The generation of volatile organic compounds (VOCs), hydrogen, and carbon oxides due to alpha particle irradiation of polyethylene, polyvinylchloride, hypalon, and neoprene, is being investigated. A wide diversity of VOCs was found including alkenes, alkanes, alcohols, ketones, benzene derivatives, and nitro compounds. Their yields however, were quite low. The relative amounts of these compounds depended on the material, atmosphere present, and the absorbed dose. This investigation will help evaluate the effect of ionizing radiation on the long-term performance assessment and regulatory compliance issues related to the Waste Isolation Pilot Plant (WIPP).

Design-basis transuranic (TRU) waste to be emplaced in the Waste Isolation Pilot Plant (WIPP) in southeastern New Mexico may generate significant quantities of gas, which may affect the performance of the WIPP with respect to regulations for radioactive and/or chemically hazardous waste constituents. We are developing a model to predict gas generation in WIPP disposal rooms during and after filling and sealing. Currently, the model includes: (1) oxic and anoxic corrosion of steels and other Fe-base alloys, including passivation and depassivation; (2) microbial degradation of cellulosics with O2, NO3-, FeO(OH) , SO42-, or CO2 as the electron acceptor; (3) α radiolysis of brine; (4) consumption of CO2 and, perhaps, H2S by Ca(OH)2 (in cementitious materials) and CaO (a potential backfill additive). The code simulates these processes and interactions among them by converting reactants (steels, cellulosics, etc.) to gases and other products at experimentally observed or estimated rates and plotting temporal reaction paths in three-dimensional phase diagrams for solids in the Fe-H2O-CO2-H2-H2S system.

Finite element calculations of the porosity history of a nuclear waste disposal room for transuranic (TRU) waste in a bedded salt formation have been completed. The analyses include a reduced elastic/secondary creep model for the host halite and a nonlinear consolidation model for the crushed salt backfill. Separate gas generation and constitutive models were used for three distinct hypothetical waste forms: 1) unaltered defense-related, contact-handled transuranic (CH-TRU) waste, 2) shredded and cemented CH-TRU waste, and 3) incinerated and vitrified CH-TRU waste. Histories were determined for a 2000 year time period starting from the emplacement of room seals. The resulting final room porosities varied from roughly 55% to less than 10%, according to waste forms.

Preliminary work suggests that Acoustic Resonance Spectroscopy (ARS) can non-invasively determine gas pressurization of waste drums. This is done by exciting the various mechanical vibrational modes of the test drum and analyzing the characteristics of the resulting resonance frequencies and their higher harmonics. If the contained waste is a liquid, the liquid level and also some properties of the liquid including density can be determined. Additional work is needed to exploit this novel acoustic technique for noninvasive characterization of waste storage drums.

Monoliths of SRL 131 borosilicate glass were irradiated in a saturated air-steam environment, at temperatures of 150°C, to examine the effects of radiation on nuclear waste glass behavior. Half of the tests used actinide and Tc-99 doped glass and were exposed to an external ionizing gamma source, while the remaining glass samples were doped only with uranium and were reacted without any external radiation exposure. The effects of radiation exposure on glass alteration and secondary phase formation were determined by comparing the reaction rates and mineral paragenesis of the two sets of samples.

All glass samples readily reacted with the water that condensed on their surfaces, producing two types of smectite clay within the first three days of testing. Additional crystalline phases precipitated on the altered glass surface with increasing reaction times, including zeolites, smectite, calcium and sodium silicates, phosphates, evaporitic salts, and uranyl silicates. Similar phases were produced on both the nonirradiated and irradiated samples; however, the quantity of precipitates was increased and the rate of paragenetic sequence development was accelerated in the latter. After 56 days of testing, the composite smectite layer developed at an average rate of ~0.16 and 0.63 µm/day for the nonirradiated and irradiated samples, respectively. These comparisons indicate that layer development is accelerated approximately four-fold due to the radiation exposure at high glass surface area/liquid volume (SA/V) conditions. This increase apparently occurs in response to the rapid concentration of radiolytic products, including nitric acid, in the thin films of water contacting the sample monoliths.

A new class of inelastic collisional processes accounts for the electron spectra of ion-bombarded sodium halide surfaces. These spectra indicate that the deexcitation process consists of a sequence of lattice-ion collisions in which localized electron transfer occurs. Such collisional processes demonstrate how collisions can initiate electronic change in molecular compounds and that electron-transfer processes must play an essential role in chemistry activated by energetic collisions.

Pyrochemical processing of spent fuel from the Integral Fast Reactor (IFR) yields a salt waste of LiCI-KCI that contains approximately 6 wt% fission products, primarily as CsCI and SrCl2. Past work has shown that zeolite A will preferentially sorb cesium and strontium and will encapsulate the salt waste in a leach-resistant, radiation-resistant aluminosilicate matrix. However, a method is still needed to convert the salt-occluded zeolite powders into a monolith suitable for geologic disposal. We are thus investigating a method that forms bonded zeolite by hot pressing a mixture of glass frit and sait-occluded zeolite powders at 990 K (717°C) and 28 MPa. The leach resistance of the bonded zeolite was measured in static leach tests run for 28 days in 363 K (90°C) deionized water. Normalized release rates of all elements in the bonded zeolite were low, <1 g/m2d. Thus, the bonded zeolite may be a suitable waste form for IFR salt waste.

The Environmental Restoration and Waste Management Applied Technology Program at LLNL is developing a Mixed Waste Management Facility to demonstrate treatment technologies that provide an alternative to incineration. As part of that program, we are developing final waste forms using ceramic processing methods for the immobilization of the treatment process residues. The ceramic phase assemblages are based on using Synroc D as a starting point and varying the phase assemblage to accommodate the differences in chemistry between the treatment process residues and the defense waste for which Synroc D was developed. Two basic formulations are used, one for low ash residues resulting from treatment of organic materials contaminated with RCRA metals, and one for high ash residues generated from the treatment of plastics and paper products. Treatment process residues are mixed with ceramic precursor materials, dried, calcined, formed into pellets at room temperature, and sintered at 1150 to 1200°C to produce the final waste form. This paper discusses the chemical composition of the waste streams and waste forms, the phase assemblages that serve as hosts for inorganic waste elements, and the changes in waste form characteristics as a function of variation in process parameters.

New diffraction and microstructural data on the incorporation of rare earths, Np and Pu in nominally single-phase zirconolite (CaZrTi2O7) are presented. Such data are vital for the crystal-chemical design of zirconolite-rich titanate ceramics of the Synroc family, for the immobilization of up to 30 wt% of actinide wastes from advanced reprocessing/partitioning schemes. Some trial zirconolite-rich ceramics were made with U and Nd simulating tetravalent and trivalent actinides respectively. Some preliminary results to examine chemical flexibility with regards to variations in waste/precursor are given, for different actinide waste compositions.

Powder X-ray diffraction patterns were collected from three titanate waste forms - a calcine powder, a prototype ceramic without waste, and a ceramic containing 10 wt% JW-A simulated waste - and interpreted quantitatively using the Rietveld method. The calcine consisted of fluorite, pyrochlore, rutile, and amorphous material. The prototype waste form contained rutile, hollandite, zirconolite and perovskite. The phase constitution of the JW-A ceramic was freudenbergite, loveringite, hollandite, zirconolite, perovskite and baddeleyite. Procedures for the collection of X-ray data are described, as are assumptions inherent in the Rietveld approach. A selection of refined crystal data are presented.

Synroc is a multi-component titanate ceramic, designed to immobilise High Level Waste (HLW) from nuclear fuel reprocessing plants [1]. Synroc precursor powders have been previously produced by various methods, such as oxide and alkoxide-hydrolysis routes [2]. However, various technological aspects of HLW processing make the use of free-flowing, dustfree, highly sinterable precursor powders desirable. Such powders have been produced by spray-drying colloidal precursors, yielding microspherical particles with controlled porosity.

The behaviour and performance of cement barriers in the repository have been modelled. Since the models employ thermodynamic calculations, an important task is the development of a validated database. This must include phases and aqueous species appropriate to cement and modified cements, containing tly ash and slag. Closed system mineralogy and chemistry are explored as a function of cure duration and thermal excursions. Work has only just begun on open system behaviour: some aspects of reactions with chloride, sulfate and carbonate are described. Future progress depends on database extensions, validation experiments and the coupling of equilibrium and kinetic aspects of reaction.

An experimental and modeling program has been designed to elucidate the structural and thermodynamic response of cement minerals to elevated temperature. The components of the program include: a) synthesis of hydrated Ca-silicates; b) structural analysis of cement phases during heating and dehydration/rehydration; c) mechanistic and thermodynamic descriptions of the hydration/dehydration behavior of hydrated Ca-silicates as a function of temperature, pressure and relative humidity; d) study of naturally occurring hydrated Ca-silicates; and e) measurements of thermodynamic data for hydrated Ca-silicates.

The history of cemenütious materials goes back to ancient times. The Greeks and Romans used calcined limestone and later developed pozzolanic cement by grinding together lime and volcanic ash called “pozzolan” which was first found near Port Pozzuoli, Italy. The ancient Chinese used lime-pozzolanic mixes to build the Great Wall. The ancient Egyptians used calcined impure gypsum to build the Great Pyramid of Cheops. The extraordinary stability and durability of these materials has impressed us, when so much dramatically damaged infrastructure restored by using modern portland cement now requires rebuilding. Stability and durability of cementitious materials have attracted intensive research interest and contractors’ concerns, as does immobilization of radioactive and hazardous industrial waste in cementitious materials. Nuclear waste pollution of the environment and an acceptable solution for waste management and disposal constitute among the most important public concerns. The analogy of ancient cementitious materials to modern Portland cement could give us some clues to study their stability and durability. This present study examines selected results of studies of ancient building materials from France, Italy, China, and Egypt, combined with knowledge obtained from the behavior of modern portland cement to evaluate the potential for stability and durability of such materials in nuclear waste forms.

An accelerated experimental technique using high wateπcement ratios has been developed to study the long term hydration of blended cements that may be used in a repository for the disposal of radioactive waste. This technique has been used to investigate the hydration reactions of Ordinary Portland Cement (OPC) blended with ground granulated blastfurnace slag (ggbs) or pulverised fuel ash (pfa). The effects of high sulphate-bearing and high carbonate-bearing ground waters on the compounds formed on hydration were investigated. Solid/solution compositional data were collected during the course of the hydration process for periods up to 2 years. Thomsonite, thaumasite, afwillite and a tobermorite-like phase were found in addition to the expected cement hydration products. The pH of the aqueous solution in contact with 60 pfa:40 OPC blends hydrated at 90°C fell to below 8. This is lower than the value required to inhibit the corrosion of steel canisters in a repository. The pH of the aqueous solution in contact with OPC and 75 ggbs:25 OPC blends remained above 11, although if the ground waters in contact with the OPC/ggbs blends were periodically replaced the pH eventually fell below 10.

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 that the effects of microbial action on waste form integrity must be addressed.

This paper provides present results from an ongoing program that addresses the effects of microbially influenced degradation (MID) on cement-solidified LLW. Data are provided on the development of an evaluation method using acid-producing bacteria. Results are from work with one type of these bacteria, the sulfur-oxidizing Thiobacillus. This work involved the use of a system in which laboratory- and vendor-manufactured, simulated waste forms were exposed on an intermittent basis to media containing thiobacilli. Testing demonstrated that MID has the potential to severely compromise the structural integrity of ion-exchange resin and evaporator-bottoms waste that is solidified with cement. In addition, it was found that a significant percentage of calcium and other elements were leached from the treated waste forms. Also, the surface pH of the treated specimens decreased to below 2. These conditions apparently contributed to the physical deterioration of simulated waste forms after 60 days of exposure to the thiobacilli.

We are investigating the possibility of monitoring the degree of immobilization of the low level radioactive waste by embedding an array of long lasting electrodes in grout. This work describes our ongoing attempts to understand the physics and chemistry of charge carrier in the grout under various load conditions.

Disposal of low-level radioactive waste in concrete vaults buried underground is being considered in the United States and a potentially significant degradation process is sulfate attack. The results of experimental studies and of modeling of the kinetics of sulfate attack of cement-based materials are given. It was concluded that the expansion process involves a two-stage mechanism. The first stage is likely to be diffusion controlled and the second stage appears to be acceleratory due to the development and spread of cracks.

The fraction released of radioactive cesium from waste forms (obtained by incorporating a given amount of simulated labelled waste materials into cement) was greater than that released when introducing the same amount of waste into a preformed cement container. Moreover, improving cement containers using chemical treatment or additives (organic or inorganic) lowers the fraction released. The effect of waste materials on cement hydration and on the properties of the obtained products has been studied. To improve repository performance it is therefore preferable to ameliorate the confinement properties of preformed cement containers rather than radioactive waste forms. In improving waste forms we are dealing with a complicated system resulting from the interaction of waste materials with cement besides allowing extra doses to workers and finally increasing the final total coasts.

Series of mortar-radwaste mixture samples, orthocylinder shaped, with the external surface completely open to contact with leachant medium, were investigated in experiments that tried to characterize long-term physico-chemical and mechanical properties of the matrix material. After curing period of 28 days in the atmosphere of defined parameters, one group of the samples was treated on leaching in the distilled water for a one year period, to establish the 137Cs cumulative leached fraction. The second group of the samples of the same batch as the first one, after one year curing period , was investigated under temperature stress conditions. The values of the temperature climaxes were: Tmin=-20°C and Tmax=+70°C. Samples were periodically immersed in distilled water at the room temperature, after each freezing and heating treatment. The experimental goal was to bring mortar-radwaste mixture forms, during a short period of experimental time, into the state that has nearly the same physico-chemical properties, mainly radionuclide leach-rate, as the first sample group has, which was treated in the normal one-year leaching process. Data obtained from these investigations, using accelerated ageing processes, do not correspond completely to the appropriate values that could define the state of the samples treated excusively in the leachant medium. Nevertheless, accelerated ageing processes are suitable for the relatively fast estimation of immobilizing properties of mortar matrices, and their behaviour in the extended experimental time.