japanese HLW Repository -, which describes the repository design process in more detail and illustrates results of first dry runs of the methodology.

The volunteer siting of disposal facilities for vitrified high-level radioactive waste (HLW) and TRU waste in Japan results in a particular need for flexibility to allow repositories to be tailored to specific sites. Instead of a single reference concept, a “catalogue” of designs for individual repository components is being developed, which can then be combined to form optimum solutions for particular site boundary conditions. As highlighted in a companion paper – Makino et al: Supporting Development of Practical Designs for a Japanese HLW Repository –, which puts the repository design process in the context of the Japanese HLW programme, the complexity of this work justifies use of advanced Knowledge Engineering approaches. In this paper three components are described – development of the knowledge base, assessing tacit system understanding and production of innovative solutions to resolve conflicts between requirements.