Neutron irradiation was used to simulate alpha-decay damage in zirconolite, resulting in a transformation from the crystalline to the amorphous state at doses of 4–25 × 1019 n/cm2 (E ≥ 1 MeV). With increasing dose, the radiation damage microstructures resemble damage caused by: 1) alpha-decay of 232Th and 238U in natural zirconolites, 2) alpha-decay of 238Pu or 244Cm in synthetic samples, and 3) collision cascades in samples irradiated with heavy ions. Heavily damaged zirconolite recovers to a defect fluorite phase on annealing at temperatures up to 1000 °C. The main stage of structural recovery was found to occur at temperatures of 600–800 °C. The microstructures after heating depend on the initial level of damage: zirconolite grains with low to moderate levels of damage anneal to imperfect single crystals, whereas heavily damaged grains recrystallize to a polycrystalline microstructure. Complications encountered in this work include the production of fission tracks (due to trace amounts of U) and a non-uniform distribution of damage at higher dose levels (possibly due to electron beam heating).

Phase equilibria are established in the CaO-TiO2-ZrO2 system at 1200 °C, using X-ray diffraction and electron probe microanalysis. The existence of two previously reported ternary phases, zirconolite (CaZrTi2O7) and calzirtite (Ca2Zr5Ti2O16), is confirmed. Each of these phases exhibits a significant range of homogeneity between TiO2 and ZrO2 while maintaining a nearly constant concentration of CaO. The ternary solubilities of the constituent binary phases are found to be negligible, with the exceptions of the perovskites, which display mutual solubility of at least 22 mol.% and may in fact form a series of continuous solid solutions. The solubilities of Hf and Gd in zirconolite are also investigated. While Hf-bearing samples did not reach thermodynamic equilibrium under the experimental conditions employed, the existence of a Hf analog to zirconolite, CaHfTi2O7, is conclusively demonstrated. The phase is stable at the stoichiometric composition, and its lattice parameters are very close to those reported in the literature for stoichiometric zirconolite. A Gd-bearing sample of the composition Ca0.88Zr0.88Gd0.24Ti2O7 is found to be essentially single phase zirconolite, in agreement with previous investigations at higher temperatures.

This study demonstrates that glass-crystal composite waste forms can be produced from waste streams containing high proportions of phosphorus, transition metals, and/or halides. The crystalline phases produced in crucible-scale melts include apatite, monazite, spinels, and a Zr-Si-Fe-Ti phase. These phases readily incorporated radionuclide and toxic metals into their crystal structures, while corrosion tests have demonstrated that glass-crystal composites can be up to 300-fold more durable than simulated high-level nuclear waste glasses, such as SRL 202U.

The ability of the cubic phase of zirconia to accommodate in solid solution the oxides of rare earth elements with differing cationic radii has been investigated. Mixed oxide phase assemblages were prepared by hydrolysing zirconium butoxide with solutions of rare earth element nitrates followed by drying, calcining and sintering. The resulting products were characterised by X-ray diffraction and energy dispersive spectroscopy. The cubic zirconia phase can accept into solid solution the larger, non-cubic stabilising, rare earth element ions such as lanthanum in the presence of the cubic stabilising oxides of yttrium and samarium. As the proportion of the larger rare earth element ions is increased the formation of pyrochlore type compounds is favoured.

Schoepite, [(UO2)8O2 (OH)12]·12H2O, dehydrates irreversibly to a defect structure-derivative of α-UO2 (OH)2 with composition UO3·0.75H2O (dehydrated schoepite) according to the reaction.[(UO2)8O2 (OH) 12]+12H2O ↠ 8[(UO2 )O0.25 (OH)1 5.] + 12H2O.Natural crystals change from translucent yellow schoepite to opaque yellow, polycrystalline dehydrated schoepite. The complete transformation occurs in three steps: (I) loss of interlayer H20 in schoepite causing collapse of the layers; (2) atomic rearrangement within the sheets from a schoepite-type arrangement to a configuration which may be similar to that of rnetaschoepite, UO3.2H2O; (3) a second rearrangement to the defect α-UO2 (OH)2-type sheet. Although dehydrated schoepite may transform to stoichiometric α-UO 2(OH)2 in water above approximately 290°C, it does not hydrate at any temperature. Schoepite transforms slowly in air at ambient temperature to metaschoepite, although this may take geologically long periods. The transformation of schoepite to dehydrated schoepite can occur rapidly at room temperature if crystals, which have only partially converted to metaschoepite. are subjected to external stress.

The uranyl oxide hydrates (UOH) are important corrosion products of uraninite and UO 2 in spent nuclear fuel under oxidizing conditions. However, the systematics of the crystal chemistry, thermodynamic parameters, and solubilities of this mineral group are poorly understood. With the exception of the synthetic UO2 (OH)2 polymorphs, all UOH crystal structures are based on sheets of edge-sharing 5- and 4-coordinated uranyl dipyramids. This structural similarity suggests that it is possible to develop a model by which to estimate the thermodynamic behavior of UOHs from data on structural endmember phases. Toward this end, a method of quantitatively describing all known UOH sheets has been developed. Only four structural unit chains are required to construct the uranyl oxide hydrate sheets (as well as the structurally similar U3O8 sheets). The H-chain is restricted to α-UO2 (OH)2 and is made up of hexagonally coordinated uranyl ions sharing opposing edges. The “arrowhead” chain composed of pentagonal dipyramids sharing edges and alternating with trigonal vacancies is present in all other UOH sheets. These arrowhead chains are directed and can occur in both an ↑ and ⇓ “sense” within a single sheet. The P-chain consists of edge-sharing pentagonal dipyramids forming a zigzag chain. The P-chain is flanked on both sides by arrowhead chains of the same “sense”. The remaining structural unit is a discontinuous “chain” of rhombic dipyramids. This “R-chain” is produced when nested adjacent ⇑ and ⇓arrowhead chains are translated by a diagonal shift. This “chain” occurs in sheets which contain only 4-coordinate uranyl ion and those containing both 4- and 5-coordinate uranyl ions.

Aeschynite and euxenite structure-types are proposed as host phases for rare earth elements and actinides. Aeschynite and euxenite structure types are represented by the general formula: (REE,Y,Th,U,Ca)(Ti,Nb,Ta)2O6. In both structure-types, a variety of large cations such as rareearths and actinides can be accommodated in the tunnels of connected holes in aeschynite or the interlayer space in euxenite. The waste loadings (rare-earths and actinides) in aeschynite and euxenite are at least 40 wt. % based on chemical compositions of naturally occurring aeschynite and euxenite.

The sol-gel method has been successfully used to prepare aeschynite and euxenite as preferred crystalline ceramics. The aged gels and finally calcined products have been characterized by infrared and Raman spectroscopy, differential thermal analysis, and X-ray diffraction analysis. The fabrication technology of aeschynite and euxenite type ceramics is feasible by using hot-pressing as a final processing stage. The mechanistic processes of aeschynite and euxenite formation by the sol-gel method as described as REE(OH)3 · Nb(OH)5·Ti(OH)4 coprecipitation; the polymerization of Nb and Ti hydroxides to form isolated (Nb,Ti)2O10 di-octahedra through oxygen and hydroxyl bridges accompanied by a dehydration reaction; welding together of isolated “pieces” of (Nb,Ti) 2010 to form longer (Nb,Ti)2O10 di-octahedral chains; REE being connected to the dioctahedral chains through oxygen and hydroxyl bridges; and to gradually form small threedimensional nucleates of aeschynite and euxenite. Critical data from the literature on radiation effects and secondary alteration are reviewed as a part of the evaluation of the long-term performance of these phases.

Sulfur polymer cement (SPC) is a candidate material to encapsulate low-level waste (LLW) glass. Molten SPC will be poured into a LLW glass cullet-filled canister, surrounding the glass to act as an additional barrier to groundwater intrusion. This paper covers the first part of a study performed at Pacific Northwest National Laboratory concerned with the fundamental aspects of embedding LLW glass in SPC. Part one is a study of the SPC itself. Variations in SPC properties are discussed, especially in relation to long-term stability and controlling crystallization in a cooling canister.

A cement capable of being Hot Isostatically Pressed (HIP 'ed) into a glass-ceramic has been proposed for use as the waste form for SNF reprocessing wastes at the Idaho National Engineering Laboratories. Such an “intermediate” cement, with a composition based on that of common glasses, has been designed and tested. The cement formulations included mixed I.N.E.L. wastes, blast furnace slag, reactive silica, alumina, and I.N.E.L. soil or vermiculite, which was activated with potassium or sodium hydroxide. Modified FUETAP processing was performed and the cement was subsequently characterized. Results of compressive strength testing ranged from 1452 psi to 4163 psi, exceeding the NRC-suggested standard of >500 psi. Total dissolved solids concentrations in waste form leachates were calculated from a static leach test in which leachate conductivity was measured. Effective diffusivities for radioisotopes Cs and Sr were calculated from leachate analysis data. Diffusivity values were on the order of 10-15 to 10-10 cm2/sec, which compare favorably with diffusivities in other materials.

Properly prepared cementitious waste forms can be hot-isostatically-pressed into materials that exhibit performance equivalent to typical radwaste-type glasses. The HIPing conditions (temperature/pressure) required to “vitrify” these concretes are quite mild and therefore consistent with both safety and good productivity. This paper describes both the process and its products with reference to potential application to Idaho Chemical Processing Plant (ICPP) reprocessing wastes.

Expressed pore solutions from simulated low level liquid waste cement grouts cured at room temperature, 50°C and 90°C for various duration were analyzed by standard chemical methods and ion chromatography. The solid portions of the grouts were formulated with portland cement, fly ash, slag, and attapulgite clay in the ratios of 3:3:3:1. Two different solutions simulating off-gas condensates expected from vitrification of Hanford low level tank wastes were made. One is highly alkaline and contains the species Na+, PO43-, NO2-, NO3- and OH-. The other is carbonated and contains the species Na+, PO43-, NO2- NO3- and CO32- In both cases phosphate rapidly disappeared from the pore solution, leaving behind sodium in the form of hydroxide. The carbonates were also removed from the pore solution to form calcium carbonate and possibly calcium monocarboaluminate. These reactions resulted in the increase of hydroxide ion concentration in the early period. Subsequently there was a significant reduction OH- and Na+ ion concentrations. In contrast high concentration of NO2- and NO3- were retained in the pore solution indefinitely.

The ability of grouts formulated from mixtures of cementitious materials and attapulgite clay to immobilize various chemical species in the projected off-gas waste stream from vitrification of Hanford low level tank wastes was studied. Three different solid blends were evaluated, with cement :fly ash : slag clay weight ratios of 3:3:3:1, 3:0:6:1, and 0:0:9:1. The blended solids were mixed with a simulated low level liquid waste solution containing Na+, NO2-, NO3-, PO43- and OH- ions, in the proportion of 1 liter of solution to 1 kg of solid blend, and were cured either at 22°C (room temperature), 50°C or 90°C. Pore solutions were expressed at various ages and were analyzed to determine the reductions in concentrations of the individual ionic species. The results were expressed in the form of immobilization index (I) calculated for each species. The immobilization indices for Na+ (I Na+ ) and for OH- (IOH-)were similar in each case, and were found to be highest when only slag and clay was present (blend 0:0:9:1). The immobilization index for phosphate, , was 1 in all cases, i.e. phosphate was completely removed from solution. On the other hand removal of NO2- and NO3- ions was generally ineffective.

The rutile form of titanium dioxide (TiO2) was added to mortar matrix preparations with the objective to improve the mechanical and physico-chemical characteristics of the radwaste-mortar matrix mixtures, in particular the leach-rate of the immobilized radionuclide. The final solidified radwaste form was made with high water-to-cement ratio (0.36) for easy leaching of the immobilized radionuclide, 37Cs. TiO2was added to the mortar formulation, replacing the appropriate amount of cement, in the amounts of 1,2,5,8 and 10 weight percents of total cement weight. In the highly basic environment of the mortar (pH ≅ 12), the titanium will form a HTiO-type membrane, which is semi-permeable and selective for the cations like Cs+ in the pH range above 5.5. Only the rutile form of TiO2was observed in the prospected radwastemortar mixture samples, using X-ray Fluorescence Spectrometry. Nevertheless, 137Cs leach-rate for the matrix formulations prepared with TiO2 was notably lower than the normally prepared (TiO2 free) samples. Furthermore, the mechanical strength of the samples prepared with TiO2 was higher than that of the TiO 2-free samples, and the correlation between the mechanical strength and TiO2 content appears to be exponential over the composition range explored here. Improvement of the physico-chemical properties of the titanium prepared formulations, is a topic of further investigations.

Carbon-14 sorption by cementitious materials should be enhanced to ensure the long term safety of radioactive waste repositories. The sorption mechanism of inorganic C- 14 (CO32- was investigated using batch sorption experiments and zeta potential measurements. The results suggested that C-14 was adsorbed onto the cement surface by an electrostatic force, due to the reaction between SiO2 and CaO contained in the cementitious composition. That is, SiO2 was originally negatively charged (SiO-) in cement, but became positively charged through the interaction of Ca2+. These positive sites on the SiO2 surface adsorbed inorganic C-14. Ordinary Portland cement (OPC) did not contain enough SiO2 compared with its CaO content to produce sufficient numbers of C-14 adsorption sites. The C-14 distribution coefficient (Kd) was increased from 2,000 to 7,000 mL/g by adding SiO2 to OPC.

A model has been developed for the sorption of radioelements onto cementitious materials based on the diffuse-layer modelling approach. The model assumes that silicon sites (>SiOH) and calcium sites (>CaOH) dominate the surface chemistry and the sorption of radioelements onto the cementitious materials. Both types of site may undergo surface protonation and deprotonation reactions. Cement-based systems vary greatly in their chemistry depending on their calcium-tosilicon molar ratio, and the corresponding variation in the surface chemistry has been incorporated by allowing sorption of calcium ions onto silicon sites. This process results in a change from a silica-type surface, at very low calcium-silicon ratios, to a calcium hydroxide-type surface for high-calcium cement-based materials. The predicted variation in the surface chemistry is consistent with literature data on measured zeta potentials of cements. The model has been applied successfully to describe the sorption of simple caesium and iodide ions at varying calciumsilicon ratios. In a Nirex repository for low and intermediate level wastes, a high-calcium cementitious backfill would be specified. This model has allowed a consistent interpretation of experimental data for sorption of key radioelements, including uranium and plutonium, onto the backfill, under saline and non-saline conditions.

Many phases in the calcium-silicate-hydrate system can develop in cement exposed over long periods of time to temperatures above 25°C. As a consequence, chemical reactions involving these phases can affect the relative humidity and water chemistry of a radioactive waste repository that contains significant amounts of cement. In order to predict and simulate these chemical reactions, we are developing an internally consistent database of crystalline Ca-Si-hydrate structures. The results of first principles electronic structure calculations on two such phases, wollastonite (CaSiO3) and xonotlite (Ca6Si6O17(OH)2), are reported here. The calculated ground state properties are in very good agreement with experiment, providing equilibrium lattice parameters within about 1–1.4% of the experimentally reported values. The roles of the different types of oxygen atoms, which are fundamental to understanding the energetics of crystalline Ca-Si-hydrates are briefly discussed in terms of their electronic state densities. The good agreement with experiment for the lattice parameters and the consistency of the electronic density of states features for the two structures demonstrate the applicability of these electronic structure methods in calculating the fundamental properties of these phases.

To determine the leachability and concentration-controlling solids for nuclides contained in solidified low-level nuclear wastes, cement doped with neodymium, uranium, thorium and strontium was prepared and cured in a CO2-free atmosphere. Leaching tests were conducted by equilibrating aliquots of the crushed doped cement with CO2-free deionized water at pcH+ values ranging from 7.5 to 12.5 in an effort to determine if solubility constraints could be used to predict leachable concentrations of Nd (an analog of trivalent actinides), U, Th, and Sr. Steady state concentrations of Nd, U, Th and Sr were attained within a few days. The observed Nd concentrations continually decreased with the increase in pcH+ from about 7.5 to 9.0, and were at detection limit at pcH+ values > 9.0. The comparison of these results with thermodynamic data suggests that Nd(OH)3(am) is the controlling solid phase. The observed U concentrations decreased by two orders of magnitude with a unit increase in pcH+ in the range of 7.5 to 9.0 and were at the detection limit at pcH+ > 9.0. From the multitude of U(VI) solids that can form under the test conditions, the data suggest that CaUO4 is the likely controlling solid. The concentration of Th was at the detection limit over the entire pcH+ range covered. Based on previous work, ThO2(am) is likely to be the controlling solid. In the absence of CO2, no solubility-controlling solid was found for Sr.

Corrosion of massive concrete structures is a costly and environmentally dangerous problem. Many low level waste disposal vaults, both above and below ground, are constructed of concrete. When the integrity of these structures is compromised, potential for soil and water contamination is significantly increased. There are literally thousands of massive concrete structures within the U.S. infrastructure that are in a serious state of disrepair due to corrosion. One potential contributing agent to the destruction of concrete structures is microbially-influenced degradation (MID). MID occurs when microorganisms in the environment produce mineral or organic acids that dissolve or disintegrate the cement matrix. Three groups of bacteria are known to create conditions that are conducive to destroying concrete integrity. They are sulfur oxidizing bacteria, nitrifying bacteria, and heterotrophic bacteria.

Research is being conducted at the Idaho National Engineering Laboratory to assess the extent of naturally occurring microbially influenced degradation (MID) and its contribution to the deterioration of massive concrete structures. The preliminary steps to understanding the extent of MID, require assessing the microbial communities present on degrading concrete surfaces. Ultimately such information can be used to develop guidelines for preventive or corrective treatments for MID and aid in formulation of new materials to resist corrosion. An environmental study was conducted to determine the presence and activity of potential MID bacteria on degrading concrete surfaces of massive concrete structures. Scanning electron microscopy detected bacteria on the surfaces of concrete structures such as bridges and dams, where corrosion was evident. Enumeration of sulfur oxidizing thiobacilli and nitrogen oxidizing Nitrosomonas sp. and Nitrobacter sp. from surface samples was conducted. Bacterial community composition varied between sampling locations, and generally the presence of either sulfur oxidizers or nitrifiers dominated, although instances of both types of bacteria occurring together were encountered. No clear correlation between bacterial numbers and degree of degradation was exhibited.

Because of its apparent structural integrity, cement has been widely used in the United States as a binder to solidify Class B and C low-level radioactive waste (LLW). However, the resulting cement preparations are susceptible to failure due to the actions of stress and environment. An environmentally mediated process that could affect cement stability is the action of naturally occurring microorganisms. The U.S. Nuclear Regulatory Commission (NRC), recognizing this eventuality, stated in their Technical Position on Waste Form, Revision 1, that the effects of microbial action on waste form integrity must be addressed. This paper provides recent results from a program that examined the effects of microbially influenced degradation (MID) on cement-solidified LLW. Data are provided which were obtained during the development of an evaluation method using acid-producing bacteria. Results presented here are from work with one type of these bacteria, the sulfur-oxidizing Thiobacillus. Commercially prepared, cement-solidified, low-level radioactive waste form samples made from power reactor wastes were evaluated using a new biodegradation test developed for the NRC. Testing demonstrated that MID has the potential to severely compromise the structural integrity and nuclide retentiveness of ion-exchange resin and evaporator-bottoms wastes that have been solidified with cement. It was found that the waste form specimens physically deteriorated after 60 days of exposure to the thiobacilli. Also, the data show that significant amounts of Cs-137, Cs-134, Co-60, C-14, Tc-99, and Sr-90 contained in the waste forms were leached in the presence of Thiobacillus.

The importance of the large volumes of cement and concrete in the planned Swiss L/ILW repository in influencing the chemical environment of the repository and the sorption behaviour of the radionuclides has lead to an interest the long-term evolution of cement-groundwater systems for safety assessment purposes.

A model is presented which simulates the leaching of a column hydrated cement/concrete with fresh groundwater, using a sequential mixing tank approach. The chemical model includes the incongruent dissolution of the CSH-gel phase and reaction of three other cement phases which make up over 90 wt. % of the hydrated cement paste. The model takes account of porosity changes due to dissolution and precipitation. Different concrete compositions can be included in the column in order to represent, for example, a section through a repository cavern. The model can be run as a 1-D flow-through system or as a quasi-1-D system in which flow in any segment is proportional to porosity. Although the model does not include the sophisticated transport models of currently available coupled codes, it does allow the possibility of assessing the duration of high pH conditions within the repository and the influence of different concrete/cement types or different waste type/loading as a first stage to determining long-term transport of radionuclides from the near field.