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Sample records for cemented waste forms

  1. LEACHING BOUNDARY IN CEMENT-BASED WASTE FORMS

    Cement-based fixation systems are among the most commonly employed stabilization/solidification techniques. These cement haste mixtures, however, are vulnerable to ardic leaching solutions. Leaching of cement-based waste forms in acetic acid solutions with different acidic streng...

  2. Comparison of the leachability of three TRU cement waste forms

    Ross, W.A.; Westsik, J.H. Jr.; Roberts, F.P.; Harvey, C.O.

    1982-11-01

    Cement waste forms prepared by three processes, casting, cold pressing, and FUETAP (Formed Under Elevated Temperatures and Pressure) have been compared for their leachability by using the MCC-1 leach test. The results indicate that releases of plutonium are not controlled by the waste form matrix and that there is no significant overall advantage to any of the three cement processes from a leachability viewpoint

  3. Research needs in cement-based waste forms

    McDaniel, E.W.; Spence, R.D.; Tallent, O.K.

    1990-01-01

    Cement-based waste forms are one of the most widely used waste disposal options, yet definitive knowledge of the fate of the waste species inside the waste form is lacking. A fundamental understanding of the chemistry and microstructure of the waste forms would lead to a better understanding of the mass transfer of the waste species, more confidence in predicting and extrapolating waste form performance, and design of better waste forms. Better and cheaper leach tests would lead to quicker and more cost effective screening of waste form alternatives. In addition, assessment of durability may be important to predicting waste form performance in the field. It should be noted that the research needs discussed in this report are from the perspective of investigators working in applied waste management areas, while the proposed investigations are fundamental or basic. Details as to experimental methods and tools to be used in achieving the objectives of the proposed are research beyond the scope of this paper and are better filled in by others. In broad terms, the research topics discussed are correlation of cement-based waste form physical properties to performance, waste-form fundamental chemistry and microstructure, and product performance testing

  4. Compressive strength test for cemented waste forms: validation process

    Haucz, Maria Judite A.; Candido, Francisco Donizete; Seles, Sandro Rogerio

    2007-01-01

    In the Cementation Laboratory (LABCIM), of the Development Centre of the Nuclear Technology (CNEN/CDTN-MG), hazardous/radioactive wastes are incorporated in cement, to transform them into monolithic products, preventing or minimizing the contaminant release to the environment. The compressive strength test is important to evaluate the cemented product quality, in which it is determined the compression load necessary to rupture the cemented waste form. In LABCIM a specific procedure was developed to determine the compressive strength of cement waste forms based on the Brazilian Standard NBR 7215. The accreditation of this procedure is essential to assure reproductive and accurate results in the evaluation of these products. To achieve this goal the Laboratory personal implemented technical and administrative improvements in accordance with the NBR ISO/IEC 17025 standard 'General requirements for the competence of testing and calibration laboratories'. As the developed procedure was not a standard one the norm ISO/IEC 17025 requests its validation. There are some methodologies to do that. In this paper it is described the current status of the accreditation project, especially the validation process of the referred procedure and its results. (author)

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

    Lee, Yoon-ji; Lee, Ki-Won; Min, Byung-Youn; Hwang, Doo-Seong; Moon, Jei-Kwon

    2015-01-01

    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 m 3 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. The characterization of cement waste form for final disposal of decommissioned concrete waste

    Lee, K.W.; Lee, Y.J.; Hwang, D.S.; Moon, J.K.

    2015-01-01

    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. In addition, 83 drums of 200 l, and 41 containers of 4 m 3 of concrete waste were generated. Conditioning of concrete waste is needed for final disposal. Concrete waste is conditioned as follows: mortar using coarse and fine aggregates is filled into 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 compressive strength test, leaching test, and thermal cycling test of cement waste forms conclude that the concrete waste, water, and cement have been suggested to have 75:15:10 as the optimized mixing ratio. In addition, the compressive strength of cement waste form was satisfied, including fine powder up to a maximum 40 wt% in concrete debris waste of about 75%. (authors)

  7. Characterization of low and intermediate level cemented waste forms

    Koester, R.; Vejmelka, P.; Brunner, H.; Ganser, B.

    1985-01-01

    The main objective of the characterization work was to establish source term formulations for the cemented waste forms as input for safety analysis. For the operation phase of a repository radionuclide mobilization from the waste packages via the gas phase, caused by mechanical or thermal impact has to be considered. For this reason, besides laboratory tests, experiments with inactive full scale samples were performed to determine quantitatively the activity release from the waste packages under defined thermal and mechanical stresses. In order to evaluate source terms for the mobilization of relevant radionuclides via the liquid phase as a function of time due to leaching and corrosion, detailed experimental work with simulated inactive and dopted laboratory samples and with inactive full scale samples was performed. The experimental work was accompanied by theoretical investigations to establish an improved basis for long term predictions. (orig./PW)

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

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

    1997-01-01

    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

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

    Westsik, J.H. Jr.

    1984-08-01

    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 10 8 Rad and exposure to 31 thermal cycles ranging from +60 0 ) to -30 0 C 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

  10. A novel waste form for disposal of spent-nuclear-fuel reprocessing waste: A vitrifiable cement

    Gougar, M.L.D.; Scheetz, B.E.; Siemer, D.D.

    1999-01-01

    A cement capable of being hot isostatically pressed into a glass ceramic has been proposed as the waste form for spent-nuclear-fuel reprocessing wastes at the Idaho National Engineering and Environmental Laboratory (INEEL). This intermediate cement, with a composition based on that of common glasses, has been designed and tested. The cement formulations included mixed INEEL wastes, blast furnace slag, reactive silica, and INEEL soil or vermiculite, which were activated with potassium or sodium hydroxide. Following autoclave processing, the cements were characterized. X-ray diffraction analysis revealed three notable crystalline phases: quartz, calcite, and fluorite. Results of compressive strength testing ranged from 1452 and 4163 psi, exceeding the US Nuclear Regulatory Commission (NRC)-suggested standard of >500 psi. From American National Standards Institute/American Nuclear Society 16.1-1986 leach testing, effective diffusivities for Cs were determined to be on the order of 10 -11 to 10 -10 cm 2 /s and for Sr were 10 -12 cm 2 /s, which are four orders of magnitude less than diffusivities in some other radwaste materials. Average leach indices (LI) were 9.6 and 11.9 for Cs and Sr, respectively, meeting the NRC Standard of LI > 6. The 28-day Materials Characterization Center-1 leach testing resulted in normalized elemental mass losses between 0.63 and 28 g/(m 2 ·day) for Cs and between 0.34 and 0.70 g/(m 2 ·day) industry-accepted standard while Cs losses indicate a process sensitive parameter

  11. Leaching experiment of cement solidified waste form under unsaturated condition

    Wang Zhiming; Yao Laigen; Li Shushen; Zhao Yingjie; Cai Yun; Li Dan; Han Xinsheng; An Yongfeng

    2003-01-01

    A device for unsaturated leaching experiments was designed and built up. 8 different sizes, ranging from 40.2 cm 3 to 16945.5 cm 3 , of solidified waste form were tested in the experiment. 5 different water contents, from 0.15 to 0.40, were used for the experiment. The results show that the cumulative leaching fraction increases with water content when the sizes of the forms are equal to and less than 4586.7 cm 3 , for example, the ratios of the cumulative leaching fractions are between 1.24-1.41 under water content of 0.35 and 0.15 on 360 day of Teaching. It can also be seen that the cumulative leaching fraction under higher water content is close to that under saturated condition. The cumulative leaching fraction decreases with size of the form. Maximum leached depth of the solidified waste forms is about 2.25 cm after one year Teaching. Moreover, it has no clear effect on cumulative leaching fraction that sampling or non-sampling during the experiment

  12. The effect of cure conditions on the stability of cement waste forms after immersion in water

    Siskind, B.; Adams, J.W.; Clinton, J.H.; Piciulo, P.L.; McDaniel, K.

    1988-01-01

    We investigated the effects of curing conditions on the stability of cement-solidified ion-exchange resins after immersion in water. The test specimens consisted of partially depleted mixed-bed bead resins solidified in one of three vendor-supplied Portland I cement formulations, in a reference cement formulation, or in a gypsum-based binder formulation. We cured samples prepared using each formulation in sealed containers for periods of 7, 14, or 28 days as well as in air or with an accelerated heat cure prior to 90-day immersion in water. Two cement formulations exhibited apparent Portland-cement-like behavior, i.e., compressive strength increased or stabilized with increasing cure time. Two cement formulations exhibited behavior apparently unlike that of Portland cement, i.e., compressive strength decreased with increasing cure time. Such non-Portland-cement-like behavior is correlated with higher waste loadings. The gypsum-based formulation exhibited approximately constant compressive strength with cure time. Accelerated heat cures may not give compressive strengths representative of real-time cures. Some physical deterioration (cracking, spalling) of the waste form occurs during immersion

  13. The effect of cure conditions on the stability of cement waste forms after immersion in water

    Siskind, B.; Adams, J.W.; Clinton, J.H.; Piciulo, P.L.

    1988-01-01

    The authors investigated the effects of curing conditions on the stability of cement-solidified ion-exchange resins after immersion in water. The test specimens consisted of partially depleted mixed-bed bead resins solidified in one of three vendor-supplied Portland I cement formulations, in a reference cement formulation, or in a gypsum-based binder formulation. They cured samples prepared using each formulation in sealed containers for periods of 7, 14, or 28 days as well as in air or with an accelerated heat cure prior to 90-day immersion in water. Two cement formulations exhibited apparent Portland-cement-like behavior, i.e., compressive strength increased or stabilized with increasing cure time. Two cement formulations exhibited behavior apparently unlike that of Portland cement, i.e. compressive strength decreased with increasing cure time. Such non-Portland-cement-like behavior is correlated with higher waste loadings. The gypsum-based formulation exhibited approximately constant compressive strength with cure time. Accelerated heat cures may not give compressive strengths representative of real-time cures. Some physical deterioration (cracking, spalling) of the waste form occurs during immersion

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

    Mattus, C.H.; Mattus, A.J.

    1994-03-01

    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)

  15. Microbially influenced degradation of cement-solidified low-level radioactive waste forms

    Rogers, R.D.; Hamilton, M.A.; Veeh, R.H.; McConnell, J.W. Jr.

    1996-01-01

    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. This paper contains information on three groups of microoganisms that are associated with the degradation of cement materials: sulfur-oxidizing bacteria (Thiobacillus), nitrifying bacteria (Nitrosomonas and Nitrobacter), and heterotrophic bacteria, which produce organic acids. Preliminary work using laboratory- and vendor-manufactured, simulated waste forms exposed to thiobacilli has shown that microbiologically influenced degradation 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 was leached from the treated waste forms. Also, the surface pH of the treated specimens was decreased to below 2. These conditions apparently contributed to the physical deterioration of simulated waste forms after 30 to 60 days of exposure

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

    Pepper, D.W.

    1986-01-01

    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

  17. Corrosion studies on mild steel in contact with cemented waste forms

    Platts, N.

    1987-03-01

    The internal corrosion of mild steel drums has been assessed in support of the proposed cementation of low level solid and sludge wastes arising from the Winfrith SGHWR. General corrosion rates are expected to be very low but while oxygen is present pitting, at least in principle, is possible. Passive currents for mild steel in contact with the different cement formulations have therefore been determined, thereby enabling the timescale for residual oxygen within the drums to be assessed. These measurements indicate that for BFS/OPC cement formulations the oxygen will be exhausted within 37 days, though potential monitoring suggests that sulphides present in the cement may reduce this figure. For PFA/OPC cement formulations the passive currents suggest that it would take ca. one year for the oxygen present in a drum to be exhausted. Therefore though pitting should not be a problem with the BFS/OPC cement formulations, further investigation of the extent of pitting that may occur in the PFA/OPC waste form is desirable. (author)

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

    Ewart, F.T.; Howse, R.M.; Sharpe, B.M.; Smith, A.J.; Thomason, H.P.; Williams, S.J.; Young, M.

    1986-01-01

    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

  19. Chemical stability of seven years aged cement-PET composite waste form containing radioactive borate waste simulates

    Saleh, H.M., E-mail: hosamsaleh70@yahoo.com [Radioisotope Department, Atomic Energy Authority, Dokki (Egypt); Tawfik, M.E. [Department of Polymers and Pigments, National Research Center, Dokki (Egypt); Bayoumi, T.A. [Radioisotope Department, Atomic Energy Authority, Dokki (Egypt)

    2011-04-15

    Different samples of radioactive borate waste simulate [originating from pressurized water reactors (PWR)] have been prepared and solidified after mixing with cement-water extended polyester composite (CPC). The polymer-cement composite samples were prepared from recycled poly (ethylene terephthalate) (PET) waste and cement paste (water/cement ratio of 40%). The prepared samples were left to set at room temperature (25 deg. C {+-} 5) under humid conditions. After 28 days curing time the obtained specimens were kept in their molds to age for 7 years under ambient conditions. Cement-polymer composite waste form specimens (CPCW) have been subjected to leach tests for both {sup 137}Cs and {sup 60}Co radionuclides according to the method proposed by the International Atomic Energy Agency (IAEA). Leaching tests were justified under various factors that may exist within the disposal site (e.g. type of leachant, surrounding temperature, leachant behavior, the leachant volume to CPCW surface area...). The obtained data after 260 days of leaching revealed that after 7 years of aging the candidate cement-polymer composite (CPC) containing radioactive borate waste samples are characterized by adequate chemical stability required for the long-term disposal process.

  20. DEVELOPMENT and TESTING OF A CEMENT-BASED SOLID WASTE FORM USING SYNTHETIC UP-1 GROUNDWATER

    COOKE, G.A.; LOCKREM, L.L.

    2006-01-01

    The Effluent Treatment Facility (ETF) in the 200 East Area of the Hanford Site is investigating the conversion of several liquid waste streams from evaporator operations into solid cement-based waste forms. The cement/waste mixture will be poured into plastic-lined mold boxes. After solidification the bags will be removed from the molds and sealed for land disposal at the Hanford Site. The RJ Lee Group, Inc. Center for Laboratory Sciences (CLS) at Columbia Basin College (CBC) was requested to develop and test a cementitious solids (CS) formulation to solidify evaporated groundwater brine, identified as UP-1, from Basin 43. Laboratory testing of cement/simulant mixtures is required to demonstrate the viability of cement formulations that reduce the overall cost, minimize bleed water and expansion, and provide suitable strength and cure temperature. Technical support provided mixing, testing, and reporting of values for a defined composite solid waste form. In this task, formulations utilizing Basin 43 simulant at varying wt% solids were explored. The initial mixing consisted of making small (∼ 300 g) batches and casting into 500-mL Nalgene(reg s ign) jars. The mixes were cured under adiabatic conditions and checked for bleed water and consistency at recorded time intervals over a 1-week period. After the results from the preliminary mixing, four formulations were selected for further study. The testing documentation included workability, bleed water analysis (volume and pH) after 24 hours, expansivity/shrinkage, compressive strength, and selected Toxicity Characteristic Leaching Procedure (TCLP) leach analytes of the resulting solid waste form

  1. Tritium transport in subsurface soil and its retentivity in cement waste forms

    Varghese, C.; Khan, Z.A.

    1990-01-01

    Tritiated aqueous wastes generated from power reactors and reprocessing plants need fixation and containment in solid media. Incorporating tritium in inorganic hydrates like cement retards its movements substantially. Naturally occurring clay mineral like vermiculite which has high water holding capacity was used as a filler material. To improve the retention of tritium in the cement blocks, certain coating materials were used after curing the blocks for 28 days. Long time leach tests for these barrier incorporated wastes forms have shown tritium releases of 17% loaded tritium over a period of 97 days. But the selected coating reduced it to 0.9%. Investigations on the movements of tritium at Tarapur and Narora revealed that in the unsaturated zone, tritium percolation was low, whereas in the saturated zone, it was found to travel with higher velocities, limiting to ground water movement. (author). 5 refs., 3 tabs

  2. Cement waste form qualification report: WVDP [West Valley Demonstration Project] PUREX decontaminated supernatant

    McVay, C.W.; Stimmel, J.R.; Marchetti, S.

    1988-08-01

    This report provides a summary of work performed to develop a cement-based, low-level waste formulation suitable for the solidification of decontaminated high-level waste liquid produced as a by-product of PUREX spent fuel reprocessing. The resultant waste form is suitable for interim storage and is intended for ultimate disposal as low-level Class C waste; it also meets the stability requirements of the NRC Branch Technical Position on Waste Form Qualification, May 1983 and the requirements of 10 CFR 61. A recipe was developed utilizing only Portland Type I cement based on an inorganic salts simulant of the PUREX supernatant. The qualified recipe was tested full scale in the production facility and was observed to produce a product with entrained air, low density, and lower-than-expected compressive strength. Further laboratory scale testing with actual decontaminated supernatant revealed that set retarders were present in the supernatant, precluding setting of the product and allowing the production of ''bleed water.'' Calcium nitrate and sodium silicate were added to overcome the set retarding effect and produced a final product with improved performance when compared to the original formulation. This report describes the qualification process and qualification test results for the final product formulation. 7 refs., 38 figs., 21 tabs

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

    Langton, C.A.

    1984-01-01

    Defense waste processing at the Savannah River Plant will include decontamination and disposal of approximately 400 million liters of waste containing NaNO 3 , NaOH, Na 2 SO 4 , and NaNO 2 . 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 137 Cs and 90 Sr 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

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

    Saleh, H.M.

    2010-01-01

    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

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

    Ewart, F.T.; Howse, R.M.; Sharpe, B.M.; Smith, A.J.; Thomason, H.P.; Williams, S.J.; Young, M.

    1986-01-01

    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)

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

    Rogers, R.D.; Hamilton, M.A.; Veeh, R.H.; McConnell, J.W.

    1994-01-01

    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

  7. Leachability of radionuclides from cement solidified waste forms produced at operating nuclear power plants

    Croney, S.T.

    1985-03-01

    This study determined the leachability indexes of radionuclides contained in solidified liquid wastes from operating nuclear power plants. Different sizes of samples of cement-solidified liquid wastes were collected from two nuclear power plants - a pressurized water reactor and a boiling water reactor - to correlate radionuclide leaching from small- and full-sized (55-gallon) waste forms. Diffusion-based model analysis (ANS 16.1) of measured radionuclide leach data from both small- and full-sized samples was performed and indicate that leach data from small samples can be used to determine leachability indexes for full-sizes waste forms. The leachability indexes for cesium, strontium, and cobalt isotopes were determined for waste samples from both plants according to the models used for ANS 16.1. The leachability indexes for the pressurized water reactor samples were 6.4 for cesium, 7.1 for strontium, and 10.4 for cobalt. Leachability indexes for the boiling water reactor samples were 6.5 for cesium, 8.6 for strontium, and 11.1 for cobalt

  8. Equilibrium leach tests with cobalt in the system cemented waste form/container material/aqueous solution

    Vejmelka, P.; Koester, R.; Lee, M. J.; Han, K. W.

    1991-01-01

    The equilibrium concentrations of Co in the system of cemented waste form/aqueous solutions were determined including the effect of the container material and its corrosion products under the respective conditions. The chemical conditions in the near field of the waste form were characterized by measurement of the pH and E h value. As disposal relevant solutions, saturated sodium chloride, Q-brine (main constituent MgCl 2 ) and a granitic type groundwater were used. For comparison, also experiments using deionized water were performed. In all systems investigated the cemented waste form itself has a strong influence on the chemical conditions in the near field. The pH and E h values are affected in all cases by the addition of the cemented waste form. There is no or only a slight difference between the E h values if iron powder or iron hydroxide is added to the cemented waste form/solution systems, but the E h is markedly decreased when iron powder is added to the solution free of cement. The Co concentration is decreased in all solutions by the addition of the cemented waste form, the largest effect is observed in Q-brine and this can be attributed either to the sorption of the Co-ions on the corrosion products of the cement or to the coprecipitation of Co-hydroxide and Mg-hydroxide. In the other solutions the Co concentration is decreased by precipitation of Co-hydroxide due to the high pH value of 12.5, and the concentrations are comparable for the different solutions

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

    Bouniol, P.

    2014-01-01

    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 O 2 , 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 H 2 aq → H 2 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 H 2 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

  10. Test of 134Cs, 85,89Sr leaching rate in a resemble vitrifiable cement waste form

    Lin Meiqiong; Wei Feng; Yin Qi; Fan Xianhua; Xu Shengli; Li Yongde

    2003-01-01

    A novel material--resemble vitrifiable cement for conditioning low and mediate level radioactive waste has been developed. Waste form has been characterized for their physical and chemical performance, phase composition. The cement formulation has been patented. In this experiment the cement is mixed with simulated wastes spiked with 134 Cs and 85,89 Sr by 5 min at least. The Ratio of the waste to the cement is 0.45-0.55. The mixture is packed into cylindrical molds which has the same dimension of diameter and height . The grouts are cured for a period of 28 d in a room temperature curing chamber at an atmospheric pressure. The cured waste form is then completely immersed into deionized water. According to standard GB7023-86, leaching rate of 134 Cs and 85,89 Sr are measured. The result shows that the leaching rate of the species 134 Cs and 85,89 Sr is to be on the order 10 -4 and 10 -5 on the 42 d immersion, respectively and is better than that of commercial cement

  11. Waste-form development for conversion to portland cement at Los Alamos National Laboratory (LANL) Technical Area 55 (TA-55)

    Veazey, G.W.; Schake, A.R.; Shalek, P.D.; Romero, D.A.; Smith, C.A.

    1996-10-01

    The process used at TA-55 to cement transuranic (TRU) waste has experienced several problems with the gypsum-based cement currently being used. Specifically, the waste form could not reliably pass the Waste Isolation Pilot Plant (WIPP) prohibition for free liquid and the Environmental Protection Agency (EPA)-Toxicity Characteristic Leaching Procedure (TCLP) standard for chromium. This report describes the project to develop a portland cement-based waste form that ensures compliance to these standards, as well as other performance standards consisting of homogeneous mixing, moderate hydration temperature, timely initial set, and structural durability. Testing was conducted using the two most common waste streams requiring cementation as of February 1994, lean residue (LR)- and oxalate filtrate (OX)-based evaporator bottoms (EV). A formulation with a pH of 10.3 to 12.1 and a minimum cement-to-liquid (C/L) ratio of 0.80 kg/l for OX-based EV and 0.94 kg/L for LR-based EV was found to pass the performance standards chosen for this project. The implementation of the portland process should result in a yearly cost savings for raw materials of approximately $27,000 over the gypsum process

  12. The Long-Term Study of Some mechanical and Chemical Properties of Polymer Modified Cement Waste Forms

    Shatta, H.A.; Eskander, S.B.

    2002-01-01

    The present work is a new experimental trend to study the long-term durability of polymer/cement/borate waste forms in storage and disposal sites during the long period of time. The data obtained indicate that all the samples of polymer modified cement waste forms immersed in both distilled and ground water (for more than 2700 days) show a slight increase in the compressive strength values. The chemical stability of the immobilized borate waste forms in distilled and ground water was studied. The cumulative leach fraction in percentage, of Cs-137 isotope from the final solidified waste forms in case of using ground water was lower than that in distilled water for all samples used and the presence of polymer lowered the percent release of radioactive Cs-137 isotope in all samples studied . The leach coefficients (Lc) and the leach indices (Li) of radioactive cesium-137 isotope for cement borate and polymer modified cement waste forms were also calculated. It was found that the leach indices values greater than 6

  13. Evaluation on the long-term durability and leachability of cemented waste form

    Kim, Ki Hong; Lee, Jae Won; Ryue, Young Gerl [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-08-01

    The long-term durability and leachability on the cemented waste form containing boric acids produced in domestic nuclear nuclear power plants were evaluated. Compressive strength of waste form after durability test such as thermal stability and water immersion cycle was higher than before test and consistently increased with increasing of test time in range of 83 of 286 kgf/cm{sup 2}. Long-term leachability was evaluated by standard test methods, leach affecting factors, and prediction of long-term leachability with result data of short-term leach test. In all leach tests, the release of Cs-137 was controlled by diffusion, whereas release of Co-60 was not controlled by diffusion. Leach rate of Cs-137 was relatively constant at standard leach test methods such as NAS 16.1, IAEA, ISO-6961, and MCC-1, but that of Co-60 increased with leachant-renewal frequencies. The leach rate of both Cs-137 and Co-60 increased as test temperature raised. The release of Co-137 decreased in simulated seawater as leachant, but increased with increasing leachant volume. The prediciton of long-term release of Cs-137 from large-scale waste form using the results from short-term leach test of small-scale waste form were within {+-} 5% of actual release. The leachability indexes of Cs-137 were between 6.5 and 7.5 and those of Co-60 were ranged from 11.6 to 13.3, increasing as cumulative leaching time increased. (author). 22 refs., 14 figs., 15 tabs.

  14. The effects of temperature on the leaching behavior of cement waste forms: The cement/sodium sulfate system

    Fuhrmann, M.; Pietrzak, R.; Heiser, J.; Franz, E.M.; Colombo, P.

    1989-10-01

    The leaching mechanisms of simulated low-level radioactive waste forms are being determined as support for development of an accelerated leach test. Two approaches are being used: (1) comparisons of leaching data with results of a model that describes diffusion from a finite cylinder, and (2) observation of the leaching process at temperatures between 20 degree C and 65 degree C. To provide results that can be used for modeling, leaching at elevated temperatures must change neither the leaching mechanism nor the structural controls of leaching such as the porosity. Releases of 137 Cs, 85 Sr, calcium, sodium and potassium from portland cement containing sodium sulfate, as a simulated evaporator sludge, have been determined under a variety of experimental conditions. Data from the leach tests were compared to model results for diffusion from the finite cylinder. While most leaching appears to be diffusion controlled, notable exceptions occur. For all samples activation energies ranging between 6 and 11 Kcal/mole have been calculated from the relationship of the effective diffusion coefficient to increasing temperature, close to the expected value of 5 Kcal/mole for diffusion. 11 refs., 4 figs., 2 tabs

  15. Cement waste form development for ion-exchange resins and fine particles ILW of AREVA La Hague Reprocessing Plant

    Chartier, D.; Sanchez-Canet, J.; Avril, D.; Roussel, C.; Pineau, J.N.

    2015-01-01

    Wastes have been temporarily stored in dedicated silos in La Hague reprocessing plant. These wastes are to be retrieved in the near future and to be conditioned for final disposal. Some of these wastes are supposed to be encapsulated in cement matrix and, depending on the chemical composition of the waste streams, several projects are presently ongoing. The present article aims to focus on one amongst these cement encapsulation relevant projects, namely the conditioning of a mix of spent ion-exchange resins (from filtration of pool) and fine particles (insoluble fission products from spent fuel dissolution and Zircaloy and stainless steel fines from cladding shearing). The project, aims to retrieve these wastes from a silo, separate the resins and fine particles from the other waste (hulls and end pieces), in order to encapsulate the intermediate-level fines and resins in a cement matrix. The waste forms will be produced in AREVA's La Hague reprocessing plant, prior to being sent as intermediate-level waste to a long-term repository. The cement formulation developments were initially carried out at a small scale at C.E.A. Marcoule on surrogate wastes. One of the main issues that were considered was the chemical compatibility between waste and cement matrix. Indeed, swelling phenomena are sometimes reported when ion exchange resins are embedded in cement matrixes such as Portland cement. This kind of destructive phenomenon has been prevented by the use of cement containing a high amount of ground granulated blast furnace slag. The impact of the variability of ionic charge of the resins on the waste form's properties has also been addressed in order to comfort the results obtained on the reference ionic charge of resins NaNO 3 . Once the results obtained were satisfactory, intermediate scale and full scale tests were performed by AREVA. These tests have focused on adjusting the mixing process and controlling the thermal properties of the mix during setting

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

    Burbank, D.A. Jr.; Weingardt, K.M.

    1993-10-01

    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

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

    Sukhanov, L.P.

    2013-01-01

    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

  18. Thermal properties of fly ash substituted slag cement waste forms for disposal of Savannah River Plant salt waste

    Roy, D.M.; Kaushal, S.; Licastro, P.H.; Langton, C.A.

    1985-01-01

    Waste processing at the Savannah River Plant will involve reconstitution of the salts (NaNO 3 , NaNO 2 , NaOH, etc.) into a concentrated solution (32 weight percent salts) followed by solidification in a cement-based waste form for burial. The stability and mechanical durability of such a 'saltstone monolith' will depend largely on the temperature reached due to heat of hydration and the thermal properties of the waste form. Fly ash has been used as an inexpensive constituent and to moderate the hydration and setting processes so as to avoid reaching prohibitively high temperatures which could cause thermal stresses. Both high-calcium and low-calcium fly ashes have been studied for this purpose. Other constituents of these mixes include granulated blast furnace slag and finely crushed limestone. Adiabatic temperature increase and thermal conductivity of these mixes have been studied and related x-ray diffraction and scanning electron microscopy studies carried out to understand the hydration process

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

    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.

  20. Conditioning of radioactive waste solutions by cementation

    Vejmelka, P.; Rudolph, G.; Kluger, W.; Koester, R.

    1992-02-01

    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) [de

  1. Radioactive waste cementation

    Soriano B, A.

    1996-01-01

    This research was carried out to develop the most adequate technique to immobilize low and medium-activity radioactive waste. different brands of national cement were used, portland and pozzolanic cement. Prismatic and cylindrical test tubes were prepared with different water/cement (W/C) relationship. Additives such a as clay and bentonite were added in some other cases. Later, the properties of these test tubes were evaluated. Properties such as: mechanical resistance, immersion resistance, lixiviation and porosity resistance. Cement with the highest mechanical resistance values, 62,29 MPa was pozzolanic cement for a W/C relationship of 0,35. It must be mentioned that the other types of cements reached a mechanical resistance over 10 MPa, a value indicated by the international standards for transportation and storage of low and medium-activity radioactive waste at a superficial level. However, in the case of immersion resistance, Sol cement (portland type I) with a W/C relationship of 0,35 reached a compression resistance over 61,92 MPa; as in the previous cases, the other cements reached a mechanical resistance > 10 MPa. Regarding porosity, working with W/C relationships = 0,35 0,40 and 0,45, without additives and with additives, the percentage of porosity found for all cements is lower than 40% percentage indicated by international standards. With regard to the lixiviation test, pozzolanic cement best retained Cesium-137 and Cobalt-60, and increased its advantages when bentonite was added, obtaining a lixiviation rate of 2,02 x E-6 cm/day. Sol cement also improved its properties when bentonite was added and obtained a lixiviation rate of 2,84 x E-6 cm/day for Cesium-137. However, Cobalt-60 is almost completely retained with the 3 types of cement with or without additives, reaching the limits indicated by the international standards for the lixiviation rate of beta-gamma emitter < 5,00E-4 cm/day. Characterizing the final product involves the knowledge of its

  2. Cements in Radioactive Waste Disposal

    Glasser, F.P.

    2013-01-01

    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

  3. Cements in Radioactive Waste Disposal

    Glasser, F. P. [University of Aberdeen, Scotland (United Kingdom)

    2013-09-15

    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

  4. Technical justifications for the tests and criteria in the waste form technical position appendix on cement stabilization

    Siskind, B.; Cowgill, M.G.

    1992-01-01

    As part of its technical assistance to the Nuclear Regulatory Commission (NRC), Brookhaven National Laboratory (BNL) developed a background document for the cement stabilization appendix, Appendix A, to Rev. 1 of the Technical Position on Waste Form (TP). Here we present an overview of this background document, which provides technical justification for the stability tests to be performed on cement-stabilized waste forms and for the criteria posed in each test, especially for those tests which have been changed from their counterparts in the May 1983 Rev. 0 TP. We address guidelines for procedures from Appendix A which are considered in less detail or not at all in the Rev. 0 of the TP, namely, qualification specimen preparation (mixing, curing, storage), statistical sampling and analysis, process control program specimen preparation and examination, and surveillance specimens. For each waste form qualification test, criterion or procedural guidelines, we consider the reason for its inclusion in Appendix A, the changes from Rev. 0 of the TP (if applicable), and a discussion of the justification or rationale for these changes

  5. Technical justifications for the tests and criteria in the waste form Technical position appendix on cement stabilization

    Siskind, B.; Cowgill, M.G.

    1992-01-01

    As part of its technical assistance to the Nuclear Regulatory Commission (NRC), Brookhaven National Laboratory (BNL) developed a background document for the cement stabilization appendix. Appendix A, to Rev. 1 of the Technical Position on Waste Form (TP). Here we present an overview of this background document, which provides technical justification for the stability tests to be performed on cement-stabilized waste forms and for the criteria posed in each test) especially for those tests which have been changed from their counterparts in the May 1983 Rev. 0 TP. We address guidelines for procedures from Appendix A which are considered in less detail or not at all in the Rev. 0 of the TP, namely, qualification specimen preparation (mixing, curing, storage), statistical sampling and analysis, process control program specimen preparation and examination, and surveillance specimens. For each waste form qualification test, criterion or procedural guideline, we consider the reason for its inclusion in Appendix A, the changes from Rev. 0 of the TP (if applicable), and a discussion of the justification or rationale for these changes. (author)

  6. Long-time leaching and corrosion tests on full-scale cemented waste forms in the Asse salt mine. Sampling and analyses 2003

    Kienzler, B.; Schlieker, M.; Bauer, A.; Metz, V.; Meyer, H.

    2004-10-01

    The paper presents the follow-up of experimental findings from full-scale leach tests performed on simulated cemented waste forms for more than 20 years in salt brines and water. Measurements cover pH, density, and the composition of leachates as well as the release of radionuclides such as Cs, U and Np. Indicators for waste form corrosion and radionuclide release is Cs and NO 3 . Corrosion of cemented waste forms depends on the pore volume of the hardened cement which is correlated to the water/cement ratio. The release of radionuclides is evaluated and compared to small-scale laboratory tests. Excellent interpretation of observed concentrations is obtained for uranium and neptunium by comparison with model calculations. (orig.)

  7. Leach studies of chelating agents and influence on radionuclide leaching from simulated LLW/ILW cement waste forms

    Vejmelka, P.; Koester, R.; Ferrara, D.; Wacks, M.E.

    1990-01-01

    Leach studies were performed on cemented waste forms containing sodium nitrate, trace amounts of cesium-137, and cobalt-60, and a chelating agent (ethylene diamine tetraacetic acid (EDTA), nitrilotriacetic acid (NTA), or citric acid). Leaching of the chelates was measured in water and the effect of the chelates on the release of the Cs-137 and Co-60 was studied. The time dependence of the release rate of the chelates is comparable but the chelate concentration in solution and the released fractions are different. EDTA shows the highest release rate followed by NTA and citrate. The release of the non complex forming cesium is not affected by the presence of the chelates. Independent from the strong complex formation of cobalt with EDTA, NTA, and citrate in the alkaline region the cobalt release is also not affected by the presence of the chelates. The high calcium content of the system decreases the stability of the Co complexes in the high pH region (12-13). Experiments were performed to determine the equilibrium concentration of the chelates between liquid and solid phases. The liquid phases were deionized water, saturated sodium chloride, 24 percent magnesium chloride and Q-brine. The equilibrium studies are based on the assumption that in time a stable final condition is to be established in the near field of the waste form in which each compound is at chemical equilibrium between the dissolved and the various solid phases. The total release may be assessed from the concentration in solution and flow rate out of the near field. The fraction of EDTA released from the cement ranged from 0.2 in the Q-brine to 0.5 in the saturated sodium chloride. The concentration of EDSA in solution was dependent on the original amount in the cement sample, but the released fraction was independent of the initial loading. Indicating, EDTA concentration is not affected by solubility limits. 11 refs., 3 figs., 2 tabs

  8. 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

    Jardine, L J

    2003-01-01

    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 and D activities required to develop acceptable waste treatment processes for use in Russia as part of the ongoing Joint US RF Plutonium Disposition Activities

  9. The mechanical and physical behaviour on the cemented waste forms after one year disposal of real conditions

    Bucataru-Nicu, M.; Mihai, F.; Rotarescu, Gh.; Turcanu, C.N.; Lungu, L.

    1999-01-01

    To prove the durability of cemented waste forms, an experimental program started in 1997 to establish the alteration in time of mechanical and physic-chemical properties of different simulated and real matrices in laboratory and Baita-Bihor disposal conditions. The selected matrices for characterization are normal cement and concrete compositions with and without mineral additives (bentonite and volcanic tuff) as reference samples, the conditioned matrices prepared with radioactive sludge non-active components (iron hydroxides and phosphate, calcium phosphate and cooper ferrocyanide) and non-active matrices prepared with organic acids and salts (oxalic, citric, tartaric, EDTA). All samples will be positioned in five locations of Baita-Bihor repository and kept 1, 3, 5 and 10 years before mechanical and structural characterization. These locations have different humidity conditions decreasing from location 1 to 5. These results will be compared with results obtained for the same type of samples kept in laboratory conditions. The research program requires detailed identification of the complex phenomena involved, long investigation times and extending basic data determined from laboratory tests in simplified conditions. The basic information on the waste matrix are obtained by mechanical strength tests. In this program, up to now, samples were prepared and disposed as follows: 1)cement-water with cement/water ratio 2:1 and 2)cement-sand-water with a cement/sand/water ratio 1:2:0.5. For the mechanical tests the sizes of the samples were: 40 x 40 x 160 mm (tested at bending and compression) and 20 x 20 x 20 mm (tested at compression). After the preparation, the samples were kept 24 hours in pattern, then in tight plastic bag for 7, 14 and 28 days for reinforcement. The bending and compression tests were done in a construction materials laboratory, using an electric press machine. After 28 days the samples of reference were introduced in the National Repository. Some of

  10. Cementation of liquid radioactive waste

    Efremenkov, V.

    2004-01-01

    The cementation methods for immobilisation of radioactive wastes are discussed in terms of methodology, chemistry and properties of the different types of cements as well as the worldwide experience in this field. Two facilities for cementation - DEWA and MOWA - are described in details

  11. Improved cement solidification of low and intermediate level radioactive wastes

    1993-01-01

    Cementation was the first and is still the most widely applied technique for the conditioning of low and intermediate level radioactive wastes. Compared with other solidification techniques, cementation is relatively simple and inexpensive. However, the quality of the final cemented waste forms depends very much on the composition of the waste and the type of cement used. Different kinds of cement are used for different kinds of waste and the compatibility of a specific waste with a specific cement type should always be carefully evaluated. Cementation technology is continuously being developed in order to improve the characteristics of cemented waste in accordance with the increasing requirements for quality of the final solidified waste. Various kinds of additives and chemicals are used to improve the cemented waste forms in order to meet all safety requirements. This report is meant mainly for engineers and designers, to provide an explanation of the chemistry of cementation systems and to facilitate the choice of solidification agents and processing equipment. It reviews recent developments in cementation technology for improving the quality of cemented waste forms and provides a brief description of the various cement solidification processes in use. Refs, figs and tabs

  12. Study on the mechanisms of solidification for Cs+ and Sr2+ in alkali-activated slag cement waste forms

    Yan Sheng; Shen Xiaodong; Lu Linchao; Wu Xuequan; Wen Yinghui

    1994-09-01

    The mechanisms of adsorption and chemical immobilization in Alkali-Activated Slag Cement (AASC) waste forms were quantitatively studied for Cs + and Sr 2+ . Experimental results show that, hardened AASC paste with zeolite and silica fume possesses strong adsorption ability and anti-desorption ability for Cs + and Sr 2+ . The results and analyzed by XRD show that CS + and SR 2+ could be incorporated into the lattice of C-S-H by replacing Ca 2+ in C-S-H. The d 111 values of C-S-H increase with increasing CsNO 3 and Sr(NO 3 ) 2 dosages in a limited and regular step. After washed out by deionized water in a ultrasonic washer, samples were analyzed by SEM-EDS and a lot of Cs + and Sr 2+ were detected. This means that the hydration products of AASC have a high retaining ability for Cs + and Sr 2+ . (3 tabs., 8 figs.)

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

    Burbank, D.A.; Weingardt, K.M.

    1993-01-01

    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

  14. Study on cementation of simulated radioactive borated liquid wastes

    Sun Qina; Li Junfeng; Wang Jianlong

    2010-01-01

    To compare sulfoaluminate cement with ordinary Portland cement on their cementation of radioactive borated liquid waste and to provide more data for formula optimization, simulated radioactive borated liquid waste were solidified by the two cements. 28 d compressive strength and strength losses after water/freezing/irradiation resistance tests were investigated. Leaching test and X-ray diffraction analysis were also conducted. The results show that it is feasible to solidify borated liquid wastes with sulfoaluminate cement and ordinary Portland cement with formulas used in the study. The 28 d compressive strengths, strength losses after tests and simulated nuclides leaching rates of the solidified waste forms meet the demand of GB 14569.1-93. The sulfoaluminate cement formula show better retention of Cs + than ordinary Portland cement formula. Boron, in form of B (OH) 4 - , incorporate in ettringite as solid solutions. (authors)

  15. Comparative waste forms study

    Wald, J.W.; Lokken, R.O.; Shade, J.W.; Rusin, J.M.

    1980-12-01

    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

  16. Centralized cement solidification technique for low-level radioactive wastes

    Matsuda, Masami; Nishi, Takashi; Izumida, Tatsuo; Tsuchiya, Hiroyuki.

    1996-01-01

    A centralized cement solidification system has been developed to enable a single facility to solidify such low-level radioactive wastes as liquid waste, spent ion exchange resin, incineration ash, and miscellaneous solid wastes. Since the system uses newly developed high-performance cement, waste loading is raised and deterioration of waste forms after land burial prevented. This paper describes the centralized cement solidification system and the features of the high-performance cement. Results of full-scale pilot plant tests are also shown from the viewpoint of industrial applicability. (author)

  17. Cement-Based Materials for Nuclear Waste Storage

    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...

  18. Solidification of ion exchange resin wastes in hydraulic cement

    Neilson, R.M. Jr.; Kalb, P.; Fuhrmann, M.; Colombo, P.

    1982-01-01

    Work has been conducted to investigate the solidification of ion exchange resin wastes with portland cements. These efforts have been directed toward the development of acceptable formulations for the solidification of ion exchange resin wastes and the characterization of the resultant waste forms. This paper describes formulation development work and defines acceptable formulations in terms of ternary phase compositional diagrams. The effects of cement type, resin type, resin loading, waste/cement ratio and water/cement ratio are described. The leachability of unsolidified and solidified resin waste forms and its relationship to full-scale waste form behavior is discussed. Gamma irradiation was found to improve waste form integrity, apparently as a result of increased resin crosslinking. Modifications to improve waste form integrity are described. 3 tables

  19. Cementation unit for radioactive wastes

    Dellamano, Jose Claudio; Vicente, Roberto; Lima, Jose Rodrigues de

    2001-01-01

    This communication describes the waste cementation process and facility developed at Instituto de Pesquisas Energeticas e Nucleares - IPEN. The process is based on 200 litres batch operation, in drum mixing, with continuous cement feeding. The equipment is a single recoverable helicoidal mixer and a turning table that allows the drum to rotate during the mixing operation, simulating a planetary mixer. The facility was designed to treat contact handled liquids and wet solid wastes, but can be adapted for shielded equipment and remote operation. (author)

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

    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.

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

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

    2014-01-01

    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

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

    Zhou, H.; Colombo, P.

    1987-03-01

    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 (CaOH 2 ) 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 % Na 2 SO 4 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

  3. Modelling the effects of waste components on cement hydration

    van Eijk, R.J.; Brouwers, Jos

    2000-01-01

    Ordinary Portland Cement (OPC) is often used for the Solidification/Stabilization (S/S) of waste containing heavy metals and salts. These waste componenents will precipitate in the form of insoluble compounds onto unreacted cement clinker grains preventing further hydration. In this study the long

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

    Kalb, P.D.; Heiser, J.H. III; Colombo, P.

    1990-01-01

    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. Because hydraulic cement requires a chemical hydration reaction for setting and hardening, it is subject to potential interactions between elements in the waste and binder that can retard or prevent solidification. 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 its melting point, 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

  5. The influence of moisture content of sand on leach rates from cement and bitumen radioactive waste forms in a shallow ground repository

    Lovasic, Z.; Torok, J.; Buckley, L.P.

    1992-11-01

    Leaching rates from cement and bitumen-conditioned waste in contact with sand were determined with different water content of the sand up to and including saturation. A cylindrical waste form occupied one end of a column, with sand packed in the remainder. Axial radionuclide concentration profiles in the columns were determined several times during the experiment using an automated system consisting of a gamma spectrometer and an optical bench. The migration of 85 Sr, 137 Cs and 60 Co was determined from concentration profiles with time. The diffusion coefficient for 137 Cs in cement varied more than expected with the change of moisture content of sand. The leaching of bitumen waste forms is complex, and has poor reproducibility. When the water content of the sand was below 0.067 mL/mL, no leaching was determined from bitumen waste forms in all of the specimens except one. No clear relationship could be established between leach rate and the water content of sand above 0.18 mL/mL water content

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

    Shengtao, Feng; Li, Gong; Li, Cheng; Benli, Wang; Lihong, Wang [China Inst. for Radiation Protection, Taiyuan, Shanxi (China)

    1997-02-01

    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 DH{sub 4A} 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 DH{sub 4A} 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, DH{sub 4A} 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.

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

    Feng Shengtao; Gong Li; Cheng Li; Wang Benli; Wang Lihong

    1997-01-01

    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 DH 4A 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 DH 4A 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, DH 4A 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

  8. Processing method for cleaning water waste from cement kneader

    Soda, Kenzo; Fujita, Hisao; Nakajima, Tadashi.

    1990-01-01

    The present invention concerns a method of processing cleaning water wastes from a cement kneader in a case of processing liquid wastes containing radioactive wastes or deleterious materials such as heavy metals by means of cement solidification. Cleaning waste wastes from the kneader are sent to a cleaning water waste tank, in which gentle stirring is applied near the bottom and sludges are retained so as not to be coagulated. Sludges retained at the bottom of the cleaning water waste tank are sent after elapse of a predetermined time and then kneaded with cements. Thus, since the sludges in the cleaning water are solidified with cement, inhomogenous solidification products consisting only of cleaning sludges with low strength are not formed. The resultant solidification product is homogenous and the compression strength thereof reaches such a level as capable of satisfying marine disposal standards required for the solidification products of radioactive wastes. (I.N.)

  9. Safety disposal studies of radioactive and hazardous wastes using cement

    Aly, M.M.E.

    2000-01-01

    radioactive waste is generated from the production of nuclear energy and from the use of radioactive materials applications, agriculture and medicine. the important of safe management of radioactive waste for the protection of human health and the environment has long been recognized. conditioning of radioactive waste is the transform of radioactive waste into a suitable form for storage and disposal. common immobilization methods include solidification of low radioactive waste in cement or bitumen.in order to improve cement properties to decrease the release of liquid radioactive waste into the environment and its dispersion to a level where the risks to individuals, population and the environment

  10. Immobilization of radioactive waste in cement-based matrices

    Glasser, F.P.; Rahman, A.A.; Crawford, R.W.; McCulloch, C.E.; Angus, M.J.

    1984-01-01

    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)

  11. Modified sulfur cement solidification of low-level wastes

    1985-10-01

    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.

  12. Modified sulfur cement solidification of low-level wastes

    1985-10-01

    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

  13. Immobilization of radioactive waste through cementation using Cuban zeolitic rock as additive

    Chales Suarez, G.; Castillo Gomez, R.

    1997-01-01

    The cementation of both simulated and real low level aqueous wastes using Cuban zeolite as additive is described. Mechanical characteristics and leach testing of the cemented waste forms has been studied. The results obtained have shown that the presence of zeolite in the cemented waste for reduces considerably the leach rates of Cs and Co and moreover, mechanical characteristics (set time and compressive strength) are better when compared with direct cementation of aqueous wastes. (author). 13 refs, 8 tabs

  14. PURIFIED WASTE FCC CATALYST AS A CEMENT REPLACEMENT MATERIAL

    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.

  15. 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

    Atkins, M.; Beckley, N.; Carson, S.; Cowie, J.; Glasser, F.P.; Kindness, A.; Macphee, D.; Pointer, C.; Rahman, A.; Jappy, J.G.; Evans, P.A.; McHugh, G.; Natingley, N.J.; Wilding, C.

    1991-01-01

    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 ∼ 20 0 C by relatively pure water. It embraces mainly six chemical components - Na 2 O, K 2 O, CaO, MgO, SiO 2 and water - together with limited data on the effect of sulphate, SO 4 -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

  16. Survey of concrete waste forms

    Moore, J.G.

    1981-01-01

    The incorporation of radioactive waste in cement has been widely studied for many years. It has been routinely used at nuclear research and production sites for some types of nuclear waste for almost three decades and at power reactor plants for nearly two decades. Cement has many favorable characteristics that have contributed to its popularity. It is a readily available material and has not required complex and/or expensive equipment to solidify radioactive waste. The resulting solid products are noncombustible, strong, radiation resistant, and have reasonable chemical and thermal stability. As knowledge increased on the possible dangers from radioactive waste, requirements for waste fixation became more stringent. A brief survey of some of the research efforts used to extend and improve cementitious waste hosts to meet these requirements is given in this paper. Selected data are presented from the rather extensive study of the applicability of concrete as a waste form for Savannah River defense waste and the use of polymer impregnation to reduce the leachability and improve the durability of such waste forms. Hot-pressed concretes that were developed as prospective host solids for high-level wastes are described. Highlights are given from two decades of research on cementitious waste forms at Oak Ridge National Laboratory. The development of the hydrofracture process for the disposal of all locally generated radioactive waste led to a process for the disposal of I-129 and to the current research on the German in-situ solidification process for medium-level waste and the Oak Ridge FUETAP process for all classes of waste including commercial and defense high-level wastes. Finally, some of the more recent ORNL concepts are presented for the use of cement in the disposal of inorganic and biological sludges, waste inorganic salts, trash, and krypton

  17. Cementation of the solid radioactive waste with polymer-cement solutions using the method of impregnation

    Gorbunova, O.

    2015-01-01

    Cementation of solid radioactive waste (SRW), i.e. inclusion of solid radioactive waste into cement matrix without cavities - is one of the main technological processes used for conditioning low and intermediate level radioactive waste. At FSUE 'Radon' the industrialized method of impregnation has been developed and since 2003 has been using for cementation of solid radioactive waste. The technology is that the polymer-cement solution, having high penetrating properties, is supplied under pressure through a tube to the bottom of the container in which solid radioactive waste has preliminarily been placed. The polymer-cement solution is evenly moving upwards through the channels between the particles of solid radioactive waste, fills the voids in the bulk volume of the waste and hardens, forming a cement compound, the amount of which is equal to the original volume. The aim of the investigation was a selection of a cement solution suitable for SRW impregnation (including fine particles) without solution depletion and bottom layers stuffing. It has been chosen a polymer: PHMG (polyhexamethylene-guanidine), which is a stabilizing and water-retaining component of the cement solution. The experiments confirm that the polymer increases the permeability of the cement solution by a 2-2.5 factor, the viscosity by a 1.2 factor, the stability of the consistency by a 1.5-1.7 factor, and extends the operating range of the W/C ratio to 0.5-1.1. So it is possible to penetrate a volume of SRW bigger by a 1.5-2.0 factor. It has been proved, that PHMG polymer increases strength and frost-resistance of the final compounds by a 1.8-2.7 factor, and contributes to fast strength development at the beginning of hardening and it decreases Cs-137 leashing rate by a 1.5-2 factor

  18. Immobilization of radioactive waste in cement based matrices

    Glasser, F.P.; Rahman, A.A.; Macphee, D.; McCulloch, C.E.; Angus, M.J.

    1985-06-01

    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. Leach characterization of cement encapsulated wastes

    Roy, D.M.; Scheetz, B.E.; Wakeley, L.D.; Barnes, M.W.

    1982-01-01

    Matrix encapsulation of defense nuclear waste as well as intermediate-level commercial wastes within a low-temperature cementitious composite were investigated. The cements for this study included both as-received and modified calcium silicate and calcium aluminate cements. Specimens were prepared following conventional formulation techniques designed to produce dense monoliths, followed by curing at 60 0 C. An alternative preparation procedure is contrasted in which the specimens were ''warm'' pressed in a uniaxial press at 150 0 C at 50,000 psi for 0.5 h. Specimens of the waste/cement composites were leached in deionized water following three different procedures which span a wide range of temperatures and solution saturation conditions. Aluminate and compositionally adjusted silicate cements exhibited a better retentivity for Cs and Sr than did the as-received silicate cement. 15 refs

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

    Yufei Yang

    2013-01-01

    Full Text Available 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, clinker was mixed water to prepare cement paste. EPMA results showed that As was generally distributed throughout the cement paste. As content in calcium silicate hydrates gel (C-S-H was in low level, but higher than that in other cement mineral phases. This means that most of As is expected to form some compounds that disperse on the surfaces of cement mineral phases. Linear combination fitting (LCF of the X-ray absorption near edge structure spectra revealed that As in the cement paste was predominantly As(V and mainly existed as Mg3(AsO42, Ca3(AsO42, and Na2HAsO4.

  1. Waste form development/test

    Kalb, P.D.; Colombo, P.

    1983-01-01

    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

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

    Ross, W.A.; Lokken, R.O.; May, R.P.; Roberts, F.P.; Timmerman, C.L.; Treat, R.L.; Westsik, J.H. Jr.

    1982-09-01

    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 10 5 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

  3. Characterization of surrogate radioactive cemented waste: a laboratory study

    Fiset, J.F.; Lastra, R.; Bilodeau, A.; Bouzoubaa

    2011-01-01

    Portland cement is commonly used to stabilize intermediate and low level of radioactive wastes. The stabilization/solidification process needs to be well understood as waste constituents can retard or activate cement hydration. The objectives of this project were to prepare surrogate radioactive cemented waste (SRCW), develop a comminution strategy for SRCW, determine its chemical characteristics, and develop processes for long term storage. This paper emphasizes on the characterization of surrogate radioactive cemented waste. The SRCW produced showed a high degree of heterogeneity mainly due to the method used to add the solution to the host cement. Heavy metals such as uranium and mercury were not distributed uniformly in the pail. Mineralogical characterization (SEM, EDS) showed that uranium is located around the rims of hydrated cement particles. In the SRCW, uranium occurs possibly in the form of a hydrated calcium uranate.The SEM-EDS results also suggest that mercury occurs mainly in the form of HgO although some metallic mercury may be also present as a result of partial decomposition of the HgO. (author)

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

    Kalb, P.D.; Heiser, J.H. III; Pietrzak, R.; Colombo, P.

    1991-01-01

    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

  5. Cementation of wastes with boric acid

    Tello, Cledola C.O.; Haucz, Maria Judite A.; Alves, Lilian J.L.; Oliveira, Arno H.

    2000-01-01

    In nuclear power plants (PWR) are generated wastes, such as concentrate, which comes from the evaporation of liquid radioactive wastes, and spent resins. Both have boron in their composition. The cementation process is one of the options to solidify these wastes, but the boron has a negative effect on the setting of the cement mixture. In this paper are presented the experiments that are being carried out in order to overcome this problem and also to improve the efficiency of the process. Simulated wastes were cemented using additives (clays, admixtures etc.). In the process and product is being evaluated the effect of the amount, type and addition order of the materials. The mixtures were selected in accordance with their workability and incorporated waste. The solidified products are monolithic without free water with a good mechanical resistance. (author)

  6. Practical Model of Cement Based Grout Mix Design, for Use into Low Level Radiation Waste Management

    Radu Lidia

    2015-12-01

    Full Text Available The cement based grouts, as functional performance composite materials, are widely used for both immobilisation and encapsulation as well as for stabilization in the field of inorganic waste management. Also, to ensure that low level radioactive waste (LLW are contained for storage and ultimate disposal, they are encapsulated or immobilized in monolithic waste forms, with cement –based grouts.

  7. 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

    Angelis, G. de; Marchetti, A.; Balzamo, S.

    1992-01-01

    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

  8. The fixation of radioactive wastes in cement

    Kulichenko, V.V.; Dukhovich, F.S.; Volkova, O.I.; Boyarinova, M.V.

    1976-01-01

    The authors study the leaching behaviour of the main long-lived fission products 90 Sr and 137 Cs. It is found that 90 Sr and 137 Cs have high elution values, namely (2-12) x 10 -2 resp. (2-6) x 10 -2 g/cm 2 /24h, independently of the type of waste. On the basis of these results, maximum concentrations for the solutions in the cement/solution mixtures are proposed. Further studies relate to the formation of radiolysis gas in the waste fixed to cement. Experiments are described to make use of the empty space in the containers, filled with solid waste by filling them with mixtures of cement and liquid radioactive waste of 10 -4 to 1- 6 Ci. The ratio solution/cement should amount to 0.5. The containers are then buried underground. This method of combined waste storage helped to reduce the cost for the storage of liquid waste by about 40-50%. (RB) [de

  9. Portland blended cements: demolition ceramic waste management

    Trezza, M.A.; Zito, S.; Tironi, A.; Irassar, E.F.; Rahhal, V.F.

    2017-01-01

    Demolition ceramic wastes (DCWs) were investigated in order to determine their potential use as supplementary cementitious materials in Portland Blended Cements (PBCs). For this purpose, three ceramic wastes were investigated. After characterization of the materials used, the effect of ceramic waste replacement (8, 24 and 40% by mass) was analyzed. Pozzolanic activity, hydration progress, workability and compressive strength were determined at 2, 7 and 28 days. The results showed that the ground wastes behave as filler at an early age, but as hydration progresses, the pozzolanic activity of ceramic waste contributes to the strength requirement. [es

  10. Package materials, waste form

    Anon.

    1980-01-01

    The schedules for waste package development for the various host rocks were presented. The waste form subtask activities were reviewed, with the papers focusing on high-level waste, transuranic waste, and spent fuel. The following ten papers were presented: (1) Waste Package Development Approach; (2) Borosilicate Glass as a Matrix for Savannah River Plant Waste; (3) Development of Alternative High-Level Waste Forms; (4) Overview of the Transuranic Waste Management Program; (5) Assessment of the Impacts of Spent Fuel Disassembly - Alternatives on the Nuclear Waste Isolation System; (6) Reactions of Spent Fuel and Reprocessing Waste Forms with Water in the Presence of Basalt; (7) Spent Fuel Stabilizer Screening Studies; (8) Chemical Interactions of Shale Rock, Prototype Waste Forms, and Prototype Canister Metals in a Simulated Wet Repository Environment; (9) Impact of Fission Gas and Volatiles on Spent Fuel During Geologic Disposal; and (10) Spent Fuel Assembly Decay Heat Measurement and Analysis

  11. Densified waste form and method for forming

    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.

  12. Mechanisms and modelling of waste-cement and cement-host rock interactions

    2017-06-01

    Safe and sustainable disposal of hazardous and radioactive waste is a major concern in today's industrial societies. The hazardous waste forms originate from residues of thermal treatment of waste, fossil fuel combustion and ferrous/non-ferrous metal smelting being the most important ones in terms of waste production. Low- and intermediate-level radioactive waste is produced in the course of nuclear applications in research and energy production. For both waste forms encapsulation in alkaline, cement-based matrices is considered to ensure long-term safe disposal. Cementitious materials are in routine use as industrial materials and have mainly been studied with respect to their evolution over a typical service life of several decades. Use of these materials in waste management applications, however, requires assessments of their performance over much longer time periods on the order of thousands to several ten thousands of years.

  13. Basalt waste added to Portland cement

    Thiago Melanda Mendes

    2016-08-01

    Full Text Available Portland cement is widely used as a building material and more than 4.3 billion tons were produced in 2014, with increasing environmental impacts by this industry, mainly through CO2 emissions and consumption of non-removable raw materials. Several by-products have been used as raw materials or fuels to reduce environmental impacts. Basaltic waste collected by filters was employed as a mineral mixture to Portland cement and two fractions were tested. The compression strength of mortars was measured after 7 days and Scanning Electron Microscopy (SEM and Electron Diffraction Scattering (EDS were carried out on Portland cement paste with the basaltic residue. Gains in compression strength were observed for mixtures containing 2.5 wt.% of basaltic residue. Hydration products observed on surface of basaltic particles show the nucleation effect of mineral mixtures. Clinker substitution by mineral mixtures reduces CO2 emission per ton of Portland cement.

  14. Alternative solidified forms for nuclear wastes

    McElroy, J.L.; Ross, W.A.

    1976-01-01

    Radioactive wastes will occur in various parts of the nuclear fuel cycle. These wastes have been classified in this paper as high-level waste, intermediate and low-level waste, cladding hulls, and residues. Solidification methods for each type of waste are discussed in a multiple barrier context of primary waste form, applicable coatings or films, matrix encapsulation, canister, engineered structures, and geological storage. The four major primary forms which have been most highly developed are glass for HLW, cement for ILW, organics for LLW, and metals for hulls

  15. Leaching tests of cemented organic radioactive waste

    Calabria, Jaqueline A. Almeida; Haucz, Maria Judite A.; Tello, Cledola Cassia O.

    2011-01-01

    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 A cceptance 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)

  16. Leaching of nuclear power reactor wastes forms

    Endo, L.S.; Villalobos, J.P.; Miyamoto, H.

    1986-01-01

    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) [pt

  17. Measurement and control of cement set times in waste solidification

    Stone, J.A.; d'Entremont, P.D.

    1976-09-01

    Fixation of radioactive waste in concrete was investigated on laboratory scale. Some cement formulations containing simulated or actual sludges from the Savannah River Plant had set times that would be too short for reliable handling in plant equipment. Set times could be controlled by use of excess water, but the concrete forms produced had inferior strength. A commercial organic retarder was found to be effective for increasing set times of cement-sludge formulations. However, the dosage of retarder required to control set times of high-alumina cement formulations was 1.0 to 1.5 wt percent of dry solids, which is 5 to 10 times the normal dosage for Portland cements. Data were obtained to predict the optimum content of retarder and water

  18. Hospital waste ashes in Portland cement mortars

    Genazzini, C.; Zerbino, R.; Ronco, A.; Batic, O.; Giaccio, G.

    2003-01-01

    Nowadays, most concretes incorporate mineral additions such as pozzolans, fly ash, silica fume, blast furnace slag, and calcareous filler among others. Although the technological and economical benefits were the main reasons for the use of mineral additions, the prevention of environmental contamination by means of proper waste disposal becomes a priority. The chance of incorporating hospital waste ashes in Portland cement-based materials is presented here. Ash characterization was performed by chemical analysis, X-ray diffraction, radioactive material detection, and fineness and density tests. Conduction calorimetry and setting time tests were developed on pastes including ash contents from 0% to 100%. Mortars were prepared including ash contents up to 50% of cement. The results of setting time, temperature development, flexural and compressive strengths, water absorption, density, and leachability are analyzed. Results indicate that Portland cement systems could become an alternative for the disposal of this type of ashes

  19. Near-field modelling in cement environments. Task 3. Characterization of radioactive waste forms. A series of final reports (1985-89). No 32B

    Haworth, A.; Sharland, S.M.; Tweed, C.J.

    1991-01-01

    The models used in the assessment of the performance of a radioactive waste repository are necessarily simple and must make assumptions in many areas. The aim of this project is to develop research models to examine various physical and chemical processes in the nearfield of a repository in more detail to justify some of these assumptions. The Cheqmate program has been developed to study coupled chemistry and ionic-transport processes. The numerical method used in the program has recently been improved and models of sorption included in the chemical part of the code. A model of the degradation of cement in a repository has been developed using Cheqmate. In a previous study, the scale of perturbation to the ground-water chemistry surrounding a repository due to the ingress of high pH from the backfill was investigated. The thermodynamic description of clay has been improved and used in a validation exercise for Cheqmate from information provided by the British Geological Survey. A model of diffusion/sorption experiments through cement has also been developed. This model is intended to provide a framework for modelling transport of sorbing species through cement. 11 figs.; 4 tabs.; 20 refs

  20. Method of processing radioactive liquid wastes by solidification with cement

    Yasumura, Keijiro; Matsuura, Hiroyuki.

    1975-01-01

    Object: To subject radioactive liquid wastes to a cement solidification treatment after heating and drying it by a thin film scrape-off drier to render it into the form of power, and then molding it into pellets for the treatment. Structure: Radioactive liquid wastes discharged from a nuclear power plant or nuclear reactor are supplied through a storage tank into a thin film scrape-off drier. In the drier, the radioactive liquid wastes are heated to separate the liquid, and the residue is taken out as dry powder from the scrape-off apparatus. The powder obtained in this way is molded into pellets of a desired form. These pellets are then packed in a drum can or similar container, into which cement paste is then poured for solidification. (Moriyama, K.)

  1. Status of waste form testing

    Lawroski, H.

    1984-01-01

    The promulgation of the amendment of 10 CFR Part 61 by the Nuclear Regulatory Commission of December 27, 1982 by Federal Register Notice with an effective date of December 27, 1983 established the criteria for licensing requirements, paragraph 60.56, contained the description to provide adequate stability of the site through the use of suitable waste forms. In May, 1983, the NRC published a final Branch Technical Position (BTP) paper on waste form. The position taken by the BTP was considerably more severe than indicated in 10 CFR Part 61. An extensive and expensive testing program was started in 1983. As an interim measure, the presently utilized solidification processes such as cement, Dow binder, Envirostone and bitumen, and the presently qualified High Integrity containers (HICs) were considered acceptable with the caveat that acceptable process control programs were being utilized. The NRC requested that topical reports for licenses be submitted. The topical reports were to contain test results to substantiate the acceptability of the waste forms. The test results to date show that the volume of wastes will have to increase to meet the position taken by the NRC in the BTP. This position will cause more waste to be generated which is contrary to the emphasis by states and others to reduce the volume of waste. The details of testing will be discussed in the paper to be presented

  2. Immobilisation of shredded waste in a cement monolith

    James, J.M.; Smith, D.L.

    1987-11-01

    During 1983/84 work was continued on the development of the process for the encapsulation of shredded waste in cement. Using simulant shredded waste the conditions for operating the process on the 500 litres scale have been established. Evaluation of the cemented product showed that it was satisfactorily infilled with cement grout with no significant voidage. (author)

  3. Carbon-14 behavior in a cement-dominated environment: Implications for spent CANDU resin waste disposal

    Dayal, R.; Reardon, E.J.

    1994-01-01

    Cement based waste forms and concrete engineered barriers are expected to play a key role in providing 14 C waste containment and control 14 C migration for time periods commensurate with its hazardous life of about 50,000 years. The main thrust of this study was, therefore, to evaluate the performance of cement based waste forms with regard to 14 C containment. Of particular importance are the geochemical processes controlling 14 C solubility and release under anticipated cement dominated low and intermediate level waste repository conditions. Immobilization of carbonate-form exchange resin in grout involves transfer of sorbed 14 CO 3 2- ions, through exchange for hydroxyl ions from the grout slurry, followed by localized precipitation of solid calcium carbonate at the cement/resin interface in the grout matrix. Carbon-14 release behavior can be attributed to the dissolution characteristics and solubility of calcite present in the cement based waste form. The groundwater flow regime can exert a pronounced effect both on the near-field chemistry and the leaching behavior of 14 C. For a cement dominated repository, at relatively low-flow or stagnant groundwater conditions, the alkaline near-field chemical environments inhibits the release of 14 C from the cemented waste form. Under high flow conditions, the near-field environment is characterized by relatively neutral pH conditions which promote calcite dissolution, thus resulting in 14 C release from the waste form

  4. Fixation of calcined waste by bituminization or cementation

    Napravnik, J.; Kyrs, M.; Ditl, P.

    1983-01-01

    The overall concept is given of the combination of calcination with fixation into bitumen, cement etc. The design is shown of a calciner with the capacity of 10 L/h which was tested on real radioactive wastes for 2000 h. The geometrical and operating parameters of the apparatus have been optimized based on a statistical evaluation of the experiments. Wastes containing nitrates are calcined at 300-550 deg. C, yielding oxides. Wastes containing sulphates, carbonates, KMnO 4 , or borates are calcined at 150-330 deg. C, yielding soluble salts. The content of H 3 BO 4 and Na 2 B 4 O 7 and in some cases of sulphates in the calcinate retards hardening of the mixture with cement. Nitrates and detergents also interfere. The effect of the above components on the products mixed with bitumen is much less. Detergents can be decomposed at 200-300 deg. C; organic acids can be reacted with A1 salts to form insoluble substances lowering the leaching rate of Sr and Cs; small amounts of SiO 2 eliminate the effect of borates on cement hardening. The drawbacks of bituminization with bitumen emulsions are the complicated preparation of the emulsion, higher leaching rate of the product and low stability of the emulsion against breaking. The leachability was determined (1-50 days) of different products containing LWR wastes: 33% of concentrated waste in cement of calcination product stabilized with PVA exhibit approx. 8x10 - 3 g/cm 2 per day, 33% of calcine in cement approx. 3x10 - 3 ; 40% concentrate fixed with bitumen emulsion approx. 9x10 - 4 ; 50% calcine stabilized with PVA in bitumen, pilot-plant scale approx. 2x10 - 5 ; the same but on a laboratory scale approx. 1.10 - 5 . (author)

  5. Use of Incineration Solid Waste Bottom Ash as Cement Mixture in Cement Production

    Jun, N. H.; Abdullah, M. M. A. B.; Jin, T. S.; Kadir, A. A.; Tugui, C. A.; Sandu, A. V.

    2017-06-01

    Incineration solid waste bottom ash was use to examine the suitability as a substitution in cement production. This study enveloped an innovative technology option for designing new equivalent cement that contains incineration solid waste bottom ash. The compressive strength of the samples was determined at 7, 14, 28 and 90 days. The result was compared to control cement with cement mixture containing incineration waste bottom ash where the result proved that bottom ash cement mixture able achieve its equivalent performance compared to control cement which meeting the requirement of the standards according to EN 196-1. The pozzolanic activity index of bottom ash cement mixture reached 0.92 at 28 days and 0.95 at 90 and this values can be concluded as a pozzolanic material with positive pozzolanic activity. Calcium hydroxide in Portland cement decreasing with the increasing replacement of bottom ash where the reaction occur between Ca(OH)2 and active SiO2.

  6. Concrete and cement composites used for radioactive waste deposition.

    Koťátková, Jaroslava; Zatloukal, Jan; Reiterman, Pavel; Kolář, Karel

    2017-11-01

    This review article presents the current state-of-knowledge of the use of cementitious materials for radioactive waste disposal. An overview of radwaste management processes with respect to the classification of the waste type is given. The application of cementitious materials for waste disposal is divided into two main lines: i) as a matrix for direct immobilization of treated waste form; and ii) as an engineered barrier of secondary protection in the form of concrete or grout. In the first part the immobilization mechanisms of the waste by cement hydration products is briefly described and an up-to date knowledge about the performance of different cementitious materials is given, including both traditional cements and alternative binder systems. The advantages, disadvantages as well as gaps in the base of information in relation to individual materials are stated. The following part of the article is aimed at description of multi-barrier systems for intermediate level waste repositories. It provides examples of proposed concepts by countries with advanced waste management programmes. In the paper summary, the good knowledge of the material durability due to its vast experience from civil engineering is highlighted however with the urge for specific approach during design and construction of a repository in terms of stringent safety requirements. Copyright © 2017 Elsevier Ltd. All rights reserved.

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

    Rahman, A.A.; Glasser, F.P.

    1987-01-01

    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

  8. Radionuclide Retention in Concrete Waste Forms

    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.

  9. Leaching of nuclear power reactor waste forms

    Endo, L.S.; Villalobos, J.P.; Miyamoto, H.

    1987-01-01

    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 cm 2 /s for boric acid waste form and 9 x 10 -9 cm 2 /s for ion-exchange resin waste. Strontium diffusivity coefficients found are about 3 x 10 -11 cm 2 /s and 9 x 10 -11 cm 2 /s respectively. (Author) [pt

  10. Overview of waste stabilization with cement.

    Batchelor, B

    2006-01-01

    Cement can treat a variety of wastes by improving physical characteristics (solidification) and reducing the toxicity and mobility of contaminants (stabilization). Potentially adverse waste-binder interactions are an important consideration because they can limit solidification. Stabilization occurs when a contaminant is converted from the dissolved (mobile) phase to a solid (immobile) phase by reactions, such as precipitation, sorption, or substitution. These reactions are often strongly affected by pH, so the presence of components of the waste that control pH are critical to stabilization reactions. Evaluating environmental impacts can be accomplished in a tiered strategy in which simplest approach would be to measure the maximum amount of contaminant that could be released. Alternatively, the sequence of release can be determined, either by microcosm tests that attempt to simulate conditions in the disposal zone or by mechanistic models that attempt to predict behavior using fundamental characteristics of the treated waste.

  11. Waste-form development

    Neilson, R.M. Jr.; Colombo, P.

    1982-01-01

    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

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

    Shatta, H.A.; Saleh, H.M.

    2006-01-01

    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

  13. Evaluating the cement stabilization of arsenic-bearing iron wastes from drinking water treatment.

    Clancy, Tara M; Snyder, Kathryn V; Reddy, Raghav; Lanzirotti, Antonio; Amrose, Susan E; Raskin, Lutgarde; Hayes, Kim F

    2015-12-30

    Cement stabilization of arsenic-bearing wastes is recommended to limit arsenic release from wastes following disposal. Such stabilization has been demonstrated to reduce the arsenic concentration in the Toxicity Characteristic Leaching Procedure (TCLP), which regulates landfill disposal of arsenic waste. However, few studies have evaluated leaching from actual wastes under conditions similar to ultimate disposal environments. In this study, land disposal in areas where flooding is likely was simulated to test arsenic release from cement stabilized arsenic-bearing iron oxide wastes. After 406 days submersed in chemically simulated rainwater, wastes. Presenting the first characterization of cement stabilized waste using μXRF, these results revealed the majority of arsenic in cement stabilized waste remained associated with iron. This distribution of arsenic differed from previous observations of calcium-arsenic solid phases when arsenic salts were stabilized with cement, illustrating that the initial waste form influences the stabilized form. Overall, cement stabilization is effective for arsenic-bearing wastes when acidic conditions can be avoided. Copyright © 2015 Elsevier B.V. All rights reserved.

  14. Characteristics of Cement Solidification of Metal Hydroxide Waste

    Dae-Seo Koo

    2017-02-01

    Full Text Available To perform the permanent disposal of metal hydroxide waste from electro-kinetic decontamination, it is necessary to secure the technology for its solidification. The integrity tests on the fabricated solidification should also meet the criteria of the Korea Radioactive Waste Agency. We carried out the solidification of metal hydroxide waste using cement solidification. The integrity tests such as the compressive strength, immersion, leach, and irradiation tests on the fabricated cement solidifications were performed. It was also confirmed that these requirements of the criteria of Korea Radioactive Waste Agency on these cement solidifications were met. The microstructures of all the cement solidifications were analyzed and discussed.

  15. Characteristics of cement solidification of metal hydroxide waste

    Koo, Dae Seo; Sung, Hyun Hee; Kim, Seung Soo; Kim, Gye Nam; Choi, Jong Won [Dept. of Decontemination Decommission Technology Development, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-02-15

    To perform the permanent disposal of metal hydroxide waste from electro-kinetic decontamination, it is necessary to secure the technology for its solidification. The integrity tests on the fabricated solidification should also meet the criteria of the Korea Radioactive Waste Agency. We carried out the solidification of metal hydroxide waste using cement solidification. The integrity tests such as the compressive strength, immersion, leach, and irradiation tests on the fabricated cement solidifications were performed. It was also confirmed that these requirements of the criteria of Korea Radioactive Waste Agency on these cement solidifications were met. The microstructures of all the cement solidifications were analyzed and discussed.

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

    Li, J.

    2013-01-01

    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

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

    Phillips, J.A.; Semones, G.B.

    1995-01-01

    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

  18. Immobilization of radioactive waste in cement based matrices

    Glasser, F.P.; Rahman, A.A.; Macphee, S.; Atkins, M.; Beckley, N.; Carson, S.

    1986-11-01

    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-SiO 2 -H 2 O 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)

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

    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

  20. Cementation of liquid radioactive waste with high content of borate salts

    Gorbunova, O.

    2015-01-01

    The report reviews the ways of optimization of cementation of boron-containing liquid radioactive waste. The most common way to hardening the low-level liquid radioactive waste (LRW) is the cementation. However, boron-containing liquid radioactive waste with low pH values cannot be cemented without alkaline additives, to neutralize acid forms of borate compounds. Cement setting without additives happens only on 14-56 days, the compounds have low strength, and hence an insufficient reliability of radionuclides fixation in the cement matrix. The alkaline additives increase the volume of the final cement compound which enhances financial and operational costs. In order to control the speed of hardening of cement solution with a boron-containing liquid radioactive waste and to remove the components that prevent hardening of cement solution, it is proposed an electromagnetic treatment of LRW in the vortex layer of ferromagnetic particles. The results of infrared spectroscopy show, that electromagnetic treatment of liquid radioactive waste changes the ionic forms of the borates and raises the pH due to the dissociation of the oxygen and hydrogen bonds in the aqueous solutions of the boron compounds. The various types of ferromagnetic activators of the vortex layer have been investigated, including the highly dispersed nano-powders and the magnetic phases of the iron oxides. It has been determined the technological parameters of the electromagnetic treatment of liquid radioactive waste and the subsequent cementation of this type of LRW. By using the method of scanning electron microscopy it has been shown, that the nano-particles of magnetic phases of the ferric oxides are involved in phase formation of hydro-aluminum-calcium ferrites in the early stages of hardening and improving strength of the cement compounds with liquid radioactive waste. (authors)

  1. Special waste-form lysimeters: Arid

    Jones, T.L.; Serne, R.J.

    1987-08-01

    The release of contaminant from solidified low-level waste forms is being studied in a field lysimeter facility at the Hanford Site in southeastern Washington State. Duplicate samples of five different waste forms have been buried in 10 lysimeters since March 1984. Waste-form samples represent three different waste streams and four solidification agents (masonry cement, Portland III cement, Dow polymer /sup (a)/, and bitumen). Most precipitation at the Hanford Site arrives as winter snow; this contributes to a strong seasonal pattern in water storage and drainage observed in the lysimeters. The result is an annual range in the volumetric soil water content from 11% in late winter to 7% in the late summer and early fall, as well as annual changes in pore water velocities from approximately 1 cm/wk in early spring to less than 0.05 cm/wk in early fall. Measurable quantities of tritium and cobalt-60 are being collected in lysimeter drainage water. Approximately 30% of the original tritium inventory has been leached from two lysimeters originally containing tritium. Cobalt-60 is present in all waste forms; it is being collected in the leachate from five lysimeters. The total amount released varies, but in each case it is less than 0.1% of the original cobalt inventory of the waste sample. Nonradioactive constituents contained in the waste form, such as sodium, boron, and sulfate, are also being leached

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

    Walker, B.W.

    2000-01-01

    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

  3. Physicochemical changes of cements by ground water corrosion in radioactive waste storage

    Contreras R, A.; Badillo A, V. E.; Robles P, E. F.; Nava E, N.

    2009-10-01

    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)

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

    Glasser, F.P.

    1985-01-01

    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.)

  5. The effect of chemically adjusting cement compositions on leachabilities of waste ions

    Barnes, M.W.; Scheetz, B.E.; Roy, D.M.

    1986-01-01

    The chemical composition of both portland and aluminate cements was adjusted by adding amorphous silica. In the case of portland cement, the object was to react with excess portlandite and obtain an overall composition compatible with C-S-H gel or C-S-H gel + silica at low temperatures, and to obtain the tobermorite composition in order to be in equilibrium with this phase at temperatures above normal ambient. In the case of aluminate cement, the object was to be in equilibrium with more silica-rich phases. These silica-adjusted cements were used to make composites with nuclear waste forms. Leach tests showed that the silica-adjusted composites were chemically more stable than those made with as-received cement. Leach rates were lower in the case of the adjusted cements for Rb, Cs, Ca, Sr, Ba, La, Ce, Nd, Gd, Al, and Si. Only Na in the case of both portland and aluminate cements, and Mg and U in the case of aluminate cements, had greater leach rates in adjusted cements. Adjusting the composition of cements with silica is concluded to be beneficial when making composites to encapsulate nuclear waste forms

  6. Effect of olive waste (Husk on behavior of cement paste

    Sharaf Alkheder

    2016-12-01

    Full Text Available Jordan is a famous country in terms of olive trees agriculture that resulted in a mass production of olive oil products. The huge amounts of olive waste (husk that resulted from olives processing to produce olive oil represent an environmental challenge in the country. The idea in this paper comes to use olive waste as a partial replacement for Portland cement in cement paste to conserve the environment, reduce cement consumption and increase cost efficiency. The wastes were burned properly in an oven and maintained for 6 h until it was fully transformed into ashes. Then, the oven was turned off and ashes were allowed to cool. After cooling, the material passed sieve #200 were used. The sieved ashes were used in the cement mix as a partial cement replacement for making the mortar and cement paste. Normal consistency and setting time were determined as well as soundness, compressive strength. Results indicated that normal consistency of the cement pastes containing different percentage of olive waste is somehow lower than that of the ordinary cement paste and slightly decreases with increasing the percentage. The results also indicated that the compressive strength of hardened blended cement paste containing different percentages of olive waste slightly decrease with olive waste content at 3, 7, and 28 days.

  7. Waste form development

    Neilson, R.M. Jr.; Colombo, P.

    1982-01-01

    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

  8. Properties of backfilling material for solidifying miscellaneous waste using recycled cement from waste concrete

    Matsuda, Atsuo; Yamamoto, Kazuo; Konishi, Masao; Iwamoto, Yoshiaki; Yoshikane, Toru; Koie, Toshio; Nakashima, Yoshio.

    1997-01-01

    A large reduction of total radioactive waste is expected, if recycled cement from the waste concrete of decommissioned nuclear power plants would be able to be used the material for backfilling mortar among the miscellaneous waste. In this paper, we discuss the hydration, strength and consistency of recycled cement compared with normal portland cement. The strength of recycled cement mortar is lower than that of normal portland cement mortar on the same water to cement ratio. It is possible to obtain the required strength to reduce the water to cement ratio by using of high range water-reducing AE agent. According to reducing of water to cement ratio, the P-type funnel time of mortar increase with the increase of its viscosity. However, in new method of self-compactability for backfilling mortar, it became evident that there was no difference between the recycled cement and normal portland cement on the self-compactability. (author)

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

    Zhou, H.; Colombo, P.

    1984-01-01

    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

  10. Leaching properties of solidified TRU waste forms

    Colombo, P.; Neilson, R.M. Jr.

    1979-01-01

    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 239 Pu, counter efficiency, and stability of counting samples

  11. VUJE experience with cementation of liquid and wet radioactive waste

    Kravarik, Kamil; Holicka, Zuzana; Pekar, Anton; Zatkulak, Milan

    2011-01-01

    Liquid and wet LLW generated during operation as well as decommissioning of NPPs is treated with different methods and fixed in a suitable fixation matrix so that a final product meets required criteria for its disposal in a final repository. Cementation is an important process used for fixation of liquid and wet radioactive waste such as concentrate, spent resins and sludge. Active cement grout is also used for fixation of low level solid radioactive waste loaded in final packing containers. VUJE Inc. has been engaged in research of cementation for long. The laboratory for analyzing radioactive waste properties, prescription of cementation formulation and estimation of final cement product properties has been established. Experimental, semi-production cementation plant has been built to optimize operation parameters of cementation. VUJE experience with cementation of liquid and wet LLW is described in the presented paper. VUJE has assisted in commissioning of Jaslovske Bohunice Treatment Centre. Cement formulations for treatment of concentrate, spent resins and sludge have been developed. Research studies on the stability of a final concrete packaging container for disposal in repository have been performed. Gained experience has been further utilized for design and manufacture of several cementation plants for treatment of various liquid and wet LLW. Their main technological and technical parameters as well as characterization of treated waste are described in the paper. Applications include the Mochovce Final Treatment Centre, Movable Cementation Facility utilizing in-drum mixing for treatment of sludge, Cementation Facility for treatment of tritiated water in Latvia and Cementation Facility for fixation of liquid and solid institutional radioactive waste in Bulgaria, which utilizes lost stirrer mixer. (author)

  12. Summary: special waste form lysimeters - arid program

    Skaggs, R.L.; Walter, M.B.

    1987-01-01

    The purpose of the Special Waste Form Lysimeters - Arid Program is to determine the performance of solidified commercial low-level waste forms using a field-scale lysimeter facility constructed for measuring the release and migration of radionuclides from the waste forms. The performance of these waste forms, as measured by radionuclide concentrations in lysimeter effluent, will be compared to that predicted by laboratory characterization of the waste forms. Waste forms being tested include nuclear power reactor waste streams that have been solidified in cement, Dow polymer, and bitumen. To conduct the field leaching experiments a lysimeter facility was built to measure leachate under actual environmental conditions. Field-scale samples of waste were buried in lysimeters equipped to measure water balance components, effluent radionuclide concentrations, and to a limited extent, radionuclide concentrations in lysimeter soil samples. The waste forms are being characterized by standard laboratory leach tests to obtain estimates of radionuclide release. These estimates will be compared to leach rates observed in the field. Adsorption studies are being conducted to determine the amount of contaminant available for transport after the release. Theoretical solubility calculations will also be performed to investigate whether common solid phases could be controlling radionuclide release. 4 references, 8 figures, 1 table

  13. 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

    Amarantos, S.G.; Papadokostaki, K.G.; Petropoulos, J.H.

    1991-01-01

    The leaching kinetics of inactive Cs from cylindrical cement specimens containing Cs 2 SO 4 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 NaNO 3 was performed under exposure to atmospheric CO 2 . 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 Cs 2 SO 4 solutions. Cellulose films containing CaSO 4 or SrSO 4 were leach tested in frequently renewed water at 25 0 C. The elution curves follow a √t law in conformity with the Higuchi equation. Elution tests of NaCl or SrSO 4 embedded in epoxy resin were performed. The SrSO 4 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

  14. Radioactive wastes dispersed in stabilized ash cements

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

    1997-01-01

    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 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 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 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

  15. Decomposition for the analysis of radionuclides in solidified cement radioactive waste

    Lee, Jeong Jin; Pyo, Hyung Yeal; Jee, Kwang Yung; Jeon, Jong Seon

    2004-01-01

    Spent ion exchange resins make solid radioactive wastes when mixed with cement as solidifying material that was widely used in securing human environment from radionuclides for at least hundreds years. The cumulative increase of low and medium level radioactive wastes results in capacity problem of temporary storage in some NPPs (Nuclear Power Plants) of Korea around 2008. Radioactive wastes are scheduled to be disposed in a permanent disposal facility in accordance with the Korean Radioactive Wastes Management Program. It is mandatory to identify kinds and concentration of radionuclides immobilized for transporting them from temporary storage in NPPs to disposal facility. Accordingly, the effective sample decomposition prior to radiochemical separation is prerequisite to obtain the analytical data about radionuclides in cement waste forms. The closed-vessel microwave digestion technology among several sample preparation methods is taken into account to decompose cement waste forms. In this study, SRM 1880a (Portland cement) which is known for its certified values was used to optimize decomposition condition of cement waste forms containing nonradioactive ion exchange resins from NPP. With such variables as reagents, time, and power, the variation of the transparency and the color of the solution after closed-vessel microwave digestion can be examine. SRM 1880a is decomposed by suggested digestion procedure and the recoveries of constituents were investigated by ICP-AES and AAS

  16. Characterization of cement-stabilized Cd wastes

    Maria Diez, J.; Madrid, J.; Macias, A.

    1996-01-01

    Portland cement affords both physical and chemical immobilization of cadmium. The immobilization has been studied analyzing the pore fluid of cement samples and characterizing the solid pastes by X-ray diffraction. The influence of cadmium on the cement hydration and on its mechanical properties has been also studied by isothermal conduction calorimetry and by the measure of strength and setting development. Finally, the effect of cement carbonation on the immobilization of cadmium has been analyzed

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

    Kalb, P.D.; Heiser, J.H. III; Colombo, P.

    1990-01-01

    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

  18. Radioactive Wastes Cementation during Decommissioning Of Salaspils Research Reactor

    Abramenkova, G.; Klavins, M.; Abramenkovs, A.

    2009-01-01

    This paper deals with information on the radioactive wastes cementation technology for decommissioning of Salaspils Research Reactor (SRR). Dismantled radioactive materials were cemented in concrete containers using tritiated water-cement mortar. The laboratory tests system was developed to meet the waste acceptance criteria for disposal of containers with cemented radioactive wastes in near-surface repository 'Radons'. The viscosity of water-cement mortar, mechanical tests of solidified mortar's samples, change of temperature of the samples during solidification time and leakage of Cs-137 and T-3 radionuclides was studied for different water-cement compositions with different additives. The pH and electro conductivity of the solutions during leakage tests were controlled. It was shown, that water/cement ratio significantly influences on water-cement mortar's viscosity and solidified samples mechanical stability. Increasing of water ratio from 0.45 up to 0.62 decreases water-cement mortar's viscosity from 1100 mPas up to 90 mPas and decreases mechanical stability of water-cement samples from 23 N/mm 2 to the 12 N/mm 2 . The role of additives - fly ash and Penetron admix in reduction of solidification temperature is discussed. It was found, that addition of fly ash to the cement-water mortar can reduce the solidification temperature from 81 deg. C up to 62 deg. C. The optimal interval of water ratio in cement mortar is discussed. Tritium and Cs-137 leakage tests show, that radionuclides release curves has a complicate structure. The possible radionuclides release mechanisms are discussed. Experimental results indicated that addition of fly ash result in facilitation of tritium and cesium leakage in water phase. Further directions of investigations are drafted. (authors)

  19. Sulfur polymer cement for macroencapsulation of mixed waste debris

    Mattus, C.H.

    1998-01-01

    In FY 1997, the US DOE Mixed Waste Focus Area (MWFA) sponsored a demonstration of the macroencapsulation of mixed waste debris using sulfur polymer cement (SPC). Two mixed wastes were tested--a D006 waste comprised of sheets of cadmium and a D008/D009 waste comprised of lead pipes and joints contaminated with mercury. The demonstration was successful in rendering these wastes compliant with Land Disposal Restrictions (LDR), thereby eliminating one Mixed Waste Inventory Report (MWIR) waste stream from the national inventory

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

    Dragolici, F.; Nicu, M.; Lungu, L.; Turcanu, C.; Rotarescu, Gh.

    2000-01-01

    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

  1. Formulating a low-alkalinity cement for radioactive waste repositories

    Coumes, C. Cau Dit; Courtois, S.; Leclercq, S.; Bourbon, X.

    2004-01-01

    A multi-annual research program has been launched in January 2003 by CEA, EDF and ANDRA in order to formulate and characterize low-alkalinity and low-heat cements which would be compatible with an underground waste repository environment. Four types of bindings have been investigated: binary blends of Portland cement and silica fume or metakaolin, as well as ternary blends of Portland cement, fly ash and silica fume or metakaolin. Promising results have been obtained with a mixture comprising 37.5% Portland cement, 32.5% silica fume, and 30% fly ash: pH of water in equilibrium with fully hydrated cement is below 11. Moreover, silica fume compensates for the low reactivity of fly ash, while fly ash allows to reduce water demand, heat release, and dimensional variations of cement pastes and mortars. (authors)

  2. Formulating a low-alkalinity cement for radioactive waste repositories

    Coumes, C. Cau Dit; Courtois, S.; Leclercq, S.; Bourbon, X

    2004-07-01

    A multi-annual research program has been launched in January 2003 by CEA, EDF and ANDRA in order to formulate and characterize low-alkalinity and low-heat cements which would be compatible with an underground waste repository environment. Four types of bindings have been investigated: binary blends of Portland cement and silica fume or metakaolin, as well as ternary blends of Portland cement, fly ash and silica fume or metakaolin. Promising results have been obtained with a mixture comprising 37.5% Portland cement, 32.5% silica fume, and 30% fly ash: pH of water in equilibrium with fully hydrated cement is below 11. Moreover, silica fume compensates for the low reactivity of fly ash, while fly ash allows to reduce water demand, heat release, and dimensional variations of cement pastes and mortars. (authors)

  3. Transuranic waste management program waste form development

    Bennett, W.S.; Crisler, L.R.

    1981-01-01

    To ensure that all technology necessary for long term management of transuranic (TRU) wastes is available, the Department of Energy has established the Transuranic Waste Management Program. A principal focus of the program is development of waste forms that can accommodate the very diverse TRU waste inventory and meet geologic isolation criteria. The TRU Program is following two approaches. First, decontamination processes are being developed to allow removal of sufficient surface contamination to permit management of some of the waste as low level waste. The other approach is to develop processes which will allow immobilization by encapsulation of the solids or incorporate head end processes which will make the solids compatible with more typical waste form processes. The assessment of available data indicates that dewatered concretes, synthetic basalts, and borosilicate glass waste forms appear to be viable candidates for immobilization of large fractions of the TRU waste inventory in a geologic repository

  4. Alkali-slag cements for the immobilization of radioactive wastes

    Shi, C.; Day, R.L.

    1996-01-01

    Alkali-slag cements consist of glassy slag and an alkaline activator and can show both higher early and later strengths than Type III Portland cement, if a proper alkaline activator is used. An examination of microstructure of hardened alkali-slag cement pastes with the help of XRD and SEM with EDAX shows that the main hydration product is C-S-H (B) with low C/S ratio and no crystalline substances exist such as Ca(OH) 2 , Al (OH) 3 and sulphoaluminates. Mercury intrusion tests indicate that hardened alkali-slag cement pastes have a lower porosity than ordinary Portland cement, and contain mainly gel pores. The fine pore structure of hardened alkali-slag cement pastes will restrict the ingress of deleterious substances and the leaching of harmful species such as radionuclides. The leachability of Cs + from hardened alkali-slag cement pastes is only half of that from hardened Portland cement. From all these aspects, it is concluded that alkali-slag cements are a better solidification matrix than Portland cement for radioactive wastes

  5. Design and application of environmentally effective concrete with usage of chrysotile-cement waste

    Egorova Lada; Semenov Vyacheslav; Pligina Anna; Askhadullin Aizat

    2016-01-01

    Construction is resource-demanding industry, characterized by a large volume of waste. Particularly chrysotile cement waste obtained both in production and in dismantling over age chrysotile-cement products: corrugated asbestos boards and flat sheets, chrysotile-cement tubes. We propose to use dry chrysotile-cement waste as recycled aggregate for concrete. Based on developed compositions and identified properties of heavy concrete with chrysotile-cement waste introduce this technology to the ...

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

    Naish, C.C.; Buttle, D.; Wallace-Sims, R.; O'Brien, T.M.

    1991-01-01

    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)

  7. Investigations on cement/polymer Waste packages containing intermediate level waste and organic exchange resins

    ELsourougy, M R; Zaki, A A; Aly, H F [Atomic energy authority, hot laboratory center, Cairo, (Egypt); Khalil, M Y [Nuclear engineering department, Alexandria university. Alexandria, (Egypt)

    1995-10-01

    Polymers can be added to cements to improve its nuclear waste immobilization properties. This trend in cementation processes is attracting attention and requiring through investigations. In this work, polymers of different kinds were added to ordinary portland cement for the purpose of solidifying intermediate level liquid wastes and organic ion exchange resins. Epoxy polymer such as Kemapoxy-150 reduced the leaching rate of cesium compared to cement alone. Latex to cement ratio less than 4% caused an increase in leaching rate of cesium. When cesium was absorbed to an organic resin its leachability was improved. 5 figs., 4 tabs.

  8. Investigations on cement/polymer Waste packages containing intermediate level waste and organic exchange resins

    ELsourougy, M.R.; Zaki, A.A.; Aly, H.F.; Khalil, M.Y.

    1995-01-01

    Polymers can be added to cements to improve its nuclear waste immobilization properties. This trend in cementation processes is attracting attention and requiring through investigations. In this work, polymers of different kinds were added to ordinary portland cement for the purpose of solidifying intermediate level liquid wastes and organic ion exchange resins. Epoxy polymer such as Kemapoxy-150 reduced the leaching rate of cesium compared to cement alone. Latex to cement ratio less than 4% caused an increase in leaching rate of cesium. When cesium was absorbed to an organic resin its leachability was improved. 5 figs., 4 tabs

  9. Leaching studies of low-level radioactive waste forms

    Dayal, R.; Arora, H.; Milian, L.; Clinton, J.

    1985-01-01

    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

  10. Device for cementing radioactive of toxic waste into barrels

    Hempelmann, W.; Waldenmeier, G.; Mathis, P.; Mathis, B.; Mathis, F.

    1987-01-01

    The plant consists of conventional means, such as through mixers and dosing and transport spirals, which are accommodated in a glovebox. The inactive additives cement and sand and the active materials evaporation concentrates and sludges are mixed by them, and are then filled into a waste barrel which is empty or already filled with solid waste. Liquid radioactive wastes are used to concrete over the solid waste. (DG) [de

  11. Compatibility of cement with low and medium level wastes generated at nuclear power plants

    De Angelis, G.

    1987-01-01

    The incorporation of low- and medium level wastes in cement can give rise to serious problems due to the interaction between radwastes and cement components. Some troubles can occur during the first stages of the cementation process while deleterious phenomena come into evidence in the medium and long term. With the aim to improve the quality of the final waste forms a pretreatment of the waste streams is often necessary, as well as the addition of special agents to the cement matrix. The prediction of the long-term behaviour of the final products is of main importance in this research field. To this purpose some accelerated tests have been proposed, but much effort should be further devoted to the comprehension of the waste/matrix interaction. In this paper the principal waste streams are reviewed together with the possible remedies which seem to be more suitable for their cementation. Thus the qpportunity of separatly handling and solidifying different types of waste is once more supported

  12. Characterization of low and medium level radioactive waste forms

    Sambell, R.A.J.

    1983-01-01

    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

  13. Corrosion of metal containers containing cemented radioactive wastes

    Duffo, G.S.; Farina, S.B.; Schulz, F.M.; Marotta, F

    2010-01-01

    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

  14. Heat of hydration measurements on cemented radioactive wastes. Part 1: cement-water pastes

    Lee, D.J.

    1983-12-01

    This report describes the hydration of cement pastes in terms of chemical and kinetic models. A calorimetric technique was used to measure the heat of hydration to develop these models. The effects of temperature, water/cement ratio and cement replacements, ground granulated blast furnace slag (BFS) and pulverised fuel ash (PFA) on the hydration of ordinary Portland cement (OPC) is reported. The incorporation of BFS or PFA has a marked effect on the hydration reaction. The effect of temperature is also important but changing the water/cement ratio has little effect. Results from cement pastes containing only water and cement yield total heats of reaction of 400, 200 and 100 kJ/kg for OPC, BFS and PFA respectively. Using the results from the models which have been developed, the effect of major salts present in radioactive waste streams can be assessed. Values of the total heat of reaction, the time to complete 50 percent reaction, and the energy of activation, can be compared for different waste systems. (U.K.)

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

    Rogers, R.D.; McConnell, J.W. Jr.

    1988-06-01

    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

  16. Solid recovered fuels in the cement industry with special respect to hazardous waste.

    Thomanetz, Erwin

    2012-04-01

    Cements with good technical properties have been produced in Europe since the nineteenth century and are now worldwide standardized high-quality mass products with enormous production numbers. The basic component for cement is the so-called clinker which is produced mainly from raw meal (limestone plus clay plus sands) in a rotary kiln with preheater and progressively with integrated calciner, at temperatures up to 1450 °C. This process requires large amounts of fossil fuels and is CO₂-intensive. But most CO₂ is released by lime decomposition during the burning process. In the 1980s the use of alternative fuels began--firstly in the form of used oil and waste tyres and then increasingly by pre-conditioned materials from commercial waste and from high calorific industrial waste (i.e. solid recovered fuel (SRF))--as well as organic hazardous waste materials such as solvents, pre-conditioned with sawdust. Therefore the cement industry is more and more a competitor in the waste-to-energy market--be it for municipal waste or for hazardous waste, especially concerning waste incineration, but also for other co-incineration plants. There are still no binding EU rules identifying which types of SRF or hazardous waste could be incinerated in cement kilns, but there are some well-made country-specific 'positive lists', for example in Switzerland and Austria. Thus, for proper planning in the cement industry as well as in the waste management field, waste disposal routes should be considered properly, in order to avoid surplus capacities on one side and shortage on the other.

  17. Mathematical modelling of transport phenomena in radioactive waste-cement-bentonite matrix

    Plecas, Ilija; Dimovic, Slavko

    2010-01-01

    Document available in extended abstract form only. The leaching rate of 137 Cs from spent mix bead (anion and cation) exchange resins in a cement-bentonite matrix has been studied. Transport phenomena involved in the leaching of a radioactive material from a cement-bentonite matrix are investigated using three methods based on theoretical equations. These are: the diffusion equation for a plane source an equation for diffusion coupled to a first-order equation and an empirical method employing a polynomial equation. The results presented in this paper are from a 25-year mortar and concrete testing project that will influence the design choices for radioactive waste packaging for a future Serbian radioactive waste disposal center. Radioactive waste is waste material containing radioactive chemical elements which does not have a practical purpose. It is often the product of a nuclear process, such as nuclear fission. Waste can also be generated from the processing of fuel for nuclear reactors or nuclear weapons. The main objective in managing and disposing of radioactive (or other) waste is to protect people and the environment. This means isolating or diluting the waste so that the rate or concentration of any radionuclides returned to the biosphere is harmless. Storage as the placement of waste in a nuclear facility where isolation, environmental protection and human control are provided with the intent that the waste will be retrieved at a later time. Disposal as the emplacement of waste in an approved, specified facility (e.g. near surface or geological repository) without the intention of retrieval. The processing of radioactive wastes may be done for economic reasons (e.g. to reduce the volume for storage or disposal, or to recover a 'resource' from the waste), or safety reasons (e.g. converting the waste to a more 'stable' form, such as one that will contain the radionuclide inventory for a long time). Typically processing involves reducing

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

    Naish, C.C.; Buttle, D.; Wallace-Sims, R.; O'Brien, T.M.

    1991-01-01

    This report describes work to investigate acoustic emission as a non-intrusive monitor of corrosion and degradation of cemented wasteforms where the waste is a potentially reactive metal. The acoustic data collected shows good correlation with the corrosion rate as measured by hydrogen gas evolution rates and the electrochemically measured corrosion rates post cement hardening. The technique has been shown to be sensitive in detecting stress caused by expansive corrosion product within the cemented wasteform. The attenuation of the acoustic signal by the wasteform reduced the signal received by the monitoring equipment by a factor of 10 over a distance of approximately 150-400 mm, dependent on the water level in the cement. Full size packages were successfully monitored. It is concluded that the technique offers good potential for monitoring cemented containers of the more reactive metals, for example Magnox and aluminium. (author)

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

    Mota Vieira, V.; Oliveira, C.C. de

    2015-01-01

    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)

  20. High-level radioactive waste incorporation into (special) cements

    Roy, D.M.; Gouda, G.R.

    1978-01-01

    A feasibility study has demonstrated that very strong, durable, relatively impermeable cylinders may be prepared by hot pressing combinations of cements with simulated radioactive waste solids. While the properties have not been studied exhaustively, the results suggest an optional method for immobilization and isolation of radioactive waste. Samples prepared with calcium aluminate cements appeared to have properties superior to those with Portland cements. Four simulated radioactive waste compositions having high rare-earth oxide contents, and some containing a large excess of NaNO 3 , were studied. Modest temperatures [423 to 673 K (150 to 400 0 C)] were used for hot pressing at pressures from 178 to 345 MPa. Dense strong very low porosity specimens resulted when mixtures containing from 10 to 50% waste were hot pressed, incorporating also a small percentage of water. In addition, high-strength cement cylinders were prepared with the waste solid (approximately 20 wt% waste) in a separate core and were very resistant to leaching by water near its boiling point. With this configuration, even the NaNO 3 -containing wastes were resistant to leaching by water

  1. Radioactive waste-Portland cement systems: I, radionuclide distribution

    Jantzen, C.M.; Glasser, F.P.; Lachowski, E.E.

    1984-01-01

    Crystal chemical stabilization of radioactive wastes can be achieved during clinkering of, or with, ordinary portland cement. Waste loadings of 20 to 30 wt% are achieved by dilute solid solution of waste ions into cementitious host lattices. Higher waste loadings result in compatible noncementitious radiophases. The cementitious phases hydrate without loss of compressive strength. Crystallochemical relationships predict that the radionuclide partitioning in the anhydrous clinkered phases will be maintained in the hydration products. These cementitious hydroxylated radiophases would be in internal equilibrium under anticipated repository conditions. The radionuclide distributions observed are described in the context of established phase equilibria for commercial waste cement systems, but are applicable to transuranic, medium- and low-level wastes

  2. Utilization of waste glass in ECO-cement: Strength properties and microstructural observations

    Sobolev, Konstantin; Tuerker, Pelin; Soboleva, Svetlana; Iscioglu, Gunsel

    2007-01-01

    Waste glass creates a serious environmental problem, mainly because of the inconsistency of the waste glass streams. The use of waste glass as a finely ground mineral additive (FGMA) in cement is a promising direction for recycling. Based on the method of mechano-chemical activation, a new group of ECO-cements was developed. In ECO-cement, relatively large amounts (up to 70%) of portland cement clinker can be replaced with waste glass. This report examines the effect of waste glass on the microstructure and strength of ECO-cement based materials. Scanning electron microscopy (SEM) investigations were used to observe the changes in the cement hydrates and interface between the cement matrix and waste glass particles. According to the research results, the developed ECO-cement with 50% of waste glass possessed compressive strength properties at a level similar to normal portland cement

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

    Mallaupoma, M.; Soriano, A.; Rodriguez, G.; Cruz, W.

    1996-01-01

    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

  4. Immobilization of radioactive waste in cement-based matrices

    Glasser, F.P.; Rahman, A.A.; Macphee, D.; McCulloch, C.E.; Angus, M.J.

    1984-01-01

    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-SiO 2 -H 2 O 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 CO 2 -containing environments. (author)

  5. Employing 3R Techniques in Managing Cement Industry Waste

    Lamyaa Mohammed Dawood

    2018-01-01

    Full Text Available Waste management conserves human health, ownership, environment, and keeps valuable natural resources. Lean-green waste of an organization’s operations can be decreased through implementation 3R (Reduce, Reuse, and Recycling techniques by reduction of manufacturing system wastes. This research aims to integrate lean-green waste of the manufacturing system throughout employing 3R techniques and weighted properties method in order to manage waste. Al-Kufa cement plant is employed as a case study. Results are generated using Edraw Max Version 7 and Excel. Overall results show reduce technique of lean-green waste management has major contribution of 55 % and recycling technique has minor contribution 18 %. Defects waste has major integration of lean-green waste, while air emissions waste has minor integration of lean-green waste.

  6. The thermoelectric generators use for waste heat utilization from cement plant

    Sztekler Karol

    2017-01-01

    Production often entails the formation of by-product which is waste heat. One of the equipment processing heat into electricity is a thermoelectric generator. Its operation is based on the principle of thermoelectric phenomenon, which is known as a Seebeck phenomenon. The simplicity of thermoelectric phenomena allows its use in various industries, in which the main waste product is in the form of heat with the temperature of several hundred degrees. The study analyses the possibility of the thermoelectric systems use for the waste heat utilization resulting in the cement production at the cement plant. The location and design of the thermoelectric system that could be implemented in cement plant is chosen. The analysis has been prepared in the IPSEpro software.

  7. Utilization of flotation wastes of copper slag as raw material in cement production

    Alp, I.; Deveci, H.; Suenguen, H.

    2008-01-01

    Copper slag wastes, even if treated via processes such as flotation for metal recovery, still contain heavy metals with hazardous properties posing environmental risks for disposal. This study reports the potential use of flotation waste of a copper slag (FWCS) as iron source in the production of Portland cement clinker. The FWCS appears a suitable raw material as iron source containing >59% Fe 2 O 3 mainly in the form of fayalite (Fe 2 SiO 4 ) and magnetite (Fe 3 O 4 ). The clinker products obtained using the FWCS from the industrial scale trial operations over a 4-month period were characterised for the conformity of its chemical composition and the physico-mechanical performance of the resultant cement products was evaluated. The data collected for the clinker products produced using an iron ore, which is currently used as the cement raw material were also included for comparison. The results have shown that the chemical compositions of all the clinker products including those of FWCS are typical of a Portland cement clinker. The mechanical performance of the standard mortars prepared from the FWCS clinkers were found to be similar to those from the iron ore clinkers with the desired specifications for the industrial cements e.g. CEM I type cements. Furthermore, the leachability tests (TCLP and SPLP) have revealed that the mortar samples obtained from the FWCS clinkers present no environmental problems while the FWCS could act as the potential source of heavy metal contamination. These findings suggest that flotation wastes of copper slag (FWCS) can be readily utilised as cement raw material due to its availability in large quantities at low cost with the further significant benefits for waste management/environmental practices of the FWCS and the reduced production and processing costs for cement raw materials

  8. Utilization of flotation wastes of copper slag as raw material in cement production.

    Alp, I; Deveci, H; Süngün, H

    2008-11-30

    Copper slag wastes, even if treated via processes such as flotation for metal recovery, still contain heavy metals with hazardous properties posing environmental risks for disposal. This study reports the potential use of flotation waste of a copper slag (FWCS) as iron source in the production of Portland cement clinker. The FWCS appears a suitable raw material as iron source containing >59% Fe(2)O(3) mainly in the form of fayalite (Fe(2)SiO(4)) and magnetite (Fe(3)O(4)). The clinker products obtained using the FWCS from the industrial scale trial operations over a 4-month period were characterised for the conformity of its chemical composition and the physico-mechanical performance of the resultant cement products was evaluated. The data collected for the clinker products produced using an iron ore, which is currently used as the cement raw material were also included for comparison. The results have shown that the chemical compositions of all the clinker products including those of FWCS are typical of a Portland cement clinker. The mechanical performance of the standard mortars prepared from the FWCS clinkers were found to be similar to those from the iron ore clinkers with the desired specifications for the industrial cements e.g. CEM I type cements. Furthermore, the leachability tests (TCLP and SPLP) have revealed that the mortar samples obtained from the FWCS clinkers present no environmental problems while the FWCS could act as the potential source of heavy metal contamination. These findings suggest that flotation wastes of copper slag (FWCS) can be readily utilised as cement raw material due to its availability in large quantities at low cost with the further significant benefits for waste management/environmental practices of the FWCS and the reduced production and processing costs for cement raw materials.

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

    Kienzler, Bernhard; Metz, Volker; Schlieker, Martina; Bohnert, Elke

    2015-01-01

    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) 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

  10. 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

    Wilding, C.R.; Phillips, D.C.; Burnay, S.G.; Spindler, W.E.; Lyon, C.E.; Winter, J.A.

    1991-01-01

    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 238 Pu. Irradiated materials included matrix materials, simulated waste forms and real waste forms. 2 figs.; 3 tabs.; 8 refs

  11. Treatability study of absorbent polymer waste form for mixed waste treatment

    Herrmann, S. D.; Lehto, M. A.; Stewart, N. A.; Croft, A. D.; Kern, P. W.

    2000-01-01

    A treatability study was performed to develop and characterize an absorbent polymer waste form for application to low level (LLW) and mixed low level (MLLW) aqueous wastes at Argonne National Laboratory-West (ANL-W). In this study absorbent polymers proved effective at immobilizing aqueous liquid wastes in order to meet Land Disposal Restrictions for subsurface waste disposal. Treatment of aqueous waste with absorbent polymers provides an alternative to liquid waste solidification via high-shear mixing with clays and cements. Significant advantages of absorbent polymer use over clays and cements include ease of operations and waste volume minimization. Absorbent polymers do not require high-shear mixing as do clays and cements. Granulated absorbent polymer is poured into aqueous solutions and forms a gel which passes the paint filter test as a non-liquid. Pouring versus mixing of a solidification agent not only eliminates the need for a mixing station, but also lessens exposure to personnel and the potential for spread of contamination from treatment of radioactive wastes. Waste minimization is achieved as significantly less mass addition and volume increase is required of and results from absorbent polymer use than that of clays and cements. Operational ease and waste minimization translate into overall cost savings for LLW and MLLW treatment

  12. Test plan for formulation and evaluation of grouted waste forms with shine process wastes

    Ebert, W. L. [Argonne National Lab. (ANL), Argonne, IL (United States); Jerden, J. L. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-09-01

    The objective of this experimental project is to demonstrate that waste streams generated during the production of Mo99 by the SHINE Medical Technologies (SHINE) process can be immobilized in cement-based grouted waste forms having physical, chemical, and radiological stabilities that meet regulatory requirements for handling, storage, transport, and disposal.

  13. Westinghouse integrated cementation facility. Smart process automation minimizing secondary waste

    Fehrmann, H.; Jacobs, T.; Aign, J.

    2015-01-01

    The Westinghouse Cementation Facility described in this paper is an example for a typical standardized turnkey project in the area of waste management. The facility is able to handle NPP waste such as evaporator concentrates, spent resins and filter cartridges. The facility scope covers all equipment required for a fully integrated system including all required auxiliary equipment for hydraulic, pneumatic and electric control system. The control system is based on actual PLC technology and the process is highly automated. The equipment is designed to be remotely operated, under radiation exposure conditions. 4 cementation facilities have been built for new CPR-1000 nuclear power stations in China

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

    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.)

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

    Hamlat, M.S.; Rabia, N.

    1998-01-01

    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 (α) were used. (D) is an effective diffusion coefficient that describes the kinetic behaviour and is most easily determined using Soxhlet test, whereas, (α) 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.)

  16. Cementation of biodegraded radioactive oils and organic waste

    Gorbunova, O.; Safonov, A.; Tregubova, V.; German, K.

    2015-01-01

    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 (CH 4 , H 2 O, CO 2 , N 2 ), 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 90 Sr, up to 20% of 137 Cs 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

  17. Sodium Sulphate Effect on Cement Produced with Building Stone Waste

    Emre Sancak

    2015-01-01

    Full Text Available In this study, the blended cements produced by using the building stone waste were exposed to sulphate solution and the cement properties were examined. Prepared mortar specimens were cured under water for 28 days and then they were exposed to three different proportions of sodium sulphate solution for 125 days. Performances of cements were determined by means of compressive strength and tensile strength tests. The broken parts of some mortar bars were examined with scanning electron microscope (SEM. Besides, they were left under moist atmosphere and their length change was measured and continuously monitored for period of 125 days. In blended cements, solely cements obtained by replacing 10–20% of diatomites gave similar strength values with ordinary Portland cement (CEM I 42.5R at the ages of 7, 28, and 56 days. In all mortar specimens that included either waste andesite (AP or marble powder (MP showed best performance against very severe effective sodium sulphate solutions (13500 mg/L.

  18. Sulfur polymer cement stabilization of elemental mercury mixed waste

    Melamed, D.; Fuhrmann, M.; Kalb, P.; Patel, B.

    1998-04-01

    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)

  19. 40 CFR 63.1220 - What are the replacement standards for hazardous waste burning cement kilns?

    2010-07-01

    ... hazardous waste burning cement kilns? 63.1220 Section 63.1220 Protection of Environment ENVIRONMENTAL... Waste Combustors Replacement Emissions Standards and Operating Limits for Incinerators, Cement Kilns... burning cement kilns? (a) Emission and hazardous waste feed limits for existing sources. You must not...

  20. Mixed Waste Focus Area - Waste form initiative

    Nakaoka, R.; Waters, R.; Pohl, P.; Roach, J.

    1998-01-01

    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

  1. Long-term interactions of full-scale cemented waste simulates with salt brines

    Kienzler, B.; Borkel, C.; Metz, V.; Schlieker, M.

    2016-07-01

    Since 1967 radioactive wastes have been disposed of in the Asse II salt mine in Northern Germany. A significant part of these wastes originated from the pilot reprocessing plant WAK in Karlsruhe and consisted of cemented NaNO{sub 3} solutions bearing fission products, actinides, as well as process chemicals. With respect to the long-term behavior of these wastes, the licensing authorities requested leaching experiments with full scale samples in relevant salt solutions which were performed since 1979. The experiments aimed at demonstrating the transferability of results obtained with laboratory samples to real waste forms and at the investigation of the effects of the industrial cementation process on the properties of the waste forms. This research program lasted until 2013. The corroding salt solutions were sampled several times and after termination of the experiments, the solid materials were analyzed by various methods. The results presented in this report cover the evolution of the solutions and the chemical and mineralogical characterization of the solids including radionuclides and waste components, and the paragenesis of solid phases (corrosion products). The outcome is compared to the results of model calculations. For safety analysis, conclusions are drawn on radionuclide retention, evolution of the geochemical environment, evolution of the density of solutions, and effects of temperature and porosity of the cement waste simulates on cesium mobilization.

  2. Long-term interactions of full-scale cemented waste simulates with salt brines

    Kienzler, B.; Borkel, C.; Metz, V.; Schlieker, M.

    2016-01-01

    Since 1967 radioactive wastes have been disposed of in the Asse II salt mine in Northern Germany. A significant part of these wastes originated from the pilot reprocessing plant WAK in Karlsruhe and consisted of cemented NaNO 3 solutions bearing fission products, actinides, as well as process chemicals. With respect to the long-term behavior of these wastes, the licensing authorities requested leaching experiments with full scale samples in relevant salt solutions which were performed since 1979. The experiments aimed at demonstrating the transferability of results obtained with laboratory samples to real waste forms and at the investigation of the effects of the industrial cementation process on the properties of the waste forms. This research program lasted until 2013. The corroding salt solutions were sampled several times and after termination of the experiments, the solid materials were analyzed by various methods. The results presented in this report cover the evolution of the solutions and the chemical and mineralogical characterization of the solids including radionuclides and waste components, and the paragenesis of solid phases (corrosion products). The outcome is compared to the results of model calculations. For safety analysis, conclusions are drawn on radionuclide retention, evolution of the geochemical environment, evolution of the density of solutions, and effects of temperature and porosity of the cement waste simulates on cesium mobilization.

  3. Statistical study of chemical additives effects in the waste cementation

    Tello, Cledola C.O. de; Diniz, Paula S.; Haucz, Maria J.A.

    1997-01-01

    This paper presents the statistical study, that was carried out to analyse the chemical additives effect in the waste cementation process. Three different additives from two industries were tested: set accelerator, set retarder and super plasticizers, in cemented pates with and without bentonite. The experiments were planned in accordance with the 2 3 factorial design, so that the effect of each type of additive, its quantity and manufacturer in cemented paste and specimens could be evaluated. The results showed that the use of these can improve the cementation process and the product. The admixture quantity and the association with bentonite were the most important factors affecting the process and product characteristics. (author). 4 refs., 9 figs., 4 tabs

  4. Processes and Equipment for the Cementation of Radioactive Waste

    Schaefer, S.; Studenski, J.

    2012-01-01

    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.)

  5. Using of borosilicate glass waste as a cement additive

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

    2016-01-01

    Highlights: • Borosilicate glass waste used as cement additive can improves its radiation shielding. • When content is 14.8%, the linear attenuation coefficient is 0.2457 cm"−"1 after 28 d. • From 0 to 22.2%, linear attenuation coefficient firstly increase and then decrease. - Abstract: 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.

  6. Using of borosilicate glass waste as a cement additive

    Han, Weiwei [State Key Laboratory of Silicate Materials for Architectures, Wuhan University of Technology, Wuhan, Hubei 430070 (China); School of Civil Engineering and Architecture, Wuhan University of Technology, Wuhan, Hubei 430070 (China); Sun, Tao, E-mail: sunt@whut.edu.cn [State Key Laboratory of Silicate Materials for Architectures, Wuhan University of Technology, Wuhan, Hubei 430070 (China); Key Laboratory of Roadway Bridge & Structure Engineering, Wuhan University of Technology, Wuhan, Hubei 430070 (China); Li, Xinping [Key Laboratory of Roadway Bridge & Structure Engineering, Wuhan University of Technology, Wuhan, Hubei 430070 (China); Sun, Mian [School of Materials Science and Engineering, Wuhan University of Technology, Wuhan, Hubei 430070 (China); Lu, Yani [Urban Construction Institute, Hubei Engineering University, Xiaogan, Hubei 432000 (China)

    2016-08-15

    Highlights: • Borosilicate glass waste used as cement additive can improves its radiation shielding. • When content is 14.8%, the linear attenuation coefficient is 0.2457 cm{sup −1} after 28 d. • From 0 to 22.2%, linear attenuation coefficient firstly increase and then decrease. - Abstract: 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{sup −1} after 28 d, which is beneficial for the preparation of radiation shielding concrete with high performances.

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

    Tanaka, Satoru; Nagasaki, Shinya; Ohe, Toshiaki

    2000-01-01

    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.)

  8. The effects of large scale processing on caesium leaching from cemented simulant sodium nitrate waste

    Lee, D.J.; Brown, D.J.

    1982-01-01

    The effects of large scale processing on the properties of cemented simulant sodium nitrate waste have been investigated. Leach tests have been performed on full-size drums, cores and laboratory samples of cement formulations containing Ordinary Portland Cement (OPC), Sulphate Resisting Portland Cement (SRPC) and a blended cement (90% ground granulated blast furnace slag/10% OPC). In addition, development of the cement hydration exotherms with time and the temperature distribution in 220 dm 3 samples have been followed. (author)

  9. Leaching behavior of phosphate-bonded ceramic waste forms

    Singh, D.; Wagh, A.S.; Jeong, S.Y.; Dorf, M.

    1996-04-01

    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. We have developed a magnesium phosphate ceramic 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

  10. Immobilisation of shredded soft waste in cement monolith

    Brown, D.J.; Dalton, M.J.; Smith, D.L.

    1983-04-01

    A grouting process for the immobilisation of shredded contaminated laboratory waste in a cement monolith is being developed at the Atomic Energy Establishment Winfrith. The objective is to produce a 'monolithic' type package which is acceptable both for sea and land disposal. The work carried out on this project in the period April 1982 - March 1983 is summarised in this report. (author)

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

    Dufournet, F.

    1987-01-01

    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/MgCl 2 . Then behavior with water is investigated: swelling, cracking and leaching [fr

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

    Palmer, J.D.; Smith, D.L.G.

    1986-01-01

    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

  13. Effect of Cement Replacement with Carbide Waste on the Strength of Stabilized Clay Subgrade

    Muntohar A.S.

    2016-03-01

    Full Text Available Cement is commonly used for soil stabilization and many other ground improvement techniques. Cement is believed to be very good to improve the compressive and split-tensile strength of clay subgrades. In some application cement could be partly or fully replaced with carbide waste. This research is to study the effectiveness of the cement replacement and to find the maximum carbide waste content to be allowed for a clay subgrade. The quantities of cement replaced with the carbide waste were 30, 50, 70, 90, and 100% by its mass. The results show that replacing the cement with carbide waste decreased both the compressive and split tensile strength. Replacing cement content with carbide waste reduced its ability for stabilization. The carbide waste content should be less than 70% of the cement to provide a sufficient stabilizing effect on a clay subgrade.

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

    Lee, D.J.; Brown, D.J.

    1981-08-01

    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.)

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

    Palmer, J.D.; Smith, D.L.

    1985-07-01

    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.)

  16. Cement-Polymer Composite Containers for Radioactive Wastes Disposal

    Ghattas, N.K.; Eskander, S.B.; Bayoumi, T.A.; Saleh, H.M.

    2009-01-01

    Improving cement-composite containers using polymer as organic additives was studied extensively. Both unsaturated styrenated polyester (SPE) and polymethyl methacrylate (PMMA) were used to fill the pores in cement containers that used for disposal of radioactive wastes. Two different techniques were adopted for the addition of organic polymers based on their viscosity. The low density PMMA was added using impregnation technique. On the other hand high density SPE was mixed with cement paste as a premix process. Predetermined weight of dried borate radioactive powder waste simulate was introduced into the Cement-polymer composite (CPC) container and then closed before subjecting it to leaching characterization. The effect of the organic polymers on the hydration of cement matrix and on the properties of the obtained CPC container has been studied using X-ray diffraction, IR-analysis, thermal effects and weight loss. Porosity, pore parameters and rate of release were also determined. The results obtained showed that for the candidate CPC container positive effect of polymer dominates and an improvement in the retardation rate of PMMA release radionuclides was observed

  17. Waste forms for plutonium disposition

    Johnson, S.G.; O'Holleran, T.P.; Frank, S.M.; Meyer, M.K.; Hanson, M.; Staples, B.A.; Knecht, D.A.; Kong, P.C.

    1997-01-01

    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

  18. Polymer-Cement Composites Containing Waste Perlite Powder

    Paweł Łukowski

    2016-10-01

    Full Text Available Polymer-cement composites (PCCs are materials in which the polymer and mineral binder create an interpenetrating network and co-operate, significantly improving the performance of the material. On the other hand, the need for the utilization of waste materials is a demand of sustainable construction. Various mineral powders, such as fly ash or blast-furnace slag, are successfully used for the production of cement and concrete. This paper deals with the use of perlite powder, which is a burdensome waste from the process of thermal expansion of the raw perlite, as a component of PCCs. The results of the testing of the mechanical properties of the composite and some microscopic observations are presented, indicating that there is a possibility to rationally and efficiently utilize waste perlite powder as a component of the PCC. This would lead to creating a new type of building material that successfully meets the requirements of sustainable construction.

  19. Immobilisation of MTR waste in cement (product evaluation)

    Howard, C.G.; Lee, D.J.

    1988-01-01

    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. 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. (author)

  20. Feasibility of using ceramic furnace wastes in cement composites

    Fazzan, J.V.; Sanches, A.O.; Akasaki, J.L.; Malmonge, J.A.

    2016-01-01

    Currently, the region of Epitacio-SP President is classified as Paulista West Center in the production of ceramic tiles and bricks. However, as these industries have also generated environmental impacts in the production process with the generation of waste, the construction industries presents as great potential to absorb a large portion of these materials, called Pozzolans. In this sense, the research aims to study the characterization of Ceramic Furnace Wastes (CFC) and the evaluation of their reactivity. Mortar specimens were molded with different waste percentages in partial replacement of Portland cement, for analysis of compressive strength and capillary water absorption test. The characterization results show that important properties can be obtained by the preparation conditions of ashes, besides obtaining resistant activity index higher than expected by technical standards when using the material in replacement of Portland cement. (author)

  1. Standardized waste form test methods

    Slate, S.C.

    1984-11-01

    The Materials Characterization Center (MCC) is developing standard tests to characterize nuclear waste forms. Development of the first thirteen tests was originally initiated to provide data to compare different high-level waste (HLW) forms and to characterize their basic performance. The current status of the first thirteen MCC tests and some sample test results is presented: The radiation stability tests (MCC-6 and 12) and the tensile-strength test (MCC-11) are approved; the static leach tests (MCC-1, 2, and 3) are being reviewed for full approval; the thermal stability (MCC-7) and microstructure evaluation (MCC-13) methods are being considered for the first time; and the flowing leach tests methods (MCC-4 and 5), the gas generation methods (MCC-8 and 9), and the brittle fracture method (MCC-10) are indefinitely delayed. Sample static leach test data on the ARM-1 approved reference material are presented. Established tests and proposed new tests will be used to meet new testing needs. For waste form production, tests on stability and composition measurement are needed to provide data to ensure waste form quality. In transportation, data are needed to evaluate the effects of accidents on canisterized waste forms. The new MCC-15 accident test method and some data are presented. Compliance testing needs required by the recent draft repository waste acceptance specifications are described. These specifications will control waste form contents, processing, and performance. 2 references, 2 figures

  2. Standardized waste form test methods

    Slate, S.C.

    1984-01-01

    The Materials Characterization Center (MCC) is developing standard tests to characterize nuclear waste forms. Development of the first thirteen tests was originally initiated to provide data to compare different high-level waste (HLW) forms and to characterize their basic performance. The current status of the first thirteen MCC tests and some sample test results are presented: the radiation stability tests (MCC-6 and 12) and the tensile-strength test (MCC-11) are approved; the static leach tests (MCC-1, 2, and 3) are being reviewed for full approval; the thermal stability (MCC-7) and microstructure evaluation (MCC-13) methods are being considered for the first time; and the flowing leach test methods (MCC-4 and 5), the gas generation methods (MCC-8 and 9), and the brittle fracture method (MCC-10) are indefinitely delayed. Sample static leach test data on the ARM-1 approved reference material are presented. Established tests and proposed new tests will be used to meet new testing needs. For waste form production, tests on stability and composition measurement are needed to provide data to ensure waste form quality. In transporation, data are needed to evaluate the effects of accidents on canisterized waste forms. The new MCC-15 accident test method and some data are presented. Compliance testing needs required by the recent draft repository waste acceptance specifications are described. These specifications will control waste form contents, processing, and performance

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

    Kalb, P.D.; Heiser, J.H. III; Colombo, P.

    1991-07-01

    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

  4. Laboratory procedures for waste form testing

    Mast, E.S.

    1994-01-01

    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

  5. Laboratory procedures for waste form testing

    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.

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

    Brodersen, K.; Vinther, A.

    1990-11-01

    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

  7. Industrial Wastes as Auxiliary Additives to Cement/Lime Stabilization of Soils

    Jijo James

    2016-01-01

    Full Text Available Chemical stabilization involves the use of chemical agents for initiating reactions within the soil for modification of its geotechnical properties. Cement and lime stabilization have been the most common stabilization methods adopted for soil treatment. Cement stabilization results in good compressive strengths and is preferred for cohesionless to moderately cohesive soil but loses effectiveness when the soil is highly plastic. Lime stabilization is the most preferred method for plastic clays; however, it proves to be ineffective in sulphate rich clays and performs poorly under extreme conditions. With such drawbacks, lots of researches have been undertaken to address the issues faced with each stabilization method, in particular, the use of solid wastes for soil stabilization. Solid waste reuse has gained high momentum for achieving sustainable waste management in recent times. Research has shown that the use of solid wastes as additives with and replacement for conventional stabilizers has resulted in better results than the performance of either individually. This review provides insight into some of the works done by earlier researchers on lime/cement stabilization with industrial wastes as additives and helps to form a sound platform for further research on industrial wastes as additives to conventional stabilizers.

  8. Immobilization of low and intermediate level of organic radioactive wastes in cement matrices

    Eskander, S.B. [Radioisotopes Department, Atomic Energy Authority, Dokki, Cairo (Egypt); Abdel Aziz, S.M. [Middle Eastern Regional Radioisotope Centre for the Arab Countries, Dokki, Cairo (Egypt); El-Didamony, H. [Faculty of Science, Zagazig University, Zagazig, El-Sharkia (Egypt); Sayed, M.I., E-mail: mois_161272@yahoo.com [Middle Eastern Regional Radioisotope Centre for the Arab Countries, Dokki, Cairo (Egypt)

    2011-06-15

    Highlights: {yields} Solidification/stabilization of liquid scintillation waste. {yields} Resistance to frost attack. {yields} Retarding effect of scintillator waste was overcome by adding clay. {yields} Evaluation of the suitability of cement-clay composite to solidify and stabilize scintillation waste. - Abstract: The adequacy of cement-clay composite, for solidification/stabilization of organic radioactive spent liquid scintillator wastes and its resistance to frost attack were determined by a freezing/thawing (F/T) test. Frost resistance is assessed for the candidate cement-clay composite after 75 cycles of freezing and thawing by evaluating their mass durability index, compressive strength, apparent porosity, volume of open pores, water absorption, and bulk density. Infrared (IR), X-ray diffraction (XRD), differential thermal analysis (DTA), thermal gravimetric analysis (TGA) and scanning electron microscopy (SEM) were performed for the final waste form (FWF) before and after the F/T treatment to follow the changes that may take place in its microstructure during the hydration regime. The results were obtained indicate that the candidate composite exhibits acceptable resistance to freeze/thaw treatment and has adequate suitability to solidify and stabilize organic radioactive spent liquid scintillator wastes even at very exaggerating conditions (-50{sup Degree-Sign }C and +60{sup Degree-Sign }C ).

  9. Immobilization of low and intermediate level of organic radioactive wastes in cement matrices

    Eskander, S.B.; Abdel Aziz, S.M.; El-Didamony, H.; Sayed, M.I.

    2011-01-01

    Highlights: → Solidification/stabilization of liquid scintillation waste. → Resistance to frost attack. → Retarding effect of scintillator waste was overcome by adding clay. → Evaluation of the suitability of cement-clay composite to solidify and stabilize scintillation waste. - Abstract: The adequacy of cement-clay composite, for solidification/stabilization of organic radioactive spent liquid scintillator wastes and its resistance to frost attack were determined by a freezing/thawing (F/T) test. Frost resistance is assessed for the candidate cement-clay composite after 75 cycles of freezing and thawing by evaluating their mass durability index, compressive strength, apparent porosity, volume of open pores, water absorption, and bulk density. Infrared (IR), X-ray diffraction (XRD), differential thermal analysis (DTA), thermal gravimetric analysis (TGA) and scanning electron microscopy (SEM) were performed for the final waste form (FWF) before and after the F/T treatment to follow the changes that may take place in its microstructure during the hydration regime. The results were obtained indicate that the candidate composite exhibits acceptable resistance to freeze/thaw treatment and has adequate suitability to solidify and stabilize organic radioactive spent liquid scintillator wastes even at very exaggerating conditions (-50 ° C and +60 ° C ).

  10. Differential thermal, Thermogravimetric and X-ray diffraction investigation of hydration phases in cementitious waste form

    Khalil, M.Y.; Nagy, M.E.; El-Sourougy, M.R.; Zaki, A.A.

    1996-01-01

    Hydration phases of cement determine the final properties of the product. Adding other components to the cement paste may alter the final phases formed and affect properties of the hardened products. In this work ordinary portland cement and/or blast furnace slag cement were hardened with low-or intermediate-level radioactive liquid wastes and different additives. Hydration phases were investigated using differential thermal, thermogravimetric, and X-ray diffraction techniques. Low-and intermediate-level liquid wastes were found not to affect the hydration phases of cement. The addition of inorganic exchangers and latex were found to affect the hydration properties of the cement waste system. This resulted in a reduction of compressive strength. On the contrary, addition of epoxy also affected the hydration causing increase in compressive strength. 10 figs., 2 tabs

  11. Diffusion from cylindrical waste forms

    Thomas, G.F.

    1985-05-01

    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

  12. Geochemical analysis of leachates from cement/low-level radioactive waste/soil systems

    Criscenti, L.J.; Serne, R.J.

    1988-09-01

    Laboratory experiments were conducted as part of the Special Waste Form Lysimeters/endash/Arid Program. These experiments were conducted to investigate the performance of solidified low-level nuclear waste in a typical arid, near-surface disposal site, and to evaluate the ability of laboratory tests to predict leaching in actual field conditions. Batch leaching, soil adsorption column, and soil/waste form column experiments were conducted using Portland III cement waste forms containing boiling-water reactor evaporator concentrate and ion-exchange resin waste. In order to understand the reaction chemistry of the cement waste form/soil/ground-water system, the compositions of the leachates from the laboratory experiments were studied with the aid of the MINTEQ ion speciation/solubility and mass transfer computer code. The purpose of this report is to describe the changes in leachate composition that occur during the course of the experiments, to discuss the geochemical modeling results, and to explore the factors controlling the major element chemistry of these leachates. 18 refs., 84 figs., 14 tabs

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

    Quillin, K.; Duerden, S.L.; Majumdar, A.J.

    1993-01-01

    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 90 o C 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)

  14. Characteristics of waste forms improved by using admixtures

    Rzyski, B.M.; Suarez, A.A.

    1989-06-01

    The immobilization of nitric waste streams with ordinary Portland cement can be improved by use of some admixtures. The aim of this work was to investigated how the main characteristics of waste forms prepared with Portalnd cement pastes are modified by the addition of sulphonic naphtalene acids, lignosulphonic acids and emulsified fatty acids, which are present in some commercial admixtures. The effectiveness of the admixture in reducing the pore volume, as well as improving other parameters, depends on its chemical composition and on the amount utilized as well as the water to cement ratio and salt content. The admixture which has emulsified fatty acids in its composition shows some adverse results when the samples are immersed in water. The mechanical strenght however is some what increased even when water load is increased. (author) [pt

  15. Characteristics of waste forms improved by using admixtures

    Rzyski, B.M.; Suarez, A.A.

    1989-01-01

    The immobilization of nitric waste streams with ordinary Portland cement can be improved by use of some admixtures. The aim of this work was to investigate how the main characteristics of waste forms prepared with Portland cement pastes are modified by the addition of sulphonic naphyhalene acids, lignosulphonic acids and emulsified fatty acids, which are present in some commercial admixtures. The effectiveness of the admixtures in reducing the pore volume, as well as improving other parameters, depends on its chemical composition and on the amount utilized as well as the water to cement ratio and salt content. The admixture which has emulsified fatty acids in its composition shows some adverse results when the samples are immersed in water. The mechanical strength however is somewhat increased even when water load is increased

  16. Development programs in the United States of America for the application of cement-based grouts in radioactive waste management

    Dole, L.R.; Row, T.H.

    1984-01-01

    This paper briefly reviews seven cement-based waste form development programs at six of the US Department of Energy (DOE) sites. These sites have developed a variety of processes that range from producing 25 mm (1 in.) diameter pellets in a glove box to producing 240 m (800 ft.) diameter grout sheets within the bedding planes of a deep shale formation. These successful applications of cement-based waste forms to the many radioactive waste streams from nuclear facilities bear witness to the flexibility and reliability of this class of materials. This paper also discusses the major issues regarding the application of cement-based waste forms to radioactive waste management problems. These issues are (1) leachability, (2) radiation stability, (3) thermal stability, (4) phase complexity of the matrix, and (5) effects of the waste stream composition. A cursory review of current research in each of these areas is given This paper also discusses future trends in cement-based waste form development and applications. 31 references, 11 figures.

  17. Development programs in the United States of America for the application of cement-based grouts in radioactive waste management

    Dole, L.R.; Row, T.H.

    1984-01-01

    This paper briefly reviews seven cement-based waste form development programs at six of the US Department of Energy (DOE) sites. These sites have developed a variety of processes that range from producing 25 mm (1 in.) diameter pellets in a glove box to producing 240 m (800 ft.) diameter grout sheets within the bedding planes of a deep shale formation. These successful applications of cement-based waste forms to the many radioactive waste streams from nuclear facilities bear witness to the flexibility and reliability of this class of materials. This paper also discusses the major issues regarding the application of cement-based waste forms to radioactive waste management problems. These issues are (1) leachability, (2) radiation stability, (3) thermal stability, (4) phase complexity of the matrix, and (5) effects of the waste stream composition. A cursory review of current research in each of these areas is given This paper also discusses future trends in cement-based waste form development and applications. 31 references, 11 figures

  18. Determination of quality standard of cemented radioactive wastes

    Johan, B.

    1997-01-01

    The standard determination of cemented low activity radioactive waste has been carried out. This study was intended to find the quality of immobilized product as a guaranty to the public and the environmental safety. As preliminary study, there parameters type have been studied i.e, the density, compressive strength and the leaching rate. By adopted the formulations of advanced country and field data, it can be concluded that the cementation waste should have the density, p, between 1.70 and 2.50 g.cm - 3, the compressive strength, S, between 20 and 50 N.mm - 2, the leaching rates R n between 1.70 x 10 - 2 and 2.50 x 10 - 3 g.cm - 2.day - 1 and after 150 days leaching time the leaching rate R n - 3 g.cm - 2.day - 1. This study is expected to be used to propose the Indonesian National Standard (INS) (author)

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

    Waki, Toshikazu; Yamada, Motoyuki; Horikawa, Yoshihiko; Kaneko, Masaaki; Saso, Michitaka; Haruguchi, Yoshiko; Yamashita, Yu; Sakai, Hitoshi

    2009-01-01

    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. The Influence of Temperature on Kaolinite Fired Clay- Cement Materials Used as Disposal of Radioactive Waste

    El-Dakroury, A.

    2008-01-01

    The decay of encapsulated radioactive nuclides may cause elevation of the temperature in the waste form. Cement has been used for immobilization of low and intermediated level radioactive waste. The unstable component of concrete is the ordinary Portland cement (OPC). This study aimed to investigate experimentally the change occurring in the compressive strength and physico- chemical properties of ordinary Portland cement (OPC) matrices in comparable with other matrices which (OPC) were partially substituted by 0, 10, 20 and 30% of thermally-activated kaolinite clay by weight (the kaolinite clay activated by firing at 750 degree C for 5 hr then quenched by tap water). If all matrices mentioned before, are being exposed to the treatment temperature were varied from 100 degree C to 600 degree C by increment of 100 degree C for period of 3 hr without any load. The phase composition was performed by mean of differential thermal analysis and X-R-D. The results show that the replacements of (OPC) by 20 wt % thermal-activated Kaolinite clay improve the compressive properties by 30 %. The results of this investigation cleared that the recrystallization and carbonation of Ca(OH) 2 ; they also show a deformation of C-H-S and C 4 Ah 13 phases, besides the matrices have more stable resistance at 600 degree C. Meanwhile, this new immobilization matrix 20 % by wt thermally activated Kaolinite - clay showed the lowest leaching rate of simulated radioactive waste of Sr or Cs compared to the ordinary Portland cement (OPC) matrix

  1. Solidification of intermediate level liquid waste - ILLW, CEMEX waste form qualification

    D'Andrea, V.; Guerra, M.; Pancotti, F.; Maio, V.

    2015-01-01

    In the Sogin EUREX Facility about 125 m 3 of intermediate level radioactive waste and about 113 m 3 of low level radioactive waste, produced during the re-processing of MTR and CANDU fuel, are stored. Solidification of these wastes is planned in order to fulfill the specific requirements established by the Safety Authority, taking into account the criteria set up in a Technical Guide on the issue of radioactive waste management. The design of a cementation plant (CEMEX) of all liquid radioactive wastes is currently ongoing. The process requires that the liquid waste is neutralized with NaOH (NaOH 19 M) and metered into 440 liter drum together with the cement, while the mixture is stirred by a lost paddle ('in drum mixing process'). The qualification of the Waste Form consists of all the activities demonstrating that the final cemented product has 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. All tests performed to qualify the conditioning process for immobilizing first extraction cycle (MTR and CANDU) and second extraction cycle liquid wastes, gave results in compliance with the minimum requirements established for disposal

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

    Martensson, P.; Cronstrand, P.

    2013-01-01

    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 H 3 BO 3 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 H 3 BO 3 and the studies of AEC without H 3 BO 3 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)

  3. An Experimental Study on Effect of Palm – Shell Waste Additive to Cement Strenght Enhancement

    Adi Novriansyah

    2017-03-01

    Full Text Available Enhancing the cement strength through attaching chemical additive has been popular to meet the required condition for a particular well-cementing job. However, due to a low oil-price phenomenon, pouring and additive should be reconsidered because it can raise the cost and make the project become uneconomic. Another additive material in nanocomposite form will be introduced through this experimental study. The nanocomposite material consist of silica nanoparticle, known as “Nanosilica” and a palm-shell-waste, which is abundant in Indonesia. Before making a nanocomposite, the palm-shell should be burned to obtain a charcoal form, ground and sieved to attain a uniform size.   The study focuses on the two parameters, compressive strength and shear bond strength, which can reflect the strength of the cement. These values are obtained by performing a biaxial loading test to the cement sample. Various samples with different concentration of nanocomposite should be prepared and following the mixing, drying, and hardening process before the loading test is carried out. The result from the test shows a positive indication for compressive strength and shear bond strength values, according to the representative well cementing standards. Increasing the nanocomposite concentration on the cement will increase these values. Furthermore, an investigation on the temperature effect confirms that the sample with 700oC burning temperature have highest compressive-strength and shear-bond-strength values. This is a potential opportunity utilizing a waste-based material to produce another product with higher economic value.

  4. Thermal cycling and vibration response for PREPP concrete waste forms

    Nielson, R.M.; Welch, J.M.

    1983-06-01

    The Process Experimental Pilot Plant (PREPP) will process those transuranic wastes which do not satisfy the Waste Isolation Pilot Plant (WIPP) Waste Acceptance Criteria. Since these wastes will contain considerable quantities of combustible materials, incineration will be an integral part of the treatment process. Four basic types of PREPP ash wastes have been identified. The four types are designated high metal box waste, combustible waste, average waste, and inorganic sludge. In this process, the output of the incinerator is a mixture of ash and shredded noncombustible material (principally metals) which is separated into two sizes, -1/4 inch (under-size waste) and reverse arrow 1/4 inch (oversize waste). These wastes are solidified with hydraulic cement in 55-gallon drums. Simulated PREPP waste forms prepared by Colorado School of Mines Research Institute were subjected to thermal cycling and vibration testing to demonstrate compliance with the WIPP immobilization criterion. Although actual storage and transport conditions are expected to vary somewhat from those utilized in the testing protocol, the generation of only very small amounts of particulate suggests that the immobilization criterion should be routinely met for similar waste form formulations and production procedures. However, the behavior of waste forms containing significant quantities of off-gas scrubber sludge or considerably higher waste loadings may differ. Limited thermal cycling and vibration testing of prototype waste forms should be conducted if the final formulations or production methods used for actual waste forms differ appreciably from those tested in this study. If such testing is conducted, consideration should be given to designing the experiment to accommodate a larger number of thermal cycles more representative of the duration of storage expected

  5. Polymer-Cement Mortar with Quarry Waste as Sand Replacement

    D. N. Gómez-Balbuena

    2018-01-01

    Full Text Available The activities of carved Quarry extraction generate problems of landscape pollution such is the case of solid waste discharged into open land dumps in central Mexico. This article presents the technological application of this solid waste in a new polymeric material with properties similar to those of a traditional mortar. It is concluded that the polymeric material uses low amounts of cement with respect to the traditional mortar, and it is elaborated with the recycled quarry as they are presented in its granulometry. The polymer used favored a low water/cement ratio (0.3 which did not allow to decrease resistance due to the fine nature of the materials (residues and cement in addition to maintaining the workability of the material. The quarry residue was classified as silt with low plasticity and was characterized by X-ray diffraction and Fluorescence to identify 76% of SiO2, which is why it was used as a stone aggregate even though the fines content was approximately 93%. The maximum compression resistance obtained at 28 days were 8 Mpa with the polymer/solid ratios of 0.10, water/solids of 0.30, and quarry/solids of 0.67. Linear equations were analyzed for more representative values with R squared adjustment.

  6. Physico Mechanical Properties of Irradiated Waste Rubber Cement Mortar

    Younes, M.M.

    2010-01-01

    In the present study a partial replacement of aggregate with two different ratios of waste rubber (5%, 10%) with the addition of a constant ratio of rice husk ash (RHA), 5% was carried out. The hardened cement mortar used the optimum water of consistency. The specimens were molded into 1 inch cubic moulds .The specimens were first cured for 24 hours, at 100% relative humidity and then cured under tap water for 3, 7 and 28 days followed by irradiation at different doses of gamma irradiation namely 5 and 10 kGy. The physico-chemical and mechanical properties such as compressive strength, total porosity and bulk density were studied for the three types of specimens. The results showed that the values of the compressive strength, bulk density and chemically combined water of the blended cement mortar paste (OPC-RHA) increase ,while blended cement mortar paste with 5% RHA and 5, 10% waste rubber decrease. The results were confirmed by scanning electron microscopy and thermal behavior of the specimens. Also, it was observed that the irradiated sample was thermally more stable than the unirradiated one

  7. Immobilisation of shredded waste in a cement matrix

    James, J.M.; Smith, D.L.

    1987-10-01

    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)

  8. Effect of Concrete Waste Form Properties on Radionuclide Migration

    Mattigod, Shas V.; Bovaird, Chase C.; Wellman, Dawn M.; Skinner, De'Chauna J.; Cordova, Elsa A.; Wood, Marcus I.

    2009-01-01

    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. Numerous sets of tests were initiated in fiscal years (FY) 2006-2009 to evaluate (1) diffusion of iodine (I) and technetium (Tc) from concrete into uncontaminated soil after 1 and 2 years, (2) I and rhenium (Re) diffusion from contaminated soil into fractured concrete, (3) I and Re (set 1) and Tc (set 2) diffusion from fractured concrete into uncontaminated soil, (4) evaluate the moisture distribution profile within the sediment half-cell, (5) the reactivity and speciation of uranium (VI) (U(VI)) compounds in concrete porewaters, (6) the rate of dissolution of concrete monoliths, and (7) the diffusion of simulated tank waste into concrete.

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

    Colombo, P.; Kalb, P.D.; Fuhrmann, M.

    1983-09-01

    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 % Na 2 SO 4 , 25 wt % H 3 BO 3 , 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 % Na 2 SO 4 , 40 wt % H 3 BO 3 , 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

  10. Influence of industrial solid waste addition on properties of soil-cement bricks

    Siqueira, F. B.; Amaral, M. C.; Bou-Issa, R. A.; Holanda, J. N. F.

    2016-01-01

    Abstract The reuse of pollutant solid wastes produced in distinct industrial activities (avian eggshell waste and welding flux slag waste) as a source of alternative raw material for producing soil-cement bricks for civil construction was investigated. Soil-cement bricks containing up to 30 wt% of industrial solid waste were uniaxially pressed and cured for 28 days. Special emphasis is given on the influence of solid waste addition on the technical properties (as such volumetric shrinkage, wa...

  11. Low-level waste cement solidification design, installation, and start-up

    Jezek, G.R.

    1988-08-01

    This report describes the design, installation, and start-up activities of the Cement Solidification System (CSS) at the West Valley Demonstration Project (WVDP), West Valley, New York. The CSS, designed to operate within an existing process cell, automatically and remotely solidifies low-level nuclear waste by mixing it with Portland Type I cement. The qualified waste form mixture is placed into square, 270-litre (71-gallon) metal drums. The drums have an integral polyethylene liner to protect the carbon-steel material from potential corrosion. The CSS produces drums at a continuous operation rate of four drums per hour. All system processing data is monitored by a computerized Data Acquisition System (DAS). 6 figs

  12. Testing of variables which affect stablity of cement solidified low-level waste

    Boris, G.F.

    1989-01-01

    This paper describes the test program undertaken to investigate variables which could affect the stability of cement solidified low-level waste and to evaluate the effect of these variables on certain tests prescribed in the Technical Position on Waste Form. The majority of the testing was performed on solidified undepleted bead resin, however, six additional waste types, suggested by the NRC, were tested. The tested variables included waste loading, immersion duration, depletion level, ambient cure duration, curing environment, immersion medium and waste type. Of these, lower waste loadings, longer ambient cures prior to testing and immersion in demineralized water versus simulated sea water and potable water resulted in higher compressive strengths for bead resin samples. Immersion times longer than 90 days did not affect the resin samples. Compressive strengths for other waste types varied depending upon the waste. The strengths of all waste types exceeded the minimum criterion by at least a factor of four, up to a factor of forty. The higher waste loadings exhibit strengths less than the lower waste loadings

  13. Radiolytic gas generation from cement-based waste hosts for DOE low-level radioactive wastes

    Dole, L.R.; Friedman, H.A.

    1986-01-01

    Using cement-based immobilization binders with simulated radioactive waste containing sulfate, nitrate, nitrite, phosphate, and fluoride anions, the gamma- and alpha-radiolytic gas generation factors (G/sub t/, molecules/100 eV) and gas compositions were measured on specimens of cured grouts. These tests studied the effects of; (1) waste composition; (2) the sample surface-to-volume ratio; (3) the waste slurry particle size; and (4) the water content of the waste host formula. The radiolysis test vessels were designed to minimize the ''dead'' volume and to simulate the configuration of waste packages

  14. Solidification of low-level radioactive liquid waste using a cement-silicate process

    Grandlund, R.W.; Hayes, J.F.

    1979-01-01

    Extensive use has been made of silicate and Portland cement for the solidification of industrial waste and recently this method has been successfully used to solidify a variety of low level radioactive wastes. The types of wastes processed to date include fuel fabrication sludges, power reactor waste, decontamination solution, and university laboratory waste. The cement-silicate process produces a stable solid with a minimal increase in volume and the chemicals are relatively inexpensive and readily available. The method is adaptable to either batch or continuous processing and the equipment is simple. The solid has leaching characteristics similar to or better than plain Portland cement mixtures and the leaching can be further reduced by the use of ion-exchange additives. The cement-silicate process has been used to solidify waste containing high levels of boric acid, oils, and organic solvents. The experience of handling the various types of liquid waste with a cement-silicate system is described

  15. Effect of Ground Waste Concrete Powder on Cement Properties

    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.

  16. Coated particle waste form development

    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

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

    Dalen, A. van; Rijpkema, J.E.

    1989-01-01

    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 220 0 C. The low leachability, the very slow effective diffusion of H 2 O 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

  18. Physical and thermal behavior of cement composites reinforced with recycled waste paper fibers

    Hospodarova, Viola; Stevulova, Nadezda; Vaclavik, Vojtech; Dvorsky, Tomas

    2017-07-01

    In this study, three types of recycled waste paper fibers were used to manufacture cement composites reinforced with recycled cellulosic fibers. Waste cellulosic fibers in quantity of 0.2, 0.3, and 0.5 wt.% were added to cement mixtures. Physical properties such as density, water capillarity, water absorbability and thermal conductivity of fiber cement composites were studied after 28 days of hardening. However, durability of composites was tested after their water storage up to 90 days. Final results of tested properties of fiber cement composites were compared with cement reference sample without cellulosic fibers.

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

    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.

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

    Vejmelka, P.; Johnsen, P.; Kluger, W.; Koester, R.

    1987-01-01

    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 3x10 5 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, 800 0 C) ≅ 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 NaNO 3 (175 litre drum) with thermal load is considerably higher, as due to the NaNO 3 content of the products, after an induction period of about 20 minutes there is an exothermal reaction between the bitumen and the NaNO 3 , 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) [de

  1. Recent advances in cement solidification of radioactive wastes

    Vigreux, B.; Jaouen, C.

    1987-01-01

    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

  2. Radiation effects on medium active waste forms. Annual report - 1988

    Johnson, D.; Wilding, C.; Lyon, C.

    1989-01-01

    Work has continued on measurements of dimensional changes, strength, and gas evolution on samples of several simulated waste forms under accelerated γ and α irradiation conditions. Samples of RMA5 (mixed ion exchangers in modified vinylester polymer) and RMA10 (incinerated PCM materials in cement) maintain their integrity during irradiation but samples of RMA3 (organic ion exchangers in cement) and RMA11.1 (mixed PCM materials in cement) swell and eventually disintegrate under some γ irradiation conditions. Disintegration of RMA3 samples occurred when samples were γ irradiated whilst immersed in water. Samples of RMA11.1 which cannot rapidly dry out swell, sometimes substantially, during γ irradiation. The principal gases of interest in gas evolution experiments are hydrogen and oxygen. Hydrogen is evolved under all circumstances but oxygen evolution does not always occur. Samples of RMA10 evolve oxygen when α irradiated in an inert atmosphere but oxygen concentration initially falls during α irradiation in air atmosphere. Samples of RMA11.1 absorb oxygen from an air atmosphere during both α and γ irradiation. A comparison has been carried out of the effects of γ and α irradiation on identical cement grouts using BFS/OPC mixes produced under high shear mixing conditions. In contrast to earlier results on such systems, no γ irradiated samples showed physical deterioration after irradiation to 9 MGy but the a irradiated samples all showed surface cracks after about 1 MGy. The gas evolution measurements showed that during α irradiation oxygen evolution commenced after a dose of ∼ 1 MGy whereas oxygen was completely removed from the atmosphere γ irradiation. Hydrogen was evolved under all conditions and the rate of production was found to be dependent upon the dose rate. More hydrogen was evolved during α irradiations than during γ irradiation. A technique for the measurement of hydrogen permeability through cement systems has been further developed

  3. Design and application of environmentally effective concrete with usage of chrysotile-cement waste

    Egorova Lada

    2016-01-01

    Full Text Available Construction is resource-demanding industry, characterized by a large volume of waste. Particularly chrysotile cement waste obtained both in production and in dismantling over age chrysotile-cement products: corrugated asbestos boards and flat sheets, chrysotile-cement tubes. We propose to use dry chrysotile-cement waste as recycled aggregate for concrete. Based on developed compositions and identified properties of heavy concrete with chrysotile-cement waste introduce this technology to the production of foundation wall blocks. The studies confirmed the possibility of using chrysotile-cement aggregate and fine screening of crushing as a secondary coarse and fine aggregates for concrete with proper quality without increasing the cost of the product. Environmental safety of the obtained products was ensured. The direction for implementation of the research project was proposed.

  4. Secondary Waste Cast Stone Waste Form Qualification Testing Plan

    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

  5. The effect of concentration on the structure and crystallinity of a cementitious waste form for caustic wastes

    Chung, Chul-Woo; Turo, Laura A.; Ryan, Joseph V.; Johnson, Bradley R.; McCloy, John S.

    2013-01-01

    Highlights: ► Cast Stone: Portland cement, fly ash, blast furnace slag, and simulated nuclear waste. ► Caustic secondary waste from the off-gas of a vitrification process was targeted. ► Crystallinity, micro- and mesostructure, and engineering properties characterized. ► Waste concentration varied from 0 to 2.5 M, but caused minimal changes. ► Cast Stone shows good compositional versatility as a secondary waste form. -- Abstract: Cement-based waste forms have long been considered economical technologies for disposal of various types of waste. A solidified cementitious waste form, Cast Stone, has been identified to immobilize the radioactive secondary waste from vitrification processes. In this work, Cast Stone was considered for a Na-based caustic liquid waste, and its physical properties were analyzed as a function of liquid waste loading up to 2 M Na. Differences in crystallinity (phase composition), microstructure, mesostructure (pore size distribution and surface area), and macrostructure (density and compressive strength) were investigated using various analytical techniques, in order to assess the suitability of Cast Stone as a chemically durable waste. It was found that the concentration of secondary waste simulant (caustic waste) had little effect on the relevant engineering properties of Cast Stone, showing that Cast Stone could be an effective and tolerant waste form for a wide range of concentrations of high sodium waste

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

    Brown, D.J.; Palmer, J.D.; Smith, D.L.

    1983-10-01

    The testing of cemented waste specimens containing sodium nitrate and sodium sulphate has been carried out over a period of two and a half years. One set of samples is kept under environmental conditions which simulate the anticipated storage conditions and a second series is stored under water. Additional results are reported here from the long-term stability testing of these simulated cemented wastes. Some applications of the method developed for measuring the apparent viscosity of cement pastes are described. Some preliminary information is given on the assessment of cement-waste mixing equipment. (U.K.)

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

    Gougar, M.L.D.; Scheetz, B.E.; Roy, D.M.

    1996-01-01

    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

  8. Behavior of cement paste as backfill in waste disposal boreholes

    Ferreira, Eduardo G.A.; Isiki, Vera L.K.; Miyamoto, Hissae; Marumo, Julio T.; Vicente, Roberto

    2011-01-01

    The Radioactive Waste Management Laboratory (GRR) at the Nuclear and Energy Research Institute (IPEN) in Sao Paulo, Brazil, is developing the concept a repository for disposition of disused sealed radioactive sources in a deep borehole, aiming at providing a feasible and inexpensive alternative for final disposal. A relevant fraction of the Brazilian inventory of sources has long half-life which prevents them to be disposed of in shallow ground disposal facilities. In the concept of repository under study, Portland cement paste is intended to be used as a backfill between the steel casing and the geological formation around the borehole. Cement paste will function as structural, an additional barrier against the migration of radionuclides outside the repository, and as a blockage against the transport of water between the different strata of the geological setting. The durability of cementitious materials under the conditions prevailing at the depth of disposal is as yet unknown. The objective of this research is to investigate the behavior of the cement paste and to estimate its service life. In this paper we present the results of mechanical strength measurements and chemical and mineralogical analysis of samples to detect the changes caused by radiation, temperature and aggressive chemicals present in ground water. Techniques of analysis included Inductively Coupled Plasma Atomic Emission Spectroscopy, Ion Chromatography, X-Ray Diffraction, and Thermo-Gravimetric Analysis. (author)

  9. Cement-based processes for the immobilization of intermediate level radioactive waste

    Brown, D.J.; Lee, D.J.; Price, M.S.T.; Smith, D.L.G.

    1985-01-01

    Increasing attention is being paid to the use of cement-based materials for the immobilisation of intermediate level wastes. Various cementitious materials are surveyed and the use of blast furnace slag is shown to be advantageous. The properties of cemented wastes are surveyed both during processing and as solid products. The application of Winfrith Cementation Laboratory technology to plant and flowsheet development for Winfrith Reactor sludge immobilisation is described. (author)

  10. TSA waste stream and final waste form composition

    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

  11. 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

    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)

  12. Physical, chemical and mineralogical characterization of water treatment plant waste for use in soil-cement brick

    Pessin, L.R.; Destefani, A.Z.; Holanda, J.N.F.

    2011-01-01

    The water treatment plants (WTP) for human consumption generate huge amounts of waste in the form of sludge (sludge) that have been over the years mostly inadequately prepared in water resources and the environment. Moreover, traditional methods of disposal of waste water treatment plants commonly used are generally costly activities. An alternative method for disposal of this waste abundant is its incorporation in ceramic products. This work is focused on the physical-chemical and mineralogical composition of a sample of waste water treatment plants from the region of Campos dos Goytacazes-RJ to their use in the manufacture of soil-cement brick. Several characterization techniques were used including X-ray diffraction, X-ray fluorescence, scanning electron microscopy, picnometry, particle size analysis and plasticity. The experimental results indicate that the waste water treatment plants have the potential to be used in the manufacture of ecologic soil-cement bricks. (author)

  13. Preliminary assessment of nine waste-form products/processes for immobilizing transuranic wastes

    Crisler, L.R.

    1980-09-01

    Nine waste-form processes for reduction of the present and projected Transuranic (TRU) waste inventory to an immobilized product have been evaluated. Product formulations, selected properties, preparation methods, technology status, problem areas needing resolution and location of current research development being pursued in the United States are discussed for each process. No definitive utility ranking is attempted due to the early stage of product/process development for TRU waste containing products and the uncertainties in the state of current knowledge of TRU waste feed compositional and quantitative makeup. Of the nine waste form products/processes included in this discussion, bitumen and cements (encapsulation agents) demonstrate the degree of flexibility necessary to immobilize the wide composition range present in the TRU waste inventory. A demonstrated process called Slagging Pyrolysis Incineration converts a varied compositional feed (municipal wastes) to a ''basalt'' like product. This process/product appears to have potential for TRU waste immobilization. The remaining waste forms (borosilicate glass, high-silica glass, glass ceramics, ''SYNROC B'' and cermets) have potential for immobilizing a smaller fraction of the TRU waste inventory than the above discussed waste forms

  14. Crystallization behavior of nuclear waste forms

    Rusin, J.M.; Lokken, R.O.; May, R.P.; Wald, J.W.

    1981-09-01

    Several waste form options have been or are being developed for the immobilization of high-level wastes. The final selection of a waste form must take into consideration both waste form product as well as process factors. Crystallization behavior has an important role in nuclear waste form technology. For glass or vitreous waste forms, crystallization is generally controlled to a minimum by appropriate glass formulation and heat treatment schedules. With glass ceramic waste forms, crystallization is essential to convert glass products to highly crystalline waste forms with a minimum residual glass content. In the case of ceramic waste forms, additives and controlled sintering schedules are used to contain the radionuclides in specific tailored crystalline phases

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

    Darnell, G.R.

    1992-01-01

    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

  16. Nuclear waste forms for actinides

    Ewing, Rodney C.

    1999-01-01

    The disposition of actinides, most recently 239Pu from dismantled nuclear weapons, requires effective containment of waste generated by the nuclear fuel cycle. Because actinides (e.g., 239Pu and 237Np) are long-lived, they have a major impact on risk assessments of geologic repositories. Thus, demonstrable, long-term chemical and mechanical durability are essential properties of waste forms for the immobilization of actinides. Mineralogic and geologic studies provide excellent candidate phases for immobilization and a unique database that cannot be duplicated by a purely materials science approach. The “mineralogic approach” is illustrated by a discussion of zircon as a phase for the immobilization of excess weapons plutonium. PMID:10097054

  17. Development of low-pH cements for immobilisation of intermediate level radioactive waste: achievements and challenges

    Martins de Freitas, Regeane; Al-Tabbaa, Abir [Engineering Department, University of Cambridge, Cambridge (United Kingdom)

    2013-07-01

    Although cementation is a widely recognized solidification/ stabilization process for immobilisation of Intermediate Level Radioactive Waste (ILRW), the low resistance to hyper-alkaline pore waters compromises the effectiveness of the process when Portland Cement (PC) is employed. Moreover the manufacture of PC is responsible for significant CO{sub 2} emissions. In this context, low pH cements are environmentally more suitable and have emerged as a potential alternative for obtaining secure waste forms. This paper summarises the achievements on development of low-pH cements and the challenges of using these new materials for the ILRW immobilisation. The performance of waste forms is also discussed in terms of radionuclides release. Reactive magnesium oxide and magnesium phosphate cements are emphasised as they feature important advantages such as consumption of available constituents for controlling acid-base reactions, reduced permeability and higher density. Additionally, in order to identify new opportunities for study, the long-term modelling approach is also briefly discussed. (authors)

  18. Utilization of small-angle neutron scattering to decide the maximum loading of nuclear waste in cement matrix

    Das, Avik; Mazumder, S.; Sen, D.; Yalmali, V.; Shah, J.G.

    2014-01-01

    Nuclear power plants generate many kinds of hazardous nuclear waste which are needed to be disposed in an eco-friendly manner. Many different waste incarceration techniques have been adapted for managing the nuclear waste of different category of radioactivity. Immobilisation of low and intermediate level radioactive wastes in cement matrix is one of the widely used and cost-effective techniques in waste management. However, loading of nuclear waste in cement matrix can alter the mesoscopic structure of the hydrated cement and hence, it is very important to set the maximum limit of waste loading in cement for providing proper physical isolation to the nuclear waste

  19. CERAMIC WASTE FORM DATA PACKAGE

    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.

  20. Solid forms for Savannah River Plant radioactive wastes

    Wallace, R.M.; Hale, W.H.; Bradley, R.F.; Hull, H.L.; Kelley, J.A.; Stone, J.A.; Thompson, G.H.

    1976-01-01

    Methods are being developed to immobilize Savannah River Plant wastes in solid forms such as cement, asphalt, or glass. 137 Cs and 90 Sr are the major biological hazards and heat producers in the alkaline wastes produced at SRP. In the conceptual process being studied, 137 Cs removed from alkaline supernates, together with insoluble sludges that contain 90 Sr, 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 137 Cs, 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

  1. Synroc tailored waste forms for actinide immobilization

    Gregg, Daniel J.; Vance, Eric R. [Australian Nuclear Science and Technology Organisation, Kirrawee (Australia). ANSTOsynroc, Inst. of Materials Engineering

    2017-07-01

    Since the end of the 1970s, Synroc at the Australian Nuclear Science and Technology Organisation (ANSTO) has evolved from a focus on titanate ceramics directed at PUREX waste to a platform waste treatment technology to fabricate tailored glass-ceramic and ceramic waste forms for different types of actinide, high- and intermediate level wastes. The particular emphasis for Synroc is on wastes which are problematic for glass matrices or existing vitrification process technologies. In particular, nuclear wastes containing actinides, notably plutonium, pose a unique set of requirements for a waste form, which Synroc ceramic and glass-ceramic waste forms can be tailored to meet. Key aspects to waste form design include maximising the waste loading, producing a chemically durable product, maintaining flexibility to accommodate waste variations, a proliferation resistance to prevent theft and diversion, and appropriate process technology to produce waste forms that meet requirements for actinide waste streams. Synroc waste forms incorporate the actinides within mineral phases, producing products which are much more durable in water than baseline borosilicate glasses. Further, Synroc waste forms can incorporate neutron absorbers and {sup 238}U which provide criticality control both during processing and whilst within the repository. Synroc waste forms offer proliferation resistance advantages over baseline borosilicate glasses as it is much more difficult to retrieve the actinide and they can reduce the radiation dose to workers compared to borosilicate glasses. Major research and development into Synroc at ANSTO over the past 40 years has included the development of waste forms for excess weapons plutonium immobilization in collaboration with the US and for impure plutonium residues in collaboration with the UK, as examples. With a waste loading of 40-50 wt.%, Synroc would also be considered a strong candidate as an engineered waste form for used nuclear fuel and highly

  2. Structural and microstructural aspects of asbestos-cement waste vitrification

    Iwaszko, Józef; Zawada, Anna; Przerada, Iwona; Lubas, Małgorzata

    2018-04-01

    The main goal of the work was to evaluate the vitrification process of asbestos-cement waste (ACW). A mixture of 50 wt% ACW and 50 wt% glass cullet was melted in an electric furnace at 1400 °C for 90 min and then cast into a steel mold. The vitrified product was subjected to annealing. Optical microscopy, scanning electron microscopy (SEM), Fourier transform infrared spectroscopy (FT-IR) and X-ray diffraction (XRD) were used to evaluate the effects of the vitrification. The chemical constitution of the material before and after the vitrification process was also analyzed. It was found that the vitrified product has an amorphous structure in which the components of asbestos-cement waste are incorporated. MIR spectroscopy showed that the absorption bands of chrysotile completely disappeared after the vitrification process. The results of the spectroscopic studies were confirmed by X-ray studies - no diffraction reflections from the chrysotile crystallographic planes were observed. As a result of the treatment, the fibrous asbestos construction, the main cause of its pathogenic properties, completely disappeared. The vitrified material was characterized by higher resistance to ion leaching in an aquatic environment than ACW and a smaller volume of nearly 72% in relation to the apparent volume of the substrates. The research has confirmed the high effectiveness of vitrification in neutralizing hazardous waste containing asbestos and the FT-IR spectroscopy was found to be useful to identify asbestos varieties and visualizing changes caused by the vitrification process. The work also presents the current situation regarding the utilization of asbestos-containing products.

  3. Secondary waste form testing: ceramicrete phosphate bonded ceramics

    Singh, D.; Ganga, R.; Gaviria, J.; Yusufoglu, Y.

    2011-01-01

    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 3 , and Class C fly ash + slag. Optimum waste loadings were as high as 20 wt.% for the fly ash and CaSiO 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 3 filler waste forms appeared to be superior to the waste forms with fly ash + slag as a filler. Waste form weight loss was ∼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 binder components from

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

    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

  5. Influence of superplasticizers on the long-term properties of cement pastes and possible impact on radionuclide uptake in a cement-based repository for radioactive waste

    Wieland, E.; Lothenbach, B.; Glaus, M.A.; Thoenen, T.; Schwyn, B.

    2014-01-01

    Highlights: • We investigate the hydration of different cement mixes containing concrete admixtures. • The concentration of concrete admixtures decreases with time due to sorption on cement phases. • We observe no influence on the phase composition of cement paste and the ion composition of pore fluids. • Uptake of 63 Ni, 152 Eu and 228 Th by cement paste is not affected by the concrete admixtures. - Abstract: Cementitious materials will be used for the construction of the engineered barrier of the planned repositories for radioactive waste in Switzerland. Superplasticizers (SPs) are commonly used to improve the workability of concretes and, along with a set accelerator (Acc), to produce shotcrete. In this study the influence of a polycarboxylate- (PCE) and a polynaphthalene-sulphonate-based (PNS) SP on the hydration process, mineral composition and the sorption behaviour of metal cations has been investigated using an ordinary Portland cement (OPC), a low-alkali cement mix (LAC) consisting of CEM III-type cement and nanosilica, and a shotcrete-type cement mix (ESDRED) consisting of a CEM I-type cement and silica fume prepared in the presence of an alkali-free set accelerator. Both the PCE and PNS SP do not significantly influence the amount and quantity of hydrates formed during hydration. The concentration of both SPs decreased rapidly in the early stage of the hydration process for all cements due to sorption onto cement phases. After 28 days of hydration and longer, the concentration of the PNS SP in the pore fluids of all cements was generally lower than that of the PCE SP, indicating stronger uptake of the PNS SP. The formate present in the Acc sorbs only weakly onto the cement phases, which led to higher aqueous concentration of organics in the ESDRED cement than in OPC and LAC. Sorption experiments with 63 Ni, 152 Eu and 228 Th on a cation exchange resin indicate that, at concentrations above 0.1 g L −1 , the two SPs could reduce sorption of metal

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

    Li, Guanshu

    1994-01-01

    inter-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)

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

    Brodersen, K.

    1992-01-01

    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......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....... A model of the quantitative aspects for the equilibrium situation is presented. Experiments with hygroscopic water uptake support the model and give indications about the rate of the process. The release mechanism is only thought to be important for radionuclides which are not fixed in a low...

  8. Increased Coal Replacement in a Cement Kiln Burner by Feeding a Mixture of Solid Hazardous Waste and Shredded Plastic Waste

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

    2013-01-01

    The present study aims to find the maximum possible replacement of coal by combined feeding of plastic waste and solid hazardous waste mixed with wood chips (SHW) in rotary kiln burners used in cement kiln systems. The coal replacement should be achieved without negative impacts on product quality, emissions or overall operation of the process. A full-scale experiment was carried out in the rotary kiln burner of a cement kiln by varying SHW and plastic waste feeding rates. Experimental result...

  9. Mechanical and chemical properties of polyvinyl alcohol modified cement mortar with silica fume used as matrix including radioactive waste

    Dakroury, A. M.

    2007-01-01

    This paper discussed the mechanical and chemical properties of polyvinyl alcohol - modified cement mortar with silica fume to assess the safety for disposal of radioactive waste. The modified cement mortars containing polyvinyl alcohol (PVA) in the presence of 10 % silica fume (SF) .The chemical reaction between polymer and cement - hydrated product were investigated by the Infrared Spectral Technology, Differential Thermal Analysis and X-ray diffraction. The leaching of 137Cs from a waste composite into a surrounding fluid has been studied .The results shown that PVA increases the strength and decreases the porosity. The increase in strength duo to the interaction of PVA with cement , may be forming some new compound that fill the pores or improve the bond between the cement . The pozzolanic reaction of the SF increases the calcium silicate hydrates in the hardening matrix composites. There is distinct change in the refinement of the pore structure in cement composites giving fewer capillary pores and more of the finer gel pores

  10. Special waste form lysimeters-arid. Annual report, 1985

    Walter, M.B.; Graham, M.J.

    1985-09-01

    The Special Waste Form Lysimeters-Arid program was initiated to determine typical source terms generated by commercial solidified low-level nuclear waste in an arid climate. Waste-form leaching tests are being conducted at a field facility at the Hanford site near Richland, Washington. A similar program is being conducted at a humid site. The field facility consists of 10 lysimeters placed around a central instrument caisson. The waste samples from boiling water and pressurized water reactors were emplaced in 1984, and the lysimeters are being monitored for movement of contaminants and water. Solidifying agents being tested include vinyl ester-styrene, bitumen, and cement. Laboratory leaching and geochemical modeling studies are being conducted to predict expected leach rates at the field site and to aid field-data interpretation. Small samples of the solidified waste forms were made for use in the laboratory leaching studies that include standard leach tests and leaching of solidified waste forms in soil columns. Complete chemical and radionuclide analyses are being conducted on the solid and liquid portions of the wastes. 2 refs

  11. Special Waste Form Lysimeters-Arid: annual report 1985

    Walter, M.B.; Graham, M.J.

    1986-01-01

    The Special Waste Form Lysimeters-Arid program was initiated to determine typical source terms generated by commercial solidified low-level nuclear waste in an arid climate. Waste-form leaching tests are being conducted at a field facility at the Hanford site near Richland, Washington. A similar program is being conducted at a humid site. The field facility consists of 10 lysimeters placed around a central instrument caisson. The waste samples from boiling water and pressurized water reactors were emplaced in 1984, and the lysimeters are being monitored for movement of contaminants and water. Solidifying agents being tested include vinyl ester-styrene, bitumen, and cement. Laboratory leaching and geochemical modeling studies are being conducted to predict expected leach rates at the field site and to aid field-data interpretation. Small samples of the solidified waste forms were made for use in the laboratory leaching studies that include standard leach tests and leaching of solidified waste forms in soil columns. Complete chemical and radionuclide analyses are being conducted on the solid and liquid portions of the wastes

  12. An examination of sulfur polymer cement as a waste encapsulation agent

    McNew, E.B.

    1995-01-01

    Sulfur polymer cement (SPC) is a unique material having potential applications for hazardous and radioactive waste encapsulation. This material was originally developed by the US Bureau of Mines as an acid and chemical resistant construction cement and has since been applied in tie waste encapsulation field. The material is easily prepared from elemental sulfur and organic dienes. It is an easy to use low-viscosity thermoplastic, and has many favorable properties such as low porosity, high compressive strength, and resistance to chemical attack. The results of several invetigations on this material will be discussed, and include: (1) the chemical form and physical structure of the material, (2) the compressive strength of cylindrical test samples after gamma radiation testing, (3) the aqueous leaching behavior of lead, cerium, cesium, cobalt, and strontium from SPC-ash mixtures at room and elevated temperatures, (4) the casting compatibility of mixtures of SPC with different waste materials, (5) the ability of SPC to encapsulate elemental mercury contaminated soils, (6) laboratory and field studies of SPC biocorrosion by Thiobacillus bacteria, (7) small scale (10 kg) SPC-ash monolith casting studies, and (8) methods for the formulation of a grade of SPC more applicable to the encapsulation of aggregate waste materials

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

    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.

  14. Processes for production of alternative waste forms

    Ross, W.A.; Rusin, J.M.; McElroy, J.L.

    1979-01-01

    During the past 20 years, numerous waste forms and processes have been proposed for solidification of high-level radioactive wastes (HLW). The number has increased significantly during the past 3 to 4 years. At least five factors must be considered in selecting the waste form and process method: 1) processing flexibility, 2) waste loading, 3) canister size and stability, 4) waste form inertness and stability, and 5) processing complexity. This paper describes various waste form processes and operations, and a simple system is proposed for making comparisons. This system suggests that one goal for processes would be to reduce the number of process steps, thereby providing less complex processing systems

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

    Dragolici, F.

    2013-01-01

    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)

  16. Formulation of portland composite cement using waste glass as a supplementary cementitious material

    Manullang, Ria Julyana; Samadhi, Tjokorde Walmiki; Purbasari, Aprilina

    2017-09-01

    Utilization of waste glass in cement is an attractive options because of its pozzolanic behaviour and the market of glass-composite cement is potentially available. The objective of this research is to evaluate the formulation of waste glass as supplementary cementitious material (SCM) by an extreme vertices mixture experiment, in which clinker, waste glass and gypsum proportions are chosen as experimental variables. The composite cements were synthesized by mixing all of powder materials in jar mill. The compressive strength of the composite cement mortars after being cured for 28 days ranges between 229 to 268 kg/cm2. Composite cement mortars exhibit lower compressive strength than ordinary Portland cement (OPC) mortars but is still capable of meeting the SNI 15-7064-2004 standards. The highest compressive strength is obtained by shifting the cement blend composition to the direction of increasing clinker and gypsum proportions as well as reducing glass proportion. The lower compressive strength of composite cement is caused by expansion due to ettringite and ASR gel. Based on the experimental result, the composite cement containing 80% clinker, 15% glass and 5% gypsum has the highest compressive strength. As such, the preliminary technical feasibility of reuse of waste glass as SCM has been confirmed.

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

    Onofrei, M.

    1996-01-01

    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 -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

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

    Vejmelka, P.; Sambell, R.A.J.

    1984-01-01

    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

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

    Kelly, K.E.; Beeh, J.

    1994-01-01

    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

  20. Secondary Waste Form Down Selection Data Package – Ceramicrete

    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

  1. Establishment of cementation parameters of dried waste from evaporation coming from NPP operation

    Faria, Érica R.; Tello, Clédola C.O.; Costa, Bruna S.

    2017-01-01

    The radioactive wastes generated in Brazil are treated and sent to initial and intermediate storages. The 'Project RBMN' proposes the implantation of the Brazilian repository to receive and permanently dispose the low and intermediate level radioactive wastes. The CNEN NN 6.09 standard - Acceptance Criteria for Disposal of Low and Intermediate Radioactive Wastes (LIRW) - establishes the fundamental requirements to accept the wastes packages in the repository. The evaporator concentrate is one of liquid wastes generated in a Nuclear Power Plant (NPP) operation and usually it is cemented directly inside the packing. The objective of this research is to increase the amount of the incorporated waste in each package, using the drying process before the cementation, consequently reducing the volume of the waste to be disposed. Drying and cementation parameters were established in order to scale-up the process aiming at waste products that comply with the requirements of CNEN standard. The cementation of the resulting dry wastes was carried out with different formulations, varying the amount of cement, dry waste and water. These tests were analyzed in order to select the best products, with higher waste incorporation than current process and its complying the requirements of the standard CNEN NN 6.09. (author)

  2. Establishment of cementation parameters of dried waste from evaporation coming from NPP operation

    Faria, Érica R.; Tello, Clédola C.O., E-mail: erica.engqui@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte/MG (Brazil); Costa, Bruna S., E-mail: brusilveirac@gmail.com [Universidade Federal de Minas Gerais, Belo Horizonte, MG (Brazil)

    2017-07-01

    The radioactive wastes generated in Brazil are treated and sent to initial and intermediate storages. The 'Project RBMN' proposes the implantation of the Brazilian repository to receive and permanently dispose the low and intermediate level radioactive wastes. The CNEN NN 6.09 standard - Acceptance Criteria for Disposal of Low and Intermediate Radioactive Wastes (LIRW) - establishes the fundamental requirements to accept the wastes packages in the repository. The evaporator concentrate is one of liquid wastes generated in a Nuclear Power Plant (NPP) operation and usually it is cemented directly inside the packing. The objective of this research is to increase the amount of the incorporated waste in each package, using the drying process before the cementation, consequently reducing the volume of the waste to be disposed. Drying and cementation parameters were established in order to scale-up the process aiming at waste products that comply with the requirements of CNEN standard. The cementation of the resulting dry wastes was carried out with different formulations, varying the amount of cement, dry waste and water. These tests were analyzed in order to select the best products, with higher waste incorporation than current process and its complying the requirements of the standard CNEN NN 6.09. (author)

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

    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

    2004-01-01

    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 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)

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

    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)

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

    Marques, M.A.; Antunes, M.L.P.; Montagnoli, R.M.; Mancini, S.D.

    2012-01-01

    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)

  6. Method of cement-solidification of radioactive liquid wastes containing surfactant

    Sugimoto, Y; Yusa, H

    1979-04-10

    Purpose: To provide the subject method comprising the steps of adjusting the concentration of the surfactant to a value less than the predetermined value even when the concentration of the surfactant is high, and rendering the uniaxial compression strength of the cement-solidification body into more than the defined fabrication reference value. Method: To radioactive liquid wastes there are applied means for boiling and heating liquid wastes by addition of sulfuric acid, means for cracking surfactants by the addition of oxidants and means for precipitating and arresting surfactants. After suppressing the hindrance of the cement hydration reaction by surfactants, the radioactive liquid wastes are cement-solidified. (Nakamura, S.).

  7. Miscellaneous Waste-Form FEPs

    Schenker, A.

    2000-01-01

    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

  8. Testing protocols for evaluating monolithic waste forms containing mixed wastes

    Gilliam, T.M.; Sams, T.L.; Pitt, W.W.

    1986-01-01

    Test protocols have been presented which can be used as a guide in cement-based grout formulation development studies. Based on experience at ORNL, these six tests are generally sufficient to develop a grout product which will meet all applicable DOE, NRC, and EPA performance criteria. As such, these tests can be used to minimize the time required to tailor a grout to be compatible with both the waste stream and the process disposal scenario. 9 refs

  9. Formulation development for PREPP concreted waste forms

    Neilson, R.M. Jr.; Welch, J.M.

    1984-05-01

    Analysis of variance and logistic regression techniques have been used to develop models describing the effects of formulation variables and their interactions on compressive strength, solidification, free-standing water, and workability of hydraulic cement grouts incorporating simulated Process Experimental Pilot Plant (PREPP) wastes. These models provide the basis for specifications of grout formulations to solidify these wastes. The experimental test matrix, formulation preparation, and test methods employed are described. The development of analytical models for formulation behavior and the conclusions drawn regarding appropriate formulation variable ranges are discussed. 13 references, 9 figures, 15 tables

  10. The sorption and mechanical properties of the modified cement matrix used for conditioning of radioactive waste

    Dogaru, Daniela; Nuculae, Ortenzia; Jinescu, Gheorghita; Duliu, Octavian; Dogaru, Gheorghe

    2008-01-01

    Full text: Radioactive contaminant sorption onto concrete represents one of the most important retardation mechanisms in engineered barriers such as the conditioning matrix itself, concrete walls and concrete floors. During the life of a disposal facility for radioactive waste, the sorption properties as well as the mechanical properties of the cement are affected by both external and internal processes. The most important sorbing material present in concrete is the hydrated cement. The sorption data obtained for specific cement or cement mixes in general may be used to characterize a given cement type. In order to improve the mechanical and sorption properties of the cement matrix, different additives were used in the laboratory tests. The used additives are known to have good sorption properties. The paper describes the influence of the concentration of additives on the mechanical and sorption properties of the cement matrix. As radioactive contaminants 134 Cs, 60 Co, 3 H, 241 Am were used. (authors)

  11. Joint Cementation of liquid and solid radioactive waste in decommissioning of atomic objects

    Varlakov, A.; Varlakova, G.; Germanov, A.; Sukhanov, L.

    2015-01-01

    Joint cementation of liquid and solid radioactive waste (RW) directly in the containers that serve as the final packaging allows the reduction of waste amounts sent for storage and disposal. In the A.A. Bochvar Research Institute we have created a mobile unit and conducted tests on joint grouting of solid RW with the use of cement mixtures prepared on the basis of liquid RW. Two variants of void filling between the fragments of solid RW in a container: pouring and impregnation were examined. Impregnation is the so-called method suggested to be used for the solid RW with small fragments and dense filling layer. In this case the gaps between the waste are filled with the cement compound by using special technological procedures, in particular, the vibration action on the container filled with waste and the use of cement mixture with high level of penetration. It was observed that the pouring method is right for the cementation of solid RW with fragment sizes not exceeding 100-150 mm. Impregnation method can be used for the joint grouting of practically all types of solid RW regardless their size and fragments homogeneity. Cementation of densely packed and large size solid RW by impregnation guarantees the joint grouting of the lower layers of the waste in the container and can be controlled by determination of the impregnation degree and density of the cement mixture that passed through the waste layer

  12. Utilization of municipal solid waste incineration (MSWI) fly ash in blended cement

    Aubert, J.E.; Husson, B.; Sarramone, N.

    2006-01-01

    This paper is the first of a series of two articles dealing with the processes applied to MSWI fly ash with a view to reusing it safely in cement-based materials. Part 1 presents two stabilization processes and Part 2 deals with the use of the two treated fly ashes (TFA) in mortars. Two types of binder were used: an Ordinary Portland Cement (OPC) containing more than 95% clinker (CEM I 52.5R) and a binary blend cement composed of 70% ground granulated blast furnace slag and 30% clinker (CEM III-B 42.5N). In this first part, two stabilization processes are presented: the conventional process, called 'A', based on the washing, phosphation and calcination of the ash, and a modified process, called 'B', intended to eliminate metallic aluminum and sulfate contained in the ash. The physical, chemical and mineralogical characteristics of the two TFA were comparable. The main differences observed were those expected, i.e. TFA-B was free of metallic aluminum and sulfate. The mineralogical characterization of the two TFAs highlighted the presence of large amounts of a calcium aluminosilicate phase taking two forms, a crystalline form (gehlenite) and an amorphous form. Hydration studies on pastes containing mixed TFA and calcium hydroxide showed that this phase reacted with calcium hydroxide to form calcium aluminate hydrates. This formation of hydrates was accompanied by a hardening of the pastes. These results are very encouraging for the reuse of such TFA in cement-based materials because they can be considered as pozzolanic additions and could advantageously replace a part of the cement in cement-based materials. Finally, leaching tests were carried out to evaluate the environmental impact of the two TFAs. The elements which were less efficiently stabilized by process A were zinc, cadmium and antimony but, when the results of the leaching tests were compared with the thresholds of the European landfill directive, TFA-A could nevertheless be accepted at landfills for non

  13. Utilization of municipal solid waste incineration (MSWI) fly ash in blended cement

    Aubert, J.E. [Laboratoire Materiaux et Durabilite des Constructions (L.M.D.C.), INSA-UPS, 135 avenue de Rangueil, 31077 Toulouse cedex 4 (France)]. E-mail: aubert@insa-toulouse.fr; Husson, B. [Laboratoire Materiaux et Durabilite des Constructions (L.M.D.C.), INSA-UPS, 135 avenue de Rangueil, 31077 Toulouse cedex 4 (France); Sarramone, N. [Laboratoire Materiaux et Durabilite des Constructions (L.M.D.C.), INSA-UPS, 135 avenue de Rangueil, 31077 Toulouse cedex 4 (France)

    2006-08-25

    This paper is the first of a series of two articles dealing with the processes applied to MSWI fly ash with a view to reusing it safely in cement-based materials. Part 1 presents two stabilization processes and Part 2 deals with the use of the two treated fly ashes (TFA) in mortars. Two types of binder were used: an Ordinary Portland Cement (OPC) containing more than 95% clinker (CEM I 52.5R) and a binary blend cement composed of 70% ground granulated blast furnace slag and 30% clinker (CEM III-B 42.5N). In this first part, two stabilization processes are presented: the conventional process, called 'A', based on the washing, phosphation and calcination of the ash, and a modified process, called 'B', intended to eliminate metallic aluminum and sulfate contained in the ash. The physical, chemical and mineralogical characteristics of the two TFA were comparable. The main differences observed were those expected, i.e. TFA-B was free of metallic aluminum and sulfate. The mineralogical characterization of the two TFAs highlighted the presence of large amounts of a calcium aluminosilicate phase taking two forms, a crystalline form (gehlenite) and an amorphous form. Hydration studies on pastes containing mixed TFA and calcium hydroxide showed that this phase reacted with calcium hydroxide to form calcium aluminate hydrates. This formation of hydrates was accompanied by a hardening of the pastes. These results are very encouraging for the reuse of such TFA in cement-based materials because they can be considered as pozzolanic additions and could advantageously replace a part of the cement in cement-based materials. Finally, leaching tests were carried out to evaluate the environmental impact of the two TFAs. The elements which were less efficiently stabilized by process A were zinc, cadmium and antimony but, when the results of the leaching tests were compared with the thresholds of the European landfill directive, TFA-A could nevertheless be accepted at

  14. Study of incorporation of casting wastes in soil-cement blocks

    Folmann, R.; Malkowski, W.; Valentina, L.V.D.; Folgureras, M.V.

    2011-01-01

    Soil-cement bricks are an ecological alternative to conventional ones because they don't need to be burned to gain strength. This study examines the technical possibility of reuse of foundry sand as fine aggregate in soil-cement. The raw materials were characterized by X-ray diffraction, thermal analysis and particle size. The foundry sand wastes were mixed at different levels to soil and cement. Compaction tests were performed to define the mixtures at optimum moisture content corresponding to the maximum density. Isothermal Conduction Calorimeter was performed to study the influence of waste in the cement hydration reaction. The results indicate that the addition of foundry sand waste decreases the optimum moisture content and increases the maximum density without significantly affecting the cement set. (author)

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

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

    2011-01-01

    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

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

    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

  17. Pelletized waste form demonstration program, October 1980-March 1981

    Lewis, E.L.; Herbert, R.F. Jr.

    1981-01-01

    During the last six months, performance testing of waste/cement pellets was continued. These evaluations included leachability tests and compressive strength tests of cold soil/cement pellets of various compositions. Fractional leach rates (g/cm 2 /day) after 21 months of testing were, in all cases -5 g/cm 2 /day (Mound Acceptance Value). Based upon these recent data, the pressed waste/cement pellets appeared to be a suitable matrix for the immobilization of low-level transuranic wastes. The installation of the Carver custom pellet press was completed. Plutonium-238 contaminated (< 100 nCi/g) ash/cement pellets were produced at a rate of 360 pellets/hr. Pellets of two different compositions were produced, 50% ash/50% cement and 65% ash/35% cement. The compressive strength of sample pellets was slightly lower than expected. Static MCC-1 leachability testing as well as long-term radiolysis testing of sample pellets are scheduled

  18. Development of solid radionuclide waste forms in the United States

    Crandall, J.L.

    1979-01-01

    New ways of reworking the wastes require a new classification in terms of the final waste forms. This paper surveys the candidate forms: encapsulation binders, in-place solidification waste forms, glass and ceramic waste forms, mineral waste forms, matrix waste forms, gaseous waste forms (fixation), and canisters and engineered barriers. Participants in the US-high-level waste form development program are listed. Requirements and selection of waste forms are also discussed. 26 references

  19. Analysis of Cement-Based Pastes Mixed with Waste Tire Rubber

    Sola, O. C.; Ozyazgan, C.; Sayin, B.

    2017-03-01

    Using the methods of thermal gravimetry, differential thermal analysis, Furier transform infrared analysis, and capillary absorption, the properties of a cement composite produced by introducing waste tyre rubber into a cement mixture were investigated. It was found that the composite filled with the rubber had a much lower water absorption ability than the unfilled one.

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

    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

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

    Naceri, Abdelghani; Hamina, Makhloufi Chikouche

    2009-08-01

    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.

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

    Naceri, Abdelghani; Hamina, Makhloufi Chikouche

    2009-01-01

    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.

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

    El-Dakrory, A.M.; Sayed, M.S.; Adham, K.

    1999-01-01

    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

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

    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.

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

    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.

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

    Sikora, Pawel; Augustyniak, Adrian; Cendrowski, Krzysztof; Horszczaruk, Elzbieta; Rucinska, Teresa; Nawrotek, Pawel; Mijowska, Ewa

    2016-01-01

    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. PMID:28773823

  7. Valorisation of waste plastic bags in cement-mortar composites as ...

    2015-01-07

    Jan 7, 2015 ... Keywords: Waste plastic bags, cement-plastic-mortar composite, aggregates coating ..... and closely attached to the aggregate by physical bonds and ... transformation steps, known as fusing material behaviour. In fact,.

  8. Sustainability of cement kiln co-processing of wastes in India: a pilot study.

    Baidya, Rahul; Ghosh, Sadhan Kumar; Parlikar, Ulhas V

    2017-07-01

    Co-processing in cement kiln achieves effective utilization of the material and energy value present in the wastes, thereby conserving the natural resources by reducing the use of virgin material. In India, a number of multifolded initiatives have been taken that take into account the potential and volume of waste generation. This paper studies the factors which might influence the sustainability of co-processing of waste in cement kilns as a business model, considering the issues and challenges in the supply chain framework in India in view of the four canonical pillars of sustainability. A pilot study on co-processing was carried out in one of the cement plant in India to evaluate the environmental performance, economical performance, operational performance and social performance. The findings will help India and other developing countries to introduce effective supply chain management for co-processing while addressing the issues and challenges during co-processing of different waste streams in the cement kilns.

  9. Analysis by X-Ray images of wind blandes waste incorporated in Portland cement

    Marques, M.A.

    2011-01-01

    The wind blandes wastes can be reused in the incorporation in Portland cement, to be used in non-structural constructions. This work shows 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 mass relation of sand, pebbles and cement was 3:2:1 and 10%, 20% and 50% of waste was incorporated in cement specimens. Frontal and upper projections were obtained in X-Rays images. 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)

  10. Influence of chemical composition of civil construction waste in the cement paste

    Cunha, G.A.; Andrade, A.C.D.; Souza, J.M.M.; Evangelista, A.C.J.; Almeida, V.C.

    2009-01-01

    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)

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

    Gilliam, T.M.; McDaniel, E.W.; Dole, L.R.; Friedman, H.A.; Loflin, J.A.; Mattus, A.J.; Morgan, I.L.; Tallent, O.K.; West, G.A.

    1987-09-01

    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. Hanford Waste Vitrification Plant Project Waste Form Qualification Program Plan

    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

  13. Pore solution chemistry of simulated low-level liquid waste incorporated in cement grouts

    Kruger, A.A.

    1995-12-01

    Expressed pore solutions from simulated low level liquid waste cement grouts cured at room temperature, 50 degree C and 90 degree C for various duration were analyzed by standard chemical methods and ion chromatography. The solid portions of the grouts were formulated with portland cement, fly ash, slag, and attapulgite clay in the ratios of 3:3:3:1. Two different solutions simulating off-gas condensates expected from vitrification of Hanford low level tank wastes were made. One is highly alkaline and contains the species Na + , P0 4 3- , N0 2 - , NO 3 - and OH - . The other is carbonated and contains the species, Na + , PO 4 3- , NO 2 - , NO 3 - , and CO 3 2- . In both cases phosphate rapidly disappeared from the pore solution, leaving behind sodium in the form of hydroxide. The carbonates were also removed from the pore solution to form calcium carbonate and possibly calcium monocarboaluminate. These reactions resulted in the increase of hydroxide ion concentration in the early period. Subsequently there was a significant reduction OH - and Na + ion concentrations. In contrast high concentration of N0 2 - and N0 3 - were retained in the pore solution indefinitely

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

    Glasser, F.P.; Atkins, M.; Kindness, A.

    1991-01-01

    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)

  15. A generalized definition for waste form durability

    Fanning, T. H.; Bauer, T. H.; Morris, E. E.; Wigeland, R. A.

    2002-01-01

    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

  16. Cement plate slab production with the incorporation of glass wool waste ground

    Nathany Angélica dos Santos

    Full Text Available Abstract One of the characteristics of Civil Construction is its ability to absorb a wide range of wastes to produce new products. Therefore, the objective of this study was to evaluate the behavior of cement plates produced with the addition of ground glass wool waste and in natura, in shredded form. The tensile strength tests in bending and permeability were performed, according to ABNT NBR 15498: 2007 as well as environmental ones according to ABNT NBR 10004: 2004. The plates produced in this study were characterized as being impermeable, with a satisfactory tensile strength in bending and being classified as plates for indoor and outdoor use, and as products that do not offer immediate risk to health and to the environment; however, it should be discarded in landfill.

  17. Liquid secondary waste: Waste form formulation and qualification

    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); Nichols, R. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-07-31

    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, including Direct Feed Low Activity Waste (DFLAW) vitrification, 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. The powdered salt waste form produced by the ETF will be replaced by a stabilized solidified waste form for disposal in Hanford’s Integrated Disposal Facility (IDF). 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 IDF. Waste form testing to support this plan is composed of work in the near term to provide data as input to a performance assessment (PA) for Hanford’s IDF. In 2015, three Hanford Liquid Secondary Waste simulants were developed based on existing and projected waste streams. Using these waste simulants, fourteen mixes of Hanford Liquid Secondary Waste were prepared and tested varying the waste simulant, the water-to-dry materials ratio, and the dry materials blend composition.1 In FY16, testing was performed using a simulant of the EMF process condensate blended with the caustic scrubber—from the Low Activity Waste (LAW) melter—, processed through the ETF. The initial EMF-16 simulant will be based on modeling efforts performed to determine the mass balance of the ETF for the DFLAW.2 The compressive strength of all of the mixes exceeded the target of 3.4 MPa (500 psi) to meet the requirements identified as potential IDF Waste Acceptance Criteria in Table 1 of the Secondary Liquid Waste Immobilization Technology Development Plan.3 The hydraulic properties of the waste forms tested (hydraulic conductivity

  18. Miscellaneous Waste-Form FEPs

    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.

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

    Mattus, C.H.

    2001-01-01

    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)

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

    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

  1. Liquid secondary waste. Waste form formulation and qualification

    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.

  2. Influence of industrial solid waste addition on properties of soil-cement bricks

    F. B. Siqueira

    Full Text Available Abstract The reuse of pollutant solid wastes produced in distinct industrial activities (avian eggshell waste and welding flux slag waste as a source of alternative raw material for producing soil-cement bricks for civil construction was investigated. Soil-cement bricks containing up to 30 wt% of industrial solid waste were uniaxially pressed and cured for 28 days. Special emphasis is given on the influence of solid waste addition on the technical properties (as such volumetric shrinkage, water absorption, bulk density, durability, and compressive strength, microstructure and mineral phases of soil-cement bricks. Microstructural evolution was evaluated via confocal microscopy. The experimental results showed that the solid wastes behave as charge material and influenced both technical properties and microstructure of the soil-cement bricks. It was found that up to 15 wt% of welding flux slag waste and up to 30 wt% of avian eggshell waste could be added into the soil-cement bricks for use as building material.

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

    Stockman, H.W.; Nagy, K.L.; Morris, C.E.

    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

  4. Combined Waste Form Cost Trade Study

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

    2008-01-01

    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

  5. Utilization of the national Portland cement for immobilizing radioactive wastes - Physical characteristics

    Rzyski, B.M.; Suarez, A.A.

    1988-01-01

    This paper shows the results obtained in the study of the national Portland cement, P320, as matrix for radioactive nitric waste incorporation. Cement use practice in other countries is common for this purposes and demonstrates to be cheap and accessible when low and medium level wastes are immobilized. Some of physical characteristics as: homogeneity,mechanical strenght, setting and porosity are analysed due to water-cement ratio and salt contents. Those characteristics which are proper of the final product, must be controlled in such way to assure a long time integrity of the wasteform. The establishment of process and quality control criteria are based in such kind of data. (author) [pt

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

    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.

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

    O'Holleran, T.O.; Feng, X.; Kalb, P.

    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 large volume casting cementation process for intermediate level radioactive waste

    Chen Zhuying; Chen Baisong; Zeng Jishu; Yu Chengze

    1988-01-01

    The recent tendency of radioactive waste treatment and disposal both in China and abroad is reviewed. The feasibility of the large volume casting cementation process for treating and disposing the intermediate level radioactive waste from spent fuel reprocessing plant in shallow land is assessed on the basis of the analyses of the experimental results (such as formulation study, solidified radioactive waste properties measurement ect.). It can be concluded large volume casting cementation process is a promising, safe and economic process. It is feasible to dispose the intermediate level radioactive waste from reprocessing plant it the disposal site chosen has resonable geological and geographical conditions and some additional effective protection means are taken

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

    Trezza, M.A.; Ferraiuelo, M.F.

    2003-01-01

    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

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

    McConnell, J.W. Jr.; Rogers, R.D.; Jastrow, J.D.; Sanford, W.E.; Sullivan, T.M.

    1995-01-01

    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 nuclear power station were solidified into waste forms using Portland cement and vinyl ester-styrene. These waste forms are being tested to develop a low-level waste data base and to obtain information on survivability of waste forms in a disposal environment. This paper reviews radionuclide releases from those waste forms in the first 9 years of sampling. Included is a discussion of the recently discovered upward migration of radionuclides. Also, lysimeter data are applied to a performance assessment source term model, and initial results are presented

  11. Using cement, lignite fly ash and baghouse filter waste for solidification of chromium electroplating treatment sludge

    Wantawin, C.

    2004-02-01

    Full Text Available The objective of the study is to use baghouse filter waste as a binder mixed with cement and lignite fly ash to solidify sludge from chromium electroplating wastewater treatment. To save cost of solidification, reducing cement in binder and increasing sludge in the cube were focused on. Minimum percent cement in binder of 20 for solidification of chromium sludge was found when controlling lignite fly ash to baghouse filter waste at the ratio of 30:70, sludge to binder ratio of 0.5, water to mixer ratio of 0.3 and curing time of 7 days. Increase of sludge to binder ratio from 0.5 to 0.75 and 1 resulted in increase in the minimum percent cement in binder up to 30 percent in both ratios. With the minimum percent cement in binder, the calculated cement to sludge ratios for samples with sludge to binder ratios of 0.5, 0.75 and 1 were 0.4, 0.4 and 0.3 respectively. Leaching chromium and compressive strength of the samples with these ratios could achieve the solidified waste standard by the Ministry of Industry. For solidification of chromium sludge at sludge to binder ratio of 1, the lowest cost binder ratio of cement to lignite fly ash and baghouse filter waste in this study was 30:21:49. The cost of binder in this ratio was 718 baht per ton dry sludge.

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

    Mayberry, J.L.; DeWitt, L.M.; Darnell, R.

    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. Volume 1

    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.

  14. Results of field testing of waste forms using lysimeters

    McConnell, J.W. Jr.; Rogers, R.D.

    1988-01-01

    The purpose of the field testing task, using lysimeter arrays, is to expose samples of solidified resin waste to the actual physical, chemical, and microbiological conditions of disposal enviroment. Wastes used in the experiment include a mixture of synthetic organic ion exchange resins and a mixture of organic exchange resins and an inorganic zeolite. Solidification agents used to produce the 4.8-by 7.6-cm cylindrical waste forms used in the study were Portland Type I-II cement and Dow vinyl ester-styrene. Seven of these waste forms were stacked end-to-end and inserted into each lysimeter to provide a 1-L volume. There are 10 lysimeters, 5 at ORNL and 5 at ANL-E. Lysimeters used in this study were designed to be self-contained units which will be disposed at the termination of the 20-year study. Each is a 0.91-by 3.12-m right-circular cylinder divided into an upper compartment, which contains fill material, waste forms, and instrumentation, and an empty lower compartment, which collects leachate. Four lysimeters at each site are filled with soil, while a fifth (used as a control) is filled with inert silica oxide sand. Instrumentation within each lysimeter includes porous cup soil-water samplers and soil moisture/temperature probes. The probes are connected to an on-site data acquisition and storage system (DAS) which also collects data from a field meteorological station located at each site. 9 refs

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

    Butler, L.G.

    1999-01-01

    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 29 Si MAS NMR spectroscopy, which shows a reduction in chain silicates with added toluene. Also, 2 H NMR of d 8 -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)

  16. Final waste classification and waste form technical position papers

    1983-05-01

    The waste classification technical position paper describes overall procedures acceptable to NRC staff which may be used by licensees to determine the presence and concentrations of the radionuclides listed in section 61.55, and thereby classifying waste for near-surface disposal. This technical position paper also provides guidance on the types of information which should be included in shipment manifests accompanying waste shipments to near-surface disposal facilities. The technical position paper on waste form provides guidance to waste generators on test methods and results acceptable to NRC staff for implementing the 10 CFR Part 61 waste form requirements. It can be used as an acceptable approach for demonstrating compliance with the 10 CFR Part 61 waste structural stability criteria. This technical position paper includes guidance on processing waste into an acceptable stable form, designing acceptable high-integrity containers, packaging cartridge filters, and minimizing radiation effects on organic ion-exchange resins. The guidance in the waste form technical position paper may be used by licensees as the basis for qualifying process control programs to meet the waste form stability requirements, including tests which can be used to demonstrate resistance to degradation arising from the effects of compression, moisture, microbial activity, radiation, and chemical changes. Generic test data (e.g., topical reports prepared by vendors who market solidification technology) may be used for process control program qualification where such generic data is applicable to the particular types of waste generated by a licensee

  17. Cementation of wastes with boric acid; Cimentacao de rejeitos contendo acido borico

    Tello, Cledola C.O.; Haucz, Maria Judite A. [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil); Alves, Lilian J.L.; Oliveira, Arno H. [Minas Gerais Univ., Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

    2000-07-01

    In nuclear power plants (PWR) are generated wastes, such as concentrate, which comes from the evaporation of liquid radioactive wastes, and spent resins. Both have boron in their composition. The cementation process is one of the options to solidify these wastes, but the boron has a negative effect on the setting of the cement mixture. In this paper are presented the experiments that are being carried out in order to overcome this problem and also to improve the efficiency of the process. Simulated wastes were cemented using additives (clays, admixtures etc.). In the process and product is being evaluated the effect of the amount, type and addition order of the materials. The mixtures were selected in accordance with their workability and incorporated waste. The solidified products are monolithic without free water with a good mechanical resistance. (author)

  18. Physical and mechanical properties of cement mortar made with brick waste

    Mohamed Guendouz

    2018-01-01

    Full Text Available The development of new building materials is a current problem where researchers are trying to find the right materials for each region and returning cheapest countries. Recycling and recovery of waste are now considered as a promising solution to meet the deficit between production and consumption and protecting the environment. This work focused on the study of the effect of brick waste in the production of cement mortar with substitution rates ranging from 5-30% by weight of cement and to compare its performance with fresh and hardened state with ordinary mortar considered as control mortar. Compressive and tensile strengths up to 28 days of age were compared with those of controlled mortar. Water absorption was also measured at 28 days of age. The test results indicate the beneficial effect of brick waste powder on performance of cement mortar with an optimum of 15% of cement weight substitution.

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

    Zhang, T.; Vandeperre, L.J.; Cheeseman, C.R.

    2012-01-01

    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 MgCO 3 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 H 2 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)

  20. Obtaining a sulfoaluminate belite cement by industrial waste

    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

  1. Safety evaluation of the radioactive waste-cement composites, (4)

    Matsuzuru, Hideo; Wadachi, Yoshiki; Ito, Akihiko

    1976-10-01

    The leaching behavior of 137 Cs has been studied to evaluate safety of sea and ground disposal of the cement composites. The rate depends on flow rate of the external solution, particle radius and composition of the cement composite. The rate-determining step of the leaching in the dynamic condition is the internal diffusion through the matrix cement composite. The rate in the static condition, on the other hand is controlled by external diffusion through the interface layer between solid and liquid. The cement composites containing mineral zeolite(25%) give very low leachability; the leaching fraction is 0.001 - 0.02 for the portland cement and 0.001 - 0.002 for the slag cement. (auth.)

  2. Influence of natural sorbents in immobilization of radioactive waste in cement

    Plecas, I.; Dimovic, S.

    2006-01-01

    Leach characteristics of 137 Cs and 60 Co radionuclides from spent mix bead ion exchange resins and both ordinary Portland cement and cement mixed with two kind of natural sorbents, (bentonite and clinoptilolite) have been studied using International Atomic Energy's (IAEA) standard leach method. A study is undertaken to determine the waste immobilization performance of low-level wastes in cement-natural sorbents mixtures. The solidification matrix was a standard Portland cement mixed with 290-350 (kg/m 3 ) spent mix bead exchange resins, with or without 1-10 % of bentonite or/and clinoptilolite The leaching rates from the cement-bentonite matrix as 60 Co: (1.20-9.72)x10 -5 (cm/d) and for 137 Cs: (1.00-9.22)x10 -4 (cm/d), after 300 days were measured. From the leaching data the apparent diffusivity of cobalt and cesium in cement bentonite or/and clinoptilolite matrix with a waste load of 350 (kg/m 3 ) spent mix bead exchange resins was measured as 60 Co: (1.0-5.9)x10 -6 (cm 2 /d) and for 137 Cs: (0.48-2.4)x10 -4 (cm 2 /d) after 300 days. The compressive strength of these samples is determined following the ASTM standards. These results are part of a 30-year mortar and concrete testing project which will influence the design of radioactive waste management for a future Serbian radioactive waste disposal center. (author)

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

    Eskander, S.B.; Saleh, H.M.

    2002-01-01

    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

  4. Investigating the Influence of Waste Basalt Powder on Selected Properties of Cement Paste and Mortar

    Dobiszewska, Magdalena; Beycioğlu, Ahmet

    2017-10-01

    Concrete is the most widely used man-made construction material in civil engineering applications. The consumption of cement and thus concrete, increases day by day along with the growth of urbanization and industrialization and due to new developments in construction technologies, population growing, increasing of living standard. Concrete production consumes much energy and large amounts of natural resources. It causes environmental, energy and economic losses. The most important material in concrete production is cement. Cement industry contributes to production of about 7% of all CO2 generated in the world. Every ton of cement production releases nearly one ton of CO2 to atmosphere. Thus the concrete and cement industry changes the environment appearance and influences it very much. Therefore, it has become very important for construction industry to focus on minimizing the environmental impact, reducing energy consumption and limiting CO2 emission. The need to meet these challenges has spurred an interest in the development of a blended Portland cement in which the amount of clinker is reduced and partially replaced with mineral additives - supplementary cementitious materials (SCMs). Many researchers have studied the possibility of using another mineral powder in mortar and concrete production. The addition of marble dust, basalt powder, granite or limestone powder positively affects some properties of cement mortar and concrete. This paper presents an experimental study on the properties of cement paste and mortar containing basalt powder. The basalt powder is a waste emerged from the preparation of aggregate used in asphalt mixture production. Previous studies have shown that analysed waste used as a fine aggregate replacement, has a beneficial effect on some properties of mortar and concrete, i.e. compressive strength, flexural strength and freeze resistance also. The present study shows the results of the research concerning the modification of cement

  5. Preliminary Hanford Waste Vitrification Plan Waste Form Qualification Plan

    Nelson, J.L.

    1987-09-01

    This Waste Form Qualification Plan describes the waste form qualification activities that will be followed during the design and operation of the Hanford Waste Vitrification Plant to ensure that the vitrified Hanford defense high-level wastes will meet the acceptance requirements of the candidate geologic repositories for nuclear waste. This plan is based on the defense waste processing facility requirements. The content of this plan is based on the assumption that the Hanford Waste Vitrification Plant high-level waste form will be disposed of in one of the geologic repository projects. Proposed legislation currently under consideration by Congress may change or delay the repository site selection process. The impacts of this change will be assessed as details of the new legislation become available. The Plan describes activities, schedules, and programmatic interfaces. The Waste Form Qualification Plan is updated regularly to incorporate Hanford Waste Vitrification Plant-specific waste acceptance requirements and to serve as a controlled baseline plan from which changes in related programs can be incorporated. 10 refs., 5 figs., 5 tabs

  6. Novel waste forms for HLW and ILW immobilisation

    Lee, William E.; Milestone, Neil B.; Ojovan, Michael I.; Hyatt, Neil C.; Stennett, Martin C.; Setiadi, Anthony; Zhou, Qizhi

    2006-01-01

    The complex nature and heterogeneity of legacy wastes means that a toolbox of different host systems must be developed in which to immobilize them. New zirconolite ceramic, glass composite materials and novel cement systems including calcium sulpho aluminate cements and alkali activated slags being examined in the Immobilisation Science Laboratory at the University of Sheffield are described. (authors)

  7. The transformation of waste Bakelite to replace natural fine aggregate in cement mortar

    Nopagon Usahanunth

    2017-06-01

    Full Text Available Bakelite material has been used to produce the various components for cars and consumer goods industry in Thailand. The growth of Bakelite consumption increases Bakelite waste. Bakelite waste is prohibited from disposing of direct landfilling and open burning because of the improper disposal and emission reasons. A large amount of this waste needs the large safe space of warehouse area for keeping which becomes a waste management problem. Size reduction by milling machine is helpful for waste handling and storing, however, the post-milling waste Bakelite plastic utilization shall be studied to maintain the waste storing capacity. There are some studies of the milling machine used for waste plastic size reduction. However, the particular study of milling machine application for waste size reduction and its milling waste utilization is still insufficient in Thailand. The purpose of this research is the use of waste Bakelite aggregate milling machine for Bakelite waste size reduction and use of the post-milling waste Bakelite as a fine aggregate to replace natural sand material in cement mortar. The waste Bakelite fine aggregate (WBFA was mixed in cement mortar mixture with the proportion 0% 20% 40% 60% 80% and 100% by volume for cement mortar sample preparation. The mortar sample was tested for compressive strength follow ASTM standard. The compressive test result of mortar samples will be compared between conventional mortar (0% WBFA and waste Bakelite mortar (WBM as well as comparing with the mortar standard. From an analysis of the sample test data found that the WBFA content in cement mortar mixture can predict the strength of WBM. The compressive strength of WBM at 28 days age with the fraction of WBFA is not exceeded 11.03%, and 23.08% respectively can be met the mortar standard according to the equation. The utilization of WBM to develop mortar non-structural mortar product can be usable from a technical point of view.

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

    Singh, D.; Wagh, A.S.; Jeong, S.Y.

    1996-01-01

    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

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

    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.

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

    Bai, Y.; Bao, Y.B.; Cai, X.L.; Chen, C.H.; Ye, X.C.

    2014-01-01

    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

  11. INEL studies concerning solidification of low-level waste in cement

    Mandler, J.W.

    1989-01-01

    The Idaho National Engineering Laboratory (INEL) has performed numerous studies addressing issues concerning the solidification of low-level radioactive waste in cement. These studies have been performed for both the Nuclear Regulatory Commission (NRC) and the Department of Energy (DOE). This short presentation will only outline the major topics addressed in some of these studies, present a few conclusions, and identify some of the technical concerns we have. More details of the work and pertinent results will be given in the Working Group sessions. The topics that have been addressed at the INEL which are relevant to this Workshop include (1) solidification of ion-exchange resins and evaporator waste in cement at commercial nuclear power plants, (2) leachability and compressive strength of power plant waste solidified in cement, (3) suggested guidelines for preparation of a solid waste process control program (PCP), (4) cement solidification of EPICOR-II resin wastes, and (5) performance testing of cement-solidified EPICOR-II resin wastes

  12. DWPF waste form compliance plan (Draft Revision)

    Plodinec, M.J.; Marra, S.L.

    1991-01-01

    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

  13. Investigation on the potential of waste cooking oil as a grinding aid in Portland cement.

    Li, Haoxin; Zhao, Jianfeng; Huang, Yuyan; Jiang, Zhengwu; Yang, Xiaojie; Yang, Zhenghong; Chen, Qing

    2016-12-15

    Although there are several methods for managing waste cooking oil (WCO), a significant result has not been achieved in China. A new method is required for safe WCO management that minimizes the environmental threat. In this context, this work was developed in which cement clinker and gypsum were interground with various WCOs, and their properties, such as grindability, water-cement ratio required to achieve a normal consistency, setting times, compressive strength, contents of calcium hydroxide and ettringite in the hardened paste, microstructure and economic and environmental considerations, were addressed in detail. The results show that, overall, WCO favorably improves cement grinding. WCO prolonged the cement setting times and resulted in longer setting times. Additionally, more remarkable effects were found in cements in which WCO contained more unsaturated fatty acid. WCOs increased the cement strength. However, this enhancement was rated with respect to the WCO contents and components. WCOs decreased the CH and AFt contents in the cement hardened paste. Even the AFt content at later ages was reduced when WCO was used. WCO also densify microstructure of the hardened cement paste. It is economically and environmentally feasible to use WCOs as grinding aids in the cement grinding process. These results contribute to the application of WCOs as grinding aids and to the safe management of WCO. Copyright © 2016 Elsevier Ltd. All rights reserved.

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

    Richet, C.; Galle, C.; Le Bescop, P.; Peycelon, H.; Bejaoui, S.; Tovena, I.; Pointeau, I.; L'Hostis, V.; Levera, P.

    2004-03-01

    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)

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

    Dragolici, Felicia; Lungu, Laura; Nicu, Mihaela; Rotarescu, Gheorghe; Turcanu, Corneliu

    2003-01-01

    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

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

    Badillo A, V.E.; Almazan T, M.G.; Alonso V, G.; Palacios H, J.C.

    2008-01-01

    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)

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

    Vieira, Vanessa Mota; Tello, Cledola Cassia Oliveira de

    2011-01-01

    Cementation is a very useful process to solidify radioactive wastes. Depending on the waste it can be necessary to use of chemical additives (admixtures) to improve the cementation process and its product. Admixtures are materials, other than cement, aggregate and water, that are added either before or during the mixing to alter some properties, such as workability, curing temperature range, and setting time. However there are a large variety of these materials that are frequently changed or taken out of the market. In this changeable scenario it is essential to know the commercially available materials and their characteristics. In this research the effects of chemical admixtures in the solidification process has been studied. For the tests it was prepared a solution simulating the evaporator concentrate waste, cemented by two different formulations, and three chemical admixtures from two manufacturers. The tested admixtures were accelerators, set retarders and super plasticizers. The experiments were organized by a planning factorial 23 to quantify the effects of formulations, of the admixtures, its quantity and manufacturer in properties of the paste and products. The measured parameters were the density, the viscosity and the setting time of the paste, and the product 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)

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

    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

  19. A analysis of cementation technology for liquid radioactive-waste in PWR NPPs

    Chen Liang; Chen Li; Li Junhua

    2009-01-01

    Cementation is one of the most popular solidification technology for the low-and-intermediate level liquid radioactive waste. It has been applied in all of domestic PWR NPPs. The process characteristics and operation of the cementations in the different NPPs are introduced,and the advantage and disadvantage of the cementation are analyzed in this paper. A drum and a cask are compared as a package of the solidified waste, the drum can decrease over 50% final volume of the waste, furthermore the cost for manufacture and transportation for this drum is more cheaper than the cask, but an additional shielding may be necessary for the waste with higher level radioactivity that is packed in drum. More waste can be contained if an appropriate in-drum mixer is used while secondary waste will be unavoidable if the out-drum mixing is adopted. A carriage can make it easier to decontaminate on the surface of equipment and on the floor, furthermore the carriage is more economical than a roller conveyor in manufacture and maintenance. The cementation recipe for the waste should be optimized and additive material should be as less as possible to increase the containing rate of the waste. (authors)

  20. Thermal conductivity of multibarrier waste form components

    Lokken, R.O.

    1981-01-01

    The multiple barrier concept of radioactive waste immobilization under investigation at Pacific Northwest Laboratory (PNL) uses composite waste forms which exhibit enhanced inertness through improvements in thermal stability, mechanical strength, and leachability by the use of coatings and metal matrices. Since excessive heat may be generated by radioactive decay of the waste, the thermal conductivity of the various barriers, and more importantly of the composite, becomes an important parameter in design criteria. This report presents results of thermal conductivity measurements on 21 various glass, ceramic, metal and composite materials used in multibarrier waste forms development

  1. Ceramic and glass radioactive waste forms

    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.

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

    Cunha, G.A.; Evangelista, A.C.J.; Almeida, V.C. de

    2009-01-01

    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)

  3. Carbonation-Induced Mineralogical Changes in Coal Mining Waste Blended Cement Pastes and Their Influence on Mechanical and Microporosity Properties

    Moisés Frías

    2018-04-01

    Full Text Available The worldwide pursuit of new eco-efficient pozzolans is ongoing. Kaolinite-based waste is an eco-friendly source of recycled metakaolinite, a highly pozzolanic product. In this study, a blended cement paste containing 20% activated coal waste (ACW was exposed to a 100% CO2 atmosphere at 65% RH for 7 days. The variations in its phase composition and strength were studied and compared to an OPC control. Both pastes were cured for 28 days prior to the carbonation test. Reaction kinetics were assessed using XRD, SEM/EDX, TG/DTG, FT-IR, Micro-Raman spectroscopy, pore solution pH and the cumulative carbonated fraction. The blended cement carbonated 68% faster than the control. While portlandite carbonation was the main reaction in both cements, decalcification was also observed (more intensely in the 20% ACW paste in other hydraulic calcium phases (C-S-H gel, monocarboaluminate (C4AcH12, ettringite and tetracalcium aluminate (C4AH13. The end product of this reaction was calcium carbonate, mainly in the form of calcite, although traces of aragonite and amorphous carbonate were also detected. Compressive strength values rose with accelerated carbonation time and pore size reduction in both cement pastes.

  4. Geo-environmental application of municipal solid waste incinerator ash stabilized with cement

    Davinder Singh

    2017-04-01

    Full Text Available The behavior of soluble salts contained in the municipal solid waste incinerator (MSWI ash significantly affects the strength development and hardening reaction when stabilized with cement. The present study focuses on the compaction and strength behavior of mixed specimens of cement and MSWI ash. A series of indices such as unconfined compressive strength, split tensile strength, California bearing ratio (CBR and pH value was examined. Prior to this, the specimens were cured for 7 d, 14 d, and 28 d. The test results depict that the maximum dry density (MDD decreases and the optimum moisture content (OMC increases with the addition of cement. The test results also reveal that the cement increases the strength of the mixed specimens. Thus, the combination of MSWI ash and cement can be used as a lightweight filling material in different structures like embankment and road construction.

  5. Cement stabilization of hazardous and radioactive electroplating sludge

    Langton, C.A.; Pickett, J.B.; Martin, M.L.

    1991-01-01

    Cement stabilization was evaluated for treatment of nickel and uranium in electroplating sludge at the Savannah River Site. Waste forms were prepared by pretreating the sludge and the solidifying it in a variety of cement, cement plus flyash, and cement-flyash-slag mixes. The sludge was also treated by one-step filtration-solidification. Leaching results and processing data indicate the cement solidification is an effective method of treating hazardous-low-level electroplating waste

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

    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.

  7. Stainless steel-zirconium alloy waste forms

    McDeavitt, S.M.; Abraham, D.P.; Keiser, D.D. Jr.; Park, J.Y.

    1996-01-01

    An electrometallurgical treatment process has been developed by Argonne National Laboratory to convert various types of spent nuclear fuels into stable storage forms and waste forms for repository disposal. The first application of this process will be to treat spent fuel alloys from the Experimental Breeder Reactor-II. Three distinct product streams emanate from the electrorefining process: (1) refined uranium; (2) fission products and actinides extracted from the electrolyte salt that are processed into a mineral waste form; and (3) metallic wastes left behind at the completion of the electrorefining step. The third product stream (i.e., the metal waste stream) is the subject of this paper. The metal waste stream contains components of the chopped spent fuel that are unaffected by the electrorefining process because of their electrochemically ''noble'' nature; this includes the cladding hulls, noble metal fission products (NMFP), and, in specific cases, zirconium from metal fuel alloys. The selected method for the consolidation and stabilization of the metal waste stream is melting and casting into a uniform, corrosion-resistant alloy. The waste form casting process will be carried out in a controlled-atmosphere furnace at high temperatures with a molten salt flux. Spent fuels with both stainless steel and Zircaloy cladding are being evaluated for treatment; thus, stainless steel-rich and Zircaloy-rich waste forms are being developed. Although the primary disposition option for the actinides is the mineral waste form, the concept of incorporating the TRU-bearing product into the metal waste form has enough potential to warrant investigation

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

    Vieira, Vanessa Mota; Tello, Cledola Cassia Oliveira de

    2012-01-01

    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)

  9. Advanced waste forms from spent nuclear fuel

    Ackerman, J.P.; McPheeters, C.C.

    1995-01-01

    More than one hundred spent nuclear fuel types, having an aggregate mass of more than 5000 metric tons (2700 metric tons of heavy metal), are stored by the United States Department of Energy. This paper proposes a method for converting this wide variety of fuel types into two waste forms for geologic disposal. The method is based on a molten salt electrorefining technique that was developed for conditioning the sodium-bonded, metallic fuel from the Experimental Breeder Reactor-II (EBR-II) for geologic disposal. The electrorefining method produces two stable, optionally actinide-free, high-level waste forms: an alloy formed from stainless steel, zirconium, and noble metal fission products, and a ceramic waste form containing the reactive metal fission products. Electrorefining and its accompanying head-end process are briefly described, and methods for isolating fission products and fabricating waste forms are discussed

  10. Immobilization of radioactive waste in cement-based matrices

    Glasser, F.P.; Rahman, A.A.; Macphee, D.; Angus, M.J.; Atkins, M.; Dept. of Chemistry)

    1985-01-01

    A solubility model of the system CaO-SiO 2 -H 2 O 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 U 6+ is precipitated as Ca-U-O-H 2 O phases. Lowering the internal Esub(eta) reduces U solubility. Studies of the carbonation of slag-cement blends are reported. (author)

  11. Mixed low-level waste form evaluation

    Pohl, P.I.; Cheng, Wu-Ching; Wheeler, T.; Waters, R.D.

    1997-01-01

    A scoping level evaluation of polyethylene encapsulation and vitreous waste forms for safe storage of mixed low-level waste was performed. Maximum permissible radionuclide concentrations were estimated for 15 indicator radionuclides disposed of at the Hanford and Savannah River sites with respect to protection of the groundwater and inadvertent intruder pathways. Nominal performance improvements of polyethylene and glass waste forms relative to grout are reported. These improvements in maximum permissible radionuclide concentrations depend strongly on the radionuclide of concern and pathway. Recommendations for future research include improving the current understanding of the performance of polymer waste forms, particularly macroencapsulation. To provide context to these estimates, the concentrations of radionuclides in treated DOE waste should be compared with the results of this study to determine required performance

  12. Hydroceramics, a ''new'' cementitious waste form material for U.S. defense-type reprocessing waste

    Siemer, Darryl D.

    2002-01-01

    A ''hydroceramic'' (HC) is a concrete which possesses mineralogy similar to the zeolitized rock indigenous to the USA's current ''basis'' high level radioactive waste (HLW) repository site, Yucca Mountain (YM). It is made by curing a mixture of inorganic waste, calcined clay, vermiculite, Na 2 S, NaOH, plus water under hydrothermal conditions. The product differs from conventional Portland cement and/or slag-based concretes (''grouts'') in that it is primarily comprised of alkali aluminosilicate ''cage minerals'' (cancrinites, sodalites, and zeolites)rather than hydrated calcium silicates (C-S-H in cement-chemistry shorthand). Consequently it microencapsulates individual salt molecules thereby rendering them less leachable than they are from conventional grouts. A fundamental difference between the formulations of HCs and radwaste-type glasses is that the latter contain insufficient aluminum to form insoluble minerals with all of the alkali metals in them. This means that the imposition of worst-case ''repository failure'' (hydrothermal) conditions would cause a substantial fraction of such glasses to alter to water-soluble forms. Since the same conditions tend to reduce the solubility of HC concretes, they constitute a more rugged immobilization sub-system. This paper compares leach characteristics of HCs with those of radwaste-type glasses and points out why hydroceramic solidification makes more sense than vitrification for US defense-type reprocessing waste. (orig.)

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

    Brown, D.J.; Smith, D.L.

    1982-06-01

    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 dm 3 scale are described. (author)

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

    Brown, D.J.; Capp, P.D.; Smith, D.L.

    1982-08-01

    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.)

  15. Iodine waste form summary report (FY 2007)

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

    2007-01-01

    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

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

    Abdel Raouf, M.W.; Husain, A.I.; El-Gammal, B.

    2005-01-01

    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 Na 2 S (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/m 2 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

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

    Howard, C.G.; Smith, D.L.G.; Williams, J.R.A.

    1986-01-01

    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 dm 3 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)

  18. Proposal of Solidification/Stabilisation Process of Chosen Hazardous Waste by Cementation

    Bozena Dohnalkova

    2015-01-01

    This paper presents a part of the project solving which is dedicated to the identification of the hazardous waste with the most critical production within the Czech Republic with the aim to study and find the optimal composition of the cement matrix that will ensure maximum content disposal of chosen hazardous waste. In the first stage of project solving – which represents this paper – a specific hazardous waste was chosen, its properties were identified and suitable soli...

  19. Actinides in intermediate-level liquid waste: removal by oxalic acid precipitation followed by cement incorporation and characterization of the final product

    Bokelund, H.; Lebrun, M.; Ougier, M.; de Caritat de Peruzzis, G.

    1991-01-01

    The purpose of this study was to investigate the conditions for the provision of an alpha free waste form (non-TRU waste with 5000) and adequate (70) DF-values were found for americium and for plutonium, respectively, with calcium as the preferred carrier. No difference between simulated and genuine ILLW was found. The final cement product was investigated by measurements of its mechanical and chemical properties. The compressive strength was evaluated as functions of the ageing time and the salt content of the waste incorporated. Furthermore, the change of porosity of the product and its resistance to water leaching were tested. The study was carried out on both simulated and genuine ILLW samples. The use of microsilica as an additive to the cement gave significant improvements in the performance of the matrix: the compressive strength was increased and, more pronounced, the leachability was decreased by up to 50%. No detrimental effects of oxalates on the cement matrix were found

  20. Immobilization of radioactive waste in cement based matrices

    Glasser, F.P.; Macphee, D.; Atkins, M.; Beckley, N.; Carson, S.; McHugh, G.; Mattingley, N.J.; Naish, C.C.; Wilding, C.R.

    1988-01-01

    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)

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

    Abotsi, G.M.K.; Bostick, D.T.; Beck, D.E.

    1996-05-01

    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

  2. Stability of High-Level Waste Forms

    Besmann, Theodore M.; Vienna, John D.

    2005-09-30

    The objective of the proposed effort is to use a new approach to develop solution models of complex waste glass systems and spent fuel that are predictive with regard to composition, phase separation, and volatility. The effort will also yield thermodynamic values for waste components that are fundamentally required for corrosion models used to predict the leaching/corrosion behavior for waste glass and spent fuel material. This basic information and understanding of chemical behavior can subsequently be used directly in computational models of leaching and transport in geologic media, in designing and engineering waste forms and barrier systems, and in prediction of chemical interactions.

  3. CRYSTALLINE CERAMIC WASTE FORMS: REFERENCE FORMULATION REPORT

    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

  4. SEPARATIONS AND WASTE FORMS CAMPAIGN IMPLEMENTATION PLAN

    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.

  5. Effect of curing time on the fraction of Cs137 from cement-waste matrix

    Plecas, I.; Pavlovic, R.; Pavlovic, S.

    2003-01-01

    To assess the safety of disposal of radioactive waste material in cement, curing conditions and time of leaching radionuclides 137 Cs have been studied. Leaching tests in cement-waste matrix, were carried out in accordance with a method recommended by IAEA. Curing conditions and curing time prior to commencing the leaching test are critically important in leach studies since the extent of hydration of the cement materials determines how much hydration product develops and whether it is available to block the pore network, thereby reducing leaching. Results presented in this paper are examples of results obtained in a 10-year concrete testing project which will influence the design of the engineer trenches system for future central Yugoslav radioactive waste storing center. (orig.)

  6. X-ray diffraction of slag-based sodium salt waste forms

    Langton, C. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Missimer, D. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2014-09-30

    The attached report documents sample preparation and x-ray diffraction results for a series of cement and blended cement matrices prepared with either water or a 4.4 M Na salt solution. The objective of the study was to provide initial phase characterization for the Cementitious Barriers Partnership reference case cementitious salt waste form. This information can be used to: 1) generate a base line for the evolution of the waste form as a function of time and conditions, 2) potentially to design new binders based on mineralogy of the binder, 3) understand and predict anion and cation leaching behavior of contaminants of concern, and 4) predict performance of the waste forms for which phase solubility and thermodynamic data are available.

  7. Chemical compatibility of DWPF canistered waste forms

    Harbour, J.R.

    1993-01-01

    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. Comparison of bitumen and cement immobilization of intermediate- and low-level radioactive waste

    Voss, J.W.

    1979-01-01

    This paper discusses a systems comparison of two available immobilization processes for intermediate- and low-level radioactive wastes -- bitumen and cement. This study examines a conceptual coprocessed UO 2 - PuO 2 fuel cycle. Radioactive wastes are generated at each stage of this fuel cycle. This study focuses on these transuranic (TRU) wastes generated at a conceptual Fuel Coprocessing Facility. In this report, these wastes are quantified, the immobilization systems conceptualized to process these wastes are presented, and a comparison of the systems is made

  9. Processability analysis of candidate waste forms

    Gould, T.H. Jr.; Dunson, J.B. Jr.; Eisenberg, A.M.; Haight, H.G. Jr.; Mello, V.E.; Schuyler, R.L. III.

    1982-01-01

    A quantitative merit evaluation, or processability analysis, was performed to assess the relative difficulty of remote processing of Savannah River Plant high-level wastes for seven alternative waste form candidates. The reference borosilicate glass process was rated as the simplest, followed by FUETAP concrete, glass marbles in a lead matrix, high-silica glass, crystalline ceramics (SYNROC-D and tailored ceramics), and coated ceramic particles. Cost estimates for the borosilicate glass, high-silica glass, and ceramic waste form processing facilities are also reported

  10. Igneous Intrusion Impacts on Waste Packages and Waste Forms

    P. Bernot

    2004-01-01

    The purpose of this model report is to assess the potential impacts of igneous intrusion on waste packages and waste forms in the emplacement drifts at the Yucca Mountain Repository. The model is based on conceptual models and includes an assessment of deleterious dynamic, thermal, hydrologic, and chemical impacts. This constitutes the waste package and waste form impacts submodel of the Total System Performance Assessment for the License Application (TSPA-LA) model assessing the impacts of a hypothetical igneous intrusion event on the repository total system performance. This submodel is carried out in accordance with Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of SR and LA (BSC 2003a) and Total System Performance Assessment-License Application Methods and Approaches (BSC 2002a). The technical work plan is governed by the procedures of AP-SIII.10Q, Models. Any deviations from the technical work plan are documented in the TSPA-LA approach to implementing the models for waste package and waste form response during igneous intrusion is based on identification of damage zones. Zone 1 includes all emplacement drifts intruded by the basalt dike, and Zone 2 includes all other emplacement drifts in the repository that are not in Zone 1. This model report will document the following model: (1) Impacts of magma intrusion on the components of engineered barrier system (e.g., drip shields and cladding) of emplacement drifts in Zone 1, and the fate of waste forms. (2) Impacts of conducting magma heat and diffusing magma gases on the drip shields, waste packages, and cladding in the Zone 2 emplacement drifts adjacent to the intruded drifts. (3) Impacts of intrusion on Zone 1 in-drift thermal and geochemical environments, including seepage hydrochemistry. The scope of this model only includes impacts to the components stated above, and does not include impacts to other engineered barrier system (EBS) components such as the invert and

  11. Igneous Intrusion Impacts on Waste Packages and Waste Forms

    P. Bernot

    2004-08-16

    The purpose of this model report is to assess the potential impacts of igneous intrusion on waste packages and waste forms in the emplacement drifts at the Yucca Mountain Repository. The model is based on conceptual models and includes an assessment of deleterious dynamic, thermal, hydrologic, and chemical impacts. This constitutes the waste package and waste form impacts submodel of the Total System Performance Assessment for the License Application (TSPA-LA) model assessing the impacts of a hypothetical igneous intrusion event on the repository total system performance. This submodel is carried out in accordance with Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of SR and LA (BSC 2003a) and Total System Performance Assessment-License Application Methods and Approaches (BSC 2002a). The technical work plan is governed by the procedures of AP-SIII.10Q, Models. Any deviations from the technical work plan are documented in the TSPA-LA approach to implementing the models for waste package and waste form response during igneous intrusion is based on identification of damage zones. Zone 1 includes all emplacement drifts intruded by the basalt dike, and Zone 2 includes all other emplacement drifts in the repository that are not in Zone 1. This model report will document the following model: (1) Impacts of magma intrusion on the components of engineered barrier system (e.g., drip shields and cladding) of emplacement drifts in Zone 1, and the fate of waste forms. (2) Impacts of conducting magma heat and diffusing magma gases on the drip shields, waste packages, and cladding in the Zone 2 emplacement drifts adjacent to the intruded drifts. (3) Impacts of intrusion on Zone 1 in-drift thermal and geochemical environments, including seepage hydrochemistry. The scope of this model only includes impacts to the components stated above, and does not include impacts to other engineered barrier system (EBS) components such as the invert and

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

    Sami, N.M.

    2013-01-01

    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 CO 2 emission. Thus, the cement industry has a crucial role in the global warming. The formation of alite (Ca 3 SiO 5 ), which is the main component of the Portland cement clinker, produces a greater amount of CO 2 emission than the formation of belite (Ca 2 SiO 4 ). The proportion of alite to belite is about 3 in ordinary Portland clinker. Therefore, by decreasing this proportion less CO 2 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 CO 2 used as raw materials and alternative hydrothermal calcination routes are employed for belite clinker production, CO 2 emission can be strongly reduced or even totally avoided. The availability of fly ash will help in reducing the CO 2 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.

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

    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...

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

    Susetyo Haria Putero; Nunung Prabaningrum; Widya Rosita

    2007-01-01

    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)

  15. Designing Advanced Ceramic Waste Forms for Electrochemical Processing Salt Waste

    Ebert, W. L.; Snyder, C. T.; Frank, Steven; Riley, Brian

    2016-01-01

    This report describes the scientific basis underlying the approach being followed to design and develop ''advanced'' glass-bonded sodalite ceramic waste form (ACWF) materials that can (1) accommodate higher salt waste loadings than the waste form developed in the 1990s for EBR-II waste salt and (2) provide greater flexibility for immobilizing extreme waste salt compositions. This is accomplished by using a binder glass having a much higher Na_2O content than glass compositions used previously to provide enough Na+ to react with all of the Cl- in the waste salt and generate the maximum amount of sodalite. The phase compositions and degradation behaviors of prototype ACWF products that were made using five new binder glass formulations and with 11-14 mass% representative LiCl/KCl-based salt waste were evaluated and compared with results of similar tests run with CWF products made using the original binder glass with 8 mass% of the same salt to demonstrate the approach and select a composition for further studies. About twice the amount of sodalite was generated in all ACWF materials and the microstructures and degradation behaviors confirmed our understanding of the reactions occurring during waste form production and the efficacy of the approach. However, the porosities of the resulting ACWF materials were higher than is desired. These results indicate the capacity of these ACWF waste forms to accommodate LiCl/KCl-based salt wastes becomes limited by porosity due to the low glass-to-sodalite volume ratio. Three of the new binder glass compositions were acceptable and there is no benefit to further increasing the Na content as initially planned. Instead, further studies are needed to develop and evaluate alternative production methods to decrease the porosity, such as by increasing the amount of binder glass in the formulation or by processing waste forms in a hot isostatic press. Increasing the amount of binder glass to eliminate porosity will decrease the waste

  16. Designing Advanced Ceramic Waste Forms for Electrochemical Processing Salt Waste

    Ebert, W. L. [Argonne National Lab. (ANL), Argonne, IL (United States); Snyder, C. T. [Argonne National Lab. (ANL), Argonne, IL (United States); Frank, Steven [Argonne National Lab. (ANL), Argonne, IL (United States); Riley, Brian [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-03-01

    This report describes the scientific basis underlying the approach being followed to design and develop “advanced” glass-bonded sodalite ceramic waste form (ACWF) materials that can (1) accommodate higher salt waste loadings than the waste form developed in the 1990s for EBR-II waste salt and (2) provide greater flexibility for immobilizing extreme waste salt compositions. This is accomplished by using a binder glass having a much higher Na2O content than glass compositions used previously to provide enough Na+ to react with all of the Cl– in the waste salt and generate the maximum amount of sodalite. The phase compositions and degradation behaviors of prototype ACWF products that were made using five new binder glass formulations and with 11-14 mass% representative LiCl/KCl-based salt waste were evaluated and compared with results of similar tests run with CWF products made using the original binder glass with 8 mass% of the same salt to demonstrate the approach and select a composition for further studies. About twice the amount of sodalite was generated in all ACWF materials and the microstructures and degradation behaviors confirmed our understanding of the reactions occurring during waste form production and the efficacy of the approach. However, the porosities of the resulting ACWF materials were higher than is desired. These results indicate the capacity of these ACWF waste forms to accommodate LiCl/KCl-based salt wastes becomes limited by porosity due to the low glass-to-sodalite volume ratio. Three of the new binder glass compositions were acceptable and there is no benefit to further increasing the Na content as initially planned. Instead, further studies are needed to develop and evaluate alternative production methods to decrease the porosity, such as by increasing the amount of binder glass in the formulation or by processing waste forms in a hot isostatic press. Increasing the amount of binder glass to eliminate porosity will decrease

  17. Corrosion studies on PREPP waste form

    Welch, J.M.; Neilson, R.M. Jr.

    1984-05-01

    Deformation or Failure Test and Accelerated Corrosion Test procedures were conducted to investigate the effect of formulation variables on the corrosion of oversize waste in Process Experimental Pilot Plant (PREPP) concrete waste forms. The Deformation or Failure Test did not indicate substantial waste form swelling from corrosion. The presence or absence of corrosion inhibitor was the most significant factor relative to measured half-cell potentials identified in the Accelerated Corrosion Test. However, corrosion inhibitor was determined to be only marginally beneficial. While this study produced no evidence that corrosion is of sufficient magnitude to produce serious degradation of PREPP waste forms, the need for corrosion rate testing is suggested. 11 references, 4 figures, 8 tables

  18. Liquid Secondary Waste Grout Formulation and Waste Form Qualification

    Um, Wooyong [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Williams, B. D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Snyder, Michelle M. V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wang, Guohui [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-05-23

    This report describes the results from liquid secondary waste (LSW) grout formulation and waste form qualification tests performed at Pacific Northwest National Laboratory (PNNL) for Washington River Protection Solutions (WRPS) to evaluate new formulations for preparing a grout waste form with high-sulfate secondary waste simulants and the release of key constituents from these grout monoliths. Specific objectives of the LSW grout formulation and waste form qualification tests described in this report focused on five activities: 1.preparing new formulations for the LSW grout waste form with high-sulfate LSW simulants and solid characterization of the cured LSW grout waste form; 2.conducting the U.S. Environmental Protection Agency (EPA) Method 1313 leach test (EPA 2012) on the grout prepared with the new formulations, which solidify sulfate-rich Hanford Tank Waste Treatment and Immobilization Plant (WTP) off-gas condensate secondary waste simulant, using deionized water (DIW); 3.conducting the EPA Method 1315 leach tests (EPA 2013) on the grout monoliths made with the new dry blend formulations and three LSW simulants (242-A evaporator condensate, Environmental Restoration Disposal Facility (ERDF) leachate, and WTP off-gas condensate) using two leachants, DIW and simulated Hanford Integrated Disposal Facility (IDF) Site vadose zone pore water (VZPW); 4.estimating the 99Tc desorption Kd (distribution coefficient) values for 99Tc transport in oxidizing conditions to support the IDF performance assessment (PA); 5.estimating the solubility of 99Tc(IV)-bearing solid phases for 99Tc transport in reducing conditions to support the IDF PA.

  19. Transuranic contaminated waste form characterization and data base

    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

  20. Implementation of radon measurements to evaluate the suitability of using cement containers for storing radioactive waste containing Ra-226

    Shweikani, R.; Kheituo, M.; Hushari, M.; Ali, A. F.

    2003-12-01

    This work aimed at studying radon diffusion through walls of cubic cement containers containing inside radioactive waste rich in Radium-226. In addition, the effect of the wall thickness on radon exhalation and external gamma exposure were also studded. Cubic cement molds were prepared with different dimensions ranged from 5 to 11 cm containing central cubic holes to contain the radioactive materials with dimensions ranged from 2 to 7 cm. The thicknesses of the walls were varied from 1 to 4 cm. Radon exhalation was studied by placing each pre-prepared cement specimen in a tightly closed glass container (desiccators, volume 7 liters) provided with input and output gas system circulation for one week. Active method (Lucas cell) was used to measure the concentration of radon in the container. It was noticed that radon concentration increased with the increase of the radioactive materials inside the specimens. This was simply explained as it is due to the increase of the amount of radium-226 in the specimen with will definitely lead to the increase of radon production. In addition, it was noticed that radon concentration were increased by increasing the thickness of the specimen wall for fixed amount of the radioactive materials inside. This result was unexpected. Therefore, many attempts were performed to explain it. For that, the mechanism of cement solidifications and structure of cement after solidification were studied. The conditions which affect the size and number of the formed pores in the specimens were also studied assuming that increasing the wall thickness will increase porosity and lead to the increase diffusion paths. It was concluded that it is possible to use the cubic cement containers to stop gamma radiation from the radioactive materials, but it is not possible to use them to stop radon unless special arrangements are performed. (author)

  1. Immobilization of radioactive waste in cement based matrices

    Glasser, F.P.; Rahman, A.A.; Macphee, D.; Atkins, M.; Beckley, N.; Lachowski, E.E.

    1986-04-01

    A mathematical and thermodynamic model of the Ca0-Si0 2 -H 2 0 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)

  2. Behaviour of intermediate-level waste forms in an aqueous environment

    Amarantos, S.; DeBatist, R.; Brodersen, K.; Glasser, F.P.; Pottier, P.E.; Vejmelka, R.; Zamorani, E.

    1985-01-01

    Under Action 1 of the Second Community Programme (1980-1984), study continued of the behavoiur of low and medium activity waste matrices using 10 reference waste forms (RWFs) representative of the main waste packages produced in the Community. The aim of this paper is to outline the main results for three types of matrix: cement and derived forms, organic polymers and bitumens. The results include data on diffusion coefficients, leach rates and waste form volume changes and mass losses. They constitute a considerable advance in knowledge of confinement properties but bring to light the need for further study of radionuclide release mechanisms for the purpose of constructing long-term models of waste form behaviour in the presence of water

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

    Singh, Satish S., E-mail: sss42@pitt.edu [Department of Chemical & Petroleum Engineering, University of Pittsburgh, Pittsburgh, PA 15261 (United States); Roy, Abhijit, E-mail: abr20@pitt.edu [Department of Bioengineering, University of Pittsburgh, Pittsburgh, PA 15261 (United States); Lee, Boeun, E-mail: bol11@pitt.edu [Department of Bioengineering, University of Pittsburgh, Pittsburgh, PA 15261 (United States); Parekh, Shrey, E-mail: smp116@pitt.edu [Department of Bioengineering, University of Pittsburgh, Pittsburgh, PA 15261 (United States); Kumta, Prashant N., E-mail: pkumta@pitt.edu [Department of Chemical & Petroleum Engineering, University of Pittsburgh, Pittsburgh, PA 15261 (United States); Department of Bioengineering, University of Pittsburgh, Pittsburgh, PA 15261 (United States); Center for Craniofacial Regeneration, University of Pittsburgh, Pittsburgh, PA 15261 (United States); Department of Mechanical Engineering and Materials Science, University of Pittsburgh, PA 15261 (United States); Center for Complex Engineered Multifunctional Materials, University of Pittsburgh, PA 15261 (United States)

    2016-06-01

    Ionic substitutions in hydroxyapatite (HA) scaffolds and self-setting cements containing Sr{sup 2+} ions incorporated are particularly of interest in bone regeneration. To date, the approach widely used to incorporate Sr{sup 2+} ions into HA cements has been the addition of Sr{sup 2+} containing salts, such as SrCO{sub 3}, SrCl{sub 2} ∙ 6H{sub 2}O, or SrHPO{sub 4}. However, this approach is dependent upon the relative solubility of Sr{sup 2+} containing salts with respect to calcium phosphate (CaP) precursors. Therefore, in the current study Sr{sup 2+} substituted dicalcium phosphate dihydrate (DCPD) was first synthesized and directly reacted with tetracalcium phosphate (TTCP) to form Sr{sup 2+} substituted HA forming cements. Rietveld refinement indicated that after one week of aging in phosphate buffered saline, cements prepared with and without Sr{sup 2+} were composed of 75% HA and 25% unreacted TTCP by weight. Cements prepared with 10% Sr{sup 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{sup 2+} content. It was concluded that both the scaffold microstructure and Sr{sup 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{sup 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{sup 2+} substituted DCPD precursors rather than individually separate Sr{sup 2+} containing salts may be a useful approach to prepare Sr{sup 2+} containing HA cements. - Highlights: • Sr{sup 2+} containing HA cement was prepared by reacting TTCP with Sr{sup 2+} substituted DCPD. • Cements with increased Sr{sup 2+} supported increased

  4. Characterizing cemented TRU waste for RCRA hazardous constituents

    Yeamans, D.R.; Betts, S.E.; Bodenstein, S.A.

    1996-01-01

    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

  5. Characterization of radioactive waste forms and packages

    1997-01-01

    This publication provides a compendium of waste form, container and waste package properties which are potential importance for waste characterization to support approval for treatment/conditioning, storage and disposal methods and for predicting both short and long term waste behaviour in the repository environment. The properties to be characterized are defined in terms of the technical rationale for their control and characterization. Characterization methods for each property are described in general with reference to detailed discussions existing in the literature. Guidance as to the advantages and disadvantages of individual methods from a technical perspective is also provided where appropriate. This report deals with the characterization of all types of radioactive wastes except spent fuel intended for direct disposal. 115 refs, 17 figs, 12 tabs

  6. Assessment of geotechnical properties of uncemented/cemented clayey soil incorporated with waste crumb rubber

    JITENDRA SINGH YADAV

    2017-03-01

    Full Text Available Discarded waste tires are becoming a serious threat to health, environment, and ecological systems worldwide if it is not disposed of properly. Every year 1000 million of waste tires are discarded. This number may grow up to 1200 million by the year 2030. On the other hand, construction of civil engineering structures such as buildings, dams, highways, etc. are at high risk of differential settlement, especially in the case of weak or soft clay, which is due to its low shear strength and high compressibility. The paper aims to assess geotechnical properties of uncemented/cemented clayey soil incorporated with waste crumb rubber. Compaction parameters, unconfined compressive strength, split tensile strength, toughness index, CBR values and swelling pressure of rubberized uncemented/cemented clayey soil (3%, and 6% cement and different percentage of crumb rubber have been obtained. After rigorous extensive study, it has been concluded that unconfined compressive strength and split tensile strength of rubberized cemented clayey soil decreases with the increase of the percentage of crumb rubber whereas the axial, and diametral strain are found to be increased with the addition of crumb rubber up to 5% after that it starts to decrease. The CBR values, swelling pressure, and toughness index of uncemented/cemented clayey soil was significantly affected by incorporation of crumb rubber. SEM studies have also been incorporated in this investigation.

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

    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.

  8. Results of the freeze resistance test, swelling index and coefficient of permeability of finegrained mining waste reinforced with cements

    Morman, Justyna

    2018-04-01

    The article presents the result of laboratory tests for mining waste with grain size of 0 to 2 mm stabilized with cement. Used for stabilization of cement CEM I 42.5 R and blast furnace cement CEM III / A 42.5N - LH / HSR / NA and a plasticizer sealant. Cement was added to the mining waste test in the proportions of 5 - 8% in relation to the skeleton's weight. For the cemented samples, the freeze resistance test, swelling index, coefficient of permeability and pH of water leachate were tested. The addition of a cement binder resulted in diminishing the water permeability of mining waste and limiting the leaching of fine particles from the material.

  9. Leaching studies of low-level radioactive waste forms

    Dayal, R.; Arora, H.; Clinton, J.C.; Milian, L.

    1985-01-01

    A research program has been under way at the Brookhaven National Laboratory to investigate the radionuclide release behavior of ion exchange bead resin waste solidified in Portland cement. An important aspect of this program is to develop and evaluate testing procedures and methodologies which enable the long-term performance evaluation of waste forms under simulated field conditions. Cesium and strontium release behavior using a range of testing procedures, including intermittent leachant flow conditions, has been investigated. For cyclic wet/dry leaching tests, extended dry periods tend to enhance the release of Cs and suppress the release of Sr. Under extended wet period leaching conditions, however, both Cs and Sr exhibit suppressed releases. In contrast, radionuclide releases observed under continuously saturated leaching conditions, as represented by conventional leaching tests, are significantly different. The relevance and aplicability of these laboratory data obtained under a wide range of leaching conditions to the performance evaluation of waste forms under anticipated field conditions is discussed. 12 refs., 9 figs., 3 tabs

  10. Construction of solid waste form test facility

    Park, Hyun Whee; Lee, Kang Moo; Koo, Jun Mo; Jung, In Ha; Lee, Jong Ryeul; Kim, Sung Whan; Bae, Sang Min; Cho, Kang Whon; Sung, Suk Jong

    1989-02-01

    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)

  11. Determining leach rates of monolithic waste forms

    Gilliam, T.M.; Dole, L.R.

    1986-01-01

    The ANS 16.1 Leach Procedure provides a conservative means of predicting long-term release from monolithic waste forms, offering a simple and relatively quick means of determining effective solid diffusion coefficients. As presented here, these coefficients can be used in a simple model to predict maximum release rates or be used in more complex site-specific models to predict actual site performance. For waste forms that pass the structural integrity test, this model also allows the prediction of EP-Tox leachate concentrations from these coefficients. Thus, the results of the ANS 16.1 Leach Procedure provide a powerful tool that can be used to predict the waste concentration limits in order to comply with the EP-Toxicity criteria for characteristically nonhazardous waste. 12 refs., 3 figs

  12. Multibarrier waste forms. Part III: Process considerations

    Lokken, R.O.

    1979-10-01

    The multibarrier concept for the solidification and storage of radioactive waste utilizes up to three barriers to isolate radionuclides from the environment: a solidified waste inner core, an impervious coating, and a metal matrix. The coating and metal matrix give the composite waste form enhanced inertness with improvements in thermal stability, mechanical strength, and leach resistance. Preliminary process flow rates and material costs were evaluated for four multibarrier waste forms with the process complexity increasing thusly: glass marbles, uncoated supercalcine, glass-coated supercalcine, and PyC/Al 2 O 3 -coated supercalcine. This report discusses the process variables and their effect on optimization of product quality, processing simplicity, and material cost. 11 figures, 2 tables

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

    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

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

    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 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

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

    Saradhi, I.V.; Mahadevan, T.N.; Krishnamoorthy, T.M.

    1999-01-01

    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

  16. IGNEOUS INTRUSION IMPACTS ON WASTE PACKAGES AND WASTE FORMS

    Bernot, P.

    2004-01-01

    The purpose of this model report is to assess the potential impacts of igneous intrusion on waste packages and waste forms in the emplacement drifts at the Yucca Mountain Repository. The models are based on conceptual models and includes an assessment of deleterious dynamic, thermal, hydrologic, and chemical impacts. The models described in this report constitute the waste package and waste form impacts submodel of the Total System Performance Assessment for the License Application (TSPA-LA) model assessing the impacts of a hypothetical igneous intrusion event on the repository total system performance. This submodel is carried out in accordance with Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of LA (BSC 2004 [DIRS:167796]) and Total System Performance Assessment-License Application Methods and Approaches (BSC 2003 [DIRS: 166296]). The technical work plan was prepared in accordance with AP-2.27Q, Planning for Science Activities. Any deviations from the technical work plan are documented in the following sections as they occur. The TSPA-LA approach to implementing the models for waste package and waste form response during igneous intrusion is based on identification of damage zones. Zone 1 includes all emplacement drifts intruded by the basalt dike, and Zone 2 includes all other emplacement drifts in the repository that are not in Zone 1. This model report will document the following model assessments: (1) Mechanical and thermal impacts of basalt magma intrusion on the invert, waste packages and waste forms of the intersected emplacement drifts of Zone 1. (2) Temperature and pressure trends of basaltic magma intrusion intersecting Zone 1 and their potential effects on waste packages and waste forms in Zone 2 emplacement drifts. (3) Deleterious volatile gases, exsolving from the intruded basalt magma and their potential effects on waste packages of Zone 2 emplacement drifts. (4) Post-intrusive physical

  17. Preparation techniques for ceramic waste form powder

    Hash, M.C.; Pereira, C.; Lewis, M.A.

    1997-01-01

    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

  18. NNWSI waste form performance test development

    Bates, J.K.; Gerding, T.J.

    1984-01-01

    A test method has been developed to measure the release of radionuclides from the waste package under simulated NNWSI repository conditions, and to provide information concerning materials interactions that may occur in the repository. Data from 13 weeks of unsaturated testing are discussed and compared to that from a 13-week analog test. The data indicate that the waste form test is capable of producing consistent, reproducible results that will be useful in evaluating the role of the waste in the long-term performance of the repository. 6 references, 3 figures

  19. Solidification of Waste Steel Foudry Dust with Portland Cement

    Škvára, F.; Kaštánek, František; Pavelková, I.; Šolcová, Olga; Maléterová, Ywetta; Schneider, Petr

    B89, č. 1 (2001), s. 67-81 ISSN 0304-3894 R&D Projects: GA ČR GA104/99/0440 Institutional research plan: CEZ:AV0Z4072921; CEZ:MSM 223100002 Keywords : solidification, * foundry dust * cement Subject RIV: CI - Industrial Chemistry, Chemical Engineering Impact factor: 0.497, year: 2001

  20. Solidification process for toxic and hazardous wastes. Second part: Cement solidification matrices

    Donato, A.; Arcuri, L.; Dotti, M.; Pace, A.; Pietrelli, L.; Ricci, G.; Basta, M.; Cali, V.; Pagliai, V.

    1989-05-01

    This paper reports the second part of a general study carried out at the Nuclear Fuel Division aiming at verifying the possible application of the radioactive waste solidification processes to industrial hazardous wastes (RTN). The cement solidification of several RTN types has been taken into consideration, both from the technical and from the economic point of view. After a short examination of the Italian juridical and economical situation in the field, which demonstrates the need of the RTN solidification, the origin and characteristics of the RTN considered in the study and directly provided by the producing industries are reviewed. The laboratory experimental results of the cementation of RTN produced by gold manufacturing industries and by galvanic industries are reported. The cementation process can be considered a very effective mean for reducing both the RTN management costs and the environmental impact of RTN disposal. (author)

  1. Partial replacement of Portland cement by red ceramic waste in mortars: study of pozzolanic activity

    Silva, A.R. da; Cabral, K.C.; Pinto, E.N. de M.G.l.

    2016-01-01

    The objective of this study is to analyze the pozzolanic activity of red ceramic residue on the partial replacement of Portland cement in mortars. The mortars were prepared by substituting 25% of the Portland cement for ground of ceramic residue with water cement’s factor of 0.48. The concrete used to construct the reference mortars and those with addiction was CPII-Z-32 (compound of Portland pozzolana cement). The chemical analysis and physical ceramic waste showed that this meets the requirements of NBR12653 (2014) for use as pozzolanic material. The pozzolanic activity index (IAP) obtained for the ceramic waste to twenty-eight days cure rate was 80.28%. (author)

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

    Darnell, G.R.

    1991-01-01

    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

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

    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.

  4. Essays of leaching in cemented products containing simulated waste from evaporator concentrated of PWR reactor

    Haucz, Maria Judite A.; Calabria, Jaqueline A. Almeida; Tello, Cledola Cassia O.; Candido, Francisco Donizete; Seles, Sandro Rogerio Novaes

    2011-01-01

    This paper evaluated the results from leaching resistance essays of cemented products, prepared from three distinct formulations, containing simulated waste of concentrated from the PWR reactor evaporator. The leaching rate is a parameter of qualification of solidified products containing radioactive waste and is determined in accordance with regulation ISO 6961. This procedure evaluates the capacity of transfer organic and inorganic substances presents in the waste through dissolution in the extractor medium. For the case of radioactive waste it is reached the more retention of contaminants in the cemented product, i.e.the lesser value of lixiviation rate. Therefore, this work evaluated among the proposed formulation that one which attend the criterion established in the regulation CNEN-NN-6.09

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

    Howard, C.G.; Lee, D.J.

    1988-01-01

    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)

  6. 40 CFR 63.1204 - What are the standards for hazardous waste burning cement kilns that are effective until...

    2010-07-01

    ... Principal Organic Hazardous Constituents (POHCs) in the waste feed that you specify under paragraph (c)(3... in the hazardous waste feed, considering the results of hazardous waste analyses or other data and... waste burning cement kilns that are effective until compliance with the standards under § 63.1220? 63...

  7. Alkali activated slag cements using waste glass as alternative activators. Rheological behaviour

    Manuel Torres-Carrasco

    2015-03-01

    The findings show that AAS paste behaviour of rheology when the activator was a commercial waterglass solution or NaOH/Na2CO3 with waste glass was similar, fit the Herschel-Bulkley model. The formation of primary C-S-H gel in both cases were confirmed. However, the rheological behaviour in standard cements fit the Bingham model. The use of the waste glass may be feasible from a rheological point of view in pastes can be used.

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

    Sams, T.L.; Gilliam, T.M.

    1987-01-01

    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

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

    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

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

    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

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

    Karellas, S.; Leontaritis, A.-D.; Panousis, G.; Bellos, E.; Kakaras, E.

    2013-01-01

    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. Electrochemical/Pyrometallurgical Waste Stream Processing and Waste Form Fabrication

    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.

  13. The immobilisation of shredded waste in a cement matrix

    James, J.M.; Smith, D.L.

    1985-10-01

    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)

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

    Duffó, Gustavo S.; Farina, Silvia B.; Schulz, Fátima M.; Marotta, Francesca

    2010-10-01

    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.

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

    Duffo, Gustavo S.; Farina, Silvia B.; Schulz, Fatima M.; Marotta, Francesca

    2010-01-01

    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. Environmental impact of incineration of calorific industrial waste: rotary kiln vs. cement kiln.

    Vermeulen, Isabel; Van Caneghem, Jo; Block, Chantal; Dewulf, Wim; Vandecasteele, Carlo

    2012-10-01

    Rotary kiln incinerators and cement kilns are two energy intensive processes, requiring high temperatures that can be obtained by the combustion of fossil fuel. In both processes, fossil fuel is often substituted by high or medium calorific waste to avoid resource depletion and to save costs. Two types of industrial calorific waste streams are considered: automotive shredder residue (ASR) and meat and bone meal (MBM). These waste streams are of current high interest: ASR must be diverted from landfill, while MBM can no longer be used for cattle feeding. The environmental impact of the incineration of these waste streams is assessed and compared for both a rotary kiln and a cement kiln. For this purpose, data from an extensive emission inventory is applied for assessing the environmental impact using two different modeling approaches: one focusing on the impact of the relevant flows to and from the process and its subsystems, the other describing the change of environmental impact in response to these physical flows. Both ways of assessing emphasize different aspects of the considered processes. Attention is paid to assumptions in the methodology that can influence the outcome and conclusions of the assessment. It is concluded that for the incineration of calorific wastes, rotary kilns are generally preferred. Nevertheless, cement kilns show opportunities in improving their environmental impact when substituting their currently used fuels by more clean calorific waste streams, if this improvement is not at the expense of the actual environmental impact. Copyright © 2012 Elsevier Ltd. All rights reserved.

  17. The role of cement to be expected in radioactive waste disposal system

    Tanaka, Satoru; Nagasaki, Shinya; Ohe, Toshiaki

    1997-01-01

    Based on the present states of cement in radioactive waste disposal system, its roles and functions to be further expected were discussed diming at safety evaluation of wastes. In the present waste disposal system, cement has two important roles as the structural materials for the system and as the barrier materials for protecting from various radiations. In order to enhance the durability of those materials, it is needed to improve them in respects of acid resistance and repression of the reactions with radioactive wastes. Generally development of cracks in concrete structures is inevitable and the repairs become necessary in old structures. Therefore, it is desirable that cement for a disposal system has self-diagnostic and self-repairing abilities to keep the efficiency for a long period. The compressive strength of ordinary high-strength concrete is around 50-60 Nmm -2 . However, it is needed to further increase the strength to decrease the amount of hardening materials and the width of concrete vessel for wastes. In addition, it is desirable to develop new techniques to recycle concrete wastes involving radioactive materials at ultra-low levels. (M.N.)

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

    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.

  19. High-level waste-form-product performance evaluation

    Bernadzikowski, T.A.; Allender, J.S.; Stone, J.A.; Gordon, D.E.; Gould, T.H. Jr.; Westberry, C.F. III.

    1982-01-01

    Seven candidate waste forms were evaluated for immobilization and geologic disposal of high-level radioactive wastes. The waste forms were compared on the basis of leach resistance, mechanical stability, and waste loading. All forms performed well at leaching temperatures of 40, 90, and 150 0 C. Ceramic forms ranked highest, followed by glasses, a metal matrix form, and concrete. 11 tables

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

    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)

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

    Jones, T.L.; Serne, R.J.; Toste, A.P.

    1988-04-01

    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

  2. Review of glass ceramic waste forms

    Rusin, J.M.

    1981-01-01

    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 (BaAl 3 Si 2 O 8 ) and a fresnoite-based (Ba 2 TiSi 2 O 8 ) glass ceramic, have been studied at Pacific Northwest Laboratory (PNL). A basalt-based glass ceramic primarily containing diopsidic augite (CaMgSi 2 O 6 ) 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 (CaMgSi 2 O 6 ). 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

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

    Eskander, S.B.; Bayoumi, T.A.; Saleh, H.M.

    2012-01-01

    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. Characterization of radioactive waste forms. Volume 1

    Brodersen, K.; Nilsson, K.

    1989-01-01

    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

  5. Characterization of radioactive waste forms. Volume 2

    Smith, D.L.; Green, T.H.

    1989-01-01

    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

  6. Formulation of criteria for pollution control on cement products produced from solid wastes in China.

    Yang, Yufei; Huang, Qifei; Yang, Yu; Huang, Zechun; Wang, Qi

    2011-08-01

    The process of producing cement products from solid waste can increase the level of pollutants in the cement products. Therefore, it is very important to establish a pollution control standard for cement products to protect the environment and human health. This paper presents acceptance limits for the availability of heavy metals in cement products which have been produced from solid wastes and explains how the limits have been calculated. The approach and method used to formulate these criteria were based on EN 12920. The typical exposure scenarios used in this paper involve concrete being used for drinking water supply pipelines and concrete pavements and are based on an analysis of typical applications of cement in China, and the potential for contact with water. The parameters of a tank test which was based on NEN 7375 were set in accordance with the environmental conditions of typical scenarios in China. Mechanisms controlling the release of heavy metals in concrete and a model for that release were obtained using the leaching test. Finally, based on acceptance criteria for drinking water and groundwater quality in China, limit values for the availability of heavy metals in concrete were calculated. Copyright © 2011 Elsevier Ltd. All rights reserved.

  7. Elution behavior of heavy metals from cement solidified products of incinerated ash waste - 59102

    Meguro, Yoshihiro; Kawato, Yoshimi; Nakayama, Takuya; Tomioka, Osamu; Mitsuda, Motoyuki

    2012-01-01

    A method, in which incinerated ash is solidified with a cement material, has been developed to dispose radioactive incinerated ash waste. In order to bury the solidified product, it is required that elution of hazardous heavy metals included in the ash from the solidified products is inhibited. In this study, the elution behavior of the heavy metals from the synthetic solidified products, which included Pb(II), Cd(II), and Cr(VI) and were prepared using ordinary portland cement (OPC), blast furnace slag cement (BFS), or a cement material that showed low alkalinity (LA-Cement), was investigated. Several chemicals and materials were added as additive agents to prevent the elution of the heavy metals. When OPC was used, Cd elution was inhibited, but Pb and Cr were not enough even using the additive agent examined. FeSO 4 and Na 2 S additive agents worked effective to inhibit elution of Cr. When BFS was used, the elution of Pb, Cd and Cr was inhibited for the all products prepared. In the case of LA-Cement, the elution of Pb and Cd was inhibited for the all products, but only the product that was added FeSO 4 showed good result of the elution of Cr. (authors)

  8. Alternative High-Performance Ceramic Waste Forms

    Sundaram, S. K. [Alfred Univ., NY (United States)

    2017-02-01

    This final report (M5NU-12-NY-AU # 0202-0410) summarizes the results of the project titled “Alternative High-Performance Ceramic Waste Forms,” funded in FY12 by the Nuclear Energy University Program (NEUP Project # 12-3809) being led by Alfred University in collaboration with Savannah River National Laboratory (SRNL). The overall focus of the project is to advance fundamental understanding of crystalline ceramic waste forms and to demonstrate their viability as alternative waste forms to borosilicate glasses. We processed single- and multiphase hollandite waste forms based on simulated waste streams compositions provided by SRNL based on the advanced fuel cycle initiative (AFCI) aqueous separation process developed in the Fuel Cycle Research and Development (FCR&D). For multiphase simulated waste forms, oxide and carbonate precursors were mixed together via ball milling with deionized water using zirconia media in a polyethylene jar for 2 h. The slurry was dried overnight and then separated from the media. The blended powders were then subjected to melting or spark plasma sintering (SPS) processes. Microstructural evolution and phase assemblages of these samples were studied using x-ray diffraction (XRD), scanning electron microscopy (SEM), energy dispersion analysis of x-rays (EDAX), wavelength dispersive spectrometry (WDS), transmission electron spectroscopy (TEM), selective area x-ray diffraction (SAXD), and electron backscatter diffraction (EBSD). These results showed that the processing methods have significant effect on the microstructure and thus the performance of these waste forms. The Ce substitution into zirconolite and pyrochlore materials was investigated using a combination of experimental (in situ XRD and x-ray absorption near edge structure (XANES)) and modeling techniques to study these single phases independently. In zirconolite materials, a transition from the 2M to the 4M polymorph was observed with increasing Ce content. The resulting

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

    Reynier, N.; Riveros, P.; Lastra, R.; Laviolette, C.; Bouzoubaa, N.; Chapman, M.

    2015-01-01

    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, CaUO 4 , 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 H 2 SO 4 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 CaSO 4 . 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

  10. PWR Users Group 10 CFR 61 Waste Form Requirements Compliance Test Program

    Rosenlof, R.C.

    1985-01-01

    In January of 1984, a PWR Users Group was formed to initiate a 10 CFR 61 Waste Form Requirements Compliance Test Program on a shared cost basis. The original Radwaste Solidification Systems sold by ATCOR ENGINEERED SYSTEMS, INC. to the utilities were required to produce a free-standing monolith with no free water. None of the other requirements of 10 CFR 61 had to be met. Current regulations, however, have substantially expanded the scope of the waste form acceptance criteria. These new criteria required that generators of radioactive waste demonstrate the ability to produce waste forms which meet certain chemical and physical requirements. This paper will present the test program used and the results obtained to insure 10 CFR 61 compliance of the three (3) typical waste streams generated by the ATCOR PWR Users Group's plants. The primary objective of the PWR Users Group was not to maximize waste loading within the masonry cement solidification media, but to insure that the users Radwaste Solidification System is capable of producing waste forms which meet the waste form criteria of 10 CFR 61. A description of the laboratory small sample certification program and the actual full scale pilot plant verification approach used is included in this paper. Also included is a discussion of the development of a Process Control Program to ensure the reproducibility of the test results with actual waste

  11. Effect of borate concentration on solidification of radioactive wastes by different cements

    Sun Qina; Li Junfeng; Wang Jianlong

    2011-01-01

    Highlights: → The effect of borate on cementation of radioactive borate evaporator concentrates by sulfoaluminate cement (SAC) and Portland cement (PC) was compared. → The X-ray diffraction (XRD) revealed that borate did not interfere with the formation of main hydration products of SAC and PC. → Borate, in the form of B(OH) 4- , incorporated in ettringite as solid solution phase. - Abstract: To investigate the effect of borate on the cementation of radioactive evaporator concentrates, and to provide more data for solidification formula optimization, the simulated borate evaporator concentrates with different borate concentrations (as B) and Na/B ratio (molar ratio) were solidified by sulfoaluminate cement (SAC) and Portland cement (PC), with addition of Ca(OH) 2 , zeolite and accelerator or water reducer. The hydration products of solidified matrices were characterized by X-ray diffraction (XRD). The experimental results showed that borate retarded the cement setting for both SAC and PC formulas, and the final setting time prolonged with decrease of Na/B ratio. Borate could enhance the fluidity of the cement mixture. The 28 d compressive strengths of the solidified matrices for both SAC and PC formulas decreased with increase of borate concentration. The XRD patterns suggested that, in the matrices maintained for 28 d, borate did not interfere with the formation of main hydration products of SAC and PC. Borate, in the form of B(OH) 4- , incorporated in ettringite (3CaO.Al 2 O 3 .3CaSO 4 .32H 2 O) as solid solution phase. The formula of SAC and PC developed in this study was effective for cementation of the simulated borate evaporator concentrates. However further optimization was required to reduce retarding effect of higher borate concentrations and to extend the practical feasibility for actual evaporator concentrates.

  12. Permeability of Consolidated Incinerator Facility Wastes Stabilized with Portland Cement

    Walker, B.W.

    1999-01-01

    The Consolidated Incinerator Facility (CIF) at the Savannah River Site (SRS) burns low-level radioactive wastes and mixed wastes as method of treatment and volume reduction. The CIF generates secondary waste, which consists of ash and off-gas scrubber solution. Currently the ash is stabilized/solidified in the Ashcrete process. The scrubber solution (blowdown) is sent to the SRS Effluent Treatment Facility (ETF) for treatment as waste water. In the past, the scrubber solution was also stabilized/solidified in the Ashcrete process as blowcrete and will continue to be treated this way for listed waste burns and scrubber solution that do not meet the Effluent Treatment Facility (ETF) Waste Acceptance Criteria (WAC)

  13. Cement encapsulation of low-level waste liquids. Final report

    Baker, M.N.; Houston, H.M.

    1999-01-01

    Pretreatment of liquid high-level radioactive waste at the West Valley Demonstration Project (WVDP) was essential to ensuring the success of high-level waste (HLW) vitrification. By chemically separating the HLW from liquid waste, it was possible to achieve a significant reduction in the volume of HLW to be vitrified. In addition, pretreatment made it possible to remove sulfates, which posed several processing problems, from the HLW before vitrification took place

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

    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.

  15. Detection of free liquid in cement-solidified radioactive waste drums using computed tomography

    Steude, J.S.; Tonner, P.D.

    1991-01-01

    Acceptance criteria for disposal of radioactive waste drums require that the cement-solidified material in the drum contain minimal free liquid after the cement has hardened. Free liquid is to be avoided because it may corrode the drum, escape and cause environmental contamination. The DOE has requested that a nondestructive evaluation method be developed to detect free liquid in quantities in excess of 0.5% by volume. This corresponds to about 1 liter in a standard 208 liter (55 gallon) drum. In this study, the detection of volumes of free liquid in a 57 cm (2 ft.) diameter cement-solidified drum is demonstrated using high-energy X-ray computed tomography (CT0. In this paper it is shown that liquid concentrations of simulated radioactive waste inside glass tubes imbedded in cement can easily be detected, even for tubes with inner diameters less than 2 mm (0.08 in.). Furthermore, it is demonstrated that tubes containing water and liquid concentrations of simulated radioactive waste can be distinguished from tubes of the same size containing air. The CT images were obtained at a rate of about 6 minutes per slice on a commercially available CT system using a 9 MeV linear accelerator source

  16. Industrial Wastes as Alternative Mineral Addition in Portland Cement and as Aggregate in Coating Mortars

    Oliveira, Kamilla Almeida; Nazário, Bruna Inácio; Oliveira, Antonio Pedro Novaes de; Hotza, Dachamir; Raupp-Pereira, Fabiano

    2017-01-01

    This paper presents an evaluation study of wastes from pulp and paper as well as construction and demolition industries for application in cement-based materials. The alternative raw materials were used as a source of calcium carbonate (CaCO3) and as pozzolanic material (water-reactive SiO2) in partial replacement of Portland cement. In addition to the hydraulic binder, coating mortars were composed by combining the pulp and paper fluidized bed sand residue with construction and demolition wa...

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

    Singh, Sarbjit; Mhatre, Amol; Sagar, Veena

    2012-05-01

    A method was developed for the estimation of 241 Am 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 241 Am. Also in such large samples, error due to the distribution of activity is more. Estimation of 241 Am in the cemented sample was carried out by applying corrections for attenuation and for the sample geometry. (author)

  18. Immobilization of strontium and cesium in intermediate-level liquid wastes by solidification in cements

    Rudolph, G.; Koester, R.

    1979-01-01

    An accelerated leach test at elevated temperature has been developed which gives intercomparable results within one day. It is very useful for product quality control at large throughputs. Using this test, it has been shown that cesium leachabilities from cement products containing a simulated waste typical of fuel reprocessing plants can be reduced by addition of a bentonite. Addition of barium silicate hydrate retards strontium leaching in these cements. Leach rates in tap water and in salt brine are lower than in distilled water and sodium chloride solution

  19. High level waste fixation in cermet form

    Kobisk, E.H.; Aaron, W.S.; Quinby, T.C.; Ramey, D.W.

    1981-01-01

    Commercial and defense high level waste fixation in cermet form is being studied by personnel of the Isotopes Research Materials Laboratory, Solid State Division (ORNL). As a corollary to earlier research and development in forming high density ceramic and cermet rods, disks, and other shapes using separated isotopes, similar chemical and physical processing methods have been applied to synthetic and real waste fixation. Generally, experimental products resulting from this approach have shown physical and chemical characteristics which are deemed suitable for long-term storage, shipping, corrosive environments, high temperature environments, high waste loading, decay heat dissipation, and radiation damage. Although leach tests are not conclusive, what little comparative data are available show cermet to withstand hydrothermal conditions in water and brine solutions. The Soxhlet leach test, using radioactive cesium as a tracer, showed that leaching of cermet was about X100 less than that of 78 to 68 glass. Using essentially uncooled, untreated waste, cermet fixation was found to accommodate up to 75% waste loading and yet, because of its high thermal conductivity, a monolith of 0.6 m diameter and 3.3 m-length would have only a maximum centerline temperature of 29 K above the ambient value

  20. Vitrified medical wastes bottom ash in cement clinkerization. Microstructural, hydration and leaching characteristics.

    Papamarkou, S; Christopoulos, D; Tsakiridis, P E; Bartzas, G; Tsakalakis, K

    2018-04-19

    The present investigation focuses on the utilization of medical wastes incineration bottom ash (MBA), vitrified with soda lime recycled glass (SLRG), as an alternative raw material in cement clinkerization. Bottom ash is recovered from the bottom of the medical wastes incineration chamber, after being cooled down through quenching. It corresponds to 10-15 wt% of the initial medical wastes weight and since it has been classified in the category of hazardous wastes, its safe management has become a major environmental concern worldwide. MBA glasses of various syntheses were initially obtained during the MBA vitrification simultaneously with various amounts of silica scrap (20, 25 and 30 wt% correspondingly). The produced MBA glasses were in turn used for the production of Portland cement clinker, after sintering at 1400 °C, thus substituting traditional raw materials. Both evaluation of vitrification and sintering products was carried out by chemical and mineralogical analyses along with microstructure examination. The final cements were prepared by clinkers co-grinding in a laboratory ball mill with appropriate amounts of gypsum (≈5.0 wt%) and the evaluation of their quality was carried out by determining setting times, standard consistency, expansibility and compressive strength at 2, 7, 28 and 90 days. Finally, the leaching behaviour of the vitrified MBA and hydrated cements, together with the corresponding of the "as received" MBA, was further examined using the standard leaching tests of the Toxicity Characteristic Leaching Procedure (TCLP) and the EN 12457-2. According to the obtained results, the quality of the produced cement clinkers was not affected by the addition of the vitrified MBA in the raw meal, with the trace elements detected in all leachates measured well below the corresponding regulatory limits. Copyright © 2018 Elsevier B.V. All rights reserved.

  1. Radionuclide migration through porous cement-waste composition in semi-real conditions

    Plecas, I.; Peric, A.; Kostadinovic, A.

    1989-01-01

    In this paper, result of examination of Leakage rate or radionuclides Co-60 and Cs-137 in semi-real conditions are given. Radionuclides Co-60 and Cs-137 were immobilized by cement process and conditioned in concrete containers trying to make similar scenario for storing radioactive waste materials as in engineering trench system, repository. Experiments were realized with two waste water, evaporator bottom and reactor cooling system, (EB) and (RCS), from Nuclear Power Plants Krsko, in which the main radionuclides are Co-60 and Cs-137. These results will be used for future Yugoslav radioactive waste storing center (author)

  2. Solidification of radioactive waste resins using cement mixed with organic material

    Laili, Zalina; Yasir, Muhamad Samudi; Wahab, Mohd Abdul

    2015-01-01

    Solidification of radioactive waste resins using cement mixed with organic material i.e. biochar is described in this paper. Different percentage of biochar (0%, 5%, 8%, 11%, 14% and 18%) was investigated in this study. The characteristics such as compressive strength and leaching behavior were examined in order to evaluate the performance of solidified radioactive waste resins. The results showed that the amount of biochar affect the compressive strength of the solidified resins. Based on the data obtained for the leaching experiments performed, only one formulation showed the leached of Cs-134 from the solidified radioactive waste resins

  3. Solidification of radioactive waste resins using cement mixed with organic material

    Laili, Zalina, E-mail: liena@nm.gov.my [Nuclear Science Programme, School of Applied Physics, Faculty of Science and Technology, Universiti Kebangsaan Malaysia (UKM), Bangi, 43600, Selangor Malaysia (Malaysia); Waste and Environmental Technology Division, Malaysian Nuclear Agency (Nuclear Malaysia), Bangi, 43000 Kajang, Selangor (Malaysia); Yasir, Muhamad Samudi [Nuclear Science Programme, School of Applied Physics, Faculty of Science and Technology, Universiti Kebangsaan Malaysia (UKM), Bangi, 43600, Selangor Malaysia (Malaysia); Wahab, Mohd Abdul [Waste and Environmental Technology Division, Malaysian Nuclear Agency (Nuclear Malaysia), Bangi, 43000 Kajang, Selangor (Malaysia)

    2015-04-29

    Solidification of radioactive waste resins using cement mixed with organic material i.e. biochar is described in this paper. Different percentage of biochar (0%, 5%, 8%, 11%, 14% and 18%) was investigated in this study. The characteristics such as compressive strength and leaching behavior were examined in order to evaluate the performance of solidified radioactive waste resins. The results showed that the amount of biochar affect the compressive strength of the solidified resins. Based on the data obtained for the leaching experiments performed, only one formulation showed the leached of Cs-134 from the solidified radioactive waste resins.

  4. Development of Concrete Paving Blocks Prepared from Waste Materials without Portland Cement

    Charin NAMARAK

    2018-02-01

    Full Text Available This experiment used three types of waste materials: calcium carbide residue, fly ash, and recycled concrete aggregate to develop concrete paving blocks. The blocks had calcium carbide residue and fly ash as a binder without ordinary Portland cement (OPC and combined with 100 % of recycled concrete aggregate. The concrete paving blocks were 10 × 10 × 20 cm and were formed using a pressure of 6 or 8 MPa. The binder-to-aggregate ratio was held constant at 1:3 by weight, while the water-to-binder ratios were 0.30, 0.35, and 0.40. The effects of the water-to-binder ratios and fineness of the binder on the compressive strength, flexural strength, abrasion resistance, and water absorption of the concrete paving blocks were determined and compared with those of TIS 827 and ASTM C1319 standards. The results revealed that by applying this procedure, we were able to produce an excellence concrete paving block without using OPC. The compressive strength of the concrete paving blocks made from these waste materials was 41.4 MPa at 28 days and increased to 45.3 MPa at 60 days. Therefore, these waste materials can be used as raw materials to manufacture concrete paving blocks without OPC that meet the requirements of 40 MPa and 35 MPa specified by the TIS 827 and ASTM C1319 standards, respectively.DOI: http://dx.doi.org/10.5755/j01.ms.24.1.17566

  5. Impact test for solid waste forms

    Wallace, R.M.; Kelley, J.A.

    1976-03-01

    Samples of concretes and glasses being considered for incorporation of radioactive waste sludge were subjected to impact tests to determine the relationship between the energy of the impact and the resulting increase in surface area of the damaged sample. Test results indicate that the increased surface area per unit of energy input for glass waste forms is less by a factor of about three than that for concretes containing 40 wt percent simulated sludge (average values of 9.6 cm 2 /Joule and 24.7 cm 2 /Joule for glass and concrete, respectively)

  6. Use of the “red gypsum” industrial waste as substitute of natural gypsum for commercial cements manufacturing

    Gázquez, M. J.

    2012-06-01

    Full Text Available The main objective of this research has been the valorisation of a waste from the TiO2 production process (sulphate method, called red gypsum, in the production of cements. This waste is mainly formed by di-hydrate calcium sulphate and iron hydroxides. To cover this objective it has been necessary to perform the physico-chemical characterisation of the red gypsum as well as the main components in the production of cements and of the new cements generated. Moreover, for the red gypsum, has been analyzed its radioactive content because it is generated in a NORM (Naturally Occurring Radioactive Materials industry. Finally, the most important properties of the obtained cements with different proportions of red gypsum in their composition have been studied by comparing them with the standard ones obtained in a Portland cement. Lastly, we have demonstrated that the new cements fulfil all the quality tests imposed by the European legislation.

    El objetivo de esta investigación ha sido analizar la valorización de un residuo generado en el proceso de producción de dióxido de titanio (vía sulfato, denominado yeso rojo, en la producción de cementos. Dicho residuo está compuesto fundamentalmente por sulfato de calcio di-hidratado e hidróxidos de hierro. Para ello, ha sido necesaria la caracterización físico-química del yeso rojo, así como la de los otros componentes fundamentales en la fabricación de cementos y de los cementos generados con el mencionado residuo. Además, en el caso del yeso rojo, se ha analizado su contenido radiactivo al generarse éste en una industria NORM (Natural Occurring Radioactive Materials. Posteriormente, se han estudiado las propiedades más importantes de los cementos producidos con diferentes porcentajes de yeso rojo añadido, comparando estas mezclas con las propiedades de un cemento Portland comercial, comprobándose que se cumplen todas las normas Europeas de calidad exigibles.

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

    Angus, M.; Borwick, J.; Cann, G.; Hayes, M.; McLuckie, B.; Jowsey, J.

    2012-01-01

    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)

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

    Beijer, O.

    1980-01-01

    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 cm 2 /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)

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

    Watson, Claire; Benbow, Steven; Savage, David

    2007-05-01

    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

  10. Preparation and leaching of radioactive INEL waste forms

    Schuman, R.P.; Welch, J.M.; Staples, B.A.

    1982-01-01

    The purpose of this study is to prepare and leach test ceramic and glass waste form specimens produced from actual transuranic waste sludges and high-level waste calcines, respectively. Description of wastes, specimen fabrication, leaching procedure, analysis of leachates and results are discussed. The conclusion is that radioactive waste stored at INEL can be readily incorporated in fused ceramic and glass forms. Initial leach testing results indicate that these forms show great promise for safe long-term containment of radioactive wastes

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

    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

  12. Waste Form Features, Events, and Processes

    R. Schreiner

    2004-01-01

    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

    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. Investigation of Waste Paper Cellulosic Fibers Utilization into Cement Based Building Materials

    Viola Hospodarova

    2018-03-01

    Full Text Available Recently, the utilization of renewable natural cellulosic materials, such as wood, plants, and waste paper in the preparation of building materials has attracted significant interest. This is due to their advantageous properties, low environmental impact and low cost. The objective of this paper is to investigate the influence of recycled cellulosic fibers (in the amount 0.5 wt % of the filler and binder weight and superplasticizer (in the amount 0.5 wt % of the cement weight on the resulting properties of cement composites (consistency of fresh mixture, density, thermal conductivity, and compressive and flexural strength for hardening times of 1, 3, 7, 28, and 90 days. Plasticizer use improved the workability of fresh cement mixture. In comparison to the reference sample, the results revealed a decrease in density of 6.8% and in the thermal conductivity of composites with cellulosic fibers of 34%. The highest values of compressive (48.4 MPa and flexural (up to 7 MPa strength were achieved for hardened fiber cement specimens with plasticizer due to their significantly better dispersion of cement particles and improved bond strength between fibers and matrix.

  15. Modelling of the degradation of cement in a nuclear waste repository

    Haworth, A.; Sharland, S.M.; Tweed, C.J.

    1989-01-01

    The current UK concept for a low- or intermediate-level nuclear waste repository includes a largely cementitious backfill. The cement provides a high pH environment in which the general corrosion rate of the metal canisters is reduced and the solubilities of many nuclides low. It has previously been assumed that this high pH will exist for a period of 10 7 years, however cement will degrade due to leaching of the solid components and attack from aqueous species in groundwater. In this paper the authors describe the preliminary stages of a model of the degradation of cement in a repository. The modelling involves the incorporation of a thermodynamic description of cement into the static code PHREEQE. This is then used in a coupled chemistry-transport model of simple leaching of cement using the code CHEQMATE. This preliminary modelling also provides a useful verification of CHEQMATE as a direct comparison with a THCCDM (a coupled code based on CHEMTRN) model is possible. Results from this preliminary model suggest that the fall in pH due to leaching is slow

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

    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.

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

    Ferreira, Eduardo G.A.; Isiki, Vera L.K.; Miyamoto, Hissae; Marumo, Julio T.; Vicente, Roberto

    2009-01-01

    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)

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

    Galicia A, E.; Badillo A, V. E.; Ramirez S, J. R.; Nava E, N.

    2014-10-01

    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)

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

    Sui, Xiuwen; Zhang, Yun; Shao, Shuai; Zhang, Shushen

    2014-01-01

    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 implement