Static corrosion tests were performed with an inactive HLW glass of R7T7 type in order to evaluate alteration-phase formation and associated Cs release during the glass corrosion. The tests were performed in NaOH solutions at elevated temperatures in order to accelerate the corrosion. Crystalline phases formed in the corroded glass were analyzed by use of XRD, in addition to the solution analysis. The results indicated that; 1)Analcime or Na-beidellite or both of them form during the corrosion depending on the conditions, 2)Si rich amorphous phases are contained in the alteration-phases, 3)In addition to solution pH, solution concentrations of Na and K sensitively affect formation of analcime and Na-beidellite, 4)Analcime formation accelerates the glass corrosion, 5)Most of Cs in the glass is retained in the alteration-phases by sorption onto Na-beidellite or incorporation into amorphous phases and pollucite

The chemistry of the aqueous environment inside any breached waste packages is critical to the release of radionuclides from the waste packages for high-level waste disposal. A corrosion test cell that simulates some aspects of the internal geometry of the waste packages has been designed to investigate changes in the in-package solution chemistry. A series of tests was conducted to evaluate solution chemistry variations as a function of applied potential and temperature using a specimen of Type 316L stainless steel with a predrilled hole as a simulated pit. A micro-syringe was used to extract solutions from inside and outside the pit. The solutions were analyzed for cation concentrations using capillary electrophoresis, and the pH was measured using a micro-electrode. Preliminary measurements showed substantially high cation concentrations inside the pit due to anodic dissolution of Type 316L stainless steel. The solution pH became considerably acidic, reaching a value of 2.6. These results suggest that interactions of waste package internal structural components with incoming water may have significant influence on the evolution of water chemistry and the subsequent corrosion of waste forms such as spent nuclear fuel.

At present, crystalline ceramic based on titanate pyrochlore, (Ca,Gd,Hf,Pu,U)2Ti2O7, is considered as the US candidate waste form for the immobilization of weapons grade plutonium. Naturally occuring U-bearing minerals with pyrochlore-type structure: hatchettolite, betafite, and ellsworthite, were studied in orders to understand long-term radiation damage effects in Pu ceramic waste forms. Chemical shifts (δ) of U(Lδ1)– and U(Lβ1) – X-ray emission lines were measured by X-ray spectrometry. Calculations were performed on the basis of a two-dimensional δLá1- and δLδ1- correlation diagram. It was shown that 100% of uranium in hatchettolite and, probably, 95-100% of uranium in betafite are in the form of (UO2)2+. formal calculation shows that in ellsworthite only 20% of uranium is in the form of U4+ and 80% of the rest is in the forms of U5+ and U6+. The conversion of the initial U4+ ion originally occurring in the pyrochlore structure of natural minerals to (UO2)2+ due to metamict decay causes a significant increase in uranium mobility.

Ceramic samples with formulations Ca1-xGdxZrTi2-xAlxO7(0 ≤ x ≤1) were prepared by cold pressing in pellets at 200 MPa followed by their sintering at either 1400 or 1500 °C for 5 hours. All the samples produced were single phase with rare grains of cubic zirconia based solid solution in the samples with high Gd and Al contents (x = 0.9 and 1.0). The samples produced at 1500 °C were more homogeneous and less porous than the samples sintered at 1400 °C. From SEM/EDS data average compositions of zirconolite in the samples are very close to specified formulation (within experimental error), for example, the zirconolite formula in the ceramic with x = 1 was found to be Gd1.04Zr1.02Ti1.07Al0.85O7. XRD and TEM patterns showed monoclinic (2M) structure for zirconolites, which transforms to higher symmetry modification within 0 ≤ x ≤ 0.5 compositional range. Zirconolite with high Gd content x →1 forms two varieties with trigonal and orthorhombic structure.

The lower limit of the size of the octahedral A4+ ion in the ATi2O6 brannerite structure is just smaller than that of Ce/Pu. Attempts to expand the A ion size beyond that of Th by (a) substituting a Ba ion plus two U5+ ions for three A ions or (b) substituting one Ba plus one hexavalent ion for two A ions did not succeed. Ge, Sn and Zr substitutions in the Ti site of ThTi2O6 do not exceed 0.2 formula unit in ceramic preparations. These and other coupled substitutions in the B site of ThTi2O6 showed that the average B site size could tolerate deviations of < 1%. Ce4+ is unusually stabilised in air atmospheres at temperatures close to the melting point of 1400°C in the A site of brannerite. Lattice parameter data on different endmember ATi2O6 brannerites are given. The lower and upper size limits for the eightfold A ions in the pyrochlore structure are around 0.100 and 0.117 nm respectively. A BaUTi2O7 stoichiometry did not produce a pyrochlore structure, and when fired in either argon or air yielded a mixture of BaUTiO6, whose structure is still uncertain, plus brannerite and rutile.

Leach testing is an important tool for the investigation of chemical durability of radioactive waste forms. The data obtained from these tests allows for the selection of the most durable ceramics for the immobilization of weapons-grade plutonium. Ceramics based on cubic zirconia, (Zr,Gd,Pu)O2, doped with 10 wt.% 239Pu, zircon, (Zr,Pu)SiO4, doped with separately 5-6 and 10 wt.% 239Pu, pyrochlore, (Ca,Pu,Gd,Hf,U)2Ti2O7, doped with 10 wt.% 239Pu were studied using MCC-1 leach test. Ceramic samples were placed in test vessels with deionized water and set temperatures of 25°C and 90°C in an oven for 28 days. The leaching solution was removed and replaced with fresh deionized water after 3 and 14 days from the start of the test. Only data obtained after 14 and 28 days were used for the interpretation. No saturation of leach solution with Pu was observed in any experiments. It was found that the (without correction for porosity) normalized Pu mass loss (in g/m2 ) after 14 and 28 days was:

- for the zirconia ceramic – 2 × 10−2 to 4 × 10−2 at 90°C and 8 × 10−3 to 9 × 10−3 at 25 °C;

- for zircon doped with 5-6 wt.% of Pu – 7.0 × 10−3 to 8.2 × 10−3 at 90°C and 1.0 × 10−3 to 1.2 × 10−3 at 25 °C;

- for zircon doped with 10 wt.% of Pu – 0.2 to 0.2 at 90°C and 3.0 × 10−2 to 4.0 × 10−2 at 25 °C;

- for pyrochlore – 1.0 × 10−3 to 1.0 × 10−3 at 90°C and 2.0 × 10−3 to 3.0 × 10−3 at 25 °C.

It was shown that the high leach rate for zircon based ceramic doped with 10 wt.% Pu was caused by the presence of the separated inclusions of PuO2 phases observed in ceramic matrix. The results obtained so far allow us to conclude that ceramics based on zircon, zirconia and pyrochlore are characterized by similar chemical resistance to leaching in deionized water. Based on its low porosity, zircon may be most durable ceramic host phase for Pu.

A glass-bonded sodalite ceramic waste form (CWF) has been developed to immobilize electrorefiner salt wastes from electrometallurgical treatment of spent sodium-bonded reactor fuel for disposal. A degradation model is being developed to support qualification of the CWF for disposal in the federal high-level waste disposal system. The parameter values in the waste form degradation model were previously determined from the dissolution rates measured in MCC-1 tests conducted at 40, 70, and 90°C. The results of several series of tests that were conducted to confirm the applicability of the dissolution rate model and model parameters are presented in this paper: (1) Series of MCC-1 tests were conducted in five dilute buffer solutions in the pH range of 4.8 – 9.8 at 20°C with hot isostatic pressing (HIP) sodalite, HIP glass, and HIP CWF. The results show that the model adequately predicts the dissolution rate of these materials at 20°C. (2) Tests at 20 and 70°C with CWF made by pressureless-consolidation (PC) indicate that the model parameters extracted from the results of tests with HIP CWF can be applied to PC CWF. (3) The dissolution rates of a glass made with a composition corresponding to 80 wt. % glass and 20 wt. % sodalite were measured at 70°C to evaluate the sensitivity of the rate to the composition of binder glass in the CWF. The dissolution rates of the modified binder glass were indistinguishable from the rates of the binder glass.

This work is a first effort to link hydrogeological measurements to structural modeling by using the mobility of rare earth element (REE), U, and Th at the Olkiluoto site. REE, U and Th concentrations were measured in 5 groundwater samples located at depths varying between 132 and 446 m in three drill holes at the Olkiluoto site. The pH of groundwater samples are all about 8 and Eh(Pt) varies from –50 to –200 mV. Also the REE, U and Th concentrations of rock samples collected from core sections in the vicinity of water conducting fractures were measured to examine rock-water interactions in the system. The rock samples were cut into slices parallel to the fracture surface. An aliquot of each slice was leached with 0.5 N acid nitric to examine the readily leachable REE, U and Th fraction of each slice. The groundwater samples were normalized to each of the analyzed rock samples. Although the REE concentrations of the water samples are overall depleted in all the REEs compared to the rocks and leachates, and whole REE spectra could not be obtained for any of the groundwater samples, enrichment in the intermediate REE (IREE) with respect to the light REE (LREE) is observed.

An increasing loss of REE and U from rock matrix towards fracture surface is observed in samples at depths of 141, 159, and 466 m. Loss of U is also observed in a sample at about 246 m depth. Loss of Th is only observed at the depths of 159 and 246 m. The loss of U (readily leachable fraction) in the rock samples was found to be linked to the hydraulic conductivity in the related water-conducting fractures.

Environmental parameters that would affect the degradation of engineered materials, including waste packages and drip shields, in the potential high level nuclear waste repository at Yucca Mountain, Nevada are being characterized as part of the Yucca Mountain Project. These parameters include: temperature, relative humidity, range of water chemistry, deliquescence of salts, pH, and electrochemical potential (Eh). The likelihood of various brine compositions forming on the engineered components under repository conditions, and the implications, is discussed. Relative humidity controls the ionic strength and composition of the aqueous solutions, and hence strongly influences the corrosion processes that could occur. Studies are underway to more fully characterize the redox state of aqueous solutions in contact with engineered system components.

The proposed high-level nuclear waste repository at Yucca Mountain relies heavily on the corrosion resistance of waste packages (WP) emplaced in tunnels bored through tuffaceous rock for adequate performance during the anticipated 10,000 years regulatory period. Present WP design uses a ∼20-mm-thick outer shell of Alloy 22 as the main corrosion-resistant barrier. The operating conditions may include an initial high-temperature (>96 °C) pulse that will last approximately several hundred to a thousand years. Ti-alloy shields are envisioned to prevent water from directly dripping on the WP. However, recent findings suggest that deliquescent salts and other contaminants on the WP surfaces may cause liquid water to form there, even at high temperatures. Current performance projections predict that during the anticipated regulatory period localized corrosion modes will be unlikely and that the Alloy 22 barrier will degrade primarily by very slow uniform dissolution, essentially under passive surface conditions. A review is presented of the assumptions and experimental findings leading to those projections, as well as a discussion of findings of a recent Workshop on the challenges involved in extrapolating limited information on corrosion behavior over an extremely long service period that extends beyond the time frame of common engineering experience. Potential mechanisms for deterioration of the passive regime that may be encountered under those circumstances are discussed.

We report results of experimental studies on the behavior of Np during aqueous corrosion of unirradiated Np-bearing U oxides. Np-doped U oxides were reacted in humid air at 90°C and 150°C for several weeks within sealed stainless-steel vessels. Reacted solids were examined by scanning and transmission electron microscopies (SEM and TEM), electron energy-loss spectroscopy (EELS), and X-ray powder diffraction (XRD). Dehydrated schoepite, (UO2)O0.25-z(OH)1.5+2z (0 ≤z ≤0.15), is the predominant U(VI) compound formed in these experiments. Preliminary EELS analysis on crushed grains verify that dehydrated schoepite formed at 150°C contains up to approximately 2 wt.% Np, corresponding to a maximum Np:U molar ratio of approximately 1:40. These are maximum values because the degree to which surface-sorbed Np is present on the grains analyzed is not yet known. Crystalline NpO2 also precipitated during these experiments, and the concentration of Np in dehydrated schoepite may represent the maximum amount of Np that can be incorporated into dehydrated schoepite under the experimental conditions.

One promising host for actinide wastes is garnet-type phases of general formula AVIII3BVI 2[XO4]3. To determine the isomorphic capacity of garnet for uranium, the CaO – Fe2O3 – Al2O3 – SiO2 – ZrO2 – Gd2O3 – UO2 system was studied. Experiments were performed in air medium at 1400 – 1500 °C and 1 atm. The garnets have high capacity for Gd and Zr, while incorporation of U was found to be greatly dependent on the phase composition. Uranium content decreased from 18 wt.% in Ca-Zr-Fe garnet to 0.6 wt.% in Si-doped phases. Heavy ion irradiation (1.0 MeV Kr++) experiments were carried out for a garnet with maximal U content, (Ca2.7U0.3)VIII(Zr1.7Fe0.3)VI(Al1.1Fe1.9)IVO12. Amorphization dose of the phase was equal to 1.63×1014 ions/cm2 that is close to the other actinide hosts, such as pyrochlore Gd2Ti2O7.

Under the geological disposal conditions, spent nuclear fuel (SNF) is expected to evolve during the first thousands years while being maintained isolated from the biosphere before water comes in. Under those circumstances, several driving forces would lead to the progressive intrinsic SNF transformations within the rod which would basically modify the physical and chemical state of the fuel and the subsequent release of radionuclides in solution. In this paper, we briefly summarize the mechanisms we estimate to be significant and propose a new framework for the quantitative assessment of the radionuclide (RN) inventory we estimate to be associated to the classically referred to “Instant Release Fraction” (IRF). We hence demonstrate that in this framework, significantly high IRF values have to be expected for the long term due mainly to the presence of athermal diffusion processes.

Zirconates and titanates, based on the nominal baseline composition developed for the Plutonium Immobilization Project (PIP), have been prepared with and without process impurities. The titanates form pyrochlore as the major phase and the zirconates form a defectfluorite. Little, if any, of each impurity is accommodated in the defect-fluorite and powellite, kimzeyite, a spinel and a silicate glass appear as extra phases in this ceramic. In the titanates the pyrochlore incorporates more impurities, with the remainder being accomodated in zirconolite and a small amount of silicate glass. At extremly high levels of impurities, traces of magnetoplumbite, perovskite, and loveringite were found. The defect-fluorite zirconate phase is more radiation damage resistant than the titanate pyrochlore, though the secondary phases in the zirconate will reduce the radiation damage resistance of zirconate monoliths. To produce a dense product the oxide-route zirconate required sintering temperatures of about 1550°C, 200°C higher than that required for the titanate. Silicate impurities reduce the sintering temperatures.

Carbon steel is considered in Japan the most promising candidate material for overpacks in high-level radioactive waste disposal. Effects of bicarbonate solutions on the corrosion behavior and corrosion products of carbon steel were investigated by electrochemical measurements; FT-IR spectra and XRD pattern analyses. The results of the anodic polarization measurements showed that bicarbonate (HCO3) accelerates the anodic dissolution and the outer layer film formation of carbon steel in the case of high concentrations, whereas it inhibits these processes in the case of low concentrations. The FTIR and XRD analyses of the anodized film showed that siderite (FeCO3) was formed in 0.5 to 1.0mol/L bicarbonate solution, and Fe2(OH)2CO3 in 0.1 to 0.2mol/L bicarbonate solution, while Fe6(OH)12CO3 was formed in 0.02 to 0.05mol/L bicarbonate solution. In all cases the pH value was around 8.3. The stability of these chemical compositions was discussed using a potential – pH diagram for the Fe-H2O-CO2 system.

Possible long term corrosion scenarios for the engineered barriers proposed for the Yucca Mountain (Nevada, USA) repository are reviewed.

A caustic-side solvent extraction (CSSX) process to remove cesium from Savannah River Site (SRS) high-level waste has been developed through a joint program with Oak Ridge National Laboratory (ORNL), the Savannah River Technical Center (SRTC), and Argonne National Laboratory (ANL). The CSSX solvent consists of four components: (1) an extractant, a calixarene crown, calix[4]arene-bis(tert-octylbenzo-crown-6) designated BOBCalixC6, (2) a modifier, an alkyl aryl polyether, 1-(2,2,3,3,-tetrafluoropropoxy)-3-(4-sec-butylphenoxy)-2-propanol, also called Cs-7SB, (3) a suppressant, an alkyl amine, trioctylamine (TOA), and (4) a diluent, Isopar®L. The solvent composition is 0.01 M BOBCalixC6, 0.50 M Cs-7SB, and 0.001 M TOA in Isopar®L. In this program we have developed and demonstrated a flowsheet that can be used to process SRS tank waste. To this end, a series of flowsheet tests were completed using simulated waste in a 2-cm centrifugal contactor at ANL. Three short-term (3-4 hours) tests were completed to demonstrate various aspects of the flowsheet. These tests were followed by a 71-h test where the solvent was recycled 42 times. In each case, we met or exceeded the key process goals: (1) cesium removal from the waste with a decontamination factor greater than 40,000, (2) concentration of cesium in the aqueous strip effluent by a factor of 15 using dilute nitric acid, and (3) stripping the solvent sufficiently to allow it to be recycled many times. The results from the 71-h test are discussed.

The kinetics of vapor phase hydration (VHT) of high-sodium waste glasses often exhibits incubation, fast-rising, and final stages. The incubation stage is characterized by the slow and gradual appearance of transformed regions on the surface of the glass, signaling the initiation of the phase transformation. This is followed by a fast-rising stage, during which most of the transformation takes place. The final stage is characterized by the slower alteration of the remaining glass phase, which can have a duration orders of magnitude longer than those of the previous stages. While the Avrami equation provides a good representation of these features, the origin of the late-stage slow-down in the Avrami model is unrealistic for the VHT process. Computer simulations based on cell models have been used to investigate possible mechanistic origins. The models consider the formation of multiple alteration phases and simulate the complex interplay among the several microscopic mechanisms involved in the hydration process. The models assume that the transformation from the original pristine glass to the final stable altered phase occurs in two steps: from glass to a metastable phase, and then from a metastable to a stable phase. The results indicate that the late-stage slow-down can be reproduced as a result of the build-up of species that are incompatible with the stable phase or by the occlusion of the surface due to the growth of a surface phase that is less permeable to water. The simulation results are in good general agreement with the features of the experimental data.

Cerium-actinide bearing natural minerals which demonstrate their long-time physicochemical durability under the environment effect would be considered as analogues of actinide ceramic waste forms. Radiation damage of crystalline materials causes oxidation of cerium from initial Ce(III) to Ce(IV). Therefore, cerium valence state in actinide-cerium bearing natural minerals in some cases reflects the resistance of such minerals to radiation damage. Cerium valence state was determined in the following natural minerals of similar age and similar U-Th-contents: monazite (four samples), britholite (two samples), and aeschynite (one sample). The method of chemical shifts of X-ray emission (CeKá1 line) was used. The following contents of Ce(IV) were observed: more then 30 % in britholite, 11 % in aeschynite, 0 % in monazite. The results obtained suggest that durability of these actinide host phases with respect to radiation damage decreases in the monazite-aeschynite-britholite series.

Methods of liquid radioactive waste (LRW) decontamination from radionuclides including their co-precipitation at specific conditions or adsorption on selective sorption materials are well known and extensively used in LRW management technologies. At the same time, it was shown in a number of papers that some forms of organic and inorganic ionexchangers react with solutions containing specific components that results in formation of virtually insoluble precipitates inside the sorbent matrix or on its surface. Here in some cases the sorbent selectivity to some radionuclides increases substantially.

The sorption-reagent materials synthesized for decontamination purposes are the most highly selective in regard to such difficult to remove radionuclides as strontium-90 and cobalt-60. It was shown by comparative analysis of radionuclide removal efficiency by traditional selective sorbents and developed sorption-reagent materials that the latter have the highest distribution coefficients in systems too complex for “pure” sorption/ion-exchange decontamination. For example, the sorption-reagent materials have strontium distribution coefficients several dozens higher than those of commercially available sodium titanates and silicotitanates.

The mechanism of radionuclide sorption on sorption-reagent materials of different types was studied. It was shown that this is a multi-stage process and the course of different stages of chemical reactions and sorption is determined by the parameters of medium from which radionuclides are sorbed. It was also shown that these materials application in liquid radioactive waste management enables one to develop simple technological setups combining the advantages of sorption and precipitation methods.

One of the main fields of the sorption-reagent materials application can be decontamination of high-salinity radioactive waste formed whether as a result of ionexchanger filters regeneration in LRW management systems or in reverse osmosis installations. Use of sorption-reagent materials for high-salinity waste management enables one to reduce several ten-fold or even hundred-fold the volume of solid radioactive waste (SRW) to be sent for final disposal and, therefore, to decrease the cost of LRW management. The approach consisting in combined application of reverse-osmosis and sorption-reagent methods for LRW decontamination is suggested.