The release of radionuclides from spent nuclear fuel in contact with water is controlled by two processes – the dissolution of the UO2 grains and the rapid release of fission products segregated either to the gap between the fuel and the cladding or to the UO2 grain boundaries. The rapid release is often referred to as the Instant Release Fraction (IRF) and is of interest for the safety assessment of geological repositories for spent fuel due to the potential dose contribution.

Previous studies have shown that the instant release fraction can be correlated to the fission gas release (FGR) from the spent fuel. Studies comparing results from samples in the form of pellets, fragments, powders and a fuel rodlet have shown that the sample preparation has a significant impact on the instant release, indicating that the differentiation between gap release and grain boundary release should be further explored.

Today, there are trends towards power uprates, longer fuel cycles and increasing burn-up putting additional requirements on the nuclear fuel. These requirements are met by the development of new fuel types, such as UO2 fuels containing dopants or additives. The additives and dopants affect fuel properties such as grain size and fission gas release. In the present study we have performed experimental leaching studies using two high burnup fuels with and without additives/dopants and compared the fuel types with respect to their instant release behavior. The results of the leaching of the samples for the 3 initial contact periods; 1, 7 and 23 days are reported here.

In this investigation, CeO2 analogues, which approximate as closely as possible the characteristics of fuel-grade UO2, were characterised after dissolution under a wide range of conditions. Powdered samples were subject to a range of aggressive and environmentally relevant alteration media with different solubility controls, and reacted at 70 °C and 90 °C. Dissolution kinetics were monitored through analysis of the coexisting aqueous solution. Monolith samples were monitored for development of surface defects such as pores and dissolution pits, in addition to morphological changes at grain boundaries and surface pores upon dissolution under aggressive conditions. The surfaces were analysed using confocal profilometry, vertical scanning interferometry and scanning electron microscopy. Dissolution rates were found to be greatest in low pH solutions and at higher temperatures. Preferential dissolution appears to occur at grain boundaries and on particular grains, suggesting a crystallographic control on dissolution.

The report presents the results of research and development on cementation technology of liquid radioactive waste (LRW) generated in spent nuclear fuel reprocessing in Experimental Demonstration Center (EDC) at the Mountain Chemical Combine, Russian Federation.

Bentonite is planned to be used in many countries as an important barrier in high-level waste repositories. Assessment of the barrier with regard to, inter alia, its ability to hinder transport of dissolved radionuclides leaking from a damaged canister containing spent nuclear fuel, requires quantitative data about the pore structure inside bentonite. The present NMR study was made in order to determine the number of distinguishable porewater phases in compacted water-saturated samples of MX-80 bentonite and Na-montmorillonite. The samples were compacted to dry densities in the interval 0.7-1.6 g/cm3 and subsequently saturated with Milli-Q water or 0.1 M NaCl solution in equilibrium cells. The NMR measurements were performed with a high-field 270 MHz NMR spectrometer using a short inter-pulse CPMG method to study proton T1ρ relaxation. The measured relaxation curves were found to consist of one faster and one slower proton relaxation. Subsequent analysis of the data indicated that the faster relaxation was associated with interlayer (IL) water between montmorillonite unit layers, while the slower one was associated with non-interlayer (non-IL) water located outside the interlayer spaces. The results indicate for compacted samples with a dry density of ≥ 1.0 g/cm3, that Na montmorillonite contains a larger relative volume of non-IL water than the corresponding MX-80 bentonite. This in turn, suggests that the stacking number in Na-montmorillonite is smaller than in MX-80 bentonite. Changing the porewater chemistry seemed to have some effect on the non-IL water content in the Na montmorillonite but not in the MX-80 bentonite.

Corrosion behavior is a key issue for the waste disposal of irradiated metals, such as hulls and endpieces, and is considered to be a leaching source of radionuclides including C-14. However, little information about Zircaloy corrosion in anticorrosive conditions has been provided.

In the present study, long-term corrosion tests of Zircaloy-4 and Zircaloy-2 were performed in assumed disposal conditions (dilute NaOH solution, pH 12.5, 303 K) by using the gas flow system for 1500 days. The corrosion rate, which was determined by measuring gaseous hydrogen and the hydrogen absorbed in Zircaloy, decreased with immersion time and was lower than the value of 2×10−2 μm/y used in performance assessment (1500-day values: 5.84×10−3 and 5.66×10−3 μm/y for Zircaloy-4, 1000-day values: 8.81×10−3 μm/y for Zircaloy-2). The difference in corrosion behavior between Zircaloy 4 and Zircaloy-2 was negligible. The average values of the hydrogen absorption ratios for Zircaloy-4 and Zircaloy-2 during corrosion were 91% and 94%, respectively.

The hydrogen generation kinetics of both gas evolution and absorption into metal can be shown by a parabolic curve. This result indicates that the diffusion process controls the Zircaloy corrosion in the early corrosion stage of the present study, and that the thickness of the oxide film in this stage is limited to approximately 25 nm and may therefore be in the form of dense tetragonal zirconia.

If groundwater enters a damaged canister and comes in contact with the spent fuel, radionuclides are released into the water in the void inside the canister when fuel dissolves. Solubility limits restrict the amount of radioelements that may migrate with the water flowing from the canister. In this study the impact of variability in groundwater chemistry compositions and the impact of uncertainties in thermodynamic data on solubility limits for Np, Pb, Pu, Ra, Se, Th, U and Zr were looked into. The solubility limits for all the studied radioelements seemed to be more sensitive to uncertainties in thermodynamic data than to differences in groundwater chemistry. The sole exception was radium, where variability in water composition has a somewhat larger impact. Radium is also the most safety critical element in the safety assessment SR-Site and groundwater compositions are expected to vary during the assessment period of one million years.

In France, nuclear glass canisters arising from spent fuel reprocessing are expected to be disposed in a deep geological repository using a multi-barrier concept (glass/canister/steel overpack and claystone). In this context, glass - iron or corrosion products interactions were investigated in a clayey environment to better understand the mechanisms and driving forces controlling the glass alteration. Integrated experiments involving glass - metallic iron or magnetite - clay stacks were run at laboratory scale in anoxic conditions for two years. The interfaces were characterized by a multiscale approach using SEM, TEM, EDX and STXM at the SLS Synchrotron. Characterization of glass alteration patterns on cross sections revealed various morphologies or microstructures and an increase of the glass alteration with the proximity between the glass and the source of iron (metallic iron or magnetite) due to the consumption of the silica coming from the glass alteration. In case of magnetite, the silica consumption is mainly driven by a sorption of silica onto the magnetite. For experiments containing metallic iron, the silica consumption seems to be strongly driven by silicates precipitation including Fe and Fe/Mg when the Fe is not enough available. Moreover, in addition to Fe-silicates observed at the surface of the gel layers, iron is incorporated within the gel probably as nanosized precipitates of Fe-silicates which could affect its physical and chemical properties. Those results highlighted the impact of the distance between glass and iron source and the nature of the iron source which drive the process consuming the silica coming from the glass alteration.

Bioapatite, found in vertebrate bones and teeth, is highly reactive and may incorporate high concentrations of some radionuclides, including U, Pu, and Sr. Therefore, bioapatite may be useful in backfill or overpack materials in nuclear waste repositories. The dissolution rate for bioapatite is constant at pH > 4 and is about 5 times faster than fluorapatite. In terrestrial environments, bioapatite recrystallizes over periods of up to ca. 40 ka.

The organization of the fuel cycle is a legacy of World War II and the cold war. Fuel cycle facilities were developed and deployed without consideration of the waste management implications. This led to the fuel cycle model of an isolated single-purpose geological repository for disposal of wastes shipped from distant processing facilities. There is an alternative: collocation and integration of reprocessing and other backend facilities with the repository. Such an option alters waste form functional requirements by reducing storage and transport requirements. This, in turn, broadens the choice of waste forms by relaxing the incentives to minimize waste volumes. Waste forms can be chosen primarily on meeting two goals: repository performance and minimizing costs. Less restrictive waste volume constraints enable termination of safeguards on all wastes, enable use of solubility-limited waste forms, and reduce radiation damage as a waste form limitation. The implications of such changes in waste form requirements are discussed.

An overview of selection criteria for waste processing and disposal technologies is given. A systematic approach for selection of an optimal technology is proposed. Optimal selection of a technical processing and disposal option is case specific to the waste management needs. Waste streams considered are from nuclear applications, research, power generation, nuclear fuel cycle activities and decommissioning of nuclear facilities as well as for NORM-containing waste.

The evaluation of the redox conditions in the Swedish ILW-LLW repository, SFR-1, is of high relevance in the performance assessment. The SFR-1 repository contains heterogeneous types of wastes, of different activity levels and with different materials in the waste and in the matrices and packaging. Steel and concrete-based materials are ubiquitous in the repository. The assessment presented in this work is based on the evaluation of the redox conditions and of the reducing capacity in 15 individual and representative waste package types in SFR-1. A combination of the individual models is used to determine the redox evolution of the different vaults in the repository. The results of the model indicate that in the initial time after repository closure, O2 is consumed through degradation of organic matter and metal corrosion during the initial time after repository closure. Afterwards, the system is kept under reducing conditions for long time periods, and H2(g) is generated due to the anoxic corrosion of steel forming magnetite as main corrosion product. The time at which steel is depleted varies with the amount and characteristics of steel and ranges from 5,000 to over 60,000 years. After complete steel corrosion, the reducing capacity of the system is mainly given by magnetite. The calculated redox potential under the chemical conditions imposed by the massive amounts of cements in the repository is in the order of -0.75 V (at pH 12.5). In case of assuming that the Eh of the system is controlled by the interaction between Fe(III)/Magnetite as a result of groundwater/magnetite interactions, redox potentials in the range -0.7 to -0.01V are calculated, considering the uncertainty in the pH prevalent in the system If the absence of oxic disturbances the Eh of the repository system would be kept reducing. In the event of oxidising and diluted glacial meltwater intrusion, magnetite would gradually convert into Fe(III) oxides, buffering the redox potential of the system and preventing it from oxidation for long time periods.

The sand of glaciolacustrine origin is offered as a major component of the backfill in repositories for solid or solidified low and intermediate level waste (LILW). Clinoptilolite, hematite, and magnesium oxide are offered as additives for increasing of sorption. In this work was carried investigation of sorption properties of sand, clinoptilolite, hematite and magnesium oxide and mixtures of sand with these mineral additives.

The main hydrate of cement is calcium silicate hydrate (CSH). Such a cement-based material is essential for constructing the geological disposal system of TRU radioactive wastes including I-129 in Japan. So far, the sorption behavior of iodine on CSH gel has been examined by using the CSH samples dried once. However, the Japan’s repository would be constructed under water table. Therefore, we must focus on also the interaction of altered cementitious material and iodine under the condition saturated with saline groundwater.

In this study, the sorption behavior of iodide ions into CSH gel, formed without dried processes, was examined in imitated saline groundwater. Ca/Si ratio was set to 0.4, 0.8, 1.2 and 1.6, and NaCl concentration of each sample also was set to 0.6 M, 0.06 M or 0.006 M. These samples were synthesized with CaO, SiO2 (fumed silica), and distilled water in a given combination of 20 ml/g in liquid/solid ratio. A NaI solution was added after curing the CSH gel (hereinafter referred to as the “Surface sorption sample”) for 7 days, setting the initial concentration of NaI to 0.5 mM in sample tube. The values of Eh and pH of each sample showed iodide ions as the chemical species of iodine in the sample tube. Furthermore, this study prepared the “Co-precipitation sample” of CSH gel with iodide ions. Here, the NaI solution was added before curing the CSH gel. For all samples, the contact time-period of the CSH gel with iodide ions was set to 7 days. After each contact time-period, each sample for analyses was separated into the solid and the liquid phases by 0.20 µm membrane filter. In the liquid phase, the concentrations of Ca, I, Si and Na ions in the liquid phase were measured by ICP-AES. Besides, the Raman spectra were obtained from the solid phases of the surface sorption sample and the co-precipitation sample without dried process.

The results showed that the sorption of iodide ions into CSH gel strongly depends on the amount of water included in the CSH gel. Such a sorption behavior was confirmed in both co-precipitation samples and the surface sorption samples, even if the Ca/Si ratio is low. This means that iodide ions can be easily immobilized through the water-molecular of CSH gel. Besides, Na concentration did not so much affect the sorption behavior of iodide ions into CSH gel. In addition, the Raman spectra showed that the degree of polymerization of SiO4 tetrahedrons in CSH gel was unaffected with increasing Na ions concentration. These results suggest that the CSH gel saturated with groundwater would retard the migration of iodide ions, even if the groundwater includes salinity.

It is necessary to assess the impact of nitrate salts and their reduction products (e.g. NH3(aq)/NH4+) contained in low-level radioactive waste generated from nuclear reprocessing process for the safety assessment of geological disposal of the waste. In the present study, sorption behavior of Ni and Pd on pumice tuff was investigated in the presence of NH3(aq)/NH4+. Under various NH3(aq)/NH4+ concentration, pH and ionic strength conditions, distribution coefficient (Kd) of Ni and Pd on pumice tuff was determined by a batch experiment. For Ni system, the Kd values showed no significant dependence on initial NH4+ concentration ([NH4+]ini < 1 M) in neutral pH region, which agreed with the prediction from thermodynamic data. For Pd system, the Kd values decreased with an increase of [NH4+]ini, suggesting the formation of stable ammine complexes (Pd(NH3)m2+ (m: 1 – 4)). The obtained Kd values for Ni and Pd were analyzed using a surface complexation model. By taking complexes predicted by thermodynamic data into account, sorption behavior of Ni and Pd in the presence of NH3(aq)/NH4+ were well explained.

The sensitive analysis of radionuclide migration for the scenarios on deterioration or loss of safety functions expected in HLW disposal system due to the human error (initial defective scenarios) is performed in this study. Release rates for Cs-135 and Se-79 are estimated from Monte Carlo-based analysis. Maximum release rates of Se-79 and Cs-135 from natural barrier in initial defective scenarios for vitrified waste and overpack are approximately equivalent to that in normal scenario on all safety function working. Maximum release rate of Se-79 in initial defective scenario of buffer under the condition of colloidal migration is about 30 times as high as that in normal scenario. Maximum release rate of Cs-135 in initial defective scenario of plugs is about two orders of magnitude higher than that in normal scenario. These results especially indicate the need to understand the feasibility on two types of initial defective scenario, leading to the loss of restraint for colloidal migration in buffer and the loss of restraint with plugs from short-circuit migration.

After the Fukushima Dai-ichi nuclear accident, Japan Atomic Energy Agency (JAEA)was chosen by the Government to conduct decontamination pilotprojects at selected sitesin the contaminated area of Fukushima. Despite tight boundary conditions in terms of timescale and resources, the projectsprovide a good basis for developing recommendations on how to assure clean-up efficiency and worker safety and reduce time, cost, subsequent waste management and environmental impact. The results of the project can be summarised in terms of site characterisation and data interpretation, clean-up and waste minimisation and storage.

The Fukushima Daiichi nuclear power station accident and restoration works have produced significant volume of radioactive waste. The waste has very different characteristics from usual radioactive waste produced in nuclear power stations and it requires extensive research and development for management of the waste. R&D works such as analysis of the waste properties, hydrogen generation by radiolysis and diffusion in a storage vessel and corrosion of storage vessels, etc. have been performed for characterization and safe storage of the waste. The detailed R&D plan for processing and disposal waste will be established by the end of FY2012.

Following the release of radionuclides into the environment as a result of the accident at Fukushima Daiichi nuclear power plant, Japan Atomic Energy Agency (JAEA) had to develop an immediate and effective method of reducing the dose rate received by students in school facilities. A demonstration of a reducing method was carried out by JAEA at a junior high school ground and kindergarten yard in the center of Fukushima-city. Dose rates of the released radionuclides are largely controlled by the ground level contamination and accumulation of mainly Cesium137 (Cs-137) and Cesium 134 (Cs-134) in populated areas. An effective means of reducing dose rate was to remove the surface soil and to bury it on-site under fresh uncontaminated soil or soil collected under deep depth at the site for shielding. The dose rate at1 m above ground level was reduced from 2.5 µSv/h to 0.15 µSv/h.

The accident of the TEPCO Fukushima Dai-ichi Nuclear Power Plant occurred by the 2011 off the Pacific coast of Tohoku Earthquake on 11 Mar. 2011. It is estimated that totally 1.2-1.5x1016 Bq for 137Cs and 1.5-1.6x1017 Bq for 131I were released until the beginning of Apr. and those radionuclides (RN) were deposited on soil surface and forest etc. widely around Fukushima Pref. This work was carried out as one of the investigations for making the distribution maps of radiation dose rate and soil contaminated by RNs which the MEXT promotes. The Geoslicer investigation on the depth distribution of RNs in soil was performed after 3 months from the accident. The investigation was conducted at 11 locations in Nihonmatsu City, Kawamata Town and Namie Town, and soil samples of depth 50 cm to 1 m were taken. Both of 134Cs and 137Cs were detected in all investigated locations, and 129mTe and 110mAg were detected only in locations where radiation dose rates are high. At many locations investigated, radiocaesium more than 99 % distributed within a depth of 10 cm in soil in the surface layer. On the other hand, RNs tended to distribute to deeper part in soil at locations that are supposed to have been used as farmland than in soil in the surface layer, and radiocaesium more than 99 % in soil at locations that are supposed to have been used as farmland also distributed within a depth of around 14 cm. The apparent diffusion coefficients (Da) of RNs derived from penetration profiles near the surface layer showed a tendency to be higher in soil at locations that are supposed to have been used as farmland (Da=0.1-1.5x10-10 m2/s) than in soil in the surface layer (Da=0.65-4.4x10-11 m2/s), and most Da-values were nearly 10-11 m2/s. The distribution coefficients (Kd) by a batch method were in the range of Kd=2,000-61,000 ml/g for Cs and Kd=0.5-140 ml/g for I. Although the Kd-values are different between cation (Cs+) and anion (I-), the Da-values (134Cs, 137Cs, 129mTe and 110mAg) were similar levels. This is considered to be due to that the Da-values were controlled by dispersion by flow of rain water.