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Sample records for cement waste forms

  1. Evaluation of solidified cement waste forms

    International Nuclear Information System (INIS)

    A part of the program of the treatment of low and intermediate-level liquid wastes, is presented. The concrete has been suggested as a encapsulation or overpack material for long-term storage of radioactive waste from the nuclear fuel cycle, particularly in conjunction with the disposal of lowand intermediate-level waste. The influence of sodium nitrates, as a possible integrating component of waste, on some properties of national Portland cement, was investigated. The waste form properties, discussed in terms of their dependency on waste type and amount, include water/cement ratio, setting times, compressive strength and homogeneity. Criteria to be applied in the assessment of the final solidified waste are proposed. (Author)

  2. Quality control of cemented waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Slate, L.J.

    1994-12-31

    To insure that cemented radwaste remains immobilized after disposal, certain standards have been set in Europe by the Commission of the European Communities. One such standard is compressive strength. If the compressive strength can be predicted during the early curing stages, time and money can be saved and the quality of the final waste form guaranteed. It was determined that the 7- and 28-day compressive strength from radwaste cementation can be predicted during the mixing and early curing stages by at least three methods. The three that were studied were maturity, rheology, and impedance. Maturity is a temperature-to-time measurement, rheology is a shear stress-to-shear rate measurement, and impedance is the opposition offered to the flow of alternating current. These three methods were employed on five different cemented radwaste concentrations with three different water-to-cement ratios; thus, a total of 15 different mix designs were considered. The results showed that the impedance was the easiest to employ for an on-line process. The results of the impedance method showed a very good relationship between impedance and water-to-cement ratio; therefore, an accurate prediction of compressive strength of cemented radwaste can be drawn from this method. The results of the theology method were very good. The method showed that concrete conforms to the Bingham plastic rheologic model, and the theology method can be used to predict the compressive strength of cemented radwaste, but may be too cumbersome. The results of the maturity method were shown to be limited in accuracy for determining compressive strength.

  3. Testing and evaluation of polyethylene and sulfur cement waste forms

    International Nuclear Information System (INIS)

    This paper discusses the results of recent studies related to the use of polyethylene and modified sulfur cement as new binder materials for the improved solidification of low-level wastes. Waste streams selected for this study include those which result from advanced volume reduction technologies (dry evaporator concentrate salts and incinerator ash) and those that remain problematic for solidification using contemporary agents (ion-exchange resins). Maximum waste loadings were determined for each waste type. Recommended waste loadings of 70 wt % sodium sulfate, 50 wt % boric acid, 40 wt % incinerator ash and 30 wt % ion exchange resins, which are based on process control and waste form performance considerations are reported for polyethylene. For sulfur cement the recommended waste loadings of 40 wt % sodium sulfate and boric acid salts and 43 wt % incinerator ash are reported. However, incorporation of ion-exchange resin waste in modified sulfur cement is not recommended due to poor waste form performance. The work presented in this paper will, in part, present data that can be used to assess the acceptability of polyethylene and modified sulfur cement waste forms to meet the requirements of 10 CFR 61. 8 refs., 10 figs., 6 tabs

  4. Testing and evaluation of polyethylene and sulfur cement waste forms

    International Nuclear Information System (INIS)

    This paper discusses the results of recent studies related to the use of polyethylene and modified sulfur cement as new binder materials for the improved solidification of low-level wastes. Waste streams selected for this study include those which result from advanced volume reduction technologies (dry evaporator concentrate salts and incinerator ash) and those that remain problematic for solidification using contemporary agents (ion exchange resins). Maximum waste loadings were determined for each waste type. Recommended waste loadings of 70 wt% sodium sulfate, 50 wt% boric acid, 40 wt% incinerator ash and 30 wt% ion exchange resins, which are based on process control and waste form performance considerations are reported for polyethylene. For sulfur cement the recommended waste loadings of 40 wt% sodium sulfate and boric acid salts and 43 wt% incinerator ash are reported. However, incorporation of ion exchange resin waste in modified sulfur cement is not recommended due to poor waste form performance. Data is presented that can be used to assess the acceptability of polyethylene and modified sulfur cement waste forms to meet the requirements of 10 CFR 61. 8 references, 10 figures, 6 tables

  5. The characterization of cement waste form for final disposal of decommissioning concrete wastes

    International Nuclear Information System (INIS)

    Highlights: • Decommissioning concrete waste recycling and disposal. • Compressive strength of cement waste form. • Characteristic of thermal resistance and leaching of cement waste form. - Abstract: In Korea, the decontamination and decommissioning of KRR-1, 2 at KAERI have been under way. The decommissioning of the KRR-2 was finished completely by 2011, whereas the decommissioning of KRR-1 is currently underway. A large quantity of slightly contaminated concrete waste has been generated from the decommissioning projects. The concrete wastes, 83ea of 200 L drums, and 41ea of 4 m3 containers, were generated in the decommissioning projects. The conditioning of concrete waste is needed for final disposal. Concrete waste is conditioned as follows: mortar using coarse and fine aggregates is filled with a void space after concrete rubble pre-placement into 200 L drums. Thus, this research developed an optimizing mixing ratio of concrete waste, water, and cement, and evaluated the characteristics of a cement waste form to meet the requirements specified in the disposal site specific waste acceptance criteria. The results obtained from a compressive strength test, leaching test, and thermal cycling test of cement waste forms conclude that the concrete waste, water, and cement have been suggested as an optimized mixing ratio of 75:15:10. In addition, the compressive strength of the cement waste form was satisfied, including a fine powder up to a maximum of 40 wt% in concrete debris waste of about 75%. According to the scale-up test, the mixing ratio of concrete waste, water, and cement is 75:10:15, which meets the satisfied compressive strength because of an increase in the particle size in the waste

  6. Enhancement of cemented waste forms by supercritical CO{sub 2} carbonation of standard portland cements

    Energy Technology Data Exchange (ETDEWEB)

    Rubin, J.B.; Carey, J.; Taylor, C.M.V.

    1997-08-01

    We are conducting experiments on an innovative transformation concept, using a traditional immobilization technique, that may significantly reduce the volume of hazardous or radioactive waste requiring transport and long-term storage. The standard practice for the stabilization of radioactive salts and residues is to mix them with cements, which may include additives to enhance immobilization. Many of these wastes do not qualify for underground disposition, however, because they do not meet disposal requirements for free liquids, decay heat, head-space gas analysis, and/or leachability. The treatment method alters the bulk properties of a cemented waste form by greatly accelerating the natural cement-aging reactions, producing a chemically stable form having reduced free liquids, as well as reduced porosity, permeability and pH. These structural and chemical changes should allow for greater actinide loading, as well as the reduced mobility of the anions, cations, and radionuclides in aboveground and underground repositories. Simultaneously, the treatment process removes a majority of the hydrogenous material from the cement. The treatment method allows for on-line process monitoring of leachates and can be transported into the field. We will describe the general features of supercritical fluids, as well as the application of these fluids to the treatment of solid and semi-solid waste forms. some of the issues concerning the economic feasibility of industrial scale-up will be addressed, with particular attention to the engineering requirements for the establishment of on-site processing facilities. Finally, the initial results of physical property measurements made on portland cements before and after supercritical fluid processing will be presented.

  7. Enhancement of cemented waste forms by supercritical CO2 carbonation of standard portland cements

    International Nuclear Information System (INIS)

    We are conducting experiments on an innovative transformation concept, using a traditional immobilization technique, that may significantly reduce the volume of hazardous or radioactive waste requiring transport and long-term storage. The standard practice for the stabilization of radioactive salts and residues is to mix them with cements, which may include additives to enhance immobilization. Many of these wastes do not qualify for underground disposition, however, because they do not meet disposal requirements for free liquids, decay heat, head-space gas analysis, and/or leachability. The treatment method alters the bulk properties of a cemented waste form by greatly accelerating the natural cement-aging reactions, producing a chemically stable form having reduced free liquids, as well as reduced porosity, permeability and pH. These structural and chemical changes should allow for greater actinide loading, as well as the reduced mobility of the anions, cations, and radionuclides in aboveground and underground repositories. Simultaneously, the treatment process removes a majority of the hydrogenous material from the cement. The treatment method allows for on-line process monitoring of leachates and can be transported into the field. We will describe the general features of supercritical fluids, as well as the application of these fluids to the treatment of solid and semi-solid waste forms. some of the issues concerning the economic feasibility of industrial scale-up will be addressed, with particular attention to the engineering requirements for the establishment of on-site processing facilities. Finally, the initial results of physical property measurements made on portland cements before and after supercritical fluid processing will be presented

  8. Characterization of cement and bitumen waste forms containing simulated low-level waste incinerator ash

    International Nuclear Information System (INIS)

    Incinerator ash from the combustion of general trash and ion exchange resins was immobilized in cement and bitumen. Tests were conducted on the resulting waste forms to provide a data base for the acceptability of actual low-level waste forms. The testing was done in accordance with the US Nuclear Regulatory Commission Technical Position on Waste Form. Bitumen had a measured compressive strength of 130 psi and a leachability index of 13 as measured with the ANS 16.1 leach test procedure. Cement demonstrated a compressive strength of 1400 psi and a leachability index of 7. Both waste forms easily exceed the minimum compressive strength of 50 psi and leachability index of 6 specified in the Technical Position. Irradiation to 108 Rad and exposure to 31 thermal cycles ranging from +600) to -300C did not significantly impact these properties. Neither waste form supported bacterial or fungal growth as measured with ASTM G21 and G22 procedures. However, there is some indication of biodegradation due to co-metabolic processes. Concentration of organic complexants in leachates of the ash, cement and bitumen were too low to significantly affect the release of radionuclides from the waste forms. Neither bitumen nor cement containing incinerator ash caused any corrosion or degradation of potential container materials including steel, polyethylene and fiberglass. However, moist ash did cause corrosion of the steel

  9. Evaluation of sulfur polymer cement as a waste form for the immobilization of low-level radioactive or mixed waste

    International Nuclear Information System (INIS)

    Sulfur polymer cement (SPC), also called modified sulphur cements, is a relatively new material in the waste immobilization field, although it was developed in the late seventies by the Bureau of Mines. The physical and chemical properties of SPC are interesting (e.g., development of high mechanical strength in a short time and high resistance to many corrosive environments). Because of its very low permeability and porosity, SPC is especially impervious to water, which, in turn, has led to its consideration for immobilization of hazardous or radioactive waste. Because it is a thermosetting process, the waste is encapsulated by the sulfur matrix; therefore, very little interaction occurs between the waste species and the sulfur (as there can be when waste prevents the set of portland cement-based waste forms)

  10. Hydrated phases and pore solution composition in cement solidified saltstone waste forms

    International Nuclear Information System (INIS)

    The mineral phases and pore solution composition of hydrated cement solidified synthetic saltstone waste forms are quantified using thermogravimetric analysis, quantitative X-ray powder diffraction, and inductively coupled plasma atomic emission spectroscopy. Although the synthetic waste contained additional sulfate, the overall chemistry of the system suppressed the formation of sulfate-bearing mineral phases. This was corroborated by the pore solution analysis that indicated very high sulfur concentrations. After one year of hydration, the mineral phases present and the composition of the pore solution are stable, and are generally consistent with expectations based on the hydration of high volume portland cement replacement mixtures. (authors)

  11. Long-term leaching experiments of full-scale cemented waste forms: Experiments and modeling

    International Nuclear Information System (INIS)

    Experimental findings of full-scale leach tests performed on simulated cemented waste forms and self-shielded concrete waste containers for periods up to 19 yr in saturated salt brines (NaCl- and Q-brine) are presented. Measurements cover the evolution of leachant composition and the release of radionuclides such as Cs, U, and Np. Performance of the waste forms and the self-shielded concrete waste containers depends on the pore volume of the hardened cement/concrete, which is correlated to the water/cement ratio of the waste forms. Cesium release follows a linear time dependence. Samples, especially those having a high pore volume, show almost complete release of Cs in the period of investigation. Uranium release is independent of the leach period. Uranium concentrations are controlled by thermodynamic equilibrium. Neptunium is released only to a small extent; concentrations are close to the detection limit. Modeling of the cement corrosion progress allows the prediction of the evolution of the brines in terms of pH, calcium concentration, etc. and the identification of solids controlling the solubilities of the main components and of uranium

  12. Long-Term Leaching Experiments of Full-Scale Cemented Waste Forms: Experiments and Modeling

    International Nuclear Information System (INIS)

    Experimental findings of full-scale leach tests performed on simulated cemented waste forms and self-shielded concrete waste containers for periods up to 19 yr in saturated salt brines (NaCl- and Q-brine) are presented. Measurements cover the evolution of leachant composition and the release of radionuclides such as Cs, U, and Np. Performance of the waste forms and the self-shielded concrete waste containers depends on the pore volume of the hardened cement/concrete, which is correlated to the water/cement ratio of the waste forms. Cesium release follows a linear time dependence. Samples, especially those having a high pore volume, show almost complete release of Cs in the period of investigation. Uranium release is independent of the leach period. Uranium concentrations are controlled by thermodynamic equilibrium. Neptunium is released only to a small extent; concentrations are close to the detection limit.Modeling of the cement corrosion progress allows the prediction of the evolution of the brines in terms of pH, calcium concentration, etc. and the identification of solids controlling the solubilities of the main components and of uranium

  13. Comparative research on leaching models of radionuclides in cement solidified waste forms

    International Nuclear Information System (INIS)

    Based on the diffusion mechanism of radionuclide leaching from cement-solidified waste forms, a two-dimension decay leaching model (T-DLM) was established, according to Fick's second law and the initial and boundary conditions. The leaching behavior of 137Cs in the lab-scale alkali-activated slag-clay minerals composite cement and full scale sulfate resistant cement waste forms was predicted using the T-DLM, via correcting its apparent diffusion coefficient. It was compared with the one-dimension leaching model and one-dimension decay leaching model available. The results indicate that the values calculated using the T-DLM are in a better agreement with the experimental values. (authors)

  14. Transport of nitrate from a large cement based waste form

    International Nuclear Information System (INIS)

    A finite-element model is used to calculate the time-dependent transport of nitrate from a cement-based (saltstone) monolith with and without a clay cap. Model predictions agree well with data from two lysimeter field experiments begun in 1984. The clay cap effectively reduces the flux of nitrate from the monolith. Predictions for a landfill monolith design show a peak concentration occurring within 25 years; however, the drinking water guideline is exceeded for 1200 years. Alternate designs and various restrictive liners are being considered

  15. Integral migration and source-term experiments on cement and bitumen waste forms

    International Nuclear Information System (INIS)

    This is the final report of a programme of research which formed a part of the CEC joint research project into radionuclide migration in the geosphere (MIRAGE). This study addressed the aspects of integral migration and source term. The integral migration experiment simulated, in the laboratory, the intrusion of water into the repository, the leaching of radionuclides from two intermediate-level waste-forms and the subsequent migration through the geosphere. The simulation consisted of a source of natural ground water which flowed over a sample of waste-form, at a controlled redox potential, and then through backfill and geological material packed in columns. The two waste forms used here were cemented waste from the WAK plant at Karlsruhe in the Federal Republic of Germany and bitumenized intermediate concentrates from the Marcoule plant in France. The soluble fission products such as caesium were rapidly released from the cemented waste but the actinides, and technetium in the reduced state, were retained in the waste-form. The released of all nuclides from the bitumenized waste was very low

  16. Mathematical modelling of the corrosion and leaching behaviour of cemented waste forms

    International Nuclear Information System (INIS)

    Essential processes which may lead to a radionuclide release from a cemented waste form, in addition to diffusion, are dissolution processes, changing of solubilities and effective diffusion constants caused by chemical reactions, as well as forced fluid movement by variation of the pore volume in the hydrated cement paste. The mobilization and the transportation process can be calculated for a homogeneous cemented waste form by means of diffusion equation extended by dissolution and fluid movement terms. By introduction of effective constants the complex structure of the hardened cement paste will be homogenised mathematically. In the case of Cs leaching the effective constants describing the release process is the diffusion constant, whereas for Sr leaching the constants are the maximum solubilities of the Sr-compounds in the pore fluid of the cement and their dissolution kinetics. Release rates computed from the solution of the differential equations are compared with experimental data to evaluate the essential constants. To do this the computer code ''DIFMOD'' was developed, which computes the concentrations of the interesting substances in a specimen as a function of different parameters of influence. The procedure used is an implicit Crank-Nicolson finite difference formula. (orig./HP)

  17. Cement waste-form development for ion-exchange resins at the Rocky Flats Plant

    International Nuclear Information System (INIS)

    This report describes the development of a cement waste form to stabilize ion-exchange resins at Rocky Flats Environmental Technology Site (RFETS). These resins have an elevated potential for ignition due to inadequate wetness and contact with nitrates. The work focused on the preparation and performance evaluation of several Portland cement/resin formulations. The performance standards were chosen to address Waste Isolation Pilot Plant and Environmental Protection Agency Resource Conservation and Recovery Act requirements, compatibility with Rocky Flats equipment, and throughput efficiency. The work was performed with surrogate gel-type Dowex cation- and anion-exchange resins chosen to be representative of the resin inventory at RFETS. Work was initiated with nonactinide resins to establish formulation ranges that would meet performance standards. Results were then verified and refined with actinide-containing resins. The final recommended formulation that passed all performance standards was determined to be a cement/water/resin (C/W/R) wt % ratio of 63/27/10 at a pH of 9 to 12. The recommendations include the acceptable compositional ranges for each component of the C/W/R ratio. Also included in this report are a recommended procedure, an equipment list, and observations/suggestions for implementation at RFETS. In addition, information is included that explains why denitration of the resin is unnecessary for stabilizing its ignitability potential

  18. Saltstone: cement-based waste form for disposal of Savannah River Plant low-level radioactive salt waste

    International Nuclear Information System (INIS)

    Defense waste processing at the Savannah River Plant will include decontamination and disposal of approximately 400 million liters of waste containing NaNO3, NaOH, Na2SO4, and NaNO2. After decontamination, the salt solution is classified as low-level waste. A cement-based waste form, saltstone, has been designed for disposal of Savannah River Plant low-level radioactive salt waste. Bulk properties of this material have been tailored with respect to salt leach rate, permeability, and compressive strength. Microstructure and mineralogy of leached and unleached specimens were characterized by SEM and x-ray diffraction analyses. The disposal system for the DWPF salt waste includes reconstitution of the crystallized salt as a solution containing 32 wt % solids. This solution will be decontaminated to remove 137Cs and 90Sr and then stabilized in a cement-based waste form. Laboratory and field tests indicate that this stabilization process greatly reduces the mobility of all of the waste constitutents in the surface and near-surface environment. Engineered trenches for subsurface burial of the saltstone have been designed to ensure compatibility between the waste form and the environment. The total disposal sytem, saltstone-trench-surrounding soil, has been designed to contain radionuclides, Cr, and Hg by both physical encapsulation and chemical fixation mechanisms. Physical encapsulation of the salts is the mechanism employed for controlling N and OH releases. In this way, final disposal of the SRP low-level waste can be achieved and the quality of the groundwater at the perimeter of the disposal site meets EPA drinking water standards

  19. Saltstone: cement-based waste form for disposal of Savannah River Plant low-level radioactive salt waste

    International Nuclear Information System (INIS)

    Defense waste processing at the Savannah River Plant will include decontamination and disposal of approximately 400 million liters of waste containing NaNO3, NaOH, Na2SO4, and NaNO2. After decontamination, the salt solution is classified as low-level waste. A cement-based waste form, saltstone, has been designed for disposal of Savannah River Plant low-level radioactive salt waste. Bulk properties of this material have been tailored with respect to salt leach rate, permeability, and compressive strength. Microstructure and mineralogy of leached and unleached specimens were characterized by SEM and x-ray diffraction analyses. The disposal system for the DWPF salt waste includes reconstitution of the crystallized salt as a solution containing 32 wt % solids. This solution will be decontaminated to remove 137Cs and 90Sr and then stabilized in a cement-based waste form. Laboratory and field tests indicate that this stabilization process greatly reduces the mobility of all of the waste constitutents in the surface and near surface environment. Engineered trenches for subsurface burial of the saltstone have been designed to ensure compatibility between the waste form and the environment. The total disposal system, saltstone-trench- surrounding soil, has been designed to contain radionuclides, Cr, and Hg by both physical encapsulation and chemical fixation mechanisms. Physical encapsulation of the salts is the mechanism employed for controlling N and OH releases. In this way, final disposal of the SRP low-level waste can be achieved and the quality of the ground water at the perimeter of the disposal site meets EPA drinking water standards

  20. Bitumen coating as a tool for improving the porosity and chemical stability of simulated cement-waste forms

    International Nuclear Information System (INIS)

    Coating process of simulated cement-based waste form with bitumen was evaluated by performing physical and chemical experimental tests. X-ray diffraction (X-RD), Fourier transform infrared spectroscopy (FT-IR) and electron microscope investigations were applied on coated and non-coated simulated waste forms. Experimental results indicated that coating process improved the applicable properties of cement-based waste form such as porosity and leachability. Diffusion coefficients and leach indecies of coated specimens were calculated and show acceptable records. It could be stated that coating cemented waste form by bitumen emulsion, isolate the radioactive contaminants, thus reduces their back release to surrounding and in consequently save the environment proper and safe

  1. Mathematical modelling of the corrosion and leaching behaviour of cemented waste forms

    International Nuclear Information System (INIS)

    A theoretical model is presented which allows to calculate the leaching of radionuclides and the corrosion of cemented waste forms in contact with water or brine. The model computes both the behaviour of specimens in laboratory-scale experiments and provides a forecast of the behaviour of waste forms in the case of an accidental drowning of a repository. The mathematical formalism employed describes leaching and corrosion on the basis of diffusion and dissolution processes and of chemical reactions. The mathematical formalism is coded in FORTRAN77. This report includes the documentation of the 'DIFMOD' computer code with the associated 'DIFPLO' plot program and the input manual of both programs. Finally application of the model is demonstrated by some examples allowing interpretation of experimental data. (orig.)

  2. Integral migration and source term experiments on cement and bitumen waste forms

    International Nuclear Information System (INIS)

    This is the final report of a programme of research which formed a part of the CEC joint research project into radionuclide migration in the geosphere (MIRAGE). This study addressed the aspects of integral migration and source term. The integral migration experiment simulated, in the laboratory, the intrusion of water into the repository, the leaching of radionuclides from two intermediate level wasteforms and the subsequent migration through the geosphere. The simulation consisted of a source of natural ground water which flowed over a sample of wasteform, at a controlled redox potential, and then through backfill and geological material packed in columns. The two wasteforms used here were cemented waste from the WAK plant at Karlsruhe, W. Germany and bitumenised intermediate concentrates from the Marcoule plant in France. The soluble fission products such as caesium wire rapidly released from the cemented waste but the actinides, and technetium in the reduced state, were retained in the wasteform. The release of all nuclides from the bitumenised waste was very low. (author)

  3. Development of methodology to evaluate microbially influenced degradation of cement-solidified low-level radioactive waste forms

    International Nuclear Information System (INIS)

    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 US 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

  4. EVALUATION OF ORGANIC VAPOR RELEASE FROM CEMENT-BASED WASTE FORMS

    International Nuclear Information System (INIS)

    A cement based waste form was evaluated to determine the rates at which various organics were released during heating caused by the cementitious heat-of-hydration reaction. Saltstone is a cement-based waste form for the disposal of low-level salt solution. Samples were prepared with either Isopar(regsign) L, a long straight chained hydrocarbon, or (Cs,K) tetraphenylborate, a solid that, upon heating, decomposes to benzene and other aromatic compounds. The saltstone samples were heated over a range of temperatures. Periodically, sample headspaces were purged and the organic constituents were captured on carbon beds and analyzed. Isopar(regsign) L was released from the saltstone in a direct relationship to temperature. An equation was developed to correlate the release rate of Isopar(regsign) L from the saltstone to the temperature at which the samples were cured. The release of benzene was more complex and relied on both the decomposition of the tetraphenylborate as well as the transport of the manufactured benzene through the curing saltstone. Additional testing with saltstone prepared with different surface area/volume also was performed

  5. Radiolysis in cement-based materials ; application to radioactive waste-forms

    International Nuclear Information System (INIS)

    Cement-based materials appear to be an original environment with respect to radiolysis, due to their intrinsic complexity (porous, multiphasic and evolutional medium) or their very specific physico-chemical conditions (hyper-alkaline medium with pH ≥ 13, high content in calcium) or by the fact of numerous couplings existing between different phenomenologies. At the level of a radioactive cemented wasteform, a high degree of complexity is reached, in particular if the system communicates with the atmosphere (open system allowing regulation of the pressures but also the admission of O2, strong reactive with regards to radiolysis). Then, the radiolysis description exceeds widely the only one aspect of the decomposition of alkaline water under irradiation and makes necessary a global phenomenological approach. In this context, some 'outlying' phenomena, highly coupled with radiation chemistry, have to be taken into account because they contribute to deeply modify the net result of the radiolysis: radioactive decay of multiple αβγ emitters with filiation, phase changes (for example H2 aq → H2 gas) within the pores, gas transport by convection (Darcy law) and by diffusion (Fick law), precipitation/dissolution of solid phases, effect of the ionic strength and the temperature, disturbances connected to the presence of some solutes with redox potentialities (iron, sulphur). The integration work carried out on the previous points leads to an operational model (DOREMI) allowing the estimate of H2 amounts produced by radiolysis in different cemented radioactive waste-forms. As the final expression of the model, numerical simulations constitute a relevant tool of expertise and prospecting, contributing to accompany the thought on radiolysis in cement matrices in general and in cemented waste-forms in particular. Starting from different examples, simulations can be so used in order to test some hypotheses or illustrate the greatest influence of gas transport, dose rate

  6. A procedure to evaluate the potential for microbially influenced degradation of cement-solidified low-level radioactive waste forms

    International Nuclear Information System (INIS)

    Because of its apparent structural integrity, cement has been widely used in the US 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 US 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

  7. Hygroscopic water uptake and dry/wet cycling compared with normal water exposure of cemented waste forms

    International Nuclear Information System (INIS)

    Considerable water uptake, formation of free liquid and leaching of activity may take place if cemented waste materials containing soluble salts are exposed to high humidity air. Another feature of importance for the structural stability of some materials is cyclic variations between water saturation and dry-out periods. The phenomena have been investigated on a laboratory scale using simulated waste in the form of cemented sodium nitrate or ion-exchange resins. Spherical samples are found to be valuable for simultaneous determination of the leaching of major components, mainly sodium, and the water uptake and swelling obtained from weight-increase measurements for the samples weighed in air and immersed in water

  8. Mixed waste solidification testing on polymer and cement-based waste forms in support of Hanford's WRAP 2A facility

    International Nuclear Information System (INIS)

    A testing program has been conducted by the Westinghouse Hanford Company to confirm the baseline waste form selection for use in Waste Receiving and Processing (WRAP) Module 2A. WRAP Module 2A will provide treatment required to properly dispose of containerized contact-handled, mixed low-level waste at the US Department of Energy Hanford Site in south-central Washington State. Solidification/stabilization has been chosen as the appropriate treatment for this waste. This work is intended to test cement-based, thermosetting polymer, and thermoplastic polymer solidification media to substantiate the technology approach for WRAP Module 2A. Screening tests were performed using the major chemical constituent of each waste type to measure the gross compatibility with the immobilization media and to determine formulations for more detailed testing. Surrogate materials representing each of the eight waste types were prepared in the laboratory. These surrogates were then solidified with the selected immobilization media and subjected to a battery of standard performance tests. Detailed discussion of the laboratory work and results are contained in this report

  9. Mixed waste solidification testing on polymer and cement-based waste forms in support of Hanford's WRAP 2A facility

    International Nuclear Information System (INIS)

    A testing program has been conducted by the Westinghouse Hanford Company to confirm the baseline waste form selection for use in Waste Receiving and Processing (WRAP) Module 2A. WRAP Module 2A will provide treatment required to properly dispose of containerized contact-handled, mixed low-level waste at the US Department of Energy Hanford Site in south-central Washington State. Solidification/stabilization has been chosen as the appropriate treatment for this waste. This work will test cement-based, thermosetting polymer, and thermoplastic polymer solidification media to substantiate the technology approach for WRAP Module 2A. Screening tests were performed using the major chemical constituent of each waste type to measure the gross compatibility with the immobilization media and to determine formulations for more detailed testing. Surrogate materials representing each of the eight waste types were prepared in the laboratory. These surrogates were then solidified with the selected immobilization media and subjected to a battery of standard performance tests. Detailed discussion of the laboratory work and results are contained in this paper

  10. Methods and Production of Cementation Materials for Immobilisation into Waste Form. Research of Cementation Processes for Specific Liquid Radioactive Waste Streams of Radiochemical Plants

    International Nuclear Information System (INIS)

    In the near future Russian Federation is planning to use industrial cementation facilities at two radiochemical combines - PA 'Mayak' and Mountain Chemical Combine. Scope of the research within the IAEA CRP contact No. 14176 included the development of cementation processes for specfic liquid radioactive waste streams that are present in these enterprisers. The research on cementation of liquid waste from spent nuclear fuel reprocessing at PA 'Mayak' allowed obtaining experimental data characterizing the technological process and basic characteristics of the produced cement compounds (e.g. mechanical strength, water resistance, frost resistance, flowability, etc.) immobilizing different streams of waste (e.g. hydrated-salt sludges, filter material pulps, mixture of hydrated salt slurries and filter material pulps, tritium liquid waste). Determined optimum technological parameters will allow industrial scale production of cement compound with required quality and higher flowability that is necessary for providing uniform filling of compartments of storage facilities at these sites. The research has been also carried out for the development of cementation technology for immobilization of pulps from storage tanks of Mountain Chemical Combine radiochemical plant. Cementation of such pulps is a difficult technological task because pulps are of complex chemical composition (e.g. hydroxides of manganese, iron, nickel, etc., as well as silicon oxide) and a relatively high activity. The research of cementation process selection for these pulps included studies of the impact of sorbing additive type and content on cement compounds leachability, flowability, impact of cement compound age to its mechanical strength, heat generation of cement compounds and others. The research results obtained allowed testing of cementation facility with a pulse type mixer on the full-scale. Use of such mixer for pulp cementation makes possible to prepare a homogeneous cement compound with the

  11. Conditioning of radioactive waste solutions by cementation

    International Nuclear Information System (INIS)

    For the cementation of the low and intermediate level evaporator concentrates resulting from the reprocessing of spent fuel numerous experiments were performed to optimize the waste form composition and to characterize the final waste form. Concerning the cementation process, properties of the waste/cement suspension were investigated. These investigations include the dependence of viscosity, bleeding, setting time and hydration heat from the waste cement slurry composition. For the characterization of the waste forms, the mechanical, thermal and chemical stability were determined. For special cases detailed investigations were performed to determine the activity release from waste packages under defined mechanical and thermal stresses. The investigations of the interaction of the waste forms with aqueous solutions include the determination of the Cs/Sr release, the corrosion resistance and the release of actinides. The Cs/Sr release was determined in dependence of the cement type, additives, setting time and sample size. (orig./DG)

  12. Cement As a Waste Form for Nuclear Fission Products: The Case of (90)Sr and Its Daughters.

    Science.gov (United States)

    Dezerald, Lucile; Kohanoff, Jorge J; Correa, Alfredo A; Caro, Alfredo; Pellenq, Roland J-M; Ulm, Franz J; Saúl, Andrés

    2015-11-17

    One of the main challenges faced by the nuclear industry is the long-term confinement of nuclear waste. Because it is inexpensive and easy to manufacture, cement is the material of choice to store large volumes of radioactive materials, in particular the low-level medium-lived fission products. It is therefore of utmost importance to assess the chemical and structural stability of cement containing radioactive species. Here, we use ab initio calculations based on density functional theory (DFT) to study the effects of (90)Sr insertion and decay in C-S-H (calcium-silicate-hydrate) in order to test the ability of cement to trap and hold this radioactive fission product and to investigate the consequences of its β-decay on the cement paste structure. We show that (90)Sr is stable when it substitutes the Ca(2+) ions in C-S-H, and so is its daughter nucleus (90)Y after β-decay. Interestingly, (90)Zr, daughter of (90)Y and final product in the decay sequence, is found to be unstable compared to the bulk phase of the element at zero K but stable when compared to the solvated ion in water. Therefore, cement appears as a suitable waste form for (90)Sr storage. PMID:26513644

  13. A mixing device for nuclear liquid effluents immobilization in a cement-waste form

    International Nuclear Information System (INIS)

    A study concerning the replacement of a mixing device for low and middle level activity aqueous effluents immobilization in a cement-waste form is presented. Currently, this operation is carried out in an agitated vessel then the concrete form is poured into the metallic drum through a glove box. Once the concrete has dried, the drum is closed and the whole forms the low-activity package that is next stored. The aim of the study is to prove that preparing the grout directly in the drum not only is possible but also brings some advantages. First of all, a comparison of three kinds of stirrer is made, and the choice of one of them explained. Then the implementation of this device in a non-nuclear experimental plant is described with all the adaptations required to a nuclear environment that is to say the necessity to keep radioactivity confined and an easy maintenance of the device. The first results regarding the quality of the final package are presented. From the present work, the following conclusions can be drawn: -) a rotor-stator agitator as been selected to meet the standards regarding the quality of the produced concrete block; -) a solution has been found to prevent the cardboard ring from collapsing or being damaged and the replacement of it by a cardboard drum with a plastic internal lining seems a valuable choice; -) a vessel head has been designed to prevent the grout from spilling over and to fill the drum as much as possible; -) the access of the different parts of the equipment is suitable for a glove box work, regarding the maintenance and cleaning aspects of the device. We started to define an experimental procedure to carry out the operation properly when the device will replace the actual mixer. Some adaptations to implement the device in a nuclear environment will be made regarding the screws and bolts of the equipment. Improvements have to be made regarding the cleaning of the inside of the vessel head. The use of a high pressure cleaning

  14. Stabilization of ZnCl2-containing wastes using calcium sulfoaluminate cement: leaching behaviour of the solidified waste form, mechanisms of zinc retention.

    Science.gov (United States)

    Berger, Stéphane; Cau Dit Coumes, Céline; Champenois, Jean-Baptiste; Douillard, Thierry; Le Bescop, Patrick; Aouad, Georges; Damidot, Denis

    2011-10-30

    To assess the potential of calcium sulfoaluminate cement to solidify and stabilize wastes containing high amounts of soluble zinc chloride (a strong inhibitor of Portland cement hydration), a simulated cemented waste form was submitted to leaching by pure water at a fixed pH of 7 for three months, according to a test designed to understand the degradation processes of cement pastes. Leaching was controlled by diffusion. The zinc concentration in the leachates always remained below the detection limit (2 μmol/L), showing the excellent confining properties of the cement matrix. At the end of the experiment, the solid sample exhibited three zones which were accurately characterized: (i) a highly porous and friable surface layer, (ii) a less porous intermediate zone in which several precipitation and dissolution fronts occurred, and (iii) the sound core. Ettringite was a good tracer for degradation. The good retention of zinc by the cement matrix was mainly attributed to the precipitation of a hydrated and well crystallized phase with platelet morphology (which may belong to the layered double hydroxide family) at early age (≤ 1 day), and to chemisorption onto aluminum hydroxide at later age. PMID:21889842

  15. Cements in Radioactive Waste Disposal

    International Nuclear Information System (INIS)

    The use of cement and concrete to immobilise radioactive waste is complicated by the wide- ranging nature of inorganic cementing agents available as well as the range of service environments in which cement is used and the different functions expected of cement. For example, Portland cement based concretes are widely used as structural materials for construction of vaults and tunnels. These constructions may experience a long pre-closure performance lifetime during which they are required to protect against collapse and ingress of water: strength and impermeability are key desirable characteristics. On the other hand, cement and concrete may be used to form backfills, ranging in permeability. Permeable formulations allow gas readily to escape, while impermeable barriers retard radionuclide transport and reduce access of ground water to the waste. A key feature of cements is that, while fresh, they pass through a fluid phase and can be formed into any shape desired or used to infiltrate other materials thereby enclosing them into a sealed matrix. Thereafter, setting and hardening is automatic and irreversible. Where concrete is used to form structural elements, it is also natural to use cement in other applications as it minimises potential for materials incompatibility. Thus cement- mainly Portland cement- has been widely used as an encapsulant for storage, transport and as a radiation shield for active wastes. Also, to form and stabilise structures such as vaults and silos. Relative to other potential matrices, cement also has a chemical immobilisation potential, reacting with and binding with many radionuclides. The chemical potential of cements is essentially sacrificial, thus limiting their performance lifetime. However performance may also be required in the civil engineering sense, where strength is important, so many factors, including a geochemical description of service conditions, may require to be assessed in order to predict performance lifetime. The

  16. Kinetic analysis of data obtained from studies on microbial degradation of cement waste forms, using shrinking core models.

    Science.gov (United States)

    Idachaba, M A; Nyavor, K; Egiebor, N O

    2003-04-01

    Model equations based on analytical solutions of two shrinking core models (acid dissolution or shrinking unreacted core (SUC) model, and bulk diffusion model), were used to analyze the kinetics of microbial degradation of cement waste forms. Two current approaches of waste form microbial stability evaluation (Nuclear Regulatory Commission (NRC) method and refined biofilm formation) were used to generate the data. Good linear correlations with R(2)>0.95 were obtained for the leaching data from both the NRC and biofilm approaches, using the model equation based on the bulk diffusion concept. Analyses using the model equation based on the acid dissolution model generally gave poor correlations except when data obtained from biofilm formation method was normalized. PMID:12686024

  17. Kinetic analysis of data obtained from studies on microbial degradation of cement waste forms, using shrinking core models

    International Nuclear Information System (INIS)

    Model equations based on analytical solutions of two shrinking core models (acid dissolution or shrinking unreacted core (SUC) model, and bulk diffusion model), were used to analyze the kinetics of microbial degradation of cement waste forms. Two current approaches of waste form microbial stability evaluation (Nuclear Regulatory Commission (NRC) method and refined biofilm formation) were used to generate the data. Good linear correlations with R2>0.95 were obtained for the leaching data from both the NRC and biofilm approaches, using the model equation based on the bulk diffusion concept. Analyses using the model equation based on the acid dissolution model generally gave poor correlations except when data obtained from biofilm formation method was normalized

  18. Sets of Reports and Articles Regarding Cement Wastes Forms Containing Alpha Emitters that are Potentially Useful for Development of Russian Federation Waste Treatment Processes for Solidification of Weapons Plutonium MOX Fuel Fabrication Wastes for

    Energy Technology Data Exchange (ETDEWEB)

    Jardine, L J

    2003-06-12

    This is a set of nine reports and articles that were kindly provided by Dr. Christine A. Langton from the Savannah River Site (SRS) to L. J. Jardine LLNL in June 2003. The reports discuss cement waste forms and primarily focus on gas generation in cement waste forms from alpha particle decays. However other items such as various cement compositions, cement product performance test results and some cement process parameters are also included. This set of documents was put into this Lawrence Livermore National Laboratory (LLNL) releasable report for the sole purpose to provide a set of documents to Russian technical experts now beginning to study cement waste treatment processes for wastes from an excess weapons plutonium MOX fuel fabrication facility. The intent is to provide these reports for use at a US RF Experts Technical Meeting on: the Management of Wastes from MOX Fuel Fabrication Facilities, in Moscow July 9-11, 2003. The Russian experts should find these reports to be very useful for their technical and economic feasibility studies and the supporting R&D activities required to develop acceptable waste treatment processes for use in Russia as part of the ongoing Joint US RF Plutonium Disposition Activities.

  19. Cementation of radioactive liquid scintillator waste simulate

    International Nuclear Information System (INIS)

    Liquid scintillation counting is an important analytical tool with extensive applications in medicine and basic applied research and used in quantification of □ -particles, weak □ and x-rays. The generated spent liquid scintillator radioactive waste should be limited and controlled to protect man and his environment. In this study, the radioactive spent liquid scintillator waste simulate (SLS) was immobilized in cement matrix using a surfactant in order to facilitate and increase the amount of SLS incorporated into the cementitious materials. Mechanical properties of the final cement waste form were acceptable for blocks containing up to 20% SLS in presence of surfactant. X-ray diffraction, IR analysis and scanning electron microscope proved that the hydration of cement materials is not significantly affected by organic scintillator waste. Therefore, the cement matrix could be recommended for solidification of SLS for the acceptable mechanical, physical and chemical characterizations reached.

  20. Cement-based radioactive waste hosts formed under elevated temperatures and pressures (FUETAP concretes) for Savannah River Plant high-level defense waste

    International Nuclear Information System (INIS)

    Concretes that are formed under elevated temperatures and pressures (called FUETAP) are effective hosts for high-level radioactive defense wastes. Tailored concretes developed at the Oak Ridge National Laboratory (ORNL) have been prepared from common Portland cements, fly ash, sand, clays, and waste products. These concretes are produced by accelerated curing under mild autoclave conditions (85 to 2000C, 0.1 to 1.5 MPa) for 24 h. The solids are subsequently dewatered (to remove unbound water) at 2500C for 24 h. The resulting products are strong (compressive strength, 40 to 100 MPa), leach resistant [plutonium leaches at the rate of 10 pg/(cm2.d)], and radiolytically stable, monolithic waste forms (total gas value = 0.005 molecule/100 eV). This report summarizes the results of a 4-year FUETAP development program for Savannah River Plant (SRP) high-level defense wastes. It addresses the major questions concerning the performance of concretes as radioactive waste forms. These include leachability, radiation stability, thermal stability, thermal conductivity, impact strength, permeability, phase complexity, and effect of waste composition

  1. Mixed waste solidification testing on thermosetting polymer and cement based waste forms in support of Hanford's WRAP Module 2A Facility

    International Nuclear Information System (INIS)

    A testing program has been conducted by the Westinghouse Hanford Co. to confirm the baseline waste form selection for use in Waste Receiving and Processing (WRAP) Module 2A. WRAP Module 2A will provide treatment required to properly dispose of containerized contact-handled, mixed low-level waste at the US DOE Hanford Site in south-central Washington State. Solidification/stabilization has been chosen as the appropriate treatment for this waste. This work is intended to test cement-based and thermosetting polymer solidification media to confirm the baseline technologies selected for WRAP Module 2A. Screening tests were performed using the major chemical constituent of each waste type to measure the gross compatibility with the immobilization media and to determine formulations for more detailed testing. Surrogate wastes representing each of the eight waste types were prepared for testing. Surrogates for polymer testing were sent to a vendor commissioned for that portion of the test work. Surrogates for the grout testing were used in the Westinghouse Hanford Co. laboratory responsible for the grout performance testing. Detailed discussion of the lab. work and results are contained in this report

  2. Mixed waste solidification testing on thermosetting polymer and cement based waste forms in support of Hanford's WRAP Module 2A Facility

    International Nuclear Information System (INIS)

    A testing program has been conducted by the Westinghouse Hanford Company to confirm the baseline waste form selection for use in Waste Receiving and Processing (WRAP) Module 2A. WRAP Module 2A will provide treatment required to properly dispose of containerized contact-handled, mixed low-level waste at the US Department of Energy Hanford Site in south-central Washington State. Solidification/stabilization has been chosen as the appropriate treatment for this waste. This work is intended to test cement-based and thermosetting polymer solidification media to confirm the baseline technologies selected for WRAP Module 2A. Screening tests were performed using the major chemical constituent of each waste type to measure the gross compatibility with the immobilization media and to determine formulations for more detailed testing. Surrogate wastes representing each of the eight waste types were prepared for testing. Surrogates for polymer testing were sent to a vendor commissioned for that portion of the test work. Surrogates for the grout testing were used in the Westinghouse Hanford Company laboratory responsible for the grout performance testing. Detailed discussion of the laboratory work and results are contained in this report

  3. Cement encapsulation of uranyl nitrate waste

    International Nuclear Information System (INIS)

    During decontamination of the former nuclear fuel reprocessing plant at West Valley, New York, low-level radioactive waste streams are being identified which require disposal in an environmentally acceptable manner. One such waste stream, consisting essentially of uranyl nitrate, has been located in one of the processing cells. A study was conducted on this waste stream to determine if it could be stably encapsulated in cement. First, a recipe was developed for cement-encapsulating this highly acidic waste. Samples were then made to perform waste qualification testing as described in the NRC Branch Technical Position-Waste Form to determine the stability of this waste form. The testing showed that the waste form had a compressive strength much greater than the 345 kPA (50 psi) minimum guideline after room-temperature cure, irradiation, thermal cycling, immersion, and biodegradation. In addition, the encapsulated waste had uranium and cerium leachability index values greater than six, which is the minimum recommended by the NRC position paper. The cement-encapsulated uranyl nitrate waste thus met the NRC stability guidelines for the disposal of Class B and Class C radioactive wastes

  4. Cement-Based Materials for Nuclear Waste Storage

    CERN Document Server

    Cau-di-Coumes, Céline; Frizon, Fabien; Lorente, Sylvie

    2013-01-01

    As the re-emergence of nuclear power as an acceptable energy source on an international basis continues, the need for safe and reliable ways to dispose of radioactive waste becomes ever more critical. The ultimate goal for designing a predisposal waste-management system depends on producing waste containers suitable for storage, transportation and permanent disposal. Cement-Based Materials for Nuclear-Waste Storage provides a roadmap for the use of cementation as an applied technique for the treatment of low- and intermediate-level radioactive wastes.Coverage includes, but is not limited to, a comparison of cementation with other solidification techniques, advantages of calcium-silicate cements over other materials and a discussion of the long-term suitability and safety of waste packages as well as cement barriers. This book also: Discusses the formulation and production of cement waste forms for storing radioactive material Assesses the potential of emerging binders to improve the conditioning of problemati...

  5. Qualification of radioactive waste cement conditioning processes

    International Nuclear Information System (INIS)

    Nucleco Qualification Process Laboratory activities are focused on qualification of cement matrix conditioning processes of Low and Intermediate Level Waste produced by the decommissioning of old Nuclear Power Plants and research centres. Radioactive waste management strategies for Second- and Third Category wastes (according to the ENEA Technical Guide n. 26), involve specific processes (treatment and conditioning) aimed at producing a final waste form in which the radionuclides are incorporated into a solid matrix in order to reduce their potential migration or dispersion. The qualification of conditioning processes consists of all those activities demonstrating that the final waste form and waste package have the minimum requirements (mechanical, chemical and physical characteristics) compliant with all the subsequent management phases: long term interim storage, transport and long term disposal of the waste (in accordance with UNI 11193- 2006 standard). First, the paper recalls the classification into 3 categories of radioactive wastes by the Italian authorities. Cementation is one of the most common method for conditioning radioactive wastes into a solid, safe form suitable for long term storage. 3 tables list the qualification tests that are assigned to waste form, containers and final packages, the minimum requirements for second category wastes and the results of qualification tests

  6. Comparative waste forms study

    International Nuclear Information System (INIS)

    A number of alternative process and waste form options exist for the immobilization of nuclear wastes. Although data exists on the characterization of these alternative waste forms, a straightforward comparison of product properties is difficult, due to the lack of standardized testing procedures. The characterization study described in this report involved the application of the same volatility, mechanical strength and leach tests to ten alternative waste forms, to assess product durability. Bulk property, phase analysis and microstructural examination of the simulated products, whose waste loading varied from 5% to 100% was also conducted. The specific waste forms investigated were as follows: Cold Pressed and Sintered PW-9 Calcine; Hot Pressed PW-9 Calcine; Hot Isostatic Pressed PW-9 Calcine; Cold Pressed and Sintered SPC-5B Supercalcine; Hot Isostatic pressed SPC-5B Supercalcine; Sintered PW-9 and 50% Glass Frit; Glass 76-68; Celsian Glass Ceramic; Type II Portland Cement and 10% PW-9 Calcine; and Type II Portland Cement and 10% SPC-5B Supercalcine. Bulk property data were used to calculate and compare the relative quantities of waste form volume produced at a spent fuel processing rate of 5 metric ton uranium/day. This quantity ranged from 3173 L/day (5280 Kg/day) for 10% SPC-5B supercalcine in cement to 83 L/day (294 Kg/day) for 100% calcine. Mechanical strength, volatility, and leach resistance tests provide data related to waste form durability. Glass, glass-ceramic and supercalcine ranked high in waste form durability where as the 100% PW-9 calcine ranked low. All other materials ranked between these two groupings

  7. The suitability of a supersulfated cement for nuclear waste immobilisation

    Energy Technology Data Exchange (ETDEWEB)

    Collier, N.C., E-mail: nick.collier@sheffield.ac.uk [Immobilisation Science Laboratory, Department of Materials Science and Engineering, The University of Sheffield, Mappin Street, Sheffield S1 3JD (United Kingdom); Milestone, N.B. [Immobilisation Science Laboratory, Department of Materials Science and Engineering, The University of Sheffield, Mappin Street, Sheffield S1 3JD (United Kingdom); Callaghan Innovation, 69 Gracefield Road, PO Box 31310, Lower Hutt 5040 (New Zealand); Gordon, L.E. [Immobilisation Science Laboratory, Department of Materials Science and Engineering, The University of Sheffield, Mappin Street, Sheffield S1 3JD (United Kingdom); Geopolymer and Minerals Processing Group, Department of Chemical and Biomolecular Engineering, University of Melbourne, Parkville, Victoria 3010 (Australia); Ko, S.-C. [Holcim Technology Ltd, Hagenholzstrasse 85, CH-8050 Zurich (Switzerland)

    2014-09-15

    Highlights: • We investigate a supersulfated cement for use as a nuclear waste encapsulant. • High powder fineness requires a high water content to satisfy flow requirements. • Heat generation during hydration is similar to a control cement paste. • Typical hydration products are formed resulting in a high potential for waste ion immobilisation. • Paste pH and aluminium corrosion is less than in a control cement paste. - Abstract: Composite cements based on ordinary Portland cement are used in the UK as immobilisation matrices for low and intermediate level nuclear wastes. However, the high pore solution pH causes corrosion of some metallic wastes and undesirable expansive reactions, which has led to alternative cementing systems being examined. We have investigated the physical, chemical and microstructural properties of a supersulfated cement in order to determine its applicability for use in nuclear waste encapsulation. The hardened supersulfated cement paste appeared to have properties desirable for use in producing encapsulation matrices, but the high powder specific surface resulted in a matrix with high porosity. Ettringite and calcium silicate hydrate were the main phases formed in the hardened cement paste and anhydrite was present in excess. The maximum rate of heat output during hydration of the supersulfated cement paste was slightly higher than that of a 9:1 blastfurnace slag:ordinary Portland cement paste commonly used by the UK nuclear waste processing industry, although the total heat output of the supersulfated cement paste was lower. The pH was also significantly lower in the supersulfated cement paste. Aluminium hydroxide was formed on the surface of aluminium metal encapsulated in the cement paste and ettringite was detected between the aluminium hydroxide and the hardened cement paste.

  8. The suitability of a supersulfated cement for nuclear waste immobilisation

    Science.gov (United States)

    Collier, N. C.; Milestone, N. B.; Gordon, L. E.; Ko, S.-C.

    2014-09-01

    Composite cements based on ordinary Portland cement are used in the UK as immobilisation matrices for low and intermediate level nuclear wastes. However, the high pore solution pH causes corrosion of some metallic wastes and undesirable expansive reactions, which has led to alternative cementing systems being examined. We have investigated the physical, chemical and microstructural properties of a supersulfated cement in order to determine its applicability for use in nuclear waste encapsulation. The hardened supersulfated cement paste appeared to have properties desirable for use in producing encapsulation matrices, but the high powder specific surface resulted in a matrix with high porosity. Ettringite and calcium silicate hydrate were the main phases formed in the hardened cement paste and anhydrite was present in excess. The maximum rate of heat output during hydration of the supersulfated cement paste was slightly higher than that of a 9:1 blastfurnace slag:ordinary Portland cement paste commonly used by the UK nuclear waste processing industry, although the total heat output of the supersulfated cement paste was lower. The pH was also significantly lower in the supersulfated cement paste. Aluminium hydroxide was formed on the surface of aluminium metal encapsulated in the cement paste and ettringite was detected between the aluminium hydroxide and the hardened cement paste.

  9. Solidification of low-level radioactive wastes in masonry cement

    International Nuclear Information System (INIS)

    Portland cements are widely used as solidification agents for low-level radioactive wastes. However, it is known that boric acid wastes, as generated at pressurized water reactors (PWR's) are difficult to solidify using ordinary portland cements. Waste containing as little as 5 wt % boric acid inhibits the curing of the cement. For this purpose, the suitability of masonry cement was investigated. Masonry cement, in the US consists of 50 wt % slaked lime (CaOH2) and 50 wt % of portland type I cement. Addition of boric acid in molar concentrations equal to or less than the molar concentration of the alkali in the cement eliminates any inhibiting effects. Accordingly, 15 wt % boric acid can be satisfactorily incorporated into masonry cement. The suitability of masonry cement for the solidification of sodium sulfate wastes produced at boiling water reactors (BWR's) was also investigated. It was observed that although sodium sulfate - masonry cement waste forms containing as much as 40 wt % Na2SO4 can be prepared, waste forms with more than 7 wt % sodium sulfate undergo catastrophic failure when exposed to an aqueous environment. It was determined by x-ray diffraction that in the presence of water, the sulfate reacts with hydrated calcium aluminate to form calcium aluminum sulfate hydrate (ettringite). This reaction involves a volume increase resulting in failure of the waste form. Formulation data were identified to maximize volumetric efficiency for the solidification of boric acid and sodium sulfate wastes. Measurement of some of the waste form properties relevant to evaluating the potential for the release of radionuclides to the environment included leachability, compression strengths and chemical interactions between the waste components and masonry cement. 15 refs., 19 figs., 9 tabs

  10. Modelling the effects of waste components on cement hydration

    NARCIS (Netherlands)

    Eijk, van R.J.; Brouwers, H.J.H.

    2001-01-01

    Ordinary Portland Cement (OPC) is often used for the solidification/stabilization (S/S) of waste containing heavy metals and salts. These waste components will precipitate in the form of insoluble compounds on to unreacted cement clinker grains preventing further hydration. In this study the long te

  11. Radiobiological waste treatment-ashing treatment and immobilization with cement

    International Nuclear Information System (INIS)

    This report describes the results of the study on the treatment of radioactive biological waste in the China Institute for Radiation Protection (CIRP). The possibility of radiobiological waste treatment was investigated by using a RAF-3 type rapid ashing apparatus together with the immobilization of the resulted ash. This rapid ashing apparatus, developed by CIRP, is usually used for pretreatment of samples prior to chemical analysis and physical measurements. The results show that it can ash 3 kg of animal carcasses a batch, the ashing time is 5-7 h and the ash content is less than 4 wt%. The ashing temperature not exceeding 450 deg. C was used without any risk of high losses of radionuclides. The ash from the rapid ashing apparatus was demonstrated to be immobilized with ordinary silicate cement. The optimum cement/ash/water formulation of the cemented waste form was 35 ± 5 wt% cement, 29 ± 2 wt% water, and 36 ± 6 wt% ash. The performance of the waste form was in compliance with the technical requirements except for impact resistance. Mixing additives in immobilization formulations can improve the performance of the cemented ash waste form. The additives chosen were DH4A flow promoter as a cement additive and vermiculite or zeolite as a supplement. The recommended formulation, i.e. an improved formulation of the cemented ash waste form is that additives DH4A flow promoter and vermiculite (or zeolite) are added on the ground of optimum cement/ash/water formulation of the cemented waste form, the dosage of water, DH4A and vermiculite (or zeolite) is 70 wt%, 0.5 wt% and ≤ 5 wt% of the cement dosage, respectively. The cemented ash waste forms obtained meet all the requirements for disposal. (author). 12 refs, 7 figs, 13 tabs

  12. Comparison of modified sulfur cement and hydraulic cement for encapsulation of radioactive and mixed wastes

    International Nuclear Information System (INIS)

    The majority of solidification/stabilization systems for low-level radioactive waste (LLW) and mixed waste, both in the commercial sector and at Department of Energy (DOE) facilities, utilize hydraulic cement (such as portland cement) to encapsulate waste materials and yield a monolithic solid waste form for disposal. A new and innovative process utilizing modified sulfur cement developed by the US Bureau of Mines has been applied at Brookhaven National Laboratory (BNL) for the encapsulation of many of these ''problem'' wastes. Modified sulfur cement is a thermoplastic material, and as such, it can be heated above it's melting point (120 degree C), combined with dry waste products to form a homogeneous mixture, and cooled to form a monolithic solid product. Under sponsorship of the DOE, research and development efforts at BNL have successfully applied the modified sulfur cement process for treatment of a range of LLWs including sodium sulfate salts, boric acid salts, and incinerator bottom ash and for mixed waste contaminated incinerator fly ash. Process development studies were conducted to determine optimal waste loadings for each waste type. Property evaluation studies were conducted to test waste form behavior under disposal conditions by applying relevant performance testing criteria established by the Nuclear Regulatory Commission (for LLW) and the Environmental Protection Agency (for hazardous wastes). Based on both processing and performance considerations, significantly greater waste loadings were achieved using modified sulfur cement when compared with hydraulic cement. Technology demonstration of the modified sulfur cement encapsulation system using production-scale equipment is scheduled for FY 1991. 12 refs., 8 figs., 3 tabs

  13. Immobilization of radioactive waste in cement-based matrices

    International Nuclear Information System (INIS)

    Tobermorite and xonotlite, two synthetic calcium silicate hydrates, improve the Cs retention of cement matrices for Cs, when incorporated at the 6 to 10% level. A kinetic and mechanistic scheme is presented for the reaction of fine grained, Cs-loaded clinoptilolite with cement. The Magnox waste form reacts quickly with cement, leading to an exchange of carbonate between waste form and cement components. Carbonation of cements leads to a marked improvement in their physical properties of Cs retentivity. Diffusion models are presented for cement systems whose variable parameters can readily be derived from experimental measurements. Predictions about scaled-up behaviour of large immobilized masses are applied to extrapolation of laboratory scale results to full-size masses. (author)

  14. Modified sulfur cement solidification of low-level wastes

    International Nuclear Information System (INIS)

    This topical report describes the results of an investigation on the solidification of low-level radioactive wastes in modified sulfur cement. The work was performed as part of the Waste Form Evaluation Program, sponsored by the US Department of Energy's Low-Level Waste Management Program. Modified sulfur cement is a thermoplastic material developed by the US Bureau of Mines. Processing of waste and binder was accomplished by means of both a single-screw extruder and a dual-action mixing vessel. Waste types selected for this study included those resulting from advanced volume reduction technologies (dry evaporator concentrate salts and incinerator ash) and those which remain problematic for solidification using contemporary agents (ion exchange resins). Process development studies were conducted to ascertain optimal process control parameters for successful solidification. Maximum waste loadings were determined for each waste type and method of processing. Property evaluation testing was carried out on laboratory scale specimens in order to compare with waste form performance for other potential matrix materials. Waste form property testing included compressive strength, water immersion, thermal cycling and radionuclide leachability. Recommended waste loadings of 40 wt. % sodium sulfate and boric acid salts and 43 wt. % incinerator ash, which are based on processing and performance considerations, are reported. Solidification efficiencies for these waste types represent significant improvements over those of hydraulic cements. Due to poor waste form performance, incorporation of ion exchange resin waste in modified sulfur cement is not recommended

  15. PURIFIED WASTE FCC CATALYST AS A CEMENT REPLACEMENT MATERIAL

    Directory of Open Access Journals (Sweden)

    Danute Vaiciukyniene

    2015-06-01

    Full Text Available Zeolites are commonly used in the fluid catalytic cracking process. Zeolite polluted with oil products and became waste after some time used. The quantity of this waste inevitably rises by expanding rapidly oil industry. The composition of these catalysts depends on the manufacturer and on the process that is going to be used. The main factors retarding hydration process of cement systems and modifying them strength are organic compounds impurities in the waste FCC catalyst. The present paper shows the results of using purified waste FCC catalyst (pFCC from Lithuania oil refinery, as Portland cement replacement material. For this purpose, the purification of waste FCC catalyst (FCC samples was treated with hydrogen peroxide. Hydrogen peroxide (H2O2 is one of the most powerful oxidizers known. By acting of waste with H2O2 it can eliminate the aforementioned waste deficiency, and the obtained product becomes one of the most promising ingredients, in new advanced building materials. Hardened cement paste samples with FCC or pFCC were formed. It was observed that the pFCC blended cements developed higher strength, after 28 days, compared to the samples with FCC or reference samples. Typical content of Portland cement substituting does not exceed 30 % of mass of Portland cement in samples. Reducing the consumption of Portland cement with utilizing waste materials is preferred for reasons of environmental protection.

  16. Medium-active waste form characterization: the performance of cement-based systems. Task 3. Characterization of radioactive waste forms. A series of final reports (1985-89) No 1

    International Nuclear Information System (INIS)

    The properties of cement systems which contribute to their immobilization potential for radwastes are characterized. In the short term, both physical and chemical properties of the matrix contribute to the immobilization potential, but in the longer term, chemical effects dominate. Before the interactions of cement with wastes can be fully assessed and data extrapolated into the future, it is necessary to be able to describe quantitatively the behaviour of cements themselves. A rigorous framework, based for the development on accessible physico-chemical variables, has been constructed. The model, as presently developed, is capable of describing the future performance of cements when leached at ∼ 200C by relatively pure water. It embraces mainly six chemical components - Na2O, K2O, CaO, MgO, SiO2 and water - together with limited data on the effect of sulphate, SO4-2. The interaction of cements with inactive waste-stream constituents is described, principally sulphate and nitrate. The interaction between steel and cement is also re-examined. As a consequence of these studies, a firm scientific basis has been laid for modelling the behaviour of cemented systems at long ages, i.e., those beyond the period for which test data can be obtained

  17. The interaction between nuclear waste glass and cement

    International Nuclear Information System (INIS)

    The interaction between simulated reference waste glasses SON68 and SM539 and cement has been studied in suspensions of Ordinary Portland Cement and synthetic young cement water with pH 13.5 at 30 C. The cement appears to trigger glass dissolution by consumption of glass matrix components. This leads to fast glass dissolution at a constant rate with formation of a porous gel layer on the glass. This is probably due mostly to the reaction of Si from the glass with portlandite, forming CSH phases. After consumption of the portlandite, the glass alteration rate is expected to decrease. (authors)

  18. Immobilization of radioactive waste in cement based matrices

    International Nuclear Information System (INIS)

    The kinetics of reaction between cement and clinoptilolite are elucidated and rate equations containing temperature dependent constants derived for this reaction. Variations in clinoptilolite particle size and their consequences to reactivity are assessed. The presence of pozzolanic agents more reactive than clinoptilolite provides sacrificial agents which are partially effective in lowering the clinoptilolite reactivity. Blast furnace slag-cements have been evaluated and the background literature summarized. Experimental studies of the pore fluid in matured slag-cements show that they provide significantly more immobilization for Cs than Portland cement. The distribution of Sr in cemented waste forms has been examined, and it is shown that most of the chemical immobilization potential in the short term is likely to be associated with the aluminate phases. The chemical and structural nature of these are described. Carbonation studies on real cements are summarized. (author)

  19. Waste form development/test

    International Nuclear Information System (INIS)

    The main objective of this study is to investigate new solidification agents relative to their potential application to wastes generated by advanced high volume reduction technologies, e.g., incinerator ash, dry solids, and ion exchange resins. Candidate materials selected for the solidification of these wastes include a modified sulfur cement and low-density polyethylene, neither of which are currently employed commerically for the solidification of low-level waste (LLW). As both the modified sulfur cement and the polyethylene are thermoplastic materials, a heated screw type extruder is utilized in the production of waste form samples for testing and evaluation. In this regard, work is being conducted to determine the range of conditions under which these solidification agents can be satisfactorily applied to the specific LLW streams and to provide information relevant to operating parameters and process control

  20. Durability of incinerator ash waste encapsulated in modified sulfur cement

    International Nuclear Information System (INIS)

    Waste form stability under anticipated disposal conditions is an important consideration for ensuring continued isolation of contaminants from the accessible environment. Modified sulfur cement is a relatively new material and has only recently been applied as a binder for encapsulation of mixed wastes. Little data are available concerning its long-term durability. Therefore, a series of property evaluation tests for both binder and waste-binder combinations have been conducted to examine potential waste form performance under storage and disposal conditions. These tests include compressive strength, biodegradation, radiation stability, water immersion, thermal cycling, and leaching. Waste form compressive strength increased with ash waste loadings to 30.5 MPa at a maximum incinerator ash loading of 43 wt %. Biodegradation testing resulted in no visible microbial growth of either bacteria or fungi. Initial radiation stability testing did not reveal statistically significant deterioration in structural integrity. Results of 90 day water immersion tests were dependent on the type of ash tested. There were no statistically significant changes in compressive strength detected after completion of thermal cycle testing. Radionuclides from ash waste encapsulated in modified sulfur cement leached between 5 and 8 orders of magnitude slower than the leach index criterion established by the Nuclear Regulatory Commission (NRC) for low-level radioactive waste. Modified sulfur cement waste forms containing up to 43 wt % incinerator fly ash passed EPA Toxicity Characteristic Leaching Procedure (TCLP) criteria for lead and cadmium leachability. 11 refs., 2 figs., 5 tabs

  1. Comparative assessment of TRU waste forms and processes. Volume I. Waste form and process evaluations

    International Nuclear Information System (INIS)

    This study provides an assesses seven waste forms and eight processes for immobilizing transuranic (TRU) wastes. The waste forms considered are cast cement, cold-pressed cement, FUETAP (formed under elevated temperature and pressure) cement, borosilicate glass, aluminosilicate glass, basalt glass-ceramic, and cold-pressed and sintered silicate ceramic. The waste-immobilization processes considered are in-can glass melting, joule-heated glass melting, glass marble forming, cement casting, cement cold-pressing, FUETAP cement processing, ceramic cold-pressing and sintering, basalt glass-ceramic processing. Properties considered included gas generation, chemical durability, mechanical strength, thermal stability, and radiation stability. The ceramic products demonstrated the best properties, except for plutonium release during leaching. The glass and ceramic products had similar properties. The cement products generally had poorer properties than the other forms, except for plutonium release during leaching. Calculations of the Pu release indicated that the waste forms met the proposed NRC release rate limit of 1 part in 105 per year in most test conditions. The cast-cement process had the lowest processing cost, followed closely by the cold-pressed and FUETAP cement processes. Joule-heated glass melting had the lower cost of the glass processes. In-can melting in a high-quality canister had the highest cost, and cold-pressed and sintered ceramic the second highest. Labor and canister costs for in-can melting were identified. The major contributor to costs of disposing of TRU wastes in a defense waste repository is waste processing costs. Repository costs could become the dominant cost for disposing of TRU wastes in a commercial repository. It is recommended that cast and FUETAP cement and borosilicate glass waste-form systems be considered. 13 figures, 16 tables

  2. Utilization of Eucalyptus Oil Refineries Waste for Cement Particle Board

    Directory of Open Access Journals (Sweden)

    Rudi Setiadji

    2012-11-01

    Full Text Available Utilization of eucalyptus oil refinery waste in the manufacture of building material component of cement particle board is expected to reduce the price of housing units. This research used laboratory experimental methods, eucalyptus oil waste in the form of branches an twigs from eucalyptus tree. The variation of the testing were mixtures composition of the particle : cement, additives as accelerators, cold press load during manufacture of cement particle board. Cold press duration of cement board was 24 hours. The size of particle boards were (40 x 40 cm2 and 13 mm thick. The samples were tested for its density, water content, water absorption, flexural strength, thickness swelling, adhesion strength, and the nails pull out strength.

  3. Defining criteria for cemented waste produced from legacy liquids

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) has several hundred cubic metres of legacy radioactive waste stored in underground tanks at the Chalk River Laboratories (CRL) site in Chalk River, Ontario. As part of a larger campaign to reduce its legacy liabilities, AECL intends to remove and immobilize this waste using a cementation system. AECL plans to hire an external contractor to design and operate a cementation skid to remove and condition the liquid wastes. Clear and measurable waste form criteria must be determined and provided to the contractor in order for the contractor to demonstrate that a safe and stable waste form has been produced. AECL has reviewed industry-standard test methods and best practices related to cementation of liquid nuclear wastes. Where suitable, these test methods and practices have been incorporated into Product Performance Criteria. An extensive test program has been performed to obtain cement formulations for the legacy wastes; the resulting sample cemented wastes have been tested and the results compared to the Product Performance Criteria. Modifications to the criteria have been made as required based on knowledge gained during this process. In addition, since no industry standards had previously been identified to measure homogeneity, 3 potential test methods have been identified. Regardless of the amount of testing performed and the stringency of the performance criteria, some risk remains that the waste will deteriorate over time. However, by performing a rigorous review of industry practice and an extensive series of tests under various conditions, AECL believes that it has addressed the risks in a reasonable and prudent manner and has selected the appropriate Product Performance Criteria to achieve a safe and stable waste product

  4. Waste Form Evaluation Program. Final report

    International Nuclear Information System (INIS)

    This report presents data that can be used to assess the acceptability of polyethylene and modified sulfur cement waste forms to meet the requirements of 10 CFR 61. The waste streams selected for this study include dry evaporator concentrate salts and incinerator ash as representative wastes which result from advanced volume reduction technologies and ion exchange resins which remain problematic for solidification using commercially available matrix materials. Property evaluation tests such as compressive strength, water immersion, thermal cycling, irradiation, biodegradation and leachability were conducted for polyethylene and sulfur cement waste forms over a range of waste-to-binder ratios. Based on the results of the tests, optimal waste loadings of 70 wt % sodium sulfate, 50 wt % boric acid, 40 wt % incinerator ash and 30 wt % ion exchange resins were established for polyethylene, although maximum loadings were considerably higher. For modified sulfur cement, optimal loadings of 40 wt % sodium sulfate, 40 wt % boric acid and 40 wt % incinerator ash are reported. Ion exchange resins are not recommended for incorporation into modified sulfur cement because of poor waste form performance even at very low waste concentrations. The results indicate that all waste forms tested within the range of optimal waste concentrations satisifed the requirements of the NRC Technical Position Paper on Waste Form

  5. Mesoscopic structure of cerium waste loaded hydrated cement by SANS

    International Nuclear Information System (INIS)

    Cementation is one of the most commonly used methods for conditioning radioactive wastes. It provides a cost-effective solution for encapsulation of low and intermediate level radioactive wastes into suitable solid form for long term safety storage. Cerium is used for decontamination of alpha contaminated metallic waste and after this decontamination process, secondary wastes with corrosion products are created, which must be managed properly and cemented for near surface disposal. In the present work, modification of mesoscopic structure in hydrated cement due to addition of simulated cerium waste at different concentrations has been investigated by small-angle neutron scattering (SANS). Structural modifications, in mesoscopic length scale, have been observed. The scattering profiles for three kinds of cement blocks (virgin, 10 g/l and 20 g/l of corrosion product (C.P.) with 4 mm thickness) are shown. Data have been analyzed in the light of polydisperse spherical particles model assuming a log-normal distribution. Widely separated bimodal particle size distributions best represent the present data. Further, it has been observed that the scattering profile obeys power-law (Q-n) behaviour in two domains of Q, which reflects the self-similar/self-affined morphology of the inhomogeneities. Estimated parameters from SANS data are tabulated. A comparison is shown mentioning the value of scattering radius of gyration, exponent values (η) and average particle size for each kind of hydrated cement sample. (author)

  6. Development of test methods for quality control of LLW and MLW in cement or polymers (Parts 1 and 2). Task 3. Characterization of radioactive waste forms. A series of final reports (1985-1989) no. 39

    International Nuclear Information System (INIS)

    This report is divided into two parts. In the first part, the qualification of samples arising from the cementation of low (LLW) and intermediate level ( MLW) radioactive wastes is studied. In particular, bead ion exchange resins, filter sludges, BWR evaporator concentrates and decontamination solutions have been taken into account. The properties of the final waste forms have been compared with the ones of laboratory scale samples. The qualification of the solidified wastes was performed according to the requirements of the Italian Regulatory Body. Particular attention is devoted to mechanical and thermal properties, biodegradability and behaviour versus water. In the second part, the influence of different parameters on the leaching of Cesium from cemented BWR evaporator concentrates (sulfates) is tested. In particular the influence of the variation of temperature, initial concentration of the tracer, renewal and chemical composition of the leachant, size of the sample, has been tested. 20 refs., 68 figs., 21 tabs

  7. Leaching tests of cemented organic radioactive waste

    International Nuclear Information System (INIS)

    The use of radioisotopes in research, medical and industrial activities generates organic liquid radioactive wastes. At Centro de Desenvolvimento da Tecnologia Nuclear (CDTN) are produced organic liquid wastes from different sources, one of these are the solvent extraction activities, whose the waste volume is the largest one. Therefore a research was carried out to treat them. Several techniques to treat organic liquid radioactive wastes have been evaluated, among them incineration, oxidation processes, alkaline hydrolysis, distillation, absorption and cementation. Laboratory experiments were accomplished to establish the most adequate process in order to obtain qualified products for storage and disposal. Absorption followed by cementation was the procedure used in this study, i.e. absorbent substances were added to the organic liquid wastes before mixing with the cement. Initially were defined the absorbers, and evaluated the formulation in relation to the compressive strength of its products. Bentonite from different suppliers (B and G) and vermiculite in two granulometries (M - medium and F - small) were tested. In order to assess the product quality the specimens were submitted to the leaching test according the Standard ISO 6961 and its results were evaluated. Then they were compared with the values established by Standard CNEN NN 6.09 Acceptance criteria for waste products to be disposed, to verify if they meet the requirements for safely storage and disposal. Through this study the best formulations to treat the organic wastes were established. (author)

  8. Densified waste form and method for forming

    Energy Technology Data Exchange (ETDEWEB)

    Garino, Terry J.; Nenoff, Tina M.; Sava Gallis, Dorina Florentina

    2015-08-25

    Materials and methods of making densified waste forms for temperature sensitive waste material, such as nuclear waste, formed with low temperature processing using metallic powder that forms the matrix that encapsulates the temperature sensitive waste material. The densified waste form includes a temperature sensitive waste material in a physically densified matrix, the matrix is a compacted metallic powder. The method for forming the densified waste form includes mixing a metallic powder and a temperature sensitive waste material to form a waste form precursor. The waste form precursor is compacted with sufficient pressure to densify the waste precursor and encapsulate the temperature sensitive waste material in a physically densified matrix.

  9. Densified waste form and method for forming

    Energy Technology Data Exchange (ETDEWEB)

    Garino, Terry J.; Nenoff, Tina M.; Sava Gallis, Dorina Florentina

    2016-05-17

    Materials and methods of making densified waste forms for temperature sensitive waste material, such as nuclear waste, formed with low temperature processing using metallic powder that forms the matrix that encapsulates the temperature sensitive waste material. The densified waste form includes a temperature sensitive waste material in a physically densified matrix, the matrix is a compacted metallic powder. The method for forming the densified waste form includes mixing a metallic powder and a temperature sensitive waste material to form a waste form precursor. The waste form precursor is compacted with sufficient pressure to densify the waste precursor and encapsulate the temperature sensitive waste material in a physically densified matrix.

  10. Leaching of nuclear power reactor wastes forms

    International Nuclear Information System (INIS)

    The leaching tests for power reactor wastes carried out at IPEN/CNEN-SP are described. These waste forms consist mainly of spent resins and boric acid concentrates solidified in ordinary Portland cement. All tests were conducted according to the ISO and IAEA recommendations. 3 years leaching results are reported, determining cesium and strontium diffusivity coefficients for boric acid waste form and ion-exchange resins. (Author)

  11. Microscale Investigation of Arsenic Distribution and Species in Cement Product from Cement Kiln Coprocessing Wastes

    OpenAIRE

    Yufei Yang; Jingchuan Xue; Qifei Huang

    2013-01-01

    To improve the understanding of the immobilization mechanism and the leaching risk of Arsenic (As) in the cement product from coprocessing wastes using cement kiln, distribution and species of As in cement product were determined by microscale investigation methods, including electron probe microanalysis (EPMA) and X-ray absorption spectroscopy. In this study, sodium arsenate crystals (Na3AsO412H2O) were mixed with cement production raw materials and calcined to produce cement clinker. Then, ...

  12. Lysimeter study of commercial reactor waste forms: waste form acquisition characterization and full-scale leaching

    International Nuclear Information System (INIS)

    This report describes work conducted at Brookhaven National Laboratory (BNL) as part of a joint program with Savannah River Laboratory. Typical full-scale (55-gallon drum size) waste forms were acquired by BNL from a boiling water reactor (BWR) and a pressurized water reactor (PWR). Liquid waste stream activity concentrations were analyzed by gamma spectroscopy. This information was used to determine the waste from activity inventory, providing the necessary source term for lysimeter and leaching experiments. Predominant radionuclides of interest include 60Co, 137Cs, 134Cs, and 54Mn. A full-scale leaching experiment was initiated by BNL encompassing four representative waste stream-solidification agent combinations. Waste streams tested include PWR evaporator concentrate (boric acid waste), BWR evaporator concentrate (sodium sulfate waste) and BWR evaporator concentrate plus ion exchange resins. Solidification agents include masonry cement, portland type III cement, and vinyl ester-styrene (Dow polymer). Analyses of leachates indicate measurable leach rates of 137Cs, 134Cs, and 60Co from both BWR and PWR cement waste forms. The leach rates for both cesium isotopes in cement are at least two orders of magnitude greater than those for cobalt. Leachates from the BWR Dow polymer waste form include the same isotopes present in cement leachates, with the addition of 54Mn. Cesium leach rates from the Dow polymer waste form are approximately one order of magnitude lower than from an equivalent cement waste form. The 60Co cumulative fraction release, however, is approximately three times greater for the Dow polymer waste form

  13. Utilization of Industrial Borax Wastes (BW) for Portland Cement Production

    OpenAIRE

    ELBEYLİ, İffet YAKAR

    2004-01-01

    Industrial borax wastes (BWs) are formed as solid waste during the production of borax from tincal [Na2B4O5(OH)4.8H2O] in Bandırma, Turkey. These wastes cause different environmental problems and lead to economic losses because of high boron oxide (B2O3) content. The primary aim of this study is the removal of B2O3 from BWs and the second aim is the usage of BWs with low boron content in cement as an additive material. For this purpose, the BW was treated with water for removal of b...

  14. Cements in radioactive waste management. Characterization requirements of cement products for acceptance and quality assurance purposes

    International Nuclear Information System (INIS)

    Cementitious materials are used as immobilizing matrices for low (LLW) and medium-level wastes (MLW) and are also components of the construction materials in the secondary barriers and the repositories. This report has concerned itself with a critical assessment of the quality assurance aspects of the immobilization and disposal of MLW and LLW cemented wastes. This report has collated the existing knowledge of the use and potential of cementitious materials in radioactive waste immobilization and highlighted the physico-chemical parameters. Subject areas include an assessment of immobilization objectives and cement as a durable material, waste stream and matrix characterization, quality assurance concepts, nature of cement-based systems, chemistry and modelling of cement hydration, role and effect of blending agents, radwaste-cement interaction, assessment of durability, degradative and radiolytic processes in cements and the behaviour of cement-based matrices and their near-field interactions with the environment and the repository conditions

  15. Development of the Use of Alternative Cements for the Treatment of Intermediate Level Waste

    International Nuclear Information System (INIS)

    This paper describes initial development studies undertaken to investigate the potential use of alternative, non ordinary Portland cement (OPC) based encapsulation matrices to treat historic legacy wastes within the UK's Intermediate Level Waste (ILW) inventory. Currently these wastes are encapsulated in composite OPC cement systems based on high replacement with blast furnace slag of pulverised fuel ash. However, the high alkalinity of these cements can lead to high corrosion rates with reactive metals found in some wastes releasing hydrogen and forming expansive corrosion products. This paper therefore details preliminary results from studies on two commercial products, calcium sulfo-aluminate (CSA) and magnesium phosphate (MP) cement which react with a different hydration chemistry, and which may allow wastes containing these metals to be encapsulated with lower reactivity. The results indicate that grouts can be formulated from both cements over a range of water contents and reactant ratios that have significantly improved fluidity in comparison to typical OPC cements. All designed mixes set in 24 hours with zero bleed and the pH values in the plastic state were in the range 10-11 for CSA and 5-7 for MP cements. In addition, a marked reduction in aluminium corrosion rate has been observed in both types of cements compared to a composite OPC system. These results therefore provide encouragement that both cement types can provide a possible alternative to OPC in the immobilisation of reactive wastes, however further investigation is needed. (authors)

  16. Portland cement conditioning of the oil radioactive wastes

    International Nuclear Information System (INIS)

    Cementation is a widely used method to immobilize radioactive wastes generated during the operation of nuclear power plants. The oily radioactive wastes resulting during the normal service of Nuclear Power Plant at Cernavoda, Romania, can be conditioned in Portland cement as emulsions. In this way the interaction with cement water is not blocked. For this purpose, four compositions for conditioning were studied, namely: i) cement-emulsion; ii) cement-emulsion-sodium silicate; iii) cement-emulsion-sodium silicate-lime; iv) cement-emulsion-sand. The apparent density of hardened binding mixtures, setting time, compressive strength of hardened samples, leaching rate of tritium were determined. The results have shown that the oily wastes can be conditioned in a good manner when using Portland cement, emulsion conditioners, sodium silicate and lime. (authors)

  17. Ordinary Portland Cement matrix for solidification of cellulosic protective clothes hazardous wastes

    International Nuclear Information System (INIS)

    The used cellulosic protective clothes constitutes considerable fraction of the hazardous and radioactive wastes accumulated during the practical daily life. The direct solidification of these wastes with ordinary Portland cement resulted in waste forms having undesired characters, therefore, it is recommended to immobilize the secondary waste solutions coming from the oxidative degradation of the used protective clothes waste simulates rather than direct imbedding. IR analyses, X-ray diffraction and thermal characteristics for products of both direct encapsulation of the waste and the cementation of its degradation products were performed to evaluate the properties of the final waste cemented form before their disposal. Based on the results reached from X-ray diffraction, IR spectrograms and thermal analyses reports, it could be stated that no detectable changes in hydration and curing coarse of ordinary Portland cement when mixing the residual secondary waste solution resulting from the oxidative degradation of the used protective clothes waste simulate compared with mixing cement with water and in reverse with imbedding the unprocessed waste in cement matrix

  18. Cementation of Radioactive Waste from a PWR with Calcium Sulfoaluminate Cement

    International Nuclear Information System (INIS)

    Spent radioactive ion-exchange resin (SIER) and evaporation concentrates are radioactive wastes that are produced at by pressurized water reactor (PWR) nuclear power stations. Borate, which is used as a retardent for cement, is also present as a moderator in a PWR, therefore, borate will be present in both ion-exchange resins and evaporation concentrates. In this study the use of Calcium sulfoaluminate cements (SAC) as encapsulation medium for these waste streams was investigated. The study involved the manufacturing of different cement test samples with different amounts of SAC cement, waste resins (50% water content) and admixtures. In order to reduce hydration heat during 200 L solidification experiments, different admixtures were investigated. Initial results based on compressive strength tests and hydration temperature studies, indicated that zeolite was the best admixture for the current waste form. Experiments indicated that the addition of resin material into the current cement matrix reduces the hydration heat during curing Experimental results indicated that a combination of SAC (35 wt. %), zeolite (7 wt. %) mix with 42 wt. % resins (50% water content) and 16 wt. % of water forms a optimum cured monolith with low hydration heat. The microstructures of hydrated OPC, SAC and SAC with zeolite addition were studied using a Scanning Electron Microscopy (SEM). SEM results indicated that the SAC matrices consist of a needle type structure that changed gradually into a flake type structure with the addition of zeolite. Additionally, the presence of zeolite material inside the SAC matrix reduced the leaching rates of radionuclides significantly. In a final 200 L grouting test, measured results indicated a hydration temperature below 90oC withno thermal cracks after solidified. The influence of radiation on the compressive strength and possible gas generation (due to radiolysis) on cement waste forms containing different concentrations ion exchange resin was

  19. Radionuclide Retention in Concrete Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Mattigod, Shas V.; Bovaird, Chase C.; Wellman, Dawn M.; Wood, Marcus I.

    2010-09-30

    Assessing long-term performance of Category 3 waste cement grouts for radionuclide encasement requires knowledge of the radionuclide-cement interactions and mechanisms of retention (i.e., sorption or precipitation); the mechanism of contaminant release; the significance of contaminant release pathways; how waste form performance is affected by the full range of environmental conditions within the disposal facility; the process of waste form aging under conditions that are representative of processes occurring in response to changing environmental conditions within the disposal facility; the effect of waste form aging on chemical, physical, and radiological properties; and the associated impact on contaminant release. This knowledge will enable accurate prediction of radionuclide fate when the waste forms come in contact with groundwater. The information presented in the report provides data that 1) quantify radionuclide retention within concrete waste form materials similar to those used to encapsulate waste in the Low-Level Waste Burial Grounds (LLBG); 2) measure the effect of concrete waste form properties likely to influence radionuclide migration; and 3) quantify the stability of uranium-bearing solid phases of limited solubility in concrete.

  20. Leaching of nuclear power reactor waste forms

    International Nuclear Information System (INIS)

    The leaching tests for immobilized power reactor wastes carried out at IPEN are described. These wastes forms consist mainly of spent resins and boric acid concentrates solidified in ordinary Portland cement. All tests were conducted according to the ISO and IAEA recommendations. Three years leaching results are reported. The cesium diffuvity coefficients determined out of these results are about 1 x 10-8 cm2/s for boric acid waste form and 9 x 10-9 cm2/s for ion-exchange resin waste. Strontium diffusivity coefficients found are about 3 x 10-11 cm2/s and 9 x 10-11 cm2/s respectively. (Author)

  1. Immobilization of radioactive waste in cement based matrices

    International Nuclear Information System (INIS)

    Experimental and theoretical studies of hydrated cement systems are described. The behaviour of slag-based cement is described with a view to predicting their long term pH, Esub(n) and mineralogical balance. Modelling studies which enable the prediction at long ages of cement composites are advanced and a base model of the CaO-SiO2-H2O system presented. The behaviour of U and I in cements is explored. The tolerance of cement systems for a wide range of miscellaneous waste stream components and environmental hazards is described. The redox potential in cements is effectively lowered by irradiation. (author)

  2. Demonstration of Mixed Waste Debris Macroencapsulation Using Sulfur Polymer Cement

    Energy Technology Data Exchange (ETDEWEB)

    Mattus, C.H.

    1998-07-01

    This report covers work performed during FY 1997 as part of the Evaluation of Sulfur Polymer Cement Fast-Track System Project. The project is in support of the ``Mercury Working Group/Mercury Treatment Demonstrations - Oak Ridge`` and is described in technical task plan (TTP) OR-16MW-61. Macroencapsulation is the treatment technology required for debris by the U.S. Environmental Protection Agency Land Disposal Restrictions (LDR) under the Resource Conservation and Recovery Act. Based upon the results of previous work performed at Oak Ridge, the concept of using sulfur polymer cement (SPC) for this purpose was submitted to the Mixed Waste Focus Area (MWFA). Because of the promising properties of the material, the MWFA accepted this Quick Win project, which was to demonstrate the feasibility of macroencapsulation of actual mixed waste debris stored on the Oak Ridge Reservation. The waste acceptance criteria from Envirocare, Utah, were chosen as a standard for the determination of the final waste form produced. During this demonstration, it was shown that SPC was a good candidate for macroencapsulation of mixed waste debris, especially when the debris pieces were dry. The matrix was found to be quite easy to use and, once the optimum operating conditions were identified, very straightforward to replicate for batch treatment. The demonstration was able to render LDR compliant more than 400 kg of mixed wastes stored at the Oak Ridge National Laboratory.

  3. Cementation and solidification of miscellaneous mixed wastes at the Rocky Flats Environmental Technology Site

    International Nuclear Information System (INIS)

    The Rocky Flats Environmental Technology Site produces a variety of wastes which are amenable to micro-encapsulation in cement Portland cement is an inexpensive and readily available material for this application. The Waste Projects (WP) group at Rocky Flats evaluated cementation to determine its effectiveness in encapsulating several wastes. These included waste analytical laboratory solutions, incinerator ash, hydroxide precipitation sludge, and an acidic solution from the Delphi process (a chemical oxidation technology being evaluated as an alternative to incineration). WP prepared surrogate wastes and conducted designed experiments to optimize the cement formulation for the waste streams. These experiments used a Taguchi or factorial experimental design, interactions between the variables were also considered in the testing. Surrogate waste samples were spiked with various levels of each of six Resource Conservation and Recovery Act (RCRA) listed metals (Cd, Cr, Ba, Pb, Ni, and Ag), cemented using the optimized formulation, and analyzed for leach resistance using the Toxicity Characteristic Leaching Procedure (TCLP). The metal spike levels chosen were based on characterization data, and also based on an estimate of the highest levels of contaminants suspected in the waste. This paper includes laboratory test results for each waste studied. These include qualitative observations as well as quantitative data from TCLP analyses and environmental cycling studies. The results from these experiments show that cement stabilization of the different wastes can produce final waste forms which meet the current RCRA Land Disposal Restriction (LDR) requirements. Formulations that resulted in LDR compliant waste forms are provided. The volume increases associated with cementation are also lower than anticipated. Future work will include verification studies with actual mixed radioactive waste as well as additional formulation development studies on other waste streams

  4. A literature review of mixed waste components: Sensitivities and effects upon solidification/stabilization in cement-based matrices

    International Nuclear Information System (INIS)

    The US DOE Oak Ridge Field Office has signed a Federal Facility Compliance Agreement (FFCA) regarding Oak Ridge Reservation (ORR) mixed wastes subject to the land disposal restriction (LDR) provisions of the Resource conservation and Recovery Act. The LDR FFCA establishes an aggressive schedule for conducting treatability studies and developing treatment methods for those ORR mixed (radioactive and hazardous) wastes listed in Appendix B to the Agreement. A development, demonstration, testing, and evaluation program has been initiated to provide those efforts necessary to identify treatment methods for all of the wastes that meet Appendix B criteria. The program has assembled project teams to address treatment development needs in a variety of areas, including that of final waste forms (i.e., stabilization/solidification processes). A literature research has been performed, with the objective of determining waste characterization needs to support cement-based waste-form development. The goal was to determine which waste species are problematic in terms of consistent production of an acceptable cement-based waste form and at what concentrations these species become intolerable. The report discusses the following: hydration mechanisms of Portland cement; mechanisms of retardation and acceleration of cement set-factors affecting the durability of waste forms; regulatory limits as they apply to mixed wastes; review of inorganic species that interfere with the development of cement-based waste forms; review of radioactive species that can be immobilized in cement-based waste forms; and review of organic species that may interfere with various waste-form properties

  5. A literature review of mixed waste components: Sensitivities and effects upon solidification/stabilization in cement-based matrices

    Energy Technology Data Exchange (ETDEWEB)

    Mattus, C.H.; Gilliam, T.M.

    1994-03-01

    The US DOE Oak Ridge Field Office has signed a Federal Facility Compliance Agreement (FFCA) regarding Oak Ridge Reservation (ORR) mixed wastes subject to the land disposal restriction (LDR) provisions of the Resource conservation and Recovery Act. The LDR FFCA establishes an aggressive schedule for conducting treatability studies and developing treatment methods for those ORR mixed (radioactive and hazardous) wastes listed in Appendix B to the Agreement. A development, demonstration, testing, and evaluation program has been initiated to provide those efforts necessary to identify treatment methods for all of the wastes that meet Appendix B criteria. The program has assembled project teams to address treatment development needs in a variety of areas, including that of final waste forms (i.e., stabilization/solidification processes). A literature research has been performed, with the objective of determining waste characterization needs to support cement-based waste-form development. The goal was to determine which waste species are problematic in terms of consistent production of an acceptable cement-based waste form and at what concentrations these species become intolerable. The report discusses the following: hydration mechanisms of Portland cement; mechanisms of retardation and acceleration of cement set-factors affecting the durability of waste forms; regulatory limits as they apply to mixed wastes; review of inorganic species that interfere with the development of cement-based waste forms; review of radioactive species that can be immobilized in cement-based waste forms; and review of organic species that may interfere with various waste-form properties.

  6. Waste-form development

    International Nuclear Information System (INIS)

    Contemporary solidification agents are being investigated relative to their applications to major fuel cycle and non-fuel cycle low-level waste (LLW) streams. Work is being conducted to determine the range of conditions under which these solidification agents can be applied to specific LLW streams. These studies are directed primarily towards defining operating parameters for both improved solidification of problem wastes and solidification of new LLW streams generated from advanced volume reduction technologies. Work is being conducted to measure relevant waste form properties. These data will be compiled and evaluated to demonstrate compliance with waste form performance and shallow land burial acceptance criteria and transportation requirements

  7. Stabilization of high and low solids Consolidated Incinerator Facility (CIF) waste with super cement

    International Nuclear Information System (INIS)

    This report details solidification activities using selected Mixed Waste Focus Area technologies with the High and Low Solid waste streams. Ceramicrete and Super Cement technologies were chosen as the best possible replacement solidification candidates for the waste streams generated by the SRS incinerator from a list of several suggested Mixed Waste Focus Area technologies. These technologies were tested, evaluated, and compared to the current Portland cement technology being employed. Recommendation of a technology for replacement depends on waste form performance, process flexibility, process complexity, and cost of equipment and/or raw materials

  8. Solubility limits of radionuclides in interstitial water. Americium in cement. Task 3. Characterization of radioactive waste forms. A series of final reports (1985-89). No 34

    International Nuclear Information System (INIS)

    The migration of actinides inside cement (or concrete) is very slow, even when the material is saturated with water: precipitation of actinide hydroxide explains this retention phenomenon. The aim of this work is to measure Am solubility in aqueous solutions equilibrated with CPA55 cement to: (i) compare it with thermodynamic predictions; and (ii) correlate it to (future) migration measurements of Am through cement discs. 12 figs.; 8 tabs.; 3 refs

  9. The role of ceramics, cement and glass in the immobilization of radioactive wastes

    International Nuclear Information System (INIS)

    A brief account is given of the constitution and origin of nuclear waste. The immobilization of wastes is discussed: borosilicate glasses are considered as possible matrices; ceramic forms are dealt with in more detail. The principles of the use of ceramics are explained, with examples of different ceramic structures; cements are mentioned as being suitable for wet, medium- to low-active wastes. The effects of radiation on cement, ceramic and glass waste forms are indicated. The account concludes with 'summary and future progress'. (U.K.)

  10. Solidification of radioactive waste in a cement/lime mixture

    International Nuclear Information System (INIS)

    The suitability of a cement/lime mixture for use as a solidification agent for different types of wastes was investigated. This work includes studies directed towards determining the wasted/binder compositional field over which successful solidification occurs with various wastes and the measurement of some of the waste from properties relevant to evaluating the potential for the release of radionuclides to the environment. In this study, four types of low-level radioactive wastes were simulated for incorporation into a cement/lime mixture. These were boric acid waste, sodium sulfate wastes, aion exchange resins and incinerator ash. 7 references, 3 figures, 2 tables

  11. Activity diagrams for calcium/hydrogen, sodium/hydrogen, and potassium/hydrogen, and H4SiO4 and their relation to reactions in systems containing radioactive waste forms, cement, and rock in the presence of water

    International Nuclear Information System (INIS)

    In order to identify reactions which can occur in systems containing nuclear waste forms, cement, and repository rock in the presence of water, activity diagrams were calculated from free energies for aluminosilicates and calcium silicates. Groundwater compositions from candidate repository sites in the Palo Duro Basin of Texas, the Delaware Basin of New Mexico, and the Nevada Test Site were plotted on these diagrams. Essentially all of these are shown to be in the calcium zeolite field as shown on the diagram for calcium in the absence of other cations. Chlorite is shown to be stable in this region at the Mg and pH level of the Ogallala if the chlorite is high in iron, and at the Mg and pH level of the Wolfcamp low- or high-Fe chlorites are stable. Potassium and sodium mineral relationships fall in two categories, dilute waters and saline waters. Boreholes at Yucca Flat and Mercury Valley at the Nevada Test Site, and shallow ground water from the Rolling Plains north and east of the Palo Duro Basin are in equilibrium with kaolinite. The brines from the Salado and Rustler formations are in equilibrium with kaolinite and possibly also with sodium-potassium zeolite and illite. Leachates of cement and water, and cement, waste, and water were plotted on the calcium silicate activity diagram. These solutions are in equilibrium with calcium silicate hydrate hydrolysis reactions, with grossular and possibly with Ca-zeolites. Among the calcium silicates, calcium-silicate-hydrate gel (C-S-H gel) and tobermorite are the most likely candidates, but the thermodynamic data are not adequate to distinguish all the possibilities. 37 references, 4 figures, 3 tables

  12. Plant Test of Industrial Waste Disposal in a Cement Kiln

    Institute of Scientific and Technical Information of China (English)

    刘阳生; 韩杰; 等

    2003-01-01

    Destruction of industrial waste in cement rotary kilins(CRKs) is an alternative technology for the treatment of certain types of industrial waste(IW).In this paper,three typical types of industrial wastes were co-incinerated in the CRK at Beijing Cement Plant to determine the effects of waste disposal(especially solid waste disposal )on the quality of clinker and the concentration of pollutants in air emission.Experimental results show that(1) waste disposal does not affect the quality of clinker and fly ash,and fly ash after the IW disposal can still be used in the cement production,(2) heavy metals from IW are immobilized and stabilized in the clinker and cement,and (3) concentration of pollutants in air emission is far below than the permitted values in the China National Standard-Air Pollutants Emission Standard(GB 16297-1996).

  13. Mixture for solidification of liquid radioactive wastes into stable forms

    International Nuclear Information System (INIS)

    A mixture is proposed for cementing liquid radioactive wastes into chemically stable, mechanically strong, transportable and storable forms. The mixture consists of 60-80 wt.% Portland cement, 5-15 wt.% flue silica dust and 15-25 wt.% zeolitic tuffite. (Z.S.)

  14. Using portland cement for encapsulation of epipremnum aureum generated from phytoremediation process of liquid radioactive wastes

    International Nuclear Information System (INIS)

    Phyto remediation process was recommended for treatment of low and intermediate level liquid radioactive waste. Epipremnum aureum (golden pothas plant) was used to bioabsorbe, bioaccumulate and biostabilize Cs-137 and Co-60 from simulated waste solution containing both radionuclides. After the phyto remediation process, the collected golden pothas was solidified using portland cement aiming at complete and safe management scheme. In this part of work x-ray diffraction , infrared analysis and electron microscope examination as non-destructive techniques were used to evaluate the characteristics of obtained final waste forms of cemented golden pothas. In addition, mechanical, porosity and chemical optimizations were performed under various experimental parameters to asses the suitability of the two processes i.e. phyto remediation and cementation for managing these wastes categories. The experimental results obtained confirmed that encapsulation of 3 % dry ground golden pothas that collected from treatment process of radioactive waste solution, in cement materials did not affect the hydration, setting and curing of the cement matrix. In addition , the obtained cemented waste form exhibits acceptable constitutions that comply with the final disposal requirements.

  15. Radioactive wastes dispersed in stabilized ash cements

    Energy Technology Data Exchange (ETDEWEB)

    Rubin, J.B.; Taylor, C.M.V.; Sivils, L.D.; Carey, J.W.

    1997-12-31

    One of the most widely-used methods for the solidification/stabilization of low-level radwaste is by incorporation into Type-I/II ordinary portland cement (OPC). Treating of OPC with supercritical fluid carbon dioxide (SCCO{sub 2}) has been shown to significantly increase the density, while simultaneously decreasing porosity. In addition, the process significantly reduces the hydrogenous content, reducing the likelihood of radiolytic decomposition reactions. This, in turn, permits increased actinide loadings with a concomitant reduction in disposable waste volume. In this article, the authors discuss the combined use of fly-ash-modified OPC and its treatment with SCCO{sub 2} to further enhance immobilization properties. They begin with a brief summary of current cement immobilization technology in order to delineate the areas of concern. Next, supercritical fluids are described, as they relate to these areas of concern. In the subsequent section, they present an outline of results on the application of SCCO{sub 2} to OPC, and its effectiveness in addressing these problem areas. Lastly, in the final section, they proffer their thoughts on why they believe, based on the OPC results, that the incorporation of fly ash into OPC, followed by supercritical fluid treatment, can produce highly efficient wasteforms.

  16. Radioactive wastes dispersed in stabilized ash cements

    International Nuclear Information System (INIS)

    One of the most widely-used methods for the solidification/stabilization of low-level radwaste is by incorporation into Type-I/II ordinary portland cement (OPC). Treating of OPC with supercritical fluid carbon dioxide (SCCO2) has been shown to significantly increase the density, while simultaneously decreasing porosity. In addition, the process significantly reduces the hydrogenous content, reducing the likelihood of radiolytic decomposition reactions. This, in turn, permits increased actinide loadings with a concomitant reduction in disposable waste volume. In this article, the authors discuss the combined use of fly-ash-modified OPC and its treatment with SCCO2 to further enhance immobilization properties. They begin with a brief summary of current cement immobilization technology in order to delineate the areas of concern. Next, supercritical fluids are described, as they relate to these areas of concern. In the subsequent section, they present an outline of results on the application of SCCO2 to OPC, and its effectiveness in addressing these problem areas. Lastly, in the final section, they proffer their thoughts on why they believe, based on the OPC results, that the incorporation of fly ash into OPC, followed by supercritical fluid treatment, can produce highly efficient wasteforms

  17. Leaching from solidified waste forms under saturated and unsaturated conditions

    International Nuclear Information System (INIS)

    The leaching behavior of nitrate ion from a cement based waste form containing low-level radioactive waste was shown to be identical under saturated and unsaturated soil conditions. Only in soils containing less than 2 wt %water did the leach rate decrease. The observation of identical leach rates under saturated and unsaturated conditions is explained by diffusion through the waste form being the limiting step. Diffusion through the soil decreases in very dry soil and the limiting step changes. These laboratory tests were verified by measurements on similar, Portland cement based waste form in a field lysimeter

  18. Waste form development

    International Nuclear Information System (INIS)

    In this program, contemporary solidification agents are being investigated relative to their applications to major fuel cycle and non-fuel cycle low-level waste (LLW) streams. Work is being conducted to determine the range of conditions under which these solidification agents can be applied to specific LLW streams. These studies are directed primarily towards defining operating parameters for both improved solidification of problem wastes and solidification of new LLW streams generated from advanced volume reduction technologies. Work is being conducted to measure relevant waste form properties. These data will be compiled and evaluated to demonstrate compliance with waste form performance and shallow land burial acceptance criteria and transportation requirements (both as they exist and as they are modified with time). 6 tables

  19. Treatment and recycling of asbestos-cement containing waste

    Energy Technology Data Exchange (ETDEWEB)

    Colangelo, F. [Department of Technology, University Parthenope, Naples (Italy); Cioffi, R., E-mail: raffaele.cioffi@uniparthenope.it [Department of Technology, University Parthenope, Naples (Italy); Lavorgna, M.; Verdolotti, L. [Institute for Biomedical and Composite Materials - CNR, Naples (Italy); De Stefano, L. [Institute for Microelectronics and Microsystems - CNR, Naples (Italy)

    2011-11-15

    Highlights: {yields} Asbestos-cement wastes are hazardous. {yields} High energy milling treatment at room temperature allows mineralogical and morphological transformation of asbestos phases. {yields} The obtained milled powders are not-hazardous. {yields} The inert powders can be recycled as pozzolanic materials. {yields} The hydraulic mortars containing the milled inert powders are good building materials. - Abstract: The remediation of industrial buildings covered with asbestos-cement roofs is one of the most important issues in asbestos risk management. The relevant Italian Directives call for the above waste to be treated prior to disposal on landfill. Processes able to eliminate the hazard of these wastes are very attractive because the treated products can be recycled as mineral components in building materials. In this work, asbestos-cement waste is milled by means of a high energy ring mill for up to 4 h. The very fine powders obtained at all milling times are characterized to check the mineralogical and morphological transformation of the asbestos phases. Specifically, after 120 min of milling, the disappearance of the chrysotile OH stretching modes at 3690 cm{sup -1}, of the main crystalline chrysotile peaks and of the fibrous phase are detected by means of infrared spectroscopy and X-ray diffraction and scanning electron microscopy analyses, respectively. The hydraulic behavior of the milled powders in presence of lime is also tested at different times. The results of thermal analyses show that the endothermic effects associated to the neo-formed binding phases significantly increase with curing time. Furthermore, the technological efficacy of the recycling process is evaluated by preparing and testing hydraulic lime and milled powder-based mortars. The complete test set gives good results in terms of the hydration kinetics and mechanical properties of the building materials studied. In fact, values of reacted lime around 40% and values of compressive

  20. Treatment and recycling of asbestos-cement containing waste

    International Nuclear Information System (INIS)

    Highlights: → Asbestos-cement wastes are hazardous. → High energy milling treatment at room temperature allows mineralogical and morphological transformation of asbestos phases. → The obtained milled powders are not-hazardous. → The inert powders can be recycled as pozzolanic materials. → The hydraulic mortars containing the milled inert powders are good building materials. - Abstract: The remediation of industrial buildings covered with asbestos-cement roofs is one of the most important issues in asbestos risk management. The relevant Italian Directives call for the above waste to be treated prior to disposal on landfill. Processes able to eliminate the hazard of these wastes are very attractive because the treated products can be recycled as mineral components in building materials. In this work, asbestos-cement waste is milled by means of a high energy ring mill for up to 4 h. The very fine powders obtained at all milling times are characterized to check the mineralogical and morphological transformation of the asbestos phases. Specifically, after 120 min of milling, the disappearance of the chrysotile OH stretching modes at 3690 cm-1, of the main crystalline chrysotile peaks and of the fibrous phase are detected by means of infrared spectroscopy and X-ray diffraction and scanning electron microscopy analyses, respectively. The hydraulic behavior of the milled powders in presence of lime is also tested at different times. The results of thermal analyses show that the endothermic effects associated to the neo-formed binding phases significantly increase with curing time. Furthermore, the technological efficacy of the recycling process is evaluated by preparing and testing hydraulic lime and milled powder-based mortars. The complete test set gives good results in terms of the hydration kinetics and mechanical properties of the building materials studied. In fact, values of reacted lime around 40% and values of compressive strength in the range of 2

  1. Study of leaching mechanisms of ions incorporated in cement or polymer Task 3 Characterization of radioactive waste forms A series of final reports (1985-89) No. 2

    International Nuclear Information System (INIS)

    The leaching kinetics of inactive Cs from cylindrical cement specimens containing Cs2SO4 was studied at different temperatures and thicknesses. In all cases the √t plots were reasonably linear, at least initially, in accordance with Fick's law, and the diffusion coefficients were estimated. Leaching of specimens containing Sr-90 and NaNO3 was performed under exposure to atmospheric CO2. Low-temperature differential scanning calorimetry measurements of hydrated cement were undertaken to obtain information about the melting behaviour, and hence the state, of water within the cement. Mercury porosimetry was also carried out using representative cement specimens which had been subjected to leaching. The sorption of Cs ion from aqueous solution by cement was studied by equilibrating cement granules with aqueous Cs2SO4 solutions. Cellulose films containing CaSO4 or SrSO4 were leach tested in frequently renewed water at 250C. The elution curves follow a √t law in conformity with the Higuchi equation. Elution tests of NaCl or SrSO4 embedded in epoxy resin were performed. The SrSO4 elution behaviour was generally similar to that exhibited by cellulose. Theoretical work involved the formulation of a new, sophisticated model capable of describing the elution of a soluble salt, with simultaneous imbition of water by the matrix. Computations more specifically representative of the cellulose acetate-NaCl system, showed that the model can interpret at least semiquantitatively the observed elution behaviour

  2. Leaching properties of solidified TRU waste forms

    International Nuclear Information System (INIS)

    Safety analysis of waste forms requires an estimate of the ability of these forms to retain activity in the disposal environment. This program of leaching tests will determine the leaching properties of TRU contaminated incinerator ash waste forms using hydraulic cement, urea--formaldehyde, bitumen, and vinyl ester--styrene as solidification agents. Three types of leaching tests will be conducted, including both static and flow rate. Five generic groundwaters will be used. Equipment and procedures are described. Experiments have been conducted to determine plate out of 239Pu, counter efficiency, and stability of counting samples

  3. Durability of cemented waste in repository and under simulated conditions

    International Nuclear Information System (INIS)

    The research activities performed by Department of Radioactive Waste Management is focused on the products obtained in the LLAW treatment by chemical precipitation and conditioning by cementation. The individual mechanisms participating in the chemical precipitation process are directly dependent on the precipitate properties and structure, which are related with the initial system composition and the precipitation procedure. In the case of conditioning by cementation, the chemical nature and proportion of the sludges or concentrates affect both the hydrolysis of the initial cement components and the reactions of metastable hydration constituents, as well as the mechanical strength and chemical resistance of the hardened cemented matrix. Generally, the study of the precipitation products and their behaviour during cementation or long-term disposal is extremely difficult because of the system complexity (phase composition and structure) and the lack of the non-destructive analytical methods. For a more detailed characterization, Moessbauer Spectroscopy as a complementary analytical method to XRD, was applied for precipitates and cemented matrices. The following systems are considered: iron precipitates obtained during LLAW treatment; iron hydrated oxides structurally modified by the foreign cations; dry and hydrated cement systems. Also considered and discussed were: - influence of precipitation procedure on the decontamination factors; cementation of sludge chemical components; influence of organic complexing agents on the cemented matrix performances and structure; influence of mineral additives on the concrete; durability of cemented waste in repository and under simulated conditions. Moessbauer investigation of iron species formed in precipitation systems simulating LLAW treatment, revealed that the iron compounds obtained by fast neutralization (as in radioactive aqueous waste treatment) have a different structure compared with iron oxides and hydroxides

  4. Municipal solid waste ash as a cement raw material substitute

    OpenAIRE

    Somnuk Tangtermsirikul; Pichaya Rachdawong; Kritsada Sisomphon

    2000-01-01

    An investigation of using municipal solid waste (MSW) ash as a cement raw material substitute was performed to evaluate the potential use of ash in construction. The use of incineratior ash in cement production would not only get rid of the ash, but also alleviate many environmental problems, for example, reducing raw materials required for cement production, reducing CO2 emission into the atmosphere, and reducing landfill space requirement for the residue ash disposal. The metallic oxide con...

  5. Encapsulation of mixed radioactive and hazardous waste contaminated incinerator ash in modified sulfur cement

    International Nuclear Information System (INIS)

    Some of the process waste streams incinerated at various Department of Energy (DOE) facilities contain traces of both low-level radioactive (LLW) and hazardous constituents, thus yielding ash residues that are classified as mixed waste. Work is currently being performed at Brookhaven National Laboratory (BNL) to develop new and innovative materials for encapsulation of DOE mixed wastes including incinerator ash. One such material under investigation is modified sulfur cement, a thermoplastic developed by the US Bureau of Mines. Monolithic waste forms containing as much as 55 wt % incinerator fly ash from Idaho national Engineering Laboratory (INEL) have been formulated with modified sulfur cement, whereas maximum waste loading for this waste in hydraulic cement is 16 wt %. Compressive strength of these waste forms exceeded 27.6 MPa. Wet chemical and solid phase waste characterization analyses performed on this fly ash revealed high concentrations of soluble metal salts including Pb and Cd, identified by the Environmental Protection Agency (EPA) as toxic metals. Leach testing of the ash according to the EPA Toxicity Characteristic Leaching Procedure (TCLP) resulted in concentrations of Pb and Cd above allowable limits. Encapsulation of INEL fly ash in modified sulfur cement with a small quantity of sodium sulfide added to enhance retention of soluble metal salts reduced TCLP leachate concentrations of Pb and Cd well below EPA concentration criteria for delisting as a toxic hazardous waste. 12 refs., 4 figs., 2 tabs

  6. Durability of cemented waste in repository and under simulated conditions

    International Nuclear Information System (INIS)

    The Romanian Radioactive Waste National Repository for low level and intermediate level radioactive waste was built in Baita - Bihor county, in an extinct uranium exploitation. The site is at 840 m above sea level and the host rock is crystalline with a low porosity, a good chemical homogeneity and impermeability, keeping these qualities over a considerable horizontal and vertical spans. To obtain the experimental data necessary for the waste form and package characterization together with the back-filling material behaviour in the repository environment, a medium term research programme (1996 - 2010) was implemented. The purpose of this experimental programme is to obtain a part of the data base necessary for the approach of medium and long term assessment of the safety and performance of Baita - Bihor Repository. The programme will provide: a deeper knowledge of the chemical species and reaction mechanisms, the structure, properties and performances of the final products. For safety reasons the behaviour of waste package, which is a main barrier, must be properly known in terms of long term durability in real repository conditions. Characterization of the behaviour includes many interactions between the waste package itself and the surrounding near field conditions such as mineralogy, hydrogeology and groundwater chemistry. To obtain a more deeper knowledge of the species and physical-chemical reactions participating in the matrix formation, as well as their future behaviour during the disposal period, a thorough XRD study started in 1998. For Romanian Radioactive Waste National Repository (DNDR) Baita - Bihor the following steps are planned for the conditioned waste matrix characterization in simulated and real conditions: - preparation and characterization of normal reference matrices based on different cement formulations; - preparation of reference simulated sludge cemented matrices containing iron hydroxide and iron phosphate; - selection of real and

  7. Cement solidification method for liquid waste generated from primary loop resin elution process of PWR

    International Nuclear Information System (INIS)

    Since primary loop resin waste is eluted by sulfuric acid in The Kansai Electric Power Co., Inc., Mihama, Takahama and Ohi nuclear power station, liquid waste containing large amounts of sodium sulfate (Na2SO4) was stored in these plants. This liquid waste is planned to be solidified with cement, thus, we have carried out the cement solidification tests by use of some cement materials, and discussed a range of chemical composition and crud concentration of waste solution from resin elution process. In cases of using alumina cement material and ordinary portland cement material for solidification, properties of solidification have been examined and leaching tests of solid form for sulfate ion has been carried out. Volume reduction ratio of over 0.5 was achieved for 5 to 25wt% of sulfate ion and <5,000ppm of borate. Lithium ion restrained the solidification delay by borate. Based on this study, we concluded that these cement materials are applicable to all range of composition of waste solution from the resin elution process. (author)

  8. Thermal stability testing of low-level waste forms

    International Nuclear Information System (INIS)

    The NRC Technical Position (TP) on Waste Form specifies that waste forms should be resistant to thermal degradation. The thermal cycle testing procedure outlined in the TP on Waste Form was carried out and is believed adequate for demonstrating the thermal stability of solidified waste forms. The inclusion of control samples and the monitoring of sample temperature are recommended additions to the test. An outline for reporting thermal cycling test results is given. To produce a data base on the applicability of the thermal cycling test, the following simulated laboratory-scale waste forms were prepared and tested: boric acid and sodium sulfate evaporator bottoms, mixed bed bead resins, and powdered resins each solidified in asphalt, cement and vinyl ester-styrene. Thermal cycling does not significantly affect the compressive strength of the solidified wastes, except powdered resins solidified in cement which disintegrated during the test and bead resins in cement which showed a loss of compressive strength. After temperature cycling, cement solidified bead resins showed areas of spalling and solidified sodium sulfate forms had surface deterioration. Asphalt solidified wastes, except powdered resins, deformed by slumping on temperature cycling. Free liquid was released from vinyl esterstyrene solidifed waste forms as a result of thermal cycling. Dewatered bead and powdered resins were also tested and no free liquid was released on temperature cycling. 11 refs., 12 figs., 4 tabs

  9. Waste form product characteristics

    International Nuclear Information System (INIS)

    The Department of Energy has operated nuclear facilities at the Idaho National Engineering Laboratory (INEL) to support national interests for several decades. Since 1953, it has supported the development of technologies for the storage and reprocessing of spent nuclear fuels (SNF) and the resultant wastes. However, the 1992 decision to discontinue reprocessing of SNF has left nearly 768 MT of SNF in storage at the INEL with unspecified plans for future dispositioning. Past reprocessing of these fuels for uranium and other resource recovery has resulted in the production of 3800 M3 calcine and a total inventory of 7600 M3 of radioactive liquids (1900 M3 destined for immediate calcination and the remaining sodium-bearing waste requiring further treatment before calcination). These issues, along with increased environmental compliance within DOE and its contractors, mandate operation of current and future facilities in an environmentally responsible manner. This will require satisfactory resolution of spent fuel and waste disposal issues resulting from the past activities. A national policy which identifies requirements for the disposal of SNF and high level wastes (HLW) has been established by the Nuclear Waste Policy Act (NWPA) Sec.8,(b) para(3)) [1982]. The materials have to be conditioned or treated, then packaged for disposal while meeting US Environmental Protection Agency (EPA) and Nuclear Regulatory Commission (NRC) regulations. The spent fuel and HLW located at the INEL will have to be put into a form and package that meets these regulatory criteria. The emphasis of Idaho Chemical Processing Plant (ICPP) future operations has shifted toward investigating, testing, and selecting technologies to prepare current and future spent fuels and waste for final disposal. This preparation for disposal may include mechanical, physical and/or chemical processes, and may differ for each of the various fuels and wastes

  10. Acoustic monitoring techniques for corrosion degradation in cemented waste canisters

    International Nuclear Information System (INIS)

    This report describes work carried out to investigate acoustic emission as a monitor of corrosion and degradation of wasteforms where the waste is potentially reactive metal. Electronic monitoring equipment has been designed, built and tested to allow long-term monitoring of a number of waste packages simultaneously. Acoustic monitoring experiments were made on a range of 1 litre cemented Magnox and aluminium samples cast into canisters comparing the acoustic events with hydrogen gas evolution rates and electrochemical corrosion rates. The attenuation of the acoustic signals by the cement grout under a range of conditions has been studied to determine the volume of wasteform that can be satisfactorily monitored by one transducer. The final phase of the programme monitored the acoustic events from full size (200 litre) cemented, inactive, simulated aluminium swarf wastepackages prepared at the AEA waste cementation plant at Winfrith. (Author)

  11. Immobilization of liquid organic radioactive waste in porous cement matrix

    International Nuclear Information System (INIS)

    The purpose of this research is to develop a technique for the immobilization of liquid organic radioactive waste through impregnation of porous cement matrices to ensure high degree of end-compound filling with waste. The technique consists in the following. First of all, the prepared porous cement matrix is placed into a primary package. At the hardening age of not less than 28 days, i.e. when main cement hydration processes are over, the waste is pumped into the cement-matrix pore space through the feeding device pre-installed in the container, causing the matrix impregnation with the waste. Cement composition provides obtaining matrices with strength of up to 9 MPa and porosity of up to 70 %. Compressive strength of the matrix after impregnation does not lower. Properties of the final compound meet the regulated requirements. Waste immobilization in the matrix is reliable. The degree of the end-product filling with organic waste ranged 56-64 % by volume or 47-52 % by weight. The developed mathematical model makes it possible to predict optimal impregnation parameters and output of dosing equipment for immobilization of various types of organic waste. (authors)

  12. Recycling of textile wastes in fibre-cement composites

    OpenAIRE

    H. Monteiro; Caldeira, F.; Pinto, J; Varum, H.

    2013-01-01

    Changing wastes into raw materials is one of the most favoured options for waste management, as it diverts wastes from landfill and saves resources. Fibres, either vegetable (cellulosic) or synthetic, may be added to cement pastes in order improve the properties of concrete or mortar by reinforcement. At the same time, if our source of fibres is wastes, then such processes make ways for recycling. In this work we studied the compatibility of residues from the nonwoven textile indu...

  13. Westinghouse Cementation Facility of Solid Waste Treatment System - 13503

    International Nuclear Information System (INIS)

    During NPP operation, several waste streams are generated, caused by different technical and physical processes. Besides others, liquid waste represents one of the major types of waste. Depending on national regulation for storage and disposal of radioactive waste, solidification can be one specific requirement. To accommodate the global request for waste treatment systems Westinghouse developed several specific treatment processes for the different types of waste. In the period of 2006 to 2008 Westinghouse awarded several contracts for the design and delivery of waste treatment systems related to the latest CPR-1000 nuclear power plants. One of these contracts contains the delivery of four Cementation Facilities for waste treatment, s.c. 'Follow on Cementations' dedicated to three locations, HongYanHe, NingDe and YangJiang, of new CPR-1000 nuclear power stations in the People's Republic of China. Previously, Westinghouse delivered a similar cementation facility to the CPR-1000 plant LingAo II, in Daya Bay, PR China. This plant already passed the hot functioning tests successfully in June 2012 and is now ready and released for regular operation. The 'Follow on plants' are designed to package three 'typical' kind of radioactive waste: evaporator concentrates, spent resins and filter cartridges. The purpose of this paper is to provide an overview on the Westinghouse experience to design and execution of cementation facilities. (authors)

  14. Corrosion of metal containers containing cemented radioactive wastes

    International Nuclear Information System (INIS)

    Nuclear activities generate different kinds of radioactive wastes. In the case of Argentina, wastes classified as low and medium level are conditioned in metal drums for final disposal in a repository whose design is based on the use of multiple and independent barriers. Nuclear energy plants generate a large volume of mid-level radioactive wastes, consisting mainly of ion-exchange resins contaminated by fission products. Other contaminated products such as gloves, papers, clothing, rubber and plastic tubing, can be incinerated and the ashes from the combustion also constitute wastes that must be disposed of. These wastes (resins and ashes) must be immobilized in order to avoid the release of radionuclides into the environment. The wastes usually undergo a process of cementing to immobilize them. This work aims to systematically study the process of degradation by corrosion of the steel drums in contact with the cemented resins and with the ashes cemented with the addition of different types and concentrations of aggressive compounds (chloride and sulfate). The specimens are configured so that the parameters of interest for the steel in contact with the cemented materials can be measured. The variables of corrosion potential, electric resistivity of the matrix and polarization resistance (PR) were monitored and show that the presence of chloride increases the susceptibility to corrosion of the drum steel that is in contact with the cement resin matrix

  15. Solidification of radioactive wastes with cement (A literature study)

    International Nuclear Information System (INIS)

    The present study is structured according to thematic point of views. Emphasis was laid on a precise description of the behaviour of cement used for the solidification of radioactive waste. By utilizing different additives, processing as well as properties like leaching mechanical strength etc. can be improved. In the same way these properties can be positively influenced by various solidification processes and hardening conditions. Chapter 4 discusses the most common kinds of wastes (ion exchange resins, concentrates, solutions etc.). Cement additives and mixtures as well as processing have to be adapted according to the kind of waste to be solidified. In the last chapter activities in different countries in the field of cementation of radioactive wastes are reviewed. (author)

  16. Leaching studies of low-level radioactive waste forms

    International Nuclear Information System (INIS)

    A research program has been underway at the Brookhaven National Laboratory to investigate the release of radionuclides from low-level waste forms under laboratory conditions. This paper describes the leaching behavior of Cs-137 from two major low-level waste streams, that is, ion exchange bead resin and boric acid concentrate, solidified in Portland cement. The resultant leach data are employed to evaluate and predict the release behavior of Cs-137 from low-level waste forms under field burial conditions

  17. Leaching studies of low-level radioactive waste forms

    International Nuclear Information System (INIS)

    A research program has been underway at the Brookhaven National Laboratory to investigate the release of radionuclides from low-level waste forms under laboratory conditions. This paper describes the leaching behaviour of Cs-137 from two major low-level waste streams, that is, ion exchange bead resin and boric acid concentrate, solidified in Portland cement. The resultant leach data are employed to evaluate and predict the release behaviour of Cs-137 from low-level waste forms under field burial conditions

  18. Study of commercial chemical additives for cementation of radioactive waste

    International Nuclear Information System (INIS)

    In this research it has been studied the effects of chemical additives (admixtures) in the cementation process of radioactive wastes, which are used to improve the properties of waste cementation process, both of the paste and of the solidified product. However there are a large variety of these materials that are frequently changed or taken out of the market, then it is essential to know the commercially available materials and their effects. The tests were carried out with a solution simulating the evaporator concentrate waste coming from PWR nuclear reactors. It was cemented using two formulations, A and B, incorporating higher or lower amount of waste, respectively. It was added chemical admixtures from two manufacturers (S and H), which were: accelerators, set retarders and superplasticizers. The experiments were organized by a factorial design 23. The measured parameters were the viscosity, the setting time, the paste and product density and the compressive strength. In this study we performed comparative analyzes of the results of compressive strength at age of 28 and 90 days and between the densities of the samples at the same ages. The compressive strength test at age of 28 days is considered a parameter essential issues related to security handling, transport and storage of cemented waste product. The results showed that the addition of accelerators improved the compressive strength of the cemented product, but presented lower values density products. (authors)

  19. Immobilization of radioactive waste in cement-based matrices

    International Nuclear Information System (INIS)

    Model studies of the behaviour of cement systems have been advanced by considering the nature of the phases formed during hydration and deriving pH-composition models for the CaO-SiO2-H2O system. Preliminary results of Esub(h) measurements are also reported. Leach tests on Sr from cements are interpreted in terms of Sr retention mechanisms. Present results indicate that the aluminate phases in OPC contribute to the chemical retentivity. Studies on cement-clinoptilolite reactions, made using coarse grained clinoptilolite are reported: ferrierite also reacts chemically with cement. Two critical surveys are presented, together with new data: one on the potential of blended cements, the other on cement durability in CO2-containing environments. (author)

  20. Treatment and recycling of asbestos-cement containing waste.

    Science.gov (United States)

    Colangelo, F; Cioffi, R; Lavorgna, M; Verdolotti, L; De Stefano, L

    2011-11-15

    The remediation of industrial buildings covered with asbestos-cement roofs is one of the most important issues in asbestos risk management. The relevant Italian Directives call for the above waste to be treated prior to disposal on landfill. Processes able to eliminate the hazard of these wastes are very attractive because the treated products can be recycled as mineral components in building materials. In this work, asbestos-cement waste is milled by means of a high energy ring mill for up to 4h. The very fine powders obtained at all milling times are characterized to check the mineralogical and morphological transformation of the asbestos phases. Specifically, after 120 min of milling, the disappearance of the chrysotile OH stretching modes at 3690 cm(-1), of the main crystalline chrysotile peaks and of the fibrous phase are detected by means of infrared spectroscopy and X-ray diffraction and scanning electron microscopy analyses, respectively. The hydraulic behavior of the milled powders in presence of lime is also tested at different times. The results of thermal analyses show that the endothermic effects associated to the neo-formed binding phases significantly increase with curing time. Furthermore, the technological efficacy of the recycling process is evaluated by preparing and testing hydraulic lime and milled powder-based mortars. The complete test set gives good results in terms of the hydration kinetics and mechanical properties of the building materials studied. In fact, values of reacted lime around 40% and values of compressive strength in the range of 2.17 and 2.29 MPa, are measured. PMID:21924550

  1. Evaluation of national cements for the conditioning of radioactive wastes

    International Nuclear Information System (INIS)

    The preliminary research studies carried out to implement the liquid radioactive waste conditioning by cementation are described. First of all, different kind of commercial cements in Peru are analyzed. In the first step, the analysis were made without using the radioactive material. The analyzed parameters were density, porosity, setting time and mechanical strength of a cement type called 'Atlas'. Samples of two geometries were used. One of them was a cylindrical sample (48mm diameter and 48mm height) and the another one was a prismatic sample (40x40x160mm). The results of the different kind of analysis are presented in this paper. (authors). 2 refs., 3 tabs

  2. Characterization of low and medium level radioactive waste forms

    International Nuclear Information System (INIS)

    The work reported was carried out during the first year of the Commission of the European Community's programme on the characterization of low and medium level waste forms. Ten reference waste forms plus others of special national interest have been identified covering PWR, BWR, GCR and reprocessing wastes. The immobilising media include the three main matrices: cement, polymers and bitumen, and a glass. Characterization is viewed as one input to quality assurance of the waste form and covers: waste-matrix compatibility, radiation effects, leaching, microbiological attack, shrinkage and swelling, ageing processes and thermal effects. The aim is a balanced programme of comparative data, predictive modelling and an undserstanding of basic mechanisms

  3. Evolution of cement based materials in a repository for radioactive waste and their chemical barrier function

    Energy Technology Data Exchange (ETDEWEB)

    Kienzler, Bernhard; Metz, Volker; Schlieker, Martina; Bohnert, Elke [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany). Inst. fuer Nukleare Entsorgung (INE)

    2015-07-01

    The use of cementitious materials in nuclear waste management is quite widespread. It covers the solidification of low/intermediate-level liquid as well as solid wastes (e.g. laboratory wastes) and serves as shielding. For both high-level and intermediate-low level activity repositories, cement/concrete likewise plays an important role. It is used as construction material for underground and surface disposals, but more importantly it serves as barrier or sealing material. For the requirements of waste conditioning, special cement mixtures have been developed. These include special mixtures for the solidification of evaporator concentrates, borate binding additives and for spilling solid wastes. In recent years, low-pH cements were strongly discussed especially for repository applications, e.g. (Celine CAU DIT COUMES 2008; Garcia-Sineriz, et al. 2008). Examples for relevant systems are Calcium Silicate Cements (ordinary Portland cement (OPC) based) or Calcium Aluminates Cements (CAC). Low-pH pore solutions are achieved by reduction of the portlandite content by partial substitution of OPC by mineral admixtures with high silica content. The blends follow the pozzolanic reaction consuming Ca(OH){sub 2}. Potential admixtures are silica fume (SF) and fly ashes (FA). In these mixtures, super plasticizers are required, consisting of polycarboxilate or naphthalene formaldehyde as well as various accelerating admixtures (Garcia-Sineriz, et al. 2008). The pH regime of concrete/cement materials may stabilize radionuclides in solution. Newly formed alteration products retain or release radionuclides. An important degradation product of celluloses in cement is iso-saccharin acid. According to Glaus 2004 (Glaus and van Loon 2004), it reacts with radionuclides forming dissolved complexes. Apart from potentially impacting radionuclide solubility limitations, concrete additives, radionuclides or other strong complexants compete for surface sites for sorbing onto cement phases. In

  4. Evolution of cement based materials in a repository for radioactive waste and their chemical barrier function

    International Nuclear Information System (INIS)

    The use of cementitious materials in nuclear waste management is quite widespread. It covers the solidification of low/intermediate-level liquid as well as solid wastes (e.g. laboratory wastes) and serves as shielding. For both high-level and intermediate-low level activity repositories, cement/concrete likewise plays an important role. It is used as construction material for underground and surface disposals, but more importantly it serves as barrier or sealing material. For the requirements of waste conditioning, special cement mixtures have been developed. These include special mixtures for the solidification of evaporator concentrates, borate binding additives and for spilling solid wastes. In recent years, low-pH cements were strongly discussed especially for repository applications, e.g. (Celine CAU DIT COUMES 2008; Garcia-Sineriz, et al. 2008). Examples for relevant systems are Calcium Silicate Cements (ordinary Portland cement (OPC) based) or Calcium Aluminates Cements (CAC). Low-pH pore solutions are achieved by reduction of the portlandite content by partial substitution of OPC by mineral admixtures with high silica content. The blends follow the pozzolanic reaction consuming Ca(OH)2. Potential admixtures are silica fume (SF) and fly ashes (FA). In these mixtures, super plasticizers are required, consisting of polycarboxilate or naphthalene formaldehyde as well as various accelerating admixtures (Garcia-Sineriz, et al. 2008). The pH regime of concrete/cement materials may stabilize radionuclides in solution. Newly formed alteration products retain or release radionuclides. An important degradation product of celluloses in cement is iso-saccharin acid. According to Glaus 2004 (Glaus and van Loon 2004), it reacts with radionuclides forming dissolved complexes. Apart from potentially impacting radionuclide solubility limitations, concrete additives, radionuclides or other strong complexants compete for surface sites for sorbing onto cement phases. In Germany

  5. Biodegradation testing of TMI-2 EPICOR-II waste forms

    International Nuclear Information System (INIS)

    ASTM biodegradation tests were conducted on waste forms containing high specific activity ion exchange resins from EPICOR-II prefilters. Those tests were part of a program to test waste forms in accordance with the NRC Branch Technical Position on Waste Form. Small waste forms were manufactured using two different solidification agents, Portland Type I-II cement and vinyl ester-styrene (VES). Ion exchange material was taken from two EPICOR-II prefilters; PF-7, which contained all organic material, and PF-20, which contained organic resins and a layer of inorganic zeolites. Test results showed that the VES waste forms supported microbial growth, while cement waste forms did not support that growth. Growth was also observed adjacent to some VES waste forms. Radiation levels found in the ion exchange resins used in this study were not found to inhibit microbial growth. The extent of degradation of the waste forms could not be determined using the ASTM tests specified by the NRC Branch Technical Position on Waste Form. As a result of this work, a different testing methodology is recommended, which would provide direct verification of waste form capabilities. That methodology would evaluate solidification materials without using the ASTM procedures or subsequent compression testing. The proposed tests would provide exposure to a wide range of microbial species, use appropriately sized specimens, provide for possible use of alternate carbon sources, and extend the test length. Degradation would be determined directly by measuring metabolic activity or specimen weight loss. 16 refs., 15 figs., 3 tabs

  6. Cementation of biodegraded radioactive oils and organic waste

    International Nuclear Information System (INIS)

    The possibility of the microbiological pre-treatment of the oil-containing organic liquid radioactive waste (LRW) before solidification in the cement matrix has been studied. It is experimentally proved that the oil containing cement compounds during long-term storage are subject to microbiological degradation due to the reaction of biogenic organic acids with the minerals of the cement matrix. We recommend to biodegrade the LRW components before their solidification, which reduces the volume of LRW and prevent the destruction of the inorganic cement matrix during the long term storage. The biodegradation of the oil containing LRW is possible by using the radioresistant microflora which oxidize the organic components of the oil to carbon dioxide and water. Simultaneously there is the bio-sorption of the radionuclides by bacteria and emulsification of oil in cement slurry due to biogenic surface-active substances of glycolipid nature. It was experimentally established that after 7 days of biodegradation of oil-containing liquid radioactive waste the volume of LRW is reduced by the factor from 2 to 10 due to the biodegradation of the organic phase to the non-radioactive gases (CH4, H2O, CO2, N2), which are excluded from the volume of the liquid radioactive waste. At the same time, the microorganisms are able to extract from the LRW up to 80-90% of alpha-radionuclides, up to 50% of 90Sr, up to 20% of 137Cs due to sorption processes at the cellular structures. The radioactive biomass is subject to dehydration and solidification in the matrix. The report presents the following experimental data: type of bacterial flora, the parameters of biodegradation, the cementing parameters, the properties of the final cement compound with oil-containing liquid radioactive waste

  7. Leachability of bentonite/cement for medium-level waste immobilisation

    Energy Technology Data Exchange (ETDEWEB)

    Hamlat, M.S.; Rabia, N. [Centre de Radioprotection et de Surete, Alger-Gare (Algeria)

    1998-12-31

    The release of radionuclides from Algerian bentonite/cement matrix has been measured experimentally using static and dynamic testing procedures. The waste forms were cement/sand and bentonite/cement matrices contaminated with Cs-137. To characterise radionuclide/waste form combination, two parameters, diffusion (D) and distribution coefficients ({alpha}) were used. (D) is an effective diffusion coefficient that describes the kinetic behaviour and is most easily determined using Soxhlet test, whereas, ({alpha}) describes the distribution of radionuclide between aqueous and solid phases at equilibrium and is best measured in static test. Leach rates obtained being very low. Distribution coefficient values have showed that the bentonite has relatively a high degree of fixation. It was concluded that the matrix under study seems play a role for the immobilisation. (orig.)

  8. The Cement Solidification of Municipal Solid Waste Incineration Fly Ash

    Institute of Scientific and Technical Information of China (English)

    HOU Haobo; HE Xinghua; ZHU Shujing; ZHANG Dajie

    2006-01-01

    The chemical composition, the content and the leachability of heavy metals in municipal solid waste incineration ( MSWI) fly ash were tested and analyzed. It is shown that the leachability of Pb and Cr exceeds the leaching toxicity standard, and so the MSWI fly ash is considered as hazardous waste and must be solidifled. The effect of solidifying the MSWI fly ash by cement was studied, and it is indicated that the heavy metals can be well immobilized if the mass fraction of the fly ash is appropriate. The heavy metals were immobilized within cement hydration products through either physical fixation, substitution, deposition or adsorption mechanisms.

  9. Seawater leachability of cement solidified heavy metal wastes

    International Nuclear Information System (INIS)

    This paper investigates the seawater leachability of portland cement, solidified cadmium, and lead wastes. The synthetic seawater leachates were analyzed for metals content using atomic absorption spectrophotometry. The pH and alkalinity of the leachate was also measured. The cumulative cadmium release after 46 days of leaching was approximately 1.0 percent of the initial total amount added to the portland cement mixture. The microstructure of the solidified waste was investigated using the SEM, XRD, MIP and helium pycnometry. Cadmium was detected as cadmium hydroxide. During the leaching process the surficial microstructure of the solidified waste exhibited a dynamic layer of calcite, paragonite and brucite while the internal structure showed large amounts of ettringite crystals in the cadmium waste only which caused excessive expansion and cracking. A proposed leaching mechanism experienced by the solidified waste is related to the microstructural characteristics of the matrix

  10. The incorporation of low and medium level radioactive wastes (solids and liquids) in cement

    International Nuclear Information System (INIS)

    Progress is reported on laboratory studies on the stability of cemented wastes samples. The rate of strength development and the temperatures measured during the setting of the cement in standard waste containers is reported. Preliminary information is given on a technique for the non-destructive testing of cement/waste specimens. (U.K.)

  11. Sulfur polymer cement stabilization of elemental mercury mixed waste

    International Nuclear Information System (INIS)

    Elemental mercury, contaminated with radionuclides, is a problem throughout the Department of Energy (DOE) complex. This report describes the development and testing of a process to immobilize elemental mercury, contaminated with radionuclides, in a form that is non-dispersible, will meet EPA leaching criteria, and has low mercury vapor pressure. In this stabilization and solidification process (patent pending) elemental mercury is mixed with an excess of powdered sulfur polymer cement (SPC) and additives in a vessel and heated to ∼35 C, for several hours, until all of the mercury is converted into mercuric sulfide (HgS). Additional SPC is then added and the mixture raised to 135 C, resulting in a homogeneous molten liquid which is poured into a suitable mold where is cools and solidifies. The final stabilized and solidified waste forms were characterized by powder X-ray diffraction, as well as tested for leaching behavior and mercury vapor pressure. During this study the authors have processed the entire inventory of mixed mercury waste stored at Brookhaven National Laboratory (BNL)

  12. Processes and Equipment for the Cementation of Radioactive Waste

    International Nuclear Information System (INIS)

    In this article a short selection of different cement mixer types provided by NUKEM Technologies is given. The variety stems on one hand from historical development, but more especially from specific customer demands to meet their local and technical requirements. The Slant Batch Mixer is successfully installed in several Waste Treatment Centers (WTC). NUKEM Technologies set up these mixers with necessary auxiliary systems to facilitate all the cementation tasks of a WTC. By the slant design of the mixer a homogeneous intermixing and a rapid and comprehensive emptying is achieved. The High Shear Mixer is a batch mixer producing a thixotropic, fast flowing colloidal cement slurry. NUKEM Technologies uses this cement slurry to bubble-free/ empty space-free grouting of pre-packed solid waste items in container. The High Throughput Continuous Mixer is a continuously operating screw mixer that provides a high throughput. One or more dry components are continuously fed to the mixer where liquid waste or water is added. The High Performance In-Drum Mixer is a combination of planetary mixer with double helical mixer. NUKEM Technologies recently has developed a new High Capacity Mixer (HCM) based on a well proven conventional concrete mixer. The HCM is the successor of the slant mixer and will expend NUKEM Technologies' portfolio of cementation units. (A.C.)

  13. The effects of radiation on intermediate-level waste forms. Task 3 characterization of radioactive waste forms a series of final reports (1985-89) no. 10

    International Nuclear Information System (INIS)

    The purpose of this programme was to determine the effects of radiation on the properties of intermediate-level waste forms relevant to their storage and disposal. It had two overall aims: to provide immediate data on the effect of radiation on important European ILW waste forms through accelerated laboratory tests; and to develop an understanding of the degradation processes so that long-term, low dose rate effects can be predicted with confidence from short-term, high dose rate experiments. The programme included cement waste forms containing inorganic wastes, organic matrix waste forms, and cement waste forms containing a substantial component of organic waste. Irradiations were carried out by external gamma sources and by the incorporation of alpha emitters, such as 238Pu. Irradiated materials included matrix materials, simulated waste forms and real waste forms. 2 figs.; 3 tabs.; 8 refs

  14. The role of cement to be expected in radioactive waste disposal system. 2. From the standpoint of materials design

    International Nuclear Information System (INIS)

    Cement materials are used at various fields because of their mechanical properties, and then a large construction without using the cement materials is impossible to suppose. For disposal of radioactive wastes, it is expected to use the cement materials for a main constitution material of artificial barrier materials such as construction materials for a disposal facility, wastes container, solidification materials for wastes, and so forth, and in fact, they are used for cement solidified matters, concrete pit as a landfill apparatus, and so forth at the Low Level Radioactive Wastes Storage Center situated in Rokkasho-mura, Aomori prefecture. For their disposal, as cement materials are expected for their property on transfer control of radioactive nuclides such as water stoppage, pH buffering of circumferential groundwater, and transfer retarding, except their mechanical properties, it must be quantitatively investigated how they change with time and if their change forms any problem on safety, because a time to consider their soundness on mechanics or nuclide conservation becomes long term such as for more than hundreds years. Under consideration on disposal and technical trends of radioactive wastes in- and out of-Japan described in previous report, after showing on direction of investigation required to make the cement materials function as an artificial material in disposal of radioactive wastes and on technical trends to it, here was summarized on positioning of studies on cement in the disposal business. (G.K.)

  15. Performance of monolithic concrete waste forms

    International Nuclear Information System (INIS)

    Liquid wastes can be made into concrete or cement waste form that can be poured into a concrete vault forming a monolith. The waste isolation performance of monolithic concrete waste forms or vaults is generally dominated by the influence of cracks through the structure. In relation to water flow rate and crack spacing, monolithic concrete vaults have three general regions of performance. At extremely low flow rates, release is strictly diffusionally limited. In most situations, flow rates will not be low enough to ensure diffusional release. At slightly greater flow rates (the magnitude of which is dependent upon the diffusion coefficients and crack spacing), release is controlled by the flow rate of water through cracks in the structure with release rate approximately proportional to Darcy flow. In this region, release is not sensitive to block size and the vault behaves as an equivalent porous medium from a mass transport perspective. At higher flow rates, release rate is controlled by diffusion out of intact blocks of waste form. In this situation the release rate is very sensitive to block size (crack spacing) but independent of flow. (author)

  16. Exergy analyses in cement production applying waste fuel and mineralizer

    International Nuclear Information System (INIS)

    Highlights: • The exergy destroyed from calciner and rotary kiln correspond 47% and 30%. • Mineralizer saves energy. • Waste fuel provides 12% of energy demand. - Abstract: The cement industry is an energy-intensive industry and emits large quantities of carbon dioxide, so waste fuels could usefully substitute part of the fossil fuels. They can also help resolve air pollution problems associated with the use of fossil fuels. Other wastes have properties of reducing the thermal energy consumption of clinker production. They are named mineralizers. Then the application of both in the cement industry contributes to the reduction of environmental liabilities and provides lower cost of acquisition of fossil fuels. The aim of the present study is confirm the advantages of the application of waste SPL (spent pot lining) as a mineralizer in clinker production from an exergetic viewpoint

  17. Adaptation of magnesian cements to underground storage of nuclear wastes

    International Nuclear Information System (INIS)

    The aim of this thesis is the experimental study of magnesium oxychloride cements as filling materials for underground granitic cavities containing high level radioactive wastes. After a bibliographic study, mechanical properties are examined before and after setting, in function of the ratio MgO/MgCl2. Then behavior with water is investigated: swelling, cracking and leaching

  18. Calcium Sulfoaluminate Eco-Cement from Industrial Waste

    OpenAIRE

    Ukrainczyk, N.; Frankoviæ Mihelj, N.; Šipušić, J.

    2013-01-01

    In this paper, the potential benefits offered by calcium sulfoaluminate cement (CSA) production from industrial wastes or by-products already present in Republic of Croatia have been addressed. A variety of industrial wastes, namely phosphogypsum (PG), coal bottom ash (BA) and electric arc furnace slag (EAFS) were used as raw materials to provide additional environmental advantages in production of CSA. Mass fraction of Ye’elimite, the principal hydraulic mineral in the prepared CSA was de...

  19. The incorporation of low and medium level radioactive wastes (solids and liquids) in cement

    International Nuclear Information System (INIS)

    The use of cement has been investigated for the immobilization of liquid and solid low and medium level radioactive waste. 220 litre mixing trials have demonstrated that the high temperatures generated during the setting of ordinary Portland cement/simulant waste mixes can be significantly reduced by the use of a blend of ground granulated blast furnace slag and ordinary Portland cement. Laboratory and 220 litre trials using simulant wastes showed that the blended cement gave an improvement in properties of the cemented waste product, e.g. stability and reduction in leach rates compared with ordinary Portland cement formulations. A range of 220 litre scale mixing systems for the incorporation of liquid and solid wastes in cement was investigated. The work has confirmed that cement-based processes can be used for the immobilization of most types of low and medium level waste

  20. Factors affecting the leachability of caesium and strontium from cemented simulant evaporator wastes

    International Nuclear Information System (INIS)

    Leach rates of stable cesium and strontium from a range of simulated evaporator waste/cement formulations have been determined. Important factors in plant operation are assessed for their effect on leach rates. Increasing the curing time and lowering the water/cement ratio has been shown to reduce leach rates by up to a factor of four. Incorporation of additives such as clays and supplementary cementatious materials can reduce leach rates by up to three orders magnitude, and coating the surface of the waste form with a neat cement grout can reduce the cesium leach rate by up to four orders of magnitude. The effects of permeability of the matrix and its cesium absorption capacity on the leach rates have been analysed qualitatively. (U.K.)

  1. The incorporation of low and medium level radioactive wastes (solids and liquids) in cement

    International Nuclear Information System (INIS)

    Experimentation has shown that high temperatures generated during the setting of ordinary Portland cement/simulant waste mixes can be significantly reduced by the use of a blend of ground granulated blast furnace slag and ordinary Portland cement. Trials on simulated waste showed that blended cement gave improved stability and a reduction in leach rates, and confirmed that the cement-based process can be used for the immobilisation of most types of low and medium level waste. (U.K.)

  2. Utilizing wood wastes as reinforcement in wood cement composite bricks

    Directory of Open Access Journals (Sweden)

    Nusirat Aderinsola Sadiku

    2015-07-01

    Full Text Available This paper presents the research work undertaken to study the properties of Wood Cement Composite Bricks (WCCB from different wood wastes and cement / wood content. The WCBBs with nominal density of 1200 kg m-3 were produced from three tropical wood species and at varying cement and wood content of 2:1, 2.5:1 and 3:1 on a weight to weight basis. The properties evaluated were compressive strength, Ultra Pulse Velocity (UPV, water absorption (WA and thickness swelling (TS. The Compressive strength values ranged from 0.25 to 1.13 N mm-2 and UPV values ranged from 18753 to 49992 m s-1. The mean values of WA after 672 hours (28 days of water soaking of the WCCBs ranged from 9.50% to 47.13% where there were no noticeable change in the TS of the bricks. The observed density (OD ranged from 627 to 1159 kg m-3. A. zygia from the three wood/cement content were more dimensionally stable and better in compressive strength than the other two species where T. scleroxylon had the best performance in terms of UPV. All the properties improved with increasing cement content. WCCBs at 3.0:1 cement/wood content are suitable for structural application such as panelling, ceiling and partitioning

  3. Cement Solidification Method For Intermediate-Level Liquid Waste Containing Sodium Sulphate (Na2SO4)

    International Nuclear Information System (INIS)

    A new cement solidification method for intermediate-level liquid waste containing large amounts of sodium sulphate (Na2SO4) has been developed. This method involves two safety concepts for disposal sites: reduction in the amount of sulphate ion (SO42-) released from solidified wastes and reduction in the amount of hydrogen gas generated due to radiolysis of the water present in the solidified waste. In order to eliminate SO42- release from solidified wastes, two chemical reactions were important in our solidification method: (1) Barium-compounds (Ba(OH)2.8H2O, etc) were reacted with SO42- to form BaSO4, and (2) using alumina cement material, SO42- was mineralized as ettringite, 3CaO.Al2O3.3CaSO3.2H2O. Based on leaching tests, the amount of SO42- released from the solidified forms into ion exchange water under anaerobic conditions was less than 1 x 10-3 mol/L. Thus, this method should be effective in preventing engineered concrete barrier layers from cracking. In order to evaluate the amount of hydrogen gas generated from cement solids due to radiolysis of hydrated and non-hydrated water in the solid, gamma-ray irradiation experiments on solidified alumina cement (ALC), solidified ordinary portland cement (OPC), solidified ordinary portland cement blended with blast-furnace slag (OPC-BFS), and synthetic ettringite were performed. As a result, the generation rate of hydrogen gas from ALC was less than those from OPC and OPC-BFS and approximately equal to that from ettringite. (authors)

  4. UTILIZATION OF AGARWOOD DISTILLATION WASTE IN OILWELL CEMENT AND ITS EFFECT ON FREE WATER AND POROSITY

    OpenAIRE

    Arina Sauki; Muhammad Hazman Md. Shahid; Ku Halim Ku Hamid; Azlinda Azizi; Siti Khatijah Jamaludin; Tengku Amran Tengku Mohd; Nur Hashimah Alias

    2013-01-01

    The intent of this research is to utilize the waste produced by distillation process of Agarwood oil and convert it into a profitable oilwell cement additive. Common problem during oilwell cementing is free wáter separation. This problem could weaken cement at the top, gas migration problem and non uniform density of cement slurry that are even worst in cementing deviated well. Another concern on cementing design is the porosity of the hardened cement. If the cement is too porous, it can lead...

  5. Mixed Waste Focus Area - Waste form initiative

    International Nuclear Information System (INIS)

    The mission of the US Department of Energy's (DOE) Mixed Waste Focus Area (MWFA) is to provide acceptable technologies that enable implementation of mixed waste treatment systems which are developed in partnership with end-users, stakeholders, tribal governments, and regulators. To accomplish this mission, a technical baseline was established in 1996 and revised in 1997. The technical baseline forms the basis for determining which technology development activities will be supported by the MWFA. The primary attribute of the technical baseline is a set of prioritized technical deficiencies or roadblocks related to implementation of mixed waste treatment systems. The Waste Form Initiative (WFI) was established to address an identified technical deficiency related to waste form performance. The primary goal of the WFI was to ensure that the mixed low-level waste (MLLW) treatment technologies being developed, currently used, or planned for use by DOE would produce final waste forms that meet the waste acceptance criteria (WAC) of the existing and/or planned MLLW disposal facilities. The WFI was limited to an evaluation of the disposal requirements for the radioactive component of MLLW. Disposal requirements for the hazardous component are dictated by the Resource Conservation and Recovery Act (RCRA), and were not addressed. This paper summarizes the technical basis, strategy, and results of the activities performed as part of the WFI

  6. Incorporation of the Spinning Wastes in Cement and Mortars

    International Nuclear Information System (INIS)

    The study falls within the scope of the general problem of management of accumulated solid cellulose based wastes originating from textile industry especially those coming from the spinning process. The present investigation aims at studying the wet oxidative degradation technique as a method for treating cellulose-based wastes generated by the textile spinning industries. The resulting treated wastes were incorporated into cement or mortar. The mechanical integrity and the weight change of the final products were evaluated at the end of setting and hardening

  7. Leaching due to hygroscopic water uptake in cemented waste containing soluble salts

    DEFF Research Database (Denmark)

    Brodersen, K.

    1992-01-01

    Considerable amounts of easily soluble salts such as sodium nitrate, sulphate, or carbonate are introduced into certain types of cemented waste. When such materials are stored in atmospheres with high relative humidity or disposed or by shallow land burial under unsaturated, but still humid...... conditions, condensation of water vapour will result in generation of a certain amount of liquid in the form of a strong salt solution. The volume of liquid may well exceed the storage capacity of the pore system in the cemented material and in the release of a limited amount of free contaminated solution. A...

  8. Characterization of waste products prepared from radioactive contaminated clayey soil cemented according to the GEODUR process

    International Nuclear Information System (INIS)

    Radioactive contaminated soil may arise due to accidents of various types or may be detected during decommisioning of nuclear installations. Ordinary surface soil cannot normally be conditioned using conventional cementation processes since the content of humic materials retards or prevents the solidification. An additive available from the Danish firm Geodur A/S makes it possible to circumvent this difficulty and to produce a monolithic, nondusting waste type using rather small amounts of cement. The report describes work on characterization of such a cemented waste product prepared on basis of clayey top soil from the Risoe area. The claimed advantages of the process was verified, and data for the compression strength (low), hydraulic conductivity (satisfactory) and other pore structure-related properties are given for the obtained products. Unfortunately the behaviour of cesium and strontium, representing two of the most relevant radionuclides, was not too promising. The retention of cesium is satisfactory, but less good than for the untreated soil. Greatly improved cesium retention after drying of the materials was noticed. Good retention of strontium is only obtained after reaction of the material with carbon dioxide from the atmosphere. The behaviour of the two isotopes in other types of cemented waste is somewhat similar, but the decrease in retention compared with untreated soil makes the process less interesting as a possibility for remedial actions after accidents, etc. Some further studies of the cemented soil waste are beeing made within the frame of the Nordic Nuclear Safety Studies. Elements forming low solublity components in the high pH environment in the cemented soil will probably be retained quite efficiently. This was demonstrated in case of Zn. (author) 11 tabs., 22 ills., 8 refs

  9. Vitreous ceramic waste form for waste immobilization

    International Nuclear Information System (INIS)

    Vitreous ceramic waste forms are being developed to complement glass waste forms in supporting DOE's environmental restoration efforts. The vitreous ceramics are composed of various metal oxide crystalline phases embedded in a silicate glass matrix. The vitreous ceramics are appropriate final waste forms for waste streams that contain large amounts of scrap metals and elements with low solubilities in glass, and have low-flux contents. Homogeneous glass waste forms are appropriate for wastes with sufficient fluxes and low metal contents. Therefore, utilization of both glass and vitreous ceramics waste forms will make vitrification technology applicable to the treatment of a much larger range of radioactive and mixed wastes. The controlled crystallization in vitreous ceramics resulted in formation of durable crystalline phases and durable residual glass matrix. The durable crystalline phases in vitreous ceramics included Ca3(PO4)2, magnetite (Fe2+Ni,Mn)Fe3+2O4, hibonite Ca(Al,Fe,Zr,Cr)12O19, baddeyelite ZrO2, zirconolite CaZrTi,O, and corundum Al2O3, which are thermodynamically more stable than normal glasses and are also less soluble in water than glasses. The durable glassy matrix in vitreous ceramics is due to the enrichment of silica and alumina during the crystallization process of vitreous ceramic formation. The vitreous ceramics showed exceptional long-term chemical durability and the processability of vitreous ceramics were also demonstrated at both bench- and pilot-scale. This paper briefly describes the use of vitreous ceramics for treating sample mixed wastes with high contents of either Cr, Fe, Zr, and Al, or alkalis

  10. Accelerated ageing of blended cements for use in radioactive waste disposal

    International Nuclear Information System (INIS)

    An accelerated experimental technique has been developed to study the long term hydration of blended cements that may be used in radioactive waste disposal. This technique has been used to investigate the hydration reactions of Ordinary Portland Cement (OPC) blended with blast furnace 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 was also investigated. Solid/solution compositional data has been collected during the course of the hydration process and this can be used in the validation of models for the properties of cements. Thomsonite, afwillite, a tobermorite-like phase and thaumasite have been found in addition to the expected cement hydration products and need to be considered in modelling studies of cement hydration. The pH of ground waters reacted with OPC/pfa blends on hydration at 90oC fell below 8. This is lower than the value required to inhibit the corrosion of steel canisters in a repository. The pH in ground waters reacted with OPC and OPC/ggbs mixes remained above 11, although if the ground waters reacted with OPC/ggbs blends were periodically replaced the pH eventually fell below 10. The experimental procedure could be adapted to test the specific cement and ground water compositions relevant to the design of an underground repository over a range of experimental conditions. (author)

  11. Using of borosilicate glass waste as a cement additive

    Science.gov (United States)

    Han, Weiwei; Sun, Tao; Li, Xinping; Sun, Mian; Lu, Yani

    2016-08-01

    Borosilicate glass waste is investigated as a cement additive in this paper to improve the properties of cement and concrete, such as setting time, compressive strength and radiation shielding. The results demonstrate that borosilicate glass is an effective additive, which not only improves the radiation shielding properties of cement paste, but also shows the irradiation effect on the mechanical and optical properties: borosilicate glass can increase the compressive strength and at the same time it makes a minor impact on the setting time and main mineralogical compositions of hydrated cement mixtures; and when the natural river sand in the mortar is replaced by borosilicate glass sand (in amounts from 0% to 22.2%), the compressive strength and the linear attenuation coefficient firstly increase and then decrease. When the glass waste content is 14.8%, the compressive strength is 43.2 MPa after 28 d and the linear attenuation coefficient is 0.2457 cm-1 after 28 d, which is beneficial for the preparation of radiation shielding concrete with high performances.

  12. Production and characteristics for concrete waste forms to stabilize concrete waste produced during decommissioning procedure

    International Nuclear Information System (INIS)

    Since the decommissioning of nuclear plants and facilities, large quantities of slightly contaminated concrete waste have been generated. In Korea, the decontamination and decommissioning of the KRR-1, 2 at the KAERI have been under way. Concrete waste was generated 83 drums of 200L and 41 containers of 4 M3. These concrete wastes consist of rubble, coarse, and fine aggregates. And also, 24 drums of concrete sludge were generated from the saw cutting of radioactive concrete. The conditioning of concrete waste is needed for final disposal. The concrete waste is conditioned as follows: mortar using coarse and fine aggregates is filled void space after concrete rubble pre-placement into 200 L drum. The mortar needs to be solidified using cement or other materials to protect from sufficient strength and harmful opening. Especially, cement was frequently used solidification/stabilization above all the other ones because of competitive prices, convenient method and excellent quality. Thus, this paper has developed an optimizing mixing ratio of concrete waste, water, and cement and has evaluated characteristics of a cement waste form containing radioactive concrete to meet the requirements specified in disposal site specific waste acceptance criteria

  13. Polyethylene encapsulatin of nitrate salt wastes: Waste form stability, process scale-up, and economics

    International Nuclear Information System (INIS)

    A polyethylene encapsulation system for treatment of low-level radioactive, hazardous, and mixed wastes has been developed at Brookhaven National Laboratory. Polyethylene has several advantages compared with conventional solidification/stabilization materials such as hydraulic cements. Waste can be encapsulated with greater efficiency and with better waste form performance than is possible with hydraulic cement. The properties of polyethylene relevant to its long-term durability in storage and disposal environments are reviewed. Response to specific potential failure mechanisms including biodegradation, radiation, chemical attack, flammability, environmental stress cracking, and photodegradation are examined. These data are supported by results from extensive waste form performance testing including compressive yield strength, water immersion, thermal cycling, leachability of radioactive and hazardous species, irradiation, biodegradation, and flammability. The bench-scale process has been successfully tested for application with a number of specific ''problem'' waste streams. Quality assurance and performance testing of the resulting waste form confirmed scale-up feasibility. Use of this system at Rocky Flats Plant can result in over 70% fewer drums processed and shipped for disposal, compared with optimal cement formulations. Based on the current Rocky Flats production of nitrate salt per year, polyethylene encapsulation can yield an estimated annual savings between $1.5 million and $2.7 million, compared with conventional hydraulic cement systems. 72 refs., 23 figs., 16 tabs

  14. Waste forms for plutonium disposition

    International Nuclear Information System (INIS)

    The field of plutonium disposition is varied and of much importance, since the Department of Energy has decided on the hybrid option for disposing of the weapons materials. This consists of either placing the Pu into mixed oxide fuel for reactors or placing the material into a stable waste form such as glass. The waste form used for Pu disposition should exhibit certain qualities: (1) provide for a suitable deterrent to guard against proliferation; (2) be of minimal volume, i.e., maximize the loading; and (3) be reasonably durable under repository-like conditions. This paper will discuss several Pu waste forms that display promising characteristics

  15. Laboratory procedures for waste form testing

    Energy Technology Data Exchange (ETDEWEB)

    Mast, E.S.

    1994-09-19

    The 100 and 300 areas of the Hanford Site are included on the US Environmental Protection Agencies (EPA) National Priorities List under the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA). Soil washing is a treatment process that is being considered for the remediation of the soil in these areas. Contaminated soil washing fines can be mixed or blended with cementations materials to produce stable waste forms that can be used for beneficial purposes in mixed or low-level waste landfills, burial trenches, environmental restoration sites, and other applications. This process has been termed co-disposal. The Co-Disposal Treatability Study Test Plan is designed to identify a range of cement-based formulations that could be used in disposal efforts in Hanford in co-disposal applications. The purpose of this document is to provide explicit procedural information for the testing of co-disposal formulations. This plan also provides a discussion of laboratory safety and quality assurance necessary to ensure safe, reproducible testing in the laboratory.

  16. Laboratory procedures for waste form testing

    International Nuclear Information System (INIS)

    The 100 and 300 areas of the Hanford Site are included on the US Environmental Protection Agencies (EPA) National Priorities List under the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA). Soil washing is a treatment process that is being considered for the remediation of the soil in these areas. Contaminated soil washing fines can be mixed or blended with cementations materials to produce stable waste forms that can be used for beneficial purposes in mixed or low-level waste landfills, burial trenches, environmental restoration sites, and other applications. This process has been termed co-disposal. The Co-Disposal Treatability Study Test Plan is designed to identify a range of cement-based formulations that could be used in disposal efforts in Hanford in co-disposal applications. The purpose of this document is to provide explicit procedural information for the testing of co-disposal formulations. This plan also provides a discussion of laboratory safety and quality assurance necessary to ensure safe, reproducible testing in the laboratory

  17. [Release amount of heavy metals in cement product from co-processing waste in cement kiln].

    Science.gov (United States)

    Yang, Yu-Fei; Huang, Qi-Fei; Zhang, Xia; Yang, Yu; Wang, Qi

    2009-05-15

    Clinker was produced by Simulating cement calcination test, and concrete samples were also prepared according to national standard GB/T 17671-1999. Long-term cumulative release amount of heavy metals in cement product from co-processing waste in cement kiln was researched through leaching test which refers to EA NEN 7371 and EA NEN 7375, and one-dimensional diffusion model which is on the base of Fick diffusion law. The results show that availabilities of heavy metals are lower than the total amounts in concrete. The diffusion coefficients of heavy metals are different (Cr > As > Ni > Cd). During 30 years service, the cumulative release amounts of Cr, As, Ni and Cd are 4.43 mg/kg, 0.46 mg/kg, 1.50 mg/kg and 0.02 mg/kg, respectively, and the ratios of release which is the division of cumulative release amount and availability are 27.0%, 18.0%, 3.0% and 0.2%, respectively. The most important influence factor of cumulative release amount of heavy metal is the diffusion coefficient, and it is correlative to cumulative release amount. The diffusion coefficient of Cr and As should be controlled exactly in the processing of input the cement-kiln. PMID:19558131

  18. Alternative waste forms: a comparative study

    International Nuclear Information System (INIS)

    A characterization study utilizing comparative tests has been conducted to assess product inertness of alternative waste form materials, having evaluated at this point four basic product types: sintered ceramics, glass ceramics, glass and concrete. The seven specific waste form materials studied represent simulated nuclear waste loading of 5% to 100%, processed between room temperature and 12000C and subjected to characterization tests including phase analysis, microstructure, compression testing, volatility and leach testing. Significant conclusions based upon the results obtained to date are: sintered calcine waste form PW-9 does not retain Na, Mo and Cs when leached 900C and, in fact, does not remain a solid; glass and supercalcine are alike under both hydrous and hydrothermal leach conditions with glass exhibiting a greater retention of sodium and molybdenum, supercalcine having a greater retention of cesium, and both forms approximately equal in strontium retention; volatility measurements indicate that an order of magnitude decrease in volatility occurs when a calcine waste form is incorporated in a crystalline or glassy host; glass 76-68 is superior to supercalcine SPC-5B in retention of volatiles below 11000C because of the high release of Na from SPC-5B, however, as the temperature approaches or exceeds the glass melt temperature, volatile losses of the glass equal or exceed that of SPC-5B; glass 76-68 and supercalcine SPC-5B have high compressive strengths when compared to sintered PW-9 and cement products. This is apparently due to a stronger continuum bond resulting from a glassy matrix or crystalline ingrowth over a simple mechanical agglomeration of particles

  19. Performance of cement solidification with barium for high activity liquid waste including sulphate

    International Nuclear Information System (INIS)

    The target liquid waste to be solidified is generated from PWR primary loop spent resin treatment with sulphate acid, so, its main constituent is sodium sulphate and the activity of this liquid is relatively high. Waste form of this liquid waste is considered to be a candidate for the subsurface disposal. The disposed waste including sulphate is anticipated to rise a concentration of sulphate ion in the ground water around the disposal facility and it may cause degradation of materials such as cement and bentonite layer and comprise the disposal facility. There could be two approaches to avoid this problem, the strong design of the disposal facility and the minimization of sulphaste ion migration from the solidified waste. In this study, the latter approach was examined. In order to keep the low concentration of sulphate ion in the ground water, it is effective to make barium sulphate by adding barium compound into the liquid waste in solidification. However, adding equivalent amount of barium compound with sulphate ion causes difficulty of mixing, because production of barium sulphate causes high viscosity. In this study, mixing condition after and before adding cement into the liquid waste was estimated. The mixing condition was set with consideration to keep anion concentration low in the ground water and of mixing easily enough in practical operation. Long term leaching behavior of the simulated solidified waste was also analyzed by PHREEQC. And the concentration of the constitution affected to the disposal facility was estimated be low enough in the ground water. (author)

  20. Mathematical modeling in radionuclide migration in cement-waste composition

    International Nuclear Information System (INIS)

    The intermediate level radioactive wastes produced as a result of the operation of reprocessing plants and nuclear power station are currently being considered for immobilization in a number of possible matrix materials, including cement, bitumen, and plastics. In the storage or disposal of these solidified wastes, some radioactive products may be leached into contacting water and this water might enter into the environment. Therefore, an estimation of the quantities of radionuclides that ca be leached from waste composites is essential in radioactive waste management. At our laboratory, a promising composite for solidification of radioactive wastes has been developed by incorporation of spent ion exchange resin in cement. Leaching of 137 Cs was studied using the method recommended by the IAEA. Transport phenomena involved in the leaching of a radioactive material from a composite matrix into surrounding water were investigated using three methods on theoretical equations. These were: Method I - using a diffusion equation derived for a plane source model, Method II - using a rate equation for diffusion coupled with a first-order reaction, and Method III - in which the leaching data were also analysed by an empirical method employing a polynomial equation. These three methods are compared with respect to their applicability to the 147 Cs leaching data. (authors)

  1. Cement Waste Matrix Evaluation and Modelling of the Long Term Stability of Cementitious Waste Matrices

    International Nuclear Information System (INIS)

    Cement based materials are often used as a solidification matrix for wet radioactive waste from nuclear power plants such as ion exchange resins, sludge and evaporator concentrates. The mechanical and chemical properties of the cement-waste matrix are affected by the type and the concentration of the waste. For this reason the recipe used in the solidification process has to be carefully adjusted to respond to the variations of the waste. At the Ringhals Nuclear Power Plant (RNPP) an evaporator was to be taken into operation during the mid 2005. As a result of this process an evaporator concentrate containing boric acid was expected. The aims of the present study were to develop a recipe for the solidification of artificial evaporator concentrates, (AEC), containing H3BO3 and measure the compressive strength of the waste/cement matrix over a period of 4 years. The confirmation of the previously reported retarding properties of H3BO3 and the studies of AEC without H3BO3 were also included as a part of this work. Finally, thermodynamic calculations were used as a tool in order to predict the evolution of the mineralogy and integrity for the different cement-waste specimens over very long periods of time, i.e. up to about 100 000 years. The most important finding was that when an optimized waste/cement matrix recipe was used the compressive strength increased during the entire 4 year period and no signs of degradation were noticed. It was also found that the long-term performance of the waste matrices is to a large extent site-specific. In general, the composition of the infiltrating water is more influential than the waste matrices, both on the degradation of the waste matrices itself as well as on the engineered barriers. (author)

  2. Diffusion from cylindrical waste forms

    International Nuclear Information System (INIS)

    The diffusion of a single component material from a finite cylindrical waste form, initially containing a uniform concentration of the material, is investigated. Under the condition that the cylinder is maintained in a well-stirred bath, expressions for the fractional inventory leached and the leach rate are derived with allowance for the possible permanent immobilization of the diffusant through its decay to a stable product and/or its irreversible reaction with the waste form matrix. The usefulness of the reported results in nuclear waste disposal applications is emphasized. The results reported herein are related to those previously derived at Oak Ridge National Laboratory by Bell and Nestor. A numerical scheme involving the partial decoupling of nested infinite summations and the use of rapidly converging rational approximants is recommended for the efficient implementation of the expressions derived to obtain reliable estimates of the bulk diffusion constant and the rate constant describing the diffusant-waste form interaction from laboratory data

  3. Immobilisation of shredded waste in a cement matrix

    International Nuclear Information System (INIS)

    The work covered in the period of this report was aimed at proving the infilling capabilities of waste packages containing shredded paper and plastic simulant waste material held in a basket. The programme required the production of 200 and 500 litre packages and a demonstration that infilling could be attained to give a minimum of voidage in the completed cemented product. The procurement, testing and fitting of level detectors was an important part of this work to demonstrate a means of controlling the process to prevent overfilling of the packages. Evaluation of full-scale cemented products was required to confirm previously reported properties of density and homogeneity in packages produced by the reference encapsulation process and to demonstrate package integrity under sea-disposal conditions. A standard feedstock for the continuity of a long-term programme was required. Such a product, based on an analysis of arisings from plutonium gloveboxes, was produced in bulk and characterised. The previously observed movement of waste during infilling, due to its low density compared with that of the infill grout, required further assessment. During the period, 200, 400 and 500 litre drums required for future active infilling trials were modified and despatched to AERE Harwell for waste loading. These drums were fitted with level detectors and with grout spreader troughs which had been identified during the development programme. A prototype automated Grout Infill Test Rig designed by BNF plc was delivered to Winfrith towards the end of the period for practical assessment trials. (author)

  4. Study on immobilization of high level waste with alkali-activated slag cement

    International Nuclear Information System (INIS)

    A new solidification process using AASC (alkali-activated slag cement) to immobilize HLW (high level waste) is introduced in the article. The AASC incorporated with zeolite and condensed silica fume is used as the matrix of solidified product. High temperature and pressure are not necessary in this process. Under a confined volume the pressure caused by hydrolysis of still explosion to density the matrix of cement. It has high compressive strength, high density, low leachability and high resistance to heat. When the salt loading of HLW in the product is equal to 25 wt% the compressive strength of waste product is 65 ∼ 100 MPa, the porosity is less than 10%, the leachability for Cs is 10-5 g · d-1 and for Sr is 10-7 g · cm-2 · d-1. The performance of this kind cement is comparable with the vitrifying of HLW, but the process is simpler than vitrifying of waste method. The mechanism of the immobilization and the forms of nuclides in AASC waste product are also discussed

  5. Effect of Ground Waste Concrete Powder on Cement Properties

    Directory of Open Access Journals (Sweden)

    Xianwei Ma

    2013-01-01

    Full Text Available The paste/mortar attached to the recycled aggregate decreases the quality of the aggregate and needs to be stripped. The stripped paste/mortar is roughly 20% to 50% in waste concrete, but relevant research is very limited. In this paper, the effects of ground waste concrete (GWC powder, coming from the attached paste/mortar, on water demand for normal consistency, setting time, fluidity, and compressive strength of cement were analyzed. The results show that the 20% of GWC powder (by the mass of binder has little effect on the above properties and can prepare C20 concrete; when the sand made by waste red clay brick (WRB replaces 20% of river sand, the strength of the concrete is increased by 17% compared with that without WRB sand.

  6. Properties of lightweight cement-based composites containing waste polypropylene

    Science.gov (United States)

    Záleská, Martina; Pavlíková, Milena; Pavlík, Zbyšek

    2016-07-01

    Improvement of buildings thermal stability represents an increasingly important trend of the construction industry. This work aims to study the possible use of two types of waste polypropylene (PP) for the development of lightweight cement-based composites with enhanced thermal insulation function. Crushed PP waste originating from the PP tubes production is used for the partial replacement of silica sand by 10, 20, 30, 40 and 50 mass%, whereas a reference mixture without plastic waste is studied as well. First, basic physical and thermal properties of granular PP random copolymer (PPR) and glass fiber reinforced PP (PPGF) aggregate are studied. For the developed composite mixtures, basic physical, mechanical, heat transport and storage properties are accessed. The obtained results show that the composites with incorporated PP aggregate exhibit an improved thermal insulation properties and acceptable mechanical resistivity. This new composite materials with enhanced thermal insulation function are found to be promising materials for buildings subsoil or floor structures.

  7. Modified sulphur cement: A low porosity encapsulation material for low, medium and alpha waste

    International Nuclear Information System (INIS)

    Modified sulphur cement, available under the trade name Chement 2000, is a thermoplastic candidate material for the matrix of low, intermediate and alpha radioactive waste. The main source of sulphur is the desulphurization of fossil fuels. In view of the future increase of this product a modified compound of sulphur has been developed at the US Bureau of Mines. Modified sulphur cement as matrix material has properties in common with Portland or blast furnace cement and bitumen. The mechanical strength is comparable to hydraulic cement products. The process to incorporate waste materials is identical to bitumization. The leachability and the resistance to attack by chemicals is nearly the same as for bituminized products. This study showed also that the radiation resistance is high without radiolytic gas production and without change in dimensions (swelling). The rigidity of the matrix is a disadvantage when internal pressures are built up. The thermal conductivity and the heat of combustion of sulphur is low resulting in slow damage to the waste form under fire conditions, even when the temperature of self ignition in air is 2200C. The low leachability, the very slow effective diffusion of H2O and HTO, and the low permeability is due to the small pore diameters in the modified sulphur matrix. The loading capacity of modified sulphur cement depends on grain size and distribution and is for ungraded ashes, precipitates, dried sludges, etc., in the order of 40-50% of weight. The price of Chement 2000 per tonne is equal to those of blown bitumen

  8. MICROBIAL LEACHING OF CHROMIUM FROM SOLIDIFIED WASTE FORMS – A KINETIC STUDY

    Directory of Open Access Journals (Sweden)

    Carmalin Sophia Ayyappan

    2015-06-01

    Full Text Available In this study, Thiobacillus thiooxidans (T. thiooxidans was used to study the microbial stability / degradation of cement-based waste forms. The waste forms contained a chromium salt (CrCl3·6H2O, cement and other additives viz., lime and gypsum in two different proportions. The experimental samples of all the simulated waste forms showed evidence of microbial growth as indicated by substantial increase in sulfate. Chromium leached from the waste forms was found to be lowest in cement – lime solidified waste forms (0.061 mg·l-1 and highest in cement gypsum waste forms (0.22 mg·l-1 after 30 days of exposure. These values were lower than the toxicity characteristic leaching procedure (TCLP, regulatory limit (5 mg·l-1. Model equations based on two shrinking core models (acid dissolution and bulk diffusion model, were used to analyze the kinetics of microbial degradation of cement based waste forms. The bulk diffusion model was observed to fit the data better than the acid dissolution model, as indicated by good correlation coefficients.

  9. Demonstration of Macroencapusulation of Mixed Waste Debris Using Sulfur Polymer Cement

    International Nuclear Information System (INIS)

    This report covers work performed during FY 1997 as part of the Evaluation of Sulfur Polymer Cement Fast-Track System Project. The project is in support of the ''Mercury Working Group/Mercury Treatment Demonstrations - Oak Ridge'' and is described in technical task plan (TTP) OR-16MW-61. Macroencapsulation is the treatment technology required for debris by the U.S. Environmental Protection Agency Land Disposal Restrictions (LDR) under the Resource Conservation and Recovery Act. Based upon the results of previous work performed at Oak Ridge, the concept of using sulfur polymer cement (SPC) for this purpose was submitted to the Mixed Waste Focus Area (MWFA). Because of the promising properties of the material, the MWFA accepted this Quick Win project, which was to demonstrate the feasibility of macroencapsulation of actual mixed waste debris stored on the Oak Ridge Reservation. The waste acceptance criteria from Envirocare, Utah, were chosen as a standard for the determination of the final waste form produced. During this demonstration, it was shown that SPC was a good candidate for macroencapsulation of mixed waste debris, especially when the debris pieces were dry. The matrix was found to be quite easy to use and, once the optimum operating conditions were identified, very straightforward to replicate for batch treatment. The demonstration was able to render LDR compliant more than 400 kg of mixed wastes stored at the Oak Ridge National Laboratory

  10. The Transformation of Coal-Mining Waste Minerals in the Pozzolanic Reactions of Cements

    Directory of Open Access Journals (Sweden)

    Rosario Giménez-García

    2016-06-01

    Full Text Available The cement industry has the potential to become a major consumer of recycled waste materials that are transformed and recycled in various forms as aggregates and pozzolanic materials. These recycled waste materials would otherwise have been dumped in landfill sites, leaving hazardous elements to break down and contaminate the environment. There are several approaches for the reuse of these waste products, especially in relation to clay minerals that can induce pozzolanic reactions of special interest in the cement industry. In the present paper, scientific aspects are discussed in relation to several inert coal-mining wastes and their recycling as alternative sources of future eco-efficient pozzolans, based on activated phyllosilicates. The presence of kaolinite in this waste indicates that thermal treatment at 600 °C for 2 h transformed these minerals into a highly reactive metakaolinite over the first seven days of the pozzolanic reaction. Moreover, high contents of metakaolinite, together with silica and alumina sheet structures, assisted the appearance of layered double hydroxides through metastable phases, forming stratlingite throughout the main phase of the pozzolanic reaction after 28 days (as recommended by the European Standard as the reaction proceeded.

  11. Characterisation of cemented/bituminized LAW and MAW waste products

    International Nuclear Information System (INIS)

    In the context of work for characterising low and medium activity waste products, investigations were carried out to determine the release of radioactivity from binding waste in given accidents, such as mechanical and thermal loading for the operating phase of a final store. The effects of mechanical loads on MAW cement products and the effects of thermal laods on MAW cement and MAW bitumen products were examined. The release of fine dust reaching the lungs, with a particle size of ≤10 μm from a 200 litre roller seam cement binder with a maximum mechanical load of 3x105 Nm covering the accident case is about 1.5 g and therefore corresponds to ≅ 10-4% of the total radio-activity inventory for homogeneous products. With thermal loading (60 minute oil fire, 8000C) ≅ 10-3% of the radioactivity inventory is released via the release of water from the waste binder. The activity release of MAW bitumen products containing NaNO3 (175 litre drum) with thermal load is considerably higher, as due to the NaNO3 content of the products, after an induction period of about 20 minutes there is an exothermal reaction between the bitumen and the NaNO3, which leads to burning of the bitumen with considerable aerosol formation. The Na losses are about 32% and the Pu losses, derived from the results of laboratory experiments with samples containing Eu and Pu and samples containing Eu on the original size, are only 15% maximum, even with complete burn up. It was shown for all the investigations with samples of the original size that the effects of the load cases considered can be reduced or completely avoided by additional packing (concrete shielding). (orig./RB)

  12. Development of test methods for assessing microbial influenced degradation of cement-solidified radioactive and industrial waste

    International Nuclear Information System (INIS)

    This paper reports on the development of accelerated tests for evaluating microbial influenced degradation (MID) of cement-solidified wastes. An existing U.S. Nuclear Regulatory Commission accelerated test cannot distinguish between degradation caused by biogenic acid produced under optimal conditions in a bioreactor and that caused by active biofilms formed on the waste materials. Nutrient limitations were also observed that would significantly limit the activity of any developing biofilm. Results from this work have shown that it is possible to modify this test to remove nutrient limitations and enable the effects of MID resulting from active biofilms to be examined. Aggressive MID microorganisms can form a biofilm on the surface of cement-solidified waste so that when nutrients are provided the microbes remain active. Elemental mass loss data from exposed solidified waste forms indicate the continued development and growth of microbes on the surface of samples

  13. Mechanism and Preventive Technology of the Thaumasite Form of Sulfate Attack on Cement Mortars

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    The deterioration mechanism of thaumasite towards cement or concrete structure and the deterioration pattern of in-situ construction caused by the formation of thaumasite were studied in this paper. To improve the TSA (the thaumasite form of sulfate attack) resistance, the cement type, water to cement ratios, the mineral admixture and the circumstance factors should be taken into consideration.

  14. Reuse of grits waste for the production of soil--cement bricks.

    Science.gov (United States)

    Siqueira, F B; Holanda, J N F

    2013-12-15

    This investigation focuses on the reuse of grits waste as a raw material for replacing Portland cement by up to 30 wt.% in soil-cement bricks. The grits waste was obtained from a cellulose factory located in south-eastern Brazil. We initially characterized the waste sample with respect to its chemical composition, X-ray diffraction, fineness index, morphology, pozzolanic activity, and pollution potential. Soil-cement bricks were then prepared using the waste material and were tested to determine their technological properties (e.g., water absorption, apparent density, volumetric shrinkage, and compressive strength). Microstructural evolution was accompanied by confocal microscopy. It was found that the grits waste is mainly composed of calcite (CaCO3) particles. Our results indicate that grits waste can be used economically, safely, and sustainably at weight percentages of up to 20% to partially replace Portland cement in soil-cement bricks. PMID:24140481

  15. Recent advances in cement solidification of radioactive wastes

    International Nuclear Information System (INIS)

    Advanced cement solidification processes and systems have been developed by SGN to meet changing requirements in radioactive waste processing and packaging and to avoid the difficulties often encountered in waste concreting on an industrial scale. SGN applies a strict development methodology to ensure integration of the most recent information on chemical behavior of solidified wastes plus compliance with the precise needs of waste producers and evolving regulatory requirements concerning waste package storage and disposal. Based on a hierarchical definition of objectives, this methodology was implemented following an overall study on radwaste concreting performed in 1983 and 1984 for Electricite de France (EdF), France's national electric power utility. It ensures that industrial and regulatory factors are fully considered from the start of development work. It also constrains development in the direction of true process optimization and guarantees compliance with defined objectives. The methodology has helped SGN develop concreting processes adapted to various types of radioactive waste. The most widely employed processes are first briefly described in this paper. It then presents continuous and batch systems using these processes, focusing on technological features chosen at a very early stage in development

  16. Waste form development program. Annual report, October 1982-September 1983

    International Nuclear Information System (INIS)

    This report provides a summary of the work conducted for the Waste Form Development/Test Program at Brookhaven National Laboratory in FY 1983 under the sponsorship of the US Department of Energy's Low-Level Waste Management Program. The primary focus of this work is the investigation of new solidification agents which will provide improved immobilization of low-level radioactive wastes in an efficient, cost-effective manner. A working set of preliminary waste form evaluation criteria which could impact upon the movement of radionuclides in the disposal environment was developed. The selection of potential solidification agents for further investigation is described. Two thermoplastic materials, low-density polyethylene and a modified sulfur cement were chosen as primary candidates for further study. Three waste types were selected for solidification process development and waste form property evaluation studies which represent both new volume reduction wastes (dried evaporator concentrates and incinerator ash) and current problem wastes (ion exchange resins). Preliminary process development scoping studies were conducted to verify the compatibility of selected solidification agents and waste types and the potential for improved solidification. Waste loadings of 60 wt % Na2SO4, 25 wt % H3BO3, 25 wt % incinerator ash and 50 wt % dry ion exchange resin were achieved using low density polyethylene as a matrix material. Samples incorporating 65 wt % Na2SO4, 40 wt % H3BO3, 20 wt % incinerator ash and 40 wt % dry ion exchange resin were successfully solidified in modified sulfur cement. Additional improvements are expected for both matrix materials as process parameters are optimized. Several preliminary property evaluation studies were performed to provide the basis for an initial assessment of waste form acceptability. These included a two-week water immersion test and compressive load testing

  17. Stabilization/solidification of hazardous and radioactive wastes with alkali-activated cements

    International Nuclear Information System (INIS)

    This paper reviews progresses on the use of alkali-activated cements for stabilization/solidification of hazardous and radioactive wastes. Alkali-activated cements consist of an alkaline activator and cementing components, such as blast furnace slag, coal fly ash, phosphorus slag, steel slag, metakaolin, etc., or a combination of two or more of them. Properly designed alkali-activated cements can exhibit both higher early and later strengths than conventional portland cement. The main hydration product of alkali-activated cements is calcium silicate hydrate (C-S-H) with low Ca/Si ratios or aluminosilicate gel at room temperature; C-S-H, tobmorite, xonotlite and/or zeolites under hydrothermal condition, no metastable crystalline compounds such as Ca(OH)2 and calcium sulphoaluminates exist. Alkali-activated cements also exhibit excellent resistance to corrosive environments. The leachability of contaminants from alkali-activated cement stabilized hazardous and radioactive wastes is lower than that from hardened portland cement stabilized wastes. From all these aspects, it is concluded that alkali-activated cements are better matrix for solidification/stabilization of hazardous and radioactive wastes than Portland cement

  18. Ettringite and C-S-H Portland cement phases for waste ion immobilization: A review

    International Nuclear Information System (INIS)

    The formation, structure and chemistry of the ettringite and C-S-H phases of Portland cement have been reviewed as they relate to waste ion immobilization. The purpose of this review was to investigate the use of Portland cement as a host for priority metallic pollutants as identified by the Environmental Protection Agency and as a host for radioactive waste ions as identified in 40 CFR 191. Ettringite acts as host to a number of these ions in both the columnar and channel sections of the crystal structure. Substitutions have been made at the calcium, aluminum, hydroxide and sulfate sites. C-S-H also hosts a number of the waste species in both ionic and salt form. Immobilization mechanisms for C-S-H include sorption, phase mixing and substitution. The following ions have not apparently been reported as specifically immobilized by one of these phases: Ag, Am, Np, Pu, Ra, Tc, Th and Sn; however, some of these ions are immobilized by Portland cement

  19. Environmental production : use of waste materials in cement kilns in China

    OpenAIRE

    Wang, Ning

    2008-01-01

    This report mainly talks about utilizing the cement kiln to dispose wastes. In China, there are huge amounts of wastes can be produced every year. China government pays more attention to the environmental protection. The government wants to dispose the wastes securely. The cement kiln is a good ‘place’ to take the wastes. The cement kiln has a high temperature, long remaining time, and can solidify the heavy metals, dispose the solid, semi-solid or liquid wastes. To dispose the wa...

  20. Correlation of 137Cs leachability from small-scale to large-scale waste forms

    International Nuclear Information System (INIS)

    A study correlating the leachability of 137Cs from small-scale to large-scale cement forms was performed. The waste forms consisted of (a) organic ion exchange resins incorporated in Portland I cement, with a waste-to-cement ratio of 0.6 and a water-to-cement ratio of 0.4 (as free water) and (b) boric acid waste (12% solution) incorporated in Portland III cement, with a waste-to-cement ratio of 0.7. 137Cs was added to both waste types prior to solidification. The sample dimensions varied from 1 in. x 1 in. to 22 in. x 22 in. (diameter x height). Leach data extending over a period of 260 days were obtained using a modified IAEA leach test. A method based on semi-infinite plane source diffusion model was applied to interpret the leach data. A derived mathematical expression allows prediction of the amount of 137Cs leached from the forms as a function of leaching time and waste form dimensions. A reasonably good agreement between the experimental and calculated data was obtained. 4 figures, 6 tables

  1. Physicochemical changes of cements by ground water corrosion in radioactive waste storage; Evolucion fisicoquimica de los cementos por corrosion de aguas subterraneas en un almacen de desechos radioactivos

    Energy Technology Data Exchange (ETDEWEB)

    Contreras R, A.; Badillo A, V. E.; Robles P, E. F. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Nava E, N. [Instituto Mexicano del Petroleo, Eje Central Lazaro Cardenas No. 152, Col. San Bartolo Atepehuacan, 07730 Mexico D. F. (Mexico)], e-mail: aida.contreras@inin.gob.mx

    2009-10-15

    Knowing that the behavior of cementations materials based on known hydraulic cement binder is determined essentially by the physical and chemical transformation of cement paste (water + cement) that is, the present study is essentially about the cement paste evolution in contact with aqueous solutions since one of principal risks in systems security are the ground and surface waters, which contribute to alteration of various barriers and represent the main route of radionuclides transport. In this research, cements were hydrated with different relations cement-aqueous solution to different times. The pastes were analyzed by different solid observation techniques XRD and Moessbauer with the purpose of identify phases that form when are in contact with aqueous solutions of similar composition to ground water. The results show a definitive influence of chemical nature of aqueous solution as it encourages the formation of new phases like hydrated calcium silicates, which are the main phases responsible of radionuclides retention in a radioactive waste storage. (Author)

  2. NNWSI waste form testing program

    International Nuclear Information System (INIS)

    A waste form testing program has been developed to ensure that the release rate of radionuclides from the engineered barrier system will meet NRC and EPA regulatory requirements. Waste form performance testing will be done under unsaturated, low water availability conditions which represent the expected repository conditions. Testing will also be done under conditions of total immersion of the waste form in repository-type water to cover the possibility that localized portions of the repository might contain standing water. Testing of reprocesses waste forms for CHLW and DHLW will use reaction vessels fabricated from Topopah Spring tuff. Chemical elements which are expected to show the highest release rates in the mildly oxidizing environment of the Topopah Spring tuff horizon at Yucca Mountain are Np and Tc. To determine the effect of residual canister material and of corrosion products from the canister/overpack, waste form testing will be done in the presence of these materials. The release rate of all radionuclides which are subject to NRC and EPA regulations will be measured, and the interactive effects of the released radionuclide and the rock reaction vessels will be determined. The testing program for spent fuel will determine the release rate from bare spent fuel pellets and from Zircaloy clad spent fuel where the cladding contains minor defects. A metal testing program for Zircaloy will establish the expected lifetime of the cladding material. Estimation of the state of cladding for fuel presently in reactor pool storage will provide baseline data for Zircaloy containment credit. 9 references, 4 figures

  3. Influence of grain size variation of rice husk ash and cement water phase on radioactive waste cementation

    International Nuclear Information System (INIS)

    This research was conducted to determine grain size of rice husk ash (RHA) and cement water phase (CWP) be agreeable, so it would be to get result strength and leach rate maximum. The cementation process has be done by mixing liquid waste simulation ware made of strontium nitrate with concentration 65 ppm, type I Portland cement, filler material, rice husk ash and water. Mixture of cement mortar was casted on cylinder with diameter of 28 mm and height of 56 mm. the addition of cement water phase was varied 20v/o, 30v/o and 40v/o of cement volume. Variation of rice husk ash grain size was -40, -100, and -200 mesh as many as 25v/o from mixing volume total. After being casted, cement mortars were dried naturally for 28 days. Than the cement mortars were tested with compressive strength test by Universal Testing Machine and analysis of strontium concentration using Atomic Absorption Spectrometer to determine the leach rate after 21 days. (author)

  4. Coated particle waste form development

    International Nuclear Information System (INIS)

    Coated particle waste forms have been developed as part of the multibarrier concept at Pacific Northwest Laboratory under the Alternative Waste Forms Program for the Department of Energy. Primary efforts were to coat simulated nuclear waste glass marbles and ceramic pellets with low-temperature pyrolytic carbon (LT-PyC) coatings via the process of chemical vapor deposition (CVD). Fluidized bed (FB) coaters, screw agitated coaters (SAC), and rotating tube coaters were used. Coating temperatures were reduced by using catalysts and plasma activation. In general, the LT-PyC coatings did not provide the expected high leach resistance as previously measured for carbon alone. The coatings were friable and often spalled off the substrate. A totally different concept, thermal spray coating, was investigated at PNL as an alternative to CVD coating. Flame spray, wire gun, and plasma gun systems were evaluated using glass, ceramic, and metallic coating materials. Metal plasma spray coatings (Al, Sn, Zn, Pb) provided a two to three orders-of-magnitude increase in chemical durability. Because the aluminum coatings were porous, the superior leach resistance must be due to either a chemical interaction or to a pH buffer effect. Because they are complex, coated waste form processes rank low in process feasibility. Of all the possible coated particle processes, plasma sprayed marbles have the best rating. Carbon coating of pellets by CVD ranked ninth when compared with ten other processes. The plasma-spray-coated marble process ranked sixth out of eleven processes

  5. Coated particle waste form development

    Energy Technology Data Exchange (ETDEWEB)

    Oma, K.H.; Buckwalter, C.Q.; Chick, L.A.

    1981-12-01

    Coated particle waste forms have been developed as part of the multibarrier concept at Pacific Northwest Laboratory under the Alternative Waste Forms Program for the Department of Energy. Primary efforts were to coat simulated nuclear waste glass marbles and ceramic pellets with low-temperature pyrolytic carbon (LT-PyC) coatings via the process of chemical vapor deposition (CVD). Fluidized bed (FB) coaters, screw agitated coaters (SAC), and rotating tube coaters were used. Coating temperatures were reduced by using catalysts and plasma activation. In general, the LT-PyC coatings did not provide the expected high leach resistance as previously measured for carbon alone. The coatings were friable and often spalled off the substrate. A totally different concept, thermal spray coating, was investigated at PNL as an alternative to CVD coating. Flame spray, wire gun, and plasma gun systems were evaluated using glass, ceramic, and metallic coating materials. Metal plasma spray coatings (Al, Sn, Zn, Pb) provided a two to three orders-of-magnitude increase in chemical durability. Because the aluminum coatings were porous, the superior leach resistance must be due to either a chemical interaction or to a pH buffer effect. Because they are complex, coated waste form processes rank low in process feasibility. Of all the possible coated particle processes, plasma sprayed marbles have the best rating. Carbon coating of pellets by CVD ranked ninth when compared with ten other processes. The plasma-spray-coated marble process ranked sixth out of eleven processes.

  6. Sulfur polymer cement, a solidification and stabilization agent for hazardous and radioactive wastes

    International Nuclear Information System (INIS)

    Hydraulic cements have been the primary radioactive waste stabilization agents in the United States for 50 years. Twelve years ago, Brookhaven National Laboratory was funded by the Department of Energy's Defense Low-Level Waste Management Program to test and develop sulfur polymer cement (SPC). It has stabilized routine wastes as well as some troublesome wastes with high waste-to-agent ratios. The Department of Energy's Hazardous Waste Remedial Action Program joined the effort by providing funding for testing and developing sulfur polymer cement as a hazardous-waste stabilization agent. Sulfur polymer cement has passed all the laboratory scale tests required by the US Environmental Protection Agency and US Nuclear Regulatory Commission. Two decades of tests by the US Bureau of Mines and private concrete contractors indicate this agent is likely to exceed other agents in longevity. This bulletin provides technical data from pertinent tests conducted by these various entities

  7. Leaching of nuclear waste glass in cement pore water: effect of calcium in solution

    International Nuclear Information System (INIS)

    In the French geological repository concept, intermediate-level vitrified wastes could be disposed of in a cement medium. The glass dissolution mechanisms and kinetics, expected to depend strongly on the chemical composition and pH of the leaching water, were studied in various cement pore water compositions corresponding to different stages of cement aging. In this study, we focused on the effects induced by cement pore water at equilibrium with respect to Portlandite (pH[25 C] = 12.4). A decrease in the maximum glass dissolution rate due to the effect of calcium was clearly observed compared to the reference medium, i.e. at the same pH in KOH solution. At higher reaction progress, calcium in solution was almost totally consumed after a few days, probably due to the formation of Calcium Silicate Hydrate (C-S-H) phases with silicon leached from the glass. Two assumptions can be proposed to explain the effect of calcium on the initial regime: either the calcium from solution reacts with silicon released to form a C-S-H passivating layer at the glass/solution interface, or calcium compensates two non-bridging oxygen (Si-O-) in the altered layer, which could decrease the hydrolysis of silicon bonds. (authors)

  8. Secondary Waste Cast Stone Waste Form Qualification Testing Plan

    Energy Technology Data Exchange (ETDEWEB)

    Westsik, Joseph H.; Serne, R. Jeffrey

    2012-09-26

    The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is being constructed to treat the 56 million gallons of radioactive waste stored in 177 underground tanks at the Hanford Site. The WTP includes a pretreatment facility to separate the wastes into high-level waste (HLW) and low-activity waste (LAW) fractions for vitrification and disposal. The LAW will be converted to glass for final disposal at the Integrated Disposal Facility (IDF). Cast Stone – a cementitious waste form, has been selected for solidification of this secondary waste stream after treatment in the ETF. The secondary-waste Cast Stone waste form must be acceptable for disposal in the IDF. This secondary waste Cast Stone waste form qualification testing plan outlines the testing of the waste form and immobilization process to demonstrate that the Cast Stone waste form can comply with the disposal requirements. Specifications for the secondary-waste Cast Stone waste form have not been established. For this testing plan, Cast Stone specifications are derived from specifications for the immobilized LAW glass in the WTP contract, the waste acceptance criteria for the IDF, and the waste acceptance criteria in the IDF Permit issued by the State of Washington. This testing plan outlines the testing needed to demonstrate that the waste form can comply with these waste form specifications and acceptance criteria. The testing program must also demonstrate that the immobilization process can be controlled to consistently provide an acceptable waste form product. This testing plan also outlines the testing needed to provide the technical basis for understanding the long-term performance of the waste form in the disposal environment. These waste form performance data are needed to support performance assessment analyses of the long-term environmental impact of the secondary-waste Cast Stone waste form in the IDF

  9. Study of the sulphate expansion phenomenon in concrete: behaviour of the cemented radioactive wastes containing sulphate

    International Nuclear Information System (INIS)

    -particulate electric repulsion of colloidal ettringite formed from the transformation of the phase U and in presence of the lime. The kinetics of the transformation process is analysed based upon the diffusion theory. The transformation from thenardite to mirabilite is not justified. In summary, the U phase can induce expansion mainly through two mechanisms which act individually or simultaneously: the formation of the secondary phase U and the transformation of the phase U to ettringite. This work constitutes the first step necessary for predicting the service life of the structures storing the cemented radioactive wastes. (authors)

  10. HYDRATION AND PROPERTIES OF BLENDED CEMENT SYSTEMS INCORPORATING INDUSTRIAL WASTES

    Directory of Open Access Journals (Sweden)

    Heikal M.

    2013-06-01

    Full Text Available This paper aims to study the characteristics of ternary blended system, namely granulated blast-furnace slag (WCS, from iron steel company and Homra (GCB from Misr Brick (Helwan, Egypt and silica fume (SF at 30 mass % pozzolanas and 70 mass % OPC. The required water of standard consistency and setting times were measured as well as physico-chemical and mechanical characteristics of the hardened cement pastes were investigated. Some selected cement pastes were tested by TGA, DTA and FT-IR techniques to investigate the variation of hydrated products of blended cements. The pozzolanic activity of SF is higher than GCB and WCS. The higher activity of SF is mainly due to its higher surface area than the other two pozzolanic materials. On the other side, GCB is more pozzolanic than WCS due to GCB containing crystalline silica quartz in addition to an amorphous phase. The silica quartz acts as nucleating agents which accelerate the rate of hydration in addition to its amorphous phase, which can react with liberating Ca(OH2 forming additional hydration products.

  11. TSA waste stream and final waste form composition

    Energy Technology Data Exchange (ETDEWEB)

    Grandy, J.D.; Eddy, T.L.; Anderson, G.L.

    1993-01-01

    A final vitrified waste form composition, based upon the chemical compositions of the input waste streams, is recommended for the transuranic-contaminated waste stored at the Transuranic Storage Area of the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory. The quantities of waste are large with a considerable uncertainty in the distribution of various waste materials. It is therefore impractical to mix the input waste streams into an ``average`` transuranic-contaminated waste. As a result, waste stream input to a melter could vary widely in composition, with the potential of affecting the composition and properties of the final waste form. This work examines the extent of the variation in the input waste streams, as well as the final waste form under conditions of adding different amounts of soil. Five prominent Rocky Flats Plant 740 waste streams are considered, as well as nonspecial metals and the ``average`` transuranic-contaminated waste streams. The metals waste stream is the most extreme variation and results indicate that if an average of approximately 60 wt% of the mixture is soil, the final waste form will be predominantly silica, alumina, alkaline earth oxides, and iron oxide. This composition will have consistent properties in the final waste form, including high leach resistance, irrespective of the variation in waste stream. For other waste streams, much less or no soil could be required to yield a leach resistant waste form but with varying properties.

  12. TSA waste stream and final waste form composition

    International Nuclear Information System (INIS)

    A final vitrified waste form composition, based upon the chemical compositions of the input waste streams, is recommended for the transuranic-contaminated waste stored at the Transuranic Storage Area of the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory. The quantities of waste are large with a considerable uncertainty in the distribution of various waste materials. It is therefore impractical to mix the input waste streams into an ''average'' transuranic-contaminated waste. As a result, waste stream input to a melter could vary widely in composition, with the potential of affecting the composition and properties of the final waste form. This work examines the extent of the variation in the input waste streams, as well as the final waste form under conditions of adding different amounts of soil. Five prominent Rocky Flats Plant 740 waste streams are considered, as well as nonspecial metals and the ''average'' transuranic-contaminated waste streams. The metals waste stream is the most extreme variation and results indicate that if an average of approximately 60 wt% of the mixture is soil, the final waste form will be predominantly silica, alumina, alkaline earth oxides, and iron oxide. This composition will have consistent properties in the final waste form, including high leach resistance, irrespective of the variation in waste stream. For other waste streams, much less or no soil could be required to yield a leach resistant waste form but with varying properties

  13. UTILIZATION OF AGARWOOD DISTILLATION WASTE IN OILWELL CEMENT AND ITS EFFECT ON FREE WATER AND POROSITY

    Directory of Open Access Journals (Sweden)

    Arina Sauki

    2013-10-01

    Full Text Available The intent of this research is to utilize the waste produced by distillation process of Agarwood oil and convert it into a profitable oilwell cement additive. Common problem during oilwell cementing is free wáter separation. This problem could weaken cement at the top, gas migration problem and non uniform density of cement slurry that are even worst in cementing deviated well. Another concern on cementing design is the porosity of the hardened cement. If the cement is too porous, it can lead to gas migration and casing corrosion. All tests were conducted according to API Specification-10B. Free water test was determined at different concentrations of Agarwood Waste Additive (AWA, different inclination angles and different temperatures. Based on the findings, it was observed that zero free water was produced when 2% BWOC of AWA was used at all angles. The findings also revealed that AWA can maintain good thermal stability as it could maintain zero free water at increased temperature up to 60˚C.  The porosity of AWA cement was comparable with standard API neat cement as the porosity did not differ much at 2% BWOC of AWA. Therefore, it can be concluded that the AWA is suitable to  be used as an additive in oil well cement (OWC  with 2% BWOC is taken as the optimum concentration.

  14. Long Term Behaviour Evaluation of Cement Conditioning Matrices Used for Management of Radioactive Wastes at IFIN-HH

    International Nuclear Information System (INIS)

    The mechanical and structural characterization of the radioactive waste conditioning matrix is very important during the final disposal stage in the radioactive waste management cycle. The conditioning products should be a monolith with acceptable mechanical, chemical and physical properties that are maintained over an appropriate time such that the release of radioactivity from the waste form in the environment is minimized. The aim of this work is the XRD application for the phase identification of matrix that simulates the conditioned radioactive waste and the correlation with mechanical performance. The selected matrices for the study are normal cement with iron precipitates (hydroxide and phosphate), mineral additives (bentonite and volcanic tuff) and complexing agents (tartaric, citric and oxalic acids). The results obtained by this analysis give information about the chemical reactions between the radioactive precipitates and the hydrates, hydrolysis products of the cement. (author)

  15. Cement-based grouts in geological disposal of radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Onofrei, M. [AECL Research, Pinnawa, Manitoba (Canada)

    1996-04-01

    The behavior and performance of a specially developed high-performance cement-based grout has been studied through a combined laboratory and in situ research program conducted under the auspices of the Canadian Nuclear Fuel Waste Management Program (CNFWMP). A new class of cement-based grouts - high-performance grouts-with the ability to penetrate and seal fine fractures was developed and investigated. These high-performance grouts, which were injected into fractures in the granitic rock at the Underground Research Laboratory (URL) in Canada, are shown to successfully reduce the hydraulic conductivity of the rock mass from <10{sup -7} m s{sup -1} to 10{sup -9} m s{sup -1} and to penetrate fissures in the rock with apertures as small as 10 {mu}m. Furthermore, the laboratory studies have shown that this high - performance grout has very low hydraulic conductivity and is highly leach resistant under repository conditions. Microcracks generated in this materials from shrinkage, overstressing or thermal loads are likely to self-seal. The results of these studies suggest that the high-performance grouts can be considered as viable materials in disposal-vault sealing applications. Further work is needed to fully justify extrapolation of the results of the laboratory studies to time scales relevant to performance assessment.

  16. Cement-based grouts in geological disposal of radioactive waste

    International Nuclear Information System (INIS)

    The behavior and performance of a specially developed high-performance cement-based grout has been studied through a combined laboratory and in situ research program conducted under the auspices of the Canadian Nuclear Fuel Waste Management Program (CNFWMP). A new class of cement-based grouts - high-performance grouts-with the ability to penetrate and seal fine fractures was developed and investigated. These high-performance grouts, which were injected into fractures in the granitic rock at the Underground Research Laboratory (URL) in Canada, are shown to successfully reduce the hydraulic conductivity of the rock mass from -7 m s-1 to 10-9 m s-1 and to penetrate fissures in the rock with apertures as small as 10 μm. Furthermore, the laboratory studies have shown that this high - performance grout has very low hydraulic conductivity and is highly leach resistant under repository conditions. Microcracks generated in this materials from shrinkage, overstressing or thermal loads are likely to self-seal. The results of these studies suggest that the high-performance grouts can be considered as viable materials in disposal-vault sealing applications. Further work is needed to fully justify extrapolation of the results of the laboratory studies to time scales relevant to performance assessment

  17. Comparative life cycle analysis of cement made with coal vs hazardous waste as fuel

    International Nuclear Information System (INIS)

    The purpose of this life cycle analysis (LCA) is to compare the life cycle of cement made with coal, the standard fuel used in a cement kiln, versus cement made with hazardous waste-derived fuels. The intent of the study is to determine whether the use of hazardous waste as a fuel in the production of cement could result in an increase in detrimental effects to either health or environment. Those evaluated for potential adverse effect include cement kiln workers, waste transporters, and consumers using the final product for private use. The LCA stages included all the processes involved with cement, including raw materials acquisition, transportation, manufacturing, packaging, distribution, use, recycling, and disposal. The overall conclusions of the LCA are that use of waste fuels instead of coal to make cement: (1) does not increase, and may reduce, the concentration of contaminants in the cement product due to the reduction or elimination of the use of coal; (2) reduces or eliminates use of non-renewable fossil fuels, such as coal, as well as the environmental damage and impacts associated with coal mining; (3) provides a more environmentally beneficial means of destroying many types of wastes than alternative treatment methods, including incineration, thus decreasing the need for waste treatment facilities and capacity; (4) decreases overall emissions during transportation but may increase the overall consequences of accidents or spills; (5) results in cement product which may be packaged, transported, distributed and used in the same manner as cement product made with coal; (6) lowers the cost of cement production; and (7) overall appears to result in less health and environmental impacts

  18. Solid forms for Savannah River Plant radioactive wastes

    International Nuclear Information System (INIS)

    Methods are being developed to immobilize Savannah River Plant wastes in solid forms such as cement, asphalt, or glass. 137Cs and 90Sr are the major biological hazards and heat producers in the alkaline wastes produced at SRP. In the conceptual process being studied, 137Cs removed from alkaline supernates, together with insoluble sludges that contain 90Sr, will be incorporated into solid forms of high integrity and low volume suitable for storage in a retrievable surface storage facility for about 100 years, and for eventual shipment to an off-site repository. Mineralization of 137Cs, or its fixation on zeolite prior to incorporation into solid forms, is also being studied. Economic analyses to reduce costs and fault-tree analyses to minimize risks are being conducted. Methods are being studied for removal of sludge from (and final decontamination of) waste tanks

  19. Analysis by X-Ray images of EVA waste incorporated in Portland Cement

    International Nuclear Information System (INIS)

    The EVA is a copolymer used by Brazilian shoes industries. This material is cut for the manufacture of insoles. This operation generates about 18% of waste. The EVA waste can be reused in incorporation in Portland cement to construction without structural purposes. The aim of this work is to show X-rays images to assessment the space distribution of the wastes in the cement and to evaluate the use of this methodology. Cylindrical specimens were produced according to ABNT - NBR 5738 standards. The volume relation of sand and cement was 3:1, 10% and 30% of waste was incorporated in cement specimens. X-Rays images were obtained of cylindrical specimens in front projection. The images showed that the distribution of the waste is homogeneous, consistent with what was intended in this type of incorporation, which can provide uniformity in test results of compressive strength. (author)

  20. CERAMIC WASTE FORM DATA PACKAGE

    Energy Technology Data Exchange (ETDEWEB)

    Amoroso, J.; Marra, J.

    2014-06-13

    The purpose of this data package is to provide information about simulated crystalline waste forms that can be used to select an appropriate composition for a Cold Crucible Induction Melter (CCIM) proof of principle demonstration. Melt processing, viscosity, electrical conductivity, and thermal analysis information was collected to assess the ability of two potential candidate ceramic compositions to be processed in the Idaho National Laboratory (INL) CCIM and to guide processing parameters for the CCIM operation. Given uncertainties in the CCIM capabilities to reach certain temperatures throughout the system, one waste form designated 'Fe-MP' was designed towards enabling processing and another, designated 'CAF-5%TM-MP' was designed towards optimized microstructure. Melt processing studies confirmed both compositions could be poured from a crucible at 1600{degrees}C although the CAF-5%TM-MP composition froze before pouring was complete due to rapid crystallization (upon cooling). X-ray diffraction measurements confirmed the crystalline nature and phase assemblages of the compositions. The kinetics of melting and crystallization appeared to vary significantly between the compositions. Impedance spectroscopy results indicated the electrical conductivity is acceptable with respect to processing in the CCIM. The success of processing either ceramic composition will depend on the thermal profiles throughout the CCIM. In particular, the working temperature of the pour spout relative to the bulk melter which can approach 1700{degrees}C. The Fe-MP composition is recommended to demonstrate proof of principle for crystalline simulated waste forms considering the current configuration of INL's CCIM. If proposed modifications to the CCIM can maintain a nominal temperature of 1600{degrees}C throughout the melter, drain, and pour spout, then the CAF-5%TM-MP composition should be considered for a proof of principle demonstration.

  1. Secondary waste form testing : ceramicrete phosphate bonded ceramics.

    Energy Technology Data Exchange (ETDEWEB)

    Singh, D.; Ganga, R.; Gaviria, J.; Yusufoglu, Y. (Nuclear Engineering Division); ( ES)

    2011-06-21

    The cleanup activities of the Hanford tank wastes require stabilization and solidification of the secondary waste streams generated from the processing of the tank wastes. The treatment of these tank wastes to produce glass waste forms will generate secondary wastes, including routine solid wastes and liquid process effluents. Liquid wastes may include process condensates and scrubber/off-gas treatment liquids from the thermal waste treatment. The current baseline for solidification of the secondary wastes is a cement-based waste form. However, alternative secondary waste forms are being considered. In this regard, Ceramicrete technology, developed at Argonne National Laboratory, is being explored as an option to solidify and stabilize the secondary wastes. The Ceramicrete process has been demonstrated on four secondary waste formulations: baseline, cluster 1, cluster 2, and mixed waste streams. Based on the recipes provided by Pacific Northwest National Laboratory, the four waste simulants were prepared in-house. Waste forms were fabricated with three filler materials: Class C fly ash, CaSiO{sub 3}, and Class C fly ash + slag. Optimum waste loadings were as high as 20 wt.% for the fly ash and CaSiO{sub 3}, and 15 wt.% for fly ash + slag filler. Waste forms for physical characterizations were fabricated with no additives, hazardous contaminants, and radionuclide surrogates. Physical property characterizations (density, compressive strength, and 90-day water immersion test) showed that the waste forms were stable and durable. Compressive strengths were >2,500 psi, and the strengths remained high after the 90-day water immersion test. Fly ash and CaSiO{sub 3} filler waste forms appeared to be superior to the waste forms with fly ash + slag as a filler. Waste form weight loss was {approx}5-14 wt.% over the 90-day immersion test. The majority of the weight loss occurred during the initial phase of the immersion test, indicative of washing off of residual unreacted

  2. Establishment of PCP composition diagrams for cementations of borate wastes using response surface methodology

    International Nuclear Information System (INIS)

    For the purpose of quality assurance, it is requested by the regulatory authority in the solidification of low-level radioactive waste (LLRW) to implement the 'process control program (PCP)', in which the condition of solidification should be decided in advance of solidification according to the so-called 'PCP composition diagrams', to assure that the condition of solidification is within established process parameters and that the quality of solidified LLRW meets quality criteria. In this paper, PCP composition diagrams for the cementation of radioactive liquid borate wastes were established with response surface methodology, including using the simplex centroid design to allocate experimental conditions and statistical curve-fitting techniques to construct quality models. The constructed models were verified to have a confidence level higher than 95% by doing the lack-of-fit test on them. Quality contours of solidified borate wastes were thus established based on mathematical and statistical principles and have been used as composition diagrams in the process control program of radioactive borate wastes solidification at Ma-An-Shan nuclear power station of Taiwan Power Company. The solidification agent used in this study was a mixture of Portland type-II cement and lime powder with a weight ratio of 1/0.22. Quality contours for solidified borate wastes including free-standing water content, compressive strength, water resistance, thawing and freezing resistance, irradiation resistance, and leaching resistance were established. Characteristics of these quality contours were also discussed. With these contours, the performance of the final waste form can be assured and consequently a volume reduction will also be achieved, when the PCP is implemented. (author)

  3. Fracture analysis of cement treated demolition waste using a lattice model

    NARCIS (Netherlands)

    Xuan, D.; Schlangen, H.E.J.G.; Molenaar, A.A.A.; Houben, L.J.M.

    2013-01-01

    Fracture properties of cement treated demolition waste were investigated using a lattice model. In practice the investigated material is applied as a cement treated road base/subbase course. The granular aggregates used in this material were crushed recycled concrete and masonry. This results in six

  4. Use of waste brick as a partial replacement of cement in mortar

    International Nuclear Information System (INIS)

    The aim of this study is to investigate the use of waste brick as a partial replacement for cement in the production of cement mortar. Clinker was replaced by waste brick in different proportions (0%, 5%, 10%, 15% and 20%) by weight for cement. The physico-chemical properties of cement at anhydrous state and the hydrated state, thus the mechanical strengths (flexural and compressive strengths after 7, 28 and 90 days) for the mortar were studied. The microstructure of the mortar was investigated using scanning electron microscopy (SEM), the mineralogical composition (mineral phases) of the artificial pozzolan was investigated by the X-ray diffraction (XRD) and the particle size distributions was obtained from laser granulometry (LG) of cements powders used in this study. The results obtained show that the addition of artificial pozzolan improves the grinding time and setting times of the cement, thus the mechanical characteristics of mortar. A substitution of cement by 10% of waste brick increased mechanical strengths of mortar. The results of the investigation confirmed the potential use of this waste material to produce pozzolanic cement.

  5. A Thermoelectric Waste-Heat-Recovery System for Portland Cement Rotary Kilns

    Science.gov (United States)

    Luo, Qi; Li, Peng; Cai, Lanlan; Zhou, Pingwang; Tang, Di; Zhai, Pengcheng; Zhang, Qingjie

    2015-06-01

    Portland cement is produced by one of the most energy-intensive industrial processes. Energy consumption in the manufacture of Portland cement is approximately 110-120 kWh ton-1. The cement rotary kiln is the crucial equipment used for cement production. Approximately 10-15% of the energy consumed in production of the cement clinker is directly dissipated into the atmosphere through the external surface of the rotary kiln. Innovative technology for energy conservation is urgently needed by the cement industry. In this paper we propose a novel thermoelectric waste-heat-recovery system to reduce heat losses from cement rotary kilns. This system is configured as an array of thermoelectric generation units arranged longitudinally on a secondary shell coaxial with the rotary kiln. A mathematical model was developed for estimation of the performance of waste heat recovery. Discussions mainly focus on electricity generation and energy saving, taking a Φ4.8 × 72 m cement rotary kiln as an example. Results show that the Bi2Te3-PbTe hybrid thermoelectric waste-heat-recovery system can generate approximately 211 kW electrical power while saving 3283 kW energy. Compared with the kiln without the thermoelectric recovery system, the kiln with the system can recover more than 32.85% of the energy that used to be lost as waste heat through the kiln surface.

  6. Studies Involving Immobilization Of Hazardous Wastes In Cement-ilmenite Matrix

    International Nuclear Information System (INIS)

    Ilmenite was added to Ordinary Portland Cement to Modify the characteristic properties of the matrix as density, compressive strength and thermal stability . Coal tar and radiocesium were solidified as hazardous waste in cement-ilmenite matrix. The physical properties as density, sitting times and porosity were studied. The mechanical properties as compressive strength values and the chemical properties as leaching were measured

  7. ASHES AS AN AGENT FOR CEMENT-LIME BASED SOLIDIFICATION/STABILIZATION OF THE HAZARDOUS WASTE

    Directory of Open Access Journals (Sweden)

    Barbora Lyčkova

    2008-12-01

    Full Text Available One of the common treatment methods for the hazardous waste is the cement and cement-lime based solidification/stabilization (S/S. This article deals with the possibility of currently used recipe modification using fluidized bed heating plant ashes as an agent.

  8. ASHES AS AN AGENT FOR CEMENT-LIME BASED SOLIDIFICATION/STABILIZATION OF THE HAZARDOUS WASTE

    OpenAIRE

    Barbora Lyčkova; Vladimir Huda

    2008-01-01

    One of the common treatment methods for the hazardous waste is the cement and cement-lime based solidification/stabilization (S/S). This article deals with the possibility of currently used recipe modification using fluidized bed heating plant ashes as an agent.

  9. Stability evaluation for cement package containing radioactive waste

    International Nuclear Information System (INIS)

    In order to provide stable cement packages, ettringite formation, a major cause of cement deterioration, was studied theoretically and experimentally. A computer program was developed to calculate the chemical equilibrium compositions of a complex cement system. Higher curing temperature and the addition of NaOH were identified as effective methods to avoid ettringite formation. These findings were confirmed by measuring the amount of ettringite in solidified cement by an X-ray diffraction method

  10. Characterization of low and medium-level radioactive waste forms. Joint annual progress report 1982

    International Nuclear Information System (INIS)

    The work reported was carried out during the second year of the Commission of the European Communities programme on the characterization of low and medium-level waste forms. Ten reference waste forms plus others of special national interest have been identified covering PWR, BWR, GCR and reprocessing wastes. The immobilizing media include the three main matrices: cement, polymers and bitumen, and a glass. Characterization is viewed as one input to quality assurance of the waste form and covers: waste-matrix compatibility, radiation effects, leaching, microbiological attack, shrinkage and swelling, ageing processes and thermal effects. The aim is a balanced programme of comparative data, predictive modelling and an understanding of basic mechanisms

  11. Summary report on the development of a cement-based formula to immobilize Hanford facility waste

    International Nuclear Information System (INIS)

    This report recommends a cement-based grout formula to immobilize Hanford Facility Waste in the Transportable Grout Facility (TGF). Supporting data confirming compliance with all TGF performance criteria are presented. 9 refs., 24 figs., 50 tabs

  12. Zeolite P in cements: Its potential for immobilizing toxic and radioactive waste species

    International Nuclear Information System (INIS)

    Zeolite P, approximately (Na2O,CaO)Al2O3(SiO2)2(H2O)4, has been shown to develop spontaneously in appropriate cement formulations at >40 C, and to be a stable phase. Suitable composites can be made from mixtures containing Ca(OH)2 or Portland cement, with high proportions of the pozzolans, metakaolin or class F fly ash. Alternatively, zeolite P is easily prepared in phase-pure form using laboratory chemicals. The latter method was used to obtain zeolite P of composition (CaO)0.9(Na2O)0.1(Al2O3)(SiO2)2.66(H2O)4, on which characterization studies were performed for its sorption potential in cement-analogue environments. RD values are reported for the 25 and 85 C isotherms for a range of initial sorbate concentration (10--10,000 micromol/l). The sorbates investigated were: Cs, Sr, Ba, Pb and U(VI). In water media, zeolite P shows good selectivity for Cs, Sr, Ba and Pb, at 25 and 85 C. The highest RD recorded was for Pb2+, at >800,000 ml/g (1,000 micromol/l initial concentration). In NaOH media, Ba and Sr sorption values remained high. Cs and Pb show a marked decrease in sorption, although RDS are still reasonable, at ∼ 750 and ∼ 400 ml/g, respectively. On account of its large ion size, UO22+ uptake into zeolite P is negligible, remaining in solution or precipitating as soddyite or Na uranate. Cements conditioned to form stable zeolites offer great potential in the treatment of hazardous waste streams

  13. Reactive transport modelling of organic complexing agents in cement stabilized low and intermediate level waste

    Science.gov (United States)

    von Schenck, Henrik; Källström, Klas

    The Swedish final repository for short-lived radioactive waste (SFR 1) is located at Forsmark in Sweden. It holds low and intermediate-level operational waste from the Swedish nuclear power plants, as well as industrial, research-related, and medical waste. A variety of low molecular weight organic compounds are present in the waste or in its matrix. Such compounds can also be formed by chemical degradation of organic macromolecules. These organics can ligate to metal atoms forming stable complexes and also adsorb to the surface of cement, thereby influencing the net release of radionuclides from the repository. This motivates the study of the concentration distribution of complexing agents in the repository as a function of time. The following paper reports the results of mass transport modelling, describing the transport of complexing agents through the cementitous matrix in the rock vault for intermediate-level waste in the SFR 1 repository. Nitrilotriacetate (NTA) and isosaccharinate (ISA) have been investigated, where the former is considered to be non-sorbing and non-reacting, while the latter is produced from cellulose degradation and adsorbs strongly to cement. The 3D model considers advection, diffusion, and sorption of solvated species in cement pore water over a time period of 20,000 years. The model accounts for the spatial distribution of the flow field in the repository structure and also considers changing groundwater flow during the investigated time period. It is found that 99% of the NTA is removed after approximately 4000 years, while 90% of the ISA is retained in the rock vault after 20,000 years. The maximum pore water concentration of ISA is found to be 8.6 mol/m3 after approximately 2300 years, based on the degradation of the deposited amounts of cellulose. Over the investigated time scale, the ligands retained in the repository can redistribute across several waste compartments where the organic compounds were not originally deposited. In

  14. Characterization of Mechanical and Bactericidal Properties of Cement Mortars Containing Waste Glass Aggregate and Nanomaterials

    Directory of Open Access Journals (Sweden)

    Pawel Sikora

    2016-08-01

    Full Text Available The recycling of waste glass is a major problem for municipalities worldwide. The problem concerns especially colored waste glass which, due to its low recycling rate as result of high level of impurity, has mostly been dumped into landfills. In recent years, a new use was found for it: instead of creating waste, it can be recycled as an additive in building materials. The aim of the study was to evaluate the possibility of manufacturing sustainable and self-cleaning cement mortars with use of commercially available nanomaterials and brown soda-lime waste glass. Mechanical and bactericidal properties of cement mortars containing brown soda-lime waste glass and commercially available nanomaterials (amorphous nanosilica and cement containing nanocrystalline titanium dioxide were analyzed in terms of waste glass content and the effectiveness of nanomaterials. Quartz sand is replaced with brown waste glass at ratios of 25%, 50%, 75% and 100% by weight. Study has shown that waste glass can act as a successful replacement for sand (up to 100% to produce cement mortars while nanosilica is incorporated. Additionally, a positive effect of waste glass aggregate for bactericidal properties of cement mortars was observed.

  15. Characteristics solidified cement waste using heavy concrete and light concrete paste generated from KRR-2 and UCP

    International Nuclear Information System (INIS)

    As the number of obsolete research reactors and nuclear facilities increases, dismantling nuclear facilities has become an influential issue. During the decommissioning of nuclear plants and facilities, large quantities of slightly contaminated concrete wastes are generated. In Korea, the decontamination and decommissioning of the retired TRIGA MARK II and III research reactors and a uranium conversion plant at KAERI has been under way. By dismantling KRR-2, more than 260 tons of radioactive concrete wastes were generated among the total 2,000 tons of concrete wastes and more than 60 tons of concrete wastes contaminated with uranium compounds have been generated. Typically, the contaminated layer is only 1∼10mm thick because cement materials are porous media, the penetration of radionuclides may occur up to several centimeters from the surface of a material. Concrete is a structural material which generally consists of a binder (cement), water, and aggregate. The binder is typically a portland cement which comprises the four principal clinker phases tricalcium silicate (Ca3SiO5) and constitutes 50-70%, decalcium silicate (Ca2SiO4), tricalcium aluminate (Ca3Al2O6), and calcium aluminoferrite (Ca4Al2Fe2O10). Cement powder (anhydrous cement) created from the co-grinding of clinkers and gypsum is mixed with waster and hydrate phase are formed. The interaction between highly charged C-S-H particles in the presence of divalent calcium counter ions is strongly attractive because of ion-ion correlations and a negligible entropic repulsion. In the temperature range 100-300 .deg. C, these evolutions are mainly attributed to the loss of the bound water from the C-S-H gel. Similar consequences have been reported for mortars and concretes enhanced sometimes by the appearance of micro-cracks related to the strain incompatibilities between the aggregates and the cement paste. Concrete aggregates are combined mutually strongly by hydrated cement paste. Radionuclides may be found

  16. Characterisation and modelling of blended cements and their application to radioactive waste immobilisation

    International Nuclear Information System (INIS)

    Various aspects of the chemistry of cements, including blends with FA and BFS, pertinent to the immobilization of radioactive waste are described. The methodology and development of a model for predicting the solid and liquid phase composition in aged cement blends are given. Experimental work, backed up by thermodynamic calculations (where possible), has given valuable insight into some of the important interactions between selected (active and inactive) radwaste components and cements. The effects of elevated pressure and temperature on blended cement are also investigated. (author)

  17. Secondary Waste Form Down Selection Data Package – Ceramicrete

    Energy Technology Data Exchange (ETDEWEB)

    Cantrell, Kirk J.; Westsik, Joseph H.

    2011-08-31

    As part of high-level waste pretreatment and immobilized low activity waste processing, liquid secondary wastes will be generated that will be transferred to the Effluent Treatment Facility on the Hanford Site for further treatment. These liquid secondary wastes will be converted to stable solid waste forms that will be disposed in the Integrated Disposal Facility. Currently, four waste forms are being considered for stabilization and solidification of the liquid secondary wastes. These waste forms are Cast Stone, Ceramicrete, DuraLith, and Fluidized Bed Steam Reformer. The preferred alternative will be down selected from these four waste forms. Pacific Northwest National Laboratory is developing data packages to support the down selection process. The objective of the data packages is to identify, evaluate, and summarize the existing information on the four waste forms being considered for stabilization and solidification of the liquid secondary wastes. The information included will be based on information available in the open literature and from data obtained from testing currently underway. This data package is for the Ceramicrete waste form. Ceramicrete is a relatively new engineering material developed at Argonne National Laboratory to treat radioactive and hazardous waste streams (e.g., Wagh 2004; Wagh et al. 1999a, 2003; Singh et al. 2000). This cement-like waste form can be used to treat solids, liquids, and sludges by chemical immobilization, microencapsulation, and/or macroencapsulation. The Ceramicrete technology is based on chemical reaction between phosphate anions and metal cations to form a strong, dense, durable, low porosity matrix that immobilizes hazardous and radioactive contaminants as insoluble phosphates and microencapsulates insoluble radioactive components and other constituents that do not form phosphates. Ceramicrete is a type of phosphate-bonded ceramic, which are also known as chemically bonded phosphate ceramics. The Ceramicrete

  18. Recycling of Glass Wastes in Latvia - Its Application as Cement Substitute in Self-Compacting Concrete

    OpenAIRE

    Kara, P

    2014-01-01

    The rate of treated waste glass within the last 10 years has risen up to 25% due to improved waste glass collecting system in Latvia, however the glass waste recycling infrastructure is still up to date issue due to absence of local recycling companies. Application of glass wastes as cement substitute has significant effect on the properties of concrete making it eco-friendly construction material with several benefits like: decrease of accumulated glass wastes in landfills, the reduction of ...

  19. Physico-chemical transformations of sulfated compounds during the leaching of highly sulfated cemented wastes

    International Nuclear Information System (INIS)

    Cementation of sulfated evaporator concentrates leads to highly sulfated low level wastes, (ca. 25% w/w sodium sulfate solution as mix water), which exhibit the presence of U-phase, a sodium-bearing calcium monosulfphoaluminate-like phase. During the leaching of simulated highly sulfated OPC/BFS cements, cured at room temperature and containing U-phase, sodium sulfate, and ettringite, physico-chemical transformations have been pointed out (transformation of U-phase into ettringite). Samples having the same chemical composition, but cured at high temperature (maximal temperature during curing: 120 C), do not contain ettringite initially, but secondary ettringite is formed during leaching. XRD spectra point out the existence of precipitation fronts (or of phase formation fronts) varying linearly versus the square root of time. The analysis of leaching solutions has provided complementary data used in a code, the aim of which is to assess cement degradation, based on coupling between transport by diffusion and chemical reactions (DIFFUZON code). The U-phase-ettringite transformation is confirmed

  20. Optimisation and adoption of slag based cement for conditioning of intermediate level alkaline radioactive liquid waste in CLEAR-V campaign

    International Nuclear Information System (INIS)

    The ILW is normally treated by resorcinol formaldehyde special type of resin. Another method for management of ILW is by conditioning in cement matrix. Various waste to cement ratios have been tried at lab and plant scale by taking slag based cement and ordinary portland cement. The cement waste products were evaluated for various properties. The final selected waste to cement ratio has been successfully adopted on the plant scale for conditioning of 140 m3 of ILW at SWMF. (author)

  1. Low temperature waste form process intensification

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Cozzi, A. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hansen, E. K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hill, K. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-09-30

    This study successfully demonstrated process intensification of low temperature waste form production. Modifications were made to the dry blend composition to enable a 50% increase in waste concentration, thus allowing for a significant reduction in disposal volume and associated costs. Properties measurements showed that the advanced waste form can be produced using existing equipment and processes. Performance of the waste form was equivalent or better than the current baseline, with approximately double the amount of waste incorporation. The results demonstrate the feasibility of significantly accelerating low level waste immobilization missions across the DOE complex and at environmental remediation sites worldwide.

  2. ANSTO's waste forms for the 31. century

    International Nuclear Information System (INIS)

    ANSTO waste form development for high-level radioactive waste is directed towards practical applications, particularly problematic niche wastes that do not readily lend themselves to direct vitrification. Integration of waste form chemistry and processing method is emphasised. Some longstanding misconceptions about titanate ceramics are dealt with. We have a range of titanate-bearing waste form products aimed at immobilisation of tank wastes and sludges, actinide-rich wastes, INEEL calcines and Na-bearing liquid wastes, Al-rich wastes arising from reprocessing of Al-clad fuels, Mo-rich wastes arising from reprocessing of U-Mo fuels, partitioned Cs-rich wastes, and 99Tc. Waste form production techniques cover hot isostatic and uniaxial pressing, sintering, and cold-crucible melting, and these are strongly integrated into waste form design. Speciation and leach resistance of Cs and alkalis in cementitious products and geo-polymers are being studied. Recently we have embarked on studies of candidate inert matrix fuels for Pu burning. We also have a considerable program directed at basic understanding of the waste forms in regard to crystal chemistry, dissolution behaviour in aqueous media, radiation damage effects and optimum processing techniques. (authors)

  3. Immobilisation/solidification of hazardous toxic waste in cement matrices

    Directory of Open Access Journals (Sweden)

    Macías, A.

    1999-06-01

    Full Text Available Immobilization and solidification of polluting waste, introduced into the industrial sector more than 20 years ago, and throughout last 10 years is being the object of a growing interest for engineers and environment scientists, has become a remarkable standardized process for treatment and management of toxic and hazardous liquid wastes, with special to those containing toxic metals. Experimental monitorization of the behaviour of immobilized waste by solidification and stabilisation in life time safe deposits is not possible, reason why it is essential to develop models predicting adequately the behaviour of structures that have to undergo a range of conditions simulating the environment where they are to be exposed. Such models can be developed only if the basic physical and chemical properties of the system matrix/solidifying-waste are known. In this work immobilization/solidification systems are analyzed stressing out the formulation systems based on Portland cement. Finally, some examples of the results obtained from the study of interaction of specific species of wastes and fixation systems are presented.

    La inmovilización y solidificación de residuos contaminantes, implantada en el sector comercial desde hace más de 20 años y que desde hace diez es objeto de creciente interés por parte de ingenieros y científicos medioambientales, se ha convertido en un proceso estandarizado único para el tratamiento y gestión de residuos tóxicos y peligrosos líquidos y, en especial, de los que contienen metales pesados. La monitorización experimental del comportamiento de un residuo inmovilizado por solidificación y estabilización en el tiempo de vida de un depósito de seguridad no es posible, por lo que es imprescindible desarrollar modelos que predigan satisfactoriamente el comportamiento del sistema bajo un rango representativo de condiciones del entorno de exposición. Tales modelos sólo pueden ser desarrollados si se

  4. Long-term modeling of glass waste in portland cement- and clay-based matrices

    International Nuclear Information System (INIS)

    A set of ''templates'' was developed for modeling waste glass interactions with cement-based and clay-based matrices. The templates consist of a modified thermodynamic database, and input files for the EQ3/6 reaction path code, containing embedded rate models and compositions for waste glass, cement, and several pozzolanic materials. Significant modifications were made in the thermodynamic data for Th, Pb, Ra, Ba, cement phases, and aqueous silica species. It was found that the cement-containing matrices could increase glass corrosion rates by several orders of magnitude (over matrixless or clay matrix systems), but they also offered the lowest overall solubility for Pb, Ra, Th and U. Addition of pozzolans to cement decreased calculated glass corrosion rates by up to a factor of 30. It is shown that with current modeling capabilities, the ''affinity effect'' cannot be trusted to passivate glass if nuclei are available for precipitation of secondary phases that reduce silica activity

  5. Long-term modeling of glass waste in portland cement- and clay-based matrices

    Energy Technology Data Exchange (ETDEWEB)

    Stockman, H.W.; Nagy, K.L. [Sandia National Labs., Albuquerque, NM (United States); Morris, C.E. [Wollongong Univ., NSW (Australia). Dept. of Civil and Mining Engineering

    1995-12-01

    A set of ``templates`` was developed for modeling waste glass interactions with cement-based and clay-based matrices. The templates consist of a modified thermodynamic database, and input files for the EQ3/6 reaction path code, containing embedded rate models and compositions for waste glass, cement, and several pozzolanic materials. Significant modifications were made in the thermodynamic data for Th, Pb, Ra, Ba, cement phases, and aqueous silica species. It was found that the cement-containing matrices could increase glass corrosion rates by several orders of magnitude (over matrixless or clay matrix systems), but they also offered the lowest overall solubility for Pb, Ra, Th and U. Addition of pozzolans to cement decreased calculated glass corrosion rates by up to a factor of 30. It is shown that with current modeling capabilities, the ``affinity effect`` cannot be trusted to passivate glass if nuclei are available for precipitation of secondary phases that reduce silica activity.

  6. Hydration study of limestone blended cement in the presence of hazardous wastes containing Cr(VI)

    International Nuclear Information System (INIS)

    Considering the increasing use of limestone cement manufacture, the present paper tends to characterize limestone behavior in the presence of Cr(VI). The research reported herein provides information regarding the effect of Cr(VI) from industrial wastes in the limestone cement hydration. The cementitious materials were ordinary Portland cement, as reference, and limestone blended cement. The hydration and physicomechanical properties of cementitious materials and the influence of chromium at an early age were studied with X-ray diffraction (XRD), infrared spectroscopy (FTIR), conductimetric and mechanical tests. Portland cement pastes with the addition of Cr(VI) were examined and leaching behavior with respect to water and acid solution were investigated. This study indicates that Cr(VI) modifies the rate and the components obtained during the cement hydration

  7. Performance Test on Polymer Waste Form - 12137

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Se Yup [Korea Nuclear Engineering Co., Ltd., C-504 Bundang Techno-Park 145, Yatap-dong, Bundang-gu, Seongnam-si, Gyeonggi-do, 463-760 (Korea, Republic of)

    2012-07-01

    Polymer solidification was attempted to produce stable waste form for the boric acid concentrates and the dewatered spent resins. The polymer mixture was directly injected into the mold or drum which was packed with the boric acid concentrates and the dewatered spent resins, respectively. The waste form was produced by entirely curing the polymer mixture. A series of performance tests was conducted including compressive strength test, water immersion test, leach test, thermal stability test, irradiation stability test and biodegradation stability test for the polymer waste forms. From the results of the performance tests for the polymer waste forms, it is believed that the polymer waste form is very stable and can satisfy the acceptance criteria for permanent disposal. At present, performance tests with full scale polymer waste forms are being carried out in order to obtain qualification certificate by the regulatory institute in Korea. Polymer waste forms were prepared with the surrogate of boric acid concentrates and the surrogate of spent ion exchange resins respectively. Waste forms were also made in lab scale and in full scale. Lab. scale waste forms were directly subjected to a series of the performance tests. In the case of full scale waste form, the test specimens for the performance test were taken from a part of waste form by coring. A series of performance tests was conducted including compressive strength test, thermal stability test, irradiation stability test and biodegradation stability test, water immersion test, leach test, and free standing water for the polymer waste forms. In addition, a fire resistance test was performed on the waste forms by the requirement of the regulatory institute in Korea. Every polymer waste forms containing the boric acid concentrates and the spent ion exchange resins had exhibited excellent structural integrity of more than 27.58 MPa (4,000 psi) of compressive strength. On thermal stability testing, biodegradation

  8. Performance Test on Polymer Waste Form - 12137

    International Nuclear Information System (INIS)

    Polymer solidification was attempted to produce stable waste form for the boric acid concentrates and the dewatered spent resins. The polymer mixture was directly injected into the mold or drum which was packed with the boric acid concentrates and the dewatered spent resins, respectively. The waste form was produced by entirely curing the polymer mixture. A series of performance tests was conducted including compressive strength test, water immersion test, leach test, thermal stability test, irradiation stability test and biodegradation stability test for the polymer waste forms. From the results of the performance tests for the polymer waste forms, it is believed that the polymer waste form is very stable and can satisfy the acceptance criteria for permanent disposal. At present, performance tests with full scale polymer waste forms are being carried out in order to obtain qualification certificate by the regulatory institute in Korea. Polymer waste forms were prepared with the surrogate of boric acid concentrates and the surrogate of spent ion exchange resins respectively. Waste forms were also made in lab scale and in full scale. Lab. scale waste forms were directly subjected to a series of the performance tests. In the case of full scale waste form, the test specimens for the performance test were taken from a part of waste form by coring. A series of performance tests was conducted including compressive strength test, thermal stability test, irradiation stability test and biodegradation stability test, water immersion test, leach test, and free standing water for the polymer waste forms. In addition, a fire resistance test was performed on the waste forms by the requirement of the regulatory institute in Korea. Every polymer waste forms containing the boric acid concentrates and the spent ion exchange resins had exhibited excellent structural integrity of more than 27.58 MPa (4,000 psi) of compressive strength. On thermal stability testing, biodegradation

  9. SYNCHROTRON X-RAY MICROTOMOGRAPHY, ELECTRON PROBE MICROANALYSIS, AND NMR OF TOLUENE WASTE IN CEMENT

    International Nuclear Information System (INIS)

    Synchrotron X-ray microtomography shows vesicular structures for toluene/cement mixtures, prepared with 1.22 to 3.58 wt% toluene. Three-dimensional imaging of the cured samples shows spherical vesicles, with diameters ranging from 20 to 250 microm; a search with EPMA for vesicles in the range of 1-20 microm proved negative. However, the total vesicle volume, as computed from the microtomography images, accounts for less than 10% of initial toluene. Since the cements were cured in sealed bottles, the larger portion of toluene must be dispersed within the cement matrix. Evidence for toluene in the cement matrix comes from 29Si MAS NMR spectroscopy, which shows a reduction in chain silicates with added toluene. Also, 2H NMR of d8-toluene/cement samples shows high mobility for all, toluene and thus no toluene/cement binding. A model that accounts for all observations follows: For loadings below about 3 wt%, most toluene is dispersed in the cement matrix, with a small fraction of the initial toluene phase separating from the cement paste and forming vesicular structures that are preserved in the cured cement. Furthermore, at loadings above 3 wt%, the abundance of vesicles formed during toluene/cement paste mixing leads to macroscopic phase separation (most toluene floats to the surface of the cement paste)

  10. Waste Acceptance Testing of Secondary Waste Forms: Cast Stone, Ceramicrete and DuraLith

    Energy Technology Data Exchange (ETDEWEB)

    Mattigod, Shas V.; Westsik, Joseph H.; Chung, Chul-Woo; Lindberg, Michael J.; Parker, Kent E.

    2011-08-12

    To support the selection of a waste form for the liquid secondary wastes from WTP, Washington River Protection Solutions has initiated secondary-waste-form testing work at Pacific Northwest National Laboratory (PNNL). In anticipation of a down-selection process for a waste form for the Solidification Treatment Unit to be added to the ETF, PNNL is conducting tests on four candidate waste forms to evaluate their ability to meet potential waste acceptance criteria for immobilized secondary wastes that would be placed in the IDF. All three waste forms demonstrated compressive strengths above the minimum 3.45 MPa (500 psi) set as a target for cement-based waste forms. Further, none of the waste forms showed any significant degradation in compressive strength after undergoing thermal cycling (30 cycles in a 10 day period) between -40 C and 60 C or water immersion for 90 days. The three leach test methods are intended to measure the diffusion rates of contaminants from the waste forms. Results are reported in terms of diffusion coefficients and a leachability index (LI) calculated based on the diffusion coefficients. A smaller diffusion coefficient and a larger LI are desired. The NRC, in its Waste Form Technical Position (NRC 1991), provides recommendations and guidance regarding methods to demonstrate waste stability for land disposal of radioactive waste. Included is a recommendation to conduct leach tests using the ANS 16.1 method. The resulting leachability index (LI) should be greater than 6.0. For Hanford secondary wastes, the LI > 6.0 criterion applies to sodium leached from the waste form. For technetium and iodine, higher targets of LI > 9 for Tc and LI > 11 for iodine have been set based on early waste-disposal risk and performance assessment analyses. The results of these three leach tests conducted for a total time between 11days (ASTM C1308) to 90 days (ANS 16.1) showed: (1) Technetium diffusivity: ANSI/ANS 16.1, ASTM C1308, and EPA 1315 tests indicated that

  11. Waste Acceptance Testing of Secondary Waste Forms: Cast Stone, Ceramicrete and DuraLith

    International Nuclear Information System (INIS)

    To support the selection of a waste form for the liquid secondary wastes from WTP, Washington River Protection Solutions has initiated secondary-waste-form testing work at Pacific Northwest National Laboratory (PNNL). In anticipation of a down-selection process for a waste form for the Solidification Treatment Unit to be added to the ETF, PNNL is conducting tests on four candidate waste forms to evaluate their ability to meet potential waste acceptance criteria for immobilized secondary wastes that would be placed in the IDF. All three waste forms demonstrated compressive strengths above the minimum 3.45 MPa (500 psi) set as a target for cement-based waste forms. Further, none of the waste forms showed any significant degradation in compressive strength after undergoing thermal cycling (30 cycles in a 10 day period) between -40 C and 60 C or water immersion for 90 days. The three leach test methods are intended to measure the diffusion rates of contaminants from the waste forms. Results are reported in terms of diffusion coefficients and a leachability index (LI) calculated based on the diffusion coefficients. A smaller diffusion coefficient and a larger LI are desired. The NRC, in its Waste Form Technical Position (NRC 1991), provides recommendations and guidance regarding methods to demonstrate waste stability for land disposal of radioactive waste. Included is a recommendation to conduct leach tests using the ANS 16.1 method. The resulting leachability index (LI) should be greater than 6.0. For Hanford secondary wastes, the LI > 6.0 criterion applies to sodium leached from the waste form. For technetium and iodine, higher targets of LI > 9 for Tc and LI > 11 for iodine have been set based on early waste-disposal risk and performance assessment analyses. The results of these three leach tests conducted for a total time between 11days (ASTM C1308) to 90 days (ANS 16.1) showed: (1) Technetium diffusivity: ANSI/ANS 16.1, ASTM C1308, and EPA 1315 tests indicated that

  12. Selection of compositions for the cementation of liquid radioactive waste of Kudankulam NPP and Volgodonskaya NPP

    International Nuclear Information System (INIS)

    The purpose of this work is the selection of formulations for the cementation of liquid radioactive waste of Kudankulam NPP and Volgodonskaya NPP. The simulators of the following radioactive waste have been used for the works: concentrated still bottoms (CSB) with saline content 600-800 g/l, sludge, pulps of ion-exchange resins (IER), activated carbon, titanium and ion-exchange sorbents of Kudankulam NPP and concentrated still bottoms with saline content 900 g/l, pulps of ion-exchange resins, sludge of Volgodonskaya NPP. For Kudankulam NPP there was made a separate research of the cementation of each type of waste and also joint cementation of concentrated still bottoms and ion-exchange resin. For Volgodonskaya NPP - joint cementation of CSB and IER or sludge. The properties of the compounds were determined, which are regulated by GOST R 51883-2002, spread ability and setting time of the cement grouts. The study has shown that as a main component of the combined binding material for the cementation of low level radioactive waste (LRW) of Kudankulam NPP and Volgodonskaya NPP, the usage of Portland cement is preferable. As additives for the binding materials it is better to use lime and bentonite clay powder. Maximal inclusion of LRW into the compound when using these materials will be (% of the compound weight): CSB: 30%, sludge - 14%, IER - 14%, activated carbon - 18%, titanium sorbent - 20%, ion-selective sorbent - 14%

  13. Study on safety evaluation of monolithic cement packages of radioactive wastes under deep-sea conditions

    International Nuclear Information System (INIS)

    For sea disposal of the low-level radioactive wastes, the safety of monolithic cement solidified products to be the main solidified waste for dumping was evaluated. Safety evaluation covers the results of integrity test under deep-sea conditions, development of nondestructive inspection and leaching test of nuclides of the above solidified waste. It is concluded that previous evaluation for the sea disposal of radioactive wastes should be more conservative than the real situation, because the cement solidified products have appreciable retardation effect for dispersion of radionuclides and thus the effect of containment is enhanced. (author)

  14. Stabilization of ZnCl2-containing wastes using calcium sulfoaluminate cement: cement hydration, strength development and volume stability.

    Science.gov (United States)

    Berger, Stéphane; Cau Dit Coumes, Céline; Le Bescop, Patrick; Damidot, Denis

    2011-10-30

    The potential of calcium sulfoaluminate (CSA) cement was investigated to solidify and stabilize wastes containing large amounts of soluble zinc chloride (a strong inhibitor of Portland cement hydration). Hydration of pastes and mortars prepared with a 0.5 mol/L ZnCl(2) mixing solution was characterized over one year as a function of the gypsum content of the binder and the thermal history of the material. Blending the CSA clinker with 20% gypsum enabled its rapid hydration, with only very small delay compared with a reference prepared with pure water. It also improved the compressive strength of the hardened material and significantly reduced its expansion under wet curing. Moreover, the hydrates assemblage was less affected by a thermal treatment at early age simulating the temperature rise and fall occurring in a large-volume drum of cemented waste. Fully hydrated materials contained ettringite, amorphous aluminum hydroxide, strätlingite, together with AFm phases (Kuzel's salt associated with monosulfoaluminate or Friedel's salt depending on the gypsum content of the binder), and possibly C-(A)-S-H. Zinc was readily insolubilized and could not be detected in the pore solution extracted from cement pastes. PMID:21889260

  15. Application of Recycled Concrete Aggregates Containing Waste Glass Powder/Suspension and Bottom Ash as a Cement Component in Concrete

    OpenAIRE

    Kara, P

    2013-01-01

    The growing environmental concerns and the increasing scarcity of landfills encourage the recycling of industrial wastes and adopting environmentally friendly practices by rational usage of natural resources. The production of concrete with recycled aggregate and reduced cement volume is the most desirable form of achieving a closed life cycle as an ecological constructional material. This paper describes results of a study undertaken to examine the influence of recycled aggregates obta...

  16. Obtaining a sulfoaluminate belite cement by industrial waste

    Directory of Open Access Journals (Sweden)

    Elkhadiri, L.

    2003-06-01

    Full Text Available Sulfoaluminate belite clinkers by burning raw at moderate temperatures near 1250 °C were synthesized. The used mixtures were made by calcium carbonate blended to two industrial wastes: low calcium fly ash and phosphogypsum. The clinkers were characterised by X-Ray Diffraction (XRD, Infrared Spectroscopy (FTIR and free lime. The hydraulic behaviour of the obtained cements, by adequate clinkers with 7% of added gypsum, was followed by XRD, scaning electronic microscopy (SEM, FTIR and NMR.

    Los clínkeres belíticos de sulfoaluminatos se obtienen por cocción de crudos a temperaturas moderadas, hacia 1.250 ºC. Esos crudos se componen de carbonato de calcio mezclados con dos subproductos industriales: cenizas volantes pobres en óxido de calcio y fosfoyeso. Los clínkeres obtenidos se caracterizaron a través de Difracción de Rayos X (DRX, Espectroscopia Infrarroja por Transformada de Fourier (FTIR y por la determinación de CaO libre. El comportamiento hidráulico de los cementos elaborados de los clínkeres con el 7% de yeso se estudió por DRX, Microscopía Electrónica de Barrido (SEM, FTIR y Resonancia Magnética Nuclear (RMN

  17. Development of Magnesium Silicate Hydrate cement system for nuclear waste encapsulation

    International Nuclear Information System (INIS)

    A novel low pH cement system for encapsulating nuclear industry wastes containing aluminium has been developed using blends of MgO and silica fume (SF). Identification of the hydrated phases in MgO/silica fume samples showed that brucite formed in early stages of hydration and then reacted with the silica fume to produce a magnesium silicate hydrate (M-S-H) gel phase. When all brucite reacts with silica fume a cement system with an equilibrium pH just below 10 was achieved. Selected mixes have been characterized for hydration reactions, setting time and strength development. Mortar samples with w/s ratios of 0.5 and 50% by weight of sand added achieved compressive strengths in excess of 95 MPa after 28 days. The addition of MgCO3 buffered the early pH and the addition of fine sand particles eliminated shrinkage cracking. The interaction of the optimised mortar with Al metal has been investigated. Al metal strips were firmly bound into the MgO:SF:sand samples and no H2 gas detected, and this indicates that the novel systems developed in this work have potential for encapsulating certain types of problematic legacy wastes from the nuclear industry. (authors)

  18. Radioactive waste forms for the future

    International Nuclear Information System (INIS)

    This volume presents a compilation of important information on the full range of radioactive waste forms that have been developed, or at least suggested, for the incorporation of high-level nuclear waste. Many of the results were published in the 'gray literature' of final reports of national laboratories or in various, generally less available, proceedings volumes. This is the first publication to draw information on nuclear waste forms for high-level wastes togehter into a single volume. A detailed presentation is given on the properties and performance of non-crystalline waste forms (borosilicate glass, sintered glass, sintered glass, and lead-iron phosphate glass), and crystalline waste forms (Synroc, tailored ceramics, TiO2-ceramic matrix, glass-ceramics and concrete). A chapter on Novel waste forms reviews a number of methods that warrant further development because of their potential superior performance and unique applications. The final chapter includes a tabulated comparison of important waste form properties and an extended discussion on the corrosion process and radiation damage effects for each waste form. Of particular interest is a performance assessment of nuclear waste borosilicate glass and the crystalline ceramic Synroc. This is the first detailed attempt to compare these two important waste forms on the basis of their materials properties. The discussion emphasizes the difficulties in making such a comparison and details the types of data that are required. (author). refs.; figs.; tabs

  19. Radioactive waste forms: A review and comparison

    International Nuclear Information System (INIS)

    Borosilicate glass is, at present, the waste form of choice for most countries and for most compositions. The selection of borosilicate glass is based mainly on an anticipated ease of processing (glass frit and the waste are mixed, melted at relatively low temperatures, and poured into canisters), the fact that the technology is well demonstrated for actual (radioactive) waste, and finally the assumption that the glass as an aperiodic solid will easily accommodate wide variation in waste stream compositions which are extremely complex and varied. There are, however, alternative waste forms which may be single or polyphase crystalline ceramics. Principal ceramic nuclear waste forms include: Synroc, tailored ceramics (= supercalcine), TiO2-matrix ceramics, glass ceramics, monazite, synthetic ''basalt'', cementitious materials, and FUETAP concrete. In addition, there are a number of ''novel'' ceramic waste forms which have been developed to only the most preliminary stages (e.g., crichtonite and cesium-Titanates), and there are several multi-barrier strategies which encapsulate one ceramic waste form in another. Finally, in recent years, spent fuel has become an important waste form. Finally, in recent years, spent fuel has become an important waste form. This paper will briefly describe the importance and types of ceramic waste forms that have been developed and review their advantages and disadvantages. (author). 9 refs

  20. Hydration of blended cement pastes containing waste ceramic powder as a function of age

    Science.gov (United States)

    Scheinherrová, Lenka; Trník, Anton; Kulovaná, Tereza; Pavlík, Zbyšek; Rahhal, Viviana; Irassar, Edgardo F.; Černý, Robert

    2016-07-01

    The production of a cement binder generates a high amount of CO2 and has high energy consumption, resulting in a very adverse impact on the environment. Therefore, use of pozzolana active materials in the concrete production leads to a decrease of the consumption of cement binder and costs, especially when some type of industrial waste is used. In this paper, the hydration of blended cement pastes containing waste ceramic powder from the Czech Republic and Portland cement produced in Argentina is studied. A cement binder is partially replaced by 8 and 40 mass% of a ceramic powder. These materials are compared with an ordinary cement paste. All mixtures are prepared with a water/cement ratio of 0.5. Thermal characterization of the hydrated blended pastes is carried out in the time period from 2 to 360 days. Simultaneous DSC/TG analysis is performed in the temperature range from 25 °C to 1000 °C in an argon atmosphere. Using this thermal analysis, we identify the temperature, enthalpy and mass changes related to the liberation of physically bound water, calcium-silicate-hydrates gels dehydration, portlandite, vaterite and calcite decomposition and their changes during the curing time. Based on thermogravimetry results, we found out that the portlandite content slightly decreases with time for all blended cement pastes.

  1. Hanford Waste Vitrification Plant Project Waste Form Qualification Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    Randklev, E.H.

    1993-06-01

    The US Department of Energy has created a waste acceptance process to help guide the overall program for the disposal of high-level nuclear waste in a federal repository. This Waste Form Qualification Program Plan describes the hierarchy of strategies used by the Hanford Waste Vitrification Plant Project to satisfy the waste form qualification obligations of that waste acceptance process. A description of the functional relationship of the participants contributing to completing this objective is provided. The major activities, products, providers, and associated scheduling for implementing the strategies also are presented.

  2. Hanford Waste Vitrification Plant Project Waste Form Qualification Program Plan

    International Nuclear Information System (INIS)

    The US Department of Energy has created a waste acceptance process to help guide the overall program for the disposal of high-level nuclear waste in a federal repository. This Waste Form Qualification Program Plan describes the hierarchy of strategies used by the Hanford Waste Vitrification Plant Project to satisfy the waste form qualification obligations of that waste acceptance process. A description of the functional relationship of the participants contributing to completing this objective is provided. The major activities, products, providers, and associated scheduling for implementing the strategies also are presented

  3. Measurements of Mercury Released from Solidified/Stabilized Waste Forms

    International Nuclear Information System (INIS)

    This report covers work performed during FY 1999-2000 in support of treatment demonstrations conducted for the Mercury Working Group of the U.S. Department of Energy (DOE) Mixed Waste Focus Area. In order to comply with the requirements of the Resource Conservation and Recovery Act, as implemented by the U.S. Environmental Protection Agency (EPA), DOE must use one of these procedures for wastes containing mercury at levels above 260 ppm: a retorting/roasting treatment or an incineration treatment (if the wastes also contain organics). The recovered radioactively contaminated mercury must then be treated by an amalgamation process prior to disposal. The DOE Mixed Waste Focus Area and Mercury Working Group are working with the EPA to determine if some alternative processes could treat these types of waste directly, thereby avoiding for DOE the costly recovery step. They sponsored a demonstration in which commercial vendors applied their technologies for the treatment of two contaminated waste soils from Brookhaven National Laboratory. Each soil was contaminated with ∼4500 ppm mercury; however, one soil had as a major radioelement americium-241, while the other contained mostly europium-152. The project described in this report addressed the need for data on the mercury vapor released by the solidified/stabilized mixed low-level mercury wastes generated during these demonstrations as well as the comparison between the untreated and treated soils. A related work began in FY 1998, with the measurement of the mercury released by amalgamated mercury, and the results were reported in ORNL/TM-13728. Four treatments were performed on these soils. The baseline was obtained by thermal treatment performed by SepraDyne Corp., and three forms of solidification/stabilization were employed: one using sulfur polymer cement (Brookhaven National Laboratory), one using portland cement [Allied Technology Group (ATG)], and a third using proprietary additives (Nuclear Fuel Services)

  4. A generalized definition for waste form durability

    International Nuclear Information System (INIS)

    When evaluating waste form performance, the term ''durability'' often appears in casual discourse, but in the technical literature, the focus is often on waste form ''degradation'' in terms of mass lost per unit area per unit time. Waste form degradation plays a key role in developing models of the long-term performance in a repository environment, but other factors also influence waste form performance. These include waste form geometry; density, porosity, and cracking; the presence of cladding; in-package chemistry feedback; etc. The paper proposes a formal definition of waste form ''durability'' which accounts for these effects. Examples from simple systems as well as from complex models used in the Total System Performance Assessment of Yucca Mountain are provided. The application of ''durability'' in the selection of bounding models is also discussed

  5. Miscellaneous Waste-Form FEPs

    Energy Technology Data Exchange (ETDEWEB)

    A. Schenker

    2000-12-08

    The US DOE must provide a reasonable assurance that the performance objectives for the Yucca Mountain Project (YMP) potential radioactive-waste repository can be achieved for a 10,000-year post-closure period. The guidance that mandates this direction is under the provisions of 10 CFR Part 63 and the US Department of Energy's ''Revised Interim Guidance Pending Issuance of New US Nuclear Regulatory Commission (NRC) Regulations (Revision 01, July 22, 1999), for Yucca Mountain, Nevada'' (Dyer 1999 and herein referred to as DOE's Interim Guidance). This assurance must be demonstrated in the form of a performance assessment that: (1) identifies the features, events, and processes (FEPs) that might affect the performance of the potential geologic repository; (2) examines the effects of such FEPs on the performance of the potential geologic repository; (3) estimates the expected annual dose to a specified receptor group; and (4) provides the technical basis for inclusion or exclusion of specific FEPs.

  6. Proposed research and development plan for mixed low-level waste forms

    International Nuclear Information System (INIS)

    The objective of this report is to recommend a waste form program plan that addresses waste form issues for mixed low-level waste (MLLW). The report compares the suitability of proposed waste forms for immobilizing MLLW in preparation for permanent near-surface disposal and relates them to their impact on the U.S. Department of Energy's mixed waste mission. Waste forms are classified into four categories: high-temperature waste forms, hydraulic cements, encapsulants, and specialty waste forms. Waste forms are evaluated concerning their ability to immobilize MLLW under certain test conditions established by regulatory agencies and research institutions. The tests focused mainly on leach rate and compressive strength. Results indicate that all of the waste forms considered can be tailored to give satisfactory performance immobilizing large fractions of the Department's MLLW inventory. Final waste form selection will ultimately be determined by the interaction of other, often nontechnical factors, such as economics and politics. As a result of this report, three top-level programmatic needs have been identified: (1) a basic set of requirements for waste package performance and disposal; (2) standardized tests for determining waste form performance and suitability for disposal; and (3) engineering experience operating production-scale treatment and disposal systems for MLLW

  7. Proposed research and development plan for mixed low-level waste forms

    Energy Technology Data Exchange (ETDEWEB)

    O`Holleran, T.O.; Feng, X.; Kalb, P. [and others

    1996-12-01

    The objective of this report is to recommend a waste form program plan that addresses waste form issues for mixed low-level waste (MLLW). The report compares the suitability of proposed waste forms for immobilizing MLLW in preparation for permanent near-surface disposal and relates them to their impact on the U.S. Department of Energy`s mixed waste mission. Waste forms are classified into four categories: high-temperature waste forms, hydraulic cements, encapsulants, and specialty waste forms. Waste forms are evaluated concerning their ability to immobilize MLLW under certain test conditions established by regulatory agencies and research institutions. The tests focused mainly on leach rate and compressive strength. Results indicate that all of the waste forms considered can be tailored to give satisfactory performance immobilizing large fractions of the Department`s MLLW inventory. Final waste form selection will ultimately be determined by the interaction of other, often nontechnical factors, such as economics and politics. As a result of this report, three top-level programmatic needs have been identified: (1) a basic set of requirements for waste package performance and disposal; (2) standardized tests for determining waste form performance and suitability for disposal; and (3) engineering experience operating production-scale treatment and disposal systems for MLLW.

  8. Feasibility of disposing waste glyphosate neutralization liquor with cement rotary kiln

    Energy Technology Data Exchange (ETDEWEB)

    Bai, Y.; Bao, Y.B.; Cai, X.L.; Chen, C.H. [College of Materials Science and Engineering, Nanjing Tech University, Nanjing 210009 (China); State Key Laboratory of Materials-Oriented Chemical Engineering, Nanjing Tech University, Nanjing 210009 (China); Ye, X.C., E-mail: yexuchu@njtech.edu.cn [College of Materials Science and Engineering, Nanjing Tech University, Nanjing 210009 (China); State Key Laboratory of Materials-Oriented Chemical Engineering, Nanjing Tech University, Nanjing 210009 (China)

    2014-08-15

    Highlights: • The waste neutralization liquor was injected directly into the kiln system. • No obvious effect on the quality of cement clinker. • The disposing method was a zero-discharge process. • The waste liquor can be used as an alternative fuel to reduce the coal consumption. - Abstract: The waste neutralization liquor generated during the glyphosate production using glycine-dimethylphosphit process is a severe pollution problem due to its high salinity and organic components. The cement rotary kiln was proposed as a zero discharge strategy of disposal. In this work, the waste liquor was calcinated and the mineralogical phases of residue were characterized by scanning electron microscope (SEM) and X-ray diffraction (XRD). The mineralogical phases and the strength of cement clinker were characterized to evaluate the influence to the products. The burnability of cement raw meal added with waste liquor and the calorific value of waste liquor were tested to evaluate the influence to the thermal state of the kiln system. The results showed that after the addition of this liquor, the differences of the main phases and the strength of cement clinker were negligible, the burnability of raw meal was improved; and the calorific value of this liquor was 6140 J/g, which made it could be considered as an alternative fuel during the actual production.

  9. Feasibility of disposing waste glyphosate neutralization liquor with cement rotary kiln

    International Nuclear Information System (INIS)

    Highlights: • The waste neutralization liquor was injected directly into the kiln system. • No obvious effect on the quality of cement clinker. • The disposing method was a zero-discharge process. • The waste liquor can be used as an alternative fuel to reduce the coal consumption. - Abstract: The waste neutralization liquor generated during the glyphosate production using glycine-dimethylphosphit process is a severe pollution problem due to its high salinity and organic components. The cement rotary kiln was proposed as a zero discharge strategy of disposal. In this work, the waste liquor was calcinated and the mineralogical phases of residue were characterized by scanning electron microscope (SEM) and X-ray diffraction (XRD). The mineralogical phases and the strength of cement clinker were characterized to evaluate the influence to the products. The burnability of cement raw meal added with waste liquor and the calorific value of waste liquor were tested to evaluate the influence to the thermal state of the kiln system. The results showed that after the addition of this liquor, the differences of the main phases and the strength of cement clinker were negligible, the burnability of raw meal was improved; and the calorific value of this liquor was 6140 J/g, which made it could be considered as an alternative fuel during the actual production

  10. Liquid secondary waste. Waste form formulation and qualification

    Energy Technology Data Exchange (ETDEWEB)

    Cozzi, A. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Dixon, K. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hill, K. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); King, W. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Nichols, R. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-03-01

    The Hanford Site Effluent Treatment Facility (ETF) currently treats aqueous waste streams generated during Site cleanup activities. When the Hanford Tank Waste Treatment and Immobilization Plant (WTP) begins operations, a liquid secondary waste (LSW) stream from the WTP will need to be treated. The volume of effluent for treatment at the ETF will increase significantly. Washington River Protection Solutions is implementing a Secondary Liquid Waste Immobilization Technology Development Plan to address the technology needs for a waste form and solidification process to treat the increased volume of waste planned for disposal at the Integrated Disposal Facility IDF). Waste form testing to support this plan is composed of work in the near term to demonstrate the waste form will provide data as input to a performance assessment (PA) for Hanford’s IDF.

  11. Liquid secondary waste. Waste form formulation and qualification

    International Nuclear Information System (INIS)

    The Hanford Site Effluent Treatment Facility (ETF) currently treats aqueous waste streams generated during Site cleanup activities. When the Hanford Tank Waste Treatment and Immobilization Plant (WTP) begins operations, a liquid secondary waste (LSW) stream from the WTP will need to be treated. The volume of effluent for treatment at the ETF will increase significantly. Washington River Protection Solutions is implementing a Secondary Liquid Waste Immobilization Technology Development Plan to address the technology needs for a waste form and solidification process to treat the increased volume of waste planned for disposal at the Integrated Disposal Facility IDF). Waste form testing to support this plan is composed of work in the near term to demonstrate the waste form will provide data as input to a performance assessment (PA) for Hanford's IDF.

  12. Technical area status report for low-level mixed waste final waste forms. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Mayberry, J.L.; DeWitt, L.M. [Science Applications International Corp., Idaho Falls, ID (United States); Darnell, R. [EG and G Idaho, Inc., Idaho Falls, ID (United States)] [and others

    1993-08-01

    The Final Waste Forms (FWF) Technical Area Status Report (TASR) Working Group, the Vitrification Working Group (WG), and the Performance Standards Working Group were established as subgroups to the FWF Technical Support Group (TSG). The FWF TASR WG is comprised of technical representatives from most of the major DOE sites, the Nuclear Regulatory Commission (NRC), the EPA Office of Solid Waste, and the EPA`s Risk Reduction Engineering Laboratory (RREL). The primary activity of the FWF TASR Working Group was to investigate and report on the current status of FWFs for LLNM in this TASR. The FWF TASR Working Group determined the current status of the development of various waste forms described above by reviewing selected articles and technical reports, summarizing data, and establishing an initial set of FWF characteristics to be used in evaluating candidate FWFS; these characteristics are summarized in Section 2. After an initial review of available information, the FWF TASR Working Group chose to study the following groups of final waste forms: hydraulic cement, sulfur polymer cement, glass, ceramic, and organic binders. The organic binders included polyethylene, bitumen, vinyl ester styrene, epoxy, and urea formaldehyde. Section 3 provides a description of each final waste form. Based on the literature review, the gaps and deficiencies in information were summarized, and conclusions and recommendations were established. The information and data presented in this TASR are intended to assist the FWF Production and Assessment TSG in evaluating the Technical Task Plans (TTPs) submitted to DOE EM-50, and thus provide DOE with the necessary information for their FWF decision-making process. This FWF TASR will also assist the DOE and the MWIP in establishing the most acceptable final waste forms for the various LLMW streams stored at DOE facilities.

  13. Technical area status report for low-level mixed waste final waste forms

    International Nuclear Information System (INIS)

    The Final Waste Forms (FWF) Technical Area Status Report (TASR) Working Group, the Vitrification Working Group (WG), and the Performance Standards Working Group were established as subgroups to the FWF Technical Support Group (TSG). The FWF TASR WG is comprised of technical representatives from most of the major DOE sites, the Nuclear Regulatory Commission (NRC), the EPA Office of Solid Waste, and the EPA's Risk Reduction Engineering Laboratory (RREL). The primary activity of the FWF TASR Working Group was to investigate and report on the current status of FWFs for LLNM in this TASR. The FWF TASR Working Group determined the current status of the development of various waste forms described above by reviewing selected articles and technical reports, summarizing data, and establishing an initial set of FWF characteristics to be used in evaluating candidate FWFS; these characteristics are summarized in Section 2. After an initial review of available information, the FWF TASR Working Group chose to study the following groups of final waste forms: hydraulic cement, sulfur polymer cement, glass, ceramic, and organic binders. The organic binders included polyethylene, bitumen, vinyl ester styrene, epoxy, and urea formaldehyde. Section 3 provides a description of each final waste form. Based on the literature review, the gaps and deficiencies in information were summarized, and conclusions and recommendations were established. The information and data presented in this TASR are intended to assist the FWF Production and Assessment TSG in evaluating the Technical Task Plans (TTPs) submitted to DOE EM-50, and thus provide DOE with the necessary information for their FWF decision-making process. This FWF TASR will also assist the DOE and the MWIP in establishing the most acceptable final waste forms for the various LLMW streams stored at DOE facilities

  14. Model Analysis of Initial Hydration and Structure Forming of Portland Cement

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

    The auto efficiently hydration heat arrangement and the non-contacting electrical resistivity device were used to test the thermology effect and the resistivity variation of Portland cement hydration.The structure forming model of Portland cement initial hydration was established through the systematical experiments with different cements, the amount of mixing water and the chemical admixture. The experimental results show that, the structure forming model of cement could be divided into three stages, i e, solution-solution equilibrium period, structure forming period and structure stabilizing period. Along with the increase of mixing water, the time of inflexion appeared is in advance for thermal process of cement hydration and worsened for the structure forming process. Comparison with the control specimen, adding Na2SO4 makes the minimum critical point lower, the flattening period shorter and the growing slope after stage one steeper. So the hydration and structure forming process of Portland cement could be described more exactly by applying the thermal model and the structure-forming model.

  15. The use of mexican cements in the low and medium radioactive wastes confinement

    International Nuclear Information System (INIS)

    Inside the relative mark to the radioactive waste confinement, minerals of great fixation capacity like clays, apatites and diverse oxides are studied as matrixes, components and/or additives of the active barriers that separate the barrier geologic and the nuclear wastes. In this case, the cements intervene in those different stages of the waste management, since its are used for the immobilization of radioactive waste in the container, for the production of containers as well as filler of the spaces among the containers of the vaults, and also as engineering barrier and construction material in the civil work. For the above mentioned, it is particularly useful to characterize the Portland cements with at least 97% of clinker, since they are most recommended for this type of applications. Presently investigation work is carried out a preliminary chemical characterization, based on the mineralogical composition, of the Portland Mexican cement. Results are shown by the X-ray Diffraction technique when immobilizing a rich solution in sulfates to 5%, using two Portland commercial cements APASCO and TOLTECA, without observing the significant appearance of new phases. The cements besides incorporating the chemical species in the breast of the matrix, are also present as barriers of civil engineering in the facilities located only some meters deep for the storage of radioactive waste of low and intermediate level, for that the study of the radionuclides fixation, in the cements is of supreme importance to evaluate the safety of a nuclear repository with the help of cements; the retention of the iodine-131 in a limited interval of pH in the commercial APASCO and TOLTECA it was studied, being observed a scarce retention of this homologous of fission products, what indicates the necessity to use additives to improve the retention properties of the Mexican commercial cements for some radionuclides. (Author)

  16. Studies on cement matrix used at the radioactive waste treatment plant for radwaste conditioning

    International Nuclear Information System (INIS)

    Full text: The research activities performed by the Department of Radioactive Waste Management is focused on the LLAW treatment products obtained by chemical precipitation and on the conditioning of these products by cementation. The individual mechanisms involved in the chemical precipitation process are directly dependent on the precipitate properties and structure, which are connected with the initial system composition and the precipitation procedure, i.e. reagent concentration, rate and orders of chemical addition, mixing rate and time and ageing conditions. In the case of conditioning by cementation, the chemical nature and proportion of the sludges or concentrates affect both the hydrolysis of the initial cement components and the reactions of metastable hydration constituents, as well as the mechanical strength and chemical resistance of the hardened cemented matrix. Generally, the study of the precipitation products and their behaviour during cementation and the long-term disposal is extremely difficult because of the system complexity (phase composition and structure) and the lack of non-destructive analytical methods. The experience accumulated by the countries who developed nuclear programs in military and socioeconomic fields and which produced important volumes of radioactive wastes, lead us to study some of mineral additives to be used in the conditioning and disposal technology. It is well known that mineral additives are diminishing the leaching rate of the radionuclides in the disposal environment. The studies have the purpose to obtain the most propitious mixture of cement-bentonite and cement-volcanic tuff which have the mechanical properties similar to the cement paste used for the conditioning of radioactive waste. Taking into consideration the characteristics of these mineral binders, namely a very good plasticity and capacity of adsorption, which lead to the decrease of porosity, in the future, the mixture is planned to be used at the

  17. Studies on cement matrix materials used at the Radioactive Waste Treatment Plant for radwaste conditioning

    International Nuclear Information System (INIS)

    The research activities performed by Department of Radioactive Waste Management is focused on the treatment of LLAW products obtained by chemical precipitation and on the conditioning of these products by cementation. The individual mechanisms implied in the chemical precipitation processes are directly dependent on the precipitate properties and structure, which in turn are connected with the initial system composition and the precipitation procedure, i.e. reagent concentration, rate and orders of chemical addition, mixing rate and time and ageing conditions. In case of conditioning by cementation, the chemical nature and proportion of the sludges or concentrates affect both the hydrolysis of the initial cement components and the reactions of metastable hydration constituents, as well as the mechanical strength and chemical resistance of the hardened cemented matrix.Generally, the study of the precipitation products and their behaviour during cementation and the long-term disposal is extremely difficult because of the system complexity (phase composition and structure) and the lack of the non-destructive analytical methods. The experience accumulated by the countries who advanced nuclear programmes in military and socio-economic fields and which produced important volumes of radioactive wastes, leads us to study some of mineral additives to be used in the conditioning and disposal technology. Is well known that some mineral additives can diminish the leaching rate of the radionuclides in the disposal environment.The studies have the purpose to obtain the most propitious mixture of cement-bentonite and cement-volcanic tuff, which have the mechanical properties similar to the cement paste used for the conditioning of radioactive waste.Taking into account the characteristics of these mineral binders, namely a very good plasticity and capacity of adsorption, which lead at the decrease of porosity, the mixture is planned to be used in the future, at the Radioactive

  18. Combined Waste Form Cost Trade Study

    Energy Technology Data Exchange (ETDEWEB)

    Dirk Gombert; Steve Piet; Timothy Trickel; Joe Carter; John Vienna; Bill Ebert; Gretchen Matthern

    2008-11-01

    A new generation of aqueous nuclear fuel reprocessing, now in development under the auspices of the DOE Office of Nuclear Energy (NE), separates fuel into several fractions, thereby partitioning the wastes into groups of common chemistry. This technology advance enables development of waste management strategies that were not conceivable with simple PUREX reprocessing. Conventional wisdom suggests minimizing high level waste (HLW) volume is desirable, but logical extrapolation of this concept suggests that at some point the cost of reducing volume further will reach a point of diminishing return and may cease to be cost-effective. This report summarizes an evaluation considering three groupings of wastes in terms of cost-benefit for the reprocessing system. Internationally, the typical waste form for HLW from the PUREX process is borosilicate glass containing waste elements as oxides. Unfortunately several fission products (primarily Mo and the noble metals Ru, Rh, Pd) have limited solubility in glass, yielding relatively low waste loading, producing more glass, and greater disposal costs. Advanced separations allow matching the waste form to waste stream chemistry, allowing the disposal system to achieve more optimum waste loading with improved performance. Metals can be segregated from oxides and each can be stabilized in forms to minimize the HLW volume for repository disposal. Thus, a more efficient waste management system making the most effective use of advanced waste forms and disposal design for each waste is enabled by advanced separations and how the waste streams are combined. This trade-study was designed to juxtapose a combined waste form baseline waste treatment scheme with two options and to evaluate the cost-benefit using available data from the conceptual design studies supported by DOE-NE.

  19. Characterization of different types of ceramic waste and its incorporation to the cement paste

    International Nuclear Information System (INIS)

    The porcelain tike is a product resulting from the technological development of ceramic plating industry. Its large acceptation by the consumer market is probably linked with certain properties, such as low porosity, high mechanical resistance, facility in maintenance, besides being a material of modern and versatile characteristics. The aim of this work was characterizing the different ceramic wastes (enameled and porcelain tike) and evaluating its influence on the mechanical behavior in cement pastes. The wastes were characterized through the determination of its chemical composition, size particle distribution and X-ray diffraction. Cement pastes + wastes were prepared in 25% and 50% proportions and glue time determination, water absorption and resistance to compression assays were taken. The results indicate that although the wastes don't show any variation in the elementary chemical composition, changes in the cement paste behavior related to the values of resistance to compression were observed. (author)

  20. Calcium aluminate cements for nuclear wastes conditioning: literature review and new approaches

    International Nuclear Information System (INIS)

    Encapsulate the diverse wastes produced by nuclear activities in cementitious binders may be very complex due to the adverse cement-waste interactions. Consequences are for example: strong delay, poor mechanical strength or low resistance to leaching. In this case, pure or blended calcium aluminate cements (CACs) may be valuable alternatives. This paper summarises the properties of CAC and blended CAC system and gives some examples from literature where calcium aluminate cements are used for conventional wastes or nuclear wastes conditioning. Moreover, it proposes another approach: using CAC not only as a binder, but also as a chemical reactant. After dissolution calcium aluminates ions can combine with many chemical species (sulphates, nitrates, chlorides, alkali metals, heavy metals) to precipitate specific hydrates allowing chemical trapping of these species. An example is given for the purification of Ni and Zn nitrates solutions. (authors)

  1. Synthesis of knowledge on the long-term behaviour of concretes. Applications to cemented waste packages

    International Nuclear Information System (INIS)

    As stipulated in the former law of December 91 relating to 'concrete waste package', a progress report (phenomenological reference document) was first provided in 1999. The objective was to make an assessment of the knowledge acquired on the long-term behaviour of cement-based waste packages in the context of deep disposal and/or interim storage. The present document is an updated summary report. It takes into account a new knowledge assessment, considers coupled mechanisms and should contribute to the first performance studies (operational calculations). Handling and radio-nuclides (RN) confinement are the two major functional properties requested from the concrete used for the waste packages. In unsaturated environment (interim storage/disposal prior to closing), the main problem is the generation of cracks in the material. This aspect is a key parameter from the mechanical point of view (retrievability). It can have a major impact on the disposal phase (confinement). In saturated environment (disposal post-closing phase), the main concern is the chemical degradation of the waste package concrete submitted to underground waters leaching. In this context, the major thema are: the durability of the concretes under water (chemical degradation) and in unsaturated medium (corrosion of reinforcement), matter transport, RN retention, chemistry / transport / mechanical couplings. On the other hand, laboratory data on the behaviour of concretes are used to evaluate the RN source term of waste packages in function of time (concrete waste package OPerational Model, i.e. 'Concrete MOP'). The 'MOP' provides the physico-chemical description of the RN release in relationship with the waste package degradation itself. This description is based on simplified phenomenology for which only dimensioning mechanisms are taken into account. The use of Diffu-Ca code (basic module for the MOP) on the CASTEM numerical plate-form, already allows operational predictions. (authors)

  2. Wet oxidative degradation of cellulosic wastes 5- chemical and thermal properties of the final waste forms

    International Nuclear Information System (INIS)

    In this study, the residual solution arising from the wet oxidative degradation of solid organic cellulosic materials, as one of the component of radioactive solid wastes, using hydrogen peroxide as oxidant. Were incorporated into ordinary Portland cement matrix. Leaching as well as thermal characterizations of the final solidified waste forms were evaluated to meet the final disposal requirements. Factors, such as the amount of the residual solution incorporated, types of leachant. Release of different radionuclides and freezing-thaw treatment, that may affect the leaching characterization. Were studied systematically from the data obtained, it was found that the final solid waste from containing 35% residual solution in tap water is higher than that in ground water or sea water. Based on the data obtained from thermal analysis, it could be concluded that incorporating the residual solution form the wet oxidative degradation of cellulosic materials has no negative effect on the hydration of cement materials and consequently on the thermal stability of the final solid waste from during the disposal process

  3. Mechanisms and modelling of waste/cement interactions - Survey of topics presented at the Meiringen Workshop

    International Nuclear Information System (INIS)

    Cementitious matrices are being used worldwide as a containment medium for radioactive and non-radioactive waste in order to retard the mobility of contaminants. The present thrust of research is to further the understanding of contaminant binding in the cementitious matrix in order to predict the long-term behaviour and the potential impact of the waste on the environment. The workshop 'Mechanisms and Modelling of Waste/Cement Interactions', held in Meiringen, Switzerland, between May 8 and 12, 2005, focused on the chemical understanding and thermodynamic modelling of the processes responsible for the retention of radioactive and non-radioactive species in cementitious systems. The objectives of the workshop were to bring together scientists from different disciplines, i.e. cement chemistry, radioactive and non-radioactive hazardous waste disposal, to stimulate discussions on current developments and to identify future needs in this field of research. The topics treated in the workshop were chosen to maximize the benefit to the different fields of research. Cement chemists reported on developments in the understanding of cement mineralogy and thermodynamic modelling of cement systems. The hazardous and radioactive waste management communities presented their ideas on the mechanisms of contaminant binding to cement minerals as well as field, laboratory and modelling results from practical applications. In this paper important areas of research on waste/cement interactions presented in the workshop will be outlined and briefly discussed. The following overview reflects a subjective perception of the workshop and does not lay claim to deal comprehensively with all the papers that were presented in the workshop. (author)

  4. Cement solidification processing method for incombustible miscellaneous solid waste and device therefor

    International Nuclear Information System (INIS)

    When wastes mixed with light metals are solidified, H2 is generated at an alkalinity higher than a predetermined value in a case where a solidifying material is alkaline. Then, in the present invention, less alkaline cement is used as solidifying cement and, only when pH of the mixed paste is greater than 13.3, a pH reducing material is charged to lower the pH value of the mixed paste to less than 13.3, and the wastes are solidified by using the same. As the pH reducing agent, a powdery material, for example, white cement, fly ash and blast furnace slag may be used. This can avoid generation of H2 during a cement solidification processing. (T.M.)

  5. The Effect Of Felspar On the Results Of Low Level Radioactive Waste Treatment By Cementation

    International Nuclear Information System (INIS)

    The aim of this study was to determine an optimum composition of the mixture for waste treatment by cementation in the environmental condition. Mixture of cement and water by weight ratio of water/cement (w/c) = 0.35 was added by waste from various concentration of 5; 7.5; and 12.5 % weight from cement-water mixture. The mixture of water /cement was filled into polyethylene tube cover which has diameter = height = 3.7. Firstly none radioactive waste contained 300 g/l NaNO3 was added. The same work was done again for the edition of NaNO3 waste. After having cured for 28 days the mixture was the tube. Further examination was done by Paul Wreber compressive strength, irradiation at dose rate of 2250.900 rad/hr from 60Co sources, thermal test by Sykron furnace, and leaching rate test of active nuclide by Ortec analyzer, and non active nuclide was analyzed in BTKL Yogyakarta. In leach rate test 90Sr with a final activity of 1.10-4μCi/ml was added. The compressive strength (CS) of solid water-cement = 0.35 was 40/07 N/mm2. Solidification waste cement was 38.05 N/mm2 at pH 6, and was 39,10 N/mm at pH 10, because NaOH supported to inhibit cracking. After felspar mineral has been added CS was 39,95 N/mm2, because the function of additive was to close all pores and to inhibit cracking and bleeding . Thermal tested at 600oC during 1 hour indicated that CS was 24.65 N/mm2 and the beginning of cracking occurred at 400oC, and at 900oC . The CS was 44,95 N/mm2 Irradiation effect on CS of solid waste was not significant . Leaching test indicated that active nitrate waste cured into water at last three months were between 2.12x10-5 until 9.54x10-5 g.cm2 day1. Composition of the optimum solid content 1:65.625 wt % cement 22.968 wt % low level active nitrate waste; and 4,763 wt % felspar additive. Key words : cementation, low level radioactive waste,felspar

  6. Sulfur polymer cement as a low-level waste glass matrix encapsulant

    Energy Technology Data Exchange (ETDEWEB)

    Sliva, P.; Peng, Y.B.; Peeler, D.K. [and others

    1996-01-01

    Sulfur polymer cement (SPC) is being considered as a matrix encapsulant for the Hanford low-level (activity) waste glass. SPC is an elemental sulfur polymer-stabilized thermoplastic that is fluid at 120 {degrees}C to 140{degrees}C. The candidate process would encapsulate the waste glass by mixing the glass cullet with the SPC and casting it into the container. As the primary barrier to groundwater and a key factor in controlling the local environment of the disposal system after it has been compromised, SPC plays a key role in the waste form`s long-term performance assessment. Work in fiscal year 1995 targeted several technical areas of matrix encapsulation involving SPC. A literature review was performed to evaluate potential matrix-encapsulant materials. The dissolution and corrosion behavior of SPC under static conditions was determined as a function of temperature, pH, and sample surface area/solution volume. Preliminary dynamic flow-through testing was performed. SPC formulation and properties were investigated, including controlled crystallization, phase formation, modifying polymer effects on crystallization, and SPC processibility. The interface between SPC and simulated LLW glass was examined. Interfacial chemistry and stability, the effect of water on the glass/SPC interface, and the effect of molten sulfur on the glass surface chemistry were established. Preliminary scoping experiments, involving SPC`s Tc gettering capabilities were performed. Compressive strengths of SPC and SPC/glass composites, both before and after lifetime radiation dose exposure, were determined.

  7. Properties of low-ph cement grout as a sealing material for the geological disposal of radioactive waste

    International Nuclear Information System (INIS)

    The current solution to the problem of using cementitious material for sealing purposes in a final radioactive waste repository is to develop a low-pH cement grout. In this study, the material properties of a low-pH cement grout based on a recipe used at ONKALO are investigated by considering such factors as pH variation, compressive strength, dynamic modulus, and hydraulic conductivity by using silica fume and micro-cement. From the pH measurements of the hardened cement grout, the required pH (< pH 11) is obtained after 130 days of curing. Although the engineering properties of the low-pH cement grout used in this study are inferior to those of conventional high-pH cement grout, the utilization of silica fume and micro-cement effectively meets the long-term environmental and durability requirements for cement grout in a radioactive waste repository

  8. Immobilization of radioactive waste in cement-based matrices

    International Nuclear Information System (INIS)

    A solubility model of the system CaO-SiO2-H2O is developed which takes account of the state of Si polymerization in the solid. Free energies of formations of its bonding hydrogel are tabulated. The internal redox conditions in cements have been measured; in particular, slags lower the Esub(eta) relative to OPC. The fate of Sr and U in cement systems has been determined; Sr is incorporated in the aluminate phases, while U6+ is precipitated as Ca-U-O-H2O phases. Lowering the internal Esub(eta) reduces U solubility. Studies of the carbonation of slag-cement blends are reported. (author)

  9. Release of radionuclides and chelating agents from cement-solidified decontamination low-level radioactive waste collected from the Peach Bottom Atomic Power Station Unit 3

    International Nuclear Information System (INIS)

    As part of a study being performed for the Nuclear Regulatory Commission (NRC), small-scale waste-form specimens were collected during a low oxidation-state transition-metal ion (LOMI)-nitric permanganate (NP)-LOMI solidification performed in October 1989 at the Peach Bottom Atomic Power Station Unit 3. The purpose of this program was to evaluate the performance of cement-solidified decontamination waste to meet the low-level waste stability requirements defined in the NRC's ''Technical Position on Waste Form,'' Revision 1. The samples were acquired and tested because little data have been obtained on the physical stability of actual cement-solidified decontamination ion-exchange resin waste forms and on the leachability of radionuclides and chelating agents from those waste forms. The Peach Bottom waste-form specimens were subjected to compressive strength, immersion, and leach testing in accordance with the NRC's ''Technical Position on Waste Form,'' Revision 1. Results of this study indicate that the specimens withstood the compression tests (>500 psi) before and after immersion testing and leaching, and that the leachability indexes for all radionuclides, including 14C, 99 Tc, and 129I, are well above the leachability index requirement of 6.0, required by the NRC's ''Technical Position on Waste Form,'' Revision 1

  10. Study of chemical additives in the cementation of radioactive waste of PWR reactors

    International Nuclear Information System (INIS)

    In this research it has been studied the effects of chemical admixtures in the cementation process of radioactive wastes. These additives are used to improve the properties of waste cementation process, both of the paste and of the solidified product. However there are a large variety of these materials that are frequently changed or taken out of the market. Then it is essential to know the commercially available materials and their effects. The tests were carried out with a solution simulating the evaporator concentrate waste coming from PWR nuclear reactors. It was cemented using two formulations, A and B, incorporating higher or lower amount of waste, respectively. It was added chemical admixtures from two manufacturers (S and H), which were: accelerators, set retarders and superplasticizers. The experiments were organized by a factorial design 23. The measured parameters were: the viscosity, the setting time, the paste and product density and the compressive strength. The parameter evaluated in this study was the compressive strength at age of 28 days, is considered essential security issues relating to the handling, transport and storage of cemented waste product. The results showed that the addition of accelerators improved the compressive strength of the cemented products. (author)

  11. The incorporation of low and medium level radioactive wastes (solids and liquids) in cement

    International Nuclear Information System (INIS)

    Laboratory studies and mixing plant trials on simulated radioactive waste formulations are reported. Long term stability testing of various formulations including those containing blast furnace slag-ordinary Portland cement, sodium nitrate, ion exchange resins, and sodium nitrate-tributyl phosphate, are reported and some results are given. Mixing plant trials with a high shear cement mixer are reported. An outline of future work is presented. (U.K.)

  12. The use of cement grouts for the immobilisation of solid radioactive waste

    International Nuclear Information System (INIS)

    The use of cement grouts is being considered for the immobilisation of solid items of radioactive waste. In this report the factors which influence the selection of a grout for use in an active plant are identified. The properties and limitations of standard cement grouts are summarised. Inactive grouting trials carried out in the period September 1981 to June 1982 on the 220 dm3 scale are described. (author)

  13. Chemical and mechanical performance properties for various final waste forms -- PSPI scoping study

    Energy Technology Data Exchange (ETDEWEB)

    Farnsworth, R.K.; Larsen, E.D.; Sears, J.W.; Eddy, T.L.; Anderson, G.L.

    1996-09-01

    The US DOE is obtaining data on the performance properties of the various final waste forms that may be chosen as primary treatment products for the alpha-contaminated low-level and transuranic waste at the INEL`s Transuranic Storage Area. This report collects and compares selected properties that are key indicators of mechanical and chemical durability for Portland cement concrete, concrete formed under elevated temperature and pressure, sulfur polymer cement, borosilicate glass, and various forms of alumino-silicate glass, including in situ vitrification glass and various compositions of iron-enriched basalt (IEB) and iron-enriched basalt IV (IEB4). Compressive strength and impact resistance properties were used as performance indicators in comparative evaluation of the mechanical durability of each waste form, while various leachability data were used in comparative evaluation of each waste form`s chemical durability. The vitrified waste forms were generally more durable than the non-vitrified waste forms, with the iron-enriched alumino-silicate glasses and glass/ceramics exhibiting the most favorable chemical and mechanical durabilities. It appears that the addition of zirconia and titania to IEB (forming IEB4) increases the leach resistance of the lanthanides. The large compositional ranges for IEB and IEB4 more easily accommodate the compositions of the waste stored at the INEL than does the composition of borosilicate glass. It appears, however, that the large potential variation in IEB and IEB4 compositions resulting from differing waste feed compositions can impact waste form durability. Further work is needed to determine the range of waste stream feed compositions and rates of waste form cooling that will result in acceptable and optimized IEB or IEB4 waste form performance. 43 refs.

  14. Special waste-form lysimeters - arid: 1984--1992 data summary and preliminary interpretation

    International Nuclear Information System (INIS)

    A lysimeter facility constructed at the Hanford Site in south-central Washington State has been used since 1984 to monitor the leaching of buried waste forms under natural conditions. The facility is generating data that are useful in evaluating source-term models used in radioactive waste transport analyses. The facility includes ten bare-soil lysimeters (183 cm diameter by 305 cm depth) containing buried waste forms generated at nuclear reactors in the United States and solidified with Portland M cement, masonry cement, bitumen, and vinyl-ester styrene. The waste forms contained in the lysimeters have been leached under natural, semiarid conditions. In spite of the semiarid conditions, from 1984 through 1992, an average of 45 cm of water leached through the lysimeters, representing 27% of area precipitation. Leachate samples have been routinely collected and analyzed for radionuclide and chemical content. To date, tritium, cobalt-60, and cesium-137 have been identified in the lysimeter leachate samples. From 1984 through 1992, over 4000 μCi of tritium, representing 76 and 71 % of inventory (not decay corrected), have been leached from the two waste forms containing tritium. Cobalt-60 has been found in the leachate from all six of the waste forms that originally contained > 1 mCi of inventory. The leached amounts of cobalt-60 represent < 0.1 % of original cobalt inventories. Mobile cobalt is believed to be chelated with organic compounds, such as ethylenediaminetetraacetic acid (EDTA), that are present in the waste. Trace amounts of cesium-137 have occasionally been identified in leachate from two waste forms since 1991. Qualitatively, the field leaching results confirm laboratory studies suggesting that tritium is readily leached from cement, and that cobalt-60 is generally leached more easily from cement than from vinyl-ester styrene

  15. DSC and TG Analysis of a Blended Binder Based on Waste Ceramic Powder and Portland Cement

    Science.gov (United States)

    Pavlík, Zbyšek; Trník, Anton; Kulovaná, Tereza; Scheinherrová, Lenka; Rahhal, Viviana; Irassar, Edgardo; Černý, Robert

    2016-03-01

    Cement industry belongs to the business sectors characteristic by high energy consumption and high {CO}2 generation. Therefore, any replacement of cement in concrete by waste materials can lead to immediate environmental benefits. In this paper, a possible use of waste ceramic powder in blended binders is studied. At first, the chemical composition of Portland cement and ceramic powder is analyzed using the X-ray fluorescence method. Then, thermal and mechanical characterization of hydrated blended binders containing up to 24 % ceramic is carried out within the time period of 2 days to 28 days. The differential scanning calorimetry and thermogravimetry measurements are performed in the temperature range of 25°C to 1000°C in an argon atmosphere. The measurement of compressive strength is done according to the European standards for cement mortars. The thermal analysis results in the identification of temperature and quantification of enthalpy and mass changes related to the liberation of physically bound water, calcium-silicate-hydrates dehydration and portlandite, vaterite and calcite decomposition. The portlandite content is found to decrease with time for all blends which provides the evidence of the pozzolanic activity of ceramic powder even within the limited monitoring time of 28 days. Taking into account the favorable results obtained in the measurement of compressive strength, it can be concluded that the applied waste ceramic powder can be successfully used as a supplementary cementing material to Portland cement in an amount of up to 24 mass%.

  16. Murine osteoblastic and osteoclastic differentiation on strontium releasing hydroxyapatite forming cements.

    Science.gov (United States)

    Singh, Satish S; Roy, Abhijit; Lee, Boeun; Parekh, Shrey; Kumta, Prashant N

    2016-06-01

    Ionic substitutions in hydroxyapatite (HA) scaffolds and self-setting cements containing Sr(2+) ions incorporated are particularly of interest in bone regeneration. To date, the approach widely used to incorporate Sr(2+) ions into HA cements has been the addition of Sr(2+) containing salts, such as SrCO3, SrCl2∙6H2O, or SrHPO4. However, this approach is dependent upon the relative solubility of Sr(2+) containing salts with respect to calcium phosphate (CaP) precursors. Therefore, in the current study Sr(2+) substituted dicalcium phosphate dihydrate (DCPD) was first synthesized and directly reacted with tetracalcium phosphate (TTCP) to form Sr(2+) substituted HA forming cements. Rietveld refinement indicated that after one week of aging in phosphate buffered saline, cements prepared with and without Sr(2+) were composed of 75% HA and 25% unreacted TTCP by weight. Cements prepared with 10% Sr(2+) DCPD exhibited increased compressive strengths in comparison to unsubstituted cements. Increased MC3T3-E1 proliferation and differentiation were also observed on the cements prepared with increasing Sr(2+) content. It was concluded that both the scaffold microstructure and Sr(2+) ion release supported osteogenic differentiation. With respect to osteoclastic differentiation, no statistically significant differences in TRAP activity or cell morphology were observed. This suggests that the amount of Sr(2+) released may have been too low to influence osteoclast formation in comparison to unsubstituted cements. The results obtained herein demonstrate that the use of Sr(2+) substituted DCPD precursors rather than individually separate Sr(2+) containing salts may be a useful approach to prepare Sr(2+) containing HA cements. PMID:27040237

  17. Immobilization of radioactive waste in cement based matrices

    International Nuclear Information System (INIS)

    The modelling of cement behaviour at longer ages is reported. Factors studied include composition, pH and Esub(h). The stresses arising from irradiation are evaluated. The behaviour of two elements in cement - U and I has been studied; new experimental data are reported including solubility measurements. Some additional data are given on Sr. Results of desk studies relevant to lifetime predictions are presented. (author)

  18. Effects of aqueous environment on long-term durability of phosphate-bonded ceramic waste forms

    International Nuclear Information System (INIS)

    Over the last few years, Argonne National Laboratory has been developing room-temperature-setting chemically-bonded phosphate ceramics for solidifying and stabilizing low-level mixed wastes. This technology is crucial for stabilizing waste streams that contain volatile species and off-gas secondary waste streams generated by high-temperature treatment of such wastes. Magnesium phosphate ceramic has been developed to treat mixed wastes such as ash, salts, and cement sludges. Waste forms of surrogate waste streams were fabricated by acid-base reactions between the mixtures of magnesium oxide powders and the wastes, and phosphoric acid or acid phosphate solutions. Dense and hard ceramic waste forms are produced in this process. The principal advantage of this technology is that the contaminants are immobilized by both chemical stabilization and subsequent microencapsulation of the reaction products. This paper reports the results of durability studies conducted on waste forms made with ash waste streams spiked with hazardous and radioactive surrogates. Standard leaching tests such as ANS 16.1 and TCLP were conducted on the final waste forms. Fates of the contaminants in the final waste forms were established by electron microscopy. In addition, stability of the waste forms in aqueous environments was evaluated with long-term water-immersion tests

  19. Immobilization of radioactive and hazardous wastes in a newly developed sulfur polymer cement (Spc) matrix

    International Nuclear Information System (INIS)

    Low and Intermediate level radioactive wastes (LILW) and hazardous wastes, presents a waste disposal problem. In this respect, a process to immobilize different radioactive and hazardous wastes, including metals contaminated with radionuclides in a form that is non-dispersible and meet the Environmental Protection Agency (USA, EPA) leaching criteria is a must. In this stabilization and solidification process (S/S), simulated radioactive wastes of Cs, Sr, Ce, Cr, and Pb were immobilized in the stable form of sulfur polymer cement (SPC). In the present work, the mixture of the contaminant(s) and the sulfur mixture which is composed from 95% S and 5% aromatic/or aliphatic hydrocarbons used as polymerizing agents for sulfur (by weight), were added in a stainless steel vessel and primarily heated to 40 degree C for four hours, this time was sufficient for homogeneous mixing of the metals with sulfur and Na2S (to convert the metals to their corresponding sulfides). Additional SPC was then added and the temperature of the mixture was raised to 135 ±1 degree C, resulting in a molten form that was poured into a stainless steel mold where it cooled and solidified. Durability of the fabricated SPC matrices was assessed in terms of water of immersion, porosity, and compressive strength. The water absorption and open porosity were very low and didn't exceed 2.5 % for all matrices, whereas the compressive strength ranged between 7 and 14 KN/m2 depending on the matrix composition. The immobilized waste forms of SPC were characterized by X-ray diffraction (XRD) technique that showed that the different contaminants were stabilized during the solidification process to form stable sulfides. Leachability of the waste matrices was assessed by the Toxicity Characteristic Leaching Procedure (TCLP) of the EPA, optimized and compared with the new EPA Universal Treatment Standards.The TCLP results showed that the concentration of the most contaminants released were under detection limit of

  20. Immobilisation of MTR waste in cement (product evaluation). Annual report March 1985

    International Nuclear Information System (INIS)

    This report describes work performed at Winfrith under the UKAEA's research and development programme on radioactive waste management. The work carried out during April 1984 to March 1985 on the evaluation of laboratory and 200 dm3 scale products of cemented MTR waste was sponsored by the Department of the Environment as part of radioactive waste management research programme. The results will be used in the formulation of Government policy but at this stage they do not necessarily represent Government policy. (author)

  1. DWPF waste form compliance plan (Draft Revision)

    International Nuclear Information System (INIS)

    The Department of Energy currently has over 100 million liters of high-level radioactive waste in storage at the Savannah River Site (SRS). In the late 1970's, the Department of Energy recognized that there were significant safety and cost advantages associated with immobilizing the high-level waste in a stable solid form. Several alternative waste forms were evaluated in terms of product quality and reliability of fabrication. This evaluation led to a decision to build the Defense Waste Processing Facility (DWPF) at SRS to convert the easily dispersed liquid waste to borosilicate glass. In accordance with the NEPA (National Environmental Policy Act) process, an Environmental Impact Statement was prepared for the facility, as well as an Environmental Assessment of the alternative waste forms, and issuance of a Record of Decision (in December, 1982) on the waste form. The Department of Energy, recognizing that start-up of the DWPF would considerably precede licensing of a repository, instituted a Waste Acceptance Process to ensure that these canistered waste forms would be acceptable for eventual disposal at a federal repository. This report is a revision of the DWPF compliance plan

  2. The Immobilisation of Krom and Stronsium Waste Using Natural Fiber Reinforced Cement

    International Nuclear Information System (INIS)

    Cementation of hazardous liquid waste is one of the methods to minimize its detrimental effect on the environmental quality and human health. This research purpose was to study the effect of natural fiber composition and temperature on quality of the cement block reinforced by coconut (Cocos nucifera) fiber and bamboo (Bambusa vulgaris) fiber. This research was pursued by adsorbing stronsium waste and krom using zeolite. Thirteen percent volume of zeolite was mixed with 0.3 of water/cement ratio. Composition of natural fiber was varied by 0.00v/o, 0.05v/o, 0.10v/o, 0.25v/o, 0.50v/o, 0.75v/o and 1.00v/o. The cement blocks produced were heated at 0℃, 50℃, 100℃, 150℃, 200℃ and 250℃ for 10 minutes and then determined their compressive strength and leaching rate. The optimum composition of natural fiber causing increasing of mechanical strength has been founded at 0.50% v/o of fiber. On that composition the axial force resistance of fiber is higher than the radial one. The hydration reaction completely works when cement block is heated until certain temperature that results in the increasing of its compressive strength. However, the compressive strength of cement block heated up to 250℃ is still beyond the standard. Based on its compressive strength, the bamboo (Bambusa vulgaris) fiber is more feasible than coconut (Cocos nucifera) fiber for reinforcing cement block. Heating just influences on the physics properties of cement block. But, the ability of block cement to immobilize a matter is affected by properties of matters. (author)

  3. Ceramic and glass radioactive waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Readey, D.W.; Cooley, C.R. (comps.)

    1977-01-01

    This report contains 14 individual presentations and 6 group reports on the subject of glass and polycrystalline ceramic radioactive waste forms. It was the general consensus that the information available on glass as a waste form provided a good basis for planning on the use of glass as an initial waste form, that crystalline ceramic forms could also be good waste forms if much more development work were completed, and that prediction of the chemical and physical stability of the waste form far into the future would be much improved if the basic synergistic effects of low temperature, radiation and long times were better understood. Continuing development of the polycrystalline ceramic forms was recommended. It was concluded that the leach rate of radioactive species from the waste form is an important criterion for evaluating its suitability, particularly for the time period before solidified waste is permanently placed in the geologic isolation of a Federal repository. Separate abstracts were prepared for 12 of the individual papers; the remaining two were previously abstracted.

  4. Rietveld analysis of ceramic nuclear waste forms

    International Nuclear Information System (INIS)

    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

  5. Rietveld analysis of ceramic nuclear waste forms

    Energy Technology Data Exchange (ETDEWEB)

    White, T.J. [Univ. of South Australia, Ingle Farm (Australia); Mitamura, H. [Japan Atomic Energy Research Institute, Ibaraki (Japan)

    1994-12-31

    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.

  6. Immobilization of Radioactive Waste in Different Fly Ash Zeolite Cement Blends

    International Nuclear Information System (INIS)

    The problem of radioactive waste management has been raised from the beginning use of nuclear energy for different purposes. The rad waste streams produced were sufficient to cause dangerous effects to man and its environment. The ordinary portland cement is the material more extensively used in the technologies of solidification and immobilization of the toxic wastes, low and medium level radioactive wastes. The production of portland cement is one of the most energy-intensive and polluting. The use of high energy in the production causes high emission due to the nature and processes of raw materials. The cement industry is responsible for 7% of the total CO2 emission. Thus, the cement industry has a crucial role in the global warming. The formation of alite (Ca3SiO5), which is the main component of the Portland cement clinker, produces a greater amount of CO2 emission than the formation of belite (Ca2SiO4). The proportion of alite to belite is about 3 in ordinary Portland clinker. Therefore, by decreasing this proportion less CO2 would be emitted. Furthermore, if industrial byproducts such as fly ash from thermal power station or from incineration of municipal solid wastes have the potential to reduce CO2 used as raw materials and alternative hydrothermal calcination routes are employed for belite clinker production, CO2 emission can be strongly reduced or even totally avoided. The availability of fly ash will help in reducing the CO2 emissions and will also help in resolving, to a great extent, the fly ash disposal problem. This thesis is based on focusing on the possibility of using fly ash as raw materials to prepare low cost innovation matrices for immobilization of radioactive wastes by synthesizing new kind of cement of low consuming energy. The synthesis process is based on the hydrothermal-calcination route of the fly ash without extra additions.

  7. Potential Use Of Carbide Lime Waste As An Alternative Material To Conventional Hydrated Lime Of Cement-Lime Mortars

    OpenAIRE

    Al Khaja, Waheeb A.

    1992-01-01

    The present study aimed at the possibility of using the carbide lime waste as an alternative material to the conventional lime used for cement-lime mortar. The waste is a by-product obtained in the generation of acetylene from calcium carbide. Physical and chemical properties of the wastes were studied. Two cement-lime-sand mix proportions containing carbide lime waste were compared with the same mix proportions containing conventional lime along with a control mix without lime. Specimens wer...

  8. Performance Test on Polymer Waste Form

    International Nuclear Information System (INIS)

    Boric acid wastewater and spent ion exchange resins are generated as a low- and medium- level radioactive wastes from pressurized light water reactors. In Korea, boric acid wastewater is concentrated and dried in the form of granules, and finally solidified by using paraffin wax. In this study, polymer solidification was attempted to produce the stable waste form for the boric acid concentrates and the dewatered spent ion exchange resins. The polymer mixture which consists of epoxy resin, amine compounds and antimony trioxide was used to solidify the boric acid concentrates and the dewatered spent ion exchange resins. To evaluate the stability of polymer waste forms, a series of standardized performance tests was conducted. Also, by the requirement of the regulatory institute in Korea, an additional test was performed to estimate fire resistance and gas generation of the waste forms. A series of performance tests was conducted including compressive strength test, thermal stability test, irradiation stability test and biodegradation stability test, water immersion test, leach test, and free standing water for the polymer waste forms. In addition, a fire resistance test and an analysis of gas generation were performed on the waste forms by the requirement of the regulatory institute in Korea. From the results of the performance tests, it is believed that the polymer waste form is very stable and can satisfy the acceptance criteria for permanent disposal

  9. Performance Test on Polymer Waste Form

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Se Yup [Korea Nuclear Engineering Co., Ltd, Seongnam (Korea, Republic of)

    2012-07-01

    Boric acid wastewater and spent ion exchange resins are generated as a low- and medium- level radioactive wastes from pressurized light water reactors. In Korea, boric acid wastewater is concentrated and dried in the form of granules, and finally solidified by using paraffin wax. In this study, polymer solidification was attempted to produce the stable waste form for the boric acid concentrates and the dewatered spent ion exchange resins. The polymer mixture which consists of epoxy resin, amine compounds and antimony trioxide was used to solidify the boric acid concentrates and the dewatered spent ion exchange resins. To evaluate the stability of polymer waste forms, a series of standardized performance tests was conducted. Also, by the requirement of the regulatory institute in Korea, an additional test was performed to estimate fire resistance and gas generation of the waste forms. A series of performance tests was conducted including compressive strength test, thermal stability test, irradiation stability test and biodegradation stability test, water immersion test, leach test, and free standing water for the polymer waste forms. In addition, a fire resistance test and an analysis of gas generation were performed on the waste forms by the requirement of the regulatory institute in Korea. From the results of the performance tests, it is believed that the polymer waste form is very stable and can satisfy the acceptance criteria for permanent disposal.

  10. Secondary Waste Form Down-Selection Data Package - Fluidized Bed Steam Reforming Waste Form

    International Nuclear Information System (INIS)

    The Hanford Site in southeast Washington State has 56 million gallons of radioactive and chemically hazardous wastes stored in 177 underground tanks (ORP 2010). The U.S. Department of Energy (DOE), Office of River Protection (ORP), through its contractors, is constructing the Hanford Tank Waste Treatment and Immobilization Plant (WTP) to convert the radioactive and hazardous wastes into stable glass waste forms for disposal. Within the WTP, the pretreatment facility will receive the retrieved waste from the tank farms and separate it into two treated process streams. These waste streams will be vitrified, and the resulting waste canisters will be sent to offsite (high-level waste (HLW)) and onsite (immobilized low-activity waste (ILAW)) repositories. As part of the pretreatment and ILAW processing, liquid secondary wastes will be generated that will be transferred to the Effluent Treatment Facility (ETF) on the Hanford Site for further treatment. These liquid secondary wastes will be converted to stable solid waste forms that will be disposed of in the Integrated Disposal Facility (IDF). To support the selection of a waste form for the liquid secondary wastes from WTP, Washington River Protection Solutions (WRPS) has initiated secondary waste form testing work at Pacific Northwest National Laboratory (PNNL). In anticipation of a down-selection process for a waste form for the Solidification Treatment Unit to be added to the ETF, PNNL is developing data packages to support that down-selection. The objective of the data packages is to identify, evaluate, and summarize the existing information on the four waste forms being considered for stabilizing and solidifying the liquid secondary wastes. At the Hanford Site, the FBSR process is being evaluated as a supplemental technology for treating and immobilizing Hanford LAW radioactive tank waste and for treating secondary wastes from the WTP pretreatment and LAW vitrification processes.

  11. Secondary Waste Form Down-Selection Data Package—Fluidized Bed Steam Reforming Waste Form

    Energy Technology Data Exchange (ETDEWEB)

    Qafoku, Nikolla; Westsik, Joseph H.; Strachan, Denis M.; Valenta, Michelle M.; Pires, Richard P.

    2011-09-12

    The Hanford Site in southeast Washington State has 56 million gallons of radioactive and chemically hazardous wastes stored in 177 underground tanks (ORP 2010). The U.S. Department of Energy (DOE), Office of River Protection (ORP), through its contractors, is constructing the Hanford Tank Waste Treatment and Immobilization Plant (WTP) to convert the radioactive and hazardous wastes into stable glass waste forms for disposal. Within the WTP, the pretreatment facility will receive the retrieved waste from the tank farms and separate it into two treated process streams. These waste streams will be vitrified, and the resulting waste canisters will be sent to offsite (high-level waste [HLW]) and onsite (immobilized low-activity waste [ILAW]) repositories. As part of the pretreatment and ILAW processing, liquid secondary wastes will be generated that will be transferred to the Effluent Treatment Facility (ETF) on the Hanford Site for further treatment. These liquid secondary wastes will be converted to stable solid waste forms that will be disposed of in the Integrated Disposal Facility (IDF). To support the selection of a waste form for the liquid secondary wastes from WTP, Washington River Protection Solutions (WRPS) has initiated secondary waste form testing work at Pacific Northwest National Laboratory (PNNL). In anticipation of a down-selection process for a waste form for the Solidification Treatment Unit to be added to the ETF, PNNL is developing data packages to support that down-selection. The objective of the data packages is to identify, evaluate, and summarize the existing information on the four waste forms being considered for stabilizing and solidifying the liquid secondary wastes. At the Hanford Site, the FBSR process is being evaluated as a supplemental technology for treating and immobilizing Hanford LAW radioactive tank waste and for treating secondary wastes from the WTP pretreatment and LAW vitrification processes.

  12. Immobilization of radioactive waste in cement based matrices

    International Nuclear Information System (INIS)

    A mathematical and thermodynamic model of the Ca0-Si02-H20 system is presented to enable the solubility and pH relationships in cement and blended cement systems to be predicted. The Esub(h) function has been explored particularly in respect of slag rich systems. The stability of Sr in cements is shown to be influenced by both precipitation and lattice incorporation into the ettringite-like phase. Quality assurance parameters especially for aggregate materials and blast furnace slags are reviewed and recommendations made. It is shown that the latter fluctuate considerably in composition; additional measures for monitoring are recommended and additional research suggested to determine their long-term performance. (author)

  13. Development of Polymeric Waste Forms for the Encapsulation of Toxic Wastes Using an Emulsion-Encapsulation Based Process

    Energy Technology Data Exchange (ETDEWEB)

    Evans, R.; Quach, A.; Birnie, D. P.; Saez, A. E.; Ela, W. P.; Zeliniski, B. J. J.; Xia, G.; Smith, H.

    2003-01-01

    Developed technologies in vitrification, cement, and polymeric materials manufactured using flammable organic solvents have been used to encapsulate solid wastes, including low-level radioactive materials, but are impractical for high salt-content waste streams (Maio, 1998). In this work, we investigate an emulsification process for producing an aqueous-based polymeric waste form as a preliminary step towards fabricating hybrid organic/inorganic polyceram matrices. The material developed incorporates epoxy resin and polystyrene-butadiene (PSB) latex to produce a waste form that is non-flammable, light weight, of relatively low cost, and that can be loaded to a relatively high weight content of waste materials. Sodium nitrate was used as a model for the salt waste. Small-scale samples were manufactured and analyzed using leach tests designed to measure the diffusion coefficient and leachability index for the fastest diffusing species in the waste form, the salt ions. The microstructure and composition of the samples were probed using SEM/EDS techniques. The results show that some portion of the salt migrates towards the exterior surfaces of the waste forms during the curing process. A portion of the salt in the interior of the sample is contained in polymer corpuscles or sacs. These sacs are embedded in a polymer matrix phase that contains fine, well-dispersed salt crystals. The diffusion behavior observed in sections of the waste forms indicates that samples prepared using this emulsion process meet or exceed the leachability criteria suggested for low level radioactivity waste forms.

  14. Evaluation of interim and final waste forms for the newly generated liquid low-level waste flowsheet

    International Nuclear Information System (INIS)

    The purpose of this review is to evaluate the final forms that have been proposed for radioactive-containing solid wastes and to determine their application to the solid wastes that will result from the treatment of newly generated liquid low-level waste (NGLLLW) and Melton Valley Storage Tank (MVST) supernate at the Oak Ridge National Laboratory (ORNL). Since cesium and strontium are the predominant radionuclides in NGLLLW and MVST supernate, this review is focused on the stabilization and solidification of solid wastes containing these radionuclides in cement, glass, and polymeric materials-the principal waste forms that have been tested with these types of wastes. Several studies have shown that both cesium and strontium are leached by distilled water from solidified cement, although the leachabilities of cesium are generally higher than those of strontium under similar conditions. The situation is exacerbated by the presence of sulfates in the solution, as manifested by cracking of the grout. Additives such as bentonite, blast-furnace slag, fly ash, montmorillonite, pottery clay, silica, and zeolites generally decrease the cesium and strontium release rates. Longer cement curing times (>28 d) and high ionic strengths of the leachates, such as those that occur in seawater, also decrease the leach rates of these radionuclides. Lower cesium leach rates are observed from vitrified wastes than from grout waste forms. However, significant quantities of cesium are volatilized due to the elevated temperatures required to vitrify the waste. Hence, vitrification will generally require the use of cleanup systems for the off-gases to prevent their release into the atmosphere

  15. 137Cs contaminated waste disposal in cement factory: Environmental problems

    International Nuclear Information System (INIS)

    In the course of utilization (May-June '91) of aluminum slags polluted by 137Cs at the cement factory Presacementi in Robilante (Cuneo, Italy) and during the following months, samples were taken in particular points of the plant, at fixed frequencies. Samples were analyzed to determine 137Cs concentration. Collected data were used to study the behaviour of the element throughout the process. Emissions and ground level air concentrations were estimated from the available data. Contamination of the manufactured cement were monitored until negligible values of 137Cs concentration were attained

  16. Iodine waste form summary report (FY 2007).

    Energy Technology Data Exchange (ETDEWEB)

    Krumhansl, James Lee; Nenoff, Tina Maria; McMahon, Kevin A.; Gao, Huizhen; Rajan, Ashwath Natech

    2007-11-01

    This new program at Sandia is focused on Iodine waste form development for GNEP cycle needs. Our research has a general theme of 'Waste Forms by Design' in which we are focused on silver loaded zeolite waste forms and related metal loaded zeolites that can be validated for chosen GNEP cycle designs. With that theme, we are interested in materials flexibility for iodine feed stream and sequestration material (in a sense, the ability to develop a universal material independent on the waste stream composition). We also are designing the flexibility to work in a variety of repository or storage scenarios. This is possible by studying the structure/property relationship of existing waste forms and optimizing them to our current needs. Furthermore, by understanding the properties of the waste and the storage forms we may be able to predict their long-term behavior and stability. Finally, we are working collaboratively with the Waste Form Development Campaign to ensure materials durability and stability testing.

  17. Studies on the Potential of Waste Soda Lime Silica Glass in Glass Ionomer Cement Production

    Directory of Open Access Journals (Sweden)

    V. W. Francis Thoo

    2013-01-01

    Full Text Available Glass ionomer cements (GIC are produced through acid base reaction between calcium-fluoroaluminosilicate glass powder and polyacrylic acid (PAA. Soda lime silica glasses (SLS, mainly composed of silica (SiO2, have been utilized in this study as the source of SiO2 for synthesis of Ca-fluoroaluminosilicate glass. Therefore, the main objective of this study was to investigate the potential of SLS waste glass in producing GIC. Two glasses, GWX 1 (analytical grade SiO2 and GWX 2 (replacing SiO2 with waste SLS, were synthesized and then characterized using X-ray diffraction (XRD and energy dispersive X-ray (EDX. Synthesized glasses were then used to produce GIC, in which the properties were characterized using Fourier transform infrared spectroscopy (FT-IR and compressive test (from 1 to 28 days. XRD results showed that amorphous glass was produced by using SLS waste glass (GWX 2, which is similar to glass produced using analytical grade SiO2 (GWX 1. Results from FT-IR showed that the setting reaction of GWX 2 cements is slower compared to cement GWX 1. Compressive strengths for GWX 1 cements reached up to 76 MPa at 28 days, whereas GWX 2 cements showed a slightly higher value, which is 80 MPa.

  18. Activity release from waste packages containing LL and IL waste forms under mechanical and thermal stresses

    International Nuclear Information System (INIS)

    For transport and handling of radioactive waste packages in an underground repository safety assessments are being performed to keep any unacceptable radiation hazards from the operational staff and the population in the site neighborhood. Therefore experiments were carried out to determine source terms for activity release from waste packages containing cemented waste forms in case of heavy mechanical and thermal impacts. Mechanical impact was applied by drop test with a maximum energy input of 3.105 Nm. A special cage construction around the target (reinforced concrete covered by a 80 mm steel plate) allows the collection of the airborne fines with a particle size of < 10 μm by using micro filters in a defined geometry. In addition, in two experiments the particle fraction with an aerodynamic diameter between 1 μm and 20 μm was determined using a cascade impactor. Additional laboratory experiments were performed to determine comparative values for different waste forms. In case of thermal impact, the temperature profiles in the waste forms were measured and the release of added indicators (Cs, Sr, Eu) was determined. Further laboratory experiments were performed with inactive samples to determine the temperature dependence of water release (Thermogravimetric-Analysis)

  19. Characterizing cemented TRU waste for RCRA hazardous constituents

    International Nuclear Information System (INIS)

    Los Alamos National Laboratory (LANL) has characterized drums of solidified transuranic (TRU) waste from four major waste streams. The data will help the State of New Mexico determine whether or not to issue a no-migration variance of the Waste Isolation Pilot Plant (WIPP) so that WIPP can receive and dispose of waste. The need to characterize TRU waste stored at LANL is driven by two additional factors: (1) the LANL RCRA Waste Analysis Plan for EPA compliant safe storage of hazardous waste; (2) the WIPP Waste Acceptance Criteria (WAC) The LANL characterization program includes headspace gas analysis, radioassay and radiography for all drums and solids sampling on a random selection of drums from each waste stream. Data are presented showing that the only identified non-metal RCRA hazardous component of the waste is methanol

  20. Leachability characteristics of beryllium in redmud waste and its stabilization in cement

    International Nuclear Information System (INIS)

    More than 70% of the beryl ore processed by the Beryllium Metal Plant at the BARC Vashi Complex ends up as redmud waste. The presence of significant quantities (0.4 to 0.8%) of beryllium in the redmud qualifies it as hazardous requiring safe handling, storage and disposal. The waste also contains 0.09% of water soluble fluoride. The various standard protocol of procedures were employed to estimate the leachability of beryllium from redmud for both short term and long term periods. Nearly 50% of beryllium present in redmud is leachable in water. We have tried the stabilization of redmud using portland cement. The proportion of redmud to cement was in the ratio of 1:1, 1:2 and 1:4. The blocks were cast, cured and used in the leachability experiments using standard protocols as above. The results of the TCLP test gave the levels of beryllium well below the standard limits in the TCLP extract of cement stabilized waste indicating the suitability of stabilization of redmud with cement whereas that of raw waste (redmud) are much higher than the prescribed limits. The total leach percent of beryllium in 1:2 block is 0.05% over period of 164 days whereas 1:1 and 1:4 gave a leach percent of 0.26 and 0.15% respectively. The DLT results indicate, diffusion controlled release of beryllium from the cement stabilized redmud blocks. The effective diffusion coefficient of beryllium obtained from the modelling study is 10 orders of magnitude less than the molecular diffusion coefficient of beryllium indicating the effectiveness of cement stabilization. From the detailed experiments performed, it is felt that 1:2 proportion of redmud and cement will be the best suited option for stabilization of redmud waste. The 1:1 proportion of redmud to cement mixture which could not be cast into compact cement blocks also exhibited very low leachability characteristics similar to 1:2 and 1:4 and can be be favourably considered for stabilization in case of space constraints at storage sites. The

  1. CRYSTALLINE CERAMIC WASTE FORMS: REFERENCE FORMULATION REPORT

    Energy Technology Data Exchange (ETDEWEB)

    Brinkman, K.; Fox, K.; Marra, J.

    2012-05-15

    The research conducted in this work package is aimed at taking advantage of the long term thermodynamic stability of crystalline ceramics to create more durable waste forms (as compared to high level waste glass) in order to reduce the reliance on engineered and natural barrier systems. Durable ceramic waste forms that incorporate a wide range of radionuclides have the potential to broaden the available disposal options and to lower the storage and disposal costs associated with advanced fuel cycles. Assemblages of several titanate phases have been successfully demonstrated to incorporate radioactive waste elements, and the multiphase nature of these materials allows them to accommodate variation in the waste composition. Recent work has shown that they can be successfully produced from a melting and crystallization process. The objective of this report is to explain the design of ceramic host systems culminating in a reference ceramic formulation for use in subsequent studies on process optimization and melt property data assessment in support of FY13 melter demonstration testing. The waste stream used as the basis for the development and testing is a combination of the projected Cs/Sr separated stream, the Trivalent Actinide - Lanthanide Separation by Phosphorous reagent Extraction from Aqueous Komplexes (TALSPEAK) waste stream consisting of lanthanide fission products, the transition metal fission product waste stream resulting from the transuranic extraction (TRUEX) process, and a high molybdenum concentration with relatively low noble metal concentrations. In addition to the combined CS/LN/TM High Mo waste stream, variants without Mo and without Mo and Zr were also evaluated. Based on the results of fabricating and characterizing several simulated ceramic waste forms, two reference ceramic waste form compositions are recommended in this report. The first composition targets the CS/LN/TM combined waste stream with and without Mo. The second composition targets

  2. SEPARATIONS AND WASTE FORMS CAMPAIGN IMPLEMENTATION PLAN

    Energy Technology Data Exchange (ETDEWEB)

    Vienna, John D.; Todd, Terry A.; Peterson, Mary E.

    2012-11-26

    This Separations and Waste Forms Campaign Implementation Plan provides summary level detail describing how the Campaign will achieve the objectives set-forth by the Fuel Cycle Reasearch and Development (FCRD) Program. This implementation plan will be maintained as a living document and will be updated as needed in response to changes or progress in separations and waste forms research and the FCRD Program priorities.

  3. Immobilisation of MTR waste in cement (product evaluation). Final report. December 1987

    International Nuclear Information System (INIS)

    The enriched uranium/aluminium fuel used in Material Testing Reactors is reprocessed at Dounreay Nuclear Power Development Establishment (DNE). The main chemical component of the liquid waste produced by this process is acid deficient aluminium nitrate. This is stored in stainless steel tanks at DNE. As a result of work carried out under the UKAEA radioactive waste management programme a decision was taken to immobilise the waste in cement. The programme had two main components, plant design and development of the cementation process. The plant for the cementation of MTR waste is under construction and will be commissioned in 1988/9. The primary objective of this project is to find a suitable process for changing the highly mobile radioactive waste into an inert stable solid. Work carried out on the development of the immobilisation process showed that a conditioning stage (neutralisation) is required to make the acid waste compatible with cement. Small scale experiments showed that adding Ordinary Portland Cement blended with ground granulated Blast Furnace Slag to Simulant MTR Liquor produces an acceptable product. The process has been demonstrated at full scale (200 litres) and the products have been subjected to an extensive programme of destructive and non-destructive testing. Specimens have been tested up to 1200 days after manufacture and show no significant signs of deterioration even when stored underwater or when subjected to freeze thaw cycling. Development work has also shown that the process can successfully immobilise simulant MTR liquor over a wide range of liquor concentrations. The programme therefore successfully produced a formulation that met all the requirements of both the process and product specification. (author)

  4. Weathering Effect on 99Tc Leachability from Cementitious Waste Form

    International Nuclear Information System (INIS)

    The mass transfer of contaminants from the solid phase to the waste form pore water, and subsequently out of the solid waste form, is directly related to the number and size distribution of pores as well as the microstructure of the waste form. Because permeability and porosity are controlled by pore aperture size, pore volume, and pore distribution, it is important to have some indication of how these characteristics change in the waste form during weathering. Knowledge of changes in these key parameters can be used to develop predictive models that estimate diffusivity or permeability of radioactive contaminants can be used to develop predictive models that estimate diffusivity or permeability of radioactive contaminants from waste forms for long-term performance assessment. It is known that dissolution or precipitation of amorphous/crystalline phases within waste forms alters their pore structure and controls the transport of contaminants our of waste forms. One very important precipitate is calcite, which is formed as a result of carbonation reactions in cement and other high-alkalinity waste forms. Enhanced oxidation can also increase Tc leachability from the waste form. To account for these changes, weathering experiments were conducted in advance to increase our understating of the long-term Tc leachability, especially out of the cementitious waste form. Pore structure analysis was characterized using both N2 absorption analysis and XMT techniques, and the results show that cementitious waste form is a relatively highly-porous material compared to other waste forms studied in this task, Detailed characterization of Cast Stone chunks and monolith specimens indicate that carbonation reactions can change the Cast Stone pore structure, which in turn may correlate with Tc leachability. Short carbonation reaction times for the Cast Stone causes pore volume and surface area increases, while the average pore diameter decreases. Based on the changes in pore volumes

  5. Sulfur polymer cement, a new stabilization agent for mixed and low- level radioactive waste

    International Nuclear Information System (INIS)

    Solidification and stabilization agents for radioactive, hazardous, and mixed wastes are failing to pass governmental tests at alarming rates. The Department of Energy's National Low-Level Waste Management Program funded testing of Sulfur Polymer Cement (SPC) by Brookhaven National Laboratory during the 1980s. Those tests and tests by the US Bureau of Mines (the original developer of SPC), universities, states, and the concrete industry have shown SPC to be superior to hydraulic cements in most cases. Superior in what wastes can be successfully combined and in the quantity of waste that can be combined and still pass the tests established by the US Environmental Protection Agency and the US Nuclear Regulatory Commission

  6. Performance Characteristics of Waste Glass Powder Substituting Portland Cement in Mortar Mixtures

    Science.gov (United States)

    Kara, P.; Csetényi, L. J.; Borosnyói, A.

    2016-04-01

    In the present work, soda-lime glass cullet (flint, amber, green) and special glass cullet (soda-alkaline earth-silicate glass coming from low pressure mercury-discharge lamp cullet and incandescent light bulb borosilicate glass waste cullet) were ground into fine powders in a laboratory planetary ball mill for 30 minutes. CEM I 42.5N Portland cement was applied in mortar mixtures, substituted with waste glass powder at levels of 20% and 30%. Characterisation and testing of waste glass powders included fineness by laser diffraction particle size analysis, specific surface area by nitrogen adsorption technique, particle density by pycnometry and chemical analysis by X-ray fluorescence spectrophotometry. Compressive strength, early age shrinkage cracking and drying shrinkage tests, heat of hydration of mortars, temperature of hydration, X-ray diffraction analysis and volume stability tests were performed to observe the influence of waste glass powder substitution for Portland cement on physical and engineering properties of mortar mixtures.

  7. Chemical compatibility of DWPF canistered waste forms

    International Nuclear Information System (INIS)

    The Waste Acceptance Preliminary Specifications (WAPS) require that the contents of the canistered waste form are compatible with one another and the stainless steel canister. The canistered waste form is a closed system comprised of a stainless steel vessel containing waste glass, air, and condensate. This system will experience a radiation field and an elevated temperature due to radionuclide decay. This report discusses possible chemical reactions, radiation interactions, and corrosive reactions within this system both under normal storage conditions and after exposure to temperatures up to the normal glass transition temperature, which for DWPF waste glass will be between 440 and 460 degrees C. Specific conclusions regarding reactions and corrosion are provided. This document is based on the assumption that the period of interim storage prior to packaging at the federal repository may be as long as 50 years

  8. Mixture optimization of cement treated demolition waste with recycled masonry and concrete

    NARCIS (Netherlands)

    Xuan, D.X.; Houben, L.J.M.; Molenaar, A.A.A.; Shui, Z.H

    2011-01-01

    Due to environmental reasons and the shortage of natural resources, it is greatly valuable to recycle construction and demolition waste (CDW) as much as possible. One of effective ways to reuse more CDW is to produce a cemented road base material. The recycled CDW however is a mix of recycled masonr

  9. Investigation of combined effect of mixture variables on mechanical properties of cement treated demolition waste

    NARCIS (Netherlands)

    Xuan, D.; Houben, L.J.M.; Molenaar, A.A.A.; Shui, Z.

    2012-01-01

    One of high efficient ways to reuse the recycled construction and demolition waste (CDW) is to consider it as a road base material. The recycled CDW however is mainly a mix of recycled masonry and concrete with a wide variation in composition. This results that the mechanical properties of cement tr

  10. Systematic approach for the design of pumpable cement-based grouts for immobilization of hazardous wastes

    International Nuclear Information System (INIS)

    Cement-based grouts have been proven to be an economical and environmentally acceptable means of waste disposal. Costs can be reduced if the grout is pumped to the disposal site. This paper presents a systematic approach to guide the development of pumpable grouts. 20 refs., 2 figs

  11. Energetic and exergetic analysis of waste heat recovery systems in the cement industry

    International Nuclear Information System (INIS)

    In a typical cement producing procedure, 25% of the total energy used is electricity and 75% is thermal energy. However, the process is characterized by significant heat losses mainly by the flue gases and the ambient air stream used for cooling down the clinker (about 35%–40% of the process heat loss). Approximately 26% of the heat input to the system is lost due to dust, clinker discharge, radiation and convection losses from the kiln and the preheaters. A heat recovery system could be used to increase the efficiency of the cement plant and thus contribute to emissions decrease. The aim of this paper is to examine and compare energetically and exergetically, two different WHR (waste heat recovery) methods: a water-steam Rankine cycle, and an Organic Rankine Cycle (ORC). A parametric study proved that the water steam technology is more efficient than ORC in exhaust gases temperature higher than 310 °C. Finally a brief economic assessment of the most efficient solution was implemented. WHR installations in cement industry can contribute significantly in the reduction of the electrical consumptions operating cost thus being a very attractive investment with a payback period up to 5 years. - Highlights: • This paper presents waste heat recovery as a way to gain energy from the exhaust gases in a cement plant. • Water steam cycle and ORC has been analyzed for waste heat recovery. • The energetic and exergetic evaluation of the two waste heat recovery processes is presented and compared

  12. Improved cement mortars by addition of carbonated fly ash from solid waste incinerators

    OpenAIRE

    López-Zaldívar, O.; Mayor-Lobo, P. L.; Fernández-Martínez, F.; Hernández-Olivares, F.

    2015-01-01

    This article presents the results of a research developing high performance cement mortars with the addition of municipal solid waste incineration fly ash (MSWIFA) stabilized as insoluble carbonates. The encapsulation of hazardous wastes in mortar matrixes has also been achieved. The ashes present high concentrations of chlorides, Zn and Pb. A stabilization process with NaHCO3 has been developed reducing 99% the content of chlorides. Developed mortars replace 10% per weight of the aggregates ...

  13. Transuranic contaminated waste form characterization and data base

    Energy Technology Data Exchange (ETDEWEB)

    Kniazewycz, B.G.; McArthur, W.C.

    1980-07-01

    This volume contains appendices A to F. The properties of transuranium (TRU) radionuclides are described. Immobilization of TRU wastes by bituminization, urea-formaldehyde polymers, and cements is discussed. Research programs at DOE facilities engaged in TRU waste characterization and management studies are described.

  14. Transuranic contaminated waste form characterization and data base

    International Nuclear Information System (INIS)

    This volume contains appendices A to F. The properties of transuranium (TRU) radionuclides are described. Immobilization of TRU wastes by bituminization, urea-formaldehyde polymers, and cements is discussed. Research programs at DOE facilities engaged in TRU waste characterization and management studies are described

  15. Corrosion susceptibility of steel drums containing cemented intermediate level nuclear wastes

    International Nuclear Information System (INIS)

    Cementation processes are used as immobilization techniques for low or intermediate level radioactive waste for economical and safety reasons and for being a simple operation. In particular, ion-exchange resins commonly used for purification of radioactive liquid waste from nuclear reactors are immobilized before being stored to improve the leach resistance of the waste matrix and to maintain mechanical stability. Combustible solid radioactive waste can be incinerated and the resulting ashes can also be immobilized before storage. The immobilized resins and ashes are then contained in steel drums that may undergo corrosion depending on the presence of certain contaminants. The work described in this paper was aimed at evaluating the corrosion susceptibility of steel drums in contact with cemented ion-exchange resins and incineration ashes containing different concentrations of aggressive species (mostly chloride and sulphate ions). A special type of specimen was designed to simulate the cemented waste in the drum. The evolution of the corrosion potential and the corrosion current density of the steel, as well as the electrical resistivity of the matrix were monitored over a time period of 1 year. The results show the deleterious effect of chloride on the expected lifespan of the waste containers.

  16. Possibility of using waste tire rubber and fly ash with Portland cement as construction materials.

    Science.gov (United States)

    Yilmaz, Arin; Degirmenci, Nurhayat

    2009-05-01

    The growing amount of waste rubber produced from used tires has resulted in an environmental problem. Recycling waste tires has been widely studied for the last 20 years in applications such as asphalt pavement, waterproofing systems and membrane liners. The aim of this study is to evaluate the feasibility of utilizing fly ash and rubber waste with Portland cement as a composite material for masonry applications. Class C fly ash and waste automobile tires in three different sizes were used with Portland cement. Compressive and flexural strength, dry unit weight and water absorption tests were performed on the composite specimens containing waste tire rubber. The compressive strength decreased by increasing the rubber content while increased by increasing the fly ash content for all curing periods. This trend is slightly influenced by particle size. For flexural strength, the specimens with waste tire rubber showed higher values than the control mix probably due to the effect of rubber fibers. The dry unit weight of all specimens decreased with increasing rubber content, which can be explained by the low specific gravity of rubber particles. Water absorption decreased slightly with the increase in rubber particles size. These composite materials containing 10% Portland cement, 70% and 60% fly ash and 20% and 30% tire rubber particles have sufficient strength for masonry applications. PMID:19110410

  17. Energy recovery from wastes : experience with solid alternative fuels combustion in a precalciner cement kiln

    OpenAIRE

    Tokheim, Lars-André; Gautestad, Tor; Axelsen, Ernst Petter; Bjerketvedt, Dag

    2001-01-01

    Today virtually all cement clinker burning processes take place in rotary kilns. A mixture of calcareous and argilaceous materials is heated to a temperature of about 1450 °C. In this process decarbonation followed by partial fusion occurs, and nodules of so-called clinker are formed. The cooled clinker is mixed with a few percent of gypsum, and ground into a fine meal – cement. The most modern cement kilns are equipped with a precalciner, in which most of the calcium carbonate...

  18. The immobilisation of shredded waste in a cement matrix

    International Nuclear Information System (INIS)

    Progress on the preparations for the encapsulation of plutonium contaminated shredded waste is summarised. Waste drums have been modified and filled with active shredded waste. Commissioning of the grout infilling test rig was started at the end of this period. Inactive process trials have continued in support of the design of the active encapsulation plant. (author)

  19. USE OF CONSTRUCTION AND DEMOLITION WASTES AS RAW MATERIALS IN CEMENT CLINKER PRODUCTION

    Institute of Scientific and Technical Information of China (English)

    Christos-Triantafyllos Galbenis; Stamatis Tsimas

    2006-01-01

    The aim of the present paper was to investigate the possibility of utilizing Construction and Demolition(C&D) wastes as substitutes of Portland cement raw meal. The C&D wastes that were so used, were the Recycled Concrete Aggregates (RCA) and the Recycled Masonry Aggregates (RMA) derived from demolished buildings in Attica region, Greece. RCA and RMA samples were selected because of their calcareous and siliceous origin respectively,which conformed the composition of the ordinary Portland cement raw meal. For that reason, six samples of cement raw meals were prepared: one with ordinary raw materials, as a reference sample, and five by mixing the reference sample with RCA and RMA in appropriate proportions. The effect on the reactivity of the generated mixtures, was evaluated on the basis of the free lime content (fCaO) in the mixtures sintered at 1350℃, 1400℃ and 1450℃. Test showed that the added recycled aggregates improved the burnability of the cement raw meal without affecting negatively the cement clinker properties. Moreover, the formation of the major components (C3S, C2S, C3A and C4AF) of the produced clinkers(sintered at 1450℃) was corroborated by X-Ray Diffraction (XRD).

  20. Effects of sucrose and sorbitol on cement-based stabilization/solidification of toxic metal waste.

    Science.gov (United States)

    Zhang, Linghong; Catalan, Lionel J J; Larsen, Andrew C; Kinrade, Stephen D

    2008-03-01

    The effects of sucrose or sorbitol addition on the hydration, unconfined compressive strength and leachability of Portland cement pastes containing 1% Pb and 1% Zn were studied as a function of time. Whereas Pb and Zn were found to shorten the time to achieve maximum hydration of Portland cement, the combination of these metals with 0.15 wt% sucrose or 0.40 wt% sorbitol retarded the setting of cement by at least 7 and 28 days, respectively, without affecting the strength at 56 days. The leachability of Pb and Zn evaluated by the TCLP 1311 protocol at 56 and 71 days was slightly reduced or unchanged by the addition of sucrose or sorbitol. SEM-EDS and XRD analyses revealed that ettringite precipitation was favored whereas the formation of CSH gel, which accounts for most of the strength of hydrated cement, was delayed in cement pastes containing both metals and sucrose or sorbitol. These results indicate that controlled additions of sucrose or sorbitol can add flexibility to the handling of cement-treated metal waste, particularly when it needs to be transported by truck or pipeline between the treatment plant and the disposal site, without affecting its long-term performance. PMID:17629400

  1. Evaluation of final waste forms and recommendations for baseline alternatives to group and glass

    Energy Technology Data Exchange (ETDEWEB)

    Bleier, A.

    1997-09-01

    An assessment of final waste forms was made as part of the Federal Facilities Compliance Agreement/Development, Demonstration, Testing, and Evaluation (FFCA/DDT&E) Program because supplemental waste-form technologies are needed for the hazardous, radioactive, and mixed wastes of concern to the Department of Energy and the problematic wastes on the Oak Ridge Reservation. The principal objective was to identify a primary waste-form candidate as an alternative to grout (cement) and glass. The effort principally comprised a literature search, the goal of which was to establish a knowledge base regarding four areas: (1) the waste-form technologies based on grout and glass, (2) candidate alternatives, (3) the wastes that need to be immobilized, and (4) the technical and regulatory constraints on the waste-from technologies. This report serves, in part, to meet this goal. Six families of materials emerged as relevant; inorganic, organic, vitrified, devitrified, ceramic, and metallic matrices. Multiple members of each family were assessed, emphasizing the materials-oriented factors and accounting for the fact that the two most prevalent types of wastes for the FFCA/DDT&E Program are aqueous liquids and inorganic sludges and solids. Presently, no individual matrix is sufficiently developed to permit its immediate implementation as a baseline alternative. Three thermoplastic materials, sulfur-polymer cement (inorganic), bitumen (organic), and polyethylene (organic), are the most technologically developed candidates. Each warrants further study, emphasizing the engineering and economic factors, but each also has limitations that regulate it to a status of short-term alternative. The crystallinity and flexible processing of sulfur provide sulfur-polymer cement with the highest potential for short-term success via encapsulation. Long-term immobilization demands chemical stabilization, which the thermoplastic matrices do not offer. Among the properties of the remaining

  2. Evaluation of final waste forms and recommendations for baseline alternatives to grout and glass

    International Nuclear Information System (INIS)

    An assessment of final waste forms was made as part of the Federal Facilities Compliance Agreement/Development, Demonstration, Testing, and Evaluation (FFCA/DDT ampersand E) Program because supplemental waste-form technologies are needed for the hazardous, radioactive, and mixed wastes of concern to the Department of Energy and the problematic wastes on the Oak Ridge Reservation. The principal objective was to identify a primary waste-form candidate as an alternative to grout (cement) and glass. The effort principally comprised a literature search, the goal of which was to establish a knowledge base regarding four areas: (1) the waste-form technologies based on grout and glass, (2) candidate alternatives, (3) the wastes that need to be immobilized, and (4) the technical and regulatory constraints on the waste-from technologies. This report serves, in part, to meet this goal. Six families of materials emerged as relevant; inorganic, organic, vitrified, devitrified, ceramic, and metallic matrices. Multiple members of each family were assessed, emphasizing the materials-oriented factors and accounting for the fact that the two most prevalent types of wastes for the FFCA/DDT ampersand E Program are aqueous liquids and inorganic sludges and solids. Presently, no individual matrix is sufficiently developed to permit its immediate implementation as a baseline alternative. Three thermoplastic materials, sulfur-polymer cement (inorganic), bitumen (organic), and polyethylene (organic), are the most technologically developed candidates. Each warrants further study, emphasizing the engineering and economic factors, but each also has limitations that regulate it to a status of short-term alternative. The crystallinity and flexible processing of sulfur provide sulfur-polymer cement with the highest potential for short-term success via encapsulation. Long-term immobilization demands chemical stabilization, which the thermoplastic matrices do not offer. Among the properties of the

  3. Performance of aged cement - polymer composite immobilizing borate waste simulates during flooding scenarios

    International Nuclear Information System (INIS)

    An advanced composite of cement and water extended polyester based on the recycled Poly(ethylene terephthalate) waste was developed to incorporate the borate waste. Previous studies have reported the characterizations of the waste composite (cement-polymer composite immobilizing borate waste simulates) after 28 days of curing time. The current work studied the performance of waste composite aged for seven years and subjected to flooding scenario during 260 days using three types of water. The state of waste composite was assessed at the end of each definite interval of the water infiltration through visual examination and mechanical measurement. Scanning electron microscopy, infrared spectroscopy, X-ray diffraction and thermal analyses were used to investigate the changes that may occur in the microstructure of the waste composite under aging and flooding effects. The actual experimental results indicated reasonable evidence for the waste composite. Acceptable consistency was confirmed for the waste composite even after aging seven years and exposure to flooding scenario for 260 days.

  4. Optimization and validation of a chemical process for uranium, mercury and cesium leaching from cemented radioactive wastes

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) is developing a treatment and long-term management strategy for a legacy cemented radioactive waste that contains uranium, mercury and fission products. Extracting the uranium would be advantageous for decreasing the waste classification and reducing the cost of long-term management. Consequently, there are safety and economic and environmental incentives for the extraction of uranium, mercury and cesium before subjecting the cemented waste to a stabilization process. The mineralogical analysis of the surrogate cemented waste (SCW) indicated that uranium forms calcium uranate, CaUO4, occurring as layers of several millimeters or as grains of 20 μm. Hg is found mostly as large (∼50 μm) and small grains (5-8 μm) of HgO. The chemical leachability of three key elements (U, Hg, and Cs) from a SCW was studied with several leaching materials. The results showed that the most promising approach to leach and recover U, Hg, and Cs is the direct leaching of the SCW with H2SO4 in strong saline media. Operating parameters such as particle size, temperature, pulp density, leaching time, acid and salt concentrations, number of leaching/rinsing step, etc. were optimized to improve key elements solubilization. Sulfuric leaching in saline media of a SCW (U5) containing 1182 ppm of U, 1598 ppm of Hg, and 7.9 ppm of Cs in the optimized conditions allows key elements recovery of 98.5 ± 0.4%, 96.6 ± 0.1%, and 93.8 ± 1.1% of U, Hg, and Cs respectively. This solubilization process was then applied in triplicate to seven other SCW prepared with different cement, liquid ratio and at different aging time and temperature. Concentrated sulfuric acid is added to the slurry until the pH is about 2, which causes the complete degradation of cement and the formation of CaSO4. At this pH, the acid consumption is moderate and the formation of amorphous silica gel is avoided. Sulfuric acid is particularly useful because it produces a leachate that is

  5. Lysimeter tests of SRP waste forms

    International Nuclear Information System (INIS)

    A field study, estimated to last 10 years, has been started to define leaching and migration rates of radionuclides from typical SRP buried wastes. The study utilizes 42 lysimeters (6-ft or 10-ft diameter by 10-ft deep) which have been charged with soil and waste to simulate burial ground conditions. Eight waste forms were selected for the study, which represent the bulk of the wastes generated at SRP. This report describes the lysimeter design, the physical and radiological characteristics of the wastes, and the experimental approach. Calculations have also been made which predict the migration of various radionuclides in the lysimeter soil. The calculations should provide guidance during the course of the study, and are the basis of recommendations made for collecting and interpreting data so that important parameters of migration can be evaluated

  6. Construction of solid waste form test facility

    International Nuclear Information System (INIS)

    The Solid Waste Form Test Facility (SWFTF) is now construction at DAEDUCK in Korea. In SWFTF, the characteristics of solidified waste products as radiological homogeneity, mechanical and thermal property, water resistance and lechability will be tested and evaluated to meet conditions for long-term storage or final disposal of wastes. The construction of solid waste form test facility has been started with finishing its design of a building and equipments in Sep. 1984, and now building construction is completed. Radioactive gas treatment system, extinguishers, cooling and heating system for the facility, electrical equipments, Master/Slave manipulator, power manipulator, lead glass and C.C.T.V. has also been installed. SWFTF will be established in the beginning of 1990's. At this report, radiation shielding door, nondestructive test of the wall, instrumentation system for the utility supply system and cell lighting system are described. (Author)

  7. Alternative Waste Forms for Electro-Chemical Salt Waste

    Energy Technology Data Exchange (ETDEWEB)

    Crum, Jarrod V.; Sundaram, S. K.; Riley, Brian J.; Matyas, Josef; Arreguin, Shelly A.; Vienna, John D.

    2009-10-28

    This study was undertaken to examine alternate crystalline (ceramic/mineral) and glass waste forms for immobilizing spent salt from the Advanced Fuel Cycle Initiative (AFCI) electrochemical separations process. The AFCI is a program sponsored by U.S. Department of Energy (DOE) to develop and demonstrate a process for recycling spent nuclear fuel (SNF). The electrochemical process is a molten salt process for the reprocessing of spent nuclear fuel in an electrorefiner and generates spent salt that is contaminated with alkali, alkaline earths, and lanthanide fission products (FP) that must either be cleaned of fission products or eventually replaced with new salt to maintain separations efficiency. Currently, these spent salts are mixed with zeolite to form sodalite in a glass-bonded waste form. The focus of this study was to investigate alternate waste forms to immobilize spent salt. On a mole basis, the spent salt is dominated by alkali and Cl with minor amounts of alkaline earth and lanthanides. In the study reported here, we made an effort to explore glass systems that are more compatible with Cl and have not been previously considered for use as waste forms. In addition, alternate methods were explored with the hope of finding a way to produce a sodalite that is more accepting of as many FP present in the spent salt as possible. This study was done to investigate two different options: (1) alternate glass families that incorporate increased concentrations of Cl; and (2) alternate methods to produce a mineral waste form.

  8. Estimation of americium in cemented waste block using gamma ray spectrometry

    International Nuclear Information System (INIS)

    A method was developed for the estimation of 241Am present in the cemented waste block which was cylindrical in shape. In such large sample, the attenuation of gamma rays increases with size of the sample and density of the material present. Attenuation correction was incorporated using linear attenuation coefficients of 59.54 keV gamma ray of 241Am. Also in such large samples, error due to the distribution of activity is more. Estimation of 241Am in the cemented sample was carried out by applying corrections for attenuation and for the sample geometry. (author)

  9. Rock segments for reducing the amount of cement used on high-level radioactive waste disposal

    International Nuclear Information System (INIS)

    Methods for constructing tunnels using the minimum quantities of cement-type support materials in high-level radioactive waste disposal facilities have been developed. Research and development concerning the technical aspects of the formation of rock segments using low alkali mortar have been conducted. This study examined the mechanical characteristics of rock segments and backfill materials and analyzed the stability of the drift that is supported by the rock segments and gravel backfill. The results confirmed the technical aspects of the formation of the rock segments and the effectiveness of the planned efforts to further reduce the amount of cement used. (author)

  10. Uranium, Cesium, and Mercury Leaching and Recovery from Cemented Radioactive Wastes in Sulfuric Acid and Iodide Media

    Directory of Open Access Journals (Sweden)

    Nicolas Reynier

    2015-11-01

    Full Text Available The Canadian Nuclear Laboratories (CNL is developing a long-term management strategy for its existing inventory of solid radioactive cemented wastes, which contain uranium, mercury, fission products, and a number of minor elements. The composition of the cemented radioactive waste poses significant impediments to the extraction and recovery of uranium using conventional technology. The goal of this research was to develop an innovative method for uranium, mercury and cesium recovery from surrogate radioactive cemented waste (SRCW. Leaching using sulfuric acid and saline media significantly improves the solubilization of the key elements from the SRCW. Increasing the NaCl concentration from 0.5 to 4 M increases the mercury solubilization from 82% to 96%. The sodium chloride forms a soluble mercury complex when mercury is present as HgO or metallic mercury but not with HgS that is found in 60 °C cured SRCW. Several leaching experiments were done using a sulfuric acid solution with KI to leach SRCW cured at 60 °C and/or aged for 30 months. Solubilization yields are above 97% for Cs and 98% for U and Hg. Leaching using sulfuric acid and KI improves the solubilization of Hg by oxidation of Hg0, as well as HgS, and form a mercury tetraiodide complex. Hg and Cs were selectively removed from the leachate prior to uranium recovery. It was found that U recovery from sulfuric leachate in iodide media using the resin Lewatit TP260 is very efficient. Considering these results, a process including effluent recirculation was applied. Improvements of solubilization due to the recycling of chemical reagents were observed during effluent recirculation.

  11. Pore size distribution, strength, and microstructure of portland cement paste containing metal hydroxide waste

    Energy Technology Data Exchange (ETDEWEB)

    Majid, Z.A.; Mahmud, H.; Shaaban, M.G.

    1996-12-31

    Stabilization/solidification of hazardous wastes is used to convert hazardous metal hydroxide waste sludge into a solid mass with better handling properties. This study investigated the pore size development of ordinary portland cement pastes containing metal hydroxide waste sludge and rice husk ash using mercury intrusion porosimetry. The effects of acre and the addition of rice husk ash on pore size development and strength were studied. It was found that the pore structures of mixes changed significantly with curing acre. The pore size shifted from 1,204 to 324 {angstrom} for 3-day old cement paste, and from 956 to 263 {angstrom} for a 7-day old sample. A reduction in pore size distribution for different curing ages was also observed in the other mixtures. From this limited study, no conclusion could be made as to any correlation between strength development and porosity. 10 refs., 6 figs., 3 tabs.

  12. Preparation techniques for ceramic waste form powder

    International Nuclear Information System (INIS)

    The electrometallurgical treatment of spent nuclear fuels result in a chloride waste salt requiring geologic disposal. Argonne National Laboratory (ANL) is developing ceramic waste forms which can incorporate this waste. Currently, zeolite- or sodalite-glass composites are produced by hot isostatic pressing (HIP) techniques. Powder preparations include dehydration of the raw zeolite powders, hot blending of these zeolite powders and secondary additives. Various approaches are being pursued to achieve adequate mixing, and the resulting powders have been HIPed and characterized for leach resistance, phase equilibria, and physical integrity

  13. Development of thermal conditioning technology for alpha-contaminated wastes: a study on leaching characteristics and long-term safety assessment of simulated waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Yong Chil [Yonsei University, Seoul (Korea); Lee, Sang Hoon; Yoo, Jong Ik; Choi, Yong Cheol [Yonsei University, Seoul (Korea)

    2001-04-01

    Radioactive wastes should be stabilized for safe management during several hundred years. To assess stability of solidified waste forms, mechanical properties and chemical durability of the waste forms should be analyzed. Chemical durability is one of the most important factors in the assessment of waste forms, which could be examined by leaching tests. Various methods in leaching test are suggested by different organizations, but a formal test method in Korea is not ready yet. Therefore, the leaching test method applicable to various constituents is necessary for the safe management of radioactive wastes In this study, leaching behavior and characteristics of components such as solidification materials, heavy metals and radioactive nuclids were analyzed for cement waste form and glassy waste form. 58 refs., 25 figs., 8 tabs. (Author)

  14. Reductive capacity measurement of waste forms for secondary radioactive wastes

    Science.gov (United States)

    Um, Wooyong; Yang, Jung-Seok; Serne, R. Jeffrey; Westsik, Joseph H.

    2015-12-01

    The reductive capacities of dry ingredients and final solid waste forms were measured using both the Cr(VI) and Ce(IV) methods and the results were compared. Blast furnace slag (BFS), sodium sulfide, SnF2, and SnCl2 used as dry ingredients to make various waste forms showed significantly higher reductive capacities compared to other ingredients regardless of which method was used. Although the BFS exhibits appreciable reductive capacity, it requires greater amounts of time to fully react. In almost all cases, the Ce(IV) method yielded larger reductive capacity values than those from the Cr(VI) method and can be used as an upper bound for the reductive capacity of the dry ingredients and waste forms, because the Ce(IV) method subjects the solids to a strong acid (low pH) condition that dissolves much more of the solids. Because the Cr(VI) method relies on a neutral pH condition, the Cr(VI) method can be used to estimate primarily the waste form surface-related and readily dissolvable reductive capacity. However, the Cr(VI) method does not measure the total reductive capacity of the waste form, the long-term reductive capacity afforded by very slowly dissolving solids, or the reductive capacity present in the interior pores and internal locations of the solids.

  15. Reductive Capacity Measurement of Waste Forms for Secondary Radioactive Wastes

    Energy Technology Data Exchange (ETDEWEB)

    Um, Wooyong; Yang, Jungseok; Serne, R. Jeffrey; Westsik, Joseph H.

    2015-09-28

    The reductive capacities of dry ingredients and final solid waste forms were measured using both the Cr(VI) and Ce(IV) methods and the results were compared. Blast furnace slag (BFS), sodium sulfide, SnF2, and SnCl2 used as dry ingredients to make various waste forms showed significantly higher reductive capacities compared to other ingredients regardless of which method was used. Although the BFS exhibits appreciable reductive capacity, it requires greater amounts of time to fully react. In almost all cases, the Ce(IV) method yielded larger reductive capacity values than those from the Cr(VI) method and can be used as an upper bound for the reductive capacity of the dry ingredients and waste forms, because the Ce(IV) method subjects the solids to a strong acid (low pH) condition that dissolves much more of the solids. Because the Cr(VI) method relies on a neutral pH condition, the Cr(VI) method can be used to estimate primarily the waste form surface-related and readily dissolvable reductive capacity. However, the Cr(VI) method does not measure the total reductive capacity of the waste form, the long-term reductive capacity afforded by very slowly dissolving solids, or the reductive capacity present in the interior pores and internal locations of the solids.

  16. Special waste-form lysimeters-arid: Three-year monitoring report

    International Nuclear Information System (INIS)

    Regulations governing the disposal of commercial low-level waste require all liquid waste to be solidified before burial. Most waste must be solidified into a rigid matrix such as cement or plastic to prevent waste consolidation and site slumping after burial. These solidification processes affect the rate at which radionuclides and other solutes are released into the soil. In 1983, a program was initiated at Pacific Northwest Laboratory to study the release of waste from samples of low-level radioactive waste that had been commercially solidified. The primary method used by this program is to bury sample waste forms in field lysimeters and monitor leachate composition from the release and transport of solutes. The lysimeter facility consists of 10 lysimeters, each containing one sample of solidified waste. Five different waste forms are being tested, allowing duplicate samples of each one to be evaluated. The samples were obtained from operating nuclear power plants and are actual waste forms routinely generated at these facilities. All solidification was accomplished by commercial processes. Sample size is a partially filled 210-L drum. All containers were removed prior to burial leaving the bare waste form in contact with the lysimeter soil. 11 refs., 14 figs., 16 tabs

  17. Modelling the Interaction of Low pH Cements and Bentonite. Issues Affecting the Geochemical Evolution of Repositories for Radioactive Waste

    International Nuclear Information System (INIS)

    It is well known that in the hyperalkaline conditions (pH > 12) of standard cement pore fluids, there is potential for deleterious effects upon the host rock and other EBS materials, notably bentonite, in geological repositories for radioactive waste. Low pH cements are beginning to be considered as a potential alternative material that may address some of these concerns. Low pH cement (also known as low heat cement) was developed by the cement industry for use where large masses of cement (e.g. dams) could cause problems regarding heat generated during curing. In low pH cements, the amount of cement is reduced by substitution of materials such as fly ash, blast furnace slag, silica fume, and/or non pozzolanic silica flour. NUMO, Posiva and SKB have defined a pH limit ≤ 11 for cement grout leachates. To attain this pH, blending agents must comprise at least 50 wt % of dry materials. Because low pH cement has little, or no free portlandite, the cement consists predominantly of calcium silicate hydrate (CSH) gel with a Ca/Si ratio ≤ 0.8. In this report we give the results of a preliminary modelling study to investigate the potential impacts of low pH cement water. We compare the evolution of a bentonite sample under the influence of several invading cement porewaters over a pH range from 10 to 13.2. The porewater compositions are taken from published CSH gel leaching experiments and published cement-bentonite modelling studies. The models suggest that the amount of degradation that is likely to be observed when low pH cement water interacts with bentonite is likely to be much less than when OPC water is the invading fluid. Below pH 11 there was not an observable Na montmorillonite dissolution front which would tend to support the pH ≤ 11 target suggested by NUMO, Posiva and SKB. The models used in this study could be improved upon by including a cement component to the model (rather than representing cement as a fixed boundary condition). Solid-solution models

  18. Electrochemical/Pyrometallurgical Waste Stream Processing and Waste Form Fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Steven Frank; Hwan Seo Park; Yung Zun Cho; William Ebert; Brian Riley

    2015-07-01

    This report summarizes treatment and waste form options being evaluated for waste streams resulting from the electrochemical/pyrometallurgical (pyro ) processing of used oxide nuclear fuel. The technologies that are described are South Korean (Republic of Korea – ROK) and United States of America (US) ‘centric’ in the approach to treating pyroprocessing wastes and are based on the decade long collaborations between US and ROK researchers. Some of the general and advanced technologies described in this report will be demonstrated during the Integrated Recycle Test (IRT) to be conducted as a part of the Joint Fuel Cycle Study (JFCS) collaboration between US Department of Energy (DOE) and ROK national laboratories. The JFCS means to specifically address and evaluated the technological, economic, and safe guard issues associated with the treatment of used nuclear fuel by pyroprocessing. The IRT will involve the processing of commercial, used oxide fuel to recover uranium and transuranics. The recovered transuranics will then be fabricated into metallic fuel and irradiated to transmutate, or burn the transuranic elements to shorter lived radionuclides. In addition, the various process streams will be evaluated and tested for fission product removal, electrolytic salt recycle, minimization of actinide loss to waste streams and waste form fabrication and characterization. This report specifically addresses the production and testing of those waste forms to demonstrate their compatibility with treatment options and suitability for disposal.

  19. The Specification of Cement Powders for Waste Encapsulation Processes at Sellafield site

    International Nuclear Information System (INIS)

    Requirements are described for Portland Cement (CEM I), Ground Granulated Blast-furnace Slag (GGBS) and Fly Ash (FA) powders used for the encapsulation of Intermediate Level Radioactive Waste (ILW) in UK, with particular reference to Sellafield site encapsulation processes. Differences between the powders used by the UK nuclear industry and the equivalent British and European cement standards are explained. Research over the last 20 years to respond to changes in the performance of these powders is summarised and options for dealing with potential future changes are discussed. These include the use of special blends of GGBS to achieve the desired flow properties or alternatively poly-carboxylate super-plasticizers to produce grouts with consistent performance using cement powders with a wide range of composition. (authors)

  20. Casting granular ion exchange resins with medium-active waste in cement

    International Nuclear Information System (INIS)

    Medium active waste from nuclear power stations in Sweden is trapped in granular ion exchange resins. The resin is mixed with cement paste and cast in a concrete shell which is cubic and has an edge dimension of 1.2 m. In some cases the ion exchange cement mortar has cracked. The report presents laboratory sutdies of the properties of the ion exchange resin and the mortar. Also the leaching of the moulds has been investigated. It was shown that a mixture with a water cement ratio higher than about 0.5 swells considerably during the first weeks after casting. The diffusion constant for cesium 137 has been determined at 3.10-4 cm2/24-hour period in conjunction with exposure of the mould and mortar to sea water. The Swedish language report has 400 pages with 90 figures and 30 tables. (author)

  1. Review of glass ceramic waste forms

    International Nuclear Information System (INIS)

    Glass ceramics are being considered for the immobilization of nuclear wastes to obtain a waste form with improved properties relative to glasses. Improved impact resistance, decreased thermal expansion, and increased leach resistance are possible. In addition to improved properties, the spontaneous devitrification exhibited in some waste-containing glasses can be avoided by the controlled crystallization after melting in the glass-ceramic process. The majority of the glass-ceramic development for nuclear wastes has been conducted at the Hahn-Meitner Institute (HMI) in Germany. Two of their products, a celsian-based (BaAl3Si2O8) and a fresnoite-based (Ba2TiSi2O8) glass ceramic, have been studied at Pacific Northwest Laboratory (PNL). A basalt-based glass ceramic primarily containing diopsidic augite (CaMgSi2O6) has been developed at PNL. This glass ceramic is of interest since it would be in near equilibrium with a basalt repository. Studies at the Power Reactor and Nuclear Fuel Development Corporation (PNC) in Japan have favored a glass-ceramic product based upon diopside (CaMgSi2O6). Compositions, processing conditions, and product characterization of typical commercial and nuclear waste glass ceramics are discussed. In general, glass-ceramic waste forms can offer improved strength and decreased thermal expansion. Due to typcially large residual glass phases of up to 50%, there may be little improvement in leach resistance

  2. Hydraulic activity of cement mixed with slag from vitrified solid waste incinerator fly ash.

    Science.gov (United States)

    Lin, Kae-Long; Wang, Kuen-Sheng; Tzeng, Bor-Yu; Lin, Chung-Yei

    2003-12-01

    This study investigates the effects of the slag composition on the hydraulic activity in slag blended cement pastes that incorporate synthetic slag prepared by melting CaO-modified municipal solid waste incinerator fly ash. Two types of composition-modified slag were prepared for this study. First, fly ash was mixed with the modifier (CaO) at 5% and 15% (by weight) respectively, resulting in two fly ash mixtures. These mixtures were then melted at 1400 degrees C for 30 minutes and milled to produce two types of slag with different modifier contents, designated as C1-slag and C2-slag. These synthetic slags were blended with ordinary Portland cement at various weight ratios ranging from 10% to 40%. The synthetic slags presented sufficient hydraulic activity, and the heavy metal leaching concentrations all met the EPA's regulatory thresholds. The pore size distribution was determined by mercury intrusion porosimetry, and the results correlated with the compressive strength. The results also indicate that the incorporation of the 10% C1-slag tended to enhance the hydration degree of slag blended cement pastes during the early ages (3-28 days). However, at later ages, no significant difference in hydration degree was observed between ordinary Portland cement pastes and 10% C1-slag blended cement pastes. In the 10% C2-slag case, the trend was similar, but with a more limited enhancement during the early ages (3-28 days). Thus vitrified waste incinerator fly ash is a technically useful additive to cement, reducing the disposal needs for the toxic fly ash. PMID:14986718

  3. Product acceptance of a certified Class C low-level waste form at the West Valley Demonstration Project

    Energy Technology Data Exchange (ETDEWEB)

    Valenti, P.J. [West Valley Nuclear Services Co., Inc., NY (United States); Maestas, E.; Yeazel, J.A. [Dept. of Energy, West Valley, NY (United States). West Valley Project Office; McIntosh, T.W. [Dept. of Energy, Washington, DC (United States). Office of Remedial Action and Waste Technology

    1989-11-01

    The Department of Energy, is charged with the solidification of high-level liquid waste (HLW) remaining from nuclear fuel reprocessing activities, which were conducted at West Valley, New York between 1966 and 1972. One important aspect of the West Valley Demonstration Project`s fully integrated waste program is the treatment and conditioning of low-level wastes which result from processing liquid high-level waste. The treatment takes place in the project`s Integrated Radwaste Treatment System which removes Cesium-137 from the liquid or supernatant phase of the HLW by utilizing an ion exchange technique. The resulting decontaminated and conditioned liquid waste stream is solidified into a Class C low-level cement waste form that meets the waste form criteria specified in NRC 10 CFR 61. The waste matrix is placed in 71-gallon square drums, remotely handled and stored on site until determination of final disposition. This paper discusses the programs in place at West Valley to ensure production of an acceptable cement-based product. Topics include the short and long term test programs to predict product storage and disposal performance, description of the Process Control Plan utilized to control and maintain cement waste form product specifications and finally discuss the operational performance characteristics of the Integrated Radwaste Treatment System. Operational data and product statistics are provided.

  4. Compression and immersion tests and leaching of radionuclides, stable metals, and chelating agents from cement-solidified decontamination waste collected from nuclear power stations

    Energy Technology Data Exchange (ETDEWEB)

    Akers, D.W.; Kraft, N.C.; Mandler, J.W. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

    1994-06-01

    A study was performed for the Nuclear Regulatory Commission (NRC) to evaluate structural stability and leachability of radionuclides, stable metals, and chelating agents from cement-solidified decontamination ion-exchange resin wastes collected from seven commercial boiling water reactors and one pressurized water reactor. The decontamination methods used at the reactors were the Can-Decon, AP/Citrox, Dow NS-1, and LOMI processes. Samples of untreated resin waste and solidified waste forms were subjected to immersion and compressive strength testing. Some waste-form samples were leach-tested using simulated groundwaters and simulated seawater for comparison with the deionized water tests that are normally performed to assess waste-form leachability. This report presents the results of these tests and assesses the effects of the various decontamination methods, waste form formulations, leachant chemical compositions, and pH of the leachant on the structural stability and leachability of the waste forms. Results indicate that releases from intact and degraded waste forms are similar and that the behavior of some radionuclides such as {sup 55}Fe, {sup 60}Co, and {sup 99}Tc were similar. In addition, the leachability indexes are greater than 6.0, which meets the requirement in the NRC`s ``Technical Position on Waste Form,`` Revision 1.

  5. Compression and immersion tests and leaching of radionuclides, stable metals, and chelating agents from cement-solidified decontamination waste collected from nuclear power stations

    International Nuclear Information System (INIS)

    A study was performed for the Nuclear Regulatory Commission (NRC) to evaluate structural stability and leachability of radionuclides, stable metals, and chelating agents from cement-solidified decontamination ion-exchange resin wastes collected from seven commercial boiling water reactors and one pressurized water reactor. The decontamination methods used at the reactors were the Can-Decon, AP/Citrox, Dow NS-1, and LOMI processes. Samples of untreated resin waste and solidified waste forms were subjected to immersion and compressive strength testing. Some waste-form samples were leach-tested using simulated groundwaters and simulated seawater for comparison with the deionized water tests that are normally performed to assess waste-form leachability. This report presents the results of these tests and assesses the effects of the various decontamination methods, waste form formulations, leachant chemical compositions, and pH of the leachant on the structural stability and leachability of the waste forms. Results indicate that releases from intact and degraded waste forms are similar and that the behavior of some radionuclides such as 55Fe, 60Co, and 99Tc were similar. In addition, the leachability indexes are greater than 6.0, which meets the requirement in the NRC's ''Technical Position on Waste Form,'' Revision 1

  6. Characterization of radioactive waste forms. Volume 2

    International Nuclear Information System (INIS)

    This document is the second yearbook for Task 3 of the European Communities 1985-89 programme of research on radioactive waste management and disposal carried out by public organizations and private firms in the Community through cost-sharing contracts with the Commission of the European Communities. The report, in two volumes, describes progress made in 1987 within the field of Task 3: Testing and evaluation of conditioned waste and engineered barriers. The first volume of the report covers Item 3.1 Characterization of low and medium level radioactive waste forms and Item 3.5. Development of test methods for quality assurance. The second volume covers Item 3.2: High-level and alpha waste characterization and Item 3.3: Other engineered barriers. Item 3.4 on the round robin study will be treated in a separate report

  7. Characterization of radioactive waste forms. Volume 1

    International Nuclear Information System (INIS)

    This document is the second yearbook for Task 3 of the European Communities 1985-89 programme of research on radioactive waste management and disposal carried out by public organizations and private firms in the Community through costsharing contracts with the Commission of the European Communities. The report, in two volumes, describes progress made in 1987 within the field of Task 3: Testing and evaluation of conditioned waste and engineered barriers. The first volume of the report covers Item 3.1 Characterization of low and medium-level radioactive waste forms and Item 3.5 Development of test methods for quality assurance. The second volume covers Item 3.2: High-level and alpha waste characterization and Item 3.3: Other engineered barriers. Item 3.4 on the round robin study will be treated in a separate report

  8. Effect of AlF3 Production Waste on the Properties of Hardened Cement Paste

    Directory of Open Access Journals (Sweden)

    Danutė VAIČIUKYNIENĖ

    2012-06-01

    Full Text Available The possibility to use by-product SiO2·nH2O (often called AlF3 production waste in cement casting has been attracting the interest of researchers for many years, although high content of fluorine makes the use of amorphous SiO2 problematic. Finding the way of utilizing waste products is a very important research topic at the moment. In this study AlF3 production waste was investigated as the basic ingredient of a new pozzolanic material. The goal of this study is to investigate the possibilities of using AlF3 production waste, washed in ammonia solution, in cement stone specimens. Chemically treated silica gel additive was proved to reduce the amount of Ca(OH2 and CaCO3 in hardened cement paste samples. Experimental research has revealed that the density in hydrated samples reduces from 2220 kg/m3 to 2030 kg/m3 with the increase of silica gel content from 0 % to 35 %. The compressive strength of samples containing 10 % of silica gel additive increased by 8.04 % compared to the samples without the additive. SiO2 additive used at 10 % and 20 % increased the maximum hydration temperature. In this case, the additive modifies the hydration kinetics.DOI: http://dx.doi.org/10.5755/j01.ms.18.2.1925

  9. WRAP 2A Waste Form Qualification Plan

    Energy Technology Data Exchange (ETDEWEB)

    Burbank, D.A. Jr.

    1993-12-31

    WRAP Module 2A is a facility that will serve to treat retrieved, stored, and newly generated contact-handled mixed low level waste (MLLW) at the Department of Energy`s Hanford site near Richland, Washington. The treatment processes to be used are limited to non-thermal processes, defined as processes operating at a temperature less than 500{degree}F. In addition to waste pretreatment and conditioning processes including sorting, size reduction, and homogenization, the final treatment technologies will consist of immobilization, stabilization, and encapsulation to produce final waste forms that are suitable for disposal in compliance with all applicable regulatory requirements. The wide variety of chemical and physical characteristics exhibited by the WRAP 2A feed streams will necessitate the performance of a comprehensive waste form qualification (WFQ) testing program. The WFQ program will provide the technical basis supporting the process selection and will demonstrate that the selected treatment processes produce final waste forms that will meet all applicable regulatory requirements and performance specifications. This document describes the overall WRAP 2A WFQ program.

  10. WRAP 2A Waste Form Qualification Plan

    International Nuclear Information System (INIS)

    WRAP Module 2A is a facility that will serve to treat retrieved, stored, and newly generated contact-handled mixed low level waste (MLLW) at the Department of Energy's Hanford site near Richland, Washington. The treatment processes to be used are limited to non-thermal processes, defined as processes operating at a temperature less than 500 degree F. In addition to waste pretreatment and conditioning processes including sorting, size reduction, and homogenization, the final treatment technologies will consist of immobilization, stabilization, and encapsulation to produce final waste forms that are suitable for disposal in compliance with all applicable regulatory requirements. The wide variety of chemical and physical characteristics exhibited by the WRAP 2A feed streams will necessitate the performance of a comprehensive waste form qualification (WFQ) testing program. The WFQ program will provide the technical basis supporting the process selection and will demonstrate that the selected treatment processes produce final waste forms that will meet all applicable regulatory requirements and performance specifications. This document describes the overall WRAP 2A WFQ program

  11. Leaching characteristics of heavy metals during cement stabilization of fly ash from municipal solid waste incinerators

    Institute of Scientific and Technical Information of China (English)

    Shunwen LIANG; Jianguo JIANG; Yan ZHANG; Xin XU

    2008-01-01

    The leaching characteristics of heavy metals in products of cement stabilization of fly ash from a muni-cipal solid waste incinerator were investigated in this paper. The stabilization of heavy metals such as Cd, Pb, Cu, and Zn in fly ash from such incinerators was exam-ined through the national standard method in China based on the following-factors: additive quantity of cement and Na2S, curing time, and pH of leaching liquor. The results showed that as more additives were used, less heavy metals were leached except for Pb, which is sensitive to pH of the leachate, and the worse effect was observed for Cd. The mass ratio of cement to fly ash=10% is the most appropriate parameter according to the national standard method. When the hydration of cement was basically finished, stabilization of heavy metals did not vary after curing for 1 d. The mixtures of cement and fly ash had excellent adaptability to environmental pH. The pH of leachate was maintained at 7 when pH of leaching liquor varied from 3 to 11.

  12. On the Utilization of Pozzolanic Wastes as an Alternative Resource of Cement

    Directory of Open Access Journals (Sweden)

    Md. Rezaul Karim

    2014-12-01

    Full Text Available Recently, as a supplement of cement, the utilization of pozzolanic materials in cement and concrete manufacturing has increased significantly. This study investigates the scope to use pozzolanic wastes (slag, palm oil fuel ash and rice husk ash as an alkali activated binder (AAB that can be used as an alternative to cement. To activate these materials, sodium hydroxide solution was used at 1.0, 2.5 and 5.0 molar concentration added into the mortar, separately. The required solution was used to maintain the flow of mortar at 110% ± 5%. The consistency and setting time of the AAB-paste were determined. Mortar was tested for its flow, compressive strength, porosity, water absorption and thermal resistance (heating at 700 °C and investigated by scanning electron microscopy. The experimental results reveal that AAB-mortar exhibits less flow than that of ordinary Portland cement (OPC. Surprisingly, AAB-mortars (with 2.5 molar solution achieved a compressive strength of 34.3 MPa at 28 days, while OPC shows that of 43.9 MPa under the same conditions. Although water absorption and porosity of the AAB-mortar are slightly high, it shows excellent thermal resistance compared to OPC. Therefore, based on the test results, it can be concluded that in the presence of a chemical activator, the aforementioned pozzolans can be used as an alternative material for cement.

  13. Utilization of ground waste seashells in cement mortars for masonry and plastering.

    Science.gov (United States)

    Lertwattanaruk, Pusit; Makul, Natt; Siripattarapravat, Chalothorn

    2012-11-30

    In this research, four types of waste seashells, including short-necked clam, green mussel, oyster, and cockle, were investigated experimentally to develop a cement product for masonry and plastering. The parameters studied included water demand, setting time, compressive strength, drying shrinkage and thermal conductivity of the mortars. These properties were compared with those of a control mortar that was made of a conventional Portland cement. The main parameter of this study was the proportion of ground seashells used as cement replacement (5%, 10%, 15%, or 20% by weight). Incorporation of ground seashells resulted in reduced water demand and extended setting times of the mortars, which are advantages for rendering and plastering in hot climates. All mortars containing ground seashells yielded adequate strength, less shrinkage with drying and lower thermal conductivity compared to the conventional cement. The results indicate that ground seashells can be applied as a cement replacement in mortar mixes and may improve the workability of rendering and plastering mortar. PMID:22841935

  14. RECYCLED WASTE-BASED CEMENT COMPOSITE PATCH MATERIALS FOR RAPID/PERMANENT ROAD RESTORATION.

    Energy Technology Data Exchange (ETDEWEB)

    SUGAMA,T.

    2001-07-31

    Over the past year, KeySpan Energy sponsored a research program at Brookhaven National Laboratory (BNL) aimed at recycling boiler ash (BA) and waste water treatment sludge (WWTS) byproducts generated from Keyspan's power stations into potentially useful materials, and at reducing concurrent costs for their disposal. Also, KeySpan has an interest in developing strategies to explicitly integrate industrial ecology and green chemistry. From our collaborative efforts with Keyspan (Diane Blankenhom Project Manager, and Kenneth Yager), we succeeded in recycling them into two viable products; Pb-exchange adsorbents (PEAs), and high-performance cements (HpCs). These products were made from chemically bonded cement and ceramic (CBC) materials that were synthesized through two-step chemical reaction pathways, acid-base and hydration. Using this synthesis technology, both the WWTS and BA served in acting as solid base reactants, and sodium polyphosphate, [-(-NaPO{sub 3}-)-{sub n}], known as an intermediator of fertilizer, was employed as the acid solution reactant. In addition, two commercial cement additives, Secar No. 51 calcium aluminate cement (CAC) and Type I calcium silicate cement (CSC), were used to improve mechanical behavior and to promote the rate of acid-base reaction of the CBC materials.

  15. Mineralogy and microstructure of two Mexican Portland cements for the confinement of radioactive waste

    International Nuclear Information System (INIS)

    The cementitious materials are involved in the different stages of radioactive waste management because they are used for the waste immobilization in the container, as well as filling in the spaces between containers vaults and also as engineering barrier and construction material in civil construction site. Therefore, is necessary to have a study of commercial cement available nationwide involving solid timely analysis in order to identify which phases are responsible for confinement of radionuclides, if considered the most reactive phase -CSH- or called secondary phases. In this research the hydration products of cement are presented as well as its importance in the nuclear industry. The analysis and observation of the cement clinker and the hydration products on the manufactured pulps with two commercial cements resistant to sulphates was realized using the observation technique of solid X-ray diffraction and nuclear analytic techniques of Moessbauer spectroscopy and X-Ray Fluorescence. The results show the presence of calcium silicate hydrates in the amorphous phase and the presence of ettringite crystals and portlandite sheets is appreciated. The abundant iron phase called tetra calcium ferro aluminate has been identified by Moessbauer spectroscopy. (Author)

  16. Mineralogy and chemistry of cement paste in borehole radioactive waste repository

    International Nuclear Information System (INIS)

    Results of chemical characterization of cement paste samples after irradiation and immersion in salt solutions are presented. This is part of a research on cement paste behavior aiming at investigating the durability of cementitious materials in the environment of repositories for radioactive waste. Portland cement paste is intended to be used as a backfill in a deep borehole for disposal of sealed radiation sources which concept is under development. The service life of the engineered barrier materials plays an important role in the long term safety of such facilities. Accelerated tests in laboratory are being used to evaluate the performance of cement paste under the temperature expected at some hundred meters below grade, under exposure to the radiation emitted by the sources, and under the attack of aggressive chemicals dissolved in the groundwater, during the millennia necessary for the decay of the most active and long-lived radionuclides present in the waste. ICP-OES, Ion chromatography, X-ray diffraction, SEM and TGA are some techniques being employed in this research project. (author)

  17. Dissolution of tailored ceramic nuclear waste forms

    International Nuclear Information System (INIS)

    Dissolution experiments on polyphase, high alumina tailored ceramic nuclear waste forms developed for the chemical immobilization of Savannah River Plant nuclear waste are described. Three forms of leach tests have been adopted; bulk samples conforming to the Materials Characterization Center Static Leach Test (MCC-1), a powdered sample leach test, and a leach test performed on transmission electron microscope thin foil samples. From analysis of these tests the crystalline phases that preferentially dissolve on leaching and the product phases formed are identified and related to the tailoring and processing schemes used in forming the ceramics. The thin foil sample leaching enables the role of intergranular amorphous phases as short-circuit leaching paths in polyphase ceramics to be investigated

  18. Cement matrix for immobilisation of spent anionic resins in borate form arising from nuclear power plants

    International Nuclear Information System (INIS)

    In water cooled reactors boron is added as boric acid to control nuclear reactor power levels. The boric acid concentration in coolant/moderator water, is controlled by using strongly basic anionic resins in borate (H2BO3-) form. The spent anionic resins in borate form contain 131Iodine, 99Technitium and 137Cesium activities. Direct immobilisation of anionic resins in borate form in Ordinary Portland Cement (OPC) and Slag Cement was investigated using vermiculite, bentonite, calcium oxide and silica as admixtures. The cumulative fraction of 137Cesium leached and 137Cesium leach rate for slag cement matrix were 0.029 and 0.00064 g.cm2.d-1 respectively for 95 days of leaching. The volume reduction factor achieved by direct immobilisation of anionic resins in borate form was 0.48. Immobilisation of pyrolysis residues from these resins in OPC matrix was also studied. Leaching of matrix blocks was carried out for 180 days in DM water to optimise the matrix formulation. The cumulative fraction of 137Cesium leached and 137Cesium leach rate were 0.076 and 0.00054 respectively for 180 days leaching. The volume reduction factor achieved by immobilisation of pyrolysis residues was 2.4. OPC is non compatible to cationic resins loaded with alkali in absence of specific admixtures. Hence cationic resins loaded with alkali and anionic resins in borate form can not be immobilised together. (author)

  19. Exergetic life cycle assessment of cement production process with waste heat power generation

    International Nuclear Information System (INIS)

    Highlights: • Exergetic life cycle assessment was performed for the cement production process. • Each system’s efficiency before and after waste heat power generation was analyzed. • The waste heat power generation improved the efficiency of each production system. • It provided technical support for the implementation of energy-saving schemes. - Abstract: The cement industry is an industry that consumes a considerable quantity of resources and energy and has a very large influence on the efficient use of global resources and energy. In this study, exergetic life cycle assessment is performed for the cement production process, and the energy efficiency and exergy efficiency of each system before and after waste heat power generation is investigated. The study indicates that, before carrying out a waste heat power generation project, the objective energy efficiencies of the raw material preparation system, pulverized coal preparation system and rotary kiln system are 39.4%, 10.8% and 50.2%, respectively, and the objective exergy efficiencies are 4.5%, 1.4% and 33.7%, respectively; after carrying out a waste heat power generation project, the objective energy efficiencies are 45.8%, 15.5% and 55.1%, respectively, and the objective exergy efficiencies are 7.8%, 2.8% and 38.1%, respectively. The waste heat power generation project can recover 3.7% of the total input exergy of a rotary kiln system and improve the objective exergy efficiencies of the above three systems. The study can identify degree of resource and energy utilization and the energy-saving effect of a waste heat power generation project on each system, and provide technical support for managers in the implementation of energy-saving schemes

  20. The solubility limited source term for cement-conditioned wastes: A status report

    International Nuclear Information System (INIS)

    An important function of the cement backfill in a nuclear waste repository is to react with aqueous waste species and reduce their solubility. However, to quantify backfill performance it is first necessary to prove the existence and establish the nature of the chemical solubility controls. This can be done by characterizing the solubility-limiting phases, determining their solubility and thermodynamic functions, and assessing their stability and persistence and solubility during backfill degradation. Much of the necessary data must be acquired experimentally. The title paper describes briefly the progress of experimental work on selected species including nickel, chromium(III,VI), tin(IV), molybdenum(VI), uranium(VI), Ce(III,IV), thorium, actinide simulants (III,IV) and chloride. Data needs are assessed and although much experimental work remains to be done, methodologies have been developed which will expedite progress. The expectation of a more quantitative performance assessment of cement barriers is, therefore, attainable

  1. Concentration Limits in the Cement Based Swiss Repository for Long-lived, Intermediate-level Radioactive Wastes (LMA)

    Energy Technology Data Exchange (ETDEWEB)

    Berner, Urs

    1999-12-01

    The Swiss repository concept for long-lived, intermediate-level radioactive wastes (LMA), in Swiss terminology) foresees cylindrical concrete silos surrounded by a ring of granulated bentonite to deposit the waste. As one of the possible options and similar to the repository for high level wastes, the silos will be located in a deep crystalline host rock. Solidified with concrete in steel drums, the waste is stacked into a silo and the silo is then backfilled with a porous mortar. To characterize the release of radionuclides from the repository, the safety assessment considers first the dissolution into the pore water of the concrete, and then diffusion through the outer bentonite ring into the deep crystalline groundwater. For 19 safety relevant radionuclides (isotopes of U, Th, Pa, Np, Pu, Am, Ni, Zr, Mo, Nb, Se, Sr, Ra, Tc, Sn, I, C, Cs, Cl) the report recommends maximum elemental concentrations to be expected in the cement pore water of the particularly considered repository. These limits will form the parameter base for subsequent release model chains. Concentration limits in a geochemical environment are usually obtained from thermodynamic equilibrium calculations performed with geochemical speciation codes. However, earlier studies revealed that this procedure does not always lead to reliable results. Main reasons for this are the complexity of the systems considered, as well as the lacking completeness of, and the uncertainty associated with the thermodynamic data. To improve the recommended maximum concentrations for a distinct repository design, this work includes additional design- and system-dependent criteria. The following processes, inventories and properties are considered in particular: a) recent experimental investigations, particularly from cement systems, b) thermodynamic model calculations when reliable data are available, c) total inventories of radionuclides, d) sorption- and co-precipitation processes, e) dilution with stable isotopes, f

  2. Performance Characteristics of Waste Glass Powder Substituting Portland Cement in Mortar Mixtures

    OpenAIRE

    Kara, P; Csetényi, L; Borosnyói, A

    2014-01-01

    In several countries, waste glass causes environmental concerns as quantities stockpiled exceed recycling in the packaging stream. Being amorphous and having relatively high silicium and calcium contents, glass is pozzolanic or even cementitious, when finely ground. Reducing particle sizes typically to less than 100 µm may give control over the alkali-silica reaction in concrete, therefore making this material a possible substitute to Portland cement. Such use may moderate the problem of dump...

  3. Preparation of magnesium phosphate cement by recycling the product of thermal transformation of asbestos containing wastes

    OpenAIRE

    Viani, A; Gualtieri, A.F.

    2014-01-01

    Asbestos containing wastes have been employed for the first time in the formulation of magnesium phosphate cements. Two samples were mixed with magnesium carbonate and calcined at 1100 and 1300 C. Under these conditions, complete destruction of asbestos minerals is known to occur. The product, containing MgO, after reaction with water-soluble potassium di-hydrogen phosphate, led to the formation of hydrated phases at room temperature. Crystalline and amorphous reaction products were detected,...

  4. Preparation of magnesium phosphate cement by recycling the product of thermal transformation of asbestos containing wastes

    Czech Academy of Sciences Publication Activity Database

    Viani, Alberto; Gualtieri, A.F.

    2014-01-01

    Roč. 58, April (2014), s. 56-66. ISSN 0008-8846 R&D Projects: GA MŠk(CZ) LO1219 Institutional support: RVO:68378297 Keywords : cement-asbestos * chemically bonded ceramics * waste management * X-ray diffraction * amorphous material Subject RIV: JN - Civil Engineering Impact factor: 2.864, year: 2014 http://www.sciencedirect.com/science/article/pii/S000888461400012X

  5. Evaluation of batch mixing equipment for producing cement-based radioactive waste hosts

    International Nuclear Information System (INIS)

    This report summarizes the general criteria needed to evaluate processing equipment for producing grouts to serve as radioactive waste hosts. An equipment evaluation procedure is also defined by establishing a systematic approach to numerical scoring of equipment performance against specific selection criteria. As an example, this procedure is then used to evaluate cement-mixing equipment for the proposed Process Experimental Pilot Plant. 2 references, 3 figures, 2 tables

  6. Acoustic emission monitoring of cement-based structures immobilising radioactive waste

    International Nuclear Information System (INIS)

    The long term performance of cementitious structures immobilising radioactive waste can be affected by physical and chemical processes within the encapsulating materials such as formation of new phases (e.g., vaterite, brucite), degradation of cement phases (e.g., CSH gel, portlandite), degradation of some waste components (e.g., organics), corrosion of metallic constituents (aluminium, magnesium), gas emission, further hydration etc. The corrosion of metals in the high pH cementitious environment is of especial concern as it can potentially cause wasteform cracking. One of the perspective non-destructive methods used to monitor and assess the mechanical properties of materials and structures is based on an acoustic emission (AE) technique. In this study an AE non-destructive technique was used to evaluate the mechanical performance of cementitious structures with encapsulated metallic waste such as aluminium. AE signals generated as a result of aluminium corrosion in a small-size blast furnace slag (BFS)/ordinary Portland cement (OPC) sample were detected, recorded and analysed. A procedure for AE data analysis including conventional parameter-based AE approach and signal-based analysis was applied and demonstrated to provide information on the aluminium corrosion process and its impact on the mechanical performance of the encapsulating cement matrix. (authors)

  7. Some aspects about the Portland cement utilization as a matrix for radioactive waste immobilization

    International Nuclear Information System (INIS)

    More recently, the environmental policy has concentrated the focus on the study of the waste disposal environmental impact. Since Portland cement is commonly used as a matrix in the low-and intermediate-level radioactive waste immobilization, in the present work, some relationships between the structure and properties of matrix, based on available concrete technology information, has been established by using the multi-level approach analysis. The relationships were developed based on hydrating reactions, the microstructure models, the pore system. It have been verified that: a) CSH gel is responsible for the cementing action and for the strength; b) it seems that the capillary porosity is the strength limiting; c) the permeability, regarded in terms of gel porosity and reduced capillary porosity of the hardened cement paste, may not be a decisive factor for the radionuclide release; d) the shrinkage and the swelling induced cracks can enhance the diffusion mechanism for the cracks increase the exposed surface. The durability of the waste disposal matrix concerning chemical attack in the acidic environment has been considered. (author)

  8. Cementation of secondary wastes generated from carbonisation of spent organic ion exchange resins from nuclear power plants

    International Nuclear Information System (INIS)

    The spent IX resins containing radioactive fission and activation products from power reactors are highly active solid wastes generated during operations of nuclear reactors. Process for carbonization of IX resins to achieve weight and volume reduction has been optimized on 50 dm3/batch pilot test rig. The process generates carbonaceous residue, organic liquid condensates (predominantly styrene) and aqueous alkaline scrubber solutions as secondary wastes. The report discusses laboratory tests on leaching of 137Cs from cement matrix incorporating carbonaceous residues and extrapolation of results to 200 liter matrix block. The cumulative fraction of 137Cs leached from 200 liter cement matrix was estimated to be 0.0021 in 200 days and 0.0418 over a period of 30 years. Incorporation of organic liquid condensates into cement matrix has been tried out successfully. Thus two types of secondary wastes generated during carbonization of spent IX resins can be immobilized in cement matrix. (author)

  9. Experiment close out of lysimeter field testing of low-level radioactive waste forms

    International Nuclear Information System (INIS)

    The Field Lysimeter Investigations: Low-Level Waste Data Base Development Program is obtaining information on the performance of radioactive waste forms. These experiments were recently shut down and the contents of the lysimeters have been examined in accordance with a detailed waste form and soil sampling plan. Ion-exchange resins from a commercial nuclear power station were solidified into waste forms using portland cement and vinyl ester-styrene. These waste forms were tested to (a) obtain information on performance of waste forms in typical disposal environments, (b) compare field results with bench leach studies, (c) develop a low-level waste data base for use in performance assessment source term calculations, and (d) apply the DUST computer code to compare predicted cumulative release to actual field data. The program, funded by the Nuclear Regulatory Commission (NRC), includes observed radio nuclide releases from waste forms in field lysimeters at two test sites over 10 years of successful operation. The purpose of this paper is to present the results of the examination of waste forms and soils of the two lysimeter arrays after shut down. During this examination, the waste forms were characterized after removal from the lysimeters and the results compared to the findings of the original characterizations. Vertical soil cores were taken from the soil columns and analyzed with radiochemistry to define movement of radionuclides in the soils after release from the waste forms. A comparison is made of the DUST and BLT code predictions of releases and movement, using recently developed partition coefficients and leachate measurements, to actual radio nuclide movement through the soil columns as determined from these core analyses

  10. Management of cesium loaded AMP- Part I preparation of 137Cesium concentrate and cementation of secondary wastes

    International Nuclear Information System (INIS)

    Separation of 137cesium from High Level Waste can be achieved by use of composite-AMP, an engineered form of Ammonium Molybdo-Phosphate(AMP). Direct vitrification of cesium loaded composite AMP in borosilicate glass matrix leads to separation of water soluble molybdate phase. A proposed process describes two different routes of selective separation of molybdates and phosphate to obtain solutions of cesium concentrates. Elution of 137Cesium from composite-AMP by decomposing it under flow conditions using saturated barium hydroxide was investigated. This method leaves molybdate and phosphate embedded in the column but only 70% of total cesium loaded on column could be eluted. Alternatively composite-AMP was dissolved in sodium hydroxide and precipitation of barium molybdate-phosphate from the resultant solution, using barium nitrate was investigated by batch methods. The precipitation technique gave over 99.9% of 137Cesium activity in solutions, free of molybdates and phosphates, which is ideally suited for immobilization in borosilicate glass matrix. Detailed studies were carried out to immobilize secondary waste of 137Cesium contaminated barium molybdate-phosphate precipitates in the slag cement matrix using vermiculite and bentonite as admixtures. The cumulative fraction of 137Cs leached from the cement matrix blocks was 0.05 in 140 days while the 137Cs leach rate was 0.001 gm/cm2/d. (author)

  11. Experimental Limitations Regarding the Formation and Characterization of Uranium-Mineral Phases in Concrete Waste Forms

    International Nuclear Information System (INIS)

    Predicting the long-term fate of low-level radioactive waste forms requires understanding how the radionuclides interact with the waste form. Concrete encasement is one method being considered for containment of low-level radioactive wastes. The necessary data to conduct an accurate performance assessment of such a waste form requires understanding the behavior and interactions of the radionuclides with the concrete matrix. The formation of uranium mineral phases has been investigated in simulated concrete pore fluids and Ordinary Portland Cement/Pulverized Fuel Ash (fly ash) concrete waste forms. X-Ray diffraction analyses of uranium precipitates from concrete pore fluids suggest diuranate salts, uranium-oxyhydroxides, and ?silicates as solubility limiting phases. Scanning electron microscopy ? energy dispersive spectroscopic analyses of uranium-spiked concrete suggests that under conditions both under-saturated and over-saturated with respect to the formation of uranium mineral phases, uranyl-oxyhydroxide phases precipitate within the initial two weeks. Subsequently, uranyl-silicate phases form after approximately one month and uranyl-phosphate phases provide a significant contribution to the long-term control over uranium in concrete waste forms after two months. This investigation demonstrates the importance of investigating the solubility of complex contaminants such as uranium in the complete matrix (i.e. concrete matrix versus pore fluids) and suggests the importance of secondary uranium mineral phases in the long-term retention within concrete waste forms

  12. Waste Form Features, Events, and Processes

    International Nuclear Information System (INIS)

    The purpose of this report is to evaluate and document the inclusion or exclusion of the waste form features, events and processes (FEPs) with respect to modeling used to support the Total System Performance Assessment for License Application (TSPA-LA). A screening decision, either Included or Excluded, is given for each FEP along with the technical bases for screening decisions. This information is required by the Nuclear Regulatory Commission (NRC) in 10 CFR 63.114 (d, e, and f) [DIRS 156605]. The FEPs addressed in this report deal with the issues related to the degradation and potential failure of the waste form and the migration of the waste form colloids. For included FEPs, this analysis summarizes the implementation of the FEP in TSPA-LA, (i.e., how the FEP is included). For excluded FEPs, this analysis provides the technical bases for exclusion from TSPA-LA (i.e., why the FEP is excluded). This revision addresses the TSPA-LA FEP list (DTN: MO0407SEPFEPLA.000 [DIRS 170760]). The primary purpose of this report is to identify and document the analyses and resolution of the features, events, and processes (FEPs) associated with the waste form performance in the repository. Forty FEPs were identified that are associated with the waste form performance. This report has been prepared to document the screening methodology used in the process of FEP inclusion and exclusion. The analyses documented in this report are for the license application (LA) base case design (BSC 2004 [DIRS 168489]). In this design, a drip shield is placed over the waste package and no backfill is placed over the drip shield (BSC 2004 [DIRS 168489]). Each FEP may include one or more specific issues that are collectively described by a FEP name and a FEP description. The FEP description may encompass a single feature, process or event, or a few closely related or coupled processes if the entire FEP can be addressed by a single specific screening argument or TSPA-LA disposition. The FEPs are

  13. Waste Form Features, Events, and Processes

    Energy Technology Data Exchange (ETDEWEB)

    R. Schreiner

    2004-10-27

    The purpose of this report is to evaluate and document the inclusion or exclusion of the waste form features, events and processes (FEPs) with respect to modeling used to support the Total System Performance Assessment for License Application (TSPA-LA). A screening decision, either Included or Excluded, is given for each FEP along with the technical bases for screening decisions. This information is required by the Nuclear Regulatory Commission (NRC) in 10 CFR 63.114 (d, e, and f) [DIRS 156605]. The FEPs addressed in this report deal with the issues related to the degradation and potential failure of the waste form and the migration of the waste form colloids. For included FEPs, this analysis summarizes the implementation of the FEP in TSPA-LA, (i.e., how the FEP is included). For excluded FEPs, this analysis provides the technical bases for exclusion from TSPA-LA (i.e., why the FEP is excluded). This revision addresses the TSPA-LA FEP list (DTN: MO0407SEPFEPLA.000 [DIRS 170760]). The primary purpose of this report is to identify and document the analyses and resolution of the features, events, and processes (FEPs) associated with the waste form performance in the repository. Forty FEPs were identified that are associated with the waste form performance. This report has been prepared to document the screening methodology used in the process of FEP inclusion and exclusion. The analyses documented in this report are for the license application (LA) base case design (BSC 2004 [DIRS 168489]). In this design, a drip shield is placed over the waste package and no backfill is placed over the drip shield (BSC 2004 [DIRS 168489]). Each FEP may include one or more specific issues that are collectively described by a FEP name and a FEP description. The FEP description may encompass a single feature, process or event, or a few closely related or coupled processes if the entire FEP can be addressed by a single specific screening argument or TSPA-LA disposition. The FEPs are

  14. Corrosion of steel drums containing cemented ion-exchange resins as intermediate level nuclear waste

    International Nuclear Information System (INIS)

    Highlights: • There are no works related to the corrosion of drums containing radioactive waste. • Chloride induces high corrosion rate and after 1 year it drops abruptly. • Decrease in the corrosion rate is due to the lack of water to sustain the process. • Cementated ion-exchange resins do not pose risks of corrosion of the steel drums. -- Abstract: Exhausted ion-exchange resins used in nuclear reactors are immobilized by cementation before being stored. They are contained in steel drums that may undergo internal corrosion depending on the presence of certain contaminants. The objective of this work is to evaluate the corrosion susceptibility of steel drums in contact with cemented ion-exchange resins with different aggressive species. The corrosion potential and the corrosion rate of the steel, and the electrical resistivity of the matrix were monitored for 900 days. Results show that the cementation of ion-exchange resins seems not to pose special risks regarding the corrosion of the steel drums

  15. Corrosion of steel drums containing cemented ion-exchange resins as intermediate level nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    Duffó, G.S. [Departamento de Materiales, Comisión Nacional de Energía Atómica, Av. Gral. Paz 1499, 1650 Buenos Aires (Argentina); Universidad Nacional de San Martín, Av. Gral. Paz 1499, 1650 Buenos Aires (Argentina); Consejo Nacional de Investigaciones Científicas y Tecnológicas – CONICET, Av. Gral. Paz 1499, 1650 Buenos Aires (Argentina); Farina, S.B., E-mail: farina@cnea.gov.ar [Departamento de Materiales, Comisión Nacional de Energía Atómica, Av. Gral. Paz 1499, 1650 Buenos Aires (Argentina); Universidad Nacional de San Martín, Av. Gral. Paz 1499, 1650 Buenos Aires (Argentina); Consejo Nacional de Investigaciones Científicas y Tecnológicas – CONICET, Av. Gral. Paz 1499, 1650 Buenos Aires (Argentina); Schulz, F.M. [Consejo Nacional de Investigaciones Científicas y Tecnológicas – CONICET, Av. Gral. Paz 1499, 1650 Buenos Aires (Argentina)

    2013-07-15

    Highlights: • There are no works related to the corrosion of drums containing radioactive waste. • Chloride induces high corrosion rate and after 1 year it drops abruptly. • Decrease in the corrosion rate is due to the lack of water to sustain the process. • Cementated ion-exchange resins do not pose risks of corrosion of the steel drums. -- Abstract: Exhausted ion-exchange resins used in nuclear reactors are immobilized by cementation before being stored. They are contained in steel drums that may undergo internal corrosion depending on the presence of certain contaminants. The objective of this work is to evaluate the corrosion susceptibility of steel drums in contact with cemented ion-exchange resins with different aggressive species. The corrosion potential and the corrosion rate of the steel, and the electrical resistivity of the matrix were monitored for 900 days. Results show that the cementation of ion-exchange resins seems not to pose special risks regarding the corrosion of the steel drums.

  16. Durability of cement paste as engineered barrier in borehole waste repository

    International Nuclear Information System (INIS)

    The Radioactive Waste Management Laboratory of the Nuclear and Energy Research Institute, in Sao Paulo, Brazil, is developing the concept of a repository for disposal of sealed radioactive sources. The concept is a deep borehole drilled a few hundred meters below surface in a granite batholith. Portland cement paste is the material intended to backfill the annular space between the steel casing and the geological formation around the borehole. The hardened cement paste is intended to function as barrier against water flow between the different strata of the geological setting crossed by the borehole and also as an additional barrier against inflow of water and migration of the radionuclides present in the sealed sources. A service life of thousands of years is a necessary characteristic of the engineered barriers in this repository because many sealed sources are long-lived. The durability of cementitious materials is known only for short periods and must be evaluated for long periods. This research aims at evaluating the durability of Portland cement paste under the repository conditions foreseen in that disposal facility, by accelerated tests in laboratory. In this paper we present results of mechanical strength, mass, and volume variations of cement samples under irradiation, high temperature and immersion in saline solutions, as a function of time. (author)

  17. Studies of cement grouts and grouting techniques for sealing a nuclear fuel waste disposal vault

    International Nuclear Information System (INIS)

    This paper investigates a cement-based grout (90% Type 50, 10% silica fume, 0.4< water-to-cement ratio, w/c<0.6) that has been used in field trials to evaluate suitable grouts and grouting techniques that could be used for sealing a nuclear fuel waste disposal vault mined deep in granite. The authors describe laboratory studies carried out to determine the following grout properties: hydraulic conductivity (k); resistance to piping and erosion during setting; influence of group on the pH and chemical composition of water permeating grouted rock; and the ability of the grout to self-seal after fracturing. Laboratory tests have confirmed the low intrinsic k of these cement mixtures (10-14 m/s). Using a specially developed cone-in-cone apparatus, the authors have studied the effect of fracture dilation and temperature changes on the k of thin films of cement. If fractured, the grout has an ability to self-seal and the rate of self-sealing increases with increasing temperature. Test results are reported

  18. Durability of cement stabilized low-level wastes

    International Nuclear Information System (INIS)

    Cementitious materials containing high proportions of slag and fly ash have been tested for suitability to immobilize simulated alkaline and carbonated off-gas waste solutions after vitrification of low- level tank wastes stored at Hanford. To assess their performance, long-term durability was assessed by measuring stability of compressive strength and weight during leaching and exposure to sulfate and carbonate solutions. The important parameter controlling the durability is pore structure, because it affects both compressive strength and susceptibility to different kinds of chemical attack. Impedance spectroscopy was utilized to assess the connectivity of the pore system at early ages. Mercury intrusion porosimetry (MIP) and SEM were utilized to assess development of porosity at later ages. Phase alterations in the matrix exposed to aging and leaching in different media were followed using XRD. Mixtures were resistant to deterioration during immersion in solutions containing high concentrations of sulfate or carbonate ions. Mixtures were also resistant to leaching. These results are consistent with microstructural observations, which showed development of a finer pore structure and reduction in diffusivity over days or months of hydration

  19. ALKALI-ACTIVATED CEMENT MORTARS CONTAINING RECYCLED CLAY-BASED CONSTRUCTION AND DEMOLITION WASTE

    Directory of Open Access Journals (Sweden)

    F. Puertas

    2015-09-01

    Full Text Available The use of clay-based waste as an aggregate for concrete production is an amply studied procedure. Nonetheless, research on the use of this recycled aggregate to prepare alkaline cement mortars and concretes has yet to be forthcoming. The present study aimed to determine: the behaviour of this waste as a pozzolan in OPC systems, the mechanical strength in OPC, alkali-activated slag (AAS and fly ash (AAFA mortars and the effect of partial replacement of the slag and ash themselves with ground fractions of the waste. The pozzolanic behaviour of clay-based waste was confirmed. Replacing up to 20 % of siliceous aggregate with waste aggregate in OPC mortars induced a decline in 7 day strength (around 23 wt. %. The behaviour of waste aggregate in AAMs mortars, in turn, was observed to depend on the nature of the aluminosilicate and the replacement ratio used. When 20 % of siliceous aggregate was replaced by waste aggregate in AAS mortars, the 7 day strength values remained the same (40 MPa. In AAFA mortars, waste was found to effectively replace both the fly ash and the aggregate. The highest strength for AAFA mortars was observed when they were prepared with both a 50 % replacement ratio for the ash and a 20 % ratio for the aggregate.

  20. Testing and evaluation of alternative process systems for immobilizing radioactive mixed particulate waste in cement

    International Nuclear Information System (INIS)

    Radioactive and Hazardous Mixed Wastes have accumulated at the Department of Energy (DOE) Hanford Site in south-central Washington State. Ongoing operations and planned facilities at Hanford will also contribute to this waste stream. To meet the Resource Conservation and Recovery Act (RCRA) Land Disposal Restrictions most of this waste will need to be treated to permit disposal. In general this treatment will need to include stabilization/solidification either as a sole method or as part of a treatment train. A planned DOE facility, the Waste Receiving and Processing (WRAP) Module 2A, is scoped to provide this required treatment for containerized contact-handled (CH), mixed low-level waste (MLLW) at Hanford. An engineering development program has been conducted by Westinghouse Hanford Company (WHC) to select the best system for utilizing a cement based process in WRAP Module 2A. Three mixing processes were developed for analysis and testing; in-drum mixing, continuous mixing, and batch mixing. Some full scale tests were conducted and 55 gallon drums of solidified product were produced. These drums were core sampled and examined to evaluate mixing effectiveness. Total solids loading and the order of addition of waste and binder constituents were also varied. The highest confidence approach to meet the WRAP Module 2A waste immobilization system needs appears to be the out-of-drum batch mixing concept. This system is believed to offer the most flexibility and efficiency, given the highly variable and troublesome waste streams feeding the facility

  1. Waste Cellulose from Tetra Pak Packages as Reinforcement of Cement Concrete

    Directory of Open Access Journals (Sweden)

    Gonzalo Martínez-Barrera

    2015-01-01

    Full Text Available The development of the packaging industry has promoted indiscriminately the use of disposable packing as Tetra Pak, which after a very short useful life turns into garbage, helping to spoil the environment. One of the known processes that can be used for achievement of the compatibility between waste materials and the environment is the gamma radiation, which had proved to be a good tool for modification of physicochemical properties of materials. The aim of this work is to study the effects of waste cellulose from Tetra Pak packing and gamma radiation on the mechanical properties of cement concrete. Concrete specimens were elaborated with waste cellulose at concentrations of 3, 5, and 7 wt% and irradiated at 200, 250, and 300 kGy of gamma dose. The results show highest improvement on the mechanical properties for concrete with 3 wt% of waste cellulose and irradiated at 300 kGy; such improvements were related with the surface morphology of fracture zones of cement concrete observed by SEM microscopy.

  2. Resistance of class C fly ash belite cement to simulated sodium sulphate radioactive liquid waste attack.

    Science.gov (United States)

    Guerrero, A; Goñi, S; Allegro, V R

    2009-01-30

    The resistance of class C fly ash belite cement (FABC-2-W) to concentrated sodium sulphate salts associated with low level wastes (LLW) and medium level wastes (MLW) is discussed. This study was carried out according to the Koch and Steinegger methodology by testing the flexural strength of mortars immersed in simulated radioactive liquid waste rich in sulphate (48,000 ppm) and demineralised water (used as a reference), at 20 degrees C and 40 degrees C over a period of 180 days. The reaction mechanisms of sulphate ion with the mortar was carried out through a microstructure study, which included the use of Scanning electron microscopy (SEM), porosity and pore-size distribution and X-ray diffraction (XRD). The results showed that the FABC mortar was stable against simulated sulphate radioactive liquid waste (SSRLW) attack at the two chosen temperatures. The enhancement of mechanical properties was a result of the formation of non-expansive ettringite inside the pores and an alkaline activation of the hydraulic activity of cement promoted by the ingress of sulphate. Accordingly, the microstructure was strongly refined. PMID:18524482

  3. Low-level radioactive Hanford wastes immobilized by cement-based grouts

    International Nuclear Information System (INIS)

    More than 5,300,000 liters (1,400,000 gal) of phosphate/sulfate waste (PSW) grout were produced and placed in vault 101 at the Hanford Site. This waste was generated during decontamination operations and maintenance of the fuel storage basin at the N Reactor. The low-level radioactive liquid wastes were mixed with a blend of portland cement, fly ash, and clays. Through cementing and pozzolanic reactions with water, the grout was solidified to immobilize contaminants and retain low permeability to groundwater. Testing conducted before the campaign is described. The usefulness of each quality verification technique is discussed, focusing mainly on data from the core samples. These data provide the best information on PSW grout since core samples from all regions and depths in the vault were tested. The nondestructive testing data are also useful as they provide property data from broad regions of the vault. The mean compressive strength of the PSW grout cores is 4.17 MPa, much higher than the criterion value of 0.35 MPa. Results also show that the leachability indices for 137Cs, 60Co, sodium, and SO4 for PSW grout cores exceed the leachability criterion by at least one index point. This means that the ability of the grout to resist leaching of waste species is at least ten times greater than the limiting criterion

  4. Biological responses of brushite-forming Zn- and ZnSr- substituted beta-tricalcium phosphate bone cements

    Directory of Open Access Journals (Sweden)

    S Pina

    2010-09-01

    Full Text Available The core aim of this study was to investigate zinc (Zn- and zinc and strontium (ZnSr-containing brushite-forming beta-tricalcium phosphate (TCP cements for their effects on proliferation and differentiation of osteoblastic-like cells (MC3T3-E1 cell line as well as for their in vivo behaviour in trabecular bone cylindrical defects in a pilot study. In vitro proliferation and maturation responses of MC3T3-E1 osteoblastic-like cells to bone cements were studied at the cellular and molecular levels. The Zn- and Sr-containing brushite cements were found to stimulate pre-osteoblastic proliferation and osteoblastic maturation. Indeed, MC3T3-E1 cells exposed to the powdered cements had increased proliferative rates and higher adhesiveness capacity, in comparison to control cells. Furthermore, they exhibited higher alkaline phosphatase (ALP activity and increased Type-I collagen secretion and fibre deposition into the extracellular matrix. Proliferative and collagen deposition properties were more evident for cells grown in cements doped with Sr. The in vivo osteoconductive propertiesof the ZnCPC and ZnSrCPC cements were also pursued. Histological and histomorphometric analyses were performed at 1 and 2 months after implantation, using carbonated apatite cement (Norian SRS® as control. There was no evidence of cement-induced adverse foreign body reactions, and furthermore ZnCPC and ZnSrCPC cements revealed better in vivo performance in comparison to the control apatite cement. Additionally, the presence of both zinc and strontium resulted in the highest rate of new bone formation. These novel results indicate that the investigated ZnCPC and ZnSrCPC cements are both biocompatible and osteoconductive, being good candidate materials to use as bone substitutes.

  5. An Experimental Investigation of Partial Replacement of Cement by Industrial Waste (Hypo Sludge

    Directory of Open Access Journals (Sweden)

    Mr.R.Balamurugan

    2014-04-01

    Full Text Available Concrete is strength and tough material but it is porous material also which interacts with the surrounding environment. The durability of concrete depends largely on the movement of water and gas enters and moves through it. To produce low cost concrete by blending various ratios of cement with hypo sludge & to reduce disposal and pollution problems due to hypo sludge it is most essential to develop profitable building materials from hypo sludge. To make good quality paper limited number of times recycled Paper fibers can be used which produces a large amount of solid waste. The innovative use of hypo sludge in concrete formulations as a supplementary cementations material was tested as an alternative to traditional concrete.

  6. Moisture transport properties of cement-based materials for engineered barriers in radioactive waste disposal

    International Nuclear Information System (INIS)

    This paper reviews the multiphase modeling of moisture transport process in pore structure of cement-based materials used as engineered barriers in radioactive waste disposal. The emphasis is put on the fundamental relationship of moisture isotherm and the related hysteresis phenomenon. A typical cement-based material is retained for study and its properties for moisture transport were measured. The pore structure was characterized by mercury intrusion porosimetry (MIP) and gravimetry method. The moisture isotherm was measured in laboratory by humidity equilibrium method and the predicted isotherm from MIP pore structure is confronted with the measured isotherm. Afterwards, a numerical scheme is set up for the multiphase transport model and the model is applied to the moisture transport process of engineered barriers exposed to natural drying and drying-wetting cycles. It is observed that the ratio between drying and wetting periods has strong influence on the depth of surface convection zone. (authors)

  7. DSNF and other waste form degradation abstraction

    Energy Technology Data Exchange (ETDEWEB)

    Thornton, Thomas A.

    2000-12-20

    The purpose of this analysis/model report (AMR) is to select and/or abstract conservative degradation models for DOE-(US. Department of Energy) owned spent nuclear fuel (DSNF) and the immobilized ceramic plutonium (Pu) disposition waste forms for application in the proposed monitored geologic repository (MGR) postclosure Total System Performance Assessment (TSPA). Application of the degradation models abstracted herein for purposes other than TSPA should take into consideration the fact that they are, in general, very conservative. Using these models, the forward reaction rate for the mobilization of radionuclides, as solutes or colloids, away from the waste fondwater interface by contact with repository groundwater can then be calculated. This forward reaction rate generally consists of the dissolution reaction at the surface of spent nuclear fuel (SNF) in contact with water, but the degradation models, in some cases, may also include and account for the physical disintegration of the SNF matrix. The models do not, however, account for retardation, precipitation, or inhibition of the migration of the mobilized radionuclides in the engineered barrier system (EBS). These models are based on the assumption that all components of the DSNF waste form are released congruently with the degradation of the matrix.

  8. Stabilization of ZnCl2-Containing Waste Using Calcium Sulfoaluminate Cement

    International Nuclear Information System (INIS)

    The potential of calcium sulfoaluminate (CSA) cement was investigated to solidify and stabilize radwastes containing large amounts of soluble zinc chloride (a strong inhibitor of Portland cement hydration). Hydration of pastes and mortars prepared with a 0.5 mol/L ZnCl2 mixing solution was characterized over one year as a function of the gypsum content of the binder and the thermal history of the material. Blending the CSA clinker with 20% gypsum enabled rapid hydration, with only very small delay compared with a reference prepared with pure water. It also improved the compressive strength of the hardened material and significantly reduced its expansion under wet curing. Moreover, the hydrate assemblage was less affected by a thermal treatment at early age simulating the temperature rise and fall occurring in a large-volume drum of cemented waste. Fully hydrated materials contained ettringite, amorphous aluminum hydroxide, straetlingite, together with AFm phases (Kuzel's salt associated with monosulfoaluminate or Friedel's salt depending on the gypsum content of the binder), and possibly C-(A)-S-H. Zinc was readily insolubilized and could not be detected in the pore solution extracted from cement pastes, or in their leachates after 3 months of leaching by pure water at pH 7. The good retention of zinc by the cement matrix was mainly attributed to the precipitation of a hydrated and well crystallized phase with platelet morphology (which may belong to the layered double hydroxides family) at early age ≤ 1 day), and to chemisorption onto aluminum hydroxide at later age. (author)

  9. A preliminary assessment of polymer-modified cements for use in immobilisation of intermediate level radioactive waste

    International Nuclear Information System (INIS)

    A range of polymer-modified cements has been examined as candidate materials for the immobilisation of intermediate level radioactive waste. The waste streams studied were inactive simulates of real wastes and included ion-exchange resins, Magnox debris and dilute sludges. Preliminary experiments on the compatibility of the polymer-cement-waste combinations have been carried out and measurements of flexural strength before and after #betta#-irradiation to 109 rad and water immersion have been made. Soxhlet leach tests have been used to compare the leach rates of the different materials. From the results of these preliminary experiments, a limited number of polymer-modified cements have been suggested as suitable for more detailed study. (author)

  10. Results after ten years of field testing low-level radioactive waste forms using lysimeters

    International Nuclear Information System (INIS)

    The Field Lysimeter Investigations: Low-Level Waste Data Base Development Program is obtaining information on the performance of radioactive waste forms. Ion-exchange resins from a commercial nuclear power station were solidified into waste forms using portland cement and vinyl esterstyrene. These waste forms are being tested to: (a) obtain information on performance of waste forms in typical disposal environments, (b) compare field results with bench leach studies, (c) develop a low-level waste data base for use in performance assessment source term calculations, and (d) apply the DUST computer code to compare predicted cumulative release to actual field data. The program, funded by the Nuclear Regulatory Commission (NRC), includes observed radionuclide releases from waste forms in field lysimeters. The purpose of this paper is to present the experimental results of two lysimeter arrays over 10 years of operation, and to compare those results to bench test results and to DUST code predicted releases. Further analysis of soil cores taken to define the observed upward migration of radionuclides in one lysimeter is also presented

  11. Review of radiation effects in solid-nuclear-waste forms

    International Nuclear Information System (INIS)

    Radiation effects on the stability of high-level nuclear waste (HLW) forms are an important consideration in the development of technology to immobilize high-level radioactive waste because such effects may significantly affect the containment of the radioactive waste. Since the required containment times are long (103 to 106 years), an understanding of the long-term cumulative effects of radiation damage on the waste forms is essential. Radiation damage of nuclear waste forms can result in changes in volume, leach rate, stored energy, structure/microstructure, and mechanical properties. Any one or combination of these changes might significantly affect the long-term stability of the nuclear waste forms. This report defines the general radiation damage problem in nuclear waste forms, describes the simulation techniques currently available for accelerated testing of nuclear waste forms, and reviews the available data on radiation effects in both glass and ceramic (primarily crystalline) waste forms. 76 references

  12. Review of radiation effects in solid-nuclear-waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Weber, W.J.

    1981-09-01

    Radiation effects on the stability of high-level nuclear waste (HLW) forms are an important consideration in the development of technology to immobilize high-level radioactive waste because such effects may significantly affect the containment of the radioactive waste. Since the required containment times are long (10/sup 3/ to 10/sup 6/ years), an understanding of the long-term cumulative effects of radiation damage on the waste forms is essential. Radiation damage of nuclear waste forms can result in changes in volume, leach rate, stored energy, structure/microstructure, and mechanical properties. Any one or combination of these changes might significantly affect the long-term stability of the nuclear waste forms. This report defines the general radiation damage problem in nuclear waste forms, describes the simulation techniques currently available for accelerated testing of nuclear waste forms, and reviews the available data on radiation effects in both glass and ceramic (primarily crystalline) waste forms. 76 references.

  13. Safeguards and retrievability from waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Danker, W.

    1996-05-01

    This report describes issues discussed at a session from the PLutonium Stabilization and Immobilization Workshop related to safeguards and retrievability from waste forms. Throughout the discussion, the group probed the goals of disposition efforts, particularly an understanding of the {open_quotes}spent fuel standard{close_quotes}, since the disposition material form derives from these goals. The group felt strongly that not only the disposition goals but safeguards to meet these goals could affect the material form. Accordingly, the Department was encouraged to explore and apply safeguards as early in the implementation process as possible. It was emphasized that this was particularly true for any planned use of existing facilities. It is much easier to build safeguards approaches into the development of new facilities, than to backfit existing facilities. Accordingly, special safeguards challenges are likely to be encountered, given the cost and schedule advantages offered by use of existing facilities.

  14. Standard test method for static leaching of monolithic waste forms for disposal of radioactive waste

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This test method provides a measure of the chemical durability of a simulated or radioactive monolithic waste form, such as a glass, ceramic, cement (grout), or cermet, in a test solution at temperatures <100°C under low specimen surface- area-to-leachant volume (S/V) ratio conditions. 1.2 This test method can be used to characterize the dissolution or leaching behaviors of various simulated or radioactive waste forms in various leachants under the specific conditions of the test based on analysis of the test solution. Data from this test are used to calculate normalized elemental mass loss values from specimens exposed to aqueous solutions at temperatures <100°C. 1.3 The test is conducted under static conditions in a constant solution volume and at a constant temperature. The reactivity of the test specimen is determined from the amounts of components released and accumulated in the solution over the test duration. A wide range of test conditions can be used to study material behavior, includin...

  15. Change of carcinogenic chrysotile fibers in the asbestos cement (eternit) to harmless waste by artificial carbonatization: Petrological and technological results

    International Nuclear Information System (INIS)

    Highlights: ► Carcinogenic chrysotile fibers in asbestos cement (eternit) are liquidated. ► Thermally modified eternit grist (at 650 °C, 1 h) reacts with CO2 + water. ► Carbonates hydromagnesite and magnesite are the newly formed products of artificial carbonatization. ► Neutralizing of extreme pH values (around 12) at large eternit dumps. ► An alternative methodology for permanent liquidation of a part of CO2 emissions. -- Abstract: Asbestos cement materials, mainly the eternit roof ceiling, being widely applied in the past, represent a serious environmental load. The solar radiation, rain and frost cause the deliberation of cement from the eternit roofing and consequently the wind contaminates the surrounding area by the asbestos (chrysotile) fibers. In combination with other carcinogens (e.g. smoking), or at reduced immunity of a man, they may cause serious respiratory diseases and lung cancer. The article presents the procedure and experimental results of artificial carbonatization, applied in the asbestos cement (eternit). The wet crushed and pulverized asbestos cement was thermally modified at 650 °C and then the chrysotile fibers easily and completely reacted with the mixture of CO2 and water, producing new Mg-rich carbonates – hydromagnesite and magnesite: 2Mg3Si2O5(OH)3thermallymodifiedchrysotile+5CO2+nH2O→Mg5(CO3)4(OH)2⋅4H2Ohydromagnesite+MgCO3magnesite+4SiO2 · nH2Oin amorphousphase;n=3÷9 Applying this methodology, the asbestos-bearing waste can be stabilized and environmentally friendly permanently deposited. Finding a way of neutralizing of extreme pH values (around 12) at large eternit dumps represents also an asset of presented research. Simultaneously, the artificial carbonatization of chrysotile asbestos, applying CO2, offers an alternative way for permanent liquidation of a part of industrial CO2 emissions, contributing to multiple benefit of this methodology

  16. Change of carcinogenic chrysotile fibers in the asbestos cement (eternit) to harmless waste by artificial carbonatization: Petrological and technological results

    Energy Technology Data Exchange (ETDEWEB)

    Radvanec, Martin; Tuček, Ľubomír; Derco, Ján; Čechovská, Katarína [State Geological Institute of Dionýz Štúr, Mlynská dolina 1, SK-817 04 Bratislava (Slovakia); Németh, Zoltán, E-mail: zoltan.nemeth@geology.sk [State Geological Institute of Dionýz Štúr, Mlynská dolina 1, SK-817 04 Bratislava (Slovakia)

    2013-05-15

    Highlights: ► Carcinogenic chrysotile fibers in asbestos cement (eternit) are liquidated. ► Thermally modified eternit grist (at 650 °C, 1 h) reacts with CO{sub 2} + water. ► Carbonates hydromagnesite and magnesite are the newly formed products of artificial carbonatization. ► Neutralizing of extreme pH values (around 12) at large eternit dumps. ► An alternative methodology for permanent liquidation of a part of CO{sub 2} emissions. -- Abstract: Asbestos cement materials, mainly the eternit roof ceiling, being widely applied in the past, represent a serious environmental load. The solar radiation, rain and frost cause the deliberation of cement from the eternit roofing and consequently the wind contaminates the surrounding area by the asbestos (chrysotile) fibers. In combination with other carcinogens (e.g. smoking), or at reduced immunity of a man, they may cause serious respiratory diseases and lung cancer. The article presents the procedure and experimental results of artificial carbonatization, applied in the asbestos cement (eternit). The wet crushed and pulverized asbestos cement was thermally modified at 650 °C and then the chrysotile fibers easily and completely reacted with the mixture of CO{sub 2} and water, producing new Mg-rich carbonates – hydromagnesite and magnesite: 2Mg{sub 3}Si{sub 2}O{sub 5}(OH){sub 3thermally} {sub modified} {sub chrysotile}+5CO{sub 2}+nH{sub 2}O→Mg{sub 5}(CO{sub 3}){sub 4}(OH){sub 2}⋅4H{sub 2}O{sub hydromagnesite}+MgCO{sub 3magnesite}+4SiO{sub 2} · nH{sub 2}O{sub in} a{sub morphous} {sub phase};n=3÷9 Applying this methodology, the asbestos-bearing waste can be stabilized and environmentally friendly permanently deposited. Finding a way of neutralizing of extreme pH values (around 12) at large eternit dumps represents also an asset of presented research. Simultaneously, the artificial carbonatization of chrysotile asbestos, applying CO{sub 2}, offers an alternative way for permanent liquidation of a part of

  17. Full-scale tests of sulfur polymer cement and non-radioactive waste in heated and unheated prototypical containers

    Energy Technology Data Exchange (ETDEWEB)

    Darnell, G.R.; Aldrich, W.C.; Logan, J.A.

    1992-02-01

    Sulfur polymer cement has been demonstrated to be superior to portland cement in the stabilization of numerous troublesome low- level radioactive wastes, notably mixed waste fly ash, which contains heavy metals. EG&G Idaho, Inc. conducted full-scale, waste-stabilization tests with a mixture of sulfur polymer cement and nonradioactive incinerator ash poured over simulated steel and ash wastes. The container used to contain the simulated waste for the pour was a thin-walled, rectangular, steel container with no appendages. The variable in the tests was that one container and its contents were at 65{degree}F (18{degree}C) at the beginning of the pour, while the other was preheated to 275{degree}F (135{degree}C) and was insulated before the pour. The primary goal was to determine the procedures and equipment deemed operationally acceptable and capable of providing the best probability of passing the only remaining governmental test for sulfur polymer cement, the Nuclear Regulatory Commission`s full-scale test. The secondary goal was to analyze the ability of the molten cement and ash mixture to fill different size pipes and thus eliminate voids in the resultant 24 ft{sup 3} monolith.

  18. Full-scale tests of sulfur polymer cement and non-radioactive waste in heated and unheated prototypical containers

    Energy Technology Data Exchange (ETDEWEB)

    Darnell, G.R.; Aldrich, W.C.; Logan, J.A.

    1992-02-01

    Sulfur polymer cement has been demonstrated to be superior to portland cement in the stabilization of numerous troublesome low- level radioactive wastes, notably mixed waste fly ash, which contains heavy metals. EG G Idaho, Inc. conducted full-scale, waste-stabilization tests with a mixture of sulfur polymer cement and nonradioactive incinerator ash poured over simulated steel and ash wastes. The container used to contain the simulated waste for the pour was a thin-walled, rectangular, steel container with no appendages. The variable in the tests was that one container and its contents were at 65{degree}F (18{degree}C) at the beginning of the pour, while the other was preheated to 275{degree}F (135{degree}C) and was insulated before the pour. The primary goal was to determine the procedures and equipment deemed operationally acceptable and capable of providing the best probability of passing the only remaining governmental test for sulfur polymer cement, the Nuclear Regulatory Commission's full-scale test. The secondary goal was to analyze the ability of the molten cement and ash mixture to fill different size pipes and thus eliminate voids in the resultant 24 ft{sup 3} monolith.

  19. The effect of supplementary pulverized fuel ash on calcium aluminate phosphate cement for intermediate-level waste encapsulation

    International Nuclear Information System (INIS)

    The objective of the current study was to evaluate the effects of supplementary pulverized fuel ash on phosphate-modified calcium aluminate cement. These systems are being established as part of a wider project to develop alternative cementing systems for the encapsulation of problematic low- and intermediate- level radioactive waste in the UK. The nuclear industry has established specific processing and property criteria, which must be fulfilled to ensure suitability for industrial application. In a series of studies, pulverized fuel ash was used as a partial replacement for calcium aluminate cement, to improve the fluidity of the system and increase the setting time. Properties such as slurry pH and fluidity, setting time, mechanical properties, and porosity were investigated using Vicat, Colflow, and compressive strength testing equipment and mercury intrusion porosimetry. The hardened cement pastes were also characterised using X-ray diffraction, thermogravimetric analysis and scanning electron microscopy. A formulation envelope was identified, which fulfilled the plant acceptance tests defined by industry to ensure suitability for industrial application. It was found that pH of calcium aluminate phosphate cements is lower than that of conventional cementing systems used to encapsulate radioactive waste in the UK. Hence, they have potential to be used as an alternative cementing system for the encapsulation of problematic radioactive metals. (authors)

  20. Mechanical behavior of the asbestos-cement container for geological disposal of α level technological wastes from COGEMA reprocessing plants

    International Nuclear Information System (INIS)

    For the safety assessment of the SGN asbestos cement container concept selected by COGEMA for the conditioning of cemented technological wastes from the UP3-UP2 800 reprocessing plants, a general survey has been carried out to confirm both its confinement capacity and its mechanical strength. This safety assessment relates to the latter aspect. It implies two stages: first, the material characterization of asbestos cement and epoxide resin used in sealing and assembling; second, the finite element calculation of induced stresses and strains under storage conditions with regards to the experimented mechanical characteristics. The authors infer some damage in packaging materials in case of misoperation in conditioning process

  1. Mechanical Behaviour of the asbestos-cement container for geological disposal of α level technological wastes from Cogema reprocessing plants

    International Nuclear Information System (INIS)

    For the safety assessment of the SGN asbestos cement container concept selected by COGEMA for the conditionning of cemented α technological wastes from the UP3-UP2 800 reprocessing plants, a general survey has been carried out to confirm both its confinement capacity and its mechanical strength. This safety assessment relates to the latter aspect. It implies two stages: first, the material characterization of asbestos cement and epoxide resin used in sealing and assembling; second, the finite element calculation of induced stresses and strains under storage conditions with regards to the experimented mechanical characteristics. We infer some damage in packaging materials in case of misoperation in conditionning process

  2. DuraLith Alkali-Aluminosilicate Geopolymer Waste Form Testing for Hanford Secondary Waste

    International Nuclear Information System (INIS)

    The primary objective of the work reported here was to develop additional information regarding the DuraLith alkali aluminosilicate geopolymer as a waste form for liquid secondary waste to support selection of a final waste form for the Hanford Tank Waste Treatment and Immobilization Plant secondary liquid wastes to be disposed in the Integrated Disposal Facility on the Hanford Site. Testing focused on optimizing waste loading, improving waste form performance, and evaluating the robustness of the waste form with respect to waste variability.

  3. DuraLith Alkali-Aluminosilicate Geopolymer Waste Form Testing for Hanford Secondary Waste

    Energy Technology Data Exchange (ETDEWEB)

    Gong, W. L.; Lutz, Werner; Pegg, Ian L.

    2011-07-21

    The primary objective of the work reported here was to develop additional information regarding the DuraLith alkali aluminosilicate geopolymer as a waste form for liquid secondary waste to support selection of a final waste form for the Hanford Tank Waste Treatment and Immobilization Plant secondary liquid wastes to be disposed in the Integrated Disposal Facility on the Hanford Site. Testing focused on optimizing waste loading, improving waste form performance, and evaluating the robustness of the waste form with respect to waste variability.

  4. Comparative Effect of Bio-waste Ashes on Strength Properties of Cement Mortar

    Science.gov (United States)

    Ajay, Goyal; Hattori, Kunio; Ogata, Hidehiko; Ashraf, Muhammad; Ahmed, Mohamed Anwar

    Biomass fuels produce about 400 million tonnes of ashes as waste material. This paper discusses the pozzolanic character of bio-waste ashes obtained from dry tree leaves (AML), Korai grass (KRI) and Tifton grass (TFT). Ashes were obtained by control incineration of the wastes at 600°C for 5 hours and mortar specimens were prepared by substituting cement with 10, 20 and 30% ash. Strength development of ash-blended mortar specimens was evaluated by conducting destructive tests as well as non-destructive tests till 91 days. X-ray diffraction, scanning electron microscopic and thermo-gravimetric techniques were used to analyze the influence of ash substitution on strength properties of blended-mortar. Pozzolanic reactivity of AML- and KRI-ash was confirmed, but TFT-ash did not show enough reactivity. Overall results confirmed that up to 20% substitution of cement can be made with AML- or KRI-ash with strength approaching 90% of that of control.

  5. Blast furnace slag-cement grout blends for the immobilization of technetium-containing wastes

    International Nuclear Information System (INIS)

    Mixed low-level radioactive and chemically toxic process treatment wastes from the Portsmouth Gaseous Diffusion Plant are stabilized by solidification in cement-based grouts. Conventional portland cement and fly ash grouts are shown to be very effective for retention of hydrolyzable heavy metals (including lead, cadmium, uranium, and nickel), but are marginally acceptable for retention of radioactive 99Tc (which is present in the waste as the highly mobile pertechnate anion). Addition of ground blast furnace slag to the grout is shown to reduce the effective diffusivity of technetium by several orders of magnitude; retention of technetium is improved by decreasing the waste loading in the grout or by increasing the proportion of blast furnace slag in the grout dry mix. The selective effect of slag is believed to be due to its ability to reduce Tc(VIII) to the less soluble Tc(IV) species. The addition of other reductive grout admixtures (e.g., sodium sulfide, ferrous ion, and powdered iron metal) also appear to improve the retention of technetium in grout. 31 refs., 2 figs., 25 tabs

  6. Alpha damage in non-reference waste form matrix materials

    International Nuclear Information System (INIS)

    Although bitumen is the matrix material currently used for European α-bearing intermediate level waste streams, polymer and polymer-modified cement matrices could have advantages over bitumen for such wastes. Two organic matrix systems have been studied - an epoxide resin, and an epoxide modified cement. Alpha irradiations were carried out by incorporating 241Am at approx. 0.9 Ci/l. Comparisons have been made with unirradiated material and with materials which had been γ-irradiated to the same dose as the α-irradiated samples. Measurements were made of dimensional changes, mechanical properties and the leaching behaviour of 241Am and 137Cs. A limited amount of swelling (< 3%) was observed in α-irradiated epoxide resin; none was observed in the epoxide modified cement. Gamma irradiation to 300 kGy has no significant effect on the mechanical properties of either system. However, alpha irradiation to the same dose produced significant changes in flexural strength, an increase for the polymer and a decrease for the polymer-cement. Leaching in these systems was found to be a diffusion-controlled process; alpha irradiation to approx. 250 kGy has little effect on the leaching behaviour of either system. (author)

  7. Studies on the Potential of Waste Soda Lime Silica Glass in Glass Ionomer Cement Production

    OpenAIRE

    V. W. Francis Thoo; N. Zainuddin; Matori, K. A.; S.A. Abdullah

    2013-01-01

    Glass ionomer cements (GIC) are produced through acid base reaction between calcium-fluoroaluminosilicate glass powder and polyacrylic acid (PAA). Soda lime silica glasses (SLS), mainly composed of silica (SiO2), have been utilized in this study as the source of SiO2 for synthesis of Ca-fluoroaluminosilicate glass. Therefore, the main objective of this study was to investigate the potential of SLS waste glass in producing GIC. Two glasses, GWX 1 (analytical grade SiO2) and GWX 2 (replacing Si...

  8. Stability testing of low-level waste forms

    International Nuclear Information System (INIS)

    The NRC Technical Position on Waste Form identifies methods for thermal cycle testing and biodegradation testing of low-level waste forms. These tests were carried out on low-level waste forms to establish whether the tests are reasonable and can be achieved. The thermal-cycle test is believed adequate for demonstrating the thermal stability of solidified waste forms. The biodegradation tests are sufficient for distinguishing materials that are susceptible to biodegradation. However, failure of either of these tests should not be regarded of itself as an indication that the waste form will biodegrade to an extent that the form does not meet the stability requirements of 10 CFR Part 61

  9. NDA issues with RFETS vitrified waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Hurd, J.; Veazey, G.

    1998-12-31

    A study was conducted at Los Alamos National Laboratory (LANL) for the purpose of determining the feasibility of using a segmented gamma scanner (SGS) to accurately perform non-destructive analysis (NDA) on certain Rocky Flats Environmental Technology Site (RFETS) vitrified waste samples. This study was performed on a full-scale vitrified ash sample prepared at LANL according to a procedure similar to that anticipated to be used at RFETS. This sample was composed of a borosilicate-based glass frit, blended with ash to produce a Pu content of {approximately}1 wt %. The glass frit was taken to a degree of melting necessary to achieve a full encapsulation of the ash material. The NDA study performed on this sample showed that SGSs with either {1/2}- or 2-inch collimation can achieve an accuracy better than 6 % relative to calorimetry and {gamma}-ray isotopics. This accuracy is achievable, after application of appropriate bias corrections, for transmissions of about {1/2} % through the waste form and counting times of less than 30 minutes. These results are valid for ash material and graphite fines with the same degree of plutonium particle size, homogeneity, sample density, and sample geometry as the waste form used to obtain the results in this study. A drum-sized thermal neutron counter (TNC) was also included in the study to provide an alternative in the event the SGS failed to meet the required level of accuracy. The preliminary indications are that this method will also achieve the required accuracy with counting times of {approximately}30 minutes and appropriate application of bias corrections. The bias corrections can be avoided in all cases if the instruments are calibrated on standards matching the items.

  10. Evaluation and selection of candidate high-level waste forms

    International Nuclear Information System (INIS)

    Seven candidate waste forms being developed under the direction of the Department of Energy's National High-Level Waste (HLW) Technology Program, were evaluated as potential media for the immobilization and geologic disposal of high-level nuclear wastes. The evaluation combined preliminary waste form evaluations conducted at DOE defense waste-sites and independent laboratories, peer review assessments, a product performance evaluation, and a processability analysis. Based on the combined results of these four inputs, two of the seven forms, borosilicate glass and a titanate based ceramic, SYNROC, were selected as the reference and alternative forms for continued development and evaluation in the National HLW Program. Both the glass and ceramic forms are viable candidates for use at each of the DOE defense waste-sites; they are also potential candidates for immobilization of commercial reprocessing wastes. This report describes the waste form screening process, and discusses each of the four major inputs considered in the selection of the two forms

  11. High-performance cement-based grouts for use in a nuclear waste disposal facility

    International Nuclear Information System (INIS)

    National and international agencies have identified cement-based materials as prime candidates for sealing vaults that would isolate nuclear fuel wastes from the biosphere. Insufficient information is currently available to allow a reasonable analysis of the long-term performance of these sealing materials in a vault. A combined laboratory and modelling research program was undertaken to provide the necessary information for a specially developed high-performance cement grout. The results indicate that acceptable performance is likely for at least thousands of years and probably for much longer periods. The materials, which have been proven to be effective in field applications, are shown to be virtually impermeable and highly leach resistant under vault conditions. Special plasticizing additives used in the material formulation enhance the physical characteristics of the grout without detriment to its chemical durability. Neither modelling nor laboratory testing have yet provided a definitive assessment of the grout's longevity. However, none of the results of these studies has contraindicated the use of high-performance cement-based grouts in vault sealing applications. (Author) (24 figs., 6 tabs., 21 refs.)

  12. Optimum feeding rate of solid hazardous waste in a cement kiln burner

    Directory of Open Access Journals (Sweden)

    W.K. Hiromi Ariyaratne, Morten C. Melaaen, Lars-André Tokheim

    2013-01-01

    Full Text Available Solid hazardous waste mixed with wood chips (SHW is a partly CO2 neutral fuel, and hence is a good candidate for substituting fossil fuels like pulverized coal in rotary kiln burners used in cement kiln systems. SHW is used in several cement plants, but the optimum substitution rate has apparently not yet been fully investigated. The present study aims to find the maximum possible replacement of coal by SHW, without negatively affecting the product quality, emissions and overall operation of the process. A full-scale experiment was carried out in the rotary kiln burner of a cement plant by varying the SHW substitution rate from 0 to 3 t/hr. Clinker quality, emissions and other relevant operational data from the experiment were analysed using fuel characteristics of coal and SHW. The results revealed that SHW could safely replace around 20% of the primary coal energy without giving negative effects. The limiting factor is the free lime content of the clinker. Results from the present study were also compared with results from a previous test using meat and bone meal.

  13. The effects of aging on compressive strength of low-level radioactive waste form samples

    International Nuclear Information System (INIS)

    The Field Lysimeter Investigations: Low-Level Waste Data Base Development Program, funded by the US Nuclear Regulatory Commission (NRC), is (a) studying the degradation effects in organic ion-exchange resins caused by radiation, (b) examining the adequacy of test procedures recommended in the Branch Technical Position on Waste Form to meet the requirements of 10 CFR 61 using solidified ion-exchange resins, (c) obtaining performance information on solidified ion-exchange resins in a disposal environment, and (d) determining the condition of liners used to dispose ion-exchange resins. Compressive tests were performed periodically over a 12-year period as part of the Technical Position testing. Results of that compressive testing are presented and discussed. During the study, both portland type I-II cement and Dow vinyl ester-styrene waste form samples were tested. This testing was designed to examine the effects of aging caused by self-irradiation on the compressive strength of the waste forms. Also presented is a brief summary of the results of waste form characterization, which has been conducted in 1986, using tests recommended in the Technical Position on Waste Form. The aging test results are compared to the results of those earlier tests. 14 refs., 52 figs., 5 tabs

  14. Waste form development for a DC arc furnace

    International Nuclear Information System (INIS)

    A laboratory crucible study was conducted to develop waste forms to treat nonradioactive simulated 238Pu heterogeneous debris waste from Savannah River, metal waste from the Idaho National Engineering Laboratory (INEL), and nominal waste also from INEL using DC arc melting. The preliminary results showed that the different waste form compositions had vastly different responses for each processing effect. The reducing condition of DC arc melting had no significant effects on the durability of some waste forms while it decreased the waste form durability from 300 to 700% for other waste forms, which resulted in the failure of some TCLP tests. The right formulations of waste can benefit from devitrification and showed an increase in durability by 40%. Some formulations showed no devitrification effects while others decreased durability by 200%. Increased waste loading also affected waste form behavior, decreasing durability for one waste, increasing durability by 240% for another, and showing no effect for the third waste. All of these responses to the processing and composition variations were dictated by the fundamental glass chemistry and can be adjusted to achieve maximal waste loading, acceptable durability, and desired processing characteristics if each waste formulation is designed for the result according to the glass chemistry

  15. Effects of Waste Glass (WG on the Strength Characteristics of Cement Stabilized Expansive Soil

    Directory of Open Access Journals (Sweden)

    I.A.Ikara

    2015-11-01

    Full Text Available The study investigates the suitability of using waste glass (WG as admixture to cement stabilized black cotton soil (BCS for roads, fills and embankment. The soil was classified as A-7-5 and CH according to the American Association of State Highway and Transport Officials (AASHTO and the Unified Soil Classification System (USCS Classifications. Chemical analysis revealed that WG is rich in main oxides such as Silicon Oxide (69.2, Aluminium Oxide (2.29, Iron Oxide (1.57, Calcium Oxide (15.1 and Sodium Oxide (8.75. The soil was stabilized with 0, 2, 4, 6 and 8% cement and 0, 5 10, 15 and 20% WG by weight of the dry soil. Laboratory tests were carried out using the Standard Proctor (SP compactive efforts, California Bearing Ratio (CBR, Unconfined Compressive Strength (UCS, and compaction characteristics tests to evaluate the effectiveness of WG on Ordinary Portland cement (OPC stabilized BCS. The results obtained showed a decrease in the plasticity index (PI, liquid limit (LL, plastic limit (PL and increase Maximum Dry Density (MDD with increase in WG content in all cement proportions used and as compared to the values obtained for the natural soil. The peak 7 days UCS values of 1152kN/m2 was obtained at 8% OPC and 20% WG. Similarly, highest CBR value of 53.8% was obtained at an optimum blend of 8% OPC/20%WG. The results indicate that there is a potential in the use of WG as admixture to strengthen Black cotton soils.

  16. Review of high-level waste form properties

    International Nuclear Information System (INIS)

    This report is a review of waste form options for the immobilization of high-level-liquid wastes from the nuclear fuel cycle. This review covers the status of international research and development on waste forms as of May 1979. Although the emphasis in this report is on waste form properties, process parameters are discussed where they may affect final waste form properties. A summary table is provided listing properties of various nuclear waste form options. It is concluded that proposed waste forms have properties falling within a relatively narrow range. In regard to crystalline versus glass waste forms, the conclusion is that either glass of crystalline materials can be shown to have some advantage when a single property is considered; however, at this date no single waste form offers optimum properties over the entire range of characteristics investigated. A long-term effort has been applied to the development of glass and calcine waste forms. Several additional waste forms have enough promise to warrant continued research and development to bring their state of development up to that of glass and calcine. Synthetic minerals, the multibarrier approach with coated particles in a metal matrix, and high pressure-high temperature ceramics offer potential advantages and need further study. Although this report discusses waste form properties, the total waste management system should be considered in the final selection of a waste form option. Canister design, canister materials, overpacks, engineered barriers, and repository characteristics, as well as the waste form, affect the overall performance of a waste management system. These parameters were not considered in this comparison

  17. Review of high-level waste form properties. [146 bibliographies

    Energy Technology Data Exchange (ETDEWEB)

    Rusin, J.M.

    1980-12-01

    This report is a review of waste form options for the immobilization of high-level-liquid wastes from the nuclear fuel cycle. This review covers the status of international research and development on waste forms as of May 1979. Although the emphasis in this report is on waste form properties, process parameters are discussed where they may affect final waste form properties. A summary table is provided listing properties of various nuclear waste form options. It is concluded that proposed waste forms have properties falling within a relatively narrow range. In regard to crystalline versus glass waste forms, the conclusion is that either glass of crystalline materials can be shown to have some advantage when a single property is considered; however, at this date no single waste form offers optimum properties over the entire range of characteristics investigated. A long-term effort has been applied to the development of glass and calcine waste forms. Several additional waste forms have enough promise to warrant continued research and development to bring their state of development up to that of glass and calcine. Synthetic minerals, the multibarrier approach with coated particles in a metal matrix, and high pressure-high temperature ceramics offer potential advantages and need further study. Although this report discusses waste form properties, the total waste management system should be considered in the final selection of a waste form option. Canister design, canister materials, overpacks, engineered barriers, and repository characteristics, as well as the waste form, affect the overall performance of a waste management system. These parameters were not considered in this comparison.

  18. EFFECT ON COMPRESSIVE STRENGTH OF CONCRETE WITH PARTIAL REPLACEMENT OF CEMENT BY MUNICIPAL SOLID WASTE INCINERATION ASH

    OpenAIRE

    V. Alivelu Mangamma

    2016-01-01

    The municipal solid waste incineration ash reduces are worldwide studied topic over the last decades, so that utilize the municipal solid waste is the one of the possibilities is to use MSWI in concrete production as it is done the bottom ash features the most convenient composition in concrete and it is a available in highest amounts among the MSWI ashes the bottom ash was used as partial replacement of cement of cement in concrete strength has to find ,if the prepared concrete will get suff...

  19. The effectivity of bentonites in cesium retention of cemented waste products

    International Nuclear Information System (INIS)

    The nuclear energy has been used for the human development in different areas, as in the medicine, in the agriculture, in the industry and in the environmental protection, besides the electricity generation. As in other activities, in the use of nuclear energy, residues are also generated. They are considered radioactive wastes when the contaminant content can bring a potential negative impact in the human health and in the environment. In this case they should be properly managed and should not be released without treatment. In general the waste processing consists in a volume reduction followed by solidification and/or conditioning. A number of materials can be considered as immobilisation matrices for the wastes, with the objective of maintain the radioactive material physical and chemically stable. The cement is extensively used because it is easy to obtain, there is large. experience in its use and the processing is done at room temperature. Many materials have been studied to improve the fixation characteristics of the radionuclides in the cemented product. The aim of this study was to search, among Brazilian natural materials, those that could be effective in the contaminant retention without jeopardising the process and other characteristics of the waste product. Four types of bentonite were selected to the process and product evaluation tests. Many mixtures were prepared with simulated waste, cement and bentonite in different proportions. The viscosity, set time, compressive strength and leaching were evaluated. In addition it was verified if the products were monolithic and without free water. Inactive caesium was used as tracer. The leaching resistance is the most important parameter in the product evaluation, because it indicates the retention capacity of the matrix for radionuclides when the product is in contact with the water. In 1985 leaching tests were begun and they have been continued till now and from their results it was proved that the

  20. CSNF WASTE FORM DEGRADATION: SUMMARY ABSTRACTION

    Energy Technology Data Exchange (ETDEWEB)

    J.C. CUNNANE

    2004-08-31

    The purpose of this model report is to describe the development and validation of models that can be used to calculate the release of radionuclides from commercial spent nuclear fuel (CSNF) following a hypothetical breach of the waste package and fuel cladding in the repository. The purpose also includes describing the uncertainties associated with modeling the radionuclide release for the range of CSNF types, exposure conditions, and durations for which the radionuclide release models are to be applied. This document was developed in accordance with Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package (BSC 2004 [DIRS 169944]). This document considers radionuclides to be released from CSNF when they are available for mobilization by gas-phase mass transport, or by dissolution or colloid formation in water that may contact the fuel. Because other reports address limitations on the dissolved and colloidal radionuclide concentrations (BSC 2004 [DIRS 169944], Table 2-1), this report does not address processes that control the extent to which the radionuclides released from CSNF are mobilized and transported away from the fuel either in the gas phase or in the aqueous phase as dissolved and colloidal species. The scope is limited to consideration of degradation of the CSNF rods following an initial breach of the cladding. It considers features of CSNF that limit the availability of individual radionuclides for release into the gaseous or aqueous phases that may contact the fuel and the processes and events expected to degrade these CSNF features. In short, the purpose is to describe the characteristics of breached fuel rods and the degradation processes expected to influence radionuclide release.

  1. CSNF WASTE FORM DEGRADATION: SUMMARY ABSTRACTION

    International Nuclear Information System (INIS)

    The purpose of this model report is to describe the development and validation of models that can be used to calculate the release of radionuclides from commercial spent nuclear fuel (CSNF) following a hypothetical breach of the waste package and fuel cladding in the repository. The purpose also includes describing the uncertainties associated with modeling the radionuclide release for the range of CSNF types, exposure conditions, and durations for which the radionuclide release models are to be applied. This document was developed in accordance with Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package (BSC 2004 [DIRS 169944]). This document considers radionuclides to be released from CSNF when they are available for mobilization by gas-phase mass transport, or by dissolution or colloid formation in water that may contact the fuel. Because other reports address limitations on the dissolved and colloidal radionuclide concentrations (BSC 2004 [DIRS 169944], Table 2-1), this report does not address processes that control the extent to which the radionuclides released from CSNF are mobilized and transported away from the fuel either in the gas phase or in the aqueous phase as dissolved and colloidal species. The scope is limited to consideration of degradation of the CSNF rods following an initial breach of the cladding. It considers features of CSNF that limit the availability of individual radionuclides for release into the gaseous or aqueous phases that may contact the fuel and the processes and events expected to degrade these CSNF features. In short, the purpose is to describe the characteristics of breached fuel rods and the degradation processes expected to influence radionuclide release

  2. Manufacturing of concrete with residues from iron ore exploitation using the technology of radioactive waste cementation

    International Nuclear Information System (INIS)

    Radioactive wastes from various segments of economy are immobilized by cementation, because of availability and widespread use in civil construction of cement. New cementitious materials are developed in CDTN using mining residues based on cementing techniques of radioactive wastes. Special procedures were developed to obtain concrete with the use of super plasticizers in which natural sand was totally replaced by mining residues. The motivation for this research is the exploration of banded iron formations (BIF) as iron ore in 'Quadrilatero Ferrifero' of Minas Gerais, where huge amounts of residues are generated with great concern about the environmental sustainability and safety of dams for residue storage. The exploitation of river sand causes many negative impacts, which leads to interest in its replacement by another raw material in mortar and concrete manufacturing. The use of BIF mining residues were studied for manufacturing of concrete pavers to contribute to reducing the impact caused by extraction of natural sand and use of mining residues. Previously developed procedures with total replacement of natural sand for mining residues were modified, including use of gravel to obtain pavers with improved properties. Four different mixtures were tested, in which the proportion of gravel and super plasticizer was varied. Monitored properties of pavers, among others, were compression resistance, water absorption, and void volume. With addition of gravel, the pavers had higher void index than those made only with mortar, and higher resistance to compression after 28 days of curing (an average of 18MPa of those made with mortar to 24MPa of those made with concrete). (author)

  3. Formulation and Analysis of Compliant Grouted Waste Forms for SHINE Waste Streams

    Energy Technology Data Exchange (ETDEWEB)

    Ebert, William [Argonne National Lab. (ANL), Argonne, IL (United States); Pereira, Candido [Argonne National Lab. (ANL), Argonne, IL (United States); Heltemes, Thad A. [Argonne National Lab. (ANL), Argonne, IL (United States); Youker, Amanda [Argonne National Lab. (ANL), Argonne, IL (United States); Makarashvili, Vakhtang [Argonne National Lab. (ANL), Argonne, IL (United States); Vandegrift, George F. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-01-01

    Optional grouted waste forms were formulated for waste streams generated during the production of 99Mo to be compliant with low-level radioactive waste regulations. The amounts and dose rates of the various waste form materials that would be generated annually were estimated and used to determine the effects of various waste processing options, such as the of number irradiation cycles between uranium recovery operations, different combinations of waste streams, and removal of Pu, Cs, and Sr from waste streams for separate disposition (which is not evaluated in this report). These calculations indicate that Class C-compliant grouted waste forms can be produced for all waste streams. More frequent uranium recovery results in the generation of more chemical waste, but this is balanced by the fact that waste forms for those waste streams can accommodate higher waste loadings, such that similar amounts of grouted waste forms are required regardless of the recovery schedule. Similar amounts of grouted waste form are likewise needed for the individual and combined waste streams. Removing Pu, Cs, and Sr from waste streams lowers the waste form dose significantly at times beyond about 1 year after irradiation, which may benefit handling and transport. Although these calculations should be revised after experimentally optimizing the grout formulations and waste loadings, they provide initial guidance for process development.

  4. Processing of concentrated radioactive wastes into cement and bitumens following calcination

    International Nuclear Information System (INIS)

    A brief characteristic is presented of the most frequently used processes of solidification of liquid radioactive wastes, viz., bituminization, cementation and their combination with calcination. The effect of individual parameters is assessed on the choice of the type of solidification process as is their importance in the actual process, in temporary storage, during transportation and under conditions of long-term storage. It has been found that a combination of the procedures could lead to a modular system of methods and equipment. This would allow to approach optimal solidification of wastes in the present period and to establish a research reserve for the development of more modern, economically advantageous and safer procedures. A rough estimate is made of the costs of the solidification of 1 m3 of radioactive concentrate from the V-1 power plant at a production of 380 m3/year, this for the cementation-calcination and bituminization-calcination procedures. The said rough economic analysis only serves to identify the major operating components which have the greatest effect on the economic evaluation of the solidification procedures. (Z.M.)

  5. Rock alteration in alkaline cement waters over 15 years and its relevance to the geological disposal of nuclear waste

    International Nuclear Information System (INIS)

    Highlights: • Rock alteration in chemically disturbed zone of a nuclear waste geodisposal facility. • Experiments conducted for 15 years at 70 °C. • Initially formed C–S–H phases destabilise during longer-term reaction. • Dedolomitisation leads to formation of Mg–silicates, the dominant secondary phases. • Rock alteration leads to increased U(VI) sorption capacity. - Abstract: The interaction of groundwater with cement in a geological disposal facility (GDF) for intermediate level radioactive waste will produce a high pH leachate plume. Such a plume may alter the physical and chemical properties of the GDF host rock. However, the geochemical and mineralogical processes which may occur in such systems over timescales relevant for geological disposal remain unclear. This study has extended the timescale for laboratory experiments and shown that, after 15 years two distinct phases of reaction may occur during alteration of a dolomite-rich rock at high pH. In these experiments the dissolution of primary silicate minerals and the formation of secondary calcium silicate hydrate (C–S–H) phases containing varying amounts of aluminium and potassium (C–(A)–(K)–S–H) during the early stages of reaction (up to 15 months) have been superseded as the systems have evolved. After 15 years significant dedolomitisation (MgCa(CO3)2 + 2OH− → Mg(OH)2 + CaCO3 + CO32−(aq)) has led to the formation of magnesium silicates, such as saponite and talc, containing variable amounts of aluminium and potassium (Mg–(Al)–(K)–silicates), and calcite at the expense of the early-formed C–(A)–(K)–S–H phases. This occured in high pH solutions representative of two different periods of cement leachate evolution with little difference in the alteration processes in either a KOH and NaOH or a Ca(OH)2 dominated solution but a greater extent of alteration in the higher pH KOH/NaOH leachate. The high pH alteration of the rock over 15 years also increased the

  6. ANSTO's waste forms for the 31. century

    Energy Technology Data Exchange (ETDEWEB)

    Vance, E.R.; Begg, B. D.; Day, R. A.; Moricca, S.; Perera, D. S.; Stewart, M. W. A.; Carter, M. L.; McGlinn, P. J.; Smith, K. L.; Walls, P. A.; Robina, M. La

    2004-07-01

    ANSTO waste form development for high-level radioactive waste is directed towards practical applications, particularly problematic niche wastes that do not readily lend themselves to direct vitrification. Integration of waste form chemistry and processing method is emphasised. Some longstanding misconceptions about titanate ceramics are dealt with. We have a range of titanate-bearing waste form products aimed at immobilisation of tank wastes and sludges, actinide-rich wastes, INEEL calcines and Na-bearing liquid wastes, Al-rich wastes arising from reprocessing of Al-clad fuels, Mo-rich wastes arising from reprocessing of U-Mo fuels, partitioned Cs-rich wastes, and {sup 99}Tc. Waste form production techniques cover hot isostatic and uniaxial pressing, sintering, and cold-crucible melting, and these are strongly integrated into waste form design. Speciation and leach resistance of Cs and alkalis in cementitious products and geo-polymers are being studied. Recently we have embarked on studies of candidate inert matrix fuels for Pu burning. We also have a considerable program directed at basic understanding of the waste forms in regard to crystal chemistry, dissolution behaviour in aqueous media, radiation damage effects and optimum processing techniques. (authors)

  7. Radiation effects in ceramic nuclear waste forms

    International Nuclear Information System (INIS)

    This paper reports on alpha-decay event damage (a particle and recoil-nucleus) that results in atomic-scale disorder which causes changes in the molar volume, corrosion rate, stored energy, mechanical properties, and macrostructure of ceramics. These changes particularly of volume and corrosion rate, have critical implications for the long-term durability of nuclear waste forms, such as the polyphase. Ti-based ceramic Synroc. This paper reviews data on actinide-bearing (U and Th) phases of great age (>100 m.y.) found in nature and compares these results to observation on actinide-doped phases (Pu and Cm) of nearly equivalent α-decay doses. Of particular interest is evidence for annealing of radiation damage effects over geologic periods of time under ambient conditions

  8. Effect of increasing variation of rice husk ash as pozzolanic material and cement water phase on low-level radioactive waste cementation

    International Nuclear Information System (INIS)

    Research was conducted to determine the effect of rice husk ash addition as pozzolanic and to obtained the best composition. Rice husk ask has highly SiO2 as additive cement. Cementation process has been done by mixing 65 ppm liquid waste Sr (NO3)2 with Type I Portland cement, 40 mesh rice husk ash, 40 mesh sand, and water. The size of mortars was 28 mm in diameter and 56 mm high. The variation of cement water fraction was 0.2, 0.3, and 0.4. The variation of rice husk ash fraction was 10 v/0, 25 v/0, and 40 v/0. Sand variations in composition are 30 v/o, 40 v/o and 50 v/o of the cast volume with -40 mesh grain size. Variation of rice husk ash grain size was -40, -100, and -200 mesh as many as 25 v/o from mixing volume total. After that, the mortar was cured for 28 days. That samples had been pressed to get the of data compression strength of mortar. Leaching test had been done to the mortar having the lowest compressive strength value. (author)

  9. Heat of Hydration of Low Activity Cementitious Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Nasol, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-07-23

    During the curing of secondary waste grout, the hydraulic materials in the dry mix react exothermally with the water in the secondary low-activity waste (LAW). The heat released, called the heat of hydration, can be measured using a TAM Air Isothermal Calorimeter. By holding temperature constant in the instrument, the heat of hydration during the curing process can be determined. This will provide information that can be used in the design of a waste solidification facility. At the Savannah River National Laboratory (SRNL), the heat of hydration and other physical properties are being collected on grout prepared using three simulants of liquid secondary waste generated at the Hanford Site. From this study it was found that both the simulant and dry mix each had an effect on the heat of hydration. It was also concluded that the higher the cement content in the dry materials mix, the greater the heat of hydration during the curing of grout.

  10. Transportation considerations related to waste forms and canisters for Defense TRU wastes

    International Nuclear Information System (INIS)

    This report identifies and discusses the considerations imposed by transportation on waste forms and canisters for contact-handled, solid transuranic wastes from the US Department of Energy (DOE) activities. The report reviews (1) the existing raw waste forms and potential immobilized waste forms, (2) the existing and potential future DOE waste canisters and shipping containers, (3) regulations and regulatory trends for transporting commercial transuranic wastes on the ISA, (4) truck and rail carrier requirements and preferences for transporting the wastes, and (5) current and proposed Type B external packagings for transporting wastes

  11. Application of solvlent change techniques to blended cements used to immobilize low-level radioactive liquid waste

    International Nuclear Information System (INIS)

    The microstructures of hardened portland and blended cement pastes, including those being considered for use in immobilizing hazardous wastes, have a complex pore structure that changes with time. In solvent exchange, the pore structure is examined by immersing a saturated sample in a large volume of solvent that is miscible with the pore fluid. This paper reports the results of solvent replacement measurements on several blended cements mixed at a solution:solids ratio of 1.0 with alkaline solutions from the simulation of the off- gas treatment system in a vitrification facility treating low-level radioactive liquid wastes. The results show that these samples have a lower permeability than ordinary portland cement samples mixed at a water:solids ratio of 0.70, despite having a higher volume of porosity. The microstructure is changed by these alkaline solutions, and these changes have important consequences with regard to durability

  12. Waste forms, packages, and seals working group summary

    Energy Technology Data Exchange (ETDEWEB)

    Sridhar, N. [Center Antonio, TX (United States); McNeil, M.B. [Nuclear Regulatory Commission, Washington, DC (United States)

    1995-09-01

    This article is a summary of the proceedings of a group discussion which took place at the Workshop on the Role of Natural Analogs in Geologic Disposal of High-Level Nuclear Waste in San Antonio, Texas on July 22-25, 1991. The working group concentrated on the subject of radioactive waste forms and packaging. Also included is a description of the use of natural analogs in waste packaging, container materials and waste forms.

  13. Effect of temperature on the durability of class C fly ash belite cement in simulated radioactive liquid waste: synergy of chloride and sulphate ions.

    Science.gov (United States)

    Guerrero, A; Goñi, S; Allegro, V R

    2009-06-15

    The durability of class C fly ash belite cement (FABC-2-W) in simulated radioactive liquid waste (SRLW) rich in a mixed sodium chloride and sulphate solution is presented here. The effect of the temperature and potential synergic effect of chloride and sulfate ions are discussed. This study has been carried out according to the Koch-Steinegger test, at the temperature of 20 degrees C and 40 degrees C during a period of 180 days. The durability has been evaluated by the changes of the flexural strength of mortar, fabricated with this cement, immersed in a simulated radioactive liquid waste rich in sulfate (0.5M), chloride (0.5M) and sodium (1.5M) ions--catalogued like severely aggressive for the traditional Portland cement--and demineralised water, which was used as reference. The reaction mechanism of sulphate, chloride and sodium ions with the mortar was evaluated by scanning electron microscopy (SEM), porosity and pore-size distribution, and X-ray diffraction (XRD). The results showed that the chloride binding and formation of Friedel's salt was inhibited by the presence of sulphate. Sulphate ion reacts preferentially with the calcium aluminate hydrates forming non-expansive ettringite which precipitated inside the pores; the microstructure was refined and the mechanical properties enhanced. This process was faster and more marked at 40 degrees C. PMID:19056176

  14. The construction of solid waste form test and inspection facility

    International Nuclear Information System (INIS)

    The solid waste form test and inspection facility is a facility to test and inspect the characteristics of waste forms, such as homogenity, mechanical structure, thermal behaviour, water resistance and leachability. Such kinds of characteristics in waste forms are required to meet a certain conditions for long-term storage or for final disposal of wastes. The facility will be used to evaluate safety for the disposal of wastes by test and inspection. At this moment, the efforts to search the most effective management of the radioactive wastes generated from power plants and radioisotope user are being executed by the people related to this field. Therefore, the facility becomes more significant tool because of its guidance of sucessfully converting wastes into forms to give a credit to the safety of waste disposal for managing the radioactive wastes. In addition the overall technical standards for inspecting of waste forms such as the standardized equipment and processes in the facility will be estabilished in the begining of 1990's when the project of waste management will be on the stream. Some of the items of the project have been standardized for the purpose of localization. In future, this facility will be utilized not only for the inspection of waste forms but also for the periodic decontamination apparatus by remote operation techniques. (Author)

  15. Characterization of a ceramic waste form encapsulating radioactive electrorefiner salt

    Energy Technology Data Exchange (ETDEWEB)

    Moschetti, T. L.; Sinkler, W.; DiSanto, T.; Noy, M.; Warren, A. R.; Cummings, D. G.; Johnson, S. G.; Goff, K. M.; Bateman, K. J.; Frank, S. M.

    1999-11-11

    Argonne National Laboratory has developed a ceramic waste form to immobilize radioactive waste salt produced during the electrometallurgical treatment of spent fuel. This study presents the first results from electron microscopy and durability testing of a ceramic waste form produced from that radioactive electrorefiner salt. The waste form consists of two primary phases: sodalite and glass. The sodalite phase appears to incorporate most of the alkali and alkaline earth fission products. Other fission products (rare earths and yttrium) tend to form a separate phase and are frequently associated with the actinides, which form mixed oxides. Seven-day leach test results are also presented.

  16. Properties of the cements and their use in the storage systems of low-level radioactive wastes

    International Nuclear Information System (INIS)

    The use of materials containing cement has generalized in the facilities of definitive storage of radioactive wastes due to their easy handling and availability. Besides conforming the buildings and structures, these materials are part of the barriers system that will maintain the isolated radioactive wastes of the biosphere until their activity has decayed at innocuous levels. However, to fulfill this function, the effectiveness and durability of these materials should be demonstrated fully. In Mexico the intention exists of building a definitive storehouse for the low-level radioactive wastes, however are few the studies on the behavior of the materials containing cement used in this type of facilities. With the purpose of to guide and promoting the study of the national cements, in this work is made a revision of the characteristics and properties of the cements with relationship to its use in the systems of definitive storage of low-level radioactive wastes, as well as of some studies that are realized to evaluate its acting as engineering barriers. (Author)

  17. Periodic and uniform nanogratings formed on cemented carbide by femtosecond laser scanning

    International Nuclear Information System (INIS)

    Periodic and uniform nanogratings are fabricated by femtosecond laser scanning on cemented carbide. Specifically, three experiments are designed to study the influence of single pulse energy, scanning speed, and scanning spacing on the period and the uniformity of the formed nanogratings. The results show that the sample with single pulse energy of 2 μJ, scanning speed of 1000 μm/s, and scanning spacing of 5 μm shows the best quality of nanogratings among all the tested samples at different processing parameters. The uniformity of the nanogratings is largely determined by single pulse energy, scanning speed, and scanning spacing. Single pulse energy and scanning speed significantly affect the period of the nanogratings, whereas the period of the nanogratings maintains a fixed value under different scanning spacings. The period of the nanogratings increases gradually with the decrease of the single pulse energy and the increase of the scanning speed, respectively.

  18. The chemistry of blended cements and backfills intended for use in radioactive waste disposal

    International Nuclear Information System (INIS)

    This project was initiated by Her Majesty's Inspectorate of Pollution (HMIP) at the time when UK NIREX had announced its intention to develop a repository for low and intermediate level nuclear waste in the vicinity of Sellafield. In this repository setting, two main barriers existed to the return of radio-isotopes to the biosphere: the natural, or geologic and hydrogeologic barriers, and the man-made barriers. These latter comprise relatively short-lived containers as well as an engineered backfill. The backfill was designed to condition a high pH in the repository, thereby lowering the solubility of many long-lived radionuclides yet not confine gases, which might be generated from chemical and radioactive waste within the repository vault. The Environment Agency for England and Wales had already taken independent steps to examine the suitability of alkaline backfills, based on Portland cement, limestone flour and Ca(OH)2, for the man-made barriers. Preliminary data on post-closure repository performance assessment at Sellafield suggested the importance of two additional factors which had not hitherto been considered in assessments: (i) temperature: Inclusion of heat generating waste could drive temperatures up to ∼80 deg. C in the post closure phase; (ii) salinity of deep groundwater: Much previous work has been done in initially-pure water but borehole analyses indicated high salinity at depth. Other potential deep repositories could also be saline. These impacts were likely to occur together throughout much of the post-closure phase: backfills were likely to be in prolonged contact with hot, saline groundwater. Previous studies demonstrated that cements achieve their performance by a sacrificial action. It is however essential that the cementitious materials should not dissolve too rapidly if prolonged backfill performance lifetimes are to be achieved. By dissolving cement backfills condition permeating water to a high pH and thereby lower the solubilities of

  19. Phosphate bonded ceramics as candidate final-waste-form materials

    International Nuclear Information System (INIS)

    Room-temperature setting phosphate-bonded ceramics were studied as candidate materials for stabilization of DOE low-level problem mixed wastes which cannot be treated by other established stabilization techniques. Phosphates of Mg, Mg-Na, Al and Zr were studied to stabilize ash surrogate waste containing RCRA metals as nitrates and RCRA organics. We show that for a typical loading of 35 wt.% of the ash waste, the phosphate ceramics pass the TCLP test. The waste forms have high compression strength exceeding ASTM recommendations for final waste forms. Detailed X-ray diffraction studies and differential thermal analyses of the waste forms show evidence of chemical reaction of the waste with phosphoric acid and the host matrix. The SEM studies show evidence of physical bonding. The excellent performance in the leaching tests is attributed to a chemical solidification and physical as well as chemical bonding of ash wastes in these phosphate ceramics

  20. Physical modeling of contaminant diffusion from a cementious waste form

    International Nuclear Information System (INIS)

    Cementitious materials can be used to immobilize waste materials for disposal. The Westinghouse Hanford Company is pursuing approval of disposal technologies by which hazardous and radioactive wastes are blended or packaged with cementitious materials for disposal. Of significant concern is the mobility of the waste contaminants both from the waste form and in the arid soils of the Hanford Site. A physical model has been developed to study the diffusion of waste contaminants from simulated cementitious waste forms in unsaturated Hanford Site soils. The model can be used to predict cementitious waste form performance in a representative environment, support design of waste management facilities and technologies, and provide data for environmental permitting of proposed treatment and disposal facilities

  1. Variations and factors that influence the formation of polychlorinated naphthalenes in cement kilns co-processing solid waste.

    Science.gov (United States)

    Jin, Rong; Zhan, Jiayu; Liu, Guorui; Zhao, Yuyang; Zheng, Minghui

    2016-09-01

    Pilot studies of unintentionally produced pollutants should be performed before waste being co-processed in cement kilns. Polychlorinated naphthalene (PCN) formation and emission from cement kilns co-processing sorted municipal solid waste, sewage sludge, and waste acid, however, have not previously been studied. Here, PCNs were analyzed in stack gas samples and solid samples from different stages of three cement production runs. PCN destruction efficiencies were higher when waste was co-processed (93.1% and 88.7% in two tests) than when waste was not co-processed (39.1%), so co-processing waste would not increase PCN outputs. The PCN concentrations were higher in particle samples from the C1 preheater and stages at back end of kiln than in particle samples from other stages, suggesting that cyclone preheater and back end of kiln should be focused for controlling PCN emissions. Besides that, based on the variation of PCN concentrations and corresponding operating conditions in different stages, the temperature, feeding materials, and chlorine content were suggested as the main factors influencing PCN formation. The PCN homologue and congener profiles suggested chlorination and dechlorination were the main PCN formation and decomposition pathways, and congeners CN-23, CN-46, and CN-59 appear to be appropriate indicators of PCNs emitted from coal-burning sources. PMID:27187059

  2. Alternate nuclear waste forms and interactions in geologic media

    International Nuclear Information System (INIS)

    The primary purposes of the conference on Alternate Nuclear Waste Forms and Interactions in Geologic Media were: First, to provide an opportunity for a review of the status of the research on some of the candidate alternative waste forms; second, to provide an opportunity for comparing the characteristics of alternate waste forms to those of glasses; and third, to stimulate increased interactions between those research groups that were engaged in a more basic approach to characterizing waste forms and those who were concerned with more applied aspects such as the processing of these materials. The motivating philosophy behind this third purpose of the conference was based on the idea that by operating from the soundest possible fundamental base for any of the candidate waste forms, hopefully any future unpleasant surprise - such as that alluded to earlier in the case of glass waste forms - could be avoided. Separate abstracts have been prepared for individual papers for inclusion in the Energy Data Base

  3. DURABILITY TESTING OF FLUIDIZED BED STEAM REFORMER (FBSR) WASTE FORMS

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C

    2006-01-06

    Fluidized Bed Steam Reforming (FBSR) is being considered as a potential technology for the immobilization of a wide variety of high sodium aqueous radioactive wastes. The addition of clay and a catalyst as co-reactants converts high sodium aqueous low activity wastes (LAW) such as those existing at the Hanford and Idaho DOE sites to a granular ''mineralized'' waste form that may be made into a monolith form if necessary. Simulant Hanford and Idaho high sodium wastes were processed in a pilot scale FBSR at Science Applications International Corporation (SAIC) Science and Technology Applications Research (STAR) facility in Idaho Falls, ID. Granular mineral waste forms were made from (1) a basic Hanford Envelope A low-activity waste (LAW) simulant and (2) an acidic INL simulant commonly referred to as sodium-bearing waste (SBW). The FBSR waste forms were characterized and the durability tested via ASTM C1285 (Product Consistency Test), the Environmental Protection Agency (EPA) Toxic Characteristic Leaching Procedure (TCLP), and the Single Pass Flow Through (SPFT) test. The durability of the FBSR waste form products was tested in order to compare the measured durability to previous FBSR waste form testing on Hanford Envelope C waste forms that were made by THOR Treatment Technologies (TTT) and to compare the FBSR durability to vitreous LAW waste forms, specifically the Hanford low activity waste (LAW) glass known as the Low-activity Reference Material (LRM). The durability of the FBSR waste form is comparable to that of the LRM glass for the test responses studied.

  4. Assessment of radiation effects in defense transuranic waste forms

    International Nuclear Information System (INIS)

    The actinide concentrations of the defense transuranic (TRU) wastes were analyzed and the potential effects of the radiation on the properties of the wastes after conversion to immobile forms were assessed. The study focused on the contact-handled retrievably-stored wastes. The major components of the current inventory are defense plutonium-contaminated wastes containing various amounts of 241Am. The wastes stored at Idaho National Engineering Laboratory (INEL) are typical of the wastes in this category. There is also a substantial amount of wastes contaminated with plutonium enriched in 238Pu arising from the Department of Energy's isotopic heat-source programs. Most of these wastes are stored at the Los Alamos National Laboratory (LANL) and the Savannah River Plant (SRP) sites. Four reference wastes were selected representing a credible range of actinide activities and were used for estimating radiation doses to the final waste forms based on: INEL first stage sludge, a composite of all wastes at INEL, a composite of all wastes at LANL including both defense and heat source plutonium wastes, and a composite of all heat-source plutonium wastes at SRP free from defense plutonium. From integrated alpha and beta-gamma doses over a 105y storage period, it is concluded that: accumulated beta-gamma doses of 106 to 108 rad over a 105 y storage period will not significantly change physical properties of the waste form. The alpha decay doses are 3-30 x 1016 decays/cm3 (up to 3 x 1010 rad ionizing radiation from alpha particles) accumulated over a 105 y storage period. Radiolytic gas generation can be substantial in waste forms containing water or organic materials. The effect of leachant radiolysis on the leachabilities of TRU waste forms is not sufficiently understood to rule out the need for testing with actinide-doped specimens

  5. Solidification of ion exchange resin wastes

    International Nuclear Information System (INIS)

    Solidification media investigated included portland type I, portland type III and high alumina cements, a proprietary gypsum-based polymer modified cement, and a vinyl ester-styrene thermosetting plastic. Samples formulated with hydraulic cement were analyzed to investigate the effects of resin type, resin loading, waste-to-cement ratio, and water-to-cement ratio. The solidification of cation resin wastes with portland cement was characterized by excessive swelling and cracking of waste forms, both after curing and during immersion testing. Mixed bed resin waste formulations were limited by their cation component. Additives to improve the mechanical properties of portland cement-ion exchange resin waste forms were evaluated. High alumina cement formulations dislayed a resistance to deterioration of mechanical integrity during immersion testing, thus providing a significant advantage over portland cements for the solidification of resin wastes. Properties of cement-ion exchange resin waste forms were examined. An experiment was conducted to study the leachability of 137Cs, 85Sr, and 60Co from resins modified in portland type III and high alumina cements. The cumulative 137Cs fraction release was at least an order of magnitude greater than that of either 85Sr or 60Co. Release rates of 137Cs in high alumina cement were greater than those in portland III cement by a factor of two.Compressive strength and leach testing were conducted for resin wastes solidified with polymer-modified gypsum based cement. 137Cs, 85Sr, and 60Co fraction releases were about one, two and three orders of magnitude higher, respectively, than in equivalent portland type III cement formulations. As much as 28.6 wt % dry ion exchange resin was successfully solidified using vinyl ester-styrene compared with a maximum of 25 wt % in both portland and gypsum-based cement

  6. Contribution to the study of wastes stabilization by sulfo-aluminate cement

    International Nuclear Information System (INIS)

    Calcium sulfo-aluminate cement is mainly composed of yeelimite known to be a precursor of ettringite formation. Ettringite is able to incorporate several heavy metals by isomorphous substitutions without altering its crystalline structure. The design of a binder required for immobilizing heavy metals was undertaken. The hydration study of clinker, and cement containing 4 amounts of gypsum has been carried out by means of XRD, DTA and IR spectrometry. It was pointed out that the addition of gypsum enhances hydration. Two binders were selected: 80/20 and 70/30. The immobilisation of 7 pollutants was very successful. Nevertheless, damages appeared with the binder 70/30 containing sodium chromate and dichromate: sodium caused activation of yeelimite reactivity and important dissolution of gypsum leading to important ettringite production. With a great amount of gypsum (30 %), dissolution led to secondary ettringite formation which damaged the hardened paste. Adding polyol enhances the retention of sodium chromate. On the other hand, the immobilisation of two types of weakly radioactive wastes supplied by CEA has been made. Results obtained in terms of setting time, compressive strength and leaching were excellent. (author)

  7. Improvement, characterization and use of waste corn cob ash in cement-based materials

    Science.gov (United States)

    Suwanmaneechot, P.; Nochaiya, T.; Julphunthong, P.

    2015-12-01

    This work investigates the development of waste corn cob ash as supplementary cement replacement materials. The study focused on the effects of heat treatment on chemical composition, physical properties and engineering properties of corn cob ash. The results suggest corn cob ash that was heat treated at 600°C for 4 h shows percentage of SiO2 + Al2O3 + Fe2O3 around 72%, which can be classified as Class N calcined natural pozzolan, as prescribed by ASTM C618. The X-ray diffraction patterns indicated that the amorphous silica phase increased with increasing calcining temperatures. The water requirement, initial setting time and final setting time of specimens increased with increasing replacement percentage of raw or treated corn cob ash. The morta cubes which used 20% of treated corn cob ash replaced cement showed 103% of the 28 days compressive strength as compared to reference samples. The corn cob ash that was treated at 600°C for 4 h samples shows slightly higher effectiveness for improving the splitting tensile strength and compressive strength of concrete when compared to the untreated corn cob ash.

  8. Final waste forms project: Performance criteria for phase I treatability studies

    International Nuclear Information System (INIS)

    This document defines the product performance criteria to be used in Phase I of the Final Waste Forms Project. In Phase I, treatability studies will be performed to provide open-quotes proof-of-principleclose quotes data to establish the viability of stabilization/solidification (S/S) technologies. This information is required by March 1995. In Phase II, further treatability studies, some at the pilot scale, will be performed to provide sufficient data to allow treatment alternatives identified in Phase I to be more fully developed and evaluated, as well as to reduce performance uncertainties for those methods chosen to treat a specific waste. Three main factors influence the development and selection of an optimum waste form formulation and hence affect selection of performance criteria. These factors are regulatory, process-specific, and site-specific waste form standards or requirements. Clearly, the optimum waste form formulation will require consideration of performance criteria constraints from each of the three categories. Phase I will focus only on the regulatory criteria. These criteria may be considered the minimum criteria for an acceptable waste form. In other words, a S/S technology is considered viable only if it meet applicable regulatory criteria. The criteria to be utilized in the Phase I treatability studies were primarily taken from Environmental Protection Agency regulations addressed in 40 CFR 260 through 265 and 268; and Nuclear Regulatory Commission regulations addressed in 10 CFR 61. Thus the majority of the identified criteria are independent of waste form matrix composition (i.e., applicable to cement, glass, organic binders etc.)

  9. Final waste forms project: Performance criteria for phase I treatability studies

    Energy Technology Data Exchange (ETDEWEB)

    Gilliam, T.M. [Oak Ridge National Lab., TN (United States); Hutchins, D.A. [Martin Marietta Energy Systems, Inc., Oak Ridge, TN (United States); Chodak, P. III [Massachusetts Institute of Technology (United States)

    1994-06-01

    This document defines the product performance criteria to be used in Phase I of the Final Waste Forms Project. In Phase I, treatability studies will be performed to provide {open_quotes}proof-of-principle{close_quotes} data to establish the viability of stabilization/solidification (S/S) technologies. This information is required by March 1995. In Phase II, further treatability studies, some at the pilot scale, will be performed to provide sufficient data to allow treatment alternatives identified in Phase I to be more fully developed and evaluated, as well as to reduce performance uncertainties for those methods chosen to treat a specific waste. Three main factors influence the development and selection of an optimum waste form formulation and hence affect selection of performance criteria. These factors are regulatory, process-specific, and site-specific waste form standards or requirements. Clearly, the optimum waste form formulation will require consideration of performance criteria constraints from each of the three categories. Phase I will focus only on the regulatory criteria. These criteria may be considered the minimum criteria for an acceptable waste form. In other words, a S/S technology is considered viable only if it meet applicable regulatory criteria. The criteria to be utilized in the Phase I treatability studies were primarily taken from Environmental Protection Agency regulations addressed in 40 CFR 260 through 265 and 268; and Nuclear Regulatory Commission regulations addressed in 10 CFR 61. Thus the majority of the identified criteria are independent of waste form matrix composition (i.e., applicable to cement, glass, organic binders etc.).

  10. Stabilization/solidification (S/S) of mercury-contaminated hazardous wastes using thiol-functionalized zeolite and Portland cement.

    Science.gov (United States)

    Zhang, Xin-Yan; Wang, Qi-Chao; Zhang, Shao-Qing; Sun, Xiao-Jing; Zhang, Zhong-Sheng

    2009-09-15

    Stabilization/solidification (S/S) of mercury-containing solid wastes using thiol-functionalized zeolite and cement was investigated in this study. The thiol-functionalized zeolite (TFZ) used in the study was obtained by grafting the thiol group (-SH) to the natural clinoptilolite zeolites, and the mercury adsorption by TFZ was investigated. TFZ was used to stabilize mercury in solid wastes, and then the stabilized wastes were subjected to cement solidification to test the effectiveness of the whole S/S process. The results show that TFZ has a high level of -SH content (0.562 mmol g(-1)) and the adsorption of mercury by TFZ conform to the Freundlich adsorption isotherm. The mercury adsorption capacity is greatly enhanced upon thiol grafting, the maximum of which is increased from 0.041 mmol Hg g(-1) to 0.445 mmol Hg g(-1). TFZ is found to be effective in stabilizing Hg in the waste surrogate. In the stabilization process, the optimum pH for the stabilization reaction is about 5.0. The optimum TFZ dosage is about 5% and the optimum cement dosage is about 100%. Though Cl(-) and PO(4)(3-) have negative effects on mercury adsorption by TFZ, the Portland cement solidification of TFZ stabilized surrogates containing 1000 mg Hg/kg can successfully pass the TCLP leaching test. It can be concluded that the stabilization/solidification process using TFZ and Portland cement is an effective technology to treat and dispose mercury-containing wastes. PMID:19376646

  11. Waste-Form Development Program. Annual progress report, October 1981-September 1982

    International Nuclear Information System (INIS)

    Low-level wastes (LLW) at nuclear facilities have traditionally been solidified using portland cement (with and without additives). Urea-formaldehyde has been used for LLW solidification while bitumen (asphalt) and thermosetting polymers will be applied to domestic wastes in the near future. Operational difficulties have been observed with each of these solidification agents. Such difficulties include incompatibility with waste constitutents inhibiting solidification, premature setting, free standing water and fires. Some specific waste types have proven difficult to solidify with one or more of the contemporary agents. Similar problems are also anticipated for the solidification of new wastes, which are generated using advanced volume reduction technologies, and with the application of additional agents which may be introduced in the near future for the solidification of LLW. In the Waste Form Development program, contemporary solidification agents are being investigated relative to their potential applications to major fuel cycle and non-fuel cycle LLW streams. The range of conditions under which these solidification agents can be satisfactorily applied to specific LLW streams is being determined. These studies are primarily directed towards defining operating parameters for both improved solidification of problem wastes such as ion exchange resins, organic liquids and oils for which prevailing processes, as currently employed, appear to be inadequate, and solidification of new LLW streams including high solids content evaporator concentrates, dry solids, and incinerator ash generated from advanced volume reduction technologies. Solidified waste forms are tested and evaluated to demonstrate compliance with waste form performance and shallow land burial (SLB) acceptance criteria and transportation requirements (both as they currently exist and as they are anticipated to be modified with time)

  12. Application-ready cement recipes for solidification of low and medium level radioactive wastes from nuclear power plants

    International Nuclear Information System (INIS)

    The R+D program for solidification of low and medium level wastes from Swiss power plants is carried out at EIR (Swiss Federal Institute for Reactor Research) under contract to Nagra (National Cooperative for the Storage of Radioactive Waste). Up to now, 7 practicable recipes for these waste types have been developed for use. With these recipes the main quantities of all possible reactor radwastes arising from operating Swiss power plants can be solidified with cements. The solidified matrixes then produced fulfill the conditions required by the Swiss regulatory authorities. As a first barrier, they contribute to a safe final storage of these wastes in geological formations in Switzerland. For new waste types arising, for 'special wastes', and for the wastes from power stations not yet commissioned, the R+D work is going on. Future results will probably allow further improvement of the recipes and processes already developed. (author)

  13. Hot-pressed barium sulphate ceramic waste forms for direct immobilization of medium level Magnox waste

    International Nuclear Information System (INIS)

    A possible method of treatment for Magnox cladding waste is by dissolution in nitric acid and precipitation of barium sulphate-based floc with which radioactive ions are co-precipitated. The floc could then be immobilized in a matrix material such as cement or bitumen to give the waste form, or alternatively can be converted directly into a waste form by hot pressing. This paper describes the direct conversion of barium sulphate floc, containing simulated radwaste, into a synthetic, ceramic version of the natural mineral barite by a hot-pressing route. By variation of the parameters pressure, temperature and time, optimum conditions for consolidation of the floc to > 90% theoretical density on a laboratory scale are found to be 22.5 MPa, 9000C for 10 minutes. Using a pressure of 15 MPa, at 9000C for 30 min., hot-pressed billets of BaSO4 have been made on a 5 kg scale. In going from the magnox waste to the hot-pressed barium sulphate a volume reduction factor approx. 18 is achieved. The principal phases in the product are found to be BaSO4, MgO and Fe3O4, and the degree of consolidation achieved depends on the MgO content. The leaching behaviour of the hot-pressed materials in 1000C, 3 day Soxhlet tests also depends on the MgO content, and on the consequent level of open porosity. If there is porosity accessible to the leach water, MgO at the internal surfaces is converted to Mg(OH)2, which deposits within the pores, and a weight gain is registered in the Soxhlet test. If, however, there is no open porosity, a weight loss occurs, and leach rates approx. 4 x 10-7 kg/m2/sec are found. In contrast, pure BaSO4, hot-pressed to similar densities, shows no variation in leaching behaviour over a wide range of open porosities, and gives Soxhlet leach rates approx. 8 x 10-8 kg/m2/sec. 6 figures, 2 tables

  14. Sulfur polymer cement, a solidification and stabilization agent for radioactive and hazardous wastes

    International Nuclear Information System (INIS)

    Sulfur polymer cement (SPC) is made by reacting 95% sulfur with 2.5 % dicyclopentadiene and 2.5% cyclopentadiene oligomers, to produce a product that is much better than unmodified sulfur. SPC is being tested as a solidifying and stabilizing agent for low-level radioactive and hazardous wastes. Heavy loadings (5 wt%) of eight toxic metals were combined individually with SPC and 7 wt% sodium sulfide nonahydrate. The leach rates for mercury, lead, chromium and silver oxides were reduced by six orders of magnitude, while those of arsenic and barium were reduced by four. SPC is good for stabilizing incinerator ash. Ion-exchange resins can be stabilized with SPC after heat treatment with asbestos or diatomite at 220-250 deg C. 19 refs

  15. Mobile calcination and cementation unit for solidification of concentrated radioactive wastes

    International Nuclear Information System (INIS)

    Mobile experimental unit MESA-1 was developed and manufactured for processing radioactive concentrates by direct cementation. The unit is mainly designed for processing low-level liquid wastes from nuclear power plants and other nuclear installations, in which the level of radioactivity does not exceed 1010 Bq/m3, the salt content of liquid solutions does not exceed 500 kg/m3 and the maximum amount of boric acid is 130 kg/m3. The equipment is built into three modules which may be assembled and dismantled in a short time and transported separately. The unit without the calciner module was tested in non-radioactive mode and in operation with actual radioactive wastes from the V-1 nuclear power plant. The course and results of the tests are described in detail. All project design values were achieved, a total of 18 dm3 model solutions were processed and 1 m3 of actual wastes with a salt content of 450 kg/m3. The test showed that with regard to the radiation level reached it will be necessary in the process of calcination to increase the shielding of certain exposed points. The calciner module is being assembled for completion. (Z.M.)

  16. DSNF AND OTHER WASTE FORM DEGRADATION ABSTRACTION

    International Nuclear Information System (INIS)

    Several hundred distinct types of DOE-owned spent nuclear fuel (DSNF) may potentially be disposed in the Yucca Mountain repository. These fuel types represent many more types than can be viably individually examined for their effect on the Total System Performance Assessment for the License Application (TSPA-LA). Additionally, for most of these fuel types, there is no known direct experimental test data for the degradation and dissolution of the waste form in repository groundwaters. The approach used in the TSPA-LA model is, therefore, to assess available information on each of 11 groups of DSNF, and to identify a model that can be used in the TSPA-LA model without differentiating between individual codisposal waste packages containing different DSNF types. The purpose of this report is to examine the available data and information concerning the dissolution kinetics of DSNF matrices for the purpose of abstracting a degradation model suitable for use in describing degradation of the DSNF inventory in the Total System Performance Assessment for the License Application. The data and information and associated degradation models were examined for the following types of DSNF: Group 1--Naval spent nuclear fuel; Group 2--Plutonium/uranium alloy (Fermi 1 SNF); Group 3--Plutonium/uranium carbide (Fast Flux Test Facility-Test Fuel Assembly SNF); Group 4--Mixed oxide and plutonium oxide (Fast Flux Test Facility-Demonstration Fuel Assembly/Fast Flux Test Facility-Test Demonstration Fuel Assembly SNF); Group 5--Thorium/uranium carbide (Fort St. Vrain SNF); Group 6--Thorium/uranium oxide (Shippingport light water breeder reactor SNF); Group 7--Uranium metal (N Reactor SNF); Group 8--Uranium oxide (Three Mile Island-2 core debris); Group 9--Aluminum-based SNF (Foreign Research Reactor SNF); Group 10--Miscellaneous Fuel; and Group 11--Uranium-zirconium hydride (Training Research Isotopes-General Atomics SNF). The analyses contained in this document provide an ''upper-limit'' (i

  17. Ceramic waste form qualification using results from witness tubes

    International Nuclear Information System (INIS)

    A ceramic waste form has been developed to immobilize the salt waste stream from electrometallurgical treatment of spent nuclear fuel. The ceramic waste form is prepared in a hot isostatic press (HIP). The use of small, easily fabricated HIP capsules called witness tubes has been proposed as a practical way to obtain representative samples of ceramic waste form material for process monitoring, waste form qualification, and archiving. Witness tubes are filled with the same material used to fill the corresponding HIP can, and are HIPed along with the HIP can. Relevant physical, chemical, and performance (leach test) data are analyzed and compared. Differences between witness tube and HIP can materials are shown to be statistically insignificant, demonstrating that witness tubes do provide ceramic waste form material representative of the material in the corresponding HIP can.

  18. CHEMICALLY BONDED CEMENTS FROM BOILER ASH AND SLUDGE WASTES. PHASE II REPORT, SEPT.1998-JULY 1999.

    Energy Technology Data Exchange (ETDEWEB)

    SUGAMA,T.YAGER,K.A.BLANKENHORN,D.(KEYSPAN R AND D INITIATIVE)

    1999-08-01

    Based upon the previous Phase I research program aimed at looking for ways of recycling the KeySpan-generated wastes, such as waste water treatment sludge (WWTS) and bottom ash (BA), into the potentially useful cementitious materials called chemically bonded cement (CBC) materials, the emphasis of this Phase II program done at Brookhaven National Laboratory, in a period of September 1998 through July 1999, was directed towards the two major subjects: One was to assess the technical feasibility of WWTS-based CBC material for use as Pb-exchange adsorbent (PEA) which remediates Pb-contaminated soils in the field; and the other was related to the establishment of the optimum-packaging storage system of dry BA-based CBC components that make it a promising matrix material for the steam-cured concrete products containing sand and coarse aggregate. To achieve the goal of the first subject, a small-scale field demonstration test was carried out. Using the PEA material consisting of 30 wt% WWTS, 13 wt% Type I cement and 57 wt% water, the PES slurry was prepared using a rotary shear concrete mixer, and then poured on the Pb-contaminated soil. The PEA-to-soil ratio by weight was a factor of 2.0. The placed PEA slurry was blended with soil using hand mixing tools such as claws and shovels. The wettability of soils with the PEA was very good, thereby facilitating the soil-PEA mix procedures. A very promising result was obtained from this field test; in fact, the mount of Pb leached out from the 25-day-aged PEA-treated soil specimen was only 0.74 mg/l, meeting the requirement for EPA safe regulation of < 5 mg/l. In contrast, a large amount (26.4 mg/l) of Pb was detected from the untreated soil of the same age. Thus, this finding demonstrated that the WWTS-based CBC has a potential for use as PEA material. Regarding the second subject, the dry-packed storage system consisting of 68.7 wt% BA, 13.0 wt% calcium aluminate cement (CAC), 13.0 wt% Type I portland cement and 5.3 wt

  19. Non-destructive evaluation of curing effect on the quality of cement-based engineered barrier for radioactive waste disposal

    International Nuclear Information System (INIS)

    Cement-based engineered barrier for use of geological underground disposal of high-level radioactive wastes must be assured for long-time maintenance. The present report concerns the curing effect and its adequate non-destructive evaluation methods. For barrier materials with low-heat portland cement and fly-ash (LPC-FA), retarding the demolded-material age from 7-days to 15- and further to 28-days was found to effectively improve the material quality. Furthermore, surface permeability test was found to be used for non-destructive evaluation method of the quality of the engineered-barrier. (S. Ohno)

  20. 水泥窑处理工业废物的工厂实验研究%Plant Test of Industrial Waste Disposal in a Cement Kiln

    Institute of Scientific and Technical Information of China (English)

    刘阳生; 韩杰; 白庆中

    2003-01-01

    Destruction of industrial waste in cement rotary kilns (CRKs) is an alternative technology for thetreatment of certain types of industrial waste (IW). In this paper, three typical types of industrial wastes wereco-incinerated in the CRK at Beijing Cement Plant to determine the effects of waste disposal (especially solid wastedisposal) on the quality of clinker and the concentration of pollutants in air emission. Experimental results showthat (1) waste disposal does not affect the quality of clinker and fly ash, and fly ash after the IW disposal can still beused in the cement production, (2) heavy metals from IW are immobilized and stabilized in the clinker and cement,and (3) concentration of pollutants in air emission is far below than the permitted values in the China NationalStandard-Air Pollutants Emission Standard (GB 16297-1996).

  1. Plutonium waste fixation

    International Nuclear Information System (INIS)

    A process for the containment of radioactive wastes involves fixing the wastes in a high-alumina cement, firing to form an integral solid body and glazing the surface of the solid body. A clay can be included with the cement to assist the glazing step. (author)

  2. Experimental research on the strength of cemented backfilling body of waste rocks%废石尾砂胶结充填体强度试验研究

    Institute of Scientific and Technical Information of China (English)

    罗根平; 乔登攀

    2015-01-01

    Experimental study is systematically conducted on cemented backfilling with waste rocks.The paper states the applicability and mechanism of waste rock cemented filling process and focuses on the influencing factors on the strength of cemented filling body of waste rocks,namely the water-cement ratio,cement-sand ratio,cement content, the grading and proportioning of the particle size of waste rocks.The research results show that the lager the water-ce-ment ratio and cement-sand ratio are,the less the strength of cemented backfilling body becomes,contrary to that rela-tion between cement content and the backfilling body's strength.With constant strength,cemented filling with waste rocks consumes less cement per unit volume and cost less than other filling methods.%对废石尾砂胶结充填进行了系统的试验研究。阐述了废石尾砂胶结充填工艺的工业性及原理,着重研究了废石尾砂胶结充填体强度的影响因素:水灰比、灰砂比、水泥含量、废石尾砂的粒径级配及配比。研究结果表明,废石尾砂胶结充填体强度随水灰比、灰砂比的减小而增大,随水泥含量的增加而增加。在强度一定的条件下,废石尾砂胶结充填比其他充填方式,单位体积内水泥耗量少,成本低。

  3. Characterization of composite ceramic high level waste forms

    International Nuclear Information System (INIS)

    Argonne National Laboratory has developed a composite ceramic waste form for the disposition of high level radioactive waste produced during electrometallurgical conditioning of spent nuclear fuel. The electrorefiner LiCl/KCl eutectic salt, containing fission products and transuranics in the chloride form, is contacted with a zeolite material which removes the fission products from the salt. After salt contact, the zeolite is mixed with a glass binder. The zeolite/glass mixture is then hot isostatic pressed (HIPed) to produce the composite ceramic waste form. The ceramic waste form provides a durable medium that is well suited to incorporate fission products and transuranics in the chloride form. Presented are preliminary results of the process qualification and characterization studies, which include chemical and physical measurements and product durability testing, of the ceramic waste form

  4. Crystalline Ceramic Waste Forms: Comparison Of Reference Process For Ceramic Waste Form Fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Brinkman, K. S. [Savannah River National Laboratory; Marra, J. C. [Savannah River National Laboratory; Amoroso, J. [Savannah River National Laboratory; Tang, M. [Los Alamos National Laboratory

    2013-08-22

    The research conducted in this work package is aimed at taking advantage of the long term thermodynamic stability of crystalline ceramics to create more durable waste forms (as compared to high level waste glass) in order to reduce the reliance on engineered and natural barrier systems. Durable ceramic waste forms that incorporate a wide range of radionuclides have the potential to broaden the available disposal options and to lower the storage and disposal costs associated with advanced fuel cycles. Assemblages of several titanate phases have been successfully demonstrated to incorporate radioactive waste elements, and the multiphase nature of these materials allows them to accommodate variation in the waste composition. Recent work has shown that they can be produced from a melting and crystallization process. The objective of this report is to explore the phase formation and microstructural differences between lab scale melt processing in varying gas environments with alternative densification processes such as Hot Pressing (HP) and Spark Plasma Sintering (SPS). The waste stream used as the basis for the development and testing is a simulant derived from a combination of the projected Cs/Sr separated stream, the Trivalent Actinide - Lanthanide Separation by Phosphorous reagent Extraction from Aqueous Komplexes (TALSPEAK) waste stream consisting of lanthanide fission products, the transition metal fission product waste stream resulting from the transuranic extraction (TRUEX) process, and a high molybdenum concentration with relatively low noble metal concentrations. Melt processing as well as solid state sintering routes SPS and HP demonstrated the formation of the targeted phases; however differences in microstructure and elemental partitioning were observed. In SPS and HP samples, hollandite, pervoskite/pyrochlore, zirconolite, metallic alloy and TiO{sub 2} and Al{sub 2}O{sub 3} were observed distributed in a network of fine grains with small residual pores

  5. Crystalline Ceramic Waste Forms: Comparison Of Reference Process For Ceramic Waste Form Fabrication

    International Nuclear Information System (INIS)

    The research conducted in this work package is aimed at taking advantage of the long term thermodynamic stability of crystalline ceramics to create more durable waste forms (as compared to high level waste glass) in order to reduce the reliance on engineered and natural barrier systems. Durable ceramic waste forms that incorporate a wide range of radionuclides have the potential to broaden the available disposal options and to lower the storage and disposal costs associated with advanced fuel cycles. Assemblages of several titanate phases have been successfully demonstrated to incorporate radioactive waste elements, and the multiphase nature of these materials allows them to accommodate variation in the waste composition. Recent work has shown that they can be produced from a melting and crystallization process. The objective of this report is to explore the phase formation and microstructural differences between lab scale melt processing in varying gas environments with alternative densification processes such as Hot Pressing (HP) and Spark Plasma Sintering (SPS). The waste stream used as the basis for the development and testing is a simulant derived from a combination of the projected Cs/Sr separated stream, the Trivalent Actinide - Lanthanide Separation by Phosphorous reagent Extraction from Aqueous Komplexes (TALSPEAK) waste stream consisting of lanthanide fission products, the transition metal fission product waste stream resulting from the transuranic extraction (TRUEX) process, and a high molybdenum concentration with relatively low noble metal concentrations. Melt processing as well as solid state sintering routes SPS and HP demonstrated the formation of the targeted phases; however differences in microstructure and elemental partitioning were observed. In SPS and HP samples, hollandite, pervoskite/pyrochlore, zirconolite, metallic alloy and TiO2 and Al2O3 were observed distributed in a network of fine grains with small residual pores. The titanate phases

  6. Development of a novel fluorapatite-forming calcium phosphate cement with calcium silicate: in vitro and in vivo characteristics.

    Science.gov (United States)

    Suzuki, Yusuke; Hayashi, Makoto; Yasukawa, Takuya; Kobayashi, Hiroshi; Makino, Kosuke; Hirano, Yoriyuki; Takagi, Shozo; Chow, Laurence C; Ogiso, Bunnai

    2015-01-01

    Aim of this study was to develop a novel fluorapatite-forming calcium phosphate cement (FA-CPC) with tricalcium silicate (TCS) for endodontic applications and to examine its in vitro and in vivo characteristics. The FA-CPC powder consisted of 62.8% CaHPO4, 30.8% CaCO3, and 6.4% NaF. One part of TCS was combined with 9 parts of FA-CPC powder (FA-CPC with TCS). A 1.5 M phosphate solution was used as cement liquid. Setting time (ST), diametral tensile strength (DTS), phase composition by X-ray diffraction (XRD), and cement alkalinity were analyzed. Cement biocompatibility was assessed using rat subcutaneous model. Cement ST was 10.3±0.6 min and DTS was 3.89±0.76 MPa. XRD patterns showed that highly crystalline apatitic material was the only significant phase present and pH value was approximate 11.0. FA-CPC with TCS demonstrated similar biocompatibility as that of mineral trioxide aggregate control. These results suggest that FA-CPC with TCS may be useful for endodontic applications. PMID:25740309

  7. Final report on cermet high-level waste forms

    International Nuclear Information System (INIS)

    Cermets are being developed as an alternate method for the fixation of defense and commercial high level radioactive waste in a terminal disposal form. Following initial feasibility assessments of this waste form, consisting of ceramic particles dispersed in an iron-nickel base alloy, significantly improved processing methods were developed. The characterization of cermets has continued through property determinations on samples prepared by various methods from a variety of simulated and actual high-level wastes. This report describes the status of development of the cermet waste form as it has evolved since 1977. 6 tables, 18 figures

  8. DSNF AND OTHER WASTE FORM DEGRADATION ABSTRACTION

    Energy Technology Data Exchange (ETDEWEB)

    J. CUNNANE

    2004-11-19

    Several hundred distinct types of DOE-owned spent nuclear fuel (DSNF) may potentially be disposed in the Yucca Mountain repository. These fuel types represent many more types than can be viably individually examined for their effect on the Total System Performance Assessment for the License Application (TSPA-LA). Additionally, for most of these fuel types, there is no known direct experimental test data for the degradation and dissolution of the waste form in repository groundwaters. The approach used in the TSPA-LA model is, therefore, to assess available information on each of 11 groups of DSNF, and to identify a model that can be used in the TSPA-LA model without differentiating between individual codisposal waste packages containing different DSNF types. The purpose of this report is to examine the available data and information concerning the dissolution kinetics of DSNF matrices for the purpose of abstracting a degradation model suitable for use in describing degradation of the DSNF inventory in the Total System Performance Assessment for the License Application. The data and information and associated degradation models were examined for the following types of DSNF: Group 1--Naval spent nuclear fuel; Group 2--Plutonium/uranium alloy (Fermi 1 SNF); Group 3--Plutonium/uranium carbide (Fast Flux Test Facility-Test Fuel Assembly SNF); Group 4--Mixed oxide and plutonium oxide (Fast Flux Test Facility-Demonstration Fuel Assembly/Fast Flux Test Facility-Test Demonstration Fuel Assembly SNF); Group 5--Thorium/uranium carbide (Fort St. Vrain SNF); Group 6--Thorium/uranium oxide (Shippingport light water breeder reactor SNF); Group 7--Uranium metal (N Reactor SNF); Group 8--Uranium oxide (Three Mile Island-2 core debris); Group 9--Aluminum-based SNF (Foreign Research Reactor SNF); Group 10--Miscellaneous Fuel; and Group 11--Uranium-zirconium hydride (Training Research Isotopes-General Atomics SNF). The analyses contained in this document provide an &apos

  9. Characteristics of Cast Stone cementitious waste form for immobilization of secondary wastes from vitrification process

    Science.gov (United States)

    Chung, Chul-Woo; Um, Wooyong; Valenta, Michelle M.; Sundaram, S. K.; Chun, Jaehun; Parker, Kent E.; Kimura, Marcia L.; Westsik, Joseph H.

    2012-01-01

    The high-temperature in vitrification process of radioactive wastes could cause radioactive technetium ( 99Tc) in secondary liquid wastes to become volatile. Solidified cementitious waste forms at low temperature were developed to immobilize radioactive secondary waste. This research focuses on the characterization of a cementitious waste form called Cast Stone. Properties including compressive strength, surface area, phase composition, and technetium leaching were measured. The results indicate that technetium diffusivity is affected by simulant type. Additionally, ettringite and AFm (Al 2O 3-Fe 2O 3-mono) main crystalline phases were formed during hydration. The Cast Stone waste form passed the qualification requirements for a secondary waste form, which are compressive strength of 3.45 MPa and technetium diffusivity of 10 -9 cm 2/s. Cast Stone was found to be a good candidate for immobilizing secondary waste streams.

  10. Studies on the inspection of drum-filled cement monolithic solids of radioactive liquid wastes for sea dumping, (2)

    International Nuclear Information System (INIS)

    With an ultrasonic wave test apparatus for full-size radioactive waste packages, ultrasonic wave velocities of drum-filled monolithic cement solids were measured and compared against uni-axial compressive strengths of the solids of same composition. The results are as follows; 1) There is a linear relationship between ultrasonic wave velocity and uni-axial compressive strength for the drum-filled solid. Compressive strength 150 kg/cm2 or more required of cement solid for sea dumping correspond to ultrasonic wave velocity 3200 m/sec or more. 2) Comparison between the ultrasonic test and the rebound hammer test indicates that both the methods should be used for estimation of the uni-axial compressive strength. 3) Existence and position of defects in drum-filled cement solids can be detected by the ultrasonic test. (author)

  11. Cr(VI) removal in acidic aqueous solution using iron-bearing industrial solid wastes and their stabilisation with cement.

    Science.gov (United States)

    Singh, I B; Singh, D R

    2002-01-01

    In this study, iron-bearing industrial solid wastes iron filings, ETP sludge of steel and red mud of aluminium industries; were used for Cr(VI) removal at pH 3. A complete removal of Cr(VI) was found for initial 10 mg 1(-1) of 100 ml solutions in the presence of 2.5 g iron filings, 8 g ETP sludge and 10 g red mud for up to one hour of shaking at room temperature. After Cr(VI) removal, inclusion of chromium on the reacted iron filing surface was demonstrated by EDAX analysis. Leachability of chromium and iron from the reacted wastes was determined by using Toxicity Characteristics Leaching Procedure (TCLP). This test showed a very low level of leachability of chromium as Cr(III) and iron from the reacted wastes. To minimise their leachability further, Cr(VI)-reacted solid wastes were stabilised with Portland cement in their 3:1 ratio. Leachability tests of stabilised wastes by TCLP indicated a considerable decrease in leachability of chromium and iron compared with the that of reacted wastes alone. To explore the possibility of utilisation in building materials, bricks of cement-mixed Cr(VI)-reacted wastes were made and their comprehensive strength, durability and leachability under immersion conditions were measured. PMID:11918404

  12. Challenges in Modeling the Degradation of Ceramic Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Devanathan, Ramaswami; Gao, Fei; Sun, Xin

    2011-09-01

    We identify the state of the art, gaps in current understanding, and key research needs in the area of modeling the long-term degradation of ceramic waste forms for nuclear waste disposition. The directed purpose of this report is to define a roadmap for Waste IPSC needs to extend capabilities of waste degradation to ceramic waste forms, which overlaps with the needs of the subconsinuum scale of FMM interests. The key knowledge gaps are in the areas of (i) methodology for developing reliable interatomic potentials to model the complex atomic-level interactions in waste forms; (ii) characterization of water interactions at ceramic surfaces and interfaces; and (iii) extension of atomic-level insights to the long time and distance scales relevant to the problem of actinide and fission product immobilization.

  13. Self-Shrinkage Behaviors of Waste Paper Fiber Reinforced Cement Paste considering Its Self-Curing Effect at Early-Ages

    Directory of Open Access Journals (Sweden)

    Zhengwu Jiang

    2016-01-01

    Full Text Available The aim of this paper was to study how the early-age self-shrinkage behavior of cement paste is affected by the addition of the waste paper fibers under sealed conditions. Although the primary focus was to determine whether the waste paper fibers are suitable to mitigate self-shrinkage as an internal curing agent under different adding ways, evaluating their strength, pore structure, and hydration properties provided further insight into the self-cured behavior of cement paste. Under the wet mixing condition, the waste paper fibers could mitigate the self-shrinkage of cement paste and, at additions of 0.2% by mass of cement, the waste paper fibers were found to show significant self-shrinkage cracking control while providing some internal curing. In addition, the self-curing efficiency results were analyzed based on the strength and the self-shrinkage behaviors of cement paste. Results indicated that, under a low water cement ratio, an optimal dosage and adding ways of the waste paper fibers could enhance the self-curing efficiency of cement paste.

  14. Influence of chemical composition of civil construction waste in the cement paste; Influencia da composicao quimica dos residuos da construcao civil a pasta de cimento

    Energy Technology Data Exchange (ETDEWEB)

    Cunha, G.A.; Andrade, A.C.D.; Souza, J.M.M.; Evangelista, A.C.J.; Almeida, V.C., E-mail: valeria@eq.ufrj.b [Universidade Federal do Rio de Janeiro (EQ/UFRJ), RJ (Brazil). Escola de Quimica

    2009-07-01

    The construction and demolition waste when disposed inappropriately might cause serious public health problems. Its reutilization focusing on the development of new products using simple production techniques, assuring a new product life cycle and not damaging the environment is inserted in sustainable concept. The aim of this work was identifying the characteristics of types of waste generated in a residential reform (glassy ceramic and fill dirt leftovers) verifying separately its influence on cement pastes mechanical behavior. Cement pastes + wastes were prepared in 25% and 50% proportions with an approximately 0,35 water/cement relation and, glue time determination, water absorption, resistance to compression and X-ray fluorescence assays were taken. The results indicate that the chemical composition of the waste causes changes in the behavior of cement pastes, reflecting on their resistance to compression. (author)

  15. Preparation and leaching of radioactive INEL waste forms

    International Nuclear Information System (INIS)

    Appreciable quantities of radioactive waste are in storage at the Idaho National Engineering Laboratory (INEL). Plans are being made to convert this waste into durable solid forms for final disposal in a geological repository. Part of the inventory consists of low- and intermediate-level fission, activation, and decay products and transuranic (TRU) wastes, either stored retrievably or buried at the INEL Radioactive Waste Management area. One of the TRU wastes is a sludge from the Department of Energy Rocky Flats Plant, currently stored retrievably in 55-gallon drums. Immobilizing the TRU sludge is the primary concern of the work reported here

  16. Radioactivity leach tests under high hydrostatic pressure on full size monolithic cement packages of LWR wastes for sea dumping

    International Nuclear Information System (INIS)

    Monolithic cement-solidified products of simulated evaporator concentrates from LWR packaged in 200 l drums have been subjected to leach tests of radionuclides. The leach tests were crrried out under the condition which simulates hydrostatic pressure, temperature and flow velocity on the sea bed of 5,000 m depth. The principal results are obtained as follows: (1) No radioactivity in the leachant could be detected during the test extended to 675 hr on a cement-solidified product with asphalt capping. Water ingress through the asphalt capping layer was not perceived. (2) Solidified packages with intentionally prepared bare surface were tested to find out leaching coefficients(apparent diffusion coefficients) F sub(l). Values of F sub(l) were found to be 1.6 x 10-4 for 137Cs and 1.2 x 10-9 cm2/day for 60Co, in case of the blast furnace slag cement (C-class) solidified product of simulated waste from BWR. As to the Portland cement-solidified product of simulated PWR waste, F sub(l) was 1.5 x 10-3 cm2/day for 137Cs. (3) By using these leaching coefficients, a long-term estimation of the amounts leached out was attempted considering the decay of radionuclides. (author)

  17. Application of PCT to the EBR II ceramic waste form

    International Nuclear Information System (INIS)

    We are evaluating the use of the Product Consistency Test (PCT) developed to monitor the consistency of borosilicate glass waste forms for application to the multiphase ceramic waste form (CWF) that will be used to immobilize waste salts generated during the electrometallurgical conditioning of spent sodium-bonded nuclear fuel from the Experimental Breeder Reactor No. 2 (EBR II). The CWF is a multiphase waste form comprised of about 70% sodalite, 25% borosilicate glass binder, and small amounts of halite and oxide inclusions. It must be qualified for disposal as a non-standard high-level waste (HLW) form. One of the requirements in the DOE Waste Acceptance System Requirements Document (WASRD) for HLW waste forms is that the consistency of the waste forms be monitored.[1] Use of the PCT is being considered for the CWF because of the similarities of the dissolution behaviors of both the sodalite and glass binder phases in the CWF to borosilicate HLW glasses. This paper provides (1) a summary of the approach taken in selecting a consistency test for CWF production and (2) results of tests conducted to measure the precision and sensitivity of the PCT conducted with simulated CWF

  18. The Next Generation Ecological Self Compacting Concrete with Glass Waste Powder as a Cement Component in Concrete and Recycled Concrete Aggregates

    OpenAIRE

    Kara, P

    2013-01-01

    In the present study the performance characteristics (workability, compressive strength, frost resistance, permeability and temperature of hydration) of the ecological self compacting concrete with reduced cement content and with the next generation recycled concrete aggregates which are obtained from crashed concrete specimens with cement substitution at level of 30% with waste glass powder were investigated. Waste glass as powder ground to certain fineness accelerates beneficial chemical re...

  19. Characterization of low and medium-level radioactive waste forms. Final report - 2nd Programme 1980-84

    International Nuclear Information System (INIS)

    The European Communities Second R and D Programme 1980-84 'Management and Disposal of Radioactive Waste (Shared cost action)' included a closely coordinated research activity for the 'Characterization of low and medium-level radioactive waste forms'. This report summarizes the main results obtained during the five years of the programme by laboratories in seven European countries participating in the coordinated RandD efforts. Ten reference waste forms have been selected, based on the most important types of low and medium-level waste arisings and the three commonly used immobilization matrices: cement, bitumen and polymers. The investigated properties were mainly: waste-matrix compatibility, radiation effects, leaching behaviour, leached radionuclides speciation, microbiological resistance and thermal as well as mechanical properties. Extensive experimental results relevant for the qualification of waste products and for application in performance analysis are presented in this final report. The main conclusions are drawn for the confinement properties of these different waste forms. These conclusions have also shown the necessity of selecting several other reference waste forms for the continuation of this RandD action now being launched in the Third EC Programme 1985-89

  20. Talc-silicon glass-ceramic waste forms for immobilization of high- level calcined waste

    International Nuclear Information System (INIS)

    Talc-silicon glass-ceramic waste forms are being evaluated as candidates for immobilization of the high level calcined waste stored onsite at the Idaho Chemical Processing Plant. These glass-ceramic waste forms were prepared by hot isostatically pressing a mixture of simulated nonradioactive high level calcined waste, talc, silicon and aluminum metal additives. The waste forms were characterized for density, chemical durability, and glass and crystalline phase compositions. The results indicate improved density and chemical durability as the silicon content is increased

  1. Improved cement mortars by addition of carbonated fly ash from solid waste incinerators

    Directory of Open Access Journals (Sweden)

    López-Zaldívar, O.

    2015-09-01

    Full Text Available This article presents the results of a research developing high performance cement mortars with the addition of municipal solid waste incineration fly ash (MSWIFA stabilized as insoluble carbonates. The encapsulation of hazardous wastes in mortar matrixes has also been achieved. The ashes present high concentrations of chlorides, Zn and Pb. A stabilization process with NaHCO3 has been developed reducing 99% the content of chlorides. Developed mortars replace 10% per weight of the aggregates by treated MSWIFA. Physical/mechanical properties of these mortars have been studied. Presence of Zn, Pb, Cu and Cd has been also analyzed confirming that leaching of these heavy metal ions is mitigated. Conclusions prove better behavior of CAC and CSA mortars than those of CEM-I and CEM-II cement. Results are remarkable for the CAC mortars, improving reference strengths in more than 25%, which make them a fast-curing product suitable for the repair of structures or industrial pavements.Este artículo presenta los resultados del desarrollo de morteros mejorados con la incorporación de cenizas volantes de residuos sólidos urbanos inertizadas en forma de carbonatos. Además se consigue la encapsulación de un residuo peligroso. Las cenizas presentan una alta concentración de cloruros, Zn y Pb. Se ha desarrollado un proceso de estabilización con NaHCO3 reduciendo en un 99% el contenido de cloruros. Los morteros reemplazan un 10% en peso del árido por cenizas tratadas. Se han analizado sus propiedades físico/mecánicas y la presencia de Zn, Pb, Cu y Cd. Se demuestra un mejor comportamiento de los morteros de CAC y CSA que los de CEM-I y CEM-II y se mitiga el lixiviado de metales pesados. Los resultados son significativos en los morteros CAC al mejorar las resistencias de los de referencia en un 25%. Los morteros desarrollados son de curado rápido adecuados para la reparación de estructuras o soleras industriales.

  2. The necessity for scale up in R and D: Approach for waste immobilization in cement by BNFL

    International Nuclear Information System (INIS)

    The full scale processing of nuclear wastes immobilized in cement utilizes a wide range of chemical and physical parameters. The success of this work however, involves many factors and material properties which are affected by the actual scaling up processes. The paper outlines the approach and experience gained by British Nuclear Fuels plc (BNFL) to recognize and evaluate the major factors involved in order to successfully produce large scale stable products acceptable to the appropriate regulatory bodies and suitable for long term disposal

  3. Rock alteration in alkaline cement waters over 15 years and its relevance to the geological disposal of nuclear waste

    OpenAIRE

    Moyce, Elizabeth B.A.; Rochelle, Christopher; Morris, Katherine; Milodowski, Antoni E.; Chen, Xiaohui; Thornton, Steve; Small, Joe S.; Shaw, Samuel

    2014-01-01

    The interaction of groundwater with cement in a geological disposal facility (GDF) for intermediate level radioactive waste will produce a high pH leachate plume. Such a plume may alter the physical and chemical properties of the GDF host rock. However, the geochemical and mineralogical processes which may occur in such systems over timescales relevant for geological disposal remain unclear. This study has extended the timescale for laboratory experiments and shown that, after