Dissolution of a complex, simulated Magnox Waste (MW) glass in batch dissolution experiments at 60 °C over a period of 56 days is accompanied by extensive development of secondary gels. Gel development has been followed using a range of chemical, spectroscopic and physical means. Initially, a surface layer comprising (hydr)oxides of Fe, Zr and the lanthanides develops at the glass surface. Aluminosilicate gels containing Si, Al, Mg, Sr, Cs and Rb develop in systems where sufficient quantities of glass derived solutes accumulate in the leachate. These gels are hydrous and readily soluble in acidic oxalate solutions. Solution chemistry data is consistent with the development of Cs,Sr-bearing aluminosilicates, silica gel and (hydr)oxides of hydrolysis prone waste components. The experimental results are compared with the predictions of a model that considers kinetically constrained glass dissolution and the precipitation of secondary phases, including a Cs,Sr-bearing aluminosilicate gel.

Various elements (Cs, Sr, Ba, Zr, Ru, Pd, Ce, Nd, Gd, Fe, Cr, Ni, Mo and Te) in the simulated high-level radioactive wastes generated from commercial PUREX reprocessing were immobilized by a ceramic solidification using sodium zirconium phosphate, NaZr2(PO4)3 as a host matrix. The convertibility of the elements to the specific M, A and X sites in NZP crystal structure was determined with consideration of stoichiometry, charge balance and ion size of each element. Small disk samples of NZP waste form containing the elements were prepared by the sol-gel synthesis followed by calcination and compression sinteration at high temperature. The physicochemical structures such as produced phase in the waste forms and dispersion of the embedded elements in the NZP waste forms were investigated by means of XRD and SEM/EDX. Chemical behavior of the embedded elements and the limit of substitution of the NZP structure for the elements were investigated by the series of analysis. Moreover, in order to increase the limit of substitution of NZP, the effectiveness of pretreatment with heating the sol-gel products at 473–773K was experimentally investigated. The capability of embedding of NZP for the elements in HLW has been discussed.

The effects of chloride, sulfate, and nitrate on the fluoride ion local attack of Ti-Grade 7 (Ti-7: UNS R52400) were investigated. It was observed that a chloride: fluoride ratio of as high as 10 : 1 by mole was necessary to produce visible severe attack in immersion tests. Localized attack on Ti-7 was most severe at approximately 120°C, and at neutral pH. However, electrochemical studies detected that a narrowing of the passive region can occur at chloride: fluoride ratios as small as 1.1 : 1 by mole. The addition of sulfate did not significantly inhibit pitting or stress corrosion cracking of Ti-7 U-bends. However, the addition of nitrate is suspected to act as an effective inhibitor.

Organic complexes of actinides are known to occur upon interaction of high level waste glass and Boom Clay which is a potential host rock formation for disposal of high level waste in Belgium. The solubility and mobility of 237Np, one of the most critical radionuclides, can be affected by the high dissolved organic carbon content of the Boom Clay porewater through complexation with the humic substances. The influence of humic substances on the Np behaviour is considered through dissolution tests of Np-doped glasses in Boom Clay water and through fundamental study of the specific interaction between Np(IV) and the humic acids using spectroscopic techniques. High Np(IV) concentrations are found in the glass dissolution tests. These concentrations are higher than what we should expect from the solubility of Np(OH)4, the solubility limiting solid phase predicted under the reducing conditions and pH prevailing in Boom Clay. Studying the specific interaction of Np(IV) with humic acids in Boom Clay porewater, high soluble Np concentrations are also measured and two main tetravalent Np-humate species are observed by UV-Vis spectroscopy. The two species are interpreted in terms of mixed hydroxo-humate complexes, Np(OH)xHA with x = 3 or 4. These species are the most likely species that can form according to the pH working conditions. Using thermodynamic simplified approaches, high complexation constants, i.e. log β131 and log β141 respectively equal to 46 and 51.6, are calculated for these species under the Boom Clay conditions.

Comparing the spectroscopic results of the dissolution tests with the study of the interaction of Np(IV) with humic substances, we can conclude that the complexation of Np(IV) with the humic acids may occur and increases the solubility of Np(OH)4 upon interaction of a Np-doped glass and the Boom Clay porewater.

An oxidation and dissolution study has been performed on UO2 pellets containing ∼10 and ∼0.1 wt. % 238Pu, ∼10 wt. % 239Pu and on undoped UO2 to investigate the effects of radiolysis and composition on the corrosion behavior of spent fuel. The so-called alpha-doped UO2 is used to simulate the alpha-radiation field of different types of commercial LWR spent fuel after different storage times. Leaching experiments in demineralized and carbonated water at room temperature under oxidizing conditions showed that relatively high amounts of 238Pu were released. The leached surfaces were examined with X-ray Photoemission Spectroscopy (XPS), and the progressive surface oxidation was monitored. The oxidation of the U(IV) during the leaching experiments, in the materials doped with 238Pu resulted in precipitation of U(VI) phases: enhanced formation of studtite for the strongest radiation field and shoepite at low radiation field was observed on the surface of the pellet. Essentially no precipitation of Pu-rich phases was directly observed. Leaching in carbonated water and characterization of UO2 containing 239Pu under the same experimental conditions were performed and the results compared to those for alpha-doped UO2. The chemistry effects due to the presence of Pu in addition to alpha-radiolysis were investigated.

This paper describes how the results of vapor hydration tests (VHTs) are used to model the corrosion of waste glasses exposed to humid air in the glass degradation model for total system performance assessment (TSPA) calculations for the proposed Yucca Mountain disposal system. Corrosion rates measured in VHTs conducted at 125, 150, 175, and 200°C are compared with the rate equation for aqueous dissolution to determine parameter values that are applicable to glass degradation in humid air. These will be used to determine the minimum for the range and distribution of parameter values in calculations for the Yucca Mountain disposal system license application (TSPA-LA). The rate equation for glass dissolution is rate = kE • 10 η • pH • exp(–Ea/RT). Uncertainties in the calculated rate due to the range of waste glass compositions and water exposure conditions are taken into account by using a range of values for the rate coefficient kE. The parameter values for the pH dependence (η) and temperature dependence (Ea) and the upper limit for kE are being determined with other tests. Using the values of η and Ea from the site recommendation model, the VHT results described in this paper provide a value of log kE = 5.1 as the minimum value for the rate expression. This value will change slightly if different pH-and temperature-dependencies are used for the TSPA-LA model.

Molecular imprinting techniques have shown great promise for applications in chemical separations, including those involving lanthanides and actinides. This work examines the production of a selective resin for Am separations. Due to practical difficulties, resins were not imprinted with Am, rather 3 resins were created with each of Sm, Nd, and Pr as the template ions. An analogous “blank” or unimprinted resin was also created. The nitrate salt of the target metal ion was dissolved in CH2Cl2 and the resin was created around the ions to provide a unique structure based upon each metal. These resins were synthesized by a radical polymerization method, producing a reusable organic solid. The resins were qualified by obtaining values for their proton exchange capacities and data to define their complexation thermodynamics. Proton exchange capacities were determined using an indirect titration and were found to be 10.08 meq/g for the Sm-imprinted resin, 7.25 meq/g for the Nd resin, and 7.14 meq/g for the Pr resin. Data for the resins' thermodynamics were obtained at pH 1–7 in steps of 0.5 units. Results show that the templated resins rapidly removed the target actinide from aqueous solution under experimental conditions. Best separation results for Am from Eu and Gd were obtained with the Nd imprinted resin.

Layered double hydroxides (LDHs) have high anion exchange capacities and are easily synthesized in the laboratory, giving them considerable potential as adsorbents for anionic contaminants in the environment. In this study, sorption of arsenate, selenate, and perrhenate by uncalcined Mg-Al LDH (Mg:Al ratio 2.4:1) was evaluated. To investigate sorption mechanisms, the substrate was added to the solution both as a pre-wetted slurry (allowing 24 hours for rehydration), and as a dry powder. Sorption results varied for different anionic complexes. For selenate and perrhenate, an incubation time was required for the dry material to achieve the same Kd value as the pre-wetted material. For arsenate, however, both wetted and dry materials initially had high Kds, which decreased with time. These behaviors are attributed to the structural reconstruction of the calcined LDH material in solution. Se and Re are not highly sorbed by the phases present in the calcined material, but are incorporated as interlayer anions in the LDH structure upon hydration and reconstruction, while As may be sorbing onto the dehydration products, and being partially released during rehydration.

The prediction of heavy metal solubility in the presence of phosphate is the key for designing a successful environmental remediation strategy or for the performance assessment of a nuclear waste repository. In order to evaluate the possibility for solubilities and incorporation of U and its fission products, such as Cs, Np, Pu, it is important to have the Gibbs free energies of formation of possible end-member phases. In this paper, we use a linear free energy relationship to calculate the Gibbs free energies of formation of divalent cation “anhydrous” autunite phases [M2+(UO2)2(PO4)2] and monovalent cation “anhydrous” autunite phases [M+2(UO2)2(PO4)2]. Based on predicted Gibbs free energies of formation for autunite phases, an equilibrium calculation indicates that it is feasible to incorporate NpO2+ into the autunite phases. If concentration of Cs+ is high, Cs+ will be incorporated into the autunite phases. However, it is not feasible to incorporate low concentration Cs+ released from spent fuel into the uranyl phosphate phases with layered structure and interlayer water (such as autunite phases, boltwoodite phases, and uranophane phases) as previously proposed. The predicted data can be considered as a first order approximation.

A series of iron phosphate glass waste forms with compositions of 45Fe2O3-55P2O5, 20Fe2O3-80P2O5, and 20Fe2O3-20Na2O-60P2O5, namely FeP-1, FeP-2 and FeP-3, was studied by analytical electron microscopy (AEM). Transmission electron microscopy (TEM) bright-field (BF) imaging showed that under the electron irradiation, phase segregation occurred in both the FeP-1 and FeP-2 samples at high electron doses (3.84×1026 e/m2). In contrast, bubbles formed in the FeP-3 sample, even at a relatively low dose (2.88×1025 e/m2), which may be attributed to the migration of Na under irradiation as in the case in sodium borosilicate glass. Series electron energy-loss spectroscopy (EELS) analysis showed that the glass materials experienced mass-loss and composition variation. No obvious Fe valence state changes under irradiation were observed within the irradiation period.

Common assumptions to extrapolate the lifetime of proposed high-level waste disposal containers made of Ni-Cr-Mo alloys, in the absence of environmental and electrochemical conditions leading to localized corrosion and stress corrosion cracking, are evaluated based on a mechanistic model for passive dissolution. The predominant charge conduction mechanism through the oxide film formed on Ni-Cr-Mo alloys is hypothesized to be interstitial transport of metal cations. Dissolution of the alloy and conduction of interstitial species through the film create vacancies in the alloy. The anodic current density under potentiostatic control decreases as a function of time, and potentiostatic decays in the current density are rationalized on the basis of vacancy accumulation at the metal-oxide interface. It is concluded that the dissolution process is regulated by vacancy-enhanced diffusion of the elements in the alloy. Long-term stoichiometric dissolution arises if the diffusion coefficients of the alloying elements are similar. No credible scenario is envisioned by which catastrophic failure may occur as a result of longterm passive dissolution.

Short-term static tests were conducted with a surrogate high-level waste glass to measure the effects of pH and dissolved iron on the glass dissolution rate. The tests were conducted to determine if a term to account for the effects of dissolved iron is needed in the glass degradation model developed for Total System Performance Assessment (TSPA) calculations for the Yucca Mountain disposal system license application. The glass degradation model includes terms for dependencies on temperature, pH, and chemical affinity. A series of tests was conducted at 90 °C in various pH solutions without iron and with added FeCl3, Fe2O3, Fe3O4, and FeOOH. Tests were conducted at glass surface area/solution volume (S/V) ratios about 2 and 10 m-1 for between 2 and 21 days. Solution concentrations of boron were used to measure the extent of reaction and calculate the glass dissolution rates. Similar rates were measured in tests conducted with and without iron at each pH. Both the results of the tests with and without iron showed V-shaped pH dependence curves with minima at near-neutral pH values. The pH dependencies (η) are about 0.44 in basic solutions and –0.49 in acidic solutions, based on the combined results of tests with and without iron. These are within the range of values for the pH dependence in the TSPA model for site recommendation.

Experiments were conducted to examine the corrosion behavior of fuel and cladding under test conditions selected to determine how fuel with breached cladding may behave under unsaturated repository conditions. The results discussed here were obtained from two test samples, each consisting of ∼3.5-in. segments of ATM103, a moderate-burnup (∼30 GWd/MtU) PWR fuel, exposed to humid air at 175°C. Visual examination of the samples after 540 days revealed that each had developed an axial crack that passed through a drilled hole in the cladding and ran the full length of the sample.

Destructive examination of the fuel and cladding showed that the cladding had experienced extensive fuel-side corrosion. Metallographic examination of the cladding shows the hydride distribution near the drilled hole, the fracture surfaces, and the fuel-side corrosion products. Electron microscopy (SEM, TEM) and electron diffraction analyses were used to characterize the corrosion products. The results indicate that the fuel-side corrosion of the cladding and the specific volume increase associated with the formation of the corrosion products caused the hoop stresses that resulted in the observed axial splitting. The implications of these results for the expected evolution of the spent fuel cladding after the cladding is initially breached in an unsaturated repository are discussed.

The Yucca Mountain site was designated in July 2002 as the United States' location for a geological repository for spent nuclear fuel and other high-level radioactive wastes. This site designation was a watershed event in the history of the project, enabling the U.S. Department of Energy to seek a license from the U.S. Nuclear Regulatory Commission to construct and operate a geologic repository. Summarized below are the history and technical basis for this site designation and some key anticipated future events. Many of the significant events to date have been framed by the Nuclear Waste Policy Act (and Amendments) and the requirements of the regulatory standard.

Argonne National Laboratory has developed an electrometallurgical process for conditioning spent sodium-bonded metallic reactor fuel prior to disposal. A waste stream from this process consists of stainless steel cladding hulls that contain undissolved metal fission products such as Tc, Ru, Rh, Pd, and Ag; a small amount of undissolved actinides (U, Np, Pu) also remains with the hulls. These wastes will be immobilized in a waste form whose baseline composition is stainless steel alloyed with 15 wt% Zr (SS-15Zr). Scanning electron microscope (SEM) observations of simulated metal waste forms (SS-15Zr with up to 11 wt% actinides) show eutectic intergrowths of Fe-Zr-Cr-Ni intermetallic phases with steels. The actinide elements are almost entirely in the intermetallics, where they occur in concentrations ranging from 1–20 at%. Neutron- and electron-diffraction studies of the simulated waste forms show materials with structures similar to those of Fe2Zr and Fe23Zr6.

Dissolution experiments on simulated waste forms show that normalized release rates of U, Np, and Pu differ from each other and from release rates of other elements in the sample, and that release rates for U exceed those for any other element (including Fe). This paper uses transmission electron microscope (TEM) observations and results from energy-dispersive X-ray spectroscopy (EDX) and selected-area electron-diffraction (SAED) to characterize relationships between structural and chemical data and understand possible reasons for the observed dissolution behavior.

Transmission electron microscope observations of simulated waste form samples with compositions SS-15Zr-2Np, SS-15Zr-5U, SS-15Zr-11U-0.6Rh-0.3Tc-0.2Pd, and SS-15Zr-10Pu suggest that the major actinide-bearing phase in all of the samples has a structure similar to that of the C15 (cubic, MgCu2-type) polymorph of Fe2Zr, and that materials with this structure exhibit significant variability in chemical compositions. Material whose structure is similar to that of the C36 (dihexagonal, MgNi2-type) polymorph of Fe2Zr was also observed, and it exhibits less chemical variability than that displayed by material with the C15 structure. The TEM data also demonstrate a range of actinide concentrations in materials with the Fe23Zr6 (cubic, Mn23Th6-type) structure.

Microstructures similar to those produced during experimental deformation of Fe-10 at% Zr alloys were observed in intermetallic materials in all of the simulated waste form samples. Stacking faults and associated dislocations are common in samples with U, but rarely observed in those with Np and Pu, while twins occurred in all samples. The observed differences in dissolution behavior between samples with different actinides may be related to increased defect-assisted dissolution in samples with U.

Operational radioactive waste is generated during routine operation of nuclear power plants (NPP). This waste must be solidified in order to ensure safe conditions of storage and disposal. Vitrification of NPP operational waste is a relative new solidification option being developed for last years. The vitrification technology comprises a few stages, starting with evaporation of excess water from liquid radioactive waste, followed by batch preparation, glass melting, and ending with vitrified waste blocks and some relative small amounts of secondary waste. Application of induction high frequency cold crucible type melters facilitates the melting process and significantly reduces the generation of secondary waste. Two types of glasses were designed in order to vitrify operational waste depending on the reactor type at the NPP. For the NPP with RBMK-type reactors the glass 16.2Na2O 0.5K2O 15.5CaO 2.5 Al2O3 1.7Fe2O3 7.5B2O3 48.2SiO2 1.1 Na2SO4 1.2NaCl (5.7 others) was produced. For NPP with WWER reactors the glass 24.0Na2O 1.9K2O 6.2CaO 4.3Al2O3 1.8Fe2O3 9.0B2O3 46.8SiO2 0.8Na2SO4 0.9NaCl (4.3 others) was produced. The melting temperatures of both glass formulations were 1200–1250 C, specific power consumption was 5.2 ± 0.8 kW h/kg, 137Cs loss was within the range 3 - 4 %. The specific radioactivity of glass reached 7.0 MBq/kg. Glass blocks obtained were studied both in laboratory and field conditions. Long-term studies revealed that vitrified NPP operational waste has the minimal impact onto environment. Since the glass has excellent resistance to corrosion it gives the basic possibility of maximal simplification of engineered barrier systems in a disposal facility. The simplest disposal option for vitrified NPP waste is to locate the packages directly into earthen trenches provided the host rock has the necessary sorption and confinement properties.

Sweden plans to dispose of spent nuclear fuel in a deep geologic repository in granitic rock. The disposal conditions allow water to contact the canisters by diffusion through the surrounding bentonite clay layer. Corrosion of the canister iron insert will consume oxygen and provide actively reducing conditions in the fluid phase. Experiments with spent fuel have been done to determine the dissolution behavior of the fuel matrix and associated fission products and actinides under conditions ranging from inert atmosphere to reducing conditions in solutions. Data for U, Pu, Np, Cs, Sr, Tc, Mo, and Ru have been obtained for dissolution in a dilute NaHCO3 groundwater for 3 conditions: Ar atmosphere, H2 atmosphere, and H2 atmosphere with Fe(II) in solution. Solution concentrations forU, Pu, and Mo are all significantly lower for the conditions that include Fe(II) ions in the solutions together with H2 atmosphere, while concentrations of the other elements seem to be unaffected by the change of atmospheres or presence of Fe(II). Most of the material that initially dissolved from the fuel has reprecipitated back onto the fuel surface. Very little material was recovered from rinsing and acid stripping of the reaction vessels.

Plutonium, well known for its unusual nuclear and material properties, has remained an unsolved problem for over a half century. Due to the continuing issue of safe, longterm nuclear waste storage, gaining an understanding of plutonium and its compounds cannot be overstated. Self-irradiation in plutonium leads to vacancy formation [1], and our computations indicate that the electronic structure we predict for pure delta plutonium is preserved in the presence of vacancies, and that vacancies do stabilize the delta phase. A preliminary study of self-diffusive properties of plutonium, and of diffusion at the interface of plutonium and iron indicates that plutonium atoms readily diffuse across the interface with steel. This has relevance in nuclear waste storage in steel containers for assessing the depth of penetration to be dealt with in surface treatment for decontamination. The effect of thermomigration (Soret effect) in the plutonium-steel system appears to facilitate the movement of plutonium atoms into the bulk steel.

Three metallic slag samples recovered from the site of their inadvertent creation several decades after the fact were sectioned and analyzed to determine uranium speciation to evaluate environmental behavior as well as assess proliferation resistance of the waste form. Uranium concentration in the highly inhomogeneous samples was up to 5% by weight as determined by gamma spectroscopy. Sample sections were milled in a hardened steel ball mill for x-ray absorption spectroscopy analysis (XAS). Powders were digested in a mixture of heated concentrated nitric acid and peroxide and analyzed for elemental content using ICP-AES and ICP-MS. Though elemental content of the samples varied widely, high concentrations of Al were consistently found. Other metals of significance were Ti, Fe, Ni, Cu, Zn, and Sn. High Pb concentrations were occasionally found. XAS analysis revealed the samples contained primarily uranyl and another phase identified as a uranium-aluminum melt.

Speciation calculations are one of the more common ways to describe a chemical system. Speciation diagrams is a convenient way to display such calculations and may also be the basis on which experiments in neighbouring fields of interest, such as sorption are based. In earlier days only very simple problems could be described, e.g. hydroxide species dominance diagrams versus pH. With the aid of computers, however, more complicated systems such as groundwater chemistry may be used. However, as the complexity increase, so does the effect of uncertainties. In this paper we focus on the effect uncertainties in stability constants may have on speciation calculations. The common belief is that the curves will be replaced by narrow bands, while the truth is far from that. In many cases one or several species cover the entire region between 0 and 100%.