Drip tests designed to replicate the synergistic interactions between waste glass, repository groundwater, water vapor, and sensitized 304L stainless steel in the potential Yucca Mountain Repository have been ongoing in our laboratory for over ten years. Results will be presented from three sets of these drip tests: two with actinide-doped glasses, and one with a fully-radioactive glass. Periodic sampling of these tests have revealed trends in actinide release behavior that are consistent with their entrainment in colloidal material when as-cast glass is reacted. Results from vapor hydrated glass show that initially the actinides are completely dissolved in solution, but as the reaction proceeds, the actinides become suspended in solution. Sequential filtering and alpha spectroscopy of colloid-bearing leachate solutions indicate that more than 80% of the plutonium and americium are bound to particles that are captured by a 0.1 μm filter, while less than 10% of the neptunium is stopped by a 0.1 μm filter. Analytical transmission electron microscopy has been used to examine particles from leachate solutions and to identify several actinide-bearing phases which are responsible for the majority of actinide release during glass corrosion.

Analytical round robins were established to allow laboratories responsible for the analysis of (nuclear) wastes glasses to compare their analytical capabilities and techniques with one another in a non-competitive atmosphere. In addition the quality of analyses using different analytical techniques on the same materials could be compared. Analytical Round Robin 7 had two primary objectives: 1) to evaluate the effect of using the same analytical reference glass as a reference standard for bias correcting analytical results for all the participating laboratories in making the analyses; and 2) to discuss the analytical methods and results from the participating laboratories between the participants for the purpose of evaluating quality and interlaboratory consistency. This paper will primarily address the first objective.

The purpose of an analytical reference material is to bring the analytical results from various laboratories into better agreement by providing them a common reference point No reference glass is likely to be ideal for each element in the glass being analyzed. Obviously, the closer the reference material is to the unknown, the more confidence there will be in the results because less judgement is called for. Use of the reference glass to bias correct analytical results for a standard set of analyses from several laboratories provides the data needed to determine the effectiveness of an Analytical Reference Glass (ARG) in producing interlaboratory agreement. Concurrently, whether or not the analytical results using ARG-1 are more valid or systematically biased can be evaluated using standard statistical methods. The conclusion is that bias correcting with a standard glass such as ARG-1 for results obtained by the inductively coupled plasma - atomic emission spectroscopy (ICP-AES) analytical technique produces more accurate results.

The hydration thermodynamic approach to the prediction of glass durability was originally applied to nuclear waste glasses by Jantzen and Plodinec. This approach is useful for control of the production of nuclear waste glasses. However, improvements are necessary if the approach is to be extended to different glasses, particularly those with higher alkali metal concentrations. This is of special significance for vitrification of the salt wastes at Hanford. Various methods for improving the predictive power of the approach have been examined. Combining a more accurate representation of the alkali metal species in the glass with a more rigorous thermodynamic approach is a promising avenue to improved predictive power.

Eu and Am doped glasses were interacted with synthetic interstitial clay water (SIC) and corresponding reference leachant, humic acids free interstitial solution (IS) to investigate the influence of humic acids on the leaching behaviour of the waste glass. Static leach tests were carried out at 40°C and 90°C. The release of the lanthanide Eu and the actinide Am from the glass was obviously enhanced by the presence of humic acids. The leaching of transition elements, Fe and Ti strongly depends on the humic acids concentration. The leaching of glass matrix components, Al and B was also influenced by the concentrations of humic acids. However, humic acids have little effect on the leaching of glass matrix element Si.

We present a stochastic method to model the dissolution of nuclear glass. Using this method, we solve the diffusion equation in a stochastic way. We do this by giving a large number of particles Brownian displacements. Simultaneously, these particles can participate in other processes, like a chemical reaction or convection.

We apply this method to solve the Pescatore model for the dissolution of nuclear glass in clay. This model combines diffusion of silica in the pore water of the clay with the glass dissolution rate law proposed by Grambow. We use the model for fitting the dissolution data of four glasses in clay slurries (with a high and with a low clay content) and in pure clay. We present the values of the fitting parameters. The solution of the model, obtained by the simulation method, agrees with the analytical solution. We also extend the Pescatore model with a moving boundary, taking into account the receding of the glass surface by corrosion.

The objective of the work was to evaluate the long-term capacity of sintered glass to retain high-level nuclear wastes (HLW) in near-repository conditions. We have studied the corrosion behavior of waste forms partially devitrified (43 vol.%) in different aqueous media. Devitrified samples were irradiated at doses (γ radiation from a Co 60 source) ranging from 1.4 × 106 Gy to 2.0 × 108 Gy, in order to study their aqueous corrosion resistance in simulated underground water. The results show little or no effect of irradiation on the density, microstructure and corrosion resistance. The global dissolution rate was almost constant around a value of 5×10−5 g. cm−2 d−1. Elemental dissolution rates were also unaffected by radiation.

Static corrosion tests were performed on the simulated waste glass doped with Pu and Cm (PuO2; 0.22wt%, Cm2O3; 0.09%) in deionized water at 90 °C with S/V ratio of 2600 m−1 under oxidizing and reducing conditions, respectively. The corrosion tests under oxidizing conditions were performed in air, while the corrosion tests under reducing conditions were performed in an airtight stainless steel container purged with mixed gas (Ar+5%H2), where the Eh of the solution was maintained at approximately -0.45 V vs SHE. The results showed that the redox conditions have no remarkable influence on the leaching behavior of Pu and Cm. It was also observed that the Pu and Cm concentrations in the 1.8nm filtered solutions were one or two orders of magnitude lower than those in the 450nm filtered solutions under both redox conditions, which suggests that the insoluble suspended fractions (colloidal particles) with the size from 1.8nm to 450nm are dominant phase for Pu and Cm in the solution under both redox conditions.

The V2O5-P2O5 glass system (vanadate glass) has been developed to provide low temperature vitrification of radioactive waste without volatilization of radionuclides. The dissolution behaviour of vanadate glass was determined from the results of the MCC-1 static leach test at 70 ‘C using deionized water, cement-saturated alkaline water, and cement-saturated alkaline water with 3% NaCl. The solubility of vanadate glass was remarkably reduced by the addition of Sb2O3to the glass. It was also lowered in the cement-saturated water due to formation of a precipitation layer. Although the presence of NaCl in the solution accelerated the dissolution of vanadate glass, the addition of alkaline earth metal oxide to the glass gave a dissolution rate three orders lower than non-additive vanadate glass. The results of leach tests using simulated vitrified waste form containing 5moI% of Cs2 and SrO showed there was a possibility of fixing Cs and Sr into the glass structure.

Uranium L3 X-ray absorption data were obtained from two borosilicate glasses, which are considered as models for radioactive wasteforms, both before and after leaching. Surface sensitivity to uranium speciation was attained by a novel application of simultaneous fluorescence and electron-yield detection. Changes in speciation are clearly discernible, from U(VI) in the bulk to (UO2)2+-uranyl in the corrosion layer. The uranium concentrations within the corrosion layer also show variations with leaching times that can be determined from the data.

We are characterizing the corrosion behavior of the radioactive glass that was made with sludge from Tank 51 at the Defense Waste Processing Facility (DWPF) and a nonradioactive glass having the same composition, except for the absence of radionuclides. Static dissolution tests are being conducted in a tuff groundwater solution at glass surface area/solution volume ratios (S/V) of 2000 and 20,000 m−1. These tests are being conducted to assess the relationship between the behavior of this glass in a 7-day Product Consistency Test and in long-term tests, to assess the effects of radionuclides on the glass corrosion behavior, and to measure the disposition of radionuclides that are released as the radioactive glass corrodes. The radioactive glass reacts slower than the nonradioactive glass through the longest test durations completed to date, which are 140 days for tests at 2000 m−1 and about 400 days for tests at 20,000 m−1. This is probably because radiolysis results in lower solution pH values being maintained in tests with the radioactive glass. Rate-affecting alteration phases that had formed within one year in tests with other glasses having compositions similar to the Tank 51 glass have not yet formed in tests with either glass.

The behaviour of waste glass was investigated under open site disposal conditions. This glass was produced by vitrification of intermediate level radioactive waste from nuclear power plants. Two types of borosilicate glasses were obtained for two different reactor wastes, WWER and RBM.K. Leaching and alteration mechanisms are discussed as well as the data processing technique used for these long term tests. The decay of radionuclides was accounted for in order to obtain correct results. The leaching factors obtained can be used for the assesment of radionuclide retention. Discontinuous leaching of Cs-137 has been observed during more than 8 years testing time. The fluctuating leaching rate depends on glass composition. The average leaching rate remains within (0.4 – 4) μg/sq. sm·day.

Alteration of waste glass includes the formation of surface layers and cracks on the glass surface. SEM analysis of glass was used to show these surface layers. The thickness of the layers was determined to be within 2–6 μm. The structure of these layers depends on glass composition and the interfacing environment.

A new experimental technique has been developed to test waste forms and other proposed engineered-barrier materials under unsaturated conditions. Laboratory experiments using the pressurized unsaturated flow (PUF) apparatus have been performed with a Na-Ca-Al borosilicate glass being studied for immobilization of low-activity tank wastes and a reference borosilicate glass for immobilization of high-level wastes (SRL-202). A complex coupling between glass corrosion, secondary phase precipitation, and unsaturated flow behavior was observed. Precipitation of a family of Na-Ca-Al zeolites was also found to cause an acceleration in the reaction rate of the low-activity waste glass. This same effect was observed after approximately 1 year in high solid-to-liquid ratio batch tests but after only 12 days using the PUF method at the same temperature. The onset of secondary phase precipitation was tracked by monitoring changes in volumetric water content and by inline chemical analysis. Finally, elemental release of the major glass components and solution pH were found to differ under unsaturated flow-through conditions as compared with saturated, batch tests. These findings, while preliminary, have important implications for understanding and modeling glass corrosion behavior under unsaturated conditions.

The impact of crystalline phase precipitation in glass during canister cooling on chemical durability of the waste form limits waste loading in glass, especially for vitrification of certain high-level waste (HLW) streams rich in Na2O and Al2O3. This study investigates compositional effects on nepheline precipitation in simulated Hanford HLW glasses during canister centerline cooling (CCC) heat treatment. It has been demonstrated that the nepheline primary phase field defined by the Na2O-Al2O3-SiO2 ternary system can be used as an indicator for screening HLW glass compositions that are prone to nepheline formation. Based on the CCC results, the component effects on increasing nepheline precipitation can be approximately ranked as Al2O3 > Na2O > Li2O ≈ K2O ≈ Fe2O3 > CaO > SiC2. The presence of nepheline in glass is usually detrimental to chemical durability. Using x-ray diffraction data in conjunction with a mass balance and a second-order mixture model for 7-day product consistency test (PCT) normalized B release, the effect of glass crystallization on glass durability can be predicted with an uncertainty less than 50% if the residual glass composition is within the range of the PCT model.

During the reaction of HLW glass with interstitial clay water at different temperatures with various ratios of glass surface area to solution volume (SA/V) and durations, Eu released from the glass forms predominantly Eu-humate colloids (organic colloids) by a complexation reaction. The size distribution and stability of Eu-humate colloids have been characterized. It is likely that inorganic colloids which are mainly composed of Si, Al and Ca are generated from the corrosion of waste glass by a nucleation reaction.

The solubility limit of phosphate in glass was found to decrease as fluorine increased. Amorphous phase separation was found in the glass as a result of interaction between phosphate and fluorine, in which Ce and Gd strongly partitioned. The glasses exhibiting amorphous phase separation showed higher normalized B releases according to the 31-day product consistency test (PCT) compared to glasses without phase separation. Phosphate that leached from glass into water increased the concentrations of Ce and Gd in the 7- and 31-day PCT solutions by more than an order of magnitude. The liquid state 31P-NMR results suggested that phosphate in water interacts with Ce or Gd. Therefore, the observed concentration jump for Ce and Gd in the PCT solution may be attributed to the increases in the solubility limits of Ce and Gd as a result of phosphate complexation with Ce and Gd.

The ability of NaCl and MgSO4 to impair the performance of Portland cement, blended cements containing slag and fly ash and of a permeable backfill have been measured. Performance is determined by decrease in pH, changes in mineralogy and loss of physical coherence. Experiments have been made at 25°, 55° and 85°C and extensively backed up by chemical models of cement performance. NaCl, up to 1.5M, has a comparatively slight impact on performance but MgSO4 rapidly and almost quantitatively reacts, lowering system pH's to < 10, conditioned by mixtures of Mg(OH)2 and magnesium silicates with gypsum.

Leaching of a cement paste by a demineralized solution results in the progressive dissolution of the zone in contact with the aggressive solution. In the long term, this dissolution will determine the kinetics of degradation of the material. These kinetics will depend principally on the solubility of calcium in the zone in contact with the aggressive solution.

The logical place for the US Federal government to site a permanent repository for its massive accumulation of cold-war reprocessing radwaste would be at its primary cold-war nuclear device testing reservation, the Nevada Test Site(1). Regardless of whether it eventually chooses to implement that repository by drilling lateral “drifts” into consolidated rock (e.g. its proposed Yucca Mountain facility) or by augering moderately-deep boreholes into unwelded alluvium beds (e.g. the “Greater Confinement Disposal (GCD) repository implemented at Frenchman Flats in 1984 (2), zeolitic hydroceramic materials would be more stable than would glasses. The thermodynamic rationale for this is that in such regions, soil solutions and soil gasses both tend to “weather” buried natural glasses to zeolitic materials(3). It has also been demonstrated that the same type of phases form when properly designed cementitious “grouts” are cured under mild hydrothermal conditions(4,5) - precisely those conditions assumed by DOE's repository modelers for “failed” (i.e. flooded) repositories and under which radwaste-type glasses rapidly decompose to form crystalline zeolitic phases(6,7).

The rutile form of titanium dioxide (TiO2) and granules of high density polyethylene (PEHD) and low density polyethylene (PELD) were used to prepare mortar matrices for immobilization of radioactive waste materials containing ‘137Cs. PELD,PEHD and TiO2 were added to mortar matrix preparations with the objective of improving physico-chemical characteristics of the radwaste-mortar matrix mixtures, in particular the leach-rate of the immobilized radionuclide. The diameters of the PELD and PEHD used varied from 0.2 to 2.0 mm. One type of PELD and two types of PEHD were used to replace 50 weight percent of stone granules, average diameter of 2 mm, normally used in the matrix, in order to decrease the porosity and density of the mortar matrix and to avoid segregation of the stone particles at the bottom of the immobilized radioactive waste cylindrical form. TiO2 was also added to the mortar formulation, replacing 5 and 8 weight percent of the total cement weight, for each PEHD and PELD formulation. Cured samples were investigated under temperature stress conditions, where the temperature extremes were: Tmill = -20°C,Tmax= +70°C. Samples were periodically immersed in distilled water at the ambient room temperature, after each freezing and heating treatment. Results of accelerated leaching experiments for these samples and samples prepared exclusively with polyethylenes replacing 100 percents of the stone granules and TiO2, treated in nonaccelerated leaching experiments, were compared. Even using an accelerated ageing leach test that overestimates 137Cs leach rates, it can be deduced, that radionuclide leach rates from the radioactive waste mortar mixture forms were improved. Leach rates decreased from 5 percent, for the material prepared with stone aggregate, to 3.1 to 4.0 percent, for the materials prepared solely with PEHD, PELD or TiO2, and to about 3 percents for all six types of the TiO2-PEHD and TiO2-PELD mixtures tested.

Because there are literally square kilometers of radioactively contaminated concrete surfaces within the U.S. Department of Energy (DOE) complex, the task (both scope and cost) of decontamination is staggering. Complex-wide cleanup using conventional methodology does not appear to be feasible for every facility because of prioritization, cost, and manual effort required.

We are investigating the feasibility of using microbially influenced degradation (MID) of concrete as a unique, innovative approach for the decontamination of concrete. Currently, work is being conducted to determine the practicality and cost effectiveness of using this environmentally acceptable method for decontamination of large surface concrete structures. Under laboratory conditions, the biodecontamination process has successfully been used to remove 2 mm of the surface of concrete slabs. Subsequently, initial field application data from an ongoing pilot-scale demonstration have shown that an average of 2 mm of surface can be removed from meter-square areas of contaminated concrete. The cost for the process has been estimated as $1.29/m2. Methodologies for field application of the process are being developed and will be tested. This paper provides information on the MID process, laboratory evaluation of its use for decontamination, and results from the pilot field application.