Alloy 22 belongs to the Ni-Cr-Mo family and it is highly resistant to general and localized corrosion. It may suffer crevice corrosion in aggressive environmental conditions. This alloy has been considered as a corrosion-resistant barrier for high-level nuclear waste containers. It is assumed that localized corrosion may occurs when the corrosion potential (ECORR) is equal or higher than the crevice corrosion repassivation potential (ER,CREV). The latter is measured by means of different electrochemical techniques using artificially creviced specimens. These techniques include cyclic potentiodynamic polarization (CPP) curves, Tsujikawa-Hisamatsu electrochemical (THE) method or other non-standard methods, such as the PD-GS-PD technique.

The aim of the present work was to determine reliable critical or protection potentials for crevice corrosion of Alloy 22 in pure chloride solutions at 90°C. Conservative methodologies (which include extended potentiostatic steps) were applied for determining protection potentials below which crevice corrosion cannot initiate and propagate. Results from PD-GS-PD technique were compared with those from these methodologies in order to assess their reliability. Results from the CPP and the THE methods were also considered for comparison. The repassivation potential resulting from the PD-GS-PD technique was conservative and reproducible, and it did not depend on the amount of previous crevice corrosion propagation.

Assessing the long-term behavior of nuclear glass implies the prediction of their long-term performance, and more precisely of their evolution under irradiation and during interaction with water. After briefly recalling the major characteristics of the local and medium-range structure of borosilicate glasses of nuclear interest, we will present some structural features observed under forcing conditions. Specific structural tools (EXAFS/XANES, Neutron/x-ray diffraction, solid state spectroscopic methods…) are correlated with numerical simulations to determine the local structure of glass and provide selective information on glass surface using total electron yield detection. During alteration in near- or under-saturated conditions, some elements such as Fe change coordination, as other elements such as Zr only suffer structural modifications in under-saturated conditions. These structural modifications may explain the chemical dependence of the initial alteration rate and the transition to the residual regime. They also illustrate the molecular-scale origin of the processes at the origin of the glass-to-gel transformation. Molecular scale processes help in predicting the properties of new generations of nuclear glasses required by future production of nuclear energy. Under irradiation, various structural effects are observed, including coordination change, ion migration or disorder effects. These studies show that glasses with a simplified composition do not show the same behavior as more realistic glasses. Molecular dynamics (MD) simulations provide complementary information on elastic effects. Recent direct evidence for B-coordination change under external irradiation together with structural models derived from MD sheds light on the structural mechanisms at the origin of radiation-induced modifications of glass properties, emphasizing the importance of the thermal regime in the cascade core.

Iron remains show the corrosion behavior of metal materials over a long term while buried in the soil. The data provide useful information in the study of the stability of overpack (carbon steel) under geological disposal conditions. And they also provide important information for the conservation science of important items of cultural heritage. There are two major microbial influences on the metal material surface : corrosion caused by microbes (microbially influenced corrosion, MIC) ; and mineralization by microbes (biomineralization). To observe these two roles, an iron reducing bacteria was cultured in a liquid medium with carbon steel and detection of corroded products were carried out by XRD method in this study.

Iron reducing bacteria was cultured under static conditions for 41 days with carbon steel. The result showed that a complex (biofilm, bacteria, etc.) was generated by the bacteria and covered the surface of the carbon steel. By using a microscope, the corrosion product was revealed to be formed of green and white crystal, or needle-shaped product and lozenge crystal by SEM observation. The green crystal was vivianite (Fe2+3(PO4)2・8H2O) measuring 50˜250 μm. In a corrosion process of an iron material surface, iron ion Fe2+ is dissolved from the iron in a cathode reaction, and generates Fe3+ oxide as corrosion product. It appears that vivianite can be also generated as corrosion product in an environment rich in Fe2+ and phosphate by activity of an iron reducing bacteria. Some data on the morphological feature of these corrosion products were obtained.

Zirconium is an abundant element in nuclear wastes. In this paper, we present structural and crystallization results for a simplified glass composition belonging to the SiO2-Al2O3-B2O3-Na2O-CaO-ZrO2-RE2O3 system (RE = Nd or La) developed to immobilize highly concentrated waste solutions. The effect of varying ZrO2 content on the structure and the crystallization tendency of this glass was studied using a multi-spectroscopic approach. Zr was shown to be located in six-fold coordinated sites whose charge compensation seems preferentially insured by Na+ cations. Whereas a significant decrease of the proportion of BO4 units was observed with ZrO2 content, no effect was detected on the environment of AlO4 units. However, a significant structural evolution of the silicate network occurred due to the formation of Si-O-Zr bonds. Whatever ZrO2 concentration, the crystallization of only a rare earth silicate apatite phase was observed during either slow cooling from the melt or isothermal heat treatment. Whereas nucleation mainly occurred from the surface of the glass without ZrO2, the introduction of zirconium induced apatite crystallization in the bulk. It is proposed that this nucleating effect of ZrO2 is mainly due to changes induced in the neighborhood of Nd3+ cations in glass structure.

The Japan Atomic Energy Agency (JAEA) established the Horonobe Underground Research Laboratory (URL) Project at Horonobe, in Hokkaido, Japan to enhance reliability of nuclear waste disposal technologies to be developed in deep sedimentary environments. JAEA has undertaken a number of in-situ experiments to determine changes in the properties of the host rock and the extent of the excavation disturbed zone (EDZ) created by the excavation of underground galleries for the disposal of radioactive waste. This paper reports a seismic tomography survey (using a hammer seismic source) of the “140m Gallery” at a depth of 140m below the surface of the Horonobe URL. The observation area was 3m square on the horizontal plane along the sidewall of the 140m Gallery. The measurement was repeated with the progress of excavation of a tunnel. In this experiment, the distribution of seismic velocity in the rock around the new tunnel and its decrease as the tunnel was dug, were observed using a simple small-scale seismic tomography system. The data collected show that this system can be used to capture the EDZ around tunnels.

Carbon steel overpack will corrode by consuming oxygen introduced during repository construction after closure of repository, that will keep the environment in the vicinity of repository reducing. The iron corrosion products can migrate in bentonite as ferrous cations (Fe2+) through the interlayer of montmorillonite replacing the exchangeable sodium ions in the interlayer. This replacement of sodium may affect the migration behavior in the altered bentonite not only for redox-sensitive elements but also the other ions. Therefore we have carried out electrochemical analysis, of calcium, strontium or barium with the ferrous ion supplied by anodic corrosion of iron coupons in compacted bentonite. Fifteen micro liters of tracer solution containing 8.6 M of CaCl2 or 3.0 M of SrCl2 or 1.5 M BaCl2 were sspiked on the interface between the iron coupon and bentonite, for which the dry density was in the range of 1.4 to 1.5 Mg/m3, before assembling. The iron coupons were connected as working electrodes to the potentiostat and held at a constant supplied potential between - 500 to +300 mV (vs. Ag/AgCl reference electrode) for up to 7 days. Calcium and strontium could migrate faster and deeper into the bentonite than iron in each condition, while barium could migrate slower than iron. A model using dispersion and electromigration can explain the measured profiles in the bentonite specimens. The fitted value of electromigration velocity was a function of applied electrical potential and 10 to 23 nm/s for calcium, 11 to 19 for strontium, around 4 nm/s for barium and 5 to 15 nm/s for iron, respectively. Alternatively, the fitted value of the dispersion coefficient was not a function of applied potential, and the values were 3 - 8 × 10-12 m2/s for calcium, 2 - 4 × 10-12 m2/s for strontium, 5 - 10 × 10-12m2/s for barium and 3 - 9 × 10-12 m2/s for iron, respectively.

Actinide specific chelator (che) conjugated with magnetic nanoparticles (MNPs) have been developed to separate nuclear waste in acidic conditions. Compared to the traditional nuclear waste treatments, such as solvent extraction and ion exchange, this method is a simple, compact and cost-effective process that generates minimum secondary waste. In this paper, we focus on the coating process of MNPs to achieve a combination of good acidic resistance, high chelator loading density and efficient magnetic separation. An optimized silica coating process before conjugates chelator directly onto MNPs significantly improves the acidic resistance of the MNP-che complex. Chelator loading density is significantly increased by attaching a linear polyamine polymer poly(allylamine hydrochloride) (PAH) to the surface of the MNPs using chemical and physical approaches.

The long-term selenate uptake capacity of leached cement was studied by means of a replenishment batch experiment with cement pore water (CPW) doped with selenate. The corresponding blank experiment (without Se) was also done. The systems were studied for 31 cycles (18-days each cycle) to understand the long-term selenate immobilization in leached cement. Results showed that the retention capacity of leached cement exponentially decreases with cycle evolution. Precipitation of ettringite, identified by SEM and XRD, occurred along the experiments. The characterization of the cement solid phases indicated that selenate was only retained in the precipitated ettringite.

Experimental data have been successfully modeled by assuming that selenate incorporates into the precipitating ettringite. Precipitation of ettringite is controlled by the kinetic dissolution of the initially present monocarboaluminate.

In a repository of high-level radioactive waste, radionuclides will leach from the waste glass and migrate into the surrounding bentonite after very long time. These processes occur simultaneously in the bentonite and should be evaluated to confirm the reliability of individual models and data for the performance assessment of high-level radioactive waste repository.

Previous study [1] reported the results of the in-diffusion experiments of Cs in compacted sodium bentonite (Kunigel V1®) in contact with fully radioactive waste glass for 15 to 300 days under aerobic conditions. And Cs migration was successfully interpreted using fundamental one dimensional diffusion model. However, migration of other radionuclides in fully radioactive waste glass were extremely slow because of low solubility, low effective diffusivity and high distribution ratio, especially multivalent elements of actinide and lanthanide.

In this study, the similar in-diffusion experiment reported by Ashida et al. [1] was carried out for about 15 years to evaluate the migration behavior of multivalent actinide and lanthanide elements. The bentonite was compacted into a stainless steel cell with 20 mm in diameter and 18 mm in length to produce dry densities of 0.5 and 1.0 Mg m-3saturated distilled water. The form of fully radioactive waste glass was borosilicate glass by using vitrified in CPF (Chemical Processing Facility). The glass sample was sliced into the disc with 4 mm in thickness and sandwiched by two pieces of the saturated bentonite sample in the diffusion cell under aerobic conditions. After 15 years, bentonite sample was sliced and immersed into the HNO3 to extract the radionuclides from the bentonite. Then profiles of Cs, Eu, Pu, Am and Cm in the bentonite sample were evaluated. The concentration profile of Cs in the bentonite was constant due to its high diffusivity.

The experimental concentrations of Am, Cm and Eu in contact with compacted sodium bentonite were good agreement with the solubilities calculated by thermodynamic data. On the other hand, the profiles of Am and Cm show two parts with different slopes which cannot be fitted by simple one-dimensional diffusion model considering single specie. Leaching and migration behavior of radio nuclides will be discussed based on the one-dimensional diffusion model considering other mechanism of several species.

[1] T. Ashida, et al. Migration behavior of cesium released from fully radioactive waste glass in compacted sodium bentonite. PNC Technical Report, TN8410 98-014(1998).

The waste management process of the French nuclear spent fuels is managed by a new vitrification conditioning step. High level radioactive borosilicate glasses are melted by induction in a cold crucible to accommodate a wide range of minor actinides and fission products. Among the fission products, platinoids precipitate in the form of (Pd-Te, Ru-Rh, Ru) metallic particles in the glass. The microstructure of these phases can affect the physicochemical properties of the melt during the vitrification process. To predict the thermodynamic properties of these alloys in the glass, a database is being developed on the Pd-Rh-Ru-Te system using the Calphad method. The binary (Pd-Te, Pd-Ru, Ru-Te) and ternary (Pd-Te-Ru) systems have been modelled. The Pd-Te-Ru ternary system built by extrapolation from the binaries enables to calculate ternary isothermal sections and thermodynamic properties of the platinoid phases between 873 K and 1573 K. Solidification paths are also calculated for Pd-Te alloys representative for those observed in the glasses. The phase composition at equilibrium and the Ru solubility limit in Pd-Te alloys are also calculated.

We report the synthesis and properties of two different hydrogels based on N-isopropylacrylamide/acrylic acid and copolymers of oligo-ethylene glycol methacrylates incorporating N-Aza crown ethers. Both hydrogels show rapid response to environmental stimuli and their size can be tuned by pH and temperature. Swollen states lead to high adsorption of water and high contact surface area with ions whereas in the collapsed state the material releases water and the ions not selectively retained by the polymer.

Preliminary autoradiography tests show that these materials strongly bind 90Sr and both pH and temperature can be used to fine tune binding selectivity. This results in such materials being promising candidates for use as smart scavenging agents for radioactive decontamination.

The ground state electronic structures of the actinide oxides AO, A2O3 and AO2 (A=U, Np, Pu, Am, Cm, Bk, Cf) are determined from first-principles calculations using the self-interaction corrected local spin-density approximation. Our study reveals a strong link between preferred oxidation number and degree of localization. The ionic nature of the actinide oxides emerges from the fact that those oxides where the ground state is calculated to be metallic do not exist in nature, as the corresponding delocalized f-states favour the accommodation of additional O atoms into the crystal lattice.

The Fracture-Matrix Transport (FMT) code developed at Sandia National Laboratories solves chemical equilibrium problems using the Pitzer activity coefficient model with a database containing actinide species. The code is capable of predicting actinide solubilities at 25 °C in various ionic-strength solutions from dilute groundwaters to high-ionic-strength brines. The code uses oxidation state analogies, i.e., Am(III) is used to predict solubilities of actinides in the +III oxidation state; Th(IV) is used to predict solubilities of actinides in the +IV state; Np(V) is utilized to predict solubilities of actinides in the +V state. This code has been qualified for predicting actinide solubilities for the Waste Isolation Pilot Plant (WIPP) Compliance Certification Application in 1996, and Compliance Re-Certification Applications in 2004 and 2009.

We have established revised actinide-solubility uncertainty ranges and probability distributions for Performance Assessment (PA) by comparing actinide solubilities predicted by the FMT code with solubility data in various solutions from the open literature. The literature data used in this study include solubilities in simple solutions (NaCl, NaHCO3, Na2CO3, NaClO4, KCl, K2CO3, etc.), binary mixing solutions (NaCl+NaHCO3, NaCl+Na2CO3, KCl+K2CO3, etc.), ternary mixing solutions (NaCl+Na2CO3+KCl, NaHCO3+Na2CO3+NaClO4, etc.), and multi-component synthetic brines relevant to the WIPP.

Results of microbiological and geochemical sampling in the Outokumpu deep (2.5 km) borehole are presented. The results indicate that the discharging fractures control the observed variations in the microbial populations at different depths, which evidently reflect true variations in the microbial populations in fractures of the crystalline bedrock.

Radiogenic iodine is one of the more difficult fission products to capture and immobilize during the reprocessing of spent nuclear fuel.

However, for metallic fuels reprocessed by electrometallurgical treatment, it is believed that the majority of fission-product iodine is retained during the various processing steps. Spent fuel from the Experimental Breeder Reactor II (EBR-II) at the Idaho National Laboratory (INL) is being treated by a combination of electrochemical and pyrometallurgical methods to deactivate the bond sodium of the fuel, recover uranium, and immobilize fission products for disposal. This paper discusses the progress of various strategies and experiments to confirm the expected retention of iodine during the electrometallurgical treatment of EBR-II spent fuel. This includes surveys of previous observations and measurements, and the direct measurement of iodine from various process samples. Current measurements are aimed at iodine determination in the bond sodium and plenum regions of the fuel, refined iodine measurements in electrorefiner salt, and the retention of iodine during waste form production.

Following containment failure in the scenario of geological disposal of spent nuclear fuel, the penetration rate of groundwater into the UO2 matrix could cause a rapid increase of the fraction of inventory becoming available for prompt dissolution. In this respect, oxygen and water diffusion mechanisms are key issues to investigate. In this work, secondary-ion-mass-spectrometry (SIMS) depth profiling has been applied to characterize a polycrystalline UO2 pellet exposed to 18O-labelled water at room temperature. 18O depth profiling up to 25 μm beneath the pellet surface clearly indicates a combination of oxygen diffusion into the UO2 lattice and water diffusion along grain boundaries, behaving as high diffusivity paths. The lattice diffusion coefficient of oxygen, DL , and the quantity δDB – product of the grain boundary width, δ, and the grain boundary diffusion coefficient of water, DB – have been measured, resulting in DL = (2.5 ± 0.1) × 10-24 m2 s-1 and δDB = (7.5 ± 0.3) × 10-24 m3 s-1.

In Olkiluoto Finland colloidal silica called silica sol (EKA Chemicals) will be used as a non-cementitious grout for the sealing of fractures of the hydraulic apertures of 0.05 mm or less. The use of colloidal material has to be considered in the long-term safety assessment of a spent nuclear fuel repository. The potential relevance of colloid-mediated radionuclide transport is highly dependent on their stability in different geochemical environments. Objective of this work was to study the effect of ionic strength on stability of silica colloids released from silica gel. Silica gel samples were stored in contact with NaCl and CaCl2 electrolyte solutions and in deionized water. Colloid release and stability were followed for two years by taking the samples after one month and then twice in a year. The release and stability of colloids were followed by measuring particle size, colloidal silica concentrations and zeta potential. The particle size distributions were determined applying the dynamic light scattering (DLS) method and zeta potential based on dynamic electrophoretic mobility.

In dilute NaCl (10-7–10-2 M) and CaCl2 (3 10-7– 3 10-3 M) solutions, a mean colloid diameter was less than 100 nm and high negative zeta potential values suggests the existence of stable silica colloids. After two years, the mean particle diameter was increased but it was still less than 500 nm and absolute value of zeta potential was decreased. In 0.1–1 M NaCl and 0.03–3 M CaCl2 solutions, wide particle size distribution and zeta potential values around zero suggested particle aggregation and instable colloids. In deionized water, particle size remained rather stable and zeta potential remained high negative suggests stable silica colloids. The threshold value of ionic strength was 0.03–0.1 M when salinity had an effect on the stability of colloids. In Olkiluoto, the ionic strength of saline groundwater is order of magnitude higher than the range of effect value obtained in this study. Under the prevailing conditions in Olkiluoto, silica colloids are instable, but the possible influence of glacial melt waters has to be considered.

The safe management and disposition of used nuclear fuel and/or high level nuclear waste is a fundamental aspect of the nuclear fuel cycle. The United States currently utilizes a once-through fuel cycle where used nuclear fuel is stored on-site in either wet pools or in dry storage systems with ultimate disposal in a deep mined geologic repository envisioned. However, a decision not to use the proposed Yucca Mountain Repository will result in longer interim storage at reactor sites than previously planned. In addition, alternatives to the once-through fuel cycle are being considered and a variety of options are being explored under the U.S. Department of Energy's Fuel Cycle Research and Development Program.

These two factors lead to the need to develop a credible strategy for managing radioactive wastes from any future nuclear fuel cycle in order to provide acceptable disposition pathways for all wastes regardless of transmutation system technology, fuel reprocessing scheme(s), and/or the selected fuel cycle. These disposition paths will involve both the storing of radioactive material for some period of time and the ultimate disposal of radioactive waste.

To address the challenges associated with waste management, the DOE Office of Nuclear Energy established the Used Fuel Disposition Campaign within its Fuel Cycle Research and Development Program in the summer of 2009. The mission of the Used Fuel Disposition Campaign is to identify alternatives and conduct scientific research and technology development to enable storage and disposal of used nuclear fuel and wastes generated by existing and future nuclear fuel cycles. The near-and long-term objectives of the Fuel Cycle Research and Development Program and it's Used Fuel Disposition Campaign are presented.

The suitability of Portland cement blends for encapsulation of Cs-Ionsiv in a monolithic wasteform was investigated. No evidence of reaction or dissolution of the Cs-Ionsiv in the cementitious environment was found by scanning electron microscopy and X-ray diffraction. However, a small fraction (≤1.6 wt%) of the Cs inventory was released from the encapsulated Ionsiv during leaching experiments carried out on hydrated samples. Cs release was enhanced by exchange of K and Na present in the cementitious pore water. Cement systems lower in K and Na, such as slag based blends, showed lower Cs release than the fly ash based analogues.

Trapping experiments have been performed at the Idaho National Laboratory to assess the performance of AgX sorbent media in capturing volatile iodine during the oxidation of irradiated oxide fuel. The demonstration of iodine release and capture from the used fuel has been accomplished with laboratory-scale equipment in a hot cell environment. Iodine loadings as high as 6 ug/g media have been achieved via chemical adsorption with filter efficiencies in excess of 90%. In addition to iodine, significant quantities of tritium have also been collected on the AgX filter media. Filter media loaded with radioactive iodine has been sequestered in a tin matrix by hot isostatic pressing at 200°C. The placement and encapsulation of the sorbent media was examined by neutron radiography, thus confirming the sequestration of radioactive iodine.