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Sample records for cement waste forms

  1. Quality control of cemented waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Slate, L.J.

    1994-12-31

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

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

  9. Conditioning of radioactive waste solutions by cementation

    International Nuclear Information System (INIS)

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

  10. Cements in Radioactive Waste Disposal

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Jardine, L J

    2003-06-12

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

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

    OpenAIRE

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

    2015-01-01

    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 (DF...

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

    International Nuclear Information System (INIS)

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

  14. Stabilization of NaCl-containing cuttings wastes in cement concrete by in situ formed mineral phases.

    Science.gov (United States)

    Filippov, Lev; Thomas, Fabien; Filippova, Inna; Yvon, Jacques; Morillon-Jeanmaire, Anne

    2009-11-15

    Disposal of NaCl-containing cuttings is a major environmental concern due to the high solubility of chlorides. The present work aims at reducing the solubility of chloride by encapsulation in low permeability matrix as well as lowering its solubility by trapping into low-solubility phases. Both the studied materials were cuttings from an oil-based mud in oil drillings containing about 50% of halite, and cuttings in water-based mud from gas drilling containing 90% of halite. A reduction in the amount of dissolved salt from 41 to 19% according to normalized leaching tests was obtained by addition of potassium ortho-phosphate in the mortar formula of oil-based cuttings, while the aluminium dihydrogeno-phosphate is even more efficient for the stabilization of water-based cuttings with a NaCl content of 90%. Addition of ortho-phosphate leads to form a continuous and weakly soluble network in the cement matrix, which reduces the release of salt. The formed mineralogical phases were apatite and hydrocalumite. These phases encapsulate the salt grains within a network, thus lowering its interaction with water or/and trap chloride into low-solubility phases. The tested approaches allow to develop a confinement process of NaCl-containing waste of various compositions that can be applied to wastes, whatever the salt content and the nature of the drilling fluids (water or oil). PMID:19631465

  15. Cement encapsulation of uranyl nitrate waste

    International Nuclear Information System (INIS)

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

  16. Cement-Based Materials for Nuclear Waste Storage

    CERN Document Server

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

    2013-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-15

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

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

    Science.gov (United States)

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

    2014-09-01

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

  19. Modelling the effects of waste components on cement hydration

    NARCIS (Netherlands)

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

    2001-01-01

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

  20. Modelling the effects of waste components on cement hydration

    NARCIS (Netherlands)

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

    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 t

  1. Immobilization of radioactive waste in cement-based matrices

    International Nuclear Information System (INIS)

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

  2. Modified sulfur cement solidification of low-level wastes

    International Nuclear Information System (INIS)

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

  3. PURIFIED WASTE FCC CATALYST AS A CEMENT REPLACEMENT MATERIAL

    Directory of Open Access Journals (Sweden)

    Danute Vaiciukyniene

    2015-06-01

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

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

    International Nuclear Information System (INIS)

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

  5. The interaction between nuclear waste glass and cement

    International Nuclear Information System (INIS)

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

  6. Waste form development/test

    International Nuclear Information System (INIS)

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

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

    Directory of Open Access Journals (Sweden)

    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.

  8. Defining criteria for cemented waste produced from legacy liquids

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  11. Leaching tests of cemented organic radioactive waste

    International Nuclear Information System (INIS)

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

  12. Densified waste form and method for forming

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-25

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

  13. Densified waste form and method for forming

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-17

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

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

    OpenAIRE

    Yufei Yang; Jingchuan Xue; Qifei Huang

    2013-01-01

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

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

    OpenAIRE

    ELBEYLİ, İffet YAKAR

    2004-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    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.

  19. Immobilization of radioactive waste in cement based matrices

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Mattus, C.H.

    1998-07-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-03-01

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

  3. Properties of Cement Mortar Produced from Mixed Waste Materials with Pozzolanic Characteristics.

    Science.gov (United States)

    Yen, Chi-Liang; Tseng, Dyi-Hwa; Wu, Yue-Ze

    2012-07-01

    Waste materials with pozzolanic characteristics, such as sewage sludge ash (SSA), coal combustion fly ash (FA), and granulated blast furnace slag (GBS), were reused as partial cement replacements for making cement mortar in this study. Experimental results revealed that with dual replacement of cement by SSA and GBS and triple replacement by SSA, FA, and GBS at 50% of total cement replacement, the compressive strength (Sc) of the blended cement mortars at 56 days was 93.7% and 92.9% of the control cement mortar, respectively. GBS had the highest strength activity index value and could produce large amounts of CaO to enhance the pozzolanic activity of SSA/FA and form calcium silicate hydrate gels to fill the capillary pores of the cement mortar. Consequently, the Sc development of cement mortar with GBS replacement was better than that without GBS, and the total pore volume of blended cement mortars with GBS/SSA replacement was less than that with FA/SSA replacement. In the cement mortar with modified SSA and GBS at 70% of total cement replacement, the Sc at 56 days was 92.4% of the control mortar. Modifying the content of calcium in SSA also increased its pozzolanic reaction. CaCl(2) accelerated the pozzolanic activity of SSA better than lime did. Moreover, blending cement mortars with GBS/SSA replacement could generate more monosulfoaluminate to fill capillary pores.

  4. Properties of Cement Mortar Produced from Mixed Waste Materials with Pozzolanic Characteristics

    Science.gov (United States)

    Yen, Chi-Liang; Tseng, Dyi-Hwa; Wu, Yue-Ze

    2012-01-01

    Abstract Waste materials with pozzolanic characteristics, such as sewage sludge ash (SSA), coal combustion fly ash (FA), and granulated blast furnace slag (GBS), were reused as partial cement replacements for making cement mortar in this study. Experimental results revealed that with dual replacement of cement by SSA and GBS and triple replacement by SSA, FA, and GBS at 50% of total cement replacement, the compressive strength (Sc) of the blended cement mortars at 56 days was 93.7% and 92.9% of the control cement mortar, respectively. GBS had the highest strength activity index value and could produce large amounts of CaO to enhance the pozzolanic activity of SSA/FA and form calcium silicate hydrate gels to fill the capillary pores of the cement mortar. Consequently, the Sc development of cement mortar with GBS replacement was better than that without GBS, and the total pore volume of blended cement mortars with GBS/SSA replacement was less than that with FA/SSA replacement. In the cement mortar with modified SSA and GBS at 70% of total cement replacement, the Sc at 56 days was 92.4% of the control mortar. Modifying the content of calcium in SSA also increased its pozzolanic reaction. CaCl2 accelerated the pozzolanic activity of SSA better than lime did. Moreover, blending cement mortars with GBS/SSA replacement could generate more monosulfoaluminate to fill capillary pores. PMID:22783062

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Institute of Scientific and Technical Information of China (English)

    刘阳生; 韩杰; 等

    2003-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  11. Radioactive wastes dispersed in stabilized ash cements

    International Nuclear Information System (INIS)

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

  12. Radioactive wastes dispersed in stabilized ash cements

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  13. Treatment and recycling of asbestos-cement containing waste

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-11-15

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

  14. Treatment and recycling of asbestos-cement containing waste

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    OpenAIRE

    Somnuk Tangtermsirikul; Pichaya Rachdawong; Kritsada Sisomphon

    2000-01-01

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

  17. Thermal stability testing of low-level waste forms

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  20. Recycling of textile wastes in fibre-cement composites

    OpenAIRE

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

    2013-01-01

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

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

    Science.gov (United States)

    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.

  2. Characterization of low and medium level radioactive waste forms

    International Nuclear Information System (INIS)

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

  3. Applicability of the Waste Fibres in Cement Paste

    Directory of Open Access Journals (Sweden)

    Regina KALPOKAITĖ DIČKUVIENĖ

    2013-09-01

    Full Text Available Fibres produced from waste catalyst together with commercially available polypropylene fibres were incorporated into ordinary Portland cement paste. The effects of fibre content as well as a mix of different type of fibres on mechanical and physical properties of wet and dry samples were investigated. The results showed that presence of fibres reduced compressive strength of the plain cement in wet and dry state. Contrary, when the combination of 1.5 wt% waste and 1.5 wt% polypropylene fibres was used flexural strength of cement mixture increased by up to 9 % at the age of 28 days. It was observed that addition of 1.5 wt% of only waste fibres improved flexural strength after long hydration period as well. However, the lowest mechanical strength results showed samples with 3 wt% of waste fibres. It was also observed that higher content of waste fibres reduced porosity of the cement mixture and consequently, decreased water absorption capacity. Presence of fibres reduced drying shrinkage of samples and they were lower than plain cement after 28 days of hydration. DOI: http://dx.doi.org/10.5755/j01.ms.19.3.1992

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

    Science.gov (United States)

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

    2011-11-15

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

  5. Immobilization of radioactive waste in cement-based matrices

    International Nuclear Information System (INIS)

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

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

    OpenAIRE

    Carmalin Sophia Ayyappan

    2015-01-01

    In this study, Thiobacillus thiooxidans (T. thiooxidans) was used to study the microbial stability / degradation of cement-based waste forms. The waste forms contained a chromium salt (CrCl3·6H2O), cement and other additives viz., lime and gypsum in two different proportions. The experimental samples of all the simulated waste forms showed evidence of microbial growth as indicated by substantial increase in sulfate. Chromium leached from the waste forms was found to be lowest in cement – lime...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

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

    Institute of Scientific and Technical Information of China (English)

    HOU Haobo; HE Xinghua; ZHU Shujing; ZHANG Dajie

    2006-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

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

  10. Cementation of biodegraded radioactive oils and organic waste

    International Nuclear Information System (INIS)

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

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

    Directory of Open Access Journals (Sweden)

    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.

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

    International Nuclear Information System (INIS)

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

  13. Calcium Sulfoaluminate Eco-Cement from Industrial Waste

    OpenAIRE

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

    2013-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    Directory of Open Access Journals (Sweden)

    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.

  16. Mixed Waste Focus Area - Waste form initiative

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Directory of Open Access Journals (Sweden)

    Nusirat Aderinsola Sadiku

    2015-07-01

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

  20. Vitreous ceramic waste form for waste immobilization

    International Nuclear Information System (INIS)

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

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

    OpenAIRE

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

    2013-01-01

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

  2. Polymer-Cement Composites Containing Waste Perlite Powder

    Directory of Open Access Journals (Sweden)

    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.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  5. Laboratory procedures for waste form testing

    Energy Technology Data Exchange (ETDEWEB)

    Mast, E.S.

    1994-09-19

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

  6. Alternative waste forms: a comparative study

    International Nuclear Information System (INIS)

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

  7. Using of borosilicate glass waste as a cement additive

    Science.gov (United States)

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

    2016-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  9. Waste brick's potential for use as a pozzolan in blended Portland cement.

    Science.gov (United States)

    Lin, Kae-Long; Chen, Bor-Yann; Chiou, Chyow-San; An Cheng

    2010-07-01

    This study investigated the pozzolanic reactions and engineering properties of waste brick-blended cements in relation to various replacement ratios (0-50%). The waste brick consisted of SiO(2) (63.21%), Al(2)O(3) (16.41%), Fe(2)O(3) (6.05%), Na(2)O (1.19%), K(2)O (2.83%) and MgO (1.11%), and had a pozzolanic activity index of 107%. The toxic characteristic leaching procedure (TCLP) results demonstrate that the heavy-metal content in waste bricks met the Environmental Protection Agency regulatory limits. Experimental results indicate that 10, 20, 30, 40 and 50% of cement can be replaced by waste brick, which causes the initial and final setting times to increase. Compressive strength development was slower in waste brick-blended cement (WBBC) pastes in the early ages; however, strength at the later ages increased significantly. Species analyses demonstrate that the hydrates in WBBC pastes primarily consisted of Ca(OH)(2) and calcium silicate hydrate (C-S-H) gel, like those found in ordinary Portland cement (OPC) paste. Pozzolanic reaction products formed in the WBBC pastes, in particular, various reaction products, including hydrates of calcium silicates (CSH), aluminates (CAH) and aluminosilicates (CASH), formed as expected, resulting in consumption of Ca(OH)(2) during the late ages of curing. The changes in the properties of WBBC pastes were significant as blend ratio increased, due to the pores of C-S-H gels and CAH filling via pozzolanic reactions. This filling of gel pores resulted in densification and subsequently enhanced the gel/space ratio and degree of hydration. Experimental results demonstrate waste brick can be supplementary cementitious material.

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2009-05-15

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

  12. Immobilisation of shredded waste in a cement matrix

    International Nuclear Information System (INIS)

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

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

    Directory of Open Access Journals (Sweden)

    Carmalin Sophia Ayyappan

    2015-06-01

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

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

    Science.gov (United States)

    Záleská, Martina; Pavlíková, Milena; Pavlík, Zbyšek

    2016-07-01

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

  15. Effect of Ground Waste Concrete Powder on Cement Properties

    Directory of Open Access Journals (Sweden)

    Xianwei Ma

    2013-01-01

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

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

    International Nuclear Information System (INIS)

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

  17. Effect of Concrete Waste Form Properties on Radionuclide Migration

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-09-30

    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. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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/sub 2/SO/sub 4/, 25 wt % H/sub 3/BO/sub 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/sub 2/SO/sub 4/, 40 wt % H/sub 3/BO/sub 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.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Directory of Open Access Journals (Sweden)

    Rosario Giménez-García

    2016-06-01

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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    Siqueira, F B; Holanda, J N F

    2013-12-15

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

  4. NNWSI waste form testing program

    International Nuclear Information System (INIS)

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

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

    OpenAIRE

    Wang, Ning

    2008-01-01

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

  6. Coated particle waste form development

    International Nuclear Information System (INIS)

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

  7. Properties of Injectable Apatite-Forming Premixed Cements.

    Science.gov (United States)

    Shimada, Yashushi; Chow, Laurence C; Takagi, Shozo; Tagami, Junji

    2010-07-01

    Previous studies reported premixed calcium phosphate cements (CPCs) that were stable in the package and form hydroxyapatite (HA) as the product after exposure to an aqueous environment. These cements had setting times of greater than 60 min, which are too long to be useful for some clinical applications. The present study investigated properties of fast-setting HA-forming premixed CPCs that initially consisted of two separate premixed pastes: (1) finely ground (1.0 μm in median size) dicalcium phosphate anhydrous (DCPA) mixed with an aqueous NaH(2)PO(4) solution, 1.5 mol/L or 3.0 mol/L in concentration, and (2) tetracalcium phosphate consisting of combinations of particles of two different size distributions, 5 μm (TTCP5) and 17 μm (TTCP17) in median size, mixed with glycerin. Equal volume of Pastes 1 and 2 were injected with the use of atwo-barrel syringe fitted with a static mixer into sample molds. The molar Ca/P ratio of combined paste was approximately 1.5. Cements were characterized in terms of setting time (Gilmore needle), diametral tensile strength (DTS), and phase composition (powder x-ray diffraction, XRD). Setting times were found to range from (4.3 ± 0.6 to 68 ± 3) min (mean ± sd; n = 3), and 1-d and 7-d DTS values were from (0.89 ± 0.08 to 2.44 ± 0.16) MPa (mean ± sd; n = 5). Both the NaH(2)PO(4) concentration and TTCP particle size distribution had significant (p Powder XRD analysis showed that low crystallinity HA and unreacted DCPA were present in the 1-day specimens, and the extent of HA formation increased with increasing amount of TTCP5 in the TTCP paste. CONCLUSION: Injectable HA-forming premixed CPCs with setting times from 4 to 70 min can be prepared by using DCPA and TTCP as the ingredients. Compared to the conventional powder liquid cements, these premixed CPCs have the advantages of being easy to use and having a range of hardening times. PMID:21479133

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-10-15

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

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

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

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

  10. Secondary Waste Cast Stone Waste Form Qualification Testing Plan

    Energy Technology Data Exchange (ETDEWEB)

    Westsik, Joseph H.; Serne, R. Jeffrey

    2012-09-26

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

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

    International Nuclear Information System (INIS)

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

  12. TSA waste stream and final waste form composition

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Directory of Open Access Journals (Sweden)

    Heikal M.

    2013-06-01

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

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

    International Nuclear Information System (INIS)

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

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

    Directory of Open Access Journals (Sweden)

    Arina Sauki

    2013-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-06-21

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

  19. Minerals as natural analogues for crystalline nuclear waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Giere, R. [Purdue University West Lafayette, Earth and Atmospheric Sciences (United States)

    2000-07-01

    Between the mining of uranium ore (mostly as uraninite) and the final disposal of nuclear waste, there are many processes and steps which together comprise the nuclear fuel cycle. Radioactive waste will be generated as long as nuclear reactors are in operation, but it is also produced by other means, e.g., during certain medical, scientific and industrial procedures. The most dangerous wastes are those resulting from the reprocessing of spent nuclear fuel and from some processes in the production and dismantling of nuclear weapons. A large part of this highly radioactive waste is present as a liquid and thus, its safe isolation from the biosphere requires immobilization of the radionuclides in a durable matrix (waste form). This is a solid which must be resistant to heat, radiation and corrosion over a geologic time scale. Three main categories of waste forms have been developed for the immobilization of radioactive waste, namely glasses, crystalline and multibarrier waste forms. One of the key properties of a nuclear waste form is its chemical durability (or resistance to corrosion), because the waste form represents the primary barrier to radionuclide release. The sciences of mineralogy and petrology have both contributed significantly to the development, characterization and performance assessment of such waste forms. The most important goal of safe nuclear waste disposal is to ensure that practically no radioactive materials reach the biosphere and, ultimately, human beings. Therefore, the design of final repositories is based on an approach that places several obstacles, or barriers, between waste and biosphere, whereby each barrier has a specific role in preventing or delaying migration of radioactive material. This multibarrier concept is different for each type of waste but, for the option of geological disposal, it generally comprises the following five barriers: (1) waste form (contains the actual waste); (2) canister (surrounds waste form; composed of a

  20. CERAMIC WASTE FORM DATA PACKAGE

    Energy Technology Data Exchange (ETDEWEB)

    Amoroso, J.; Marra, J.

    2014-06-13

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  4. Secondary Waste Form Down Selection Data Package – Ceramicrete

    Energy Technology Data Exchange (ETDEWEB)

    Cantrell, Kirk J.; Westsik, Joseph H.

    2011-08-31

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

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

    Science.gov (United States)

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

    2011-06-15

    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). PMID:21536381

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

    Science.gov (United States)

    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.

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

    NARCIS (Netherlands)

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

    2013-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Barbora Lyčkova

    2008-12-01

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

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

    Science.gov (United States)

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

    2015-06-01

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

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

    OpenAIRE

    Barbora Lyčkova; Vladimir Huda

    2008-01-01

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

  11. Low temperature waste form process intensification

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-30

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-08-12

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

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

    Directory of Open Access Journals (Sweden)

    Pawel Sikora

    2016-08-01

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

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

    OpenAIRE

    Kara, P

    2014-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Directory of Open Access Journals (Sweden)

    Macías, A.

    1999-06-01

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

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

  19. Advantages of using glycolic acid as a retardant in a brushite forming cement.

    Science.gov (United States)

    Mariño, Faleh Tamimi; Torres, Jesús; Hamdan, Mohammad; Rodríguez, Carmen Rueda; Cabarcos, Enrique López

    2007-11-01

    In this study we have compared the effect of using acetic, glycolic, and citric acids on the brushite cement setting reaction and the properties of the resultant cement. The cement solid phase was made by mixing beta-tricalcium phosphate (beta-TCP), monocalcium dihydrogen phosphate anhydrate (MCPA), and sodium pyrophosphate, whereas the cement liquid phase consisted of aqueous solutions of carboxy acids at concentrations ranging from 0.5 to 3.5M. Cements were prepared by mixing the solid phase with the liquid phase to form a workable paste. The cement setting time was longer for glycolic and citric acids. The best mechanical properties in dry environments were obtained using glycolic and citric acid liquid phases. In a wet environment at 37 degrees C, the cement set with glycolic acid was the strongest one. Brushite cement diametral tensile strength seems to be affected by the calcium-carboxyl phase produced in the setting reaction. The acceptable setting time and mechanical properties of cements set in glycolic acid solutions are attributed to the additional hydrophilic groups in the carboxylic acid and the low solubility in water of the calcium salt produced in the reaction. Moreover, at high concentrations, carboxylic acids add chemically to the cement matrix becoming reactants themselves.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Directory of Open Access Journals (Sweden)

    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.

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

    International Nuclear Information System (INIS)

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

  4. Liquid secondary waste. Waste form formulation and qualification

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-03-01

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

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

    OpenAIRE

    Kara, P

    2013-01-01

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

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

    Science.gov (United States)

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

    2011-10-30

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

  7. Studies on the reuse of waste printed circuit board as an additive for cement mortar.

    Science.gov (United States)

    Ban, Bong-Chan; Song, Jong-Yoon; Lim, Joong-Yeon; Wang, Soo-Kyoon; An, Kwang-Guk; Kim, Dong-Su

    2005-01-01

    The recent development in electronic industries has generated a drastic increase in production of printed circuit boards (PCB). Accordingly, the amount of waste PCB from electronic productions and waste electronics and its environmental impact such as soil and groundwater contamination have become a great concern. This study aims to propose a method for reuse of waste PCB as an additive for cement mortar. Although the expansibility of waste PCB powder finer than 0.08 mm in water was observed to be greater than 2.0%, the maximum expansion rates in water for 0.08 to approximately 0.15 and 0.15 to approximately 0.30 mm sized PCB powders were less than 2.0%, which satisfied the necessary condition as an alternative additive for cement mortar in place of sand. The difference in the compressive strength of standard mortar and waste PCB added mortar was observed to be less than 10% and their difference was expected to be smaller after prolonged aging. The durability of waste PCB added cement mortar was also examined through dry/wet conditioning cyclic tests and acidic/alkaline conditioning tests. From the tests, both weight and compressive strength of cement mortar were observed to be recovered with aging. The leaching test for heavy metals from waste PCB added mortar showed that no heavy metal ions such as copper, lead, or cadmium were detected in the leachate, which resulted from fixation effect of the cement hydrates.

  8. Miscellaneous Waste-Form FEPs

    Energy Technology Data Exchange (ETDEWEB)

    A. Schenker

    2000-12-08

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

  9. Obtaining a sulfoaluminate belite cement by industrial waste

    Directory of Open Access Journals (Sweden)

    Elkhadiri, L.

    2003-06-01

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

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-08-01

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

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

    International Nuclear Information System (INIS)

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

  14. Combined Waste Form Cost Trade Study

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-11-01

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

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

    Science.gov (United States)

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

    2016-07-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-08-15

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  1. 40 CFR 63.1220 - What are the replacement standards for hazardous waste burning cement kilns?

    Science.gov (United States)

    2010-07-01

    ... hazardous waste burning cement kilns? 63.1220 Section 63.1220 Protection of Environment ENVIRONMENTAL..., and Lightweight Aggregate Kilns § 63.1220 What are the replacement standards for hazardous waste... oxygen; (4) For arsenic, beryllium, and chromium, both: (i) Emissions in excess of 2.1 × 10−5...

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-09-01

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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    Lertwattanaruk, Pusit; Makul, Natt; Siripattarapravat, Chalothorn

    2012-11-30

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-01-01

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

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

    International Nuclear Information System (INIS)

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

  11. DWPF waste form compliance plan (Draft Revision)

    International Nuclear Information System (INIS)

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

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

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  15. Immobilization of radioactive waste in cement-based matrices

    International Nuclear Information System (INIS)

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

  16. Rietveld analysis of ceramic nuclear waste forms

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-12-31

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

  17. Ceramic and glass radioactive waste forms

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-01-01

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

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

    Science.gov (United States)

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

    2016-03-01

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

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

    International Nuclear Information System (INIS)

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

  20. Immobilization of radioactive waste in cement based matrices

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-09-12

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-01-01

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

  3. Advanced waste forms from spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-11-01

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

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

    International Nuclear Information System (INIS)

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

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

    OpenAIRE

    Al Khaja, Waheeb A.

    1992-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. CRYSTALLINE CERAMIC WASTE FORMS: REFERENCE FORMULATION REPORT

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-05-15

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

  11. SEPARATIONS AND WASTE FORMS CAMPAIGN IMPLEMENTATION PLAN

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-11-26

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

  12. Immobilization of radioactive waste in cement based matrices

    International Nuclear Information System (INIS)

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

  13. Chemical compatibility of DWPF canistered waste forms

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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 the waste

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

    Directory of Open Access Journals (Sweden)

    V. W. Francis Thoo

    2013-01-01

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

  18. Transuranic contaminated waste form characterization and data base

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-07-01

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

  19. Transuranic contaminated waste form characterization and data base

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2016-04-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  3. Utilization of ceramic waste as fine aggregate within Portland cement and fly ash concretes

    Energy Technology Data Exchange (ETDEWEB)

    Pincha Torkittikul; Arnon Chaipanich [Chiang Mai University, Chiang Mai (Thailand). Construction Materials Research Unit

    2010-07-15

    The aim of this research work was to investigate the feasibility of using ceramic waste and fly ash to produce mortar and concrete. Ceramic waste fragments obtained from local industry were crushed and sieved to produce fine aggregates. The measured concrete properties demonstrate that while workability was reduced with increasing ceramic waste content for Portland cement concrete and fly ash concrete, the workability of the fly ash concrete with 100% ceramic waste as fine aggregate remained sufficient, in contrast to the Portland cement control concrete with 100% ceramic waste where close to zero slump was measured. The compressive strength of ceramic waste concrete was found to increase with ceramic waste content and was optimum at 50% for the control concrete, dropping when the ceramic waste content was increased beyond 50%. This was a direct consequence of having a less workable concrete. However, the compressive strength in the fly ash concrete increased with increasing ceramic waste content up to 100%. The benefits of using ceramic waste as fine aggregate in concrete containing fly ash were therefore verified.

  4. Feasibility study of the Portland cement industry waste for the reduction of energy consumption

    Energy Technology Data Exchange (ETDEWEB)

    Bernardo, Ana Carla de Souza Masselli; Junqueira, Mateus Augusto F. Chaib; Jorge, Ariosto Bretanha; Silva, Rogerio Jose da [Universidade Federal de Itajuba (UNIFEI), MG (Brazil). Institute of Mechanical Engineering]. E-mails: anacarlasz@unifei.edu.br; mateus_afcj@yahoo.com.br; ariosto.b.jorge@unifei.edu.br; rogeriojs@unifei.edu.br

    2008-07-01

    The Portland cement industry demand a high specific consumption of energy for the production of the clinker. The energy consumption for clinker production varies between 3000 and 5300 kJ/kg of produced clinker. The clinker is produced by blending of different raw materials in order t o achieve precise chemical proportions of lime, silica, alumina and iron in the finished product and by burning them at high temperatures. The Portland cement is a mixture of clinker, gypsum and other materials. Due to need of high temperatures, tradition ally the fuels used in the cement industry are mineral coal, fuel oil, natural gas and petroleum coke. The fuel burning in high temperature leads to the formation of the pollutant thermal NOx. The level of emissions of this pollutant is controlled by environmental law, thus the formation of pollutants in process need be controlled. Moreover, industrial waste has been used by Portland cement industries as a secondary fuel through a technique called co -processing. Materials like waste oils, plastics, waste tyres and sewage sludge are often proposed as alternative fuels for the cement industry. The residues can be introduced as secondary fuel or secondary raw material. For energy conservation in the process, mineralizers are added during the process production of the clinker. The mineralizers promote certain reactions which decrease the temperature in the kiln and improve the quality of the clinker. The adequate quantity of constituents in production process is complex, for maintain in controlled level, the quality of final product, the operational conditions of kiln, and the pollutant emissions. The purpose of the present work is to provide an analysis of an optimal production point through of the optimization technique considering, the introduction of the fuels, industrial wastes as secondary fuels, and raw materials, for the reduction of energy in the process of Portland cement production. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Bleier, A.

    1997-09-01

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

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

    International Nuclear Information System (INIS)

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

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

    OpenAIRE

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

    2015-01-01

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

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

    NARCIS (Netherlands)

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

    2011-01-01

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

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

    NARCIS (Netherlands)

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

    2012-01-01

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

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

    DEFF Research Database (Denmark)

    Brodersen, K.

    1992-01-01

    Considerable amounts of easily soluble salts such as sodium nitrate, sulphate, or carbonate are introduced into certain types of cemented waste. When such materials are stored in atmospheres with high relative humidity or disposed or by shallow land burial under unsaturated, but still humid...

  11. Lysimeter tests of SRP waste forms

    International Nuclear Information System (INIS)

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

  12. Alternative Waste Forms for Electro-Chemical Salt Waste

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-10-28

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

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

    Science.gov (United States)

    Yilmaz, Arin; Degirmenci, Nurhayat

    2009-05-01

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

  14. The immobilisation of shredded waste in a cement matrix

    International Nuclear Information System (INIS)

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

  15. Preparation techniques for ceramic waste form powder

    International Nuclear Information System (INIS)

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

  16. Optimization of performance of cement matrices for solidification and encapsulation of radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Khotle, M.K. [South African Nuclear Energy Corporation Limited, Church Street West Ext Pelindaba, BRITS, North Western Province (South Africa)

    2008-07-01

    Full text of publication follows: Vaalputs is South Africa's near surface radioactive waste disposal site for low level and medium level waste. Metal and concrete drums are used as waste containers, placed in trenches and covered with backfill material. Due to the chemical contents of the waste material and experimental evidence regarding the expected lifetime of the current cement waste containers require a more stringent requirement on the waste matrix. This presentation will be based on the research that has gone into the development of the matrices with low sorption and low porosity for future use by waste generators. The resulting matrices were tested for Carbon-14 immobilization and other anion nuclides as well as other nuclides such as Tritium, Iodine etc. The results will be presented. (author)

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

    OpenAIRE

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

    2001-01-01

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

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

    Institute of Scientific and Technical Information of China (English)

    Christos-Triantafyllos Galbenis; Stamatis Tsimas

    2006-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-28

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

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

    Science.gov (United States)

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

    2015-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

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

    International Nuclear Information System (INIS)

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

  3. Strength of Blended Cement Sandcrete & Soilcrete Blocks Containing Cassava Waste Ash and Plantain Leaf Ash

    Directory of Open Access Journals (Sweden)

    L. O. Ettu

    2013-01-01

    Full Text Available This work investigated the compressive strength of binary and ternary blended cement sandcrete and soilcrete blocks containing cassava waste ash (CWA and plantain leaf ash (PLA. 135 solid sandcrete blocks and 135 solid soilcrete blocks of 450mm x 225mm x 125mm were produced with OPC-CWA binary blended cement, 135 with OPC-PLA binary blended cement, and 135 with OPC-CWA-PLA ternary blended cement, each at percentage OPC replacement with pozzolan of 5%, 10%, 15%, 20%, and 25%.Three sandcrete blocks and three soilcrete blocks for each OPC-pozzolan mix and the control were crushed to obtain their compressive strengths at 3, 7, 14, 21, 28, 50, 90, 120, and 150 days of curing. Sandcrete and soilcrete block strengths from binary and ternary blended cements were found to be higher than the control values beyond 90 days of hydration. The 150-day strength values for OPC-CWA-PLA ternary blended cement sandcrete and soilcrete blocks were respectively 5.90N/mm2and 5.10N/mm2for 5% replacement, 5.80N/mm2and 4.95N/mm2for 10% replacement, 5.65N/mm2and 4.85N/mm2for 15% replacement, 5.60N/mm2and 4.75N/mm2for 20% replacement, and 5.25N/mm2and 4.65N/mm2for 25% replacement; while the control values were 5.20N/mm2and 4.65N/mm2. Thus, OPC-CWA and OPC-PLA binary blended cements as well as OPC-CWA-PLA ternary blended cement could be used in producing sandcrete and soilcrete blocks with sufficient strength for use in building and minor civil engineering works where the need for high early strength is not a critical factor.

  4. Application of alkaliphilic biofilm-forming bacteria to improve compressive strength of cement-sand mortar.

    Science.gov (United States)

    Park, Sung-Jin; Chun, Woo-Young; Kim, Wha-Jung; Ghim, Sa-Youl

    2012-03-01

    The application of microorganisms in the field of construction material is rapidly increasing worldwide; however, almost all studies that were investigated were bacterial sources with mineral-producing activity and not with organic substances. The difference in the efficiency of using bacteria as an organic agent is that it could improve the durability of cement material. This study aimed to assess the use of biofilm-forming microorganisms as binding agents to increase the compressive strength of cement-sand material. We isolated 13 alkaliphilic biofilmforming bacteria (ABB) from a cement tetrapod block in the West Sea, Korea. Using 16S RNA sequence analysis, the ABB were partially identified as Bacillus algicola KNUC501 and Exiguobacterium marinum KNUC513. KNUC513 was selected for further study following analysis of pH and biofilm formation. Cement-sand mortar cubes containing KNUC513 exhibited greater compressive strength than mineral-forming bacteria (Sporosarcina pasteurii and Arthrobacter crystallopoietes KNUC403). To determine the biofilm effect, Dnase I was used to suppress the biofilm formation of KNUC513. Field emission scanning electron microscopy image revealed the direct involvement of organic-inorganic substance in cement-sand mortar.

  5. Permeability of Consolidated Incinerator Facility Wastes Stabilized with Portland Cement

    Energy Technology Data Exchange (ETDEWEB)

    Walker, B.W.

    1999-08-23

    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).

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-31

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

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

    International Nuclear Information System (INIS)

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

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

    Directory of Open Access Journals (Sweden)

    Nicolas Reynier

    2015-11-01

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

  9. Characterization of radioactive waste forms. Volume 2

    International Nuclear Information System (INIS)

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

  10. Characterization of radioactive waste forms. Volume 1

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1989-11-01

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

  12. WRAP 2A Waste Form Qualification Plan

    Energy Technology Data Exchange (ETDEWEB)

    Burbank, D.A. Jr.

    1993-12-31

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

  13. Use of the “red gypsum” industrial waste as substitute of natural gypsum for commercial cements manufacturing

    Directory of Open Access Journals (Sweden)

    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.

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

    International Nuclear Information System (INIS)

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

  15. Dissolution of tailored ceramic nuclear waste forms

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-06-01

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

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

    Science.gov (United States)

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

    2003-12-01

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

  18. Melt processed multiphase ceramic waste forms for nuclear waste immobilization

    Energy Technology Data Exchange (ETDEWEB)

    Amoroso, Jake, E-mail: jake.amoroso@srs.gov [Savannah River National Laboratory, Aiken, SC 29808 (United States); Marra, James C. [Savannah River National Laboratory, Aiken, SC 29808 (United States); Tang, Ming [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Lin, Ye; Chen, Fanglin [University of South Carolina, Columbia, SC 29208 (United States); Su, Dong [Brookhaven National Laboratory, Upton, NY 11973 (United States); Brinkman, Kyle S. [Clemson University, Clemson, SC 29634 (United States)

    2014-11-15

    Highlights: • We explored the feasibility of melt processing multiphase titanate-based ceramics. • Melt processing produced phases obtained by alternative processing methods. • Phases incorporated multiple lanthanides and transition metals. • Processing in reducing atmosphere suppressed un-desirable Cs–Mo coupling. • Cr partitions to and stabilizes the hollandite phase, which promotes Cs retention. - Abstract: Ceramic waste forms are promising hosts for nuclear waste immobilization as they have the potential for increased durability and waste loading compared with conventional borosilicate glass waste forms. Ceramics are generally processed using hot pressing, spark plasma sintering, and conventional solid-state reaction, however such methods can be prohibitively expensive or impractical at production scales. Recently, melt processing has been investigated as an alternative to solid-state sintering methods. Given that melter technology is currently in use for High Level Waste (HLW) vitrification in several countries, the technology readiness of melt processing appears to be advantageous over sintering methods. This work reports the development of candidate multi-phase ceramic compositions processed from a melt. Cr additions, developed to promote the formation and stability of a Cs containing hollandite phase were successfully incorporated into melt processed multi-phase ceramics. Control of the reduction–oxidation (Redox) conditions suppressed undesirable Cs–Mo containing phases, and additions of Al and Fe reduced the melting temperature.

  19. Melt processed multiphase ceramic waste forms for nuclear waste immobilization

    Science.gov (United States)

    Amoroso, Jake; Marra, James C.; Tang, Ming; Lin, Ye; Chen, Fanglin; Su, Dong; Brinkman, Kyle S.

    2014-11-01

    Ceramic waste forms are promising hosts for nuclear waste immobilization as they have the potential for increased durability and waste loading compared with conventional borosilicate glass waste forms. Ceramics are generally processed using hot pressing, spark plasma sintering, and conventional solid-state reaction, however such methods can be prohibitively expensive or impractical at production scales. Recently, melt processing has been investigated as an alternative to solid-state sintering methods. Given that melter technology is currently in use for High Level Waste (HLW) vitrification in several countries, the technology readiness of melt processing appears to be advantageous over sintering methods. This work reports the development of candidate multi-phase ceramic compositions processed from a melt. Cr additions, developed to promote the formation and stability of a Cs containing hollandite phase were successfully incorporated into melt processed multi-phase ceramics. Control of the reduction-oxidation (Redox) conditions suppressed undesirable Cs-Mo containing phases, and additions of Al and Fe reduced the melting temperature.

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

    Directory of Open Access Journals (Sweden)

    Md. Rezaul Karim

    2014-12-01

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

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

    Institute of Scientific and Technical Information of China (English)

    Shunwen LIANG; Jianguo JIANG; Yan ZHANG; Xin XU

    2008-01-01

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

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

    International Nuclear Information System (INIS)

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

  3. Research Status on Solid Radioactive Wastes of Magnesium Phosphate Cement%磷酸镁水泥固化放射性废物研究现状

    Institute of Scientific and Technical Information of China (English)

    赖振宇; 李倩

    2012-01-01

    Cement solidification process has significant advantage when applied in the disposal of medium and low level radioactive wastes.Cement solidification process has been widely applied in storage and transportation of medium and low level radioactive wastes.Traditional cement solidification technology has many disadvantages,such as high water to cement ratio,high nuclei leaching rate,low solidified form strength and etc.Traditional cement must cure under high pH value environment,wastes need treatment previously.The setting time and strength of cement may affected by PO43-,BO33-,Zn,Sn.Therefore,the selection of new type cement for medium and low level radioactive wastes is necessary.Considering the characteristics of magnesium phosphate cement(MPC),magnesium phosphate cement was selected as matrix material in this study for solidifying medium and low level radioactive waste,and solification of medium and low level radioactive incineration ash was also studied.%放射性废物的固化是目前有害废物的研究热点。而水泥固化是常用于中低放射性废物的固化技术,但水泥固化技术仍存在一定的问题,如水灰比高、废物包容量小、对所固化的废物pH要求较高、固化时可能产生粉尘、PO43-,BO33-,Zn,Sn等可能影响固化体凝结时间和强度发展,造成材料的后期强度不够,材料性能劣化。因此,有必要在水泥体系上进行更多的选择,从而提高水泥固化体的性能,磷酸镁水泥则是其中可用于替代的新型水泥品种,文章对磷酸镁水泥用于放射性废物固化的研究现状进行了调研,并对今后的研究做了展望。

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

    Energy Technology Data Exchange (ETDEWEB)

    Berner, Urs

    1999-12-01

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

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

    OpenAIRE

    Kara, P; Csetényi, L; Borosnyói, A

    2014-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  9. Waste Form Features, Events, and Processes

    Energy Technology Data Exchange (ETDEWEB)

    R. Schreiner

    2004-10-27

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-15

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

  13. DSNF and other waste form degradation abstraction

    Energy Technology Data Exchange (ETDEWEB)

    Thornton, Thomas A.

    2000-12-20

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

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

    Directory of Open Access Journals (Sweden)

    F. Puertas

    2015-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Weber, W.J.

    1981-09-01

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

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

    Science.gov (United States)

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

    2009-01-30

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

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

    International Nuclear Information System (INIS)

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

  18. Safeguards and retrievability from waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Danker, W.

    1996-05-01

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

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

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-21

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

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

    International Nuclear Information System (INIS)

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

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

    Directory of Open Access Journals (Sweden)

    Mr.R.Balamurugan

    2014-04-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-02-01

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

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

    International Nuclear Information System (INIS)

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

  9. NDA issues with RFETS vitrified waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Hurd, J.; Veazey, G.

    1998-12-31

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

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

    Directory of Open Access Journals (Sweden)

    S Pina

    2010-09-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-05-15

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  16. Use of disposed waste ash from landfills to replace Portland cement.

    Science.gov (United States)

    Rukzon, Sumrerng; Chindaprasirt, Prinya

    2009-09-01

    In this study, waste ash was utilized as a pozzolanic material in blended Portland cement in order to reduce negative environmental effects and landfill volume required to dispose of waste ash. The influence of waste ash, namely palm oil fuel ash, rice husk ash and fly ash on compressive strength and sulfate resistance in mortar were studied and evaluated by some accelerated short-term techniques in sodium sulfate solutions. Ordinary Portland cement (OPC) was partially replaced with ground palm oil fuel ash (POA), ground rice husk ash (RHA) and classified fly ash (FA). Single pozzolan and a blend of equal weight portions of POA, RHA and FA were also used. The resistance to sulfate attack of mortar improves substantially with partial replacement of OPC with POA, RHA and FA. The use of a blend of equal weight portions of FA and POA or RHA produced mixes with good strength and resistance to sulfate attack. POA, RHA and FA have a high potential to be used as a pozzolanic material.

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

    Energy Technology Data Exchange (ETDEWEB)

    Rusin, J.M.

    1980-12-01

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

  18. Review of high-level waste form properties

    International Nuclear Information System (INIS)

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

  19. Waste Marble Utilization from Residue Marble Industry as a Substitution of Cement and Sand within Concrete Rooftile Production

    OpenAIRE

    Candra Aditya, Abdul Halim, Chauliah Fatma Putri

    2014-01-01

    Research on alternative materials primarily from waste have been additional material at area manufacture of building materials , especially concreteroof tile [ 1 ] - [ 17 ] . This research will expand utilization of marble waste vBulletin East Java region of Indonesia in the manufacture of concrete roof tiles by combining the use of sand and waste marble powder as a substitute for riversand and portland cement .. This research creates material innovation product of...

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

    OpenAIRE

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

    2013-01-01

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

  1. Formulation and Analysis of Compliant Grouted Waste Forms for SHINE Waste Streams

    Energy Technology Data Exchange (ETDEWEB)

    Ebert, William [Argonne National Lab. (ANL), Argonne, IL (United States); Pereira, Candido [Argonne National Lab. (ANL), Argonne, IL (United States); Heltemes, Thad A. [Argonne National Lab. (ANL), Argonne, IL (United States); Youker, Amanda [Argonne National Lab. (ANL), Argonne, IL (United States); Makarashvili, Vakhtang [Argonne National Lab. (ANL), Argonne, IL (United States); Vandegrift, George F. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-01-01

    Optional grouted waste forms were formulated for waste streams generated during the production of 99Mo to be compliant with low-level radioactive waste regulations. The amounts and dose rates of the various waste form materials that would be generated annually were estimated and used to determine the effects of various waste processing options, such as the of number irradiation cycles between uranium recovery operations, different combinations of waste streams, and removal of Pu, Cs, and Sr from waste streams for separate disposition (which is not evaluated in this report). These calculations indicate that Class C-compliant grouted waste forms can be produced for all waste streams. More frequent uranium recovery results in the generation of more chemical waste, but this is balanced by the fact that waste forms for those waste streams can accommodate higher waste loadings, such that similar amounts of grouted waste forms are required regardless of the recovery schedule. Similar amounts of grouted waste form are likewise needed for the individual and combined waste streams. Removing Pu, Cs, and Sr from waste streams lowers the waste form dose significantly at times beyond about 1 year after irradiation, which may benefit handling and transport. Although these calculations should be revised after experimentally optimizing the grout formulations and waste loadings, they provide initial guidance for process development.

  2. CSNF WASTE FORM DEGRADATION: SUMMARY ABSTRACTION

    Energy Technology Data Exchange (ETDEWEB)

    J.C. CUNNANE

    2004-08-31

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

  3. PERFORMANCE OF CEMENT MORTARS REPLACED BY GROUND WASTE BRICK IN DIFFERENT AGGRESSIVE CONDITIONS

    Directory of Open Access Journals (Sweden)

    ILHAMI DEMIR

    2011-09-01

    Full Text Available This article investigates the sulphate resistance of cement mortars when subjected to different exposure conditions. Cement mortars were prepared using ground waste brick (GWB as a pozzolanic partial replacement for cement at replacement levels of 0%, 2.5%, 5%, 7.5, 10%, 12.5 and 15%. Mortar specimens were stored under three different conditions: continuous curing in lime-saturated tab water (TW, continuous exposure to 5% sodium sulphate solution (SS, and continuous exposure to 5% ammonium nitrate solution (AN, at a temperature of 20 ± 3 ºC, for 7, 28, 90, and 180 days. Prisms with dimensions of 25×25×285 mm, to determine the expansions of the mortar samples; and another set of prisms with dimensions of 40×40×160 mm, were prepared to calculate the compressive strength of the samples. It was determined that the GWB replacement ratios between 2.5% and 10% decreased the 180 days expansion values. The highest compressive strength values were found for the samples with 10% replacement ratio in the TW, SS, and AN conditions for 180 days. The microstructure of the mortars were investigated using scanning electron microscopy (SEM and the Energy dispersive X-ray (EDX.

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

    Directory of Open Access Journals (Sweden)

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

    2013-01-01

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

  5. Effect of curing time on the fraction of Cs{sup 137} from cement-waste matrix

    Energy Technology Data Exchange (ETDEWEB)

    Plecas, I.; Pavlovic, R.; Pavlovic, S. [Vinca Institute of Nuclear Sciences, Belgrade (Yugoslavia)

    2003-07-01

    To assess the safety of disposal of radioactive waste material in cement, curing conditions and time of leaching radionuclides {sup 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. Radiation effects in ceramic nuclear waste forms

    International Nuclear Information System (INIS)

    This paper reports on alpha-decay event damage (a particle and recoil-nucleus) that results in atomic-scale disorder which causes changes in the molar volume, corrosion rate, stored energy, mechanical properties, and macrostructure of ceramics. These changes particularly of volume and corrosion rate, have critical implications for the long-term durability of nuclear waste forms, such as the polyphase. Ti-based ceramic Synroc. This paper reviews data on actinide-bearing (U and Th) phases of great age (>100 m.y.) found in nature and compares these results to observation on actinide-doped phases (Pu and Cm) of nearly equivalent α-decay doses. Of particular interest is evidence for annealing of radiation damage effects over geologic periods of time under ambient conditions

  7. Transportation considerations related to waste forms and canisters for Defense TRU wastes

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, K.J.; Andrews, W.B.; Schreiber, A.M.; Rosenthal, L.J.; Odle, C.J.

    1981-09-01

    This report identifies and discusses the considerations imposed by transportation on waste forms and canisters for contact-handled, solid transuranic wastes from the US Department of Energy (DOE) activities. The report reviews (1) the existing raw waste forms and potential immobilized waste forms, (2) the existing and potential future DOE waste canisters and shipping containers, (3) regulations and regulatory trends for transporting commercial transuranic wastes on the ISA, (4) truck and rail carrier requirements and preferences for transporting the wastes, and (5) current and proposed Type B external packagings for transporting wastes.

  8. Transportation considerations related to waste forms and canisters for Defense TRU wastes

    International Nuclear Information System (INIS)

    This report identifies and discusses the considerations imposed by transportation on waste forms and canisters for contact-handled, solid transuranic wastes from the US Department of Energy (DOE) activities. The report reviews (1) the existing raw waste forms and potential immobilized waste forms, (2) the existing and potential future DOE waste canisters and shipping containers, (3) regulations and regulatory trends for transporting commercial transuranic wastes on the ISA, (4) truck and rail carrier requirements and preferences for transporting the wastes, and (5) current and proposed Type B external packagings for transporting wastes

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

    Directory of Open Access Journals (Sweden)

    I.A.Ikara

    2015-11-01

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

  10. Heat of Hydration of Low Activity Cementitious Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Nasol, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-07-23

    During the curing of secondary waste grout, the hydraulic materials in the dry mix react exothermally with the water in the secondary low-activity waste (LAW). The heat released, called the heat of hydration, can be measured using a TAM Air Isothermal Calorimeter. By holding temperature constant in the instrument, the heat of hydration during the curing process can be determined. This will provide information that can be used in the design of a waste solidification facility. At the Savannah River National Laboratory (SRNL), the heat of hydration and other physical properties are being collected on grout prepared using three simulants of liquid secondary waste generated at the Hanford Site. From this study it was found that both the simulant and dry mix each had an effect on the heat of hydration. It was also concluded that the higher the cement content in the dry materials mix, the greater the heat of hydration during the curing of grout.

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

    International Nuclear Information System (INIS)

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

  12. Manufacturing of concrete with residues from iron ore exploitation using the technology of radioactive waste cementation

    Energy Technology Data Exchange (ETDEWEB)

    Versieux, Juniara L.; Lameiras, Fernando S.; Tello, Cledola Cassia Oliveira de, E-mail: juniarani@gmail.com, E-mail: fsl@cdtn.br, E-mail: tellocc@cdtn.br [Centro de Desenvolvimento da Tecnologia Nucelar (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    Radioactive wastes from various segments of economy are immobilized by cementation, because of availability and widespread use in civil construction of cement. New cementitious materials are developed in CDTN using mining residues based on cementing techniques of radioactive wastes. Special procedures were developed to obtain concrete with the use of super plasticizers in which natural sand was totally replaced by mining residues. The motivation for this research is the exploration of banded iron formations (BIF) as iron ore in 'Quadrilatero Ferrifero' of Minas Gerais, where huge amounts of residues are generated with great concern about the environmental sustainability and safety of dams for residue storage. The exploitation of river sand causes many negative impacts, which leads to interest in its replacement by another raw material in mortar and concrete manufacturing. The use of BIF mining residues were studied for manufacturing of concrete pavers to contribute to reducing the impact caused by extraction of natural sand and use of mining residues. Previously developed procedures with total replacement of natural sand for mining residues were modified, including use of gravel to obtain pavers with improved properties. Four different mixtures were tested, in which the proportion of gravel and super plasticizer was varied. Monitored properties of pavers, among others, were compression resistance, water absorption, and void volume. With addition of gravel, the pavers had higher void index than those made only with mortar, and higher resistance to compression after 28 days of curing (an average of 18MPa of those made with mortar to 24MPa of those made with concrete). (author)

  13. Processing of concentrated radioactive wastes into cement and bitumens following calcination

    International Nuclear Information System (INIS)

    A brief characteristic is presented of the most frequently used processes of solidification of liquid radioactive wastes, viz., bituminization, cementation and their combination with calcination. The effect of individual parameters is assessed on the choice of the type of solidification process as is their importance in the actual process, in temporary storage, during transportation and under conditions of long-term storage. It has been found that a combination of the procedures could lead to a modular system of methods and equipment. This would allow to approach optimal solidification of wastes in the present period and to establish a research reserve for the development of more modern, economically advantageous and safer procedures. A rough estimate is made of the costs of the solidification of 1 m3 of radioactive concentrate from the V-1 power plant at a production of 380 m3/year, this for the cementation-calcination and bituminization-calcination procedures. The said rough economic analysis only serves to identify the major operating components which have the greatest effect on the economic evaluation of the solidification procedures. (Z.M.)

  14. Waste forms, packages, and seals working group summary

    Energy Technology Data Exchange (ETDEWEB)

    Sridhar, N. [Center Antonio, TX (United States); McNeil, M.B. [Nuclear Regulatory Commission, Washington, DC (United States)

    1995-09-01

    This article is a summary of the proceedings of a group discussion which took place at the Workshop on the Role of Natural Analogs in Geologic Disposal of High-Level Nuclear Waste in San Antonio, Texas on July 22-25, 1991. The working group concentrated on the subject of radioactive waste forms and packaging. Also included is a description of the use of natural analogs in waste packaging, container materials and waste forms.

  15. Migration of ions in cement paste as studied by SIMS

    Energy Technology Data Exchange (ETDEWEB)

    Prince, K.E.; Aldridge, L.P. [Australian Nuclear Science and Technology Organisation (ANSTO), Lucas Heights, NSW (Australia); Rougeron, P. [Electricite de France Direction des Etudes et Recherches, Les Renardiers (France)

    1998-06-01

    Cement is often used to condition and encapsulate low level radioactive waste before it is disposed of in a repository. Ground water can attack these waste-forms by transporting aggressive ions into the cement paste and by removing radioactive ions from the paste. The extent of the attack will be governed by the diffusion of the ions in the cement paste. In this study we examine the migration of aggressive carbonate ions and inactive Cs and Sr through cement pastes. The use of SIMS for establishing the penetration depths and diffusion profiles for Cs and Sr in cement will be explored. The penetration profiles of Cs and Sr in a non-zeolite cement paste were examined and compared to those of a paste made with zeolite. The effects of the non-homogeneous nature of the cement was most pronounced in the study of the zeolite rich cement; Cs being preferentially accumulated in the zeolite material. (authors). 4 refs., 2 figs.

  16. Solidification of Municipal Solid Waste Incineration Fly Ash with Cement and Its Leaching Behaviors of Heavy Metals

    Institute of Scientific and Technical Information of China (English)

    2003-01-01

    The solidifying effect of cement addition on municipal solid waste incineration fly ash (MSWFA for short,collected from the gas exhaust system of MSW incinerator),the interaction of MSWFA with cement and water and the leaching of heavy metals from cement-solidified MSWFA are investigated.The main results show that:(1) when MSWFA is mixed with cement and water,H2 evolution,the formation and volume expansion of AFt will take place,the volume expansion can be reduced by ground rice husk ash addition;(2) heavy metals do leach from cement-solidified MSWFA and at lower pH more leaching will occur;(3) compared with cement-solidified fly ash,the leachate of solidified MSWFA is with higher heavy metal contents;(4) with the increment of cement addition leached heavy metals are decreased;and (5) concentrations of Zn,Mn,Cu and Cd in all the leachates can meet the relevant Standards of Japan,but as the regulations for soil and groundwater protection of Japan are concerned,precautions against the leaching of Pb,Cl- and Cr6+ and so on are needed.

  17. Phosphate bonded ceramics as candidate final-waste-form materials

    International Nuclear Information System (INIS)

    Room-temperature setting phosphate-bonded ceramics were studied as candidate materials for stabilization of DOE low-level problem mixed wastes which cannot be treated by other established stabilization techniques. Phosphates of Mg, Mg-Na, Al and Zr were studied to stabilize ash surrogate waste containing RCRA metals as nitrates and RCRA organics. We show that for a typical loading of 35 wt.% of the ash waste, the phosphate ceramics pass the TCLP test. The waste forms have high compression strength exceeding ASTM recommendations for final waste forms. Detailed X-ray diffraction studies and differential thermal analyses of the waste forms show evidence of chemical reaction of the waste with phosphoric acid and the host matrix. The SEM studies show evidence of physical bonding. The excellent performance in the leaching tests is attributed to a chemical solidification and physical as well as chemical bonding of ash wastes in these phosphate ceramics

  18. Physical modeling of contaminant diffusion from a cementious waste form

    International Nuclear Information System (INIS)

    Cementitious materials can be used to immobilize waste materials for disposal. The Westinghouse Hanford Company is pursuing approval of disposal technologies by which hazardous and radioactive wastes are blended or packaged with cementitious materials for disposal. Of significant concern is the mobility of the waste contaminants both from the waste form and in the arid soils of the Hanford Site. A physical model has been developed to study the diffusion of waste contaminants from simulated cementitious waste forms in unsaturated Hanford Site soils. The model can be used to predict cementitious waste form performance in a representative environment, support design of waste management facilities and technologies, and provide data for environmental permitting of proposed treatment and disposal facilities

  19. Material characterization in cemented radioactive waste with the associated particle technique

    Science.gov (United States)

    Carasco, C.; Perot, B.; Mariani, A.; El Kanawati, W.; Valkovic, V.; Sudac, D.; Obhodas, J.

    2010-07-01

    The elemental characterization of materials constituting radioactive waste is of great importance for the management of storage and repository facilities. To complement the information brought by gamma or X-ray imaging, the performance of a fast neutron interrogation system based on the associated particle technique (APT) has been investigated by using MCNP simulations and by performing proof-of-principle experiments. APT provides a 3D localisation of the emission of fast neutron induced gamma rays, whose spectroscopic analysis allows to identify the elements present in specific volumes of interest in the waste package. Monte Carlo calculations show that it is possible to identify materials enclosed behind the thick outer envelop of a ≈1 m 3 cemented waste drum, provided the excited nuclei emit gamma rays with a sufficient energy to limit photon attenuation. Neutron attenuation and scattering are also predominant effects that reduce the sensitivity and spatial selectivity of APT, but it is still possible to localise items in the waste by neutron time-of-flight and gamma-ray spectroscopy. Experimental tests confirm that the elemental characterization is possible across thick mortar slabs.

  20. Alternate nuclear waste forms and interactions in geologic media

    International Nuclear Information System (INIS)

    The primary purposes of the conference on Alternate Nuclear Waste Forms and Interactions in Geologic Media were: First, to provide an opportunity for a review of the status of the research on some of the candidate alternative waste forms; second, to provide an opportunity for comparing the characteristics of alternate waste forms to those of glasses; and third, to stimulate increased interactions between those research groups that were engaged in a more basic approach to characterizing waste forms and those who were concerned with more applied aspects such as the processing of these materials. The motivating philosophy behind this third purpose of the conference was based on the idea that by operating from the soundest possible fundamental base for any of the candidate waste forms, hopefully any future unpleasant surprise - such as that alluded to earlier in the case of glass waste forms - could be avoided. Separate abstracts have been prepared for individual papers for inclusion in the Energy Data Base

  1. DURABILITY TESTING OF FLUIDIZED BED STEAM REFORMER (FBSR) WASTE FORMS

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C

    2006-01-06

    Fluidized Bed Steam Reforming (FBSR) is being considered as a potential technology for the immobilization of a wide variety of high sodium aqueous radioactive wastes. The addition of clay and a catalyst as co-reactants converts high sodium aqueous low activity wastes (LAW) such as those existing at the Hanford and Idaho DOE sites to a granular ''mineralized'' waste form that may be made into a monolith form if necessary. Simulant Hanford and Idaho high sodium wastes were processed in a pilot scale FBSR at Science Applications International Corporation (SAIC) Science and Technology Applications Research (STAR) facility in Idaho Falls, ID. Granular mineral waste forms were made from (1) a basic Hanford Envelope A low-activity waste (LAW) simulant and (2) an acidic INL simulant commonly referred to as sodium-bearing waste (SBW). The FBSR waste forms were characterized and the durability tested via ASTM C1285 (Product Consistency Test), the Environmental Protection Agency (EPA) Toxic Characteristic Leaching Procedure (TCLP), and the Single Pass Flow Through (SPFT) test. The durability of the FBSR waste form products was tested in order to compare the measured durability to previous FBSR waste form testing on Hanford Envelope C waste forms that were made by THOR Treatment Technologies (TTT) and to compare the FBSR durability to vitreous LAW waste forms, specifically the Hanford low activity waste (LAW) glass known as the Low-activity Reference Material (LRM). The durability of the FBSR waste form is comparable to that of the LRM glass for the test responses studied.

  2. Hot-pressed barium sulphate ceramic waste forms for direct immobilization of medium level Magnox waste

    International Nuclear Information System (INIS)

    A possible method of treatment for Magnox cladding waste is by dissolution in nitric acid and precipitation of barium sulphate-based floc with which radioactive ions are co-precipitated. The floc could then be immobilized in a matrix material such as cement or bitumen to give the waste form, or alternatively can be converted directly into a waste form by hot pressing. This paper describes the direct conversion of barium sulphate floc, containing simulated radwaste, into a synthetic, ceramic version of the natural mineral barite by a hot-pressing route. By variation of the parameters pressure, temperature and time, optimum conditions for consolidation of the floc to > 90% theoretical density on a laboratory scale are found to be 22.5 MPa, 9000C for 10 minutes. Using a pressure of 15 MPa, at 9000C for 30 min., hot-pressed billets of BaSO4 have been made on a 5 kg scale. In going from the magnox waste to the hot-pressed barium sulphate a volume reduction factor approx. 18 is achieved. The principal phases in the product are found to be BaSO4, MgO and Fe3O4, and the degree of consolidation achieved depends on the MgO content. The leaching behaviour of the hot-pressed materials in 1000C, 3 day Soxhlet tests also depends on the MgO content, and on the consequent level of open porosity. If there is porosity accessible to the leach water, MgO at the internal surfaces is converted to Mg(OH)2, which deposits within the pores, and a weight gain is registered in the Soxhlet test. If, however, there is no open porosity, a weight loss occurs, and leach rates approx. 4 x 10-7 kg/m2/sec are found. In contrast, pure BaSO4, hot-pressed to similar densities, shows no variation in leaching behaviour over a wide range of open porosities, and gives Soxhlet leach rates approx. 8 x 10-8 kg/m2/sec. 6 figures, 2 tables

  3. Waste-Form Development Program. Annual progress report, October 1981-September 1982

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-09-01

    Low-level wastes (LLW) at nuclear facilities have traditionally been solidified using portland cement (with and without additives). Urea-formaldehyde has been used for LLW solidification while bitumen (asphalt) and thermosetting polymers will be applied to domestic wastes in the near future. Operational difficulties have been observed with each of these solidification agents. Such difficulties include incompatibility with waste constitutents inhibiting solidification, premature setting, free standing water and fires. Some specific waste types have proven difficult to solidify with one or more of the contemporary agents. Similar problems are also anticipated for the solidification of new wastes, which are generated using advanced volume reduction technologies, and with the application of additional agents which may be introduced in the near future for the solidification of LLW. In the Waste Form Development program, contemporary solidification agents are being investigated relative to their potential applications to major fuel cycle and non-fuel cycle LLW streams. The range of conditions under which these solidification agents can be satisfactorily applied to specific LLW streams is being determined. These studies are primarily directed towards defining operating parameters for both improved solidification of problem wastes such as ion exchange resins, organic liquids and oils for which prevailing processes, as currently employed, appear to be inadequate, and solidification of new LLW streams including high solids content evaporator concentrates, dry solids, and incinerator ash generated from advanced volume reduction technologies. Solidified waste forms are tested and evaluated to demonstrate compliance with waste form performance and shallow land burial (SLB) acceptance criteria and transportation requirements (both as they currently exist and as they are anticipated to be modified with time).

  4. Waste-Form Development Program. Annual progress report, October 1981-September 1982

    International Nuclear Information System (INIS)

    Low-level wastes (LLW) at nuclear facilities have traditionally been solidified using portland cement (with and without additives). Urea-formaldehyde has been used for LLW solidification while bitumen (asphalt) and thermosetting polymers will be applied to domestic wastes in the near future. Operational difficulties have been observed with each of these solidification agents. Such difficulties include incompatibility with waste constitutents inhibiting solidification, premature setting, free standing water and fires. Some specific waste types have proven difficult to solidify with one or more of the contemporary agents. Similar problems are also anticipated for the solidification of new wastes, which are generated using advanced volume reduction technologies, and with the application of additional agents which may be introduced in the near future for the solidification of LLW. In the Waste Form Development program, contemporary solidification agents are being investigated relative to their potential applications to major fuel cycle and non-fuel cycle LLW streams. The range of conditions under which these solidification agents can be satisfactorily applied to specific LLW streams is being determined. These studies are primarily directed towards defining operating parameters for both improved solidification of problem wastes such as ion exchange resins, organic liquids and oils for which prevailing processes, as currently employed, appear to be inadequate, and solidification of new LLW streams including high solids content evaporator concentrates, dry solids, and incinerator ash generated from advanced volume reduction technologies. Solidified waste forms are tested and evaluated to demonstrate compliance with waste form performance and shallow land burial (SLB) acceptance criteria and transportation requirements (both as they currently exist and as they are anticipated to be modified with time)

  5. Final waste forms project: Performance criteria for phase I treatability studies

    International Nuclear Information System (INIS)

    This document defines the product performance criteria to be used in Phase I of the Final Waste Forms Project. In Phase I, treatability studies will be performed to provide open-quotes proof-of-principleclose quotes data to establish the viability of stabilization/solidification (S/S) technologies. This information is required by March 1995. In Phase II, further treatability studies, some at the pilot scale, will be performed to provide sufficient data to allow treatment alternatives identified in Phase I to be more fully developed and evaluated, as well as to reduce performance uncertainties for those methods chosen to treat a specific waste. Three main factors influence the development and selection of an optimum waste form formulation and hence affect selection of performance criteria. These factors are regulatory, process-specific, and site-specific waste form standards or requirements. Clearly, the optimum waste form formulation will require consideration of performance criteria constraints from each of the three categories. Phase I will focus only on the regulatory criteria. These criteria may be considered the minimum criteria for an acceptable waste form. In other words, a S/S technology is considered viable only if it meet applicable regulatory criteria. The criteria to be utilized in the Phase I treatability studies were primarily taken from Environmental Protection Agency regulations addressed in 40 CFR 260 through 265 and 268; and Nuclear Regulatory Commission regulations addressed in 10 CFR 61. Thus the majority of the identified criteria are independent of waste form matrix composition (i.e., applicable to cement, glass, organic binders etc.)

  6. Final waste forms project: Performance criteria for phase I treatability studies

    Energy Technology Data Exchange (ETDEWEB)

    Gilliam, T.M. [Oak Ridge National Lab., TN (United States); Hutchins, D.A. [Martin Marietta Energy Systems, Inc., Oak Ridge, TN (United States); Chodak, P. III [Massachusetts Institute of Technology (United States)

    1994-06-01

    This document defines the product performance criteria to be used in Phase I of the Final Waste Forms Project. In Phase I, treatability studies will be performed to provide {open_quotes}proof-of-principle{close_quotes} data to establish the viability of stabilization/solidification (S/S) technologies. This information is required by March 1995. In Phase II, further treatability studies, some at the pilot scale, will be performed to provide sufficient data to allow treatment alternatives identified in Phase I to be more fully developed and evaluated, as well as to reduce performance uncertainties for those methods chosen to treat a specific waste. Three main factors influence the development and selection of an optimum waste form formulation and hence affect selection of performance criteria. These factors are regulatory, process-specific, and site-specific waste form standards or requirements. Clearly, the optimum waste form formulation will require consideration of performance criteria constraints from each of the three categories. Phase I will focus only on the regulatory criteria. These criteria may be considered the minimum criteria for an acceptable waste form. In other words, a S/S technology is considered viable only if it meet applicable regulatory criteria. The criteria to be utilized in the Phase I treatability studies were primarily taken from Environmental Protection Agency regulations addressed in 40 CFR 260 through 265 and 268; and Nuclear Regulatory Commission regulations addressed in 10 CFR 61. Thus the majority of the identified criteria are independent of waste form matrix composition (i.e., applicable to cement, glass, organic binders etc.).

  7. Effect of temperature on the durability of class C fly ash belite cement in simulated radioactive liquid waste: Synergy of chloride and sulphate ions

    Energy Technology Data Exchange (ETDEWEB)

    Guerrero, A., E-mail: aguerrero@ietcc.csic.es [Eduardo Torroja Institute for Construction Science (CSIC), C/Serrano Galvache, 4, 28033 Madrid (Spain); Goni, S., E-mail: sgoni@ietcc.csic.es [Eduardo Torroja Institute for Construction Science (CSIC), C/Serrano Galvache, 4, 28033 Madrid (Spain); Allegro, V.R., E-mail: allegro@ietcc.csic.es [Eduardo Torroja Institute for Construction Science (CSIC), C/Serrano Galvache, 4, 28033 Madrid (Spain)

    2009-06-15

    The durability of class C fly ash belite cement (FABC-2-W) in simulated radioactive liquid waste (SRLW) rich in a mixed sodium chloride and sulphate solution is presented here. The effect of the temperature and potential synergic effect of chloride and sulfate ions are discussed. This study has been carried out according to the Koch-Steinegger test, at the temperature of 20 deg. C and 40 deg. C during a period of 180 days. The durability has been evaluated by the changes of the flexural strength of mortar, fabricated with this cement, immersed in a simulated radioactive liquid waste rich in sulfate (0.5 M), chloride (0.5 M) and sodium (1.5 M) ions - catalogued like severely aggressive for the traditional Portland cement - and demineralised water, which was used as reference. The reaction mechanism of sulphate, chloride and sodium ions with the mortar was evaluated by scanning electron microscopy (SEM), porosity and pore-size distribution, and X-ray diffraction (XRD). The results showed that the chloride binding and formation of Friedel's salt was inhibited by the presence of sulphate. Sulphate ion reacts preferentially with the calcium aluminate hydrates forming non-expansive ettringite which precipitated inside the pores; the microstructure was refined and the mechanical properties enhanced. This process was faster and more marked at 40 deg. C.

  8. Effect of temperature on the durability of class C fly ash belite cement in simulated radioactive liquid waste: synergy of chloride and sulphate ions.

    Science.gov (United States)

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

    2009-06-15

    The durability of class C fly ash belite cement (FABC-2-W) in simulated radioactive liquid waste (SRLW) rich in a mixed sodium chloride and sulphate solution is presented here. The effect of the temperature and potential synergic effect of chloride and sulfate ions are discussed. This study has been carried out according to the Koch-Steinegger test, at the temperature of 20 degrees C and 40 degrees C during a period of 180 days. The durability has been evaluated by the changes of the flexural strength of mortar, fabricated with this cement, immersed in a simulated radioactive liquid waste rich in sulfate (0.5M), chloride (0.5M) and sodium (1.5M) ions--catalogued like severely aggressive for the traditional Portland cement--and demineralised water, which was used as reference. The reaction mechanism of sulphate, chloride and sodium ions with the mortar was evaluated by scanning electron microscopy (SEM), porosity and pore-size distribution, and X-ray diffraction (XRD). The results showed that the chloride binding and formation of Friedel's salt was inhibited by the presence of sulphate. Sulphate ion reacts preferentially with the calcium aluminate hydrates forming non-expansive ettringite which precipitated inside the pores; the microstructure was refined and the mechanical properties enhanced. This process was faster and more marked at 40 degrees C.

  9. Immobilization of Technetium in a Metallic Waste Form

    Energy Technology Data Exchange (ETDEWEB)

    S.M. Frank; D. D. Keiser, Jr.; K. C. Marsden

    2007-09-01

    Fission-product technetium accumulated during treatment of spent nuclear fuel will ultimately be disposed of in a geological repository. The exact form of Tc for disposal has yet to be determined; however, a reasonable solution is to incorporate elemental Tc into a metallic waste form similar to the waste form produced during the pyrochemical treatment of spent, sodium-bonded fuel. This metal waste form, produced at the Idaho National Laboratory, has undergone extensive qualification examination and testing for acceptance to the Yucca Mountain geological repository. It is from this extensive qualification effort that the behavior of Tc and other fission products in the waste form has been elucidated, and that the metal waste form is extremely robust in the retention of fission products, such as Tc, in repository like conditions. This manuscript will describe the metal waste form, the behavior of Tc in the waste form; and current research aimed at determining the maximum possible loading of Tc into the metal waste and subsequent determination of the performance of high Tc loaded metal waste forms.

  10. TRU waste form and package criteria meeting

    Energy Technology Data Exchange (ETDEWEB)

    None

    1977-08-01

    The broad subject of the meeting is the overall ERDA TRU waste management program, although the discussions also cover performance criteria for the Waste Isolation Pilot Plant and their implications for the overall TRU program. Separate abstracts were prepared for all ten presentations. (DLC)

  11. Ceramic waste form qualification using results from witness tubes

    International Nuclear Information System (INIS)

    A ceramic waste form has been developed to immobilize the salt waste stream from electrometallurgical treatment of spent nuclear fuel. The ceramic waste form is prepared in a hot isostatic press (HIP). The use of small, easily fabricated HIP capsules called witness tubes has been proposed as a practical way to obtain representative samples of ceramic waste form material for process monitoring, waste form qualification, and archiving. Witness tubes are filled with the same material used to fill the corresponding HIP can, and are HIPed along with the HIP can. Relevant physical, chemical, and performance (leach test) data are analyzed and compared. Differences between witness tube and HIP can materials are shown to be statistically insignificant, demonstrating that witness tubes do provide ceramic waste form material representative of the material in the corresponding HIP can.

  12. Variations and factors that influence the formation of polychlorinated naphthalenes in cement kilns co-processing solid waste.

    Science.gov (United States)

    Jin, Rong; Zhan, Jiayu; Liu, Guorui; Zhao, Yuyang; Zheng, Minghui

    2016-09-01

    Pilot studies of unintentionally produced pollutants should be performed before waste being co-processed in cement kilns. Polychlorinated naphthalene (PCN) formation and emission from cement kilns co-processing sorted municipal solid waste, sewage sludge, and waste acid, however, have not previously been studied. Here, PCNs were analyzed in stack gas samples and solid samples from different stages of three cement production runs. PCN destruction efficiencies were higher when waste was co-processed (93.1% and 88.7% in two tests) than when waste was not co-processed (39.1%), so co-processing waste would not increase PCN outputs. The PCN concentrations were higher in particle samples from the C1 preheater and stages at back end of kiln than in particle samples from other stages, suggesting that cyclone preheater and back end of kiln should be focused for controlling PCN emissions. Besides that, based on the variation of PCN concentrations and corresponding operating conditions in different stages, the temperature, feeding materials, and chlorine content were suggested as the main factors influencing PCN formation. The PCN homologue and congener profiles suggested chlorination and dechlorination were the main PCN formation and decomposition pathways, and congeners CN-23, CN-46, and CN-59 appear to be appropriate indicators of PCNs emitted from coal-burning sources. PMID:27187059

  13. Variations and factors that influence the formation of polychlorinated naphthalenes in cement kilns co-processing solid waste.

    Science.gov (United States)

    Jin, Rong; Zhan, Jiayu; Liu, Guorui; Zhao, Yuyang; Zheng, Minghui

    2016-09-01

    Pilot studies of unintentionally produced pollutants should be performed before waste being co-processed in cement kilns. Polychlorinated naphthalene (PCN) formation and emission from cement kilns co-processing sorted municipal solid waste, sewage sludge, and waste acid, however, have not previously been studied. Here, PCNs were analyzed in stack gas samples and solid samples from different stages of three cement production runs. PCN destruction efficiencies were higher when waste was co-processed (93.1% and 88.7% in two tests) than when waste was not co-processed (39.1%), so co-processing waste would not increase PCN outputs. The PCN concentrations were higher in particle samples from the C1 preheater and stages at back end of kiln than in particle samples from other stages, suggesting that cyclone preheater and back end of kiln should be focused for controlling PCN emissions. Besides that, based on the variation of PCN concentrations and corresponding operating conditions in different stages, the temperature, feeding materials, and chlorine content were suggested as the main factors influencing PCN formation. The PCN homologue and congener profiles suggested chlorination and dechlorination were the main PCN formation and decomposition pathways, and congeners CN-23, CN-46, and CN-59 appear to be appropriate indicators of PCNs emitted from coal-burning sources.

  14. Improvement, characterization and use of waste corn cob ash in cement-based materials

    Science.gov (United States)

    Suwanmaneechot, P.; Nochaiya, T.; Julphunthong, P.

    2015-12-01

    This work investigates the development of waste corn cob ash as supplementary cement replacement materials. The study focused on the effects of heat treatment on chemical composition, physical properties and engineering properties of corn cob ash. The results suggest corn cob ash that was heat treated at 600°C for 4 h shows percentage of SiO2 + Al2O3 + Fe2O3 around 72%, which can be classified as Class N calcined natural pozzolan, as prescribed by ASTM C618. The X-ray diffraction patterns indicated that the amorphous silica phase increased with increasing calcining temperatures. The water requirement, initial setting time and final setting time of specimens increased with increasing replacement percentage of raw or treated corn cob ash. The morta cubes which used 20% of treated corn cob ash replaced cement showed 103% of the 28 days compressive strength as compared to reference samples. The corn cob ash that was treated at 600°C for 4 h samples shows slightly higher effectiveness for improving the splitting tensile strength and compressive strength of concrete when compared to the untreated corn cob ash.

  15. Binding Materials of Dehydrated Phases of Waste Hardened Cement Paste and Pozzolanic Admixture

    Institute of Scientific and Technical Information of China (English)

    LU Linnu; HE Yongjia; HU Shuguang

    2009-01-01

    Fly ash (FA) and ground granulated blast-furnace slag (GGBFS) were added to improve the performances of regenerated binding materials (RBM) which refer to dehydrated phases with rebinding ability of waste hardened cement paste. Flowability tests, compressive strength tests,SEM, TG-DSC, and non-evaporable water content tests were employed to study the performances of the combined binding materials and the interactions between RBM, FA, and GGBFS. Results show that adding FA or GGBFS can improve the workability of RBM paste, and GGBFS has positive effects on strength of RBM. Pozzolanic reactions happen between RBM, FA, and GGBFS. And the activation effect of RBM to FA and GGBFS is superior to that of P.O grade-32.5 cement, especially at earlier ages, because of the high reactive f-CaO existing in RBM. On the advantages of the synergetic effects of RBM and pozzolanic admixtures such as FA and GGBFS, new combined binding materials can be prepared by blending them together.

  16. Contribution to the study of wastes stabilization by sulfo-aluminate cement

    International Nuclear Information System (INIS)

    Calcium sulfo-aluminate cement is mainly composed of yeelimite known to be a precursor of ettringite formation. Ettringite is able to incorporate several heavy metals by isomorphous substitutions without altering its crystalline structure. The design of a binder required for immobilizing heavy metals was undertaken. The hydration study of clinker, and cement containing 4 amounts of gypsum has been carried out by means of XRD, DTA and IR spectrometry. It was pointed out that the addition of gypsum enhances hydration. Two binders were selected: 80/20 and 70/30. The immobilisation of 7 pollutants was very successful. Nevertheless, damages appeared with the binder 70/30 containing sodium chromate and dichromate: sodium caused activation of yeelimite reactivity and important dissolution of gypsum leading to important ettringite production. With a great amount of gypsum (30 %), dissolution led to secondary ettringite formation which damaged the hardened paste. Adding polyol enhances the retention of sodium chromate. On the other hand, the immobilisation of two types of weakly radioactive wastes supplied by CEA has been made. Results obtained in terms of setting time, compressive strength and leaching were excellent. (author)

  17. Stabilization/solidification (S/S) of mercury-contaminated hazardous wastes using thiol-functionalized zeolite and Portland cement.

    Science.gov (United States)

    Zhang, Xin-Yan; Wang, Qi-Chao; Zhang, Shao-Qing; Sun, Xiao-Jing; Zhang, Zhong-Sheng

    2009-09-15

    Stabilization/solidification (S/S) of mercury-containing solid wastes using thiol-functionalized zeolite and cement was investigated in this study. The thiol-functionalized zeolite (TFZ) used in the study was obtained by grafting the thiol group (-SH) to the natural clinoptilolite zeolites, and the mercury adsorption by TFZ was investigated. TFZ was used to stabilize mercury in solid wastes, and then the stabilized wastes were subjected to cement solidification to test the effectiveness of the whole S/S process. The results show that TFZ has a high level of -SH content (0.562 mmol g(-1)) and the adsorption of mercury by TFZ conform to the Freundlich adsorption isotherm. The mercury adsorption capacity is greatly enhanced upon thiol grafting, the maximum of which is increased from 0.041 mmol Hg g(-1) to 0.445 mmol Hg g(-1). TFZ is found to be effective in stabilizing Hg in the waste surrogate. In the stabilization process, the optimum pH for the stabilization reaction is about 5.0. The optimum TFZ dosage is about 5% and the optimum cement dosage is about 100%. Though Cl(-) and PO(4)(3-) have negative effects on mercury adsorption by TFZ, the Portland cement solidification of TFZ stabilized surrogates containing 1000 mg Hg/kg can successfully pass the TCLP leaching test. It can be concluded that the stabilization/solidification process using TFZ and Portland cement is an effective technology to treat and dispose mercury-containing wastes. PMID:19376646

  18. Mobile calcination and cementation unit for solidification of concentrated radioactive wastes

    International Nuclear Information System (INIS)

    Mobile experimental unit MESA-1 was developed and manufactured for processing radioactive concentrates by direct cementation. The unit is mainly designed for processing low-level liquid wastes from nuclear power plants and other nuclear installations, in which the level of radioactivity does not exceed 1010 Bq/m3, the salt content of liquid solutions does not exceed 500 kg/m3 and the maximum amount of boric acid is 130 kg/m3. The equipment is built into three modules which may be assembled and dismantled in a short time and transported separately. The unit without the calciner module was tested in non-radioactive mode and in operation with actual radioactive wastes from the V-1 nuclear power plant. The course and results of the tests are described in detail. All project design values were achieved, a total of 18 dm3 model solutions were processed and 1 m3 of actual wastes with a salt content of 450 kg/m3. The test showed that with regard to the radiation level reached it will be necessary in the process of calcination to increase the shielding of certain exposed points. The calciner module is being assembled for completion. (Z.M.)

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

    International Nuclear Information System (INIS)

    Sulfur polymer cement (SPC) is made by reacting 95% sulfur with 2.5 % dicyclopentadiene and 2.5% cyclopentadiene oligomers, to produce a product that is much better than unmodified sulfur. SPC is being tested as a solidifying and stabilizing agent for low-level radioactive and hazardous wastes. Heavy loadings (5 wt%) of eight toxic metals were combined individually with SPC and 7 wt% sodium sulfide nonahydrate. The leach rates for mercury, lead, chromium and silver oxides were reduced by six orders of magnitude, while those of arsenic and barium were reduced by four. SPC is good for stabilizing incinerator ash. Ion-exchange resins can be stabilized with SPC after heat treatment with asbestos or diatomite at 220-250 deg C. 19 refs

  20. Suitability of Natural Rubber Latex and Waste Foundry Sand in Cement Concrete

    Directory of Open Access Journals (Sweden)

    Samuel Idiculla Thomas

    2016-06-01

    Full Text Available Suitability of Natural Rubber Latex (NRL as an additive and Waste Foundry Sand (WFS as partial replacement to river sand, in cement concrete was investigated. Experimental study was performed with concrete mixtures containing 1% latex to water ratio, along with 5% and 10% replacement of river sand by WFS. Properties of concrete were studied in both fresh and hardened state. The results of laboratory tests indicate that WFS and NRL reduces the workability of concrete. Slight reduction in splitting tensile strength was observed for mixtures containing NRL and WFS, in comparison to conventional mix. No specific trend was observed for flexural strength at 7 days, but at 28 days the difference was within ±3%, when compared to conventional mix. Strength development for mixtures containing NRL and WFS was slightly lower than conventional mix. The limited results of this study show that concrete containing NRL and WFS do have potential for use as non- structural concrete.

  1. Crystalline Ceramic Waste Forms: Comparison Of Reference Process For Ceramic Waste Form Fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Brinkman, K. S. [Savannah River National Laboratory; Marra, J. C. [Savannah River National Laboratory; Amoroso, J. [Savannah River National Laboratory; Tang, M. [Los Alamos National Laboratory

    2013-08-22

    The research conducted in this work package is aimed at taking advantage of the long term thermodynamic stability of crystalline ceramics to create more durable waste forms (as compared to high level waste glass) in order to reduce the reliance on engineered and natural barrier systems. Durable ceramic waste forms that incorporate a wide range of radionuclides have the potential to broaden the available disposal options and to lower the storage and disposal costs associated with advanced fuel cycles. Assemblages of several titanate phases have been successfully demonstrated to incorporate radioactive waste elements, and the multiphase nature of these materials allows them to accommodate variation in the waste composition. Recent work has shown that they can be produced from a melting and crystallization process. The objective of this report is to explore the phase formation and microstructural differences between lab scale melt processing in varying gas environments with alternative densification processes such as Hot Pressing (HP) and Spark Plasma Sintering (SPS). The waste stream used as the basis for the development and testing is a simulant derived from a combination of the projected Cs/Sr separated stream, the Trivalent Actinide - Lanthanide Separation by Phosphorous reagent Extraction from Aqueous Komplexes (TALSPEAK) waste stream consisting of lanthanide fission products, the transition metal fission product waste stream resulting from the transuranic extraction (TRUEX) process, and a high molybdenum concentration with relatively low noble metal concentrations. Melt processing as well as solid state sintering routes SPS and HP demonstrated the formation of the targeted phases; however differences in microstructure and elemental partitioning were observed. In SPS and HP samples, hollandite, pervoskite/pyrochlore, zirconolite, metallic alloy and TiO{sub 2} and Al{sub 2}O{sub 3} were observed distributed in a network of fine grains with small residual pores

  2. Final report on cermet high-level waste forms

    International Nuclear Information System (INIS)

    Cermets are being developed as an alternate method for the fixation of defense and commercial high level radioactive waste in a terminal disposal form. Following initial feasibility assessments of this waste form, consisting of ceramic particles dispersed in an iron-nickel base alloy, significantly improved processing methods were developed. The characterization of cermets has continued through property determinations on samples prepared by various methods from a variety of simulated and actual high-level wastes. This report describes the status of development of the cermet waste form as it has evolved since 1977. 6 tables, 18 figures

  3. Periodic and uniform nanogratings formed on cemented carbide by femtosecond laser scanning

    International Nuclear Information System (INIS)

    Periodic and uniform nanogratings are fabricated by femtosecond laser scanning on cemented carbide. Specifically, three experiments are designed to study the influence of single pulse energy, scanning speed, and scanning spacing on the period and the uniformity of the formed nanogratings. The results show that the sample with single pulse energy of 2 μJ, scanning speed of 1000 μm/s, and scanning spacing of 5 μm shows the best quality of nanogratings among all the tested samples at different processing parameters. The uniformity of the nanogratings is largely determined by single pulse energy, scanning speed, and scanning spacing. Single pulse energy and scanning speed significantly affect the period of the nanogratings, whereas the period of the nanogratings maintains a fixed value under different scanning spacings. The period of the nanogratings increases gradually with the decrease of the single pulse energy and the increase of the scanning speed, respectively.

  4. 水泥窑处理工业废物的工厂实验研究%Plant Test of Industrial Waste Disposal in a Cement Kiln

    Institute of Scientific and Technical Information of China (English)

    刘阳生; 韩杰; 白庆中

    2003-01-01

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

  5. CHEMICALLY BONDED CEMENTS FROM BOILER ASH AND SLUDGE WASTES. PHASE II REPORT, SEPT.1998-JULY 1999.

    Energy Technology Data Exchange (ETDEWEB)

    SUGAMA,T.YAGER,K.A.BLANKENHORN,D.(KEYSPAN R AND D INITIATIVE)

    1999-08-01

    Based upon the previous Phase I research program aimed at looking for ways of recycling the KeySpan-generated wastes, such as waste water treatment sludge (WWTS) and bottom ash (BA), into the potentially useful cementitious materials called chemically bonded cement (CBC) materials, the emphasis of this Phase II program done at Brookhaven National Laboratory, in a period of September 1998 through July 1999, was directed towards the two major subjects: One was to assess the technical feasibility of WWTS-based CBC material for use as Pb-exchange adsorbent (PEA) which remediates Pb-contaminated soils in the field; and the other was related to the establishment of the optimum-packaging storage system of dry BA-based CBC components that make it a promising matrix material for the steam-cured concrete products containing sand and coarse aggregate. To achieve the goal of the first subject, a small-scale field demonstration test was carried out. Using the PEA material consisting of 30 wt% WWTS, 13 wt% Type I cement and 57 wt% water, the PES slurry was prepared using a rotary shear concrete mixer, and then poured on the Pb-contaminated soil. The PEA-to-soil ratio by weight was a factor of 2.0. The placed PEA slurry was blended with soil using hand mixing tools such as claws and shovels. The wettability of soils with the PEA was very good, thereby facilitating the soil-PEA mix procedures. A very promising result was obtained from this field test; in fact, the mount of Pb leached out from the 25-day-aged PEA-treated soil specimen was only 0.74 mg/l, meeting the requirement for EPA safe regulation of < 5 mg/l. In contrast, a large amount (26.4 mg/l) of Pb was detected from the untreated soil of the same age. Thus, this finding demonstrated that the WWTS-based CBC has a potential for use as PEA material. Regarding the second subject, the dry-packed storage system consisting of 68.7 wt% BA, 13.0 wt% calcium aluminate cement (CAC), 13.0 wt% Type I portland cement and 5.3 wt

  6. Experimental research on the strength of cemented backfilling body of waste rocks%废石尾砂胶结充填体强度试验研究

    Institute of Scientific and Technical Information of China (English)

    罗根平; 乔登攀

    2015-01-01

    Experimental study is systematically conducted on cemented backfilling with waste rocks.The paper states the applicability and mechanism of waste rock cemented filling process and focuses on the influencing factors on the strength of cemented filling body of waste rocks,namely the water-cement ratio,cement-sand ratio,cement content, the grading and proportioning of the particle size of waste rocks.The research results show that the lager the water-ce-ment ratio and cement-sand ratio are,the less the strength of cemented backfilling body becomes,contrary to that rela-tion between cement content and the backfilling body's strength.With constant strength,cemented filling with waste rocks consumes less cement per unit volume and cost less than other filling methods.%对废石尾砂胶结充填进行了系统的试验研究。阐述了废石尾砂胶结充填工艺的工业性及原理,着重研究了废石尾砂胶结充填体强度的影响因素:水灰比、灰砂比、水泥含量、废石尾砂的粒径级配及配比。研究结果表明,废石尾砂胶结充填体强度随水灰比、灰砂比的减小而增大,随水泥含量的增加而增加。在强度一定的条件下,废石尾砂胶结充填比其他充填方式,单位体积内水泥耗量少,成本低。

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

    Energy Technology Data Exchange (ETDEWEB)

    Bernadzikowski, T. A.; Allender, J. S.; Butler, J. L.; Gordon, D. E.; Gould, Jr., T. H.; Stone, J. A.

    1982-03-01

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

  8. Challenges in Modeling the Degradation of Ceramic Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Devanathan, Ramaswami; Gao, Fei; Sun, Xin

    2011-09-01

    We identify the state of the art, gaps in current understanding, and key research needs in the area of modeling the long-term degradation of ceramic waste forms for nuclear waste disposition. The directed purpose of this report is to define a roadmap for Waste IPSC needs to extend capabilities of waste degradation to ceramic waste forms, which overlaps with the needs of the subconsinuum scale of FMM interests. The key knowledge gaps are in the areas of (i) methodology for developing reliable interatomic potentials to model the complex atomic-level interactions in waste forms; (ii) characterization of water interactions at ceramic surfaces and interfaces; and (iii) extension of atomic-level insights to the long time and distance scales relevant to the problem of actinide and fission product immobilization.

  9. Preparation and leaching of radioactive INEL waste forms

    International Nuclear Information System (INIS)

    Appreciable quantities of radioactive waste are in storage at the Idaho National Engineering Laboratory (INEL). Plans are being made to convert this waste into durable solid forms for final disposal in a geological repository. Part of the inventory consists of low- and intermediate-level fission, activation, and decay products and transuranic (TRU) wastes, either stored retrievably or buried at the INEL Radioactive Waste Management area. One of the TRU wastes is a sludge from the Department of Energy Rocky Flats Plant, currently stored retrievably in 55-gallon drums. Immobilizing the TRU sludge is the primary concern of the work reported here

  10. Characterization of low and medium-level radioactive waste forms. Final report - 2nd Programme 1980-84

    International Nuclear Information System (INIS)

    The European Communities Second R and D Programme 1980-84 'Management and Disposal of Radioactive Waste (Shared cost action)' included a closely coordinated research activity for the 'Characterization of low and medium-level radioactive waste forms'. This report summarizes the main results obtained during the five years of the programme by laboratories in seven European countries participating in the coordinated RandD efforts. Ten reference waste forms have been selected, based on the most important types of low and medium-level waste arisings and the three commonly used immobilization matrices: cement, bitumen and polymers. The investigated properties were mainly: waste-matrix compatibility, radiation effects, leaching behaviour, leached radionuclides speciation, microbiological resistance and thermal as well as mechanical properties. Extensive experimental results relevant for the qualification of waste products and for application in performance analysis are presented in this final report. The main conclusions are drawn for the confinement properties of these different waste forms. These conclusions have also shown the necessity of selecting several other reference waste forms for the continuation of this RandD action now being launched in the Third EC Programme 1985-89

  11. Development of iron phosphate ceramic waste form to immobilize radioactive waste solution

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jongkwon; Um, Wooyong; Choung, Sungwook

    2014-05-09

    The objective of this research was to develop an iron phosphate ceramic (IPC) waste form using converter slag obtained as a by-product of the steel industry as a source of iron instead of conventional iron oxide. Both synthetic off-gas scrubber solution containing technetium-99 (or Re as a surrogate) and LiCl-KCl eutectic salt, a final waste solution from pyrochemical processing of spent nuclear fuel, were used as radioactive waste streams. The IPC waste form was characterized for compressive strength, reduction capacity, chemical durability, and contaminant leachability. Compressive strengths of the IPC waste form prepared with different types of waste solutions were 16 MPa and 19 MPa for LiCl-KCl eutectic salt and the off-gas scrubber simulant, respectively, which meet the minimum compressive strength of 3.45 MPa (500 psi) for waste forms to be accepted into the radioactive waste repository. The reduction capacity of converter slag, a main dry ingredient used to prepare the IPC waste form, was 4,136 meq/kg by the Ce(IV) method, which is much higher than those of the conventional Fe oxides used for the IPC waste form and the blast furnace slag materials. Average leachability indexes of Tc, Li, and K for the IPC waste form were higher than 6.0, and the IPC waste form demonstrated stable durability even after 63-day leaching. In addition, the Toxicity Characteristic Leach Procedure measurements of converter slag and the IPC waste form with LiCl-KCl eutectic salt met the universal treatment standard of the leachability limit for metals regulated by the Resource Conservation and Recovery Act. This study confirms the possibility of development of the IPC waste form using converter slag, showing its immobilization capability for radionuclides in both LiCl-KCl eutectic salt and off-gas scrubber solutions with significant cost savings.

  12. Development of iron phosphate ceramic waste form to immobilize radioactive waste solution

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jongkwon [Division of Advanced Nuclear Engineering, Pohang University of Science and Technology (POSTECH), San 31, Hyoja-Dong, Pohang (Korea, Republic of); Um, Wooyong, E-mail: wooyong.um@pnnl.gov [Division of Advanced Nuclear Engineering, Pohang University of Science and Technology (POSTECH), San 31, Hyoja-Dong, Pohang (Korea, Republic of); Pacific Northwest National Laboratory, Richland, WA 99354 (United States); Choung, Sungwook [Division of Advanced Nuclear Engineering, Pohang University of Science and Technology (POSTECH), San 31, Hyoja-Dong, Pohang (Korea, Republic of)

    2014-09-15

    The objective of this research was to develop an iron phosphate ceramic (IPC) waste form using converter slag obtained as a by-product of the steel industry as a source of iron instead of conventional iron oxide. Both synthetic off-gas scrubber solution containing technetium-99 (or Re as a surrogate) and LiCl–KCl eutectic salt, a final waste solution from pyrochemical processing of spent nuclear fuel, were used as radioactive waste streams. The IPC waste form was characterized for compressive strength, reduction capacity, chemical durability, and contaminant leachability. Compressive strengths of the IPC waste form prepared with different types of waste solutions were 16 MPa and 19 MPa for LiCl–KCl eutectic salt and the off-gas scrubber simulant, respectively, which meet the minimum compressive strength of 3.45 MPa (500 psi) for waste forms to be accepted into the radioactive waste repository. The reduction capacity of converter slag, a main dry ingredient used to prepare the IPC waste form, was 4136 meq/kg by the Ce(IV) method, which is much higher than those of the conventional Fe oxides used for the IPC waste form and the blast furnace slag materials. Average leachability indexes of Tc, Li, and K for the IPC waste form were higher than 6.0, and the IPC waste form demonstrated stable durability even after 63-day leaching. In addition, the Toxicity Characteristic Leach Procedure measurements of converter slag and the IPC waste form with LiCl–KCl eutectic salt met the universal treatment standard of the leachability limit for metals regulated by the Resource Conservation and Recovery Act. This study confirms the possibility of development of the IPC waste form using converter slag, showing its immobilization capability for radionuclides in both LiCl–KCl eutectic salt and off-gas scrubber solutions with significant cost savings.

  13. Development of iron phosphate ceramic waste form to immobilize radioactive waste solution

    Science.gov (United States)

    Choi, Jongkwon; Um, Wooyong; Choung, Sungwook

    2014-09-01

    The objective of this research was to develop an iron phosphate ceramic (IPC) waste form using converter slag obtained as a by-product of the steel industry as a source of iron instead of conventional iron oxide. Both synthetic off-gas scrubber solution containing technetium-99 (or Re as a surrogate) and LiCl-KCl eutectic salt, a final waste solution from pyrochemical processing of spent nuclear fuel, were used as radioactive waste streams. The IPC waste form was characterized for compressive strength, reduction capacity, chemical durability, and contaminant leachability. Compressive strengths of the IPC waste form prepared with different types of waste solutions were 16 MPa and 19 MPa for LiCl-KCl eutectic salt and the off-gas scrubber simulant, respectively, which meet the minimum compressive strength of 3.45 MPa (500 psi) for waste forms to be accepted into the radioactive waste repository. The reduction capacity of converter slag, a main dry ingredient used to prepare the IPC waste form, was 4136 meq/kg by the Ce(IV) method, which is much higher than those of the conventional Fe oxides used for the IPC waste form and the blast furnace slag materials. Average leachability indexes of Tc, Li, and K for the IPC waste form were higher than 6.0, and the IPC waste form demonstrated stable durability even after 63-day leaching. In addition, the Toxicity Characteristic Leach Procedure measurements of converter slag and the IPC waste form with LiCl-KCl eutectic salt met the universal treatment standard of the leachability limit for metals regulated by the Resource Conservation and Recovery Act. This study confirms the possibility of development of the IPC waste form using converter slag, showing its immobilization capability for radionuclides in both LiCl-KCl eutectic salt and off-gas scrubber solutions with significant cost savings.

  14. Reference waste forms and packing material for the Nevada Nuclear Waste Storage Investigations Project

    International Nuclear Information System (INIS)

    The Lawrence Livermore National Laboratory (LLNL), Livermore, Calif., has been given the task of designing and verifying the performance of waste packages for the Nevada Nuclear Waste Storage Investigations (NNWSI) Project. NNWSI is studying the suitability of the tuffaceous rocks at Yucca Mountain, Nevada Test Site, for the potential construction of a high-level nuclear waste repository. This report gives a summary description of the three waste forms for which LLNL is designing waste packages: spent fuel, either as intact assemblies or as consolidated fuel pins, reprocessed commercial high-level waste in the form of borosilicate glass, and reprocessed defense high-level waste from the Defense Waste Processing Facility in Aiken, S.C. Reference packing material for use with the alternative waste package design for spent fuel is also described. 14 references, 8 figures, 20 tables

  15. Self-Shrinkage Behaviors of Waste Paper Fiber Reinforced Cement Paste considering Its Self-Curing Effect at Early-Ages

    Directory of Open Access Journals (Sweden)

    Zhengwu Jiang

    2016-01-01

    Full Text Available The aim of this paper was to study how the early-age self-shrinkage behavior of cement paste is affected by the addition of the waste paper fibers under sealed conditions. Although the primary focus was to determine whether the waste paper fibers are suitable to mitigate self-shrinkage as an internal curing agent under different adding ways, evaluating their strength, pore structure, and hydration properties provided further insight into the self-cured behavior of cement paste. Under the wet mixing condition, the waste paper fibers could mitigate the self-shrinkage of cement paste and, at additions of 0.2% by mass of cement, the waste paper fibers were found to show significant self-shrinkage cracking control while providing some internal curing. In addition, the self-curing efficiency results were analyzed based on the strength and the self-shrinkage behaviors of cement paste. Results indicated that, under a low water cement ratio, an optimal dosage and adding ways of the waste paper fibers could enhance the self-curing efficiency of cement paste.

  16. Influence of chemical composition of civil construction waste in the cement paste; Influencia da composicao quimica dos residuos da construcao civil a pasta de cimento

    Energy Technology Data Exchange (ETDEWEB)

    Cunha, G.A.; Andrade, A.C.D.; Souza, J.M.M.; Evangelista, A.C.J.; Almeida, V.C., E-mail: valeria@eq.ufrj.b [Universidade Federal do Rio de Janeiro (EQ/UFRJ), RJ (Brazil). Escola de Quimica

    2009-07-01

    The construction and demolition waste when disposed inappropriately might cause serious public health problems. Its reutilization focusing on the development of new products using simple production techniques, assuring a new product life cycle and not damaging the environment is inserted in sustainable concept. The aim of this work was identifying the characteristics of types of waste generated in a residential reform (glassy ceramic and fill dirt leftovers) verifying separately its influence on cement pastes mechanical behavior. Cement pastes + wastes were prepared in 25% and 50% proportions with an approximately 0,35 water/cement relation and, glue time determination, water absorption, resistance to compression and X-ray fluorescence assays were taken. The results indicate that the chemical composition of the waste causes changes in the behavior of cement pastes, reflecting on their resistance to compression. (author)

  17. Influence of natural pozzolan, colemanite ore waste, bottom ash, and fly ash on the properties of Portland cement

    Energy Technology Data Exchange (ETDEWEB)

    Targan, S.; Olgun, A.; Erdogan, Y.; Sevinc, V. [Dumlupinar University, Kutahya (Turkey). Dept. of Chemistry

    2003-08-01

    The effect of natural pozzolan (NP), colemanite ore waste (CW), coal fly ash (FA), and coal bottom ash (BA) on the properties of cement and concrete was examined. The parameters studied included compressive strength, bending strength, volume expansion, and setting time. A number of cements were prepared (in the presence of fixed quantity of 10% FA, 10% BA, and 4% CW) by the replacement of Portland cement (PC) with NP in range of 5 - 30%. The results showed that the final setting time of cement pastes were generally accelerated when the NP replaced part of the cement. However, NP exhibited a significant retarding effect when used in combination with CW. The results also showed that the inclusion of NP at replacement levels of 5% resulted in an increase in compressive strength of the specimens compared with that of the control concrete. The replacement of PC by 10 - 15% of NP in the presence of fixed quantity of CW improves the bending strength of the specimens compared with control specimens after 60 days of curing age.

  18. Comparison of glass and crystalline nuclear waste forms

    International Nuclear Information System (INIS)

    Nuclear waste forms may be divided into two broad categories: single phase glasses with minor crystalline components (e.g., borosilicate glasses) and crystalline waste forms, either single phase (e.g., monazite) or polyphase (e.g., SYNROC). This paper reviews the materials properties data that are available for each of these two types of waste forms. The principal data include: physical, thermal and mechanical properties, chemical durability; and radiation damage effects. Complete data are only available for borosilicate glasses and SYNROC; therefore, this comparison focuses on the performance assessment of borosilicate glass and SYNROC

  19. The Next Generation Ecological Self Compacting Concrete with Glass Waste Powder as a Cement Component in Concrete and Recycled Concrete Aggregates

    OpenAIRE

    Kara, P

    2013-01-01

    In the present study the performance characteristics (workability, compressive strength, frost resistance, permeability and temperature of hydration) of the ecological self compacting concrete with reduced cement content and with the next generation recycled concrete aggregates which are obtained from crashed concrete specimens with cement substitution at level of 30% with waste glass powder were investigated. Waste glass as powder ground to certain fineness accelerates beneficial chemical re...

  20. Use of Different Barium Salts to Inhibit the Thaumasite Form of Sulfate Attack in Cement-based Materials

    Institute of Scientific and Technical Information of China (English)

    SU Ying; WEI Xiaochao; HUANG Jian; WANG Yingbin; HE Xingyang; WANG Xiongjue; MA Baoguo

    2016-01-01

    We investigated the effects of different barium compounds on the thaumasite form of sulphate attack (TSA) resistance of cement-based materials when they were used as admixtures in mortars. Moreover, we analyzed the inhibition mechanisms within different types of barium salts, namely BaCO3 and Ba(OH)2, on the thaumasite formation. The control cement mortar and mortars with barium salts to cement and limestone weight ratios of 0.5%, 1.0%, and 1.5% were immersed in 5% (by weight) MgSO4 solution at 5℃ to mimic TSA. Appearance, mass, and compressive strength of the mortar samples were monitored and measured to assess the general degradation extent of these samples. The products of sulphate attack were further analyzed by XRD, FTIR, and SEM, respectively. Experimental results show that different degradation extent is evident in all mortars cured in MgSO4 solution. However, barium salts can greatly inhibit such degradation. Barium in hydroxide form has better effectiveness in protection against TSA than carbonate form, which may be due to their solubility difference in alkaline cement pore solution, and the presence of these barium compounds can reduce the degree of TSA by comparison with the almost completely decomposed control samples.

  1. A study on the acceptance criteria of radioactive waste form

    International Nuclear Information System (INIS)

    It is essential to accept well solidified and packaged waste forms for the safety during the operational and post operational phase in the repository, and for this, waste the acceptance criterion is necessary for the distinction of the well solidified and packaged waste form. The objective of this report is to provide the preliminary acceptance criteria to help the later establishment of final acceptance criteria. The following factors were considered for establishing the preliminary waste acceptance criteria. 1) Matrix and waste form characteristics 2) the type of repository and its characteristics 3) establishment procedure of acceptance criteria and its technical background From this study, a qualitative preliminary criterion including the radionuclide contents, surface dose, surface contamination and so on was established. (Author)

  2. Disposition of excess plutonium using ''off-spec'' MOX pellets as a sintered ceramic waste form

    International Nuclear Information System (INIS)

    The authors describe a potential strategy for the disposition of excess weapons plutonium in a way that minimizes (1) technological risks, (2) implementation costs and completion schedules, and (3) requirements for constructing and operating new or duplicative Pu disposition facilities. This is accomplished by an optimized combination of (1) using existing nuclear power reactors to ''burn'' relatively pure excess Pu inventories as mixed oxide (MOX) fuel and (2) using the same MOX fuel fabrication facilities to fabricate contaminated or impure excess Pu inventories into an ''off-spec'' MOX solid ceramic waste form for geologic disposition. Diversion protection for the SCWF to meet the ''spent fuel standard'' introduced by the National Academy of Sciences can be achieved in at least three ways. (1) One can utilize the radiation field from defense high-level nuclear waste by first packaging the SCWF pellets in 2- to 4-L cans that are subsequently encapsulated in radioactive glass in the Defense Waste Processing Facility (DWPF) glass canisters (a ''can-in-canister'' approach). (2) One can add 137Cs (recovered from defense wastes at Hanford and currently stored as CsCl in capsules) to an encapsulating matrix such as cement for the SCWF pellets in a small hot-cell facility and thus fabricate large monolithic forms. (3) The SCWF can be fabricated into reactor fuel-like pellets and placed in tubes similar to fuel assemblies, which can then be mixed in sealed repository containers with irradiated spent nuclear fuel for geologic disposition

  3. The necessity for scale up in R and D: Approach for waste immobilization in cement by BNFL

    International Nuclear Information System (INIS)

    The full scale processing of nuclear wastes immobilized in cement utilizes a wide range of chemical and physical parameters. The success of this work however, involves many factors and material properties which are affected by the actual scaling up processes. The paper outlines the approach and experience gained by British Nuclear Fuels plc (BNFL) to recognize and evaluate the major factors involved in order to successfully produce large scale stable products acceptable to the appropriate regulatory bodies and suitable for long term disposal

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

    Directory of Open Access Journals (Sweden)

    López-Zaldívar, O.

    2015-09-01

    Full Text Available This article presents the results of a research developing high performance cement mortars with the addition of municipal solid waste incineration fly ash (MSWIFA stabilized as insoluble carbonates. The encapsulation of hazardous wastes in mortar matrixes has also been achieved. The ashes present high concentrations of chlorides, Zn and Pb. A stabilization process with NaHCO3 has been developed reducing 99% the content of chlorides. Developed mortars replace 10% per weight of the aggregates by treated MSWIFA. Physical/mechanical properties of these mortars have been studied. Presence of Zn, Pb, Cu and Cd has been also analyzed confirming that leaching of these heavy metal ions is mitigated. Conclusions prove better behavior of CAC and CSA mortars than those of CEM-I and CEM-II cement. Results are remarkable for the CAC mortars, improving reference strengths in more than 25%, which make them a fast-curing product suitable for the repair of structures or industrial pavements.Este artículo presenta los resultados del desarrollo de morteros mejorados con la incorporación de cenizas volantes de residuos sólidos urbanos inertizadas en forma de carbonatos. Además se consigue la encapsulación de un residuo peligroso. Las cenizas presentan una alta concentración de cloruros, Zn y Pb. Se ha desarrollado un proceso de estabilización con NaHCO3 reduciendo en un 99% el contenido de cloruros. Los morteros reemplazan un 10% en peso del árido por cenizas tratadas. Se han analizado sus propiedades físico/mecánicas y la presencia de Zn, Pb, Cu y Cd. Se demuestra un mejor comportamiento de los morteros de CAC y CSA que los de CEM-I y CEM-II y se mitiga el lixiviado de metales pesados. Los resultados son significativos en los morteros CAC al mejorar las resistencias de los de referencia en un 25%. Los morteros desarrollados son de curado rápido adecuados para la reparación de estructuras o soleras industriales.

  5. Evaluation of solidified high-level waste forms

    International Nuclear Information System (INIS)

    One of the objectives of the IAEA waste management programme is to coordinate and promote development of improved technology for the safe management of radioactive wastes. The Agency accomplished this objective specifically through sponsoring Coordinated Research Programmes on the ''Evaluation of Solidified High Level Waste Products'' in 1977. The primary objectives of this programme are to review and disseminate information on the properties of solidified high-level waste forms, to provide a mechanism for analysis and comparison of results from different institutes, and to help coordinate future plans and actions. This report is a summary compilation of the key information disseminated at the second meeting of this programme

  6. Pelleted waste form for high-level ICPP wastes

    International Nuclear Information System (INIS)

    Simulated zirconia-type calcined waste is pelletized on a 41-cm diameter disc pelletizer using 5% bentonite, 2% metakaolin, and 2% boric acid as a solid binder and 7M phosphoric plus 4M nitric acid as a liquid binder. After heat treatment at 8000C for 2 hours the pellets are impact resistant and have a leach resistance of 10-4 g/cm2 . day, based on Soxhlet leaching for 100 hours at 950C with distilled water. An integrated pilot plant is being fabricated to verify the process. 1 figure, 4 tables

  7. Engineering properties of cement mortar with pond ash in South Korea as construction materials: from waste to concrete

    Science.gov (United States)

    Jung, Sang; Kwon, Seung-Jun

    2013-09-01

    Among the wastes from coal combustion product, only fly ash is widely used for mineral mixture in concrete for its various advantages. However the other wastes including bottom ash, so called PA (pond ash) are limitedly reused for reclamation. In this paper, the engineering properties of domestic pond ash which has been used for reclamation are experimentally studied. For this, two reclamation sites (DH and TA) in South Korea are selected, and two domestic PAs are obtained. Cement mortar with two different w/c (water to cement) ratios and 3 different replacement ratios (0%, 30%, and 60%) of sand are prepared for the tests. For workability and physical properties of PA cement mortar, several tests like flow, setting time, and compressive strength are evaluated. Several durability tests including porosity measuring, freezing and thawing, chloride migration, and accelerated carbonation are also performed. Through the tests, PA (especially from DH area) in surface saturated condition is evaluated to have internal curing action which leads to reasonable strength development and durability performances. The results show a potential applicability of PA to concrete aggregate, which can reduce consuming natural resources and lead to active reutilization of coal product waste.

  8. Transuranic contaminated waste form characterization and data base

    International Nuclear Information System (INIS)

    This report outlines the sources, quantities, characteristics and treatment of transuranic wastes in the United States. This document serves as part of the data base necessary to complete preparation and initiate implementation of transuranic wastes, waste forms, waste container and packaging standards and criteria suitable for inclusion in the present NRC waste management program. No attempt is made to evaluate or analyze the suitability of one technology over another. Indeed, by the nature of this report, there is little critical evaluation or analysis of technologies because such analysis is only appropriate when evaluating a particular application or transuranic waste streams. Due to fiscal restriction, the data base is developed from a myriad of technical sources and does not necessarily contain operating experience and the current status of all technologies. Such an effort was beyond the scope of this report

  9. Transuranic contaminated waste form characterization and data base

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-07-01

    This report outlines the sources, quantities, characteristics and treatment of transuranic wastes in the United States. This document serves as part of the data base necessary to complete preparation and initiate implementation of transuranic wastes, waste forms, waste container and packaging standards and criteria suitable for inclusion in the present NRC waste management program. No attempt is made to evaluate or analyze the suitability of one technology over another. Indeed, by the nature of this report, there is little critical evaluation or analysis of technologies because such analysis is only appropriate when evaluating a particular application or transuranic waste streams. Due to fiscal restriction, the data base is developed from a myriad of technical sources and does not necessarily contain operating experience and the current status of all technologies. Such an effort was beyond the scope of this report.

  10. Actinide Waste Forms and Radiation Effects

    Science.gov (United States)

    Ewing, R. C.; Weber, W. J.

    Over the past few decades, many studies of actinides in glasses and ceramics have been conducted that have contributed substantially to the increased understanding of actinide incorporation in solids and radiation effects due to actinide decay. These studies have included fundamental research on actinides in solids and applied research and development related to the immobilization of the high level wastes (HLW) from commercial nuclear power plants and processing of nuclear weapons materials, environmental restoration in the nuclear weapons complex, and the immobilization of weapons-grade plutonium as a result of disarmament activities. Thus, the immobilization of actinides has become a pressing issue for the twenty-first century (Ewing, 1999), and plutonium immobilization, in particular, has received considerable attention in the USA (Muller et al., 2002; Muller and Weber, 2001). The investigation of actinides and

  11. Analytical electron microscopy study of radioactive ceramic waste forms

    International Nuclear Information System (INIS)

    A ceramic waste form has been developed to immobilize the halide high-level waste stream from electrometallurgical treatment of spent nuclear fuel. Analytical electron microscopy studies, using both scanning and transmission instruments, have been performed to characterize the microstructure of this material. The microstructure consists primarily of sodalite granules (containing the bulk of the halides) bonded together with glass. The results of these studies are discussed in detail. Insight into the waste form fabrication process developed as a result of these studies is also discussed

  12. Evolution of 99Tc Species in Cementitious Nuclear Waste Form

    International Nuclear Information System (INIS)

    Technetium (Tc) is produced in large quantities as a fission product during the irradiation of 235U-enriched fuel for commercial power production and plutonium genesis for nuclear weapons. The most abundant isotope of Tc present in the wastes is 99Tc because of its high fission yield (∼6%) and long half-life (2.13x105 years). During the Cold War era, generation of fissile 239Pu for use in America's atomic weapons arsenal yielded nearly 1900 kg of 99Tc at the U.S. Department of Energy's (DOE) Hanford Site in southeastern Washington State. Most of this 99Tc is present in fuel reprocessing wastes temporarily stored in underground tanks awaiting retrieval and permanent disposal. After the wastes are retrieved from the storage tanks, the bulk of the high-level waste (HLW) and lowactivity waste (LAW) stream is scheduled to be converted into a borosilicate glass waste form that will be disposed of in a shallow burial facility called the Integrated Disposal Facility (IDF) at the Hanford Site. Even with careful engineering controls, volatilization of a fraction of Tc during the vitrification of both radioactive waste streams is expected. Although this volatilized Tc can be captured in melter off-gas scrubbers and returned to the melter, some of the Tc is expected to become part of the secondary waste stream from the vitrification process. The off-gas scrubbers downstream from the melters will generate a high pH, sodium-ammonium carbonate solution containing the volatilized Tc and other fugitive species. Effective and cost-efficient disposal of Tc found in the off-gas scrubber solution remains difficult. A cementitious waste form (Cast Stone) is one of the nuclear waste form candidates being considered to solidify the secondary radioactive liquid waste that will be generated by the operation of the waste treatment plant (WTP) at the Hanford Site. Because Tc leachability from the waste form is closely related with Tc speciation or oxidation state in both the simulant and

  13. Distributions, profiles and formation mechanisms of polychlorinated naphthalenes in cement kilns co-processing municipal waste incinerator fly ash.

    Science.gov (United States)

    Liu, Guorui; Zhan, Jiayu; Zhao, Yuyang; Li, Li; Jiang, Xiaoxu; Fu, Jianjie; Li, Chunping; Zheng, Minghui

    2016-07-01

    Co-processing municipal solid waste incinerator (MSWI) fly ash in cement kilns is challenging because the unintentional production of persistent organic pollutants (POPs) during the process is not well understood. The distributions, profiles and formation mechanisms of polychlorinated naphthalenes (PCNs) as new POPs covered under Stockholm Convention in two cement kilns co-processing MSWI fly ash were studied. The average concentrations of PCNs in stack gas samples were 710 ng m(-3). The PCN concentration in particle samples collected from different process stages in the cement kilns ranged from 1.1 to 84.7 ng g(-1). Three process sites including suspension pre-heater boiler, humidifier tower, and the kiln back-end bag filter were identified to be the major formation sites of PCNs in cement kilns co-processing MSWI fly ash. The PCN distribution patterns were similar to that of polychlorinated dibenzo-p-dioxin and dibenzofuran (PCDD/Fs), which indicates the possibility for simultaneous control of PCNs and PCDD/Fs in cement kilns co-processing fly ash. Chlorination was suggested to be an important formation mechanism of PCNs, and chlorination pathways of PCN congeners are proposed based on the congener profiles. Thermodynamic calculations, including relative thermal energies (ΔE) and standard free energy of formation (ΔG), and the charge densities of the carbon atoms in PCN supported the proposed chlorination mechanisms for PCN formation. The results presented in this study might provide helpful information for developing techniques and strategies to control PCN emissions during cement kilns co-processing MSWI fly ash.

  14. Distributions, profiles and formation mechanisms of polychlorinated naphthalenes in cement kilns co-processing municipal waste incinerator fly ash.

    Science.gov (United States)

    Liu, Guorui; Zhan, Jiayu; Zhao, Yuyang; Li, Li; Jiang, Xiaoxu; Fu, Jianjie; Li, Chunping; Zheng, Minghui

    2016-07-01

    Co-processing municipal solid waste incinerator (MSWI) fly ash in cement kilns is challenging because the unintentional production of persistent organic pollutants (POPs) during the process is not well understood. The distributions, profiles and formation mechanisms of polychlorinated naphthalenes (PCNs) as new POPs covered under Stockholm Convention in two cement kilns co-processing MSWI fly ash were studied. The average concentrations of PCNs in stack gas samples were 710 ng m(-3). The PCN concentration in particle samples collected from different process stages in the cement kilns ranged from 1.1 to 84.7 ng g(-1). Three process sites including suspension pre-heater boiler, humidifier tower, and the kiln back-end bag filter were identified to be the major formation sites of PCNs in cement kilns co-processing MSWI fly ash. The PCN distribution patterns were similar to that of polychlorinated dibenzo-p-dioxin and dibenzofuran (PCDD/Fs), which indicates the possibility for simultaneous control of PCNs and PCDD/Fs in cement kilns co-processing fly ash. Chlorination was suggested to be an important formation mechanism of PCNs, and chlorination pathways of PCN congeners are proposed based on the congener profiles. Thermodynamic calculations, including relative thermal energies (ΔE) and standard free energy of formation (ΔG), and the charge densities of the carbon atoms in PCN supported the proposed chlorination mechanisms for PCN formation. The results presented in this study might provide helpful information for developing techniques and strategies to control PCN emissions during cement kilns co-processing MSWI fly ash. PMID:27135696

  15. Evaluation of the properties of bitumen and cement pastes and mortars used in the immobilization of waste radioactive

    Energy Technology Data Exchange (ETDEWEB)

    Vieira, Vanessa Mota; de Tello, Cledola Cassia Oliveira, E-mail: vanessamotavieira@gmail.com, E-mail: tellocc@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2013-07-01

    The Project RBMN was launched in November 2008 and aims to establish, manage and execute all tasks for implementing the Brazilian Repository, from its conception to its construction. The concept to be adopted will be a near-surface repository. The inventory includes wastes from the operation of nuclear power plants, fuel cycle facilities and from the use of radionuclides in medicine, industry and activities research and development. The implementation of the national repository is an important technical requirement, and a legal requirement for the entry into operation of the nuclear power plant Angra 3. In Brazil, for the immobilization and solidification of radioactive waste of low and intermediate level of radiation from NPPs are used cement, in Angra 1, and bitumen, in Angra 2. Studies indicate serious concerns about the risks associated with bituminization radioactive waste, much related to the process as the product. There are two major problems due to the presence of products bituminization in repositories, swelling of the waste products and their degradation in the long term. To accommodate the swelling, filling the drums must be limited to 70 - 90% of its volume, which reduces the structural stability of the repository and the optimization of deposition. This study aims to evaluate of the properties of bitumen and cement pastes and mortars used in the immobilization of waste radioactive. (author)

  16. An evaluation of the composition of soil cement bricks with construction and demolition waste - doi: 10.4025/actascitechnol.v33i2.9377

    Directory of Open Access Journals (Sweden)

    Antonio Anderson da Silva Segantini

    2011-04-01

    Full Text Available Sustainable development requires the existence of a production network that includes the reuse of construction waste for new materials. Current analysis investigates an optimal soil-cement composition made up of construction and demolition waste for the manufacture of pressed bricks. Soil-cement bricks were manufactured from construction and demolition wastes (CDW, A-4 classified fine sandy soil and cement CP II Z 32. Laboratory tests, comprising test compaction, optimum water content and maximum dry specific weight, consistency limits, grain size distribution and linear shrinkage, were made to characterize the materials researched. Compressive strength and absorption tests were also undertaken in different combinations of composition. Results showed that the application of CDW improved soil-cement qualities and reduced shrinkage of the material used.

  17. Comparative assessment of TRU waste forms and processes. Volume II. Waste form data, process descriptions, and costs

    International Nuclear Information System (INIS)

    This volume contains supporting information for the comparative assessment of the transuranic waste forms and processes summarized in Volume I. Detailed data on the characterization of the waste forms selected for the assessment, process descriptions, and cost information are provided. The purpose of this volume is to provide additional information that may be useful when using the data in Volume I and to provide greater detail on particular waste forms and processes. Volume II is divided into two sections and two appendixes. The first section provides information on the preparation of the waste form specimens used in this study and additional characterization data in support of that in Volume I. The second section includes detailed process descriptions for the eight processes evaluated. Appendix A lists the results of MCC-1 leach test and Appendix B lists additional cost data. 56 figures, 12 tables

  18. Carbon dioxide emission reduction by increased utilization of waste-derived fuels in the cement industry

    OpenAIRE

    Tokheim, Lars-André; Brevik, Per

    2007-01-01

    Considerable reductions in Norway's emissions of greenhouse gases like CO2 are required to meet the commitments of the Kyoto Protocol. CO2 emissions from cement clinker production originate from decarbonation of limestone as well as fuel combustion, and the cement plants in Norway have to comply with requirements given by the pollution control authorities via the national emissions trading system. There are several ways of reducing CO2 emissions from the cement industry. Utiliz...

  19. Forming artificial soils from waste materials for mine site rehabilitation

    Science.gov (United States)

    Yellishetty, Mohan; Wong, Vanessa; Taylor, Michael; Li, Johnson

    2014-05-01

    Surface mining activities often produce large volumes of solid wastes which invariably requires the removal of significant quantities of waste rock (overburden). As mines expand, larger volumes of waste rock need to be moved which also require extensive areas for their safe disposal and containment. The erosion of these dumps may result in landform instability, which in turn may result in exposure of contaminants such as trace metals, elevated sediment delivery in adjacent waterways, and the subsequent degradation of downstream water quality. The management of solid waste materials from industrial operations is also a key component for a sustainable economy. For example, in addition to overburden, coal mines produce large amounts of waste in the form of fly ash while sewage treatment plants require disposal of large amounts of compost. Similarly, paper mills produce large volumes of alkaline rejected wood chip waste which is usually disposed of in landfill. These materials, therefore, presents a challenge in their use, and re-use in the rehabilitation of mine sites and provides a number of opportunities for innovative waste disposal. The combination of solid wastes sourced from mines, which are frequently nutrient poor and acidic, with nutrient-rich composted material produced from sewage treatment and alkaline wood chip waste has the potential to lead to a soil suitable for mine rehabilitation and successful seed germination and plant growth. This paper presents findings from two pilot projects which investigated the potential of artificial soils to support plant growth for mine site rehabilitation. We found that pH increased in all the artificial soil mixtures and were able to support plant establishment. Plant growth was greatest in those soils with the greatest proportion of compost due to the higher nutrient content. These pot trials suggest that the use of different waste streams to form an artificial soil can potentially be used in mine site rehabilitation

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

    Science.gov (United States)

    Radvanec, Martin; Tuček, L'ubomír; Derco, Ján; Čechovská, Katarína; Németh, Zoltán

    2013-05-15

    Asbestos cement materials, mainly the eternit roof ceiling, being widely applied in the past, represent a serious environmental load. The solar radiation, rain and frost cause the deliberation of cement from the eternit roofing and consequently the wind contaminates the surrounding area by the asbestos (chrysotile) fibers. In combination with other carcinogens (e.g. smoking), or at reduced immunity of a man, they may cause serious respiratory diseases and lung cancer. The article presents the procedure and experimental results of artificial carbonatization, applied in the asbestos cement (eternit). The wet crushed and pulverized asbestos cement was thermally modified at 650°C and then the chrysotile fibers easily and completely reacted with the mixture of CO2 and water, producing new Mg-rich carbonates - hydromagnesite and magnesite: [Formula: see text] Applying this methodology, the asbestos-bearing waste can be stabilized and environmentally friendly permanently deposited. Finding a way of neutralizing of extreme pH values (around 12) at large eternit dumps represents also an asset of presented research. Simultaneously, the artificial carbonatization of chrysotile asbestos, applying CO2, offers an alternative way for permanent liquidation of a part of industrial CO2 emissions, contributing to multiple benefit of this methodology. PMID:23571021

  1. The Optimization of Immobilization for the Low-Activity Waste of theEvaporation Product with Cement

    International Nuclear Information System (INIS)

    The experimental investigation of immobilization the low active wasteconcentration containing 2.44x10-3 μCi/cc a great deal of NaNO3 withcement was done. The immobilization process was carried out by mixing cement,water, concentrate, and Ca-bentonite with a given ratio within a glassbeaker. The mixture was then stirred with an electrical hand mixer untilhomogeneous. The studied immobilization condition were the influences of theweight ratio water to cement, the weight ratio of concentrate to cement withwhich the concentrate pH was varied, and the influence of the addition ofCa-bentonite (% in weight) with the optimum pH of concentrate. The sample inthe container with the size of 2.54 cm in diameter and 3.0 cm in height wasmade of polyethylene and was covered by a tight lid and was cured for 28days. After the sample was cured for 28 days and then it was taken out of thecontainer. This sample quality was ready for being tested. The quality ofcementation product tested compressive strength, density, chemical stability,irradiation stability and thermal stability. The optimum results ofinvestigation were the weight ratio of water to cement = 0.30, thecompressive strength of 30.37 N/mm2. For the immobilization of the waste andcement with the optimum pH being used, yielded in the compressive strength of28.07 N/mm2. Further more from the condition of waste and cement at theoptimum pH which was added by the optimum Ca-bentonite gained the compressivestrength of 33.64 N/mm2 before irradiation, where as after irradiation thecompressive strength was 32.41 N/mm2. The optimum thermal test resultachieved was 250 oC with the compressive strength of 44.10 N/mm2. For theleaching test results after being cured for 91 days in the distilled watermedia was 0.47x10-4 gcm-2day-1, while in the sea water was 0.66x10-4gcm-2day-1. Water medium activity until 91 days = 3.1x10-7 μCi/cc,MPC from ICRP = 8.1x10-7 μCi/cc. The experimental investigation ofcemented waste monolith block resulted

  2. Immobilization in ceramic waste forms of the residues from treatment of mixed wastes

    International Nuclear Information System (INIS)

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

  3. Glass Ceramic Waste Forms for Combined CS+LN+TM Fission Products Waste Streams

    Energy Technology Data Exchange (ETDEWEB)

    Crum, Jarrod V.; Turo, Laura A.; Riley, Brian J.; Tang, Ming; Kossoy, Anna; Sickafus, Kurt E.

    2010-09-23

    In this study, glass ceramics were explored as an alternative waste form for glass, the current baseline, to be used for immobilizing alkaline/alkaline earth + lanthanide (CS+LN) or CS+LN+transition metal (TM) fission-product waste streams generated by a uranium extraction (UREX+) aqueous separations type process. Results from past work on a glass waste form for the combined CS+LN waste streams showed that as waste loading increased, large fractions of crystalline phases precipitated upon slow cooling.[1] The crystalline phases had no noticeable impact on the waste form performance by the 7-day product consistency test (PCT). These results point towards the development of a glass ceramic waste form for treating CS+LN or CS+LN+TM combined waste streams. Three main benefits for exploring glass ceramics are: (1) Glass ceramics offer increased solubility of troublesome components in crystalline phases as compared to glass, leading to increased waste loading; (2) The crystalline network formed in the glass ceramic results in higher heat tolerance than glass; and (3) These glass ceramics are designed to be processed by the same melter technology as the current baseline glass waste form. It will only require adding controlled canister cooling for crystallization into a glass ceramic waste form. Highly annealed waste form (essentially crack free) with up to 50X lower surface area than a typical High-Level Waste (HLW) glass canister. Lower surface area translates directly into increased durability. This was the first full year of exploring glass ceramics for the Option 1 and 2 combined waste stream options. This work has shown that dramatic increases in waste loading are achievable by designing a glass ceramic waste form as an alternative to glass. Table S1 shows the upper limits for heat, waste loading (based on solubility), and the decay time needed before treatment can occur for glass and glass ceramic waste forms. The improvements are significant for both combined waste

  4. Technetium Waste Form Development - Progress Report

    International Nuclear Information System (INIS)

    Analytical electron microscopy using SEM and TEM has been used to analyze a ∼5 g. ingot with composition 71.3 wt% 316SS-5.3 wt% Zr-13.2 wt% Mo-4.0 wt% Rh-6.2 wt% Re prepared at the Idaho National Laboratory. Four phase fields have been identified two of which are lamellar eutectics, with a fifth possibly present. A Zr rich phase was found distributed as fine precipitate, ∼10 (micro)m in diameter, often coating large cavities. A Mo-Fe-Re-Cr lamellar eutectic phase field appears as blocky regions ∼30 (micro)m in diameter, surrounded by a Fe-Mo-Cr lamellar eutectic phase field, and that in turn is surrounded by a Zr-Fe-Rh-Mo-Ni phase field. The eutectic phase separation reactions are different. The Mo-Fe-Re-Cr lamellar eutectic appears a result of austenitic steel forming at lower volume fraction within an Mo-Fe-Re intermetallic phase, whereas the Fe-Mo-Cr lamellar eutectic may be a result of the same intermetallic phase forming within a ferritic steel phase. Cavitation may have arisen either as a result of bubbles, or from loss of equiaxed particles during specimen preparation.

  5. Technetium Waste Form Development - Progress Report

    Energy Technology Data Exchange (ETDEWEB)

    Gelles, David S.; Ermi, Ruby M.; Buck, Edgar C.; Seffens, Rob J.; Chamberlin, Clyde E.

    2009-01-07

    Analytical electron microscopy using SEM and TEM has been used to analyze a ~5 g. ingot with composition 71.3 wt% 316SS-5.3 wt% Zr-13.2 wt% Mo-4.0 wt% Rh-6.2 wt% Re prepared at the Idaho National Laboratory. Four phase fields have been identified two of which are lamellar eutectics, with a fifth possibly present. A Zr rich phase was found distributed as fine precipitate, ~10µm in diameter, often coating large cavities. A Mo-Fe-Re-Cr lamellar eutectic phase field appears as blocky regions ~30µm in diameter, surrounded by a Fe-Mo-Cr lamellar eutectic phase field, and that in turn is surrounded by a Zr-Fe-Rh-Mo-Ni phase field. The eutectic phase separation reactions are different. The Mo-Fe-Re-Cr lamellar eutectic appears a result of austenitic steel forming at lower volume fraction within an Mo-Fe-Re intermetallic phase, whereas the Fe-Mo-Cr lamellar eutectic may be a result of the same intermetallic phase forming within a ferritic steel phase. Cavitation may have arisen either as a result of bubbles, or from loss of equiaxed particles during specimen preparation.

  6. Glass-Ceramic Waste Forms for Uranium and Plutonium Residues Wastes - 13164

    Energy Technology Data Exchange (ETDEWEB)

    Stewart, Martin W.A.; Moricca, Sam A.; Zhang, Yingjie; Day, R. Arthur; Begg, Bruce D. [Australian Nuclear Science and Technology Organisation (ANSTO), New Illawarra Road, Lucas Heights, NSW 2234 (Australia); Scales, Charlie R.; Maddrell, Ewan R. [National Nuclear Laboratory, Sellafield, Seascale, Cumbria, UK, CA20 1PG (United Kingdom); Hobbs, Jeff [Sellafield Limited, Sellafield, Seascale, Cumbria, UK, CA20 1PG (United Kingdom)

    2013-07-01

    A program of work has been undertaken to treat plutonium-residues wastes at Sellafield. These have arisen from past fuel development work and are highly variable in both physical and chemical composition. The principal radiological elements present are U and Pu, with small amounts of Th. The waste packages contain Pu in amounts that are too low to be economically recycled as fuel and too high to be disposed of as lower level Pu contaminated material. NNL and ANSTO have developed full-ceramic and glass-ceramic waste forms in which hot-isostatic pressing is used as the consolidation step to safely immobilize the waste into a form suitable for long-term disposition. We discuss development work on the glass-ceramic developed for impure waste streams, in particular the effect of variations in the waste feed chemistry glass-ceramic. The waste chemistry was categorized into actinides, impurity cations, glass formers and anions. Variations of the relative amounts of these on the properties and chemistry of the waste form were investigated and the waste form was found to be largely unaffected by these changes. This work mainly discusses the initial trials with Th and U. Later trials with larger variations and work with Pu-doped samples further confirmed the flexibility of the glass-ceramic. (authors)

  7. Consolidated waste forms: glass marbles and ceramic pellets

    International Nuclear Information System (INIS)

    Glass marbles and ceramic pellets have been developed at Pacific Northwest Laboratory as part of the multibarrier concept for immobilizing high-level radioactive waste. These consolidated waste forms served as substrates for the application of various inert coatings and as ideal-sized particles for encapsulation in protective matrices. Marble and pellet formulations were based on existing defense wastes at Savannah River Plant and proposed commercial wastes. To produce marbles, glass is poured from a melter in a continuous stream into a marble-making device. Marbles were produced at PNL on a vibratory marble machine at rates as high as 60 kg/h. Other marble-making concepts were also investigated. The marble process, including a lead-encapsulation step, was judged as one of the more feasible processes for immobilizing high-level wastes. To produce ceramic pellets, a series of processing steps are required, which include: spray calcining - to dry liquid wastes to a powder; disc pelletizing - to convert waste powders to spherical pellets; sintering - to densify pellets and cause desired crystal formation. These processing steps are quite complex, and thereby render the ceramic pellet process as one of the least feasible processes for immobilizing high-level wastes

  8. Consolidated waste forms: glass marbles and ceramic pellets

    Energy Technology Data Exchange (ETDEWEB)

    Treat, R.L.; Rusin, J.M.

    1982-05-01

    Glass marbles and ceramic pellets have been developed at Pacific Northwest Laboratory as part of the multibarrier concept for immobilizing high-level radioactive waste. These consolidated waste forms served as substrates for the application of various inert coatings and as ideal-sized particles for encapsulation in protective matrices. Marble and pellet formulations were based on existing defense wastes at Savannah River Plant and proposed commercial wastes. To produce marbles, glass is poured from a melter in a continuous stream into a marble-making device. Marbles were produced at PNL on a vibratory marble machine at rates as high as 60 kg/h. Other marble-making concepts were also investigated. The marble process, including a lead-encapsulation step, was judged as one of the more feasible processes for immobilizing high-level wastes. To produce ceramic pellets, a series of processing steps are required, which include: spray calcining - to dry liquid wastes to a powder; disc pelletizing - to convert waste powders to spherical pellets; sintering - to densify pellets and cause desired crystal formation. These processing steps are quite complex, and thereby render the ceramic pellet process as one of the least feasible processes for immobilizing high-level wastes.

  9. Characterization of different types of ceramic waste and its incorporation to the cement paste; Caracterizaco de diferentes tipos de residuos ceramicos e sua incorporacao a pasta de cimento

    Energy Technology Data Exchange (ETDEWEB)

    Cunha, G.A.; Evangelista, A.C.J.; Almeida, V.C. de, E-mail: valeria@eq.ufrj.b [Universidade Federal do Rio de Janeiro (EQ/UFRJ), RJ (Brazil). Escola de Quimica

    2009-07-01

    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)

  10. Use of Factory-Waste Shingles and Cement Kiln Dust to Enhance the Performance of Soil Used in Road Works

    Directory of Open Access Journals (Sweden)

    Aly Ahmed

    2009-01-01

    Full Text Available An experimental work was conducted to study the use of factory-waste roof shingles to enhance the properties of fine-grained soil used in road works. Cement kiln dust (CKD, a cogenerated product of Portland cement manufacturing, was used as a stabilizing agent while the processed shingles were added to enhance the soil tensile strength. The effects of shingles on strength and stability were evaluated using the unconfined compressive strength, splitting tensile strength, and California Bearing Ratio (CBR tests. The results showed that the use of CKD alone resulted in a considerable increase in the unconfined compressive strength but had a small effect on the tensile strength. The addition of shingles substantially improved the tensile strength of the stabilized soil. A significant reduction in the capillary rise and a slight decrease in the permeability were obtained as a result of shingle addition. An optimal shingle content of 10% is recommended to stabilize the soil.

  11. New Fission-Product Waste Forms: Development and Characterization

    Energy Technology Data Exchange (ETDEWEB)

    Alexandra Navrotsky

    2010-07-30

    Research performed on the program “New Fission Product Waste Forms: Development and Characterization,” in the last three years has fulfilled the objectives of the proposal which were to 1) establish ceramic waste forms for disposing of Cs, Sr and minor actinides, 2) fully characterize the phase relationships, structures and thermodynamic and kinetic stabilities of promising waste forms, 3) establish a sound technical basis for understanding key waste form properties, such as melting temperatures and aqueous durability, based on an in-depth understanding of waste form structures and thermochemistry, and 4) establish synthesis, testing, scaleup and commercialization routes for wasteform implementation through out in-kind collaborations. In addition, since Cs and Sr form new elements by radioactive decay, the behavior and thermodynamics of waste forms containing different proportions of Cs, Sr and their decay products were discovered using non-radioactive analogues. Collaborations among researchers from three institutions, UC Davis, Sandia National Laboratories, and Shott Inc., were formed to perform the primary work on the program. The unique expertise of each of the members in the areas of waste form development, structure/property relationships, hydrothermal and high temperature synthesis, crystal/glass production, and thermochemistry was critical to program success. In addition, collaborations with the Brigham Young Univeristy, Ben Gurion University, and Los Alamos National Laboratory, were established for standard entropies of ceramic waste forms, sol-gel synthesis, and high temperature synthesis. This work has had a significant impact in a number of areas. First, the studies of the thermodynamic stability of the mineral analogues provided an important technical foundation for assessment the viability of multicomponent oxide phases for Cs and Sr removal. Moreover, the thermodynamic data discovered in this program established information on the reaction

  12. Spanish LLW and MLW disposal: durability of cemented materials in (Na, K)Cl simulated radioactive liquid waste.

    Science.gov (United States)

    Goñi, S; Guerrero, A; Hernández, M S

    2001-01-01

    The microstructural stability or durability of a specific backfilling pozzolanic-cement mortar, which is employed in Spain, in concrete containers for the storage of low level liquid wastes (LLW) and medium level liquid wastes (MLW), has been studied by means of the Koch-Steinegger test at the temperatures of 20 and 40 degrees C during a period of 365 days. Mortar samples were immersed in salt solutions of 3.46 M NaCl and 3.46 M KCl to simulate the salinity of some radioactive liquid waste matrices. The resistance of the mortar to the saline solution attack is evaluated by the development of the relative flexural strength. The changes of the microstructure were followed by mercury intrusion porosimetry (MIP), scanning electron microscopy (SEM) and X-ray diffraction (XRD). Pore solution was extracted and analyzed at different periods of time to know the possible diffusion of sodium, chloride and potassium inside the microstructure. PMID:11150135

  13. Characterization of the leaching behaviour of concrete mortars and of cement-stabilized wastes with different waste loading for long term environmental assessment.

    Science.gov (United States)

    van der Sloot, H A

    2002-01-01

    The leaching behaviour of cement-based products-both construction products and cement-stabilized wastes--have been shown to be similar after assessing the leaching characteristics by means of a pH dependence leaching test. This procedure is particularly suited to identifying the chemical speciation of materials. Geochemical modelling has shown a number of solubility controlling phases in this largely inorganic matrix, that can very well explain the observed leaching patterns as a function of pH. Understanding these relationships allows the prediction of leaching behaviour under other exposure conditions and to improve the ultimate quality of products, if so desired. The role of ettringite-type phases for the binding of oxyanions in the pH range above pH 12 has been identified before and confirmed in this work. The order of incorporation follows from the ratio between the maximum leachability at mildly alkaline pH and at high pH. Increased levels of sulfate negatively influence the binding of oxyanions in cement-stabilized waste through site competition.

  14. Physical barrier effect of geopolymeric waste form on diffusivity of cesium and strontium.

    Science.gov (United States)

    Jang, J G; Park, S M; Lee, H K

    2016-11-15

    The present study investigates the physical barrier effect of geopolymeric waste form on leaching behavior of cesium and strontium. Fly ash-based geopolymers and slag-blended geopolymers were used as solidification agents. The leaching behavior of cesium and strontium from geopolymers was evaluated in accordance with ANSI/ANS-16.1. The diffusivity of cesium and strontium in a fly ash-based geopolymer was lower than that in Portland cement by a factor of 10(3) and 10(4), respectively, showing significantly improved immobilization performance. The leaching resistance of fly ash-based geopolymer was relatively constant regardless of the type of fly ash. The diffusivity of water-soluble cesium and strontium ions were highly correlated with the critical pore diameter of the binder. The critical pore diameter of the fly ash-based geopolymer was remarkably smaller than those of Portland cement and slag-blended geopolymer; consequently, its ability physically to retard the diffusion of nuclides (physical barrier effect) was superior.

  15. Electrochemical Corrosion Studies for Modeling Metallic Waste Form Release Rates

    Energy Technology Data Exchange (ETDEWEB)

    Poineau, Frederic [Univ. of Nevada, Las Vegas, NV (United States); Tamalis, Dimitri [Florida Memorial Univ., Miami Gardens, FL (United States)

    2016-08-01

    The isotope 99Tc is an important fission product generated from nuclear power production. Because of its long half-life (t1/2 = 2.13.105 years) and beta-radiotoxicity (β-= 292 keV), it is a major concern in the long-term management of spent nuclear fuel.1 In the spent nuclear fuel, Tc is present as an alloy with Mo, Ru, Rh, and Pd called the epsilon-phase, the relative amount of which increases with fuel burn-up.2 In some separation schemes for spent nuclear fuel, Tc would be separated from the spent fuel and disposed of in a durable waste form.3 Technetium waste forms under consideration include metallic alloys, oxide ceramics and borosilicate glass.4, 5 In the development of a metallic waste form, after separation from the spent fuel, Tc would be converted to the metal, incorporated into an alloy and the resulting waste form stored in a repository.6 Metallic alloys under consideration include Tc-Zr alloys, Tc-stainless-steel alloys and Tc- Inconel alloys (Inconel is an alloy of Ni, Cr and iron which is resistant to corrosion). To predict the long term behavior of the metallic Tc waste form, understanding the corrosion properties of Tc metal and Tc alloys in various chemical environments is needed but efforts to model the behavior of Tc metallic alloys are limited. 7 One parameter that should also be considered in predicting the long-term behavior of the Tc waste form, is the ingrowth of stable Ru that occurs from the radioactive decay of 99Tc (99Tc "99Ru + β-). After a geological period of time, significant amount of Ru will be present in the Tc and may affect its corrosion properties. Studying the effect of Ru on the corrosion behavior of Tc is also of importance.

  16. Research, Calculation for Designing the Technological Equipment Line of Cementation with 3,000 Tonnes of Solid Waste/Year Capacity for ZOC Plan

    International Nuclear Information System (INIS)

    Technological process and equipment line for cementation with 3,000 tons of solid wastes/year capacity for ZOC plant was studied according to the quantity, the composition and the property of solid wastes from ZOC plant. The influence of several parameters with cementation process such as cement mixing ratio, size, moisture and NaCl content in residue, which affected to mechanical and chemical strength of waste block after cementation was evaluated. The calculation and designing for cementation equipment line were thus conducted and total investment cost of construction and installation as well as operation of this equipment line was also estimated. Radiation safety and environmental protection for waste treatment facility was calculated and the design on radiation safety and environmental protection for equipment line was proposed. Technological process and design document of this production line seemed to be reasonable because it is consistent with properties of wastes as well as it has been commonly used in the world to stabilize solid residues in nuclear industry. The advantages of this propose were simple structure of the devices, locally made, easy operation, locally available and inexpensive materials resulting in low cost of investment and operation. (author)

  17. Advanced waste forms research and development. Annual report

    Energy Technology Data Exchange (ETDEWEB)

    McCarthy, G.J.

    1975-06-11

    Research and development activities on advanced (alternatives to glass) nuclear waste forms are reported. The emphasis is on two phases of the work to give essential background information on supercalcine development. The first is a report of the data obtained in the study of cesium aluminosilicate for Cs and Ru fixation. Research on the compatibility of the phases formed in the complex oxide system made up of waste and additive cations is reported. The phase stability in a number of proposed formulations was determined. (JSR)

  18. Performance testing of waste forms in a tuff environment

    International Nuclear Information System (INIS)

    This paper describes experimental work conducted to establish the chemical composition of water which will have reacted with Topopah Spring Member tuff prior to contact with waste packages. The experimental program to determine the behavior of spent fuel and borosilicate glass in the presence of this water is then described. Preliminary results of experiments using spent fuel segments with defects in the Zircaloy cladding are presented. Some results from parametric testing of a borosilicate glass with tuff and 304L stainless steel are also discussed. Experiments conducted using Topopah Spring tuff and J-13 well water have been conducted to provide an estimate of the post-emplacement environment for waste packages in a repository at Yucca Mountain. The results show that emplacement of waste packages should cause only small changes in the water chemistry and rock mineralogy. The changes in environment should not have any detrimental effects on the performance of metal barriers or waste forms. The NNWSI waste form testing program has provided preliminary results related to the release rate of radionuclides from the waste package. Those results indicate that release rates from both spent fuel and borosilicate glass should be below 1 part in 105 per year. Future testing will be directed toward making release rate testing more closely relevant to site specific conditions. 17 references, 7 figures

  19. Crystal chemistry of portland cement hydrates as radioactive waste hosts. Final report, June 15, 1983-June 14, 1984

    International Nuclear Information System (INIS)

    Portland cement hydrates have been used as encapsulant/host phases in radioactive waste management. However, their phase chemistry and stability relationships are poorly defined. Therefore, on occasion, they have not performed as well as expected. As a result, their use has been mainly limited to low-level waste disposal. Since this knowledge gap existed, we had begun to investigate the crystal chemistry of the portland cement hydrates. It was our objective to identify potential hydrate host phases which were not only suitable for isolating radioactive-waste species but also inexpensive, easily processed, low-temperature materials. Initially, we were concentrating upon two areas of interest: the fixation of iodine by the calcium aluminate hydrates and the feasibility of using Stratling's compound as a host phase for cesium and strontium fixation. In both cases, a phase equilibrium study was initiated in order to identify phase relations and consequences of adding the species of interest to the system. An iodine-containing analogue of calcium monosulfoaluminate hydrate (C3A.CaI2.xH2O) was identified as a possible host phase. CsOH was added to formulations in the neighborhood of Stratling's compound, in order to establish phase relations and identify the fixation ability of Stratling's compound and its associated hydrates. 11 figures, 9 tables

  20. Crystal chemistry of portland cement hydrates as radioactive waste hosts. Progress report, June 15, 1983-February 7, 1984

    International Nuclear Information System (INIS)

    Portland cement hydrates have been used as encapsulant/host phases in radioactive waste management. However, their phase chemistry and stability relationships are poorly defined. Therefore, on occassion, they have not performed as well as expected. As a result, their use has been mainly limited to low-level waste disposal. Since this knowledge gap exists, we have begun to investigate the crystal chemistry of the portland cement hydrates. It is our objective to identify potential hydrate host phases which are not only suitable for isolating radioactive-waste species but also inexpensive, easily processed, low-temperature materials. Initially, we have been concentrating upon two areas of interest: The fixation of iodine by the calcium aluminate hydrates and the feasibility of using Straling's compound as a host phase for cesium and strontium fixation. In both cases, a phase equilibrium study has been initiated in order to identify phase relations and consequences of adding the species of interest to the system. An iodine-containing analogue of calcium monosulfoaluminate hydrate (C3A.CaI2.xH2O) has been identified as a possible host phase. CsOH and Sr(OH)2 are being added to formulations in the neighborhood of Stratling's compound, in order to establish phase relations and identify the fixation ability of Stratling's compound and its associated hydrates. 10 figures, 6 tables

  1. The Ceramic Waste Form Process at Idaho National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Stephen Priebe

    2007-05-01

    The treatment of spent nuclear fuel for disposition using an electrometallurgical technique results in two high-level waste forms: a ceramic waste form (CWF) and a metal waste form. Reactive metal fuel constituents, including all the transuranic metals and the majority of the fission products remain in the salt as chlorides and are processed into the CWF. The solidified salt is containerized and transferred to the CWF process where it is ground in an argon atmosphere. Zeolite 4A is ground and then dried in a mechanically-fluidized dryer. The salt and zeolite are mixed in a V-mixer and heated to 500°C to occlude the salt into the structure of the zeolite. The salt-loaded zeolite is cooled, mixed with borosilicate glass frit, and transferred to a crucible, which is placed in a furnace and heated to 925°C. During this process, known as pressureless consolidation, the zeolite is converted to the final sodalite form and the glass thoroughly encapsulates the sodalite, producing a dense, leach-resistant final waste form.

  2. Fire testing of fully active medium-level waste forms

    International Nuclear Information System (INIS)

    The effect of heat on packaged intermediate level waste (ILW) has been studied. This was done in order to be able to predict the behaviour of the ILW under accident conditions involving fire during transport or at the repository. In the study, experimental data were obtained and used in the development and validation of theoretical models to describe aspects of the behaviour of the waste form when subjected to heat. The prime objective was to be able to predict the amounts of radioactive materials released from a given incident. Four ILW streams were selected for experimental study. These four were chosen as the minimum that could be studied to provide a set of data that could be used in the prediction of the behaviour of the majority of ILW produced in the UK. Heating experiments were carried out on a small scale using packaged ILW samples made from active wastes or inactive simulants. Data were obtained on temperatures in the waste form, production of volatile materials, carry-forward of solid particulate materials and carry-forward of radionuclides. The results were used, together with data from full-scale experiments with inactive simulant ILW carried out at Winfrith, to develop and validate a theoretical model. This model calculates the temperature profiles within a package of immobilized ILW as a function of the applied heating conditions. The temperature of the waste form is used to predict the release of radioactive materials from the package. 4 refs., 65 figs., 13 tabs

  3. State of the art report on bituminized waste forms of radioactive wastes

    International Nuclear Information System (INIS)

    In this report, research and development results on the bituminization of radioactive wastes are closely reviewed, especially those regarding waste treatment technologies, waste solidifying procedures and the characteristics of asphalt and solidified forms. A new concept of the bituminization method is suggested in this report which can improve the characteristics of solidified forms. Stable solid forms with high leach resistance, high thermal resistance and good compression strength were produced by the suggested bituminization method, in which spent polyethylene from agricultural farms was added. This report can help further research and development of improved bituminized forms of radioactive wastes that will maintain long term stabilities in disposal sites. (author). 59 refs., 19 tabs., 18 figs

  4. State of the art report on bituminized waste forms of radioactive wastes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Kook; Shon, Jong Sik; Kim, Kil Jeong; Lee, Kang Moo; Jung, In Ha

    1998-03-01

    In this report, research and development results on the bituminization of radioactive wastes are closely reviewed, especially those regarding waste treatment technologies, waste solidifying procedures and the characteristics of asphalt and solidified forms. A new concept of the bituminization method is suggested in this report which can improve the characteristics of solidified forms. Stable solid forms with high leach resistance, high thermal resistance and good compression strength were produced by the suggested bituminization method, in which spent polyethylene from agricultural farms was added. This report can help further research and development of improved bituminized forms of radioactive wastes that will maintain long term stabilities in disposal sites. (author). 59 refs., 19 tabs., 18 figs

  5. The effect of W/C ratio and cement type on the longevity of grouts for use in a nuclear fuel waste disposal vault

    International Nuclear Information System (INIS)

    Cement-based grouts are being considered for use as sealing materials in the Canadian concept for nuclear fuel waste disposal. This paper describes laboratory studies of the longevity of these materials, with special emphasis on the effect of hardened grout porosity and cement type. The longevity properties determined for reference grout (90% Type 50 cement, 10% silica fume and superplasticizer) are compared with those of a slag cement grout. The fractional factorial statistical method of Box-Behnken was used to design a series of leach tests, which covered a wide range of conditions that could occur in a nuclear waste disposal vault. The leach tests have been carried out to determine the effect of temperature, ionic strength of groundwater, and cation exchange capacity (CEC) of clay (which may be in contact with the grout) on the leach resistance of the cement. The temperature ranged from 25 to 150 degrees C and the ionic strength of the groundwaters from 0.0015 to 1.37 mol. Leach rates of Ca and Si were taken as the major indicators of the long-term chemical stability of the grouts. Preliminary analysis suggests that the reference grout would be more stable than slag cement in the high-temperature environment of a nuclear fuel waste disposal vault. Within the ranges investigated, decreasing the porosity appears not to significantly decrease leach rates

  6. The Evaluation of Material Properties of Low-pH Cement Grout for the Application of Cementitious Materials to Deep Radioactive Waste Repository Tunnels

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jin Seop; Kwon, S. K.; Cho, W. J.; Kim, G. W

    2009-12-15

    Considering the current construction technology and research status of deep repository tunnels for radioactive waste disposal, it is inevitable to use cementitious materials in spite of serious concern about their long-term environmental stability. Thus, it is an emerging task to develop low pH cementitious materials. This study reviews the state of the technology on low pH cements developed in Sweden, Switzerland, France, and Japan as well as in Finland which is constructing a real deep repository site for high-level radioactive waste disposal. Considering the physical and chemical stability of bentonite which acts as a buffer material, a low pH cement limits to pH {<=}11 and pozzolan-type admixtures are used to lower the pH of cement. To attain this pH requirement, silica fume, which is one of the most promising admixtures, should occupy at least 40 wt% of total dry materials in cement and the Ca/Si ratio should be maintained below 0.8 in cement. Additionally, selective super-plasticizer needs to be used because a high amount of water is demanded from the use of a large amount of silica fume. In this report, the state of the technology on application of cementitious materials to deep repository tunnels for radioactive waste disposal was analysed. And the material properties of low-pH and high-pH cement grouts were evaluated base on the grout recipes of ONKALO in Finlan.

  7. Utilization of municipal solid waste incineration (MSWI) fly ash in blended cement Part 2. Mechanical strength of mortars and environmental impact.

    Science.gov (United States)

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

    2007-07-19

    This second of two articles dealing with the utilization of MSWI fly ash in blended cement studies the effects of two variants of the stabilization process on the behavior of the treated fly ash (TFA) introduced into cement-based mortars. From a technological point of view, the modifications of the process are very efficient and eliminate the swelling produced by the introduction of MSWI fly ash in cement-based mortars. TFA has a significant activity in cement-based mortars and can also advantageously replace a part of the cement in cement-based material. From an environmental point of view, the results of traditional leaching tests on monolithic and crushed mortars highlight a poor stabilization of some harmful elements such as antimony and chromium. The use of a cement rich in ground granulated blast furnace slag (GGBFS) with a view to stabilizing the chromium is not efficient. Since neither adequate tests nor quality criteria exist to evaluate the pollutant potential of a waste with a view to reusing it, it is difficult to conclude on the environmental soundness of such a practice. Further experiments are necessary to investigate the environmental impact of TFA introduced in cement-based mortars depending on the reuse scenario.

  8. Effects on radionuclide concentrations by cement/ground-water interactions in support of performance assessment of low-level radioactive waste disposal facilities

    International Nuclear Information System (INIS)

    The US Nuclear Regulatory Commission is developing a technical position document that provides guidance regarding the performance assessment of low-level radioactive waste disposal facilities. This guidance considers the effects that the chemistry of the vault disposal system may have on radionuclide release. The geochemistry of pore waters buffered by cementitious materials in the disposal system will be different from the local ground water. Therefore, the cement-buffered environment needs to be considered within the source term calculations if credit is taken for solubility limits and/or sorption of dissolved radionuclides within disposal units. A literature review was conducted on methods to model pore-water compositions resulting from reactions with cement, experimental studies of cement/water systems, natural analogue studies of cement and concrete, and radionuclide solubilities experimentally determined in cement pore waters. Based on this review, geochemical modeling was used to calculate maximum concentrations for americium, neptunium, nickel, plutonium, radium, strontium, thorium, and uranium for pore-water compositions buffered by cement and local ground-water. Another literature review was completed on radionuclide sorption behavior onto fresh cement/concrete where the pore water pH will be greater than or equal 10. Based on this review, a database was developed of preferred minimum distribution coefficient values for these radionuclides in cement/concrete environments

  9. Effects on radionuclide concentrations by cement/ground-water interactions in support of performance assessment of low-level radioactive waste disposal facilities

    Energy Technology Data Exchange (ETDEWEB)

    Krupka, K.M.; Serne, R.J. [Pacific Northwest National Lab., Richland, WA (United States)

    1998-05-01

    The US Nuclear Regulatory Commission is developing a technical position document that provides guidance regarding the performance assessment of low-level radioactive waste disposal facilities. This guidance considers the effects that the chemistry of the vault disposal system may have on radionuclide release. The geochemistry of pore waters buffered by cementitious materials in the disposal system will be different from the local ground water. Therefore, the cement-buffered environment needs to be considered within the source term calculations if credit is taken for solubility limits and/or sorption of dissolved radionuclides within disposal units. A literature review was conducted on methods to model pore-water compositions resulting from reactions with cement, experimental studies of cement/water systems, natural analogue studies of cement and concrete, and radionuclide solubilities experimentally determined in cement pore waters. Based on this review, geochemical modeling was used to calculate maximum concentrations for americium, neptunium, nickel, plutonium, radium, strontium, thorium, and uranium for pore-water compositions buffered by cement and local ground-water. Another literature review was completed on radionuclide sorption behavior onto fresh cement/concrete where the pore water pH will be greater than or equal 10. Based on this review, a database was developed of preferred minimum distribution coefficient values for these radionuclides in cement/concrete environments.

  10. Deterioration study of a material for encapsulation of radioactive wastes, the Portland cement, by heterotrophic microorganisms isolated from natural media

    International Nuclear Information System (INIS)

    Soils and geologic formations selected for storage of radioactive waste storage contain microflora (nitrifying and sulfoxidizing bacteria, heterotrophic microorganisms) that can corrode cement through acidic metabolism products. Nutriments required for their development are also found in these biotopes. Corrosine effects of organic acids produced by heterotrophic microorganisms are: mass decrease, leaching (especially Ca), dissolution of portlandite crystals Ca (OH)2, increase of porosity and decrease of flexural strength. Excretion of corrosive organic acids by bacteria is promoted by high temperature and basic pH. Acidification by fungi requires also a high temperature but an acidic pH

  11. Characteristics of borosilicate waste glass form for high-level radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Soo; Chun, Kwan Sik; Choi, Jong Won; Kang, Chul Hyung

    2001-03-01

    Basic data, required for the design and the performance assessment of a repository of HLW, suchas the chemical composition and the characteristics of the borosilicate waste glass have been identified according to the burn-ups of spent PWR fuels. The diemnsion of waste canister is 430mm in diameter and 1135mm in length, and the canister should hold less than 2kwatts of heat from their decay of radionuclides contained in the HLW. Based on the reprocessing of 5 years-cooled spent fuel, one canister could hold about 11.5wt.% and 10.8wt.% of oxidized HLW corresponding to their burn-ups of 45,000MWD/MTU and 55,000MWD/MTU, respectively. These waste forms have been recommanded as the reference waste forms of HLW. The characteristics of these wastes as a function of decay time been evaluated. However, after a specific waste form and a specific site for the disposal would be selected, the characteristics of the waste should be reevaluated under the consideration of solidification period, loaded waste, storage condition and duration, site circumstances for the repository system and its performance assessment.

  12. Technical area status report for low-level mixed waste final waste forms. Volume 2, Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Mayberry, J.L.; Huebner, T.L. [Science Applications International Corp., Idaho Falls, ID (United States); Ross, W. [Pacific Northwest Lab., Richland, WA (United States); Nakaoka, R. [Los Alamos National Lab., NM (United States); Schumacher, R. [Westinghouse Savannah River Co., Aiken, SC (United States); Cunnane, J.; Singh, D. [Argonne National Lab., IL (United States); Darnell, R. [EG and G Idaho, Inc., Idaho Falls, ID (United States); Greenhalgh, W. [Westinghouse Hanford Co., Richland, WA (United States)

    1993-08-01

    This report presents information on low-level mixed waste forms.The descriptions of the low-level mixed waste (LLMW) streams that are considered by the Mixed Waste Integrated Program (MWIP) are given in Appendix A. This information was taken from descriptions generated by the Mixed Waste Treatment Program (MWTP). Appendix B provides a list of characteristic properties initially considered by the Final Waste Form (FWF) Working Group (WG). A description of facilities available to test the various FWFs discussed in Volume I of DOE/MWIP-3 are given in Appendix C. Appendix D provides a summary of numerous articles that were reviewed on testing of FWFS. Information that was collected by the tests on the characteristic properties considered in this report are documented in Appendix D. The articles reviewed are not a comprehensive list, but are provided to give an indication of the data that are available.

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

    International Nuclear Information System (INIS)

    This report presents information on low-level mixed waste forms.The descriptions of the low-level mixed waste (LLMW) streams that are considered by the Mixed Waste Integrated Program (MWIP) are given in Appendix A. This information was taken from descriptions generated by the Mixed Waste Treatment Program (MWTP). Appendix B provides a list of characteristic properties initially considered by the Final Waste Form (FWF) Working Group (WG). A description of facilities available to test the various FWFs discussed in Volume I of DOE/MWIP-3 are given in Appendix C. Appendix D provides a summary of numerous articles that were reviewed on testing of FWFS. Information that was collected by the tests on the characteristic properties considered in this report are documented in Appendix D. The articles reviewed are not a comprehensive list, but are provided to give an indication of the data that are available

  14. Method of making nanostructured glass-ceramic waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Gao, Huizhen; Wang, Yifeng; Rodriguez, Mark A.; Bencoe, Denise N.

    2014-07-08

    A waste form for and a method of rendering hazardous materials less dangerous is disclosed that includes fixing the hazardous material in nanopores of a nanoporous material, reacting the trapped hazardous material to render it less volatile/soluble, and vitrifying the nanoporous material containing the less volatile/soluble hazardous material.

  15. Microstructural characterization of glass and ceramic simulated waste forms

    International Nuclear Information System (INIS)

    The microstructures of three nonradioactive glass samples simulating three Hanford process waste forms were characterized. Two samples of iodine sodalite which simulate the fixation of radioactive iodine were also characterized. X-ray diffraction, electron microscopy + x-ray energy dispersive spectrometry, and electron microprobe analysis were used in the characterization

  16. Chemical composition, effective atomic number and electron density study of trommel sieve waste (TSW), Portland cement, lime, pointing and their admixtures with TSW in different proportions.

    Science.gov (United States)

    Kurudirek, Murat; Aygun, Murat; Erzeneoğlu, Salih Zeki

    2010-06-01

    The trommel sieve waste (TSW) which forms during the boron ore production is considered to be a promising building material with its use as an admixture with Portland cement and is considered to be an alternative radiation shielding material, also. Thus, having knowledge on the chemical composition and radiation interaction properties of TSW as compared to other building materials is of importance. In the present study, chemical compositions of the materials used have been determined using a wavelength dispersive X-ray fluorescence spectrometer (WDXRFS). Also, TSW, some commonly used building materials (Portland cement, lime and pointing) and their admixtures with TSW have been investigated in terms of total mass attenuation coefficients (mu/rho), photon interaction cross sections (sigma(t)), effective atomic numbers (Z(eff)) and effective electron densities (N(e)) by using X-rays at 22.1, 25keV and gamma-rays at 88keV photon energies. Possible conclusions were drawn with respect to the variations in photon energy and chemical composition. PMID:20080413

  17. The role of chemical reaction in waste-form performance

    International Nuclear Information System (INIS)

    The dissolution rate of waste solids in a geologic repository is a complex function of waste form geometry, chemical raction rate, exterior flow field, and chemical environment. We present here an analysis to determine the stady-state mass transfer rate, over the entire range of flow conditions relevant to geologic disposal of nuclear waste. The equations for steady-state mass transfer with a chemical-reaction-rate boundary condition are solved by three different mathematical techniques which supplement each other. This theory is illustrated with laboratory leach data for borosilicate-glass and a spherical spent-fuel waste form under typical repository conditions. For borosilicate glass waste in the temperature range of 57/degree/C to 250/degree/C, dissolution rate in a repository is determined for a wide range of chemical reaction rates and for Peclet numbers from zero to well over 100, far beyond any Peclet values expected in a repository. Spent-fuel dissolution in a repository is also investigated, based on the limited leach data now available. 10 refs., 4 figs., 1 tab

  18. Geotechnical properties of peat soil stabilised with shredded waste tyre chips in combination with gypsum, lime or cement

    Directory of Open Access Journals (Sweden)

    M. Saberian

    2016-07-01

    Full Text Available Peat has a high content of water and organic substances. These weak components can cause low bearing capacity and high consolidation settlement under load, which means that peat deposits must usually be stabilised if they are to bear constructions such as buildings or roads. In this study we investigated the performance of waste tyre chips (10 % by weight and sand (400 kg m-3 supplemented with a pozzolanic binder (gypsum, lime or cement at a range of dosages (5 %, 10 % or 15 % by weight as a stabiliser for peat soil. Peat samples were taken from a fen peatland at Chaghakhor Wetland in Chahar Mahal and Bakhtiari Province, Iran. In total, 162 test specimens were prepared and subjected to laboratory strength testing (unconfined compression test and direct shear test. Additionally, the pH of each admixture was recorded immediately after mixing, elemental compositions were determined by X-Ray Fluorescence (XRF, and structures were examined using Scanning Electron Microscopy (SEM. It was observed that: (1 the total percentage of pozzolanic compounds in the peat soil was well below the minimum of 70 % set by the standard ASTM C 618 (ASTM 2000, so an additive such as cement, lime or gypsum would certainly be required; (2 specimens stabilised with gypsum or lime showed improvements in unconfined compressive strength (UCS, but those stabilised with ordinary Portland cement exhibited the greatest improvement in UCS (up to 12,200 % as well as improvements in the direct shear parameters c and φ; (3 according to the XRF tests, additives such as cement, lime and gypsum introduced considerable amounts of Si, Al, Ca and O, which are important for pozzolanic reactions in peat soils; and (4 on the basis of the results of UCS and direct shear tests, the optimum percentage of the additives tested would be 5 %.

  19. Testing of high-level waste forms under repository conditions

    International Nuclear Information System (INIS)

    The workshop on testing of high-level waste forms under repository conditions was held on 17 to 21 October 1988 in Cadarache, France, and sponsored by the Commission of the European Communities (CEC), the Commissariat a l'energie atomique (CEA) and the Savannah River Laboratory (US DOE). Participants included representatives from Australia, Belgium, Denmark, France, Germany, Italy, Japan, the Netherlands, Sweden, Switzerland, The United Kingdom and the United States. The first part of the conference featured a workshop on in situ testing of simulated nuclear waste forms and proposed package components, with an emphasis on the materials interface interactions tests (MIIT). MIIT is a sevent-part programme that involves field testing of 15 glass and waste form systems supplied by seven countries, along with potential canister and overpack materials as well as geologic samples, in the salt geology at the Waste Isolation Pilot Plant (WIPP) in Carlsbad, New Mexico, USA. This effort is still in progress and these proceedings document studies and findings obtained thus far. The second part of the meeting emphasized multinational experimental studies and results derived from repository systems simulation tests (RSST), which were performed in granite, clay and salt environments

  20. Radiation damage in natural materials: implications for radioactive waste forms

    International Nuclear Information System (INIS)

    The long-term effect of radiation damage on waste forms, either crystalline or glass, is a factor in the evaluation of the integrity of waste disposal mediums. Natural analogs, such as metamict minerals, provide one approach for the evaluaton of radiation damage effects that might be observed in crystalline waste forms, such as supercalcine or synroc. Metamict minerals are a special class of amorphous materials which were initially crystalline. Although the mechanism for the loss of crystallinity in these minerals (mostly actinide-containing oxides and silicates) is not clearly understood, damage caused by alpha particles and recoil nuclei is critical to the metamictization process. The study of metamict minerals allows the evaluation of long-term radiation damage effects, particularly changes in physical and chemical properties such as microfracturing, hydrothermal alteration, and solubility. In addition, structures susceptible to metamictization share some common properties: (1) complex compositions; (2) some degree of covalent bonding, instead of being ionic close-packed MO/sub x/ structures; and (3) channels or interstitial voids which may accommodate displaced atoms or absorbed water. On the basis of these empirical criteria, minerals such as pollucite, sodalite, nepheline and leucite warrant careful scrutiny as potential waste form phases. Phases with the monazite or fluorite structures are excellent candidates

  1. Diffusion-based leaching models for glassy waste forms

    International Nuclear Information System (INIS)

    Most scenarios for the disposal of high-level nuclear wastes assume burial under conditions in which only a limited quantity of groundwater will contact the waste form. In order to model these conditions, it is necessary to describe the release of species from a waste form matrix in contact with a limited volume of leachant in which the concentration of released species is not zero and is itself a function of release rate. Eight leaching models are presented that include the cases of a dissolving and a nondissolving matrix, finite, infinite, and replenished leachant volumes, and a matrix covered by a surface layer with different properties. The equations that describe these models assume a linear concentration profile of the diffusing species within the waste form and apply Fick's first law to obtain the leach rate. In three cases a direct comparison is possible between the solutions of these equations and solutions obtained by use of the diffusion equation derived from Fick's second law. Good agreement is found. The equations given are convenient for use with programmable calculators

  2. Radiation damage studies related to nuclear waste forms

    International Nuclear Information System (INIS)

    Much of the previously reported work on alpha radiation effects on crystalline phases of importance to nuclear waste forms has been derived from radiation effects studies of composite waste forms. In the present work, two single-phase crystalline materials, Gd2Ti2O7 (pyrochlore) and CaZrTi2O7 (zirconolite), of relative importance to current waste forms were studied independently by doping with 244Cm at the 3 wt % level. Changes in the crystalline structure measured by x-ray diffraction as a function of dose show that damage ingrowth follows an expected exponential relationship of the form ΔV/V0 = A[1-exp(-BD)]. In both cases, the materials became x-ray amorphous before the estimated saturation value was reached. The predicted magnitudes of the unit cell volume changes at saturation are 5.4% and 3.5%, respectively, for Gd2Ti2O7 and CaZrTi2O7. The later material exhibited anisotropic behavior in which the expansion of the monoclinic cell in the c0 direction was over five times that of the a0 direction. The effects of transmutations on the properties of high-level waste solids have not been studied until now because of the long half-lives of the important fission products. This problem was circumvented in the present study by preparing materials containing natural cesium and then irradiating them with neutrons to produce 134Cs, which has only a 2y half-life. The properties monitored at about one year intervals following irradiation have been density, leach rate and microstructure. A small amount of x-ray diffraction work has also been done. Small changes in density and leach rate have been observed for some of the materials, but they were not large enough to be of any consequence for the final disposal of high level wastes

  3. Microbial-influenced cement degradation: Literature review

    International Nuclear Information System (INIS)

    The Nuclear Regulatory Commission stipulates that disposed low-level radioactive waste (LLW) be stabilized. Because of apparent ease of use and normal structural integrity, cement has been widely used as a binder to solidify LLW. However, the resulting waste forms are sometimes susceptible to failure due to the actions of waste constituents, stress, and environment. This report reviews literature which addresses the effect of microbiologically influenced chemical attack on cement-solidified LLW. Groups of microorganisms are identified, which are capable of metabolically converting organic and inorganic substrates into organic and mineral acids. Such acids aggressively react with concrete and can ultimately lead to structural failure. Mechanisms inherent in microbial-influenced degradation of cement-based material are the focus of this report. This report provides sufficient evidence of the potential for microbial-influenced deterioration of cement-solidified LLW to justify the enumeration of the conditions necessary to support the microbiological growth and population expansion, as well as the development of appropriate tests necessary to determine the resistance of cement-solidified LLW to microbiological-induced degradation that could impact the stability of the waste form

  4. FY 1998 annual report on power generation by waste heat from cement production in China; Chugoku ni okeru cement hainetsu hatsuden 1998 nendo chosa hokokusho

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-03-01

    This project is to implement a feasibility study for applying waste heat power generation, which have been already commercialized in Japan and producing remarkable results, to China's cement plants producing 3,500 t/d or more of clinker, and thereby to try to establish a link with the Japan's clean development mechanism. It is expected that introduction of these systems improves energy use efficiency and environments in China. The study results indicate that the project for a Tongling Conch plant could generate power of 15,000 kW, reducing CO2 emissions by 89,178 t/y and cumulatively 1,783,560 tons in the 20-year period. The results also indicate that the project will be highly profitable, with an estimated internal return rate of as high as 33.78%. The project for a Huaxin plant could generate power of 8,400 kW, reducing CO2 emissions by 48,412 t/y and cumulatively 968,240 tons in the 20-year period, annually saving power charges by 325 million yen and bringing an internal return rate of 10.72%. (NEDO)

  5. Research and development of waste forms for geological disposal

    International Nuclear Information System (INIS)

    Ceramics are candidate materials for immobilizing high-level waste (HLW) stemming from the reprocessing of spent fuels. We are proceeding with R and D on two types of ceramic waste form : a polyphase titanate ceramic named Synroc and three kinds of single-phase zirconium ceramics. The effect of self-irradiation damage on the long-term integrity of Synroc due to alpha decay was studied under a cooperative program between JAERI and ANSTO. The hot-pressed polyphase titanate ceramic (10 wt% waste loading) was doped with 244Cm to accumulate a dose of 1.6 x 1018 alpha decays/g. The phase assemblage of the curium-doped titanate ceramic included freudenbergite and loveringite in addition to three main phases: hollandite, perovskite and zirconolite. Accumulation of alpha decays was accompanied by a gradual decrease in density. The change in density was -2.7 % after an equivalent age of 45000 years. The durability of three single-phase zirconium ceramics which contained the appropriate amount of simulated high-level waste elements was examined at 90degC and 150degC in hydrochloric acid or deionized water. The waste forms examined included 10 mol% Y2O3-stabilized ZrO2, La2Zr2O7 with a pyrochlore structure, and CaZrO3 with a perovskite structure. La2Zr2O7 showed excellent durability, and leach rates of all constituents were less than about 10-4 g·m-2·day-1 at 150degC in deionized water. This suggests that La2Zr2O7 is a promising candidate material for immobilization of waste elements from HLW. (J.P.N.)

  6. Electrochemical Corrosion Studies for Modeling Metallic Waste Form Release Rates

    Energy Technology Data Exchange (ETDEWEB)

    Poineau, Frederic [Univ. of Nevada, Las Vegas, NV (United States); Tamalis, Dimitri [Florida Memorial Univ., Miami Gardens, FL (United States)

    2016-08-01

    The isotope 99Tc is an important fission product generated from nuclear power production. Because of its long half-life (t1/2 = 2.13 ∙ 105 years) and beta-radiotoxicity (β⁻ = 292 keV), it is a major concern in the long-term management of spent nuclear fuel. In the spent nuclear fuel, Tc is present as an alloy with Mo, Ru, Rh, and Pd called the epsilon-phase, the relative amount of which increases with fuel burn-up. In some separation schemes for spent nuclear fuel, Tc would be separated from the spent fuel and disposed of in a durable waste form. Technetium waste forms under consideration include metallic alloys, oxide ceramics and borosilicate glass. In the development of a metallic waste form, after separation from the spent fuel, Tc would be converted to the metal, incorporated into an alloy and the resulting waste form stored in a repository. Metallic alloys under consideration include Tc–Zr alloys, Tc–stainless steel alloys and Tc–Inconel alloys (Inconel is an alloy of Ni, Cr and iron which is resistant to corrosion). To predict the long-term behavior of the metallic Tc waste form, understanding the corrosion properties of Tc metal and Tc alloys in various chemical environments is needed, but efforts to model the behavior of Tc metallic alloys are limited. One parameter that should also be considered in predicting the long-term behavior of the Tc waste form is the ingrowth of stable Ru that occurs from the radioactive decay of 99Tc (99Tc → 99Ru + β⁻). After a geological period of time, significant amounts of Ru will be present in the Tc and may affect its corrosion properties. Studying the effect of Ru on the corrosion behavior of Tc is also of importance. In this context, we studied the electrochemical behavior of Tc metal, Tc-Ni alloys (to model Tc-Inconel alloy) and Tc-Ru alloys in acidic media. The study of Tc-U alloys has also been performed in order to better understand the

  7. Low sintering temperature glass waste forms for sequestering radioactive iodine

    Science.gov (United States)

    Nenoff, Tina M.; Krumhansl, James L.; Garino, Terry J.; Ockwig, Nathan W.

    2012-09-11

    Materials and methods of making low-sintering-temperature glass waste forms that sequester radioactive iodine in a strong and durable structure. First, the iodine is captured by an adsorbant, which forms an iodine-loaded material, e.g., AgI, AgI-zeolite, AgI-mordenite, Ag-silica aerogel, ZnI.sub.2, CuI, or Bi.sub.5O.sub.7I. Next, particles of the iodine-loaded material are mixed with powdered frits of low-sintering-temperature glasses (comprising various oxides of Si, B, Bi, Pb, and Zn), and then sintered at a relatively low temperature, ranging from 425.degree. C. to 550.degree. C. The sintering converts the mixed powders into a solid block of a glassy waste form, having low iodine leaching rates. The vitrified glassy waste form can contain as much as 60 wt % AgI. A preferred glass, having a sintering temperature of 500.degree. C. (below the silver iodide sublimation temperature of 500.degree. C.) was identified that contains oxides of boron, bismuth, and zinc, while containing essentially no lead or silicon.

  8. Crystallization behavior during melt-processing of ceramic waste forms

    Science.gov (United States)

    Tumurugoti, Priyatham; Sundaram, S. K.; Misture, Scott T.; Marra, James C.; Amoroso, Jake

    2016-05-01

    Multiphase ceramic waste forms based on natural mineral analogs are of great interest for their high chemical durability, radiation resistance, and thermodynamic stability. Melt-processed ceramic waste forms that leverage existing melter technologies will broaden the available disposal options for high-level nuclear waste. This work reports on the crystallization behavior in selected melt-processed ceramics for waste immobilization. The phase assemblage and evolution of hollandite, zirconolite, pyrochlore, and perovskite type structures during melt processing were studied using thermal analysis, x-ray diffraction, and electron microscopy. Samples prepared by melting followed by annealing and quenching were analyzed to determine and measure the progression of the phase assemblage. Samples were melted at 1500 °C and heat-treated at crystallization temperatures of 1285 °C and 1325 °C corresponding to exothermic events identified from differential scanning calorimetry measurements. Results indicate that the selected multiphase composition partially melts at 1500 °C with hollandite coexisting as crystalline phase. Perovskite and zirconolite phases crystallized from the residual melt at temperatures below 1350 °C. Depending on their respective thermal histories, different quenched samples were found to have different phase assemblages including phases such as perovskite, zirconolite and TiO2.

  9. Tailored ceramic consolidation forms for ICPP waste compositions

    International Nuclear Information System (INIS)

    This paper reports a polyphase tailored ceramic developed for the consolidation of simulated ICPP (Idaho Chemical Processing Plant)-type high Zr content high-level waste (HLW) calcines. The ceramic is specifically designed to provide chemically stable host phases for each species present in the HLW and to maximize waste volume reduction through high loadings and form density. The ceramic is designed for a 73 wt% waste loading with a density of 3.35 ± 0.05 (g/cm3). The major phase in the ceramic is a high-silica glass, which contains the neutron poison boron as well as the majority of the nonrefractory species in the waste. The primary crystalline phases are calcium fluoride, calcium-yttrium stabilized cubic zirconia, a hexagonal apatite type silicate containing the plutonium simulant Ce, and a Cd metal phase. Minor phases include zircon, zirconolite, and a sphene-type. Leaching testing and microscopic analysis shows the ceramic form to be chemically durable, with only the glass phase showing any detectable dissolution in deionized water at 90 degrees C

  10. Technical viability and development needs for waste forms and facilities

    Energy Technology Data Exchange (ETDEWEB)

    Pegg, I.; Gould, T.

    1996-05-01

    The objective of this breakout session was to provide a forum to discuss technical issues relating to plutonium-bearing waste forms and their disposal facilities. Specific topics for discussion included the technical viability and development needs associated with the waste forms and/or disposal facilities. The expected end result of the session was an in-depth (so far as the limited time would allow) discussion of key issues by the session participants. The session chairs expressed allowance for, and encouragement of, alternative points of view, as well as encouragement for discussion of any relevant topics not addressed in the paper presentations. It was not the intent of this session to recommend or advocate any one technology over another.

  11. Radiation and transmutation effects relevant to solid nuclear waste forms

    International Nuclear Information System (INIS)

    Radiation effects in insulating solids are discussed in a general way as an introduction to the quite sparse published work on radiation effects in candidate nuclear waste forms other than glasses. Likely effects of transmutation in crystals and the chemical mitigation strategy are discussed. It seems probable that radiation effects in solidified HLW will not be serious if the actinides can be wholly incorporated in such radiation-resistant phases as monazite or uraninite

  12. Preliminary waste form characteristics report Version 1.0. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Stout, R.B.; Leider, H.R. [eds.

    1991-10-11

    This report focuses on radioactive waste form characteristics that will be used to design a waste package and an engineered barrier system (EBS) for a suitable repository as part of the Yucca Mountain Project. The term waste form refers to irradiated reactor fuel, other high-level waste (HLW) in various physical forms, and other radioactive materials (other than HLW) which are received for emplacement in a geologic repository. Any encapsulating of stabilizing matrix is also referred to as a waste form.

  13. Treatment of mercury containing waste

    Science.gov (United States)

    Kalb, Paul D.; Melamed, Dan; Patel, Bhavesh R; Fuhrmann, Mark

    2002-01-01

    A process is provided for the treatment of mercury containing waste in a single reaction vessel which includes a) stabilizing the waste with sulfur polymer cement under an inert atmosphere to form a resulting mixture and b) encapsulating the resulting mixture by heating the mixture to form a molten product and casting the molten product as a monolithic final waste form. Additional sulfur polymer cement can be added in the encapsulation step if needed, and a stabilizing additive can be added in the process to improve the leaching properties of the waste form.

  14. Effects of carbonation and leaching on porosity in cement-bound waste.

    Science.gov (United States)

    Van Gerven, T; Cornelis, G; Vandoren, E; Vandecasteele, C

    2007-01-01

    Porosity is possibly an important parameter with respect to leaching of constituents from cement monoliths. During its lifetime, the pore structure of cementitious matrices changes due to carbonation and leaching. This paper discusses the effects of both accelerated carbonation and continuous leaching on the porosity, and, conversely, how porosity affects leaching properties. Two sample types are investigated: a mortar with MSWI-bottom ash substituting the sand fraction and a cement paste with 30 wt% of the cement substituted by a flue gas cleaning residue. The samples have been intensively carbonated in a 20% CO(2) atmosphere for up to 60 days and were subsequently leached. The porosity was investigated by mercury intrusion porosimetry. Accelerated carbonation decreases total porosity by 12% in the case of 60 days of treatment of bottom ash mortars, whereas continuous leaching during 225 days increases it by 16%. Both carbonation and leaching decrease the amount of smaller capillary pores. Carbonation decreases both porosity and pH. Decreasing porosity diminishes leaching of sodium and potassium, while the decrease in pH increases leaching. However, the former process dominates the latter, resulting in a net decreasing effect of carbonation on the release of sodium and potassium from these cement matrices. PMID:16843650

  15. Waste form characterization and its relationship to transportation accident analysis

    International Nuclear Information System (INIS)

    The response of potential waste forms should be determined for extreme transportation environments that must be postulated for environmental impact analysis and also for hypothetical accident conditions to which packagings and contents must be subjected for licensing purposes. The best approach may be to test materials up to and beyond their failure point; such an approach would establish failure thresholds. Specification of what denotes failure would be defined by existing or proposed regulations or dictated by requirements developed from accident analysis. Responses to physical and thermal insults are the most important for licensing or analysis and need to be thoroughly characterized. Others in need of characterization might be responses to extreme chemical environments and to intense and prolonged radiation exposure. A complete characterization of waste-form responses would be desirable for environments that are considered extreme for transportation accidents but which may be typical for processing or disposal environments. In addition, the characterizations that are performed must be completed in laboratory environments which can be readily correlated to accident environments and must be meaningfully conveyed to a transportation impact analyst. As an example, leaching data as commonly presented are not usable to the analyst and are obtained under conditions that are not directly applicable to conditions of most transportation accidents. Transportation analysts are in need of data useful for calculating environmental impacts and for licensing of packagings. Future waste form development programs and associated decisions should consider the needs of transportation analysts

  16. The Ceramic Waste Form Process at the Idaho National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Ken Bateman; Stephen Priebe

    2006-08-01

    The treatment of spent nuclear fuel for disposition using an electrometallurgical technique results in two high-level waste forms: a ceramic waste form (CWF) and a metal waste form (MWF). The CWF is a composite of sodalite and glass, which stabilizes the active fission products (alkali, alkaline earths, and rare earths) and transuranic (TRU) elements. Reactive metal fuel constituents, including all the TRU metals and the majority of the fission products remain in the salt as chlorides and are processed into the CWF. The solidified salt is containerized and transferred to the CWF process where it is ground in an argon atmosphere. Zeolite 4A is dried in a mechanically-fluidized dryer to about 0.1 wt% moisture and ground to a particle-size range of 45µ to 250µ. The salt and zeolite are mixed in a V-mixer and heated to 500°C for about 18 hours. During this process, the salt occludes into the structure of the zeolite. The salt-loaded zeolite (SLZ) is cooled and then mixed with borosilicate glass frit with a comparable particle-size range. The SLZ/glass mixture is transferred to a crucible, which is placed in a furnace and heated to 925°C. During this process, known as pressureless consolidation, the zeolite is converted to the final sodalite form and the glass thoroughly encapsulates the sodalite, producing a dense, leach-resistant final waste form. During the last several years, changes have occurred to the process, including: particle size of input materials and conversion from hot isostatic pressing to pressureless consolidation, This paper is intended to provide the current status of the CWF process focusing on the adaptation to pressureless consolidation. Discussions will include impacts of particle size on final waste form and the pressureless consolidation cycle. A model will be presented that shows the heating and cooling cycles and the effect of radioactive decay heat on the amount of fission products that can be incorporated into the CWF.

  17. Support for DOE program in mineral waste-form development

    International Nuclear Information System (INIS)

    This research investigation relates to sintered simulation ceramic waste forms of the generic SYNROC compositional type. Though they have been formulated with simulated wastes only, they serve as prototypes for potential hot, processed, crystalline waste forms whose combined thermodynamic stability and physical integrity are considered to render them capable of long-term imobilization of high-level radwastes under deep geologic disposal conditions. The problems involved are nontrivial, largely because of the very complex nature of the radwastes: a typical waste stream would contain more than 31 cation species. When the stabilizing matrix constituents are included, the final batch composition must successfully account (and find substitutional homes for some 35 different cation species. One of the important objectives of this study thus has been to develop a computer-based method for simulating these complex ion substitutions, and for calculating the resultant phase demands and batch formulations. Primary goals of the study have been (1) use of that computer simulation capability to incorporate rationally the radwaste ions from a specific waste stream (PW-7a) into the available SYNROC lattice sites and (2) utilization of existing ceramic processing and sintering methodologies to assure (and to understand) the attainment of high density, fine microstructure, full phase development and other features of the sintered product which are known to relate directly to its integrity and leach resistance. Though improved resistance to leaching has been a continuing goal, time and budget constraints have precluded initiation of any leachability studies of these new compositions during this contract period. 27 references, 15 figures, 6 tables

  18. Waste Marble Utilization from Residue Marble Industry as a Substitution of Cement and Sand within Concrete Rooftile Production

    Directory of Open Access Journals (Sweden)

    Candra Aditya, Abdul Halim, Chauliah Fatma Putri

    2014-01-01

    Full Text Available Research on alternative materials primarily from waste have been additional material at area manufacture of building materials , especially concreteroof tile [ 1 ] - [ 17 ] . This research will expand utilization of marble waste vBulletin East Java region of Indonesia in the manufacture of concrete roof tiles by combining the use of sand and waste marble powder as a substitute for riversand and portland cement .. This research creates material innovation product of environmentally friendly with relatively low prices without compromising quality. The purpose of research is to find the composition of the mixed-use waste marble tile that produces the most optimal strength . Experimental method used in this study to test the basic material and test physical and mechanical properties of concrete roof tiles ( bending loads , water absorption and resistance to water seepage in accordance with ISO 0096 : 2007 with eight variations in material composition . Concrete tile with marble waste produces a lighter weight 3.6 % - 12.3 % . Replacement PC with marble powder by 20 % qualify flexural strength , water absorption ( no more than 10 % and there is no seepage within 20 hours ± 5 minutes . Composition tile marble concrete using waste as a substitute for river sand PC and a decent and qualified SNI 0096:2007 is a composition of 0.8 PC : 0.2 SL : 1 Ps : 2 PSL and composition 0.8 PC : 0.2 SL : 3 PSL . While most optimum is 0.8 composition PC : 0.2 SL : 1 Ps : 2 PSL . which produces Flexture1141 N

  19. Physical, Chemical and Structural Evolution of Zeolite-Containing Waste Forms Produced from Metakaolinite and Calcined Sodium Bearing Waste (HLW and/or LLW)

    Energy Technology Data Exchange (ETDEWEB)

    Grutzeck, Michael W.

    2005-06-27

    Zeolites are extremely versatile. They can adsorb liquids and gases and serve as cation exchange media. They occur in nature as well cemented deposits. The ancient Romans used blocks of zeolitized tuff as a building material. Using zeolites for the management of radioactive waste is not a new idea, but a process by which the zeolites can be made to act as a cementing agent is. Zeolitic materials are relatively easy to synthesize from a wide range of both natural and man-made substances. The process under study is derived from a well known method in which metakaolin (an impure thermally dehydroxylated kaolinite heated to {approx}700 C containing traces of quartz and mica) is mixed with sodium hydroxide (NaOH) and reacted in slurry form (for a day or two) at mildly elevated temperatures. The zeolites form as finely divided powders containing micrometer ({micro}m) sized crystals. However, if the process is changed slightly and only just enough concentrated sodium hydroxide solution is added to the metakaolinite to make a thick crumbly paste and then the paste is compacted and cured under mild hydrothermal conditions (60-200 C), the mixture will form a hard ceramic-like material containing distinct crystalline tectosilicate minerals (zeolites and feldspathoids) imbedded in an X-ray amorphous hydrated sodium aluminosilicate matrix. Due to its lack of porosity and vitreous appearance we have chosen to call this composite a ''hydroceramic''.

  20. Transuranic contaminated waste form characterization and data base

    International Nuclear Information System (INIS)

    This volume contains 5 appendices. Title listing are: technologies for recovery of transuranics; nondestructive assay of TRU contaminated wastes; miscellaneous waste characteristics; acceptance criteria for TRU waste; and TRU waste treatment technologies

  1. Description of DWPF reference waste form and canister

    Energy Technology Data Exchange (ETDEWEB)

    1981-06-01

    This document describes the reference waste form and canister for the Defense Waste Processing Facility (DWPF). The facility is planned for location at the Savannah River Plant in Aiken, SC, and is scheduled for construction authorization during FY-1983. The reference canister is fabricated of 24-in.-OD 304L stainless steel pipe with a dished bottom, domed head, and lifting and welding flanges on the head neck. The overall canister length is 9 ft 10 in., with a wall thickness of 3/8-in. (schedule 20 pipe). The canister length was selected to reduce equipment cell height in the DWPF to a practical size. The canister diameter was selected to ensure that a filled canister with its shipping cask could be accommodated on a legal-weight truck. The overall dimensions and weight appear to be generally compatible with preliminary assessments of repository requirements. The reference waste form is borosilicate glass containing approximately 28 wt % sludge oxides with the balance glass frit. Borosilicate glass was chosen because of its high resistance to leaching by water, its relatively high solubility for nuclides found in the sludge, and its reasonably low melting temperature. The glass frit contains approximately 58% SiO/sub 2/ and 15% B/sub 2/O/sub 3/. This composition results in a low average leachability in the waste form of approximately 5 x 10/sup -9/ g/cm/sup 2/-day based on /sup 137/Cs over 365 days in 25/sup 0/C water. The canister is filled with 3260 lb of glass which occupies about 85% of the free canister volume. The filled canister will generate approximately 425 watts when filled with oxides from 5-year-old sludge and 15-year-old supernate from the Stage 1 and Stage 2 processes. The radionuclide content of the canister is about 150,000 curies, with a radiation level of 2 x 10/sup 4/ rem/hour at 1 cm.

  2. Immobilization of fission products in phosphate ceramic waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Singh, D.; Wagh, A. [Argonne National Lab., IL (United States)

    1997-10-01

    Chemically bonded phosphate ceramics (CBPCs) have several advantages that make them ideal candidates for containing radioactive and hazardous wastes. In general, phosphates have high solid-solution capacities for incorporating radionuclides, as evidenced by several phosphates (e.g., monazites and apatites) that are natural analogs of radioactive and rare-earth elements. The phosphates have high radiation stability, are refractory, and will not degrade in the presence of internal heating by fission products. Dense and hard CBPCs can be fabricated inexpensively and at low temperature by acid-base reactions between an inorganic oxide/hydroxide powder and either phosphoric acid or an acid-phosphate solution. The resulting phosphates are extremely insoluble in aqueous media and have excellent long-term durability. CBPCs offer the dual stabilization mechanisms of chemical fixation and physical encapsulation, resulting in superior waste forms. The goal of this task is develop and demonstrate the feasibility of CBPCs for S/S of wastes containing fission products. The focus of this work is to develop a low-temperature CBPC immobilization system for eluted {sup 99}Tc wastes from sorption processes.

  3. A Fabrication of Salt-Loaded Ceramic Waste Form

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jeong-Guk; Lee, Jae-Hee; Kim, Hwan-Yong; Lee, Sung-Ho; Kim, In-Tae [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2006-07-01

    The waste salts such as molten LiCl salt from an oxide fuel reduction process and molten LiCl-KCl eutectic salt from an electro refining process must meet the acceptance criteria for a disposal in geological repository. Two of the more important criteria in waste form containing chloride salts are known to be leach resistance and waste form durability. According to US Argonne National Laboratory (ANL), a ceramic waste form (CWF), which was prepared by pressureless consolidation (PC) of eutectic LiCl-KCl salt loaded zeolite (SLZ), has as a good quality as that of high level radioactive waste (HLW) glass. ANL has developed the CWF fabrication method as follows: The eutectic LiCl-KCl salt recovered from the electorefiner is size-reduced to facilitate occlusion in zeolite by crushing and grinding under an argon atmosphere. The crushed salt is mechanically mixed with dried zeolite 4A in a V-mixer at a salt loading about 10 mass % then heated to about 723 K for 16 h to occlude salt within the zeolite cages. And the SLZ is then mixed with a borosilicate binder glass in a V-mixer (without heating) at a 3:1 mass ratio. The mixture is loaded into fill cans the processed at about 1188 K for about 72 h. As the mixture is heated above about 1123 K during the encapsulating step, the SLZ converts to the mineral sodalite, Na{sub 8}(AlSiO{sub 4}){sub 6}Cl{sub 2}, which incorporates most of the occluded salt into its structure. The glass becomes sufficiently fluid during bonded sodalite material referred to as the CWF, which contains 8 mass % salt. We also developed the CWR fabrication technology for a waste LiCl salt from an electrolytic reduction process (ACP; advanced spent fuel conditioning process). A melting point of the LiCl salt is higher than that of the eutectic LiCl-KCl salt, therefore, a crystalline behavior during producing CWF is somewhat different from that of ANL. Therefore, the changes of immobilization media, which had been started from zeolite A, for the

  4. Preparation of plutonium waste forms with ICPP calcined high-level waste

    International Nuclear Information System (INIS)

    Glass and glass-ceramic forms developed for the immobilization of calcined high-level wastes generated by Idaho Chemical Processing Plant (ICPP) fuel reprocessing activities have been investigated for ability to immobilize plutonium and to simultaneously incorporate calcined waste as an anti-proliferation barrier. Within the forms investigated, crystallization of host phases result in an increased loading of plutonium as well as its incorporation into potentially more durable phases than the glass. The host phases were initially formed and characterized with cerium (Ce+4) as a surrogate for plutonium (Pu+4) and samarium as a neutron absorber for criticality control. Verification of the surrogate testing results were then performed replacing cerium with plutonium. All testing was performed with surrogate calcined high-level waste. The results of these tests indicated that a potentially useful host phase, based on zirconia, can be formed either by devitrification or solid state reaction in the glass studied. This phase incorporates plutonium as well as samarium and the calcined waste becomes part of the matrix. Its ease of formation makes it potentially useful in excess plutonium dispositioning. Other durable host phases for plutonium and samarium, including zirconolite and zircon have been formed from zirconia or alumina calcine through cold press-sintering techniques and hot isostatic pressing. Host phase formation experiments conducted through vitrification or by cold press-sintering techniques are described and the results discussed. Recommendations are given for future work that extends the results of this study

  5. Pyrochlore as nuclear waste form. Actinide uptake and chemical stability

    Energy Technology Data Exchange (ETDEWEB)

    Finkeldei, Sarah Charlotte

    2015-07-01

    Radioactive waste is generated by many different technical and scientific applications. For the past decades, different waste disposal strategies have been considered. Several questions on the waste disposal strategy remain unanswered, particularly regarding the long-term radiotoxicity of minor actinides (Am, Cm, Np), plutonium and uranium. These radionuclides mainly arise from high level nuclear waste (HLW), specific waste streams or dismantled nuclear weapons. Although many countries have opted for the direct disposal of spent fuel, from a scientific and technical point of view it is imperative to pursue alternative waste management strategies. Apart from the vitrification, especially for trivalent actinides and Pu, crystalline ceramic waste forms are considered. In contrast to glasses, crystalline waste forms, which are chemically and physically highly stable, allow the retention of radionuclides on well-defined lattice positions within the crystal structure. Besides polyphase ceramics such as SYNROC, single phase ceramics are considered as tailor made host phases to embed a specific radionuclide or a specific group. Among oxidic single phase ceramics pyrochlores are known to have a high potential for this application. This work examines ZrO{sub 2} based pyrochlores as potential nuclear waste forms, which are known to show a high aqueous stability and a high tolerance towards radiation damage. This work contributes to (1) understand the phase stability field of pyrochlore and consequences of non-stoichiometry which leads to pyrochlores with mixed cationic sites. Mixed cationic occupancies are likely to occur in actinide-bearing pyrochlores. (2) The structural uptake of radionuclides themselves was studied. (3) The chemical stability and the effect of phase transition from pyrochlore to defect fluorite were probed. This phase transition is important, as it is the result of radiation damage in ZrO{sub 2} based pyrochlores. ZrO{sub 2} - Nd{sub 2}O{sub 3} pellets

  6. Commercial high-level-waste management: options and economics. A comparative analysis of the ceramic and glass waste forms

    Energy Technology Data Exchange (ETDEWEB)

    McKisson, R.L.; Grantham, L.F.; Guon, J.; Recht, H.L.

    1983-02-25

    Results of an estimate of the waste management costs of the commercial high-level waste from a 3000 metric ton per year reprocessing plant show that the judicious use of the ceramic waste form can save about $2 billion during a 20-year operating campaign relative to the use of the glass waste form. This assumes PWR fuel is processed and the waste is encapsulated in 0.305-m-diam canisters with ultimate emplacement in a BWIP-type horizontal-borehole repository. The estimated total cost (capital and operating) of the management in the ceramic form is $2.0 billion, and that of the glass form is $4.0 billion. Waste loading and waste form density are the driving factors in that the low-waste loading (25%) and relatively low density (3.1 g/cm/sup 3/) characteristic of the glass form require several times as many canisters to handle a given waste throughput than is needed for the ceramic waste form whose waste loading capability exceeds 60% and whose waste density is nominally 5.2 g/cm/sup 3/) characteristic of the glass form requires several times as many canisters to handle a given waste throughput than is needed for the ceramic waste form whose waste loading capability exceeds 60% and whose waste density is nominally 5.2 g/cm/sup 3/. The minimum-cost ceramic waste form has a 60 wt. % waste loading of commercial high-level waste and requires 25 years storage before emplacement in basalt with delayed backfill. Because of the process flexibility allowed by the availability of the high-waste loading of the ceramic form, the intermediate-level liquid waste stream can be mixed with the high-level liquid waste stream and economically processed and emplaced. The cost is greater by $0.3 billion than that of the best high-level liquid waste handling process sequence ($2.3 billion vs $2.0 billion), but this difference is less than the cost of the separate disposal of the intermediate-level liquid waste.

  7. Commercial high-level-waste management: options and economics. A comparative analysis of the ceramic and glass waste forms

    International Nuclear Information System (INIS)

    Results of an estimate of the waste management costs of the commercial high-level waste from a 3000 metric ton per year reprocessing plant show that the judicious use of the ceramic waste form can save about $2 billion during a 20-year operating campaign relative to the use of the glass waste form. This assumes PWR fuel is processed and the waste is encapsulated in 0.305-m-diam canisters with ultimate emplacement in a BWIP-type horizontal-borehole repository. The estimated total cost (capital and operating) of the management in the ceramic form is $2.0 billion, and that of the glass form is $4.0 billion. Waste loading and waste form density are the driving factors in that the low-waste loading (25%) and relatively low density (3.1 g/cm3) characteristic of the glass form require several times as many canisters to handle a given waste throughput than is needed for the ceramic waste form whose waste loading capability exceeds 60% and whose waste density is nominally 5.2 g/cm3) characteristic of the glass form requires several times as many canisters to handle a given waste throughput than is needed for the ceramic waste form whose waste loading capability exceeds 60% and whose waste density is nominally 5.2 g/cm3. The minimum-cost ceramic waste form has a 60 wt. % waste loading of commercial high-level waste and requires 25 years storage before emplacement in basalt with delayed backfill. Because of the process flexibility allowed by the availability of the high-waste loading of the ceramic form, the intermediate-level liquid waste stream can be mixed with the high-level liquid waste stream and economically processed and emplaced. The cost is greater by $0.3 billion than that of the best high-level liquid waste handling process sequence ($2.3 billion vs $2.0 billion), but this difference is less than the cost of the separate disposal of the intermediate-level liquid waste

  8. High Solids Consolidated Incinerator Facility (CIF) Wastes Stabilization with Ceramicrete and Super Cement

    Energy Technology Data Exchange (ETDEWEB)

    Walker, B.W.

    1999-09-14

    High Solids ash and scrubber solution waste streams were generated at the incinerator facility at SRS by burning radioactive diatomaceous filter rolls which contained small amounts of uranium, and listed solvents (F and U). This report details solidification activities using selected Mixed Waste Focus Area (MWFA) technologies with the High Solids waste streams.

  9. Radiation damage studies related to nuclear waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Gray, W.J.; Wald, J.W.; Turcotte, R.P.

    1981-12-01

    Much of the previously reported work on alpha radiation effects on crystalline phases of importance to nuclear waste forms has been derived from radiation effects studies of composite waste forms. In the present work, two single-phase crystalline materials, Gd/sub 2/Ti/sub 2/O/sub 7/ (pyrochlore) and CaZrTi/sub 2/O/sub 7/ (zirconolite), of relative importance to current waste forms were studied independently by doping with /sup 244/Cm at the 3 wt % level. Changes in the crystalline structure measured by x-ray diffraction as a function of dose show that damage ingrowth follows an expected exponential relationship of the form ..delta..V/V/sub 0/ = A(1-exp(-BD)). In both cases, the materials became x-ray amorphous before the estimated saturation value was reached. The predicted magnitudes of the unit cell volume changes at saturation are 5.4% and 3.5%, respectively, for Gd/sub 2/Ti/sub 2/O/sub 7/ and CaZrTi/sub 2/O/sub 7/. The later material exhibited anisotropic behavior in which the expansion of the monoclinic cell in the c/sub 0/ direction was over five times that of the a/sub 0/ direction. The effects of transmutations on the properties of high-level waste solids have not been studied until now because of the long half-lives of the important fission products. This problem was circumvented in the present study by preparing materials containing natural cesium and then irradiating them with neutrons to produce /sup 134/Cs, which has only a 2y half-life. The properties monitored at about one year intervals following irradiation have been density, leach rate and microstructure. A small amount of x-ray diffraction work has also been done. Small changes in density and leach rate have been observed for some of the materials, but they were not large enough to be of any consequence for the final disposal of high level wastes.

  10. Crystalline ceramics: Waste forms for the disposal of weapons plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Ewing, R.C.; Lutze, W. [New Mexico Univ., Albuquerque, NM (United States); Weber, W.J. [Pacific Northwest Lab., Richland, WA (United States)

    1995-05-01

    At present, there are three seriously considered options for the disposition of excess weapons plutonium: (i) incorporation, partial burn-up and direct disposal of MOX-fuel; (ii) vitrification with defense waste and disposal as glass ``logs``; (iii) deep borehole disposal (National Academy of Sciences Report, 1994). The first two options provide a safeguard due to the high activity of fission products in the irradiated fuel and the defense waste. The latter option has only been examined in a preliminary manner, and the exact form of the plutonium has not been identified. In this paper, we review the potential for the immobilization of plutonium in highly durable crystalline ceramics apatite, pyrochlore, monazite and zircon. Based on available data, we propose zircon as the preferred crystalline ceramic for the permanent disposition of excess weapons plutonium.

  11. Transmission electron microscopy analysis of corroded metal waste forms.

    Energy Technology Data Exchange (ETDEWEB)

    Dietz, N. L.

    2005-04-15

    This report documents the results of analyses with transmission electron microscopy (TEM) combined with energy dispersive X-ray spectroscopy (EDS) and selected area electron diffraction (ED) of samples of metallic waste form (MWF) materials that had been subjected to various corrosion tests. The objective of the TEM analyses was to characterize the composition and microstructure of surface alteration products which, when combined with other test results, can be used to determine the matrix corrosion mechanism. The examination of test samples generated over several years has resulted in refinements to the TEM sample preparation methods developed to preserve the orientation of surface alteration layers and the underlying base metal. The preservation of microstructural spatial relationships provides valuable insight for determining the matrix corrosion mechanism and for developing models to calculate radionuclide release in repository performance models. The TEM results presented in this report show that oxide layers are formed over the exposed steel and intermetallic phases of the MWF during corrosion in aqueous solutions and humid air at elevated temperatures. An amorphous non-stoichiometric ZrO{sub 2} layer forms at the exposed surfaces of the intermetallic phases, and several nonstoichiometric Fe-O layers form over the steel phases in the MWF. These oxide layers adhere strongly to the underlying metal, and may be overlain by one or more crystalline Fe-O phases that probably precipitated from solution. The layer compositions are consistent with a corrosion mechanism of oxidative dissolution of the steel and intermetallic phases. The layers formed on the steel and intermetallic phases form a continuous layer over the exposed waste form, although vertical splits in the layer and corrosion in pits and crevices were seen in some samples. Additional tests and analyses are needed to verify that these layers passivate the underlying metals and if passivation can break

  12. Fundamental Science-Based Simulation of Nuclear Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Devanathan, Ramaswami; Gao, Fei; Sun, Xin; Khaleel, Mohammad A.

    2010-10-04

    This report presents a hierarchical multiscale modeling scheme based on two-way information exchange. To account for all essential phenomena in waste forms over geological time scales, the models have to span length scales from nanometer to kilometer and time scales from picoseconds to millenia. A single model cannot cover this wide range and a multi-scale approach that integrates a number of different at-scale models is called for. The approach outlined here involves integration of quantum mechanical calculations, classical molecular dynamics simulations, kinetic Monte Carlo and phase field methods at the mesoscale, and continuum models. The ultimate aim is to provide science-based input in the form of constitutive equations to integrated codes. The atomistic component of this scheme is demonstrated in the promising waste form xenotime. Density functional theory calculations have yielded valuable information about defect formation energies. This data can be used to develop interatomic potentials for molecular dynamics simulations of radiation damage. Potentials developed in the present work show a good match for the equilibrium lattice constants, elastic constants and thermal expansion of xenotime. In novel waste forms, such as xenotime, a considerable amount of data needed to validate the models is not available. Integration of multiscale modeling with experimental work is essential to generate missing data needed to validate the modeling scheme and the individual models. Density functional theory can also be used to fill knowledge gaps. Key challenges lie in the areas of uncertainty quantification, verification and validation, which must be performed at each level of the multiscale model and across scales. The approach used to exchange information between different levels must also be rigorously validated. The outlook for multiscale modeling of wasteforms is quite promising.

  13. Durability of class C fly ash belite cement in simulated sodium chloride radioactive liquid waste: Influence of temperature

    Energy Technology Data Exchange (ETDEWEB)

    Guerrero, A. [Eduardo Torroja Institute for Construction Science (CSIC), C/Serrano Galvache 4, 28033 Madrid (Spain)], E-mail: aguerrero@ietcc.csic.es; Goni, S. [Eduardo Torroja Institute for Construction Science (CSIC), C/Serrano Galvache 4, 28033 Madrid (Spain)], E-mail: sgoni@ietcc.csic.es; Allegro, V.R. [Eduardo Torroja Institute for Construction Science (CSIC), C/Serrano Galvache 4, 28033 Madrid (Spain)], E-mail: allegro@ietcc.csic.es

    2009-03-15

    This work is a continuation of a previous durability study of class C fly ash belite cement (FABC-2-W) in simulated radioactive liquid waste (SRLW) that is very rich in sulphate salts. The same experimental methodology was applied in the present case, but with a SRLW rich in sodium chloride. The study was carried out by testing the flexural strength of mortars immersed in simulated radioactive liquid waste that was rich in chloride (0.5 M), and demineralised water as a reference, at 20 and 40 deg. C over a period of 180 days. The reaction mechanism of chloride ions with the mortar was evaluated by scanning electron microscopy (SEM), porosity and pore-size distribution, and X-ray diffraction (XRD). The results showed that the FABC mortar was stable against simulated chloride radioactive liquid waste (SCRLW) attack at the two chosen temperatures. The enhancement of mechanical properties was a result of the formation of non-expansive Friedel's salt inside the pores; accordingly, the microstructure was refined.

  14. Immobilization of tritiated waste-water by hydraulic cements. A survey of the state-of-the-art

    International Nuclear Information System (INIS)

    An experimental research programme including as one of its major items the definition of a strategy for tritiated waste management is being prepared at the JRC-Ispra. Laboratory work will be performed in ETHEL, the European Tritium Handling Experimental Laboratory under construction at Ispra. To provide background information needed for defining items and planning the execution of such activities, a survey of the state of the art and R and D performed in the field of tritiated water immobilization by hydraulic cements and ultimate packaging by multiple containment systems has been carried out. Particular attention has been focused on results of tritium leach test programmes performed at various USA laboratories in order to verify the leach resistance properties of some tritium immobilization and containment options. Problems and draw backs associated with such options are discussed. Final conclusions are presented. 49 refs

  15. Analysis by X-Ray images of EVA waste incorporated in Portland Cement; Analise atraves de imagens de raios X da incorporacao de residuo de EVA em cimento Portland

    Energy Technology Data Exchange (ETDEWEB)

    Marques, M.A.; Antunes, M.L.P.; Montagnoli, R.M.; Mancini, S.D., E-mail: marciomq@sorocaba.unesp.br [Universidade Estadual Paulista Julio de Mesquita Filho (UNESP), Sorocaba, SP (Brazil)

    2012-07-01

    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)

  16. Naturally occurring crystalline phases: analogues for radioactive waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Haaker, R.F.; Ewing, R.C.

    1981-01-01

    Naturally occurring mineral analogues to crystalline phases that are constituents of crystalline radioactive waste forms provide a basis for comparison by which the long-term stability of these phases may be estimated. The crystal structures and the crystal chemistry of the following natural analogues are presented: baddeleyite, hematite, nepheline; pollucite, scheelite;sodalite, spinel, apatite, monazite, uraninite, hollandite-priderite, perovskite, and zirconolite. For each phase in geochemistry, occurrence, alteration and radiation effects are described. A selected bibliography for each phase is included.

  17. Naturally occurring crystalline phases: analogues for radioactive waste forms

    International Nuclear Information System (INIS)

    Naturally occurring mineral analogues to crystalline phases that are constituents of crystalline radioactive waste forms provide a basis for comparison by which the long-term stability of these phases may be estimated. The crystal structures and the crystal chemistry of the following natural analogues are presented: baddeleyite, hematite, nepheline; pollucite, scheelite;sodalite, spinel, apatite, monazite, uraninite, hollandite-priderite, perovskite, and zirconolite. For each phase in geochemistry, occurrence, alteration and radiation effects are described. A selected bibliography for each phase is included

  18. Calcium sulfoaluminate cement blended with OPC: A potential binder to encapsulate low-level radioactive slurries of complex chemistry

    International Nuclear Information System (INIS)

    Investigations were carried out in order to solidify in cement a low-level radioactive waste of complex chemistry obtained by mixing two process streams, a slurry produced by ultra-filtration and an evaporator concentrate with a salinity of 600 gxL-1. Direct cementation with Portland cement (OPC) was not possible due to a very long setting time of cement resulting from borates and phosphates contained in the waste. According to a classical approach, this difficulty could be solved by pre-treating the waste to reduce adverse cement-waste interactions. A two-stage process was defined, including precipitation of phosphates and sulfates at 60 deg. C by adding calcium and barium hydroxide to the waste stream, and encapsulation with a blend of OPC and calcium aluminate cement (CAC) to convert borates into calcium quadriboroaluminate. The material obtained with a 30% waste loading complied with specifications. However, the pre-treatment step made the process complex and costly. A new alternative was then developed: the direct encapsulation of the waste with a blend of OPC and calcium sulfoaluminate cement (CSA) at room temperature. Setting inhibition was suppressed, which probably resulted from the fact that, when hydrating, CSA cement formed significant amounts of ettringite and calcium monosulfoaluminate hydrate which incorporated borates into their structure. As a consequence, the waste loading could be increased to 56% while keeping acceptable properties at the laboratory scale.

  19. Colloid formation during waste form reaction: implications for nuclear waste disposal

    Science.gov (United States)

    Bates, J. K.; Bradley, J.; Teetsov, A.; Bradley, C. R.; ten Brink, Marilyn Buchholtz

    1992-01-01

    Insoluble plutonium- and americium-bearing colloidal particles formed during simulated weathering of a high-level nuclear waste glass. Nearly 100 percent of the total plutonium and americium in test ground water was concentrated in these submicrometer particles. These results indicate that models of actinide mobility and repository integrity, which assume complete solubility of actinides in ground water, underestimate the potential for radionuclide release into the environment. A colloid-trapping mechanism may be necessary for a waste repository to meet long-term performance specifications.

  20. Rice husk derived waste materials as partial cement replacement in lightweight concrete Utilização de resíduos derivados da casca de arroz como substitutos parciais do cimento no concreto leve

    OpenAIRE

    Celso Yoji Kawabata; Holmer Savastano Junior; Joana Sousa-Coutinho

    2012-01-01

    In this study rice husk ash (RHA) and broiler bed ash from rice husk (BBA), two agricultural waste materials, have been assessed for use as partial cement replacement materials for application in lightweight concrete. Physical and chemical characteristics of RHA and BBA were first analyzed. Three similar types of lightweight concrete were produced, a control type in which the binder was just CEMI cement (CTL) and two other types with 10% cement replacement with, respectively, RHA and BBA. All...

  1. Engineering-Scale Demonstration of DuraLith and Ceramicrete Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Josephson, Gary B.; Westsik, Joseph H.; Pires, Richard P.; Bickford, Jody; Foote, Martin W.

    2011-09-23

    To support the selection of a waste form for the liquid secondary wastes from the Hanford Waste Immobilization and Treatment Plant, Washington River Protection Solutions (WRPS) has initiated secondary waste form testing on four candidate waste forms. Two of the candidate waste forms have not been developed to scale as the more mature waste forms. This work describes engineering-scale demonstrations conducted on Ceramicrete and DuraLith candidate waste forms. Both candidate waste forms were successfully demonstrated at an engineering scale. A preliminary conceptual design could be prepared for full-scale production of the candidate waste forms. However, both waste forms are still too immature to support a detailed design. Formulations for each candidate waste form need to be developed so that the material has a longer working time after mixing the liquid and solid constituents together. Formulations optimized based on previous lab studies did not have sufficient working time to support large-scale testing. The engineering-scale testing was successfully completed using modified formulations. Further lab development and parametric studies are needed to optimize formulations with adequate working time and assess the effects of changes in raw materials and process parameters on the final product performance. Studies on effects of mixing intensity on the initial set time of the waste forms are also needed.

  2. [Concentration and form of asbestos fibers in tap drinking water contaminated from a water supply pipe with asbestos-cement].

    Science.gov (United States)

    Saitoh, K; Takizawa, Y; Muto, H; Hirano, K

    1992-10-01

    The identification and concentration of asbestos fibers in tap drinking water supplied in a central area of Akita Prefecture, Japan, were determined by phase-contrast microscopy and a scanning electron microscope equipped with an energy-dispersive X-ray microanalyzer. The following results were obtained. 1. Asbestos fibers were found in the tap water from two areas in which an asbestos-cement pipe was used for public water supply. The concentrations of asbestos fibers in the tap water were 2.7 x 10(4) to 27.0 x 10(4) fibers per liter of water in area A and 10.0 x 10(4) to 21.0 x 10(4) in area B. On the other hand, no asbestos fiber contamination was observed in tap water of area C, which shared a common water source with area A. A vinyl chloride pipe was used over the entire length of the water supply in route C. 2. Crocidolite was the predominant type of asbestos fiber detected in the tap water. Chrysotile and a mixture of chrysotile and amosite were also observed. 3. Almost all asbestos fibers detected in the tap water possessed the form of thick or sheaved fibers with lengths ranging from ca. 5 to 10 microns. Their shapes were very different from those of asbestos fibers found in the atmosphere. The typical form of the latter is short (ca. 1 micron in length) and needle-like. 4. It was suggested that the contamination of asbestos fibers in the tap water was caused by erosion and peeling off of the inner wall of the asbestos-cement pipe used as a conduit. In order to evaluate the safety of drinking water in Japan, an extensive survey on asbestos-fiber contamination in tap water is necessary. PMID:1464953

  3. Radiation and Thermal Stability of Murataite Ceramics Nuclear Waste Forms

    Science.gov (United States)

    Lian, J.; Yudintsev, S. V.; Stefanovsky, S. V.

    2006-05-01

    The wide range of complex nuclear wastes requires a variety of robust hosts for long-term storage during disposal. Wastes with high actinide and iron concentrations have generated intense interest in murataite ceramics as a candidate waste form due to its four distinct cation sites as well as cation vacancies. Critical to this application is the radiation stability of the waste host. We have determined both the radiation and thermal stabilities of murataite ceramics using in situ observations in a transmission electron microscope during ion bombardment at the Electron Microscopy Center at Argonne National Laboratory. A central issue for structural stability is radiation damage-induced crystalline-to-amorphous transformation that may result in macroscopic swelling, cracking and phase decomposition. Such a response would lead to a significant change in chemical durability and release of incorporated radionuclides. We found that, murataite ceramics are susceptible to ion beam induce ordered-disordered transition and amorphization. The ion dose required for amorphization was determined as a function of temperature and the degree of initial structural disorder. The upper temperature limit for amorphization of murataites was determined to be in the range of 860 K to 1060 K for 1 MeV Kr2+ ion irradiation. Decrease of the susceptibility to irradiation induced amorphization for disordered murataite, suggests that the amorphization susceptibility depends, in part, on the initial degree of intrinsic disorder prior to irradiation. The thermal stability of murataite polytypes was studied by in-situ TEM observation. Phase decomposition with the precipitation of Fe-rich nanocrystals was induced in the murataite structure. The phase decomposition and nanocrystal formation have no significant effects on the radiation resistance of murataite ceramics used as potential host phases for the immobilization of actinides.

  4. Garnet nuclear waste forms – Solubility at repository conditions

    Energy Technology Data Exchange (ETDEWEB)

    Caporuscio, F.A., E-mail: floriec@lanl.gov [EES-14, Los Alamos National Laboratory, NM 87545 (United States); Scott, B.L. [MPA-MSID, Los Alamos National Laboratory, NM 87545 (United States); Xu, H. [EES-14, Los Alamos National Laboratory, NM 87545 (United States); Feller, R.K. [Effect Materials Research Group, BASF Corporation, 500 White Plains Road, Tarrytown, NY 10591 (United States)

    2014-01-15

    Highlights: • Rare-earth elements are a significant waste stream produced by nuclear fuel cycles. • Suitability of garnets as potential waste forms. • Single-crystal X-ray structural refinements for grossular, LuAG and YAG. • Garnets have low solubility, flexible crystal structure to take on large cations. • Demonstrate garnets are potentially robust waste forms for radioactive REE. -- Abstract: Radioactive rare-earth elements (REEs) constitute a significant waste stream produced from modified open and full nuclear fuel cycles. Immobilization of these REE radionuclides is thus important for sustainable nuclear energy growth. In this work, we investigated the suitability of garnets as potential waste forms for REEs by measuring their aqueous stability at repository conditions. Three garnet samples, including one natural grossular (Ca{sub 3}Al{sub 2}Si{sub 3}O{sub 12}) and two synthetic phases (LuAG – Lu{sub 3}Al{sub 5}O{sub 12} and YAG – Y{sub 3}Al{sub 5}O{sub 12}), were studied. Single-crystal X-ray structural refinements show that the unit-cell volumes increase from 1657.19 Å{sup 3} for grossular to 1679.8 Å{sup 3} for LuAG and to 1721.7 Å{sup 3} for YAG. This trend is due to increases in ionic radii in both the 8-coordinated X (from Ca to Lu to Y) and 4-coordinated Z (from Si to Al) cations. Hydrothermal experiments of the three samples were performed at 200 °C and 150 bar for 4 weeks using water and brine solutions to evaluate their solubility. The natural grossular sample exhibited Al leach rates ranging from 2.5 × 10{sup −4} to 6.43 × 10{sup −5} g/L·day and Ca leach rates from 1.39 × 10{sup −3} to 4.57 × 10{sup −3} g/L·day, indicating incongruent nature of the cation dissolution. The LuAG sample exhibited Lu leach rates of 3.73 × 10{sup −4} to 2.19 × 10{sup −4} g/L·day, and the YAG sample had Y leach rates of 1.29 × 10{sup −4} to 5.64 × 10{sup −5} g/L·day. Although these samples are generally more soluble in

  5. α and long-lived βγ waste source term. A first generation model for a deep cemented waste repository

    International Nuclear Information System (INIS)

    According to the normal scenario of radioactivity release to the biosphere, only long-lived nuclides are able to migrate significantly to the surface. A first generation model, concerning a cemented waste of hulls and ends deeply disposed of in a granitic medium is in progress at CEA. Two nuclides have been selected: 237-Neptunium (as a reference of α emitters) and 135-Cesium (as a reference of long-lived β emitters). Attributing the long term activity to these both nuclides leads to a model which is conservative beyond ca. 150000 years. Principal difficulties arise from physico-chemical behaviour of Neptunium in aqueous phase, and from non-linear Cesium adsorption on various media. Condiment code (versions 2 and 3), which is developed parallely to the present model is conceived to take account for these phenomena

  6. Description of DWPF reference waste form and canister

    International Nuclear Information System (INIS)

    This document describes the reference waste form and canister for the Defense Waste Processing Facility (DWPF). The facility is planned for location at the Savannah River Plant in Aiken, SC, and is scheduled for construction authorization during FY-1983. The reference canister is fabricated of 24-in.-OD 304L stainless steel pipe with a dished bottom, domed head, and lifting and welding flanges on the head neck. The overall canister length is 9 ft 10 in., with a wall thickness of 3/8-in. (schedule 20 pipe). The canister length was selected to reduce equipment cell height in the DWPF to a practical size. The canister diameter was selected to ensure that a filled canister with its shipping cask could be accommodated on a legal-weight truck. The overall dimensions and weight appear to be generally compatible with preliminary assessments of repository requireiajps. The rabarajca saspa bkri is bkrksilicapa class cojtaining approximately 28 wt % sludge oxides with the balance glass frit. Borosilicate glass was chosen because of its high resistance to leaching by water, its relatively high solubility for nuclides found in the sludge, and its reasonably low melting temperature. The glass frit contains approximately 58% SiO2 and 15% B2O3. This composition results in a low average leachability in the waste form of approximately 5 x 10-9 g/cm2-day based on 137Cs over 365 days in 250C water. The canister is filled with 3260 lb of glass which occupies about 85% of the free canister volume. The filled canister will generate approximately 425 watts when filled with oxides from 5-year-old sludge and 15-year-old supernate from the Stage 1 and Stage 2 processes. The radionuclide content of the canister is about 150,000 curies, with a radiation level of 2 x 104 rem/hour at 1 cm

  7. Stability of High-Level Radioactive Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Besmann, T.M.

    2001-06-22

    High-level waste (HLW) glass compositions, processing schemes, limits on waste content, and corrosion/dissolution release models are dependent on an accurate knowledge of melting temperatures and thermochemical values. Unfortunately, existing models for predicting these temperatures are empirically-based, depending on extrapolations of experimental information. In addition, present models of leaching behavior of glass waste forms use simplistic assumptions or experimentally measured values obtained under non-realistic conditions. There is thus a critical need for both more accurate and more widely applicable models for HLW glass behavior, which this project addressed. Significant progress was made in this project on modeling HLW glass. Borosilicate glass was accurately represented along with the additional important components that contain iron, lithium, potassium, magnesium, and calcium. The formation of crystalline inclusions in the glass, an issue in Hanford HLW formulations, was modeled and shown to be predictive. Thus the results of this work have already demonstrated practical benefits with the ability to map compositional regions where crystalline material forms, and therefore avoid that detrimental effect. With regard to a fundamental understanding, added insights on the behavior of the components of glass have been obtained, including the potential formation of molecular clusters. The EMSP project had very significant effects beyond the confines of Environmental Management. The models developed for glass have been used to solve a very costly problem in the corrosion of refractories for glass production. The effort resulted in another laboratory, Sandia National Laboratories-Livermore, to become conversant in the techniques and to apply those through a DOE Office of Industrial Technologies project joint with PPG Industries. The glass industry as a whole is now cognizant of these capabilities, and there is a Glass Manufacturer's Research Institute

  8. Re-use of stabilised flue gas ashes from solid waste incineration in cement-treated base layers for pavements.

    Science.gov (United States)

    Cai, Zuansi; Jensen, Dorthe L; Christensen, Thomas H; Bager, Dirch H

    2003-02-01

    Fly ash from coal-burning power plants has been used extensively as a pozzolan and fine filler in concrete for many years. Laboratory experiments were performed investigating the effect of substituting the coal-based fly ash with chemically stabilised flue gas ashes (FGA) from waste incineration. Two types of FGA were treated by the Ferrox-process, which removes the majority of the easily soluble salts in the FGA and provides binding sites for heavy metals in terms of ferrihydrite. Cubes of cement treated base layer materials containing 5% stabilised FGA were cast, sealed and cured for two weeks. Cylinders (diameter 100 mm, length 150 mm) were drilled from these cubes for tank leaching experiments. Duplicate specimens were subject to compression strength testing and to tank leaching experiments. The compressive strength of the CTB fulfilled the Danish requirements for CTB, i.e. strength more than 5 MPa after 7 days. The tank leaching tests revealed that leaching of heavy metals was not significantly affected by the use of chemically stabilised flue gas ashes from waste incineration. Assuming that diffusion controls the leaching process it was calculated that less than 1% of the metals would leach during a 100-year period from a 0.5 m thick concrete slab exposed to water on one side. Leaching of the common ions Ca, Cl, Na and SO4 was increased 3-20 times from the specimens with chemically stabilised flue gas ashes from waste incineration. However, the quantities leached were still modest. These experiments suggest that FGA from waste incineration after Ferrox-treatment could be re-used in CTB without compromising the strength and leaching from the base layer.

  9. Study of Structural Performance and Durability of Concrete by Partial Replacement of Cement with Hypo Sludge (Paper waste

    Directory of Open Access Journals (Sweden)

    K. Hari kishan,

    2015-12-01

    Full Text Available Utilization of industrial waste products as supplementary cementitious materials (SCM in concrete making is very important aspect in view of economical, environmental and technical reasons. As these supplementary cementitious materials have different chemical and mineralogical composition, their effect on micro structural properties and strength performance vary considerably. While producing paper, various wastes come out from the various processes in paper industries. The preliminary waste from paper industry is named as hypo sludge. In this study the material obtained from the paper industry waste (hypo sludge is admixed with Portland cement at different replacement levels. The properties of concrete investigated include compressive strength, split tensile strength, flexural strength, sorptivity and acid effect keeping optimum percentage of hypo sludge supposedly supplementary cementitious material (SCM. In this work, M20 grade was developed using IS method of mix design. Specimens of dimensions of 150 x 300mm cylinders for split tensile strength and dimensions of 100 x 100 x 500mm prisms for flexure strength and 150*150*150mm cubes were cast, with and without hypo sludge and tested under axial compression to justify the compressive strength for 7 and 28 days. Standard cubes were immersed in 5%HCL, 5%H2SO4 for inspecting the durability properties. The Sorptivity test has also been conducted. It is concluded that hypo sludge concrete had better mechanical properties and durability properties compared to normal concrete. Test results indicate that the use of hypo sludge in concrete has improved the performance of concrete from strength as well as durability aspects. The split tensile strength is less in hypo sludge concrete compared to normal concrete.

  10. Synthesis and mechanical properties of a calcium sulphoaluminate cement made of industrial wastes

    Directory of Open Access Journals (Sweden)

    Gallardo, M.

    2014-09-01

    Full Text Available Environmentally-friendly calcium sulphoaluminate clinkers were obtained from a mixture of aluminium dross, fluorgypsum, fly ash and CaCO₃ at temperatures within the range of 1100 to 1400 °C. After the heat treatments Ca₄Al₆O₁₂SO₄ was the main phase. Three different cements were prepared using the clinkers synthesized at 1250, 1350 and 1400 °C; the clinker powders were mixed with 20 wt% of hemihydrate. Cement pastes were prepared using a water/cement ratio (w/c, 0.4 followed by curing at 20 or 40 °C for periods of time ranging from 1 to 28 days. Most of the samples showed high compression strengths 40–47 MPa after 28 days, which were comparable to the strength of Portland cement. Ettringite was the main hydration product and its morphology consisted of acicular and hexagonal plates, which is typical of this phase.Se fabricaron clinkers de bajo impacto ambiental a base de sulfoaluminato de calcio calcinando mezclas de escoria de aluminio, fluoryeso, ceniza volante y CaCO₃ a diferentes temperaturas dentro de un rango de 1100 a 1400 °C. Se observó la formación de Ca₄Al₆O₁₂SO₄ como fase principal. Para obtener los cementos, los clinkers obtenidos a 1250, 1350 y 1400 °C se mezclaron con 20% en peso de hemihidrato. Se prepararon pastas usando una relación agua/cemento, de 0.4 y se curaron a 20 y 40 °C por diferentes periodos de tiempo desde 1 hasta 28 días. Los valores de resistencia a la compresión a los 28 días de curado de la mayoría de las muestras estuvieron entre 40–47 MPa, equiparables a los de referencia de pastas de cemento Portland. La etringita fue el principal producto de hidratación y su morfología consistió de placas hexagonales y aciculares, típicas de esta fase.

  11. Material Recovery and Waste Form Development FY 2015 Accomplishments Report

    Energy Technology Data Exchange (ETDEWEB)

    Todd, Terry Allen [Idaho National Lab. (INL), Idaho Falls, ID (United States); Braase, Lori Ann [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-11-01

    The Material Recovery and Waste Form Development (MRWFD) Campaign under the U.S. Department of Energy (DOE) Fuel Cycle Technologies (FCT) Program is responsible for developing advanced separation and waste form technologies to support the various fuel cycle options defined in the DOE Nuclear Energy Research and Development Roadmap, Report to Congress, April 2010. The FY 2015 Accomplishments Report provides a highlight of the results of the research and development (R&D) efforts performed within the MRWFD Campaign in FY-14. Each section contains a high-level overview of the activities, results, technical point of contact, applicable references, and documents produced during the fiscal year. This report briefly outlines campaign management and integration activities, but primarily focuses on the many technical accomplishments made during FY-15. The campaign continued to utilize an engineering driven-science-based approach to maintain relevance and focus. There was increased emphasis on development of technologies that support near-term applications that are relevant to the current once-through fuel cycle.

  12. DuraLith geopolymer waste form for Hanford secondary waste: Correlating setting behavior to hydration heat evolution

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Hui; Gong, Weiliang, E-mail: gongw@vsl.cua.edu; Syltebo, Larry; Lutze, Werner; Pegg, Ian L.

    2014-08-15

    Highlights: • Quantitative correlations firstly established for cementitious waste forms. • Quantitative correlations firstly established for geopolymeric materials. • Ternary DuraLith geopolymer waste forms for Hanford radioactive wastes. • Extended setting times which improve workability for geopolymer waste forms. • Reduced hydration heat release from DuraLith geopolymer waste forms. - Abstract: The binary furnace slag-metakaolin DuraLith geopolymer waste form, which has been considered as one of the candidate waste forms for immobilization of certain Hanford secondary wastes (HSW) from the vitrification of nuclear wastes at the Hanford Site, Washington, was extended to a ternary fly ash-furnace slag-metakaolin system to improve workability, reduce hydration heat, and evaluate high HSW waste loading. A concentrated HSW simulant, consisting of more than 20 chemicals with a sodium concentration of 5 mol/L, was employed to prepare the alkaline activating solution. Fly ash was incorporated at up to 60 wt% into the binder materials, whereas metakaolin was kept constant at 26 wt%. The fresh waste form pastes were subjected to isothermal calorimetry and setting time measurement, and the cured samples were further characterized by compressive strength and TCLP leach tests. This study has firstly established quantitative linear relationships between both initial and final setting times and hydration heat, which were never discovered in scientific literature for any cementitious waste form or geopolymeric material. The successful establishment of the correlations between setting times and hydration heat may make it possible to efficiently design and optimize cementitious waste forms and industrial wastes based geopolymers using limited testing results.

  13. Implementation of industrial waste ferrochrome slag in conventional and low cement castables: Effect of calcined alumina

    Directory of Open Access Journals (Sweden)

    Pattem Hemanth Kumar

    2014-12-01

    Full Text Available A new class of conventional and low-cement ferrochrome slag-based castables were prepared from 40 wt.% ferrochrome slag and 45 wt.% calcined bauxite. Rest fraction varied between high alumina cement (HAC acting as hydraulic binder and calcined alumina as pore filling additive. Standard ASTM size briquettes were prepared for crushing and bending strengths evaluation, and the samples were then subjected to firing at 800, 1100 and 1300 °C for a soaking period of 3 h. The microstructure and refractory properties of the prepared castables have been investigated using X-ray diffraction (XRD, scanning electron microscopy (SEM, cold crushing strength, modulus of rupture and permanent linear changes (PLCs test. Castables show good volume stability (linear change <0.7% at 1300 °C. The outcomes of these investigations were efficacious and in accordance with previously reported data of similar compositions. High thermo-mechanical and physico-chemical properties were attained pointing out an outstanding potential to increase the refractory lining working life of non-recovery coke oven and reheating furnaces.

  14. Reference waste form, basalts, and ground water systems for waste interaction studies

    Energy Technology Data Exchange (ETDEWEB)

    Deju, R.A.; Ledgerwood, R.K.; Long, P.E.

    1978-09-01

    This report summarizes the type of waste form, basalt, and ground water compositions to be used in theoretical and experimental models of the geochemical environment to be simulated in studying a typical basalt repository. Waste forms to be used in the experiments include, and are limited to, glass, supercalcine, and spent unreprocessed fuel. Reference basalts selected for study include the Pomona member and the Umtanum Unit, Shwana Member, of the Columbia River Basalt Group. In addition, a sample of the Basalt International Geochemical Standard (BCR-1) will be used for cross-comparison purposes. The representative water to be used is of a sodium bicarbonate composition as determined from results of analyses of deep ground waters underlying the Hanford Site. 12 figures, 13 tables.

  15. A study on the technology for reducing cement-type materials used for tunnel supports at high-level radioactive waste disposal sites (Joint research)

    International Nuclear Information System (INIS)

    Cement-type materials that are used for supports or grouting at high-level radioactive waste disposal facilities leach into the groundwater and create a highly alkaline environment. Of concern in highly alkaline environments are the alteration of bentonite used as buffers or backfill materials, and of surrounding rock mass, and the increased uncertainty regarding the provision of performance of the disposal system over a long period of time. From such a background, The Japan Atomic Energy Agency and Shimizu Corporation carried out a Joint Research aimed for technology development to reduce the use of cement-type materials at high-level radioactive waste disposal facilities as much as possible. The Japan Atomic Energy Agency has been conducting research and development of low alkalinity cement with a view to controlling the effects of high alkalinity, and focusing on its use as support materials. In the meantime, Shimizu Corporation has been developing methods for constructing tunnels using the minimum quantities of cement-type support materials and suggesting the possibility of realization of support structures using greatly reduced quantities of cement. In this study, the quantities of cement used for supports were first presented under the geological condition (soft or hard rock) by the method of waste disposal (vertical or horizontal) described in the secondary report on geological disposal of high-level radioactive waste. Then, alternative supports mainly composed of rock and bentonite were proposed to ensure long-term provision of the performance of the disposal system. The mechanical characteristic values concerning the strength and deformation properties of the alternative supports and backfill materials (for filling the gap between the tunnel wall and the segmental ring) were examined. The supports were designed based on the physical properties of rock mass and earth pressure described in the secondary report, and on the physical property values of the

  16. Silica based gel as a potential waste form for high level waste from fuel reprocessing

    International Nuclear Information System (INIS)

    To assess the feasibility of safe disposal of high-level radioactive waste as synthetic clay, or material that would react with ground water to form clay, experiments have been carried out to determine the hydrothermal crystallisation and leaching behaviour of silica based gels fired at 900 deg C. Crystallisation rates at a pressure of 500 bars and at temperatures below 400 deg C are negligible and this more or less precludes pre-disposal production of synthetic clay on the scale required. Leaching experiments suggest that the leach rates of Cs from gels by distilled water are higher than those of boro-silicate glasses and SYNROC at the lower temperatures that would be preferred for geological storage. However, amounts of bulk dissolution of gels may be lower than those of boro-silicate glasses. The initial leaching behaviour of gels might be considerably improved by hot compaction at 900 to 1000 deg C. Consideration of likely waste form dissolution behaviour in a repository environment suggests that gels of appropriate composition might perform as well as, or better than, boro-silicate glasses. A novel hypothetical plant is described that could produce the gel waste form on the scale required on a more or less continuous basis. (author)

  17. Mineralogy and microstructure of two Mexican Portland cements for the confinement of radioactive waste; Mineralogia y microestructura de dos cementos mexicanos Portland para el confinamiento de desechos radiactivos

    Energy Technology Data Exchange (ETDEWEB)

    Galicia A, E. [Universidad Autonoma del Estado de Mexico, Facultad de Ciencias, Campus El Cerrillo, Piedras Blancas, Carretera Toluca-Ixtlahuaca Km. 15.5, Estado de Mexico (Mexico); Badillo A, V. E.; Ramirez S, J. R. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Nava E, N., E-mail: nasiega_181@hotmail.com [Instituto Mexicano del Petroleo, Eje Central Lazaro Cardenas No. 152, Col. San Bartolo Atepehuacan, 07730 Mexico D. F. (Mexico)

    2014-10-15

    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)

  18. Impeding 99Tc(IV) mobility in novel waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Mal Soon; Um, Wooyong; Wang, Guohui; Kruger, Albert A.; Lukens, Wayne W.; Rousseau, Roger J.; Glezakou, Vassiliki Alexandra

    2016-06-30

    Technetium (99Tc) is a long-lived radioactive fission product whose mobility in the subsurface is largely governed by its oxidation state1. Immobilization of Tc in mineral substrates is crucial for radioactive waste management and environmental remediation. Tc(IV) incorporation in spinels2, 3 has been proposed as a novel method to increase Tc retention in glass waste forms. However, experiments with Tc-magnetite under high temperature and oxic conditions showed re-oxidation of Tc(IV) to volatile pertechnetate Tc(VII)O4-.4, 5 Here we address this problem with large-scale ab initio molecular dynamics simulations and propose that elevated temperatures, 1st row transition metal dopants can significantly enhance Tc retention in the order Co > Zn > Ni. Experiments with doped spinels at T=700 ºC provided quantitative confirmation of increased Tc retention in the same order predicted by theory. This work highlights the power of modern state-of-the-art simulations to provide essential insights and generate bottom-up design criteria of complex oxide materials at elevated temperatures.

  19. Impeding 99Tc(IV) mobility in novel waste forms

    Science.gov (United States)

    Lee, Mal-Soon; Um, Wooyong; Wang, Guohui; Kruger, Albert A.; Lukens, Wayne W.; Rousseau, Roger; Glezakou, Vassiliki-Alexandra

    2016-06-01

    Technetium (99Tc) is an abundant, long-lived radioactive fission product whose mobility in the subsurface is largely governed by its oxidation state. Tc immobilization is crucial for radioactive waste management and environmental remediation. Tc(IV) incorporation in spinels has been proposed as a novel method to increase Tc retention in glass waste forms during vitrification. However, experiments under high-temperature and oxic conditions show reoxidation of Tc(IV) to volatile pertechnetate, Tc(VII). Here we examine this problem with ab initio molecular dynamics simulations and propose that, at elevated temperatures, doping with first row transition metal can significantly enhance Tc retention in magnetite in the order Co>Zn>Ni. Experiments with doped spinels at 700 °C provide quantitative confirmation of the theoretical predictions in the same order. This work highlights the power of modern, state-of-the-art simulations to provide essential insights and generate theory-inspired design criteria of complex materials at elevated temperatures.

  20. Impeding (99)Tc(IV) mobility in novel waste forms.

    Science.gov (United States)

    Lee, Mal-Soon; Um, Wooyong; Wang, Guohui; Kruger, Albert A; Lukens, Wayne W; Rousseau, Roger; Glezakou, Vassiliki-Alexandra

    2016-01-01

    Technetium ((99)Tc) is an abundant, long-lived radioactive fission product whose mobility in the subsurface is largely governed by its oxidation state. Tc immobilization is crucial for radioactive waste management and environmental remediation. Tc(IV) incorporation in spinels has been proposed as a novel method to increase Tc retention in glass waste forms during vitrification. However, experiments under high-temperature and oxic conditions show reoxidation of Tc(IV) to volatile pertechnetate, Tc(VII). Here we examine this problem with ab initio molecular dynamics simulations and propose that, at elevated temperatures, doping with first row transition metal can significantly enhance Tc retention in magnetite in the order Co>Zn>Ni. Experiments with doped spinels at 700 °C provide quantitative confirmation of the theoretical predictions in the same order. This work highlights the power of modern, state-of-the-art simulations to provide essential insights and generate theory-inspired design criteria of complex materials at elevated temperatures. PMID:27357121

  1. Impeding 99Tc(IV) mobility in novel waste forms

    Science.gov (United States)

    Lee, Mal-Soon; Um, Wooyong; Wang, Guohui; Kruger, Albert A.; Lukens, Wayne W.; Rousseau, Roger; Glezakou, Vassiliki-Alexandra

    2016-01-01

    Technetium (99Tc) is an abundant, long-lived radioactive fission product whose mobility in the subsurface is largely governed by its oxidation state. Tc immobilization is crucial for radioactive waste management and environmental remediation. Tc(IV) incorporation in spinels has been proposed as a novel method to increase Tc retention in glass waste forms during vitrification. However, experiments under high-temperature and oxic conditions show reoxidation of Tc(IV) to volatile pertechnetate, Tc(VII). Here we examine this problem with ab initio molecular dynamics simulations and propose that, at elevated temperatures, doping with first row transition metal can significantly enhance Tc retention in magnetite in the order Co>Zn>Ni. Experiments with doped spinels at 700 °C provide quantitative confirmation of the theoretical predictions in the same order. This work highlights the power of modern, state-of-the-art simulations to provide essential insights and generate theory-inspired design criteria of complex materials at elevated temperatures. PMID:27357121

  2. Implementation of industrial waste ferrochrome slag in conventional and low cement castables: Effect of microsilica addition

    Directory of Open Access Journals (Sweden)

    Pattem Hemanth Kumar

    2014-06-01

    Samples with decreasing cement content 15–05 wt.% were formulated in combination of both slag and calcined bauxite as matrix components. Effects of varying 0–10 wt.% microsilica as a micro-fine additive in these castables were investigated in this work. Pore filling properties of microsilica improved apparent porosity and bulk density. Phase analysis through X-ray diffraction techniques demonstrates successful formation of spinel and mullite crystalline phases. Mechanical behavior was evaluated through cold crushing strength and residual cold crushing strength after five consecutive water quenching cycles. Scanning electron microscopy measurements were carried out in order to better understand the packing density and reaction mechanisms of fired castables. Slag containing castables portrays good thermal properties such as thermal shock resistance, permanent linear change and pyrometric cone equivalent.

  3. Alternative Electrochemical Salt Waste Forms, Summary of FY2010 Results

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J.; Rieck, Bennett T.; Crum, Jarrod V.; Matyas, Josef; McCloy, John S.; Sundaram, S. K.; Vienna, John D.

    2010-08-01

    In FY2009, PNNL performed scoping studies to qualify two waste form candidates, tellurite (TeO2-based) glasses and halide minerals, for the electrochemical waste stream for further investigation. Both candidates showed promise with acceptable PCT release rates and effective incorporation of the 10% fission product waste stream. Both candidates received reprisal for FY2010 and were further investigated. At the beginning of FY2010, an in-depth literature review kicked off the tellurite glasses study. The review was aimed at ascertaining the state-of-the-art for chemical durability testing and mixed chloride incorporation for tellurite glasses. The literature review led the authors to 4 unique binary and 1 unique ternary systems for further investigation which include TeO2 plus the following: PbO, Al2O3-B2O3, WO3, P2O5, and ZnO. Each system was studied with and without a mixed chloride simulated electrochemical waste stream and the literature review provided the starting points for the baseline compositions as well as starting points for melting temperature, compatible crucible types, etc. The most promising glasses in each system were scaled up in production and were analyzed with the Product Consistency Test, a chemical durability test. Baseline and PCT glasses were analyzed to determine their state, i.e., amorphous, crystalline, phase separated, had undissolved material within the bulk, etc. Conclusions were made as well as the proposed direction for FY2011 plans. Sodalite was successfully synthesized by the sol-gel method. The vast majority of the dried sol-gel consisted of sodalite with small amounts of alumino-silicates and unreacted salt. Upon firing the powders made by sol-gel, the primary phase observed was sodalite with the addition of varying amounts of nepheline, carnegieite, lithium silicate, and lanthanide oxide. The amount of sodalite, nepheline, and carnegieite as well as the bulk density of the fired pellets varied with firing temperature, sol

  4. Waste form characteristics report, revision 1.3

    Energy Technology Data Exchange (ETDEWEB)

    Leider, H.R.; Stout, R.B.

    1998-07-01

    This Waste Form Characteristics Report (WFCR) update, Version 1.3, incorporates substantial additions and changes to following 10 sections of the WFCR: 2.1.3.1 Cladding Degradation; 2.1.3.2 UO2 Oxidation in Fuel; 2.1.3.5 Dissolution Release from UO{sub 2}; 2.2.1.5 Fracture /Fragmentation Studies of Glass; 2.2.2.2 Dissolution Radionuclide Release from Glass; 2.2.2.3 Soluble-Precipitated/Colloidal Species from Glass; 3.2.2 Spent-Fuel Oxidation Models; 3.4.2 Spent-Fuel Dissolution Models; 3.5.1 Glass Dissolution Experimental Parameters; and 3.5.2 Glass Dissolution Models.

  5. Microscopic characterization of crystalline phases in waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Buck, E.C.; Dietz, N.L.; Wronkiewicz, D.J.; Bates, J.K. [Argonne National Lab., IL (United States); Millar, A. [Purdue Univ., West Lafayette, IN (United States)

    1995-07-01

    Transmission electron microscopy (TEM) has been used to determine the microstructure of crystalline phases present in zirconium- and titanium-bearing glass crystalline composite (GCC) waste forms. The GCC materials were found to contain spinels (maghemite), zirconolites, perovskites (CaTiO{sub 3}) and plagiociase feldspar (anorthite) mineral phases. The structure of the uranium and cerium-bearing monoclinic zirconolite was characterized by medium resolution TEM imaging and electron and X-ray diffraction (XRD). The phase was found to contain high levels of iron in comparison to Synroc-type zirconolites. Excess zirconium in zirconolite has resulted in martensitic baddeleyite (ZrO{sub 2}) formation. Anorthite (CaAl{sub 2}Si{sub 2}O{sub 8}) was present as elongated crystallites within a calcium-rich aluminosilicate glass. Lead and iron-bearing anorthite lying along distinct precipitates were occasionally observed within the an crystallographic planes.

  6. Progress in forming bottom barriers under waste sites

    Energy Technology Data Exchange (ETDEWEB)

    Carter, E.E. [Carter Technologies, Sugar Land, TX (United States)

    1997-12-31

    The paper describes an new method for the construction, verification, and maintenance of underground vaults to isolate and contain radioactive burial sites without excavation or drilling in contaminated areas. The paper begins with a discussion of previous full-scale field tests of horizontal barrier tools which utilized high pressure jetting technology. This is followed by a discussion of the TECT process, which cuts with an abrasive cable instead of high pressure jets. The new method is potentially applicable to more soil types than previous methods and can form very thick barriers. Both processes are performed from the perimeter of a site and require no penetration or disturbance of the active waste area. The paper also describes long-term verification methods to monitor barrier integrity passively.

  7. Radionuclide Incorporation and Long Term Performance of Apatite Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jianwei [Louisiana State Univ., Baton Rouge, LA (United States); Lian, Jie [Rensselaer Polytechnic Inst., Troy, NY (United States); Gao, Fei [Univ. of Michigan, Ann Arbor, MI (United States)

    2016-01-04

    This project aims to combines state-of-the-art experimental and characterization techniques with atomistic simulations based on density functional theory (DFT) and molecular dynamics (MD) simulations. With an initial focus on long-lived I-129 and other radionuclides such as Cs, Sr in apatite structure, specific research objectives include the atomic scale understanding of: (1) incorporation behavior of the radionuclides and their effects on the crystal chemistry and phase stability; (2) stability and microstructure evolution of designed waste forms under coupled temperature and radiation environments; (3) incorporation and migration energetics of radionuclides and release behaviors as probed by DFT and molecular dynamics (MD) simulations; and (4) chemical durability as measured in dissolution experiments for long term performance evaluation and model validation.

  8. Low-risk alternative waste forms for problematic high-level and long-lived nuclear wastes

    International Nuclear Information System (INIS)

    Full text: The highest cost component the nuclear waste clean up challenge centres on high-level waste (HLW) and consequently the greatest opportunity for cost and schedule savings lies with optimising the approach to HLW cleanup. The waste form is the key component of the immobilisation process. To achieve maximum cost savings and optimum performance the selection of the waste form should be driven by the characteristics of the specific nuclear waste to be immobilised, rather than adopting a single baseline approach. This is particularly true for problematic nuclear wastes that are often not amenable to a single baseline approach. The use of tailored, high-performance, alternative waste forms that include ceramics and glass-ceramics, coupled with mature process technologies offer significant performance improvements and efficiency savings for a nuclear waste cleanup program. It is the waste form that determines how well the waste is locked up (chemical durability), and the number of repository disposal canisters required (waste loading efficiency). The use of alternative waste forms for problematic wastes also lowers the overall risk by providing high performance HLW treatment alternatives. The benefits tailored alternative waste forms bring to the HLW cleanup program will be briefly reviewed with reference to work carried out on the following: The HLW calcines at the Idaho National Laboratory; SYNROC ANSTO has developed a process utilising a glass-ceramic combined with mature hot-isostatic pressing (HIP) technology and has demonstrated this at a waste loading of 80 % and at a 30 kg HIP scale. The use of this technology has recently been estimated to result in a 70 % reduction in waste canisters, compared to the baseline borosilicate glass technology; Actinide-rich waste streams, particularly the work being done by SYNROC ANSTO with Nexia Solutions on the Plutonium-residues wastes at Sellafield in the UK, which if implemented is forecast to result in substantial

  9. Material Recovery and Waste Form Development FY 2014 Accomplishments Report

    Energy Technology Data Exchange (ETDEWEB)

    Lori Braase

    2014-11-01

    Develop advanced nuclear fuel cycle separation and waste management technologies that improve current fuel cycle performance and enable a sustainable fuel cycle, with minimal processing, waste generation, and potential for material diversion.

  10. MODELING SOLIDIFICATION-INDUCED STRESSES IN CERAMIC WASTE FORMS CONTAINING NUCLEAR WASTES

    Energy Technology Data Exchange (ETDEWEB)

    Charles W. Solbrig; Kenneth J. Bateman

    2010-11-01

    The goal of this work is to produce a ceramic waste form (CWF) that permanently occludes radioactive waste. This is accomplished by absorbing radioactive salts into zeolite, mixing with glass frit, heating to a molten state 915 C to form a sodalite glass matrix, and solidifying for long-term storage. Less long term leaching is expected if the solidifying cooling rate doesn’t cause cracking. In addition to thermal stress, this paper proposes that a stress is formed during solidification which is very large for fast cooling rates during solidification and can cause severe cracking. A solidifying glass or ceramic cylinder forms a dome on the cylinder top end. The temperature distribution at the time of solidification causes the stress and the dome. The dome height, “the length deficit,” produces an axial stress when the solid returns to room temperature with the inherent outer region in compression, the inner in tension. Large tensions will cause cracking of the specimen. The temperature deficit, derived by dividing the length deficit by the coefficient of thermal expansion, allows solidification stress theory to be extended to the circumferential stress. This paper derives the solidification stress theory, gives examples, explains how to induce beneficial stresses, and compares theory to experimental data.

  11. Immobilization of fission products in phosphate ceramic waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Singh, D. [Argonne National Lab., IL (United States)

    1996-10-01

    The goal of this project is to develop and demonstrate the feasibility of a novel low-temperature solidification/stabilization (S/S) technology for immobilizing waste streams containing fission products such as cesium, strontium, and technetium in a chemically bonded phosphate ceramic. This technology can immobilize partitioned tank wastes and decontaminate waste streams containing volatile fission products.

  12. Hanford Tank 241-C-106: Impact of Cement Reactions on Release of Contaminants from Residual Waste

    Energy Technology Data Exchange (ETDEWEB)

    Deutsch, William J.; Krupka, Kenneth M.; Lindberg, Michael J.; Cantrell, Kirk J.; Brown, Christopher F.; Schaef, Herbert T.

    2006-09-01

    The CH2M HILL Hanford Group, Inc. (CH2M HILL) is producing risk/performance assessments to support the closure of single-shell tanks at the U.S. Department of Energy's Hanford Site. As part of this effort, staff at Pacific Northwest National Laboratory were asked to develop release models for contaminants of concern that are present in residual sludge remaining in tank 241-C-106 (C-106) after final retrieval of waste from the tank. Initial work to produce release models was conducted on residual tank sludge using pure water as the leaching agent. The results were reported in an earlier report. The decision has now been made to close the tanks after waste retrieval with a cementitious grout to minimize infiltration and maintain the physical integrity of the tanks. This report describes testing of the residual waste with a leaching solution that simulates the composition of water passing through the grout and contacting the residual waste at the bottom of the tank.

  13. Waste Heat Recovery for the Cement Sector : Market and Supplier Analysis

    OpenAIRE

    International Finance Corporation; Institute for Industrial Productivity

    2014-01-01

    This report analyzes the current status of Waste Heat Recovery (WHR) technology deployment in developing countries and investigates the success factors in countries where WHR has become widely spread. The report then focuses on the in-depth analysis of WHR potential and enabling factors in eleven country markets in Africa (Nigeria, South Africa), South Asia (India, Pakistan), Middle East (...

  14. Naturally occurring glasses: analogues for radioactive waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Ewing, R.C.; Haaker, R.F.

    1979-04-01

    Volcanic glasses are very often altered by weathering and leaching and recrystallize to their fine-grained equivalents (rhyolites, felsites). The oldest volcanic glasses are dated at 40 million years before the present, but the majority are much younger. Devitrification textures was produced experimentally; and hydration rates for volcanic glasses were determined as a function of composition, temperature, and climate. Presence of water and temperature are the most important rate controlling variables. Even material that may still be described as glassy often exhibits evidence of alteration and recrystallization. Of the volcanic glasses that are preserved in the geologic record, it would be rare to describe such a glass as pristine. Despite the common alteration and recrystallization effects observed in volcanic glasses, glasses formed as a result of impact, tektites and lunar glasses, may occur in substantially unaltered form. In the case of tektites, their resistance to alteration is a result of their high SiO/sub 2/ content and low alkali content. Lunar glasses have been preserved for hundreds of millions of years because they exist in an environment with a low oxygen fugacity and an extremely low water vapor partial presssure. Thus one might expect glasses of particular compositions or in specific types of environment to be stable for long periods of time. These conclusions are applied to radioactive waste disposal over several time periods (0-30h, 30h-20y, 20-200y).

  15. Hanford Waste Vitrification Plant Quality Assurance Program description for high-level waste form development and qualification

    International Nuclear Information System (INIS)

    The Hanford Waste Vitrification Plant Project has been established to convert the high-level radioactive waste associated with nuclear defense production at the Hanford Site into a waste form suitable for disposal in a deep geologic repository. The Hanford Waste Vitrification Plant will mix processed radioactive waste with borosilicate material, then heat the mixture to its melting point (vitrification) to forin a glass-like substance that traps the radionuclides in the glass matrix upon cooling. The Hanford Waste Vitrification Plant Quality Assurance Program has been established to support the mission of the Hanford Waste Vitrification Plant. This Quality Assurance Program Description has been written to document the Hanford Waste Vitrification Plant Quality Assurance Program

  16. Characterisation of Ba(OH)(2)-Na2SO4-blast furnace slag cement-like composites for the immobilisation of sulfate bearing nuclear wastes

    OpenAIRE

    Mobasher, N.; Bernal, S. A.; Hussain, O.H.; Apperley, D.C.; Kinoshita, H.; Provis, J.L.

    2014-01-01

    Soluble sulfate ions in nuclear waste can have detrimental effects on cementitious wasteforms and disposal facilities based on Portland cement. As an alternative, Ba(OH)2–Na2SO4–blast furnace slag composites are studied for immobilisation of sulfate-bearing nuclear wastes. Calcium aluminosilicate hydrate (C–A–S–H) with some barium substitution is the main binder phase, with barium also present in the low solubility salts BaSO4 and BaCO3, along with Ba-substituted calcium sulfoaluminate hydrat...

  17. Development of a ceramic waste form for high-level waste disposal

    International Nuclear Information System (INIS)

    A ceramic waste form is being developed by Argonne National Laboratory (ANL) as part of the demonstration of the electrometallurgical treatment of spent nuclear fuel. The halide, alkaline earth, alkali, transuranic, and rare earth fission products are stabilized in zeolite which is combined with glass and processed in a hot isostatic press (HIP) to form a ceramic composite. The mineral sodalite is formed in the HIP from the zeolite precursor. The process, from starting materials to final product, is relatively simple. An overview of the processing operations is given. The metrics that have been developed to measure the success or completion of processing operations are developed and discussed. The impact of variability in processing metrics on the durability of the final product is presented

  18. Immobilization of 99Tc in low-temperature phosphate ceramic waste forms

    International Nuclear Information System (INIS)

    Radionuclides such as 99Tc are by-products of fission reactions in high-level wastes. Technetium poses a serious environmental threat because it is easily oxidized into its highly leachable pertechnetate form. Magnesium potassium phosphate ceramics have been developed to treat 99Tc that has been separated and eluted from simulated high-level tank wastes by sorption processes. Dense and hard ceramic waste forms were fabricated by acid-base reactions between mixtures of magnesium oxide powders and wastes, and acid phosphate solutions. Standard leaching tests, such as ANS 16.1 and the Product Consistency Test, were conducted on the final waste forms to establish their performance. The fate of the contaminants in the final waste forms was established with scanning electron microscopy techniques. In addition, stability of the waste forms in aqueous environments was evaluated by long-term water immersion tests

  19. DuraLith geopolymer waste form for Hanford secondary waste: correlating setting behavior to hydration heat evolution.

    Science.gov (United States)

    Xu, Hui; Gong, Weiliang; Syltebo, Larry; Lutze, Werner; Pegg, Ian L

    2014-08-15

    The binary furnace slag-metakaolin DuraLith geopolymer waste form, which has been considered as one of the candidate waste forms for immobilization of certain Hanford secondary wastes (HSW) from the vitrification of nuclear wastes at the Hanford Site, Washington, was extended to a ternary fly ash-furnace slag-metakaolin system to improve workability, reduce hydration heat, and evaluate high HSW waste loading. A concentrated HSW simulant, consisting of more than 20 chemicals with a sodium concentration of 5 mol/L, was employed to prepare the alkaline activating solution. Fly ash was incorporated at up to 60 wt% into the binder materials, whereas metakaolin was kept constant at 26 wt%. The fresh waste form pastes were subjected to isothermal calorimetry and setting time measurement, and the cured samples were further characterized by compressive strength and TCLP leach tests. This study has firstly established quantitative linear relationships between both initial and final setting times and hydration heat, which were never discovered in scientific literature for any cementitious waste form or geopolymeric material. The successful establishment of the correlations between setting times and hydration heat may make it possible to efficiently design and optimize cementitious waste forms and industrial wastes based geopolymers using limited testing results.

  20. Evaluation of compressive strength and water absorption of soil-cement bricks manufactured with addition of pet (polyethylene terephthalate wastes

    Directory of Open Access Journals (Sweden)

    João Alexandre Paschoalin Filho

    2016-04-01

    Full Text Available This paper presents the evaluation of compressive strength of soil-cement bricks obtained by the inclusion in their mixture of PET flakes through mineral water bottles grinding. The Polyethylene Terephthalate (PET has been characterized by its difficulty of disaggregation in nature, requiring a long period for this. On the other hand, with the increase in civil construction activities the demand for raw material also increases, causing considerable environmental impacts. In this context, the objective of this research is to propose a simple methodology, preventing its dumping and accumulation in irregular areas, and reducing the demand of raw materials by the civil construction industry. The results showed that compressive strengths obtained were lower than recommended by NBR 8491 (Associação Brasileira de Normas Técnicas [ABNT], 2012b at seven days of curing time. However, they may be used as an alternative solution in masonry works in order to not submit themselves to great loads or structural functions. The studied bricks also presented water absorption near to recommended values by NBR 8491 (ABNT, 2012b. Manufacturing costs were also determined for this brick, comparing it with the costs of other brick types. Each brick withdrew from circulation approximately 300 g of PET waste. Thus, for an area of 1 m2 the studied bricks can promote the withdrawal of approximately 180 beverage bottles of 2 L capacity.

  1. Alloy waste forms for metal fission products and actinides isolated by spent nuclear fuel treatment

    International Nuclear Information System (INIS)

    Waste form alloys are being developed at Argonne National Laboratory for the disposal of remnant metallic wastes from an electrometallurgical process developed to treat spent nuclear fuel. This metal waste form consists of the fuel cladding (stainless steel or Zircaloy), noble metal fission products (e.g., Ru, Pd, Mo and Tc), and other metallic wastes. The main constituents of the metal waste stream are the cladding hulls (85 to 90 wt%); using the hulls as the dominant alloying component minimizes the overall waste volume as compared to vitrification or metal encapsulation. Two nominal compositions for the waste form are being developed: (1) stainless steel-15 wt% zirconium for stainless steel-clad fuels and (2) zirconium-8 wt% stainless steel for Zircaloy-clad fuels. The noble metal fission products are the primary source of radiation in the metal waste form. However, inclusion of actinides in the metal waste form is being investigated as an option for interim or ultimate storage. Simulated waste form alloys were prepared and analyzed to determine the baseline alloy microstructures and the microstructural distribution of noble metals and actinides. Corrosion tests of the metal waste form alloys indicate that they are highly resistant to corrosion

  2. Process development for utilizing asbestos cement waste in rotary kilns for the cement industry. Final report; Erarbeitung eines Verfahrens zur stofflichen Verwertung von zementgebundenen Asbestprodukten in Drehrohroefen fuer die Zementindustrie. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Schlegel, R.; Kieser, J.; Kraehner, A.

    1999-11-01

    The law for recycling and waste demands the utilization also for waste of asbestos cement (ac). The procedure of thermal utilization of ac in the flame of a rotary cement kiln was developed and patented by the research institute IBU-tec Weimar, Germany. The ac-material has to be pre-pulverized and grinded to a degree of fineness of R{sub 90}<15%. Considerations of safety engineering lead to the idea of common fine grinding of old oil (oo) and ac. This new procedure was searched in FuE-project in 1998/99 (financial support by BMBF). A mash of ac and oo was generated as a utilization product ready for firing which was injected into the flame of the rotary cement kiln. This particles of ac smelt to spherical shaped particles at a temperature above 1500 C. They were utilized by clinker formation. The material and gas stream leaving the kiln does not contain fibres of asbestos. This was demonstrated in a small equipment burning test. The industrial realization concerning cement plant Ruedersdorf, near Berlin, was searched, technologically described and safety engineeringly and financially assessed by a project study. Process-technical and financial advantages were seen for the dry fine grinding. The wet fine grinding with old oil could be used in cement plants using old oil as fuel. (orig.) [German] Das Kreislaufwirtschafts- und Abfallgesetz (1994) fordert u.a. die stoffliche Verwertung auch fuer Asbestzementabfaelle (AZ). Das vom Institut fuer Baustoff- und Umweltschutz-Technologie Weimar 1995 entwickelte und patentierte Verfahren zur thermischen Verwertung von AZ in der Flamme eines Zementdrehrohrofens erfuellt diese Forderung. Das AZ-Material muss vorzerkleinert und bis zur Rohmehlfeinheit (R{sub 90}<15%) feingemahlen werden. Sicherheitstechnische Ueberlegungen fuehrten zu der Idee, die Feinmahlung zusammen mit Altoel (AOe) zu erproben. Diese Verfahrensvariante wurde im Rahmen eines FuE-Projektes 1998/99 untersucht (finanzielle Foerderung durch das BMBF). Als

  3. Modelling of hydrogen production from pore water radiolysis in cemented intermediate level waste

    Directory of Open Access Journals (Sweden)

    Di Giandomenico M.-V.

    2013-07-01

    Full Text Available In France, some of the intermediate and low level wastes are embedded in hydraulic binder and put into concrete canisters. They contain β and γ emitters which cause an irradiation of water present in the pores of the hydraulic binder. This is responsible for a dihydrogen (H2 production due to radiolysis. EDF R&D and CEA have collaborated since many years in order to understand this phenomenon and develop a model called DO-RE-MI which can predict such a production of dihydrogen in concrete waste packages. A parametric study, using the developed model, was implemented in order to determine the effects of each parameter on H2 production. The main results are presented in this paper.

  4. Radiation effects in glass and glass-ceramic waste forms for the immobilization of CANDU UO2 fuel reprocessing waste

    International Nuclear Information System (INIS)

    AECL has investigated three waste forms for the immobilization of high-level liquid wastes that would arise if used CANDU fuels were reprocessed at some time in the future to remove fissile materials for the fabrication of new power reactor fuel. These waste forms are borosilicate glasses, aluminosilicate glasses and titanosilicate glass-ceramics. This report discusses the potential effects of alpha, beta and gamma radiation on the releases of radionuclides from these waste forms as a result of aqueous corrosion by groundwaters that would be present in an underground waste disposal vault. The report discusses solid-state damage caused by radiation-induced atomic displacements in the waste forms as well as irradiation of groundwater solutions (radiolysis), and their potential effects on waste-form corrosion and radionuclide release. The current literature on radiation effects on borosilicate glasses and in ceramics is briefly reviewed, as are potential radiation effects on specialized waste forms for the immobilization of 129I, 85Kr and 14C. (author). 104 refs., 9 tabs., 5 figs

  5. Low pH Cements

    Energy Technology Data Exchange (ETDEWEB)

    Savage, David; Benbow, Steven [Quintessa Ltd., Henley-on-Thames (United Kingdom)

    2007-05-15

    The development of low-pH cements for use in geological repositories for radioactive waste stems from concerns over the potential for deleterious effects upon the host rock and other EBS materials (notably bentonite) under the hyperalkaline conditions (pH > 12) of cement pore fluids. 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. SKB and Posiva have ruled out the use of blast furnace slag and fly-ash and are focusing on silica fume as a blending agent. Currently, no preferred composition has been identified by these agencies. SKB and Posiva 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/